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Text

Initial Exam 2009-301 Draft SRO Written Exam
	ML092120463
Person / Time
Site:	Harris
Issue date:	07/31/2009
From:	; NRC/RGN-II
To:	; Carolina Power & Light Co
References
	50-400/09-301
	Download: ML092120463 (125)
	v • d • e

{{#Wiki_filter:QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

1. 2009A NRC SRO OOIINEW/F/3/T.S. 3.9.8.1/N/2009A NRC SRO/025AG2.2.36/

Given the following plant conditions:

             - Plant is in Mode 6
             - Refueling Cavity Level is at 23' 6"
             - Both trains of RHR are in service for Shutdown Cooling
             - 'B' EDG is under clearance for scheduled maintenance A Loss of Offsite Power occurs.

Which ONE of the following describes the action required to comply with the RHR Limiting Condition for Operation and the basis of this LCO? A. Restore power to the 'B' RHR Pump; Ensures that sufficient cooling capacity is available to maintain the RCS below 200°F B. Restore power to the 'B' RHR Pump; Ensures that sufficient cooling capacity is available to maintain the RCS below 140°F C. Start the 'A' RHR Pump after Load Block 9 on the Sequencer; Ensures that sufficient cooling capacity is available to maintain the RCS below 200°F c D~ Start the 'A' RHR Pump after Load Block 9 on the Sequencer; En~l)Jes that sufficient cooling capacity is available to maintain the RCS below 140°F Plausibility and Answer Analysis A Incorrect. This is plausible if the candidate believes two RHR pump are required to be operable which would be true in Mode 5 or if cavity level was lower. 200°F is plausible because this is Mode 4 where. addition concerns arise. B Incorrect. This is plausible if the candidate believes two RHR pump are required to be operable which would be true in Mode 5 or if cavity level was lower. 140°F is correct. C Incorrect. ~' RHR Pump must be started. 200°F is plausible because this is Mode 4 where addition concerns arise. D Correct. ~' RHR Pump must be started. 140°F is correct. ~.--------------------------------------- Friday, December 26, 2008 1:06:52 PM 1

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Loss of RHR System: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions of operations Importance Rating: 3.1 4.2 Technical

Reference:

Tech Spec 3.9.8.1 pg 3/4 9-9 (page 364) Tech Spec Bases 3/4.9.8 pg B 3/4 9-2 (page 466) References to be provided: None Learning Objective: RHR System, Obj 9f Question origin: NEW Comments: 'B' EDG under clearance is the maintenance piece of the KA and the Loss of Offsite Power produces the Loss of RHR SRO justification: Requires knowledge of Tech Spec Bases that are not system knowledge. Origin: NEW Cog Level: F Difficulty: 3 -

Reference:

T.S.3.9.8.1 Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: 025AG2.2.36 KIA 2: Friday, December 26, 2008 1:06:52 PM 2

Ref for SRO #1, 025AG2.2.36 REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation: APPLICABILITY: MODE 6. with irradiated fuel in the vessel when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet. ACTION: With no RHR loop OPERABLE and in operation. suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outSide atmosphere within 4 hours. SURVEILLANCE REQUIREMENTS A.9.8.1 At-least one RHR loop shall be verified in operation~and--circulating* reactor coolant at a flow rate of greater than or equal to 2500 gpm at least once per 12 hours.

The RHR loop may be removed from operation for up to 1 hour per 2-hour period during the performance of CORE ALTERATIONS and core loading verification in the vicinity of the reactor vessel hot legs.

-l ~---------------------- SHEARON HARRIS - UNIT 1 3/4 9-9

Ref for SRO #1, 025AG2.2.36 REFUELING OPERATIONS BASES CONTAINMENT BUILDING PENETRATIONS (Continued) The allowance to have containment penetration (including the airlock doors and equipment hatch) flow paths with direct access from the containment atmosphere to the outside atmosphere to be unisolated during fuel movement and CORE ALTERATIONS is based on (1) confirmatory dose calculations as approved by the NRC staff which indicate acceptable radiological consequences and (2) commitments from the licensee 10 implement acceptable administralive procedures that ensure in the event of a refueling accident that the airlock or equipment hatch can and will be promptly closea following containment evacuation (even though the containment fission product control function is not required to meet acceptable dose consequences) and that the open penetration(s) can and will be promptly closed. The time to close such penetrations or combination of penetrations shall be included in the confirmatory dose calculations. Containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated. or capable of isolation via administrative controls. on at least one Side of containment. Isolation may be achieved by an OPERABLE automatic isolation valve. or by a manual isolation valve. blind flange. or equivalent. Equivalent isolation methods include use of a material that can provide a temporary. atmospheric pressure. ventilation barrier for the other containment penetrations during fuel movement. 3/4.9.5 COMMUNICATIONS - DELETED 3/4.9.6 REFUELING MACHINE - DELETED 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING - DELETED 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140°F as required during the REFUELING MODE. and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification. The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange. a large heat sink is available for core cooling. Thus. in the event of a failure of the operating RHR loop. adequate time is provided to initiate emergency procedures to cool the core. The minimum RHR flow requirement is reduced to 900 gpm when the reactor water level is below the reactor vessel flange. The 900 gpm limit reduces the possibility of cavitation during operation of the RHR pumps and ensures sufficient mixing in the event of a MODE 6 boron dilution incident. 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment purge makeup and exhaust penetrations will be automatically isolated upon detectlon of high radiation levels within the containment. The OPERABILITY of this system is ______~-:,p,.;;~e,:;1§H:::...l:..:.*r...;;.:;:.Ei-~~t:;;:::;e=-:"...:re"_i5:.;;t~r:1.:..::*E:...:;t;....t.:::.;h.:..::e;....~-:...;r-:::..:l:..rec::.a5~e=---..;;:o..!..f....:r...:::a~-tl..:...:io~a~c..!;:..tl:...;*v:..:::e-,m~a;:..:;t:.::.e-,-r a:..:.l_~~..:..:fr-"o:.!!!m;....t"'-!.h~e:......~"""co"'-!.n~tc::c.a.!.!.in=m=e,-",nt,,--~_ _ _~ ~~_~~______ l.!.-"* atmosphere to the envlronment. -(~~~~-'--------------------- SHEARON HARRIS - UNIT 1 B 3/4 9-2 Amendment No. 104 I

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

2. 2009A NRC SRO 002IMODIFIEDIF/3IFRP-S.l BASIS/N/2009A NRC SRO/029EA2.091 Given the following plant conditions:

            - The plant is operating at 100% power Current conditions:
           - BOTH Main Feedwater pumps trip
           - The RO reports that the reactor has failed to trip and cannot be tripped from the MCB
           - FRP-S.1, Response to Nuclear Power Generation/ATWS, is entered
           - The BOP trips the Main Turbine
           - Emergency Boration has been established
           - The RO is checking Pressurizer Pressure and reports that it is 2385 psig and rising Which ONE of the following lists the action required in FRP-S.1 and the basis for that action?

A. Verify Normal Pressurizer Spray Valves are open; To prevent a challenge to the Pressurizer Safety Valves B. Verify Normal Pressurizer Spray Valves are open; Boration flow will be insufficient at this Pressurizer Pressure C. Verify Pressurizer PORVs and block valves are open; To prevent a challenge to the Pressurizer Safety Valves D~ Verify Pressurizer PORVs and block valves are open; Boration flow will be insufficient at this Pressurizer Pressure {------------------------------------ Friday, December 26, 2008 1 :06:52 PM 3

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 Plausibility and Answer Analysis A Incorrect. Pressurizer spray valves are the normal pressure control components and should be open but the RNO is to verify PORVs and block valves open. The function of the pressurizer control components is to prevent reaching the pressurizer safety setpoint however the background document is concerned about boration flow not the pressurizer safeties. B Incorrect. Pressurizer spray valves are the normal pressure control components and should be open but the RNO is to verify PORVs and block valves open. The background document is concerned about boration flow though. C Incorrect. With pressurizer pressure above 2335 psig, the RNO is to verify PORVs and block valves open. The function of the pressurizer control components is to prevent reaching the pressurizer safety setpoint however the background document is concerned about boration flow, not the pressurizer safeties. o Correct. With pressurizer pressure above 2335 psig the RNO is to verify PORVs and block valves open and the background document is concerned about boration flow. KIA statement - ATWS - Ability to (a) predict the impacts of the following on the (SYSTEM) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Occurrence of a main turbinelreactor trip Importance Rating: 4.4 4.5 Technical

Reference:

FRP-S.1 Rev. 15, page 5 WOG ERG FR-S.1 HP-Rev. 1C, page 79 References to be provided: None Learning Objective: EOP-LP-3.15, Obj 5d Question origin: Modified from Bank question, OIT Development Bank FRP-S.1 (03) #3 Comments: (KIA Match) The turbine trip has occurred as part of immediate actions resulting in the increased RCS Pressure when heat removal is lost and candidate must recall from FRP-S.1 the strategy to mitigate this transient. SRO justification: Requires knowledge of FRP-S.1 and the RNO action when the Desired Pressurizer Pressure is not obtained as well as reason for performing the action. Friday, December 26, 2008 1:06:52 PM 4

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 Origin: MODIFIED Cog Level: F Difficulty: 3

Reference:

FRP-S.I BASIS Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: 029EA2.09 KIA 2: Friday, December 26, 2008 1:06:52 PM 5

Ref for SRO #2, 029ES2.09 RESPONSE TO NUCLEAR POWER GENERATION/ATWS rl Instructions 1 . . - . . -_ _ _ _- - - ' ~ ____________ Response Not Obtained ~r-~ NOTE: o Actuation of the sequencer inhibits operation of the boric acid pumps. (If the sequencer runs on Program A, the pumps can be started manually after LB-9. Otherwise, the sequencer must be reset to restore operation of the pumps) o SI flow accomplishes emergency boration.

6. Initiate Emergency Boration of RCS:

a. Check SI flow - GREATER a. GO TO Step 6c.

THAN 200 GPM

b. GO TO Step 6d.

c. Emergency borate from the c. Observe CAUTION prior to BAT: Step 7 AND GO TO Step 7.

1) Start a boric acid pump.

2) Perform any of the following (listed in order of preference):

o Open Emergency Boric Acid Addition valve: lCS-278 - o Open normal boration valves: FCV-113A FCV-113B

3) Verify boric acid flow to CSIP suction - AT LEAST 30 GPM

4) Verify CSIP flow to RCS - AT LEAST 30 GPM

d. Check PRZ Pressure - LESS d. Perform the following:

THAN 2335 PSIG

1) Verify PRZ PORVs AND block valves - OPEN

2) Maintain PRZ PORVs AND block valves open until PRZ pressure less than 2135 PSIG.

e. GO TO Step 8.

Ref for SRO #2, 029ES2.09 STEP DESCRIPTION TABLE FOR FR-S.l Step _4_ STEP: Initiate Emergency Boration of RCS PURPOSE: To add negative reactivity to bring the reactor core subcritical BASIS: After control rod trip and rod insertion functions, boration is the next most direct manner of adding negative reactivity to the core. The intended borati9n path here is the most direct one available, not requiring SI initiation, but using normal charging pump(s). Pump miniflow lines are assumed to be open to protect the pumps in the event of high RCS pressure. Several plant specific means are usually available for rapid boration and should be specified here in order of preference. Methods of rapid boration include emergency boration, injecting the BIT, and safety injection actuation. It should be noted that SI actuation will trip the main feedwater pumps. If this is undesirable, the operator can manually align the system for safety injection. However, the RWST valves to the suction of the SI pumps should be opened first before opening up the BIT valves. If a safety injection is already in progress but is having no effect on nuclear flux, then the BIT and

                   . RWST are not performing their intended function, perhaps due to blockage or leakage. In this case some other alignment using the BATs and/or non-safeguards charging pump(s) is required.

The check on RCS pressure is intended to alert the operator to a condition which would reduce charging or SI pump injection into the RCS and, therefore, boration. The PRZR PORV pressure setpoint is chosen as that pressure at which flow into the RCS is insufficient. The contingent action is a rapid depressurization to a pressure which would allow increased injection flow. When primary pressure drops 200 psi below the PORV pressure setpoint, the PORVs should be closed. The operator must verify successful closure of the PORVs, closing the isolation valves, if necessary. ~_+~_~.~_~~~~~_" ___ "_~~_~~~~~~~~_r~~".'~<_"~" __ "~_~~_ ~ _ _~_~_.~ _ _ _"~_~~._~_~_" _ _ ~_~~~~.~>~ ___~'~_ _ ~.~_¥_~ __ "~~---~---_<<_~~_"_~_~,~,_,_~_v~_

       - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -.. ~--

FR-S .1 79 HP-Rev. lC HFRSI

Original for SRO #2, 029EA2.09 QUESTIONS REPORT for OIT Development Bank

1. FRP-S.1 (03) 003 Given the following:

The plant is at 100% power.
A transient on SG level required a reactor trip, but the reactor did not trip.
The crew is.performing FRP-S.1, Response To Nuclear Power Generation I ATWS.
The crew notes that the Pressurizer pressure is 2350 psig.
Control rods are inserting in AUTO.
All Pressurizer PORV valve position indicating green lights are on and red lights are off.

Which ONE of the following actions are required in relation to the PZR PORVs? A. Verify ONLY ONE Pressurizer PORV and Block valve OPEN; reduce pressure to less than 2135 psig. B~ Verify ALL Pressurizer PORVs and Block valves OPEN; reduce pressure to less than 2135 psig. C. Verify ONLY ONE Pressurizer PORV and Block valve OPEN; reduce pressure to less than 2235 psig. D. Verify ALL Pressurizer PORVs and Block valves OPEN; reduce pressure to less than 2235 psig. Ability to determine or interpret the following as they apply to a ATWS: System component valve position indications Importance Rating: 3.4 Technical

Reference:

FRP-S.1 and SD References to be provided: None Learning Objective: Question History: 10 CFR Part 55 Content: 55.41 Comments: Program: RS Cog Level: H Difficulty: 3

Reference:

FRP-S.1 Ref. Provided?: N KeyWords: K/A 1: 029EA2.05 K/A2: -(.. - - - - - - - - - - - - - - - - - - - - Friday, December 26,20081:40:46 PM 1

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

3. 2009A NRC SRO 003INEW/H/3IEPP-OOI/y/2009A NRC SRO/055EA2.031 Given the following plant conditions:

( - The plant is operating at 100% power

              - A Station Blackout occurs
              - The crew has entered EPP-001, Loss of AC Power to 1A-SA and 1B-SB Buses, and completed Steps 1 through 6
              - Method reports that Offsite Power will be unavailable for approximately 30 minutes The following annunciators are in alarm on the MCB:
              - ALB-24/3-1, Diesel Generator A Trip
              - ALB-24/3-2, Diesel Generator A Trouble
              - ALB-24/3-3, Diesel Generator A Start Failure
              - ALB-25/3-1, Diesel Generator B Trip
              - ALB-25/3-2, Diesel Generator B Trouble
              - ALB-25/3-3, Diesel Generator B Start Failure The Outside AO reports the following annunciators locally:
              - Trip Low Press Lube Oil at the 'A' ECP
              - Trip Vibration at the 'A' ECP
              - Loss Of Both Gen Pot CKS Trip at the 'B' ECP
             - Trip Low Press Jacket Water at the 'B' ECP

(~ The USCO has reached step 7 of EPP-001. What actions are required by EPP-001 to _ (estore power to an erne[gerLcy bus and what procedure should the crew transitioll to? _ (Reference provided) A'! Start the 'A' EDG; Transition to PATH-1 B. Start the 'A' EDG; Transition to EPP-004, Reactor Trip Response C. Start the 'B' EDG; Transition to PATH-1 D. Start the 'B' EDG Transition to EPP-004, Reactor Trip Response {-------------------- Friday, December 26, 2008 1:06:52 PM 6

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 Plausibility and Answer Analysis A Correct. Since ~' EDG tripped due to a non-emergency trip then it should be tried first. The correct procedure implementation then would be to verify immediate actions of Path-1 and after those are complete transition per the Path into EPP-004. B Incorrect. The ~' EDG is the one that should be started since it is tripped due to a non-emergency trip, but the procedure directs "Return to Procedure and Step in Effect". While EPP-001 is a direct-entry procedure, the correqt transition at this point would be Path-1 to verify immediate actions. EPP-004 is the procedure that will ultimately be used, but actions of Path-1 must be verified first. C Incorrect. 'B' EDG has tripped due to an Emergency Trip and should not be started until problem is resolved. Plausible if candidate fails to recognize 'B' has tripped on an Emergency Trip. Procedure transition is correct. o Incorrect. 'B' EDG has tripped due to an Emergency Trip and should not be started until problem is resolved. Plausible if candidate fails to recognize 'B' has tripped on an Emergency Trip. Procedure transition is incorrect. KIA statement - Station Blackout - Ability to determine and interpret the following as they apply to (EMERGENCY PLANT EVOLUTION): Actions necessary to restore power (~ Importance Rating: 3.9 4.7

          . TechnicaL

Reference:

           . EPP-001 Rev. 31, page 6 References to be provided:        Steps 1-7 of EPP-001 (pages 1-6)

Learning Objective: EOP-LP-3.7, Obj 3 Question origin: NEW Comments: SRO justification: Requires SRO judgement and knowledge of procedure selection to determine correct procedure transition. Origin: NEW Cog Level: H Difficulty: 3

Reference:

EPP-001 Ref. Provided?: Y KeyWords: 2009A NRC SRO KJA 1: 055EA2.03 KJA2: Friday, December 26, 2008 1:06:52 PM 7

Ref for SRO #3, 055EA2.03 (Ref provided) C Continuous Use CAROLINA POWER & LIGHT COMPANY SHEARON HARRIS NUCLEAR POWER PLANT PLANT OPERATING MANUAL VOLUME 3 PART 4 PROCEDURE TYPE: Emergency Operating Procedure (EOP) NUMBER: EOP-EPP-OOl ( TITLE: . LOSS OF AC POWER TO lA-SA AND lB-SB BUSES

 --------------------,-----------------------~                                     ._-----

EOP-EPP-OOl Rev. 31 Page 1 of 66

Ref for SRO #3, 055EA2.03 (Ref provided) LOSS OF AC POWER TO lA-SA AND lB-SB BUSES rl'------------' Instructions Response Not Obtained

CAUTION Critical Safety Function Status Trees should be monitored for information only. Function Restoration Procedures should NOT be implemented unless directed by this procedure. NOTE: Steps 1 AND 2 are immediate action steps.

1. Verify Reactor Trip: Manually trip reactor.

o Check for any of the following: o Trip breakers RTA AND BYA - OPEN o Trip breakers RTB AND BYB - OPEN o Neutron flux - DECREASING

2. Verify Turbine Trip:

a. Check for any of the a. Manually trip turbine from following: MCB.

o All turbine throttle valves - SHUT o All turbine governor valves - SHUT ~~~---------------------------------.----- EOP-EPP-OOl I Rev. 31 I Page 3 of 66

Ref for SRO #3, 055EA2.03 (Ref provided) LOSS OF AC POWER TO lA-SA AND 1B-SB BUSES ( rlI.....--._ _ _ _Instructions ~ Response Not Obtained I.....--.----~.-

3. Check If RCS Isolated:

a. Check PRZ PORVs - SHUT a. WHEN PRZ pressure less than 2335 PSIG, THEN shut PRZ PORVs.

b. Check letdown isolation b. Perform the following:

valves - SHUT:

1) Shut all orifice 1CS-1 (LCV-460) isolation valves:

1CS-2 (LCV-459) 1CS-7 1CS-8 1CS-9

2) Shut letdown isolation valves:

1CS-1 (LCV-460) 1CS-2 (LCV-459)

c. Verify excess letdown valves - SHUT:

1CS-460 1CS-461 EOP-EPP-001 I Rev. 31 I Page 4 of 66

Ref for SRO #3, 055EA2.03 (Ref provided) LOSS OF AC POWER TO lA-SA AND lB-SB BUSES

             ---L..--______
             ---1              Instructions
                                                     ---' L..--______---'1-Response Not Obtained

4. Verify AFW Flow AND Control SG Levels:

a. Verify AFW Flow - GREATER a. Perform the following:

THAN 210 KPPH

1) Verify TDAFW pump -

RUNNING

2) Adjust TDAFW pump speed controI"ler as necessary to increase flow.

3) Verify TDAFW pump discharge pressure -

GREATER THAN SG PRESSURE

4) Verify AFW valves -

PROPERLY ALIGNED

b. Any level - GREATER THAN b. Maintain AFW flow greater 25% [40%] than 210 KPPH until level greater than 25% [40%] in at least one intact SG.

C~ c. Control AFW flow to maintain all intact levels between 25% and 50% [40% and 50%]

5. Evaluate EAL Network Using Entry Point X.

6. Verify AC Emergency Bus Cross-Ties to Non-Emergency AC Buses - OPEN o Verify any cross tie to Bus lA-SA - OPEN o Breaker 104 o Breaker 105 o Verify Any cross tie to Bus lB-SB - OPEN

    ~~~~~_~~ , _           0     Breaker-lUI:

..::.:.....::~.:.......::.t:'=-----~---n-=~-rJ"-.::;-------------------;---~-~- o J)reaKer It) ~------I------------------'--- EOP-EPP-OOI I Rev. 31 I Page 5 of 66

Ref for SRO #3, 055EA2.03 (Ref provided) LOSS OF AC POWER TO lA-SA AND lB-SB BUSES

                -I" " " - - - - - - - - - ' "

Instructions Response Not Obtained

7, Energize AC Emergency Buses using EDGs:

a. Check EDGs lA-SA AND lB-SB a. Do NOT start EDG OR close

                                 - AVAILABLE                                                                output breaker until problem corrected.

o EDG emergency trips - CLEAR (NOT PRESENT) Emergency stop any running EDG with tripped output o EDG output breakers - breaker. NORMAL (NOT TRIPPED) IF NO EDG available. THEN GO TO Step 8. -- b, Check any EDG - RUNNING b. Perform the following as necessary to start EDGs (listed in order of preference):

1) Manually start EDGs.

2) Actuate SI.

GO TO Step 7d.

c. GO TO Step 7e.

d. Check any EDG - RUNNING d. GO TO Step 8.

e. Check any AC emergency bus e. Perform the following:

                                 - ENERGIZED:

Manually close running o lA-SA bus voltage EDG output breaker at o lB-SB bus voltage MCB OR locally perform at switchgear: EDG A: Breaker 106 EDG B: Breaker 126

2) GO TO Step 7h.

f. Implement function restoration procedures as required.

g. RETURN TO procedure and step in effect.

h. Check any AC emergency bus h. Emergency stop the running

                                 - ENERGIZED:                                                               EDG(s).

o lA-SA bus voltage GO TO Step 8. o lB-SB bus voltage

i. Implement function restoration procedures as required.

___=-=-...:.::.:.+____ _-lR;:IEO-ET3iU~R..;iNba-T-°!4-F;:..*g:ii-Er:~°!-:t=c;...e_d_u_r~_e_a_n_d___~_. --------------------r--..-~. .- J::.,.'_'

-r

~--~-------------------------------------Ij---- EOP-EPP-OOI 1 Rev. 31 I Page 6 of 66

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

4. 2009A NRC SRO 004INEW/F/2/T.S. 3.8.3.1/N/2009A NRC SROI057AG2.2.401 Given the following plant conditions:

             - The plant is operating at 100% power
             - The Instrument Bus S-IV Inverter failed at 1330
             - Instrument Bus S-IV was reenergized from its alternate source at 1430 Which ONE of the following identifies the required action of Tech Specs for the failed Inverter?

Energize Instrument Bus S-IV from its Inverter connected to its: Source Required tomorrow by: A AC. Bus 1330 B. AC. Bus 1430 C~ D.C. Bus 1330 D. D.C. Bus 1430 Plausibility and Answer Analysis A Incorrect. This is plausible because the normal power supply to the inverter is A. C. but T. S. require it to be connected to the D. C. bus. The time to restore is correct b8.SBd an time of loss. B Incorrect. This is plausible because the normal power supply to the inverter is A. C. but T. S. require it to be connected to the D. C. bus. The time is also incorrect. This is 24 hours from the time the bus was reenergized. This would require the candidate to believe the actions required are series vice parallel. C Correct. This is correct because the action requires it to be connected to the D. C. bus and a 24 hour action exists from the time of loss (1330). o Incorrect. This is correct because the action requires it to be connected to the D. C. bus. The time is incorrect. This is 24 hours from the time the bus was reenergized. This would require the candidate to believe the actions required are series vice parallel. {----------------------------------------- Friday, December 26, 2008 1:06:52 PM 8

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Loss of Vital AC Inst. Bus - Ability to apply technical specifications for a C' system. Importance Rating: 3.4 4.7 Technical

Reference:

Tech Spec 3.8.3.1 pg 3/4 8-16,8-17 (pages 346-347) References to be provided: None Learning Objective: ADEL-LP-2.7, Obj 1b; 120 Volt UPS, Obj 5 Question origin: NEW Comments: SRO justification: Requires detailed knowledge of greater than >1 hour Tech Spec actions and how they are applied. Origin: NEW Cog Level: F Difficulty: 2

Reference:

T.S.3.8.3.1 Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: 057AG2.2.40 KIA 2:

,~-,~

{~----------------------------------------------- Friday, December 26, 2008 1 :06:53 PM 9

Ref for SRO #4, OS7AG2.2.40 ElECTRICAL PO~ER SYSTEMS 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3~1 .The following electrical busas shall ba energized in the specified manner W1th tie breakers open between redundant buses within the unit:

a. Division A ESF A.C. Buses consisting of:

1. 6900-volt Bus lA-SA.

2. 4S0-volt Bus lA2-SA.

3. 480-volt Bus tAl-SA.

b. Division B ESF A.C. Buses cansisting_ of:

l. 6900-vott Bus lB-5B.

2. 480-volt Bus 182-SB.

3. 480-volt Bus 183-58.

c. US-vo 1t A. C. Vi tal Bus lDP-LA-SI energi ztd froll 1ts associ ated inverter connecttd to 12S-volt D.C. Bus DP-~-SA-,

d. US-volt A.C. Vital Bus lOP-LA-SlIt energize<! f1"Ol'l its associated inverter c:onnec:te4 to 1.25"v~1t- D. C. Bus DP-lA-SAx I I. US-volt A. C. Vital Bus lDP-lS-SII energize<! f1"01l its associatld f.

inverter connected to ~-volt D.C. Bus DP-lB-SS-, l18-volt A.C. Vital Bus lDP-U-SIV energized f1'"Cll its associated inverter connected to 12S-volt D.C. Bus DP-lB-5B-,

g. US-volt D.C. Bus OP-LA-SA energized froc Emergency Battery lA-SA and charger lA-SA or 18-SA, and

n. lZS-vol t. D. C. Bus DP-lB-SB energized frem Emergency Battery 18-SB and charger 18-Sa or lA-SB APPLICABILITY: MODES l, Z, 3, and 4.

              -Two.invert.rs may be disconnected from their 125-valt D.C. bus for up to 24* hours as necessary L f~r ~e Pl.trp~s._ of perform;"g an equ,l izi ngenar;_ on

___~_~_.~-~--~--~ei-r--aHOci aUG - EMrgenc:y-B-a-ttery Pl"Ovtded:*~--t1t the; r----vt~tmsfl-_an---.n.-r-~------u _,_' giz.a and (2) the vital buses Issociatad with the other Emergency Battery _ _ __

    ----an-___.rgtnct-fMllt~tnH-mv.tUI"'SCnCt*---clJfttr~.tyo-ln--a--'

c1 atlCt US-yO 1t D. C. bus. SHEARON HARRIS - UNIT 1 3/4 8-l6 wq

Ref for SRO #4, OS7AG2.2.40 ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTIQN OPERATING r, LIMITING CONDITION FOR OPERATION ACTION:

                    ~      With one of tn. required divbfons of A..c. £SF husas not fully energized, reenergizi the division within 8 hours o~ be in at least HOT STANDBY within the nut 6 hours and 1n C01.D SHUTDOWN within the fallowing 30 hours.                                                                                           .

b. With OM US-volt A.C. vital bus notenergizad froll its associatea inverter, reenergize the 118-volt A.C. vital bus within % hours ar be in at least HOT STANDBY within the next 6 hours and in CCt.D SHU'1'DOW within the tallowing.lO hours.

c. With one US-volt A. C. vital bus nat energiZed fratl its usociatea 1nvlner =nnec:tact to its Issociat.ct D. C. bus, re-Inergize the US-volt A. C. vital bus through its associatld inve1"ter connect.ed to its usociated D.C. bus within 24 hours or be in at least HOT STANDBY within the next 6 hours and 1n COLD SHUTDQWN within tne following 30 hours.

d. With lither US-volt D*. C. bus lA-SA ar 11-58 not energizad f'l"Oll fts

                        .. us<<faUd E..rgwncy &lte-ry. l"ftiWr;tu tn. &.c.                                        l'Ju6. f100a In **SD-clated c.rgency Battery within Z hours o~ be in It leut HOT STANDBY "ith1n the next 6 *hours Il1d in COLD SHUTDOWN within the following 30 haurs.

SURVEIlLANC! REQUIREMENTS 4.8.3.1 The speci f1 ad buses sna 11 be deteJ"llli ned ene1"'gi zed in the requi red sanner at least one. per 7 days by verifying correct b~.k.r alignment and indicated voltage~n the busls * ... ~~'" .. SHEARON HARRIS - UNIT 1 3/4 8-17

                                                                .. -...'-.-.s-:;e-.;.~-.....
           ,'- ----~-,-.~-.;.-.- - - -........... --:------,..~s:~                           c............. 'r; =: ~. 4::e-   .,4 ...... ',: ',.u... '"

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

5. 2009A NRC SRO 005IMODIFIEDIHI3/AOP-025/N/2009A NRC SROI058AG2.4.201 Given the following plant conditions:

( - The plant is in Mode 3

              - The 'A' MDAFW Pump is under clearance for motor replacement
              - A loss of DP-1 B-SB occurs
              - The crew enters AOP-025, Loss of One Emergency AC Bus (6.9KV) or One Emergency DC Bus (125V)

Which ONE of the following describes the operation of the TDAFW Pump if a start signal occurs and the limiting Tech Spec action required as a result of these conditions? The TDAFW Pump will: Tech Spec Action A. start and continue to run place the plant in Hot Shutdown in 7 hours B. start and continue to run ALL required Mode changes are suspended C. start and trip on overspeed place the plant in Hot Shutdown in 7 hours D~ start and trip on overspeed ALL required Mode changes are suspended Plausibility and Answer Analysis ( A Incorrect. This is plausible because if ~' Train DC had been lost this would be the response but with the Joss of 'B' Train,controJ power-is lost and an overspeed trip will occur. The Tech Spec action listed is from 3.0.3 and is plausible because all three AFW pumps are inoperable but AFW has specific actions for all three inoperable which suspends required mode changes. B Incorrect. This is plausible because if ~' Train DC had been lost this would be the response but with the loss of 'B' Train, control power is lost and an overspeed trip will occur. The Tech Spec action is correct. C Incorrect. There is a NOTE in the AOP-025 which addresses the TDAFW Overspeed Trip. The Tech Spec action listed is from 3.0.3 and is plausible because all three AFW pumps are inoperable but AFW has specific actions for all three inoperable which suspends required mode changes. D Correct. There is a NOTE in the AOP-025 which addresses the TDAFW Overspeed Trip. The Tech Spec action is correct. f--.------------------------------------~-- Friday, December 26, 2008 1:06:53 PM 10

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Loss of DC Power - Knowledge of operational implications of EOP warnings, cautions and notes. ( Importance Rating: 3.8 4.3 Technical

Reference:

AOP-025 Rev. 25, page 45 Tech Specs 3.7.1.2 pg 3/4 7-4 (page 306) References to be provided: None Learning Objective: AOP-LP-3.25, Obj 3d Question origin: Modified from bank, OIT Exam Bank DP (08) #1 Comments: (KIA Match) Matches KA because a Note in AOP-025 alerts operator that TDAFW will start and trip on overspeed in this situation. SRO justification: Requires application of notes associated with application of Tech Spec action items. Origin: MODIFIED Cog Level: H Difficulty: 3

Reference:

AOP-025 Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: 058AG2.4.20 KIA 2:

~~-------------------------------------

Friday, December 26, 2008 1:06:53 PM 11

Ref for SRO #5, 058AG2.4.20 PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two motor-driven auxiliary feedwater pumps, e~ch capable of being powered from separate emergency buses. and

b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1. 2. and 3. ACTION:

a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.

b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours.

c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. (NO-TE:

LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status. Following restoration of one AFW train, all applicable LCOs apply based on the time the LeOs initially occurred.) SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a. At least once per 92 days on a STAGGERED TEST BASIS by:

1. Demonstrating that each motor-driven pump satisfies performance requirements by either:

a) Verifying each pump develops a differential pressure that (when temperature - compensated to 70°F) is greater than or equal to 1514 psid at a recirculation flow of greater than or equal to 50 gpm (25 KPPH) , or b) Verifytng_~e~c:h_pu_rnP deyel ops§cji fferenti a1 pressure U * --lnat- hlhcn temperature compensated to 70°F) is

 /                                           greater than or equal to 1259 psid at a flow rate of

~--.~------------------------g~r~e~a+t~er~tntl~arrl~o~r~e~qmU7arl~t~o~4~3~Orngpmr~1IT(72~15~K~pmpH~)'.------------ SHEARON HARRIS - UNIT 1 3/4 7*4 Amendment No. 93

Ref forSRO #5, 058AG2.4.20 I LOSS OF ONE EMERGENCY AC BUS (S.9KV) OR ONE EMERGENCY DC BUS (125V) I (

            ~                   INSTRUCTIONS                                  RESPONSE NOT OBTAINED
                                                                                                                                    ~

3.4 Loss of DP-1 B-SB Emergency DC Bus (125V)

04. MONITOR inverter output voltage and instrument readings for indication of proper voltage on instrument bus.

05. CHECK power to ALL instrument 05. REFER TO AOP-24, Loss of buses is ON. Uninterruptible Power Supply.

os. DISPATCH operator to inspect the following for fault conditions:

DP-1 B-SB Emergency DC Bus
125V Battery Chargers 1A-SB AND 1B-SB

07. CONTACT Maintenance as necessary to initiate repair or other corrective actions.

NOTE

Loss of DP-1 B-SB will result in all equipment on that side becoming inoperable

                     " from the loss of EDG and Load Sequencer or DC control power.

Local manual operation will be necessary for any breakers that have lost DC control power. [C.1]
Loss of B-SB Emergency DC Bus will make the Turbine Driven AFW Pump inoperable due to loss of power to 1MS-72.
Loss of DP-1 B-SB results in losing the control panel for the TDAFW pump. If started, the TDAFW pump will trip on overspeed without any alarm or indication of trip and throttle valve.

08. OPEN all load breakers on DP-1 B-SB.

-,~-~.',~--- -~--,~-. -~.-~-~. - ~---------~-----~-~.--~---~-------"-~-

                                                                         .--,-.-----~-'---- ----"--'---~-----~-~,---~-'--"--- .. -"---.-.---""

AOP-025 I Rev. 25 I Page 45 of 55

Original for SRO #5, 058G2.4.20 QUESTIONS REPORT for OIT Exam Ban.k

1. DCP (08) 001 Given the following:

The unit is at 100% power.
The following alarm is received:
ALB-015, 4-4, 125 VDC EMER BUS A TROUBLE
DP-1A-SA Bus voltage indicates 65 VDC and lowering.
Battery Charger 1A-SA is tripped, and the cause has not been determined.
The USCG has declared DC Bus "A" inoperable.
The crew has entered AOP-025, Loss of ONE Emergency AC Bus (6.9 KV) or One Emergency DC Bus (125 VDC).

Which ONE of the following describes the impact on TDAFW Pump operability, and the action(s) required to restore the DC Bus in accordance with AOP-025? A. TDAFW Pump remains operable; Direct the AO to immediately place the standby battery charger in service in accordance with OP-156.01, DC Electrical Distribution. Maintenance support is NOT required prior to restoration. B. TDAFW Pump remains operable; Notify Maintenance to determine the cause and initiate repairs prior to restoring the bus in accordance with AOP-025. C. TDAFW pump is inoperable; Direct the AO to immediately place the standby battery charger in service in accordance with OP-156.01, DC Electrical Distribution. Maintenance support is NOT required prior to restoration. D~ TDAFW pump is inoperable; Notify Maintenance to determine the cause and initiate repairs prior to restoring the bus in accordance with AOP-025.

~~~--------------------

Friday, December 26, 2008 1:43:47 PM 1

QUESTIONS REPORT for OIT Exam Bank D is correct. AOP-025 states that both the EDG and the TDAFW Pump are inoperable. ( Actions for loss of DC Bus. Would not place spare charger in service at 65 volts and decreasing on battery, require greater than 105 volts to place spare battery charger in service. A and B are incorrect because TDAFW is declared inoperable, even if the other train is available to supply DC power. Actions for B are correct C is incorrect because actions to place the spare charger in service will not be performed at this battery bus voltage Conduct of Operations: Knowledge of system status criteria which require the notification of plant personnel. Tier 1 Group 1 Importance Rating: SRO 3.3 Technical

Reference:

ALB-15, 4-4, AOP-025 Proposed references to be provided to applicants during examination: None Learning Objective: DCP text Obj 8 10 CFR Part 55 Content: 43.5, 43.2 Comments: 10CFR55.43(b).5 because the SRO must assess conditions and determine appropriate course of action. 10CFR55.43(b).2 because technical specification operability must be determined Program: S Cog Level: H Difficulty: (- Ref. Provided?: N 3.75

Reference:

KeyWords: ALB-015-4-4 KIA l~ 05802.1.14 KIA 2: NO {-------------------------------------- Friday, December 26, 2008 1:43:47 PM 2

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

6. 2009A NRC SRO 006INEW/F/3/AOP-0281N12009A NRC SROI077AA2.101 Given the following plant conditions:

            - The plant is operating at 47% power
            - Method reports a large disturbance occurring on the grid
            - Efforts are in progress to stabilize the grid
            - The crew enters AOP-028, Grid Instability The following conditions are observed:

Time Grid Frequency (Hz) _ 0107 59.6 0110 59.2 0113 58.9 0116 58.7 0119 58.5 0121 58.3 Which ONE of the following describes the EARLIEST time that the Reactor must be tripped in accordance with AOP-028 and what is the basis for that Reactor Trip? A'! 0118; Continued operation in this condition could lead to high temperatures in the generator and subsequent insulation degradation B. 0121; Continued operation in this condition could lead to high temperatures in the gen~rator a_nd ~ubsequent Lnsl)latiO[tdegradation C. 0118; Provides reactor core protection against DNB as a result of underfrequency on more than one RCP D. 0121; Provides reactor core protection against DNB as a result of underfrequency on more than one RCP ~----------------------------------------- Friday, December 26, 2008 1:06:53 PM 12

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 Plausibility and Answer Analysis In AOP-028 the body of the procedure provides specific guidance on generator conditions that require a reactor trip.

               - Generator frequency less than 59 Hz for greater than or equal to 5 minutes
               - Generator frequency less than 58.4 Hz
              - Turbine speed less than or equal to 1752 RPM A reactor trip is required at 0118 when frequency is less than 59 Hz for 5 minutes. A reactor trip setpoint is also exceeded at 0121 but this is not the earliest.

A Correct. Correct time and basis B Incorrect. Rx trip was required at 0118. Basis is correct. C Incorrect. Correct time. Basis is plausible as this is the basis for RCP Underfrequency Trip. D Incorrect. Incorrect time. Basis is plausible as this is the basis for RCP Underfrequency Trip. KIA statement - Generator Voltage and Electric Grid Disturbances - Ability to determine and interpret the following as they apply to ABNORMAL PLANT EVOLUTION): Generator overheating and required actions Importance Rating: 3.6 3.8 Technical

Reference:

AOP-028 Rev. 25, page 4 AOP-028-BD Rev. 9, page 7 References to be provided: None Learning Objective: AOP-LP-3.28, Obj 6 Question origin: NEW Comments: SRO justification: Must assess plant conditions and understand overall strategy and actions required to mitigate event and the bases for those actions Origin: NEW Cog Leve1: F Difficulty: 3

Reference:

AOP-028 Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: 077AA2.10 KIA 2: {.~---------------------- Friday, December 26, 2008 1:06:53 PM 13

Ref for SRO #6, 077AA2.10 GRID INSTABILITY

  ~
  -1'-----_            INSTRUCTIONS

_ _---' RESPONSE NOT OBTAINED

                                                          '-----_           _ _-----II             L 3.0 OPERATOR ACTIONS NOTE

This procedure contains no immediate actions.
The loss of Off-Site power may require the initiation of the Emergency Plan [C.1]

01. REFER TO PEP-11 0, Emergency Classification and Protective Action Recommendations, AND enter EAL Network at entry point X. [C.1]

NOTE

If frequency drops suddenly and power is greater than P-7, the reactor will trip automatically when RCP frequency decreases to 57.5 Hz, resulting in a turbine trip.
Operation of electrical motors with voltage below the normal band will increase stator current and change torque loading. Component trips, insulation and/or bearing damage, shorts, grounds, or blown fuses may result. The probability of

(' damage is increased with lowering voltage and increased operating time. [C.2] CAUTION

Operation of the unit between 59.0 and 58.4 Hz should be limited to 5 minutes, after which time the generator must be taken off-line.
Operation below 58.4 Hz is not allowed and the generator must be taken off-line immediately.
2. CHECK Main Generator indications 02. GO TO Step 3.

for ANY of the following conditions: o

Generator frequency less than 59 Hz for greater than or equal to 5 minutes o
Generator frequency less than 58.4 Hz o
Turbine speed less than or equal to 1752 RPM o a. TRIP the Reactor, ANn ~nTn ~ IATj::J_1

-t--~---l---------------------t--- AOP-028 I Rev. 25 I Page 4 of 19

Ref for SRO #6, 077AA2.10 GRID INSTABILITY-BASIS DOCUMENT Section 3.0-0perator Actions Step Description C2

Operation of the unit between 59.0 and 58.4 Hz should be limited to 5 minutes, after which time the generator must be taken off-line.
Operation below 58.4 Hz is not allowed and the generator must be taken off-line immediately.

This caution alerts the operator to the consequences of operating equipment in an underfrequency condition. Operation outside of established limits could cause high temperatures in the generator and possibly lead to insulation degradation and generator damage. This condition could also lead to damage of major electrical equipment when operating at lower speeds and higher currents. 2 I: Check Main Generator indications for ANY of the following conditions:

Generator frequency less than 59 Hz for greater than or equal to 5 minutes
Generator frequency less than 58.4 Hz
Turbine speed less than or equal to 1752 RPM

a. Trip the Reactor and Go to EOP-Path-1.

RNO: Go to step 3. This continuous action step checks for indications of underfrequency that will require the generator to be taken off line quickly to protect the generator and major electrical equipment. If any condition applies, frequency conditions require rapidly taking generator off-line. The operator is directed to trip reactor and go to EOP-Path-1. This will result in tripping the turbine and taking the generator off-line. 3 I: Check both emergency buses energized. RNO: Refer to AOP-025, Loss of One Emergency AC Bus (6.9kV) or One Emergency DC Bus (125V). This is a continuous action step to monitor emergency buses during condition of grid instability. Emergency buses are powered from the UAT when the generator is online during normal operation, or from the SUT during startup and shutdown. Grid instability may result in loss of one or both emergency buses. AOP-025 addresses actions for loss of emergency buses.

~----l----------------t----~-

AOP-028-BD I Rev. 9 I Page 7 of 24

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

7. 2009A NRC SRO 007INEWIHI2/AOP-012/N/2009A NRC SROI051AA2.021 Given the following plant conditions:

( - The plant is operating at 79% power

              - The following alarms are received:
                   - ALB-020-2-4A, Condsr Pre Trip Low Vacuum
                   - CTMP-7-1, Cooling Tower 1 Level HIILO
                   - ALB-021-8-5, Computer Alarm Circ Water Systems
              - Computer Alarm on the Circ Water system is determined to be due to Condenser Pit High Level
              - The BOP determines that condenser backpressure is 6.6 inches Hg in Zone 2 and rising
              - The Turbine Building Operator reports that there is a failure of an expansion joint on the Circulating Water system
              - 'A' Condenser Vacuum Pump is in service and verified running
              - The crew enters AOP-O 12, Partial Loss of Condenser Vacuum Which ONE of the following actions is required?

A. Start up standby Vacuum Pump and maintain condenser backpressure less than 7.5 inches Hg B. Reduce turbine load using AOP-038, Rapid Downpower until power is less than 60% C~ Trip the reactor, .go to PATH-1, and trip ALL Circulating Water pumps as time allows D. Trip the turbine and continue with actions in AOP-012 to stabilize condenser vacuum Plausibility and Answer Analysis A Incorrect. This action would be correct if there were indications of a problem with the in service Condenser Vacuum Pump, but it is not put in service for every loss of condenser vacuum event. B Incorrect. 60% is a threshold setpoint in AOP-012, but it is for determining reactor trip setpoints, not for AOP-038 implementation. C Correct. This action is correct for CW system expansion joint failures causing a loss of condenser vacuum at > 10% power. o Incorrect. Action is directed by AOP-012, but only when less then P-10. {~------------------------------------------ Friday, December 26, 2008 1 :06:53 PM 14

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Loss of Condenser Vacuum - Ability to determine and interpret the following as they apply to ABNORMAL PLANT EVOLUTION): Conditions requiring ( reactor and/or turbine trip Importance Rating: 3.9 4.1 Technical

Reference:

AOP-012 Rev. 18, pages 4-6 References to be provided: None Learning Objective: AOP-LP-3.12, Obj 5 Question origin: NEW Comments: SRO justification: Requires assessing plant conditions and then recalling what action written into AOP-012 is required. Origin: NEW Cog Level: H Difficulty: 2

Reference:

AOP-OI2 Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: 051AA2.02 KIA 2: Friday, December 26, 2008 1 :06:53 PM 15

Ref for SRO #7, 051AA2.02 PARTIAL LOSS OF CONDENSER VACUUM ~ 11'---------.-_ INSTRUCTIONS _ _---' RESPONSE NOT OBTAINED .~

                                                     ~--------------------------~

3.0 OPERATOR ACTIONS NOTE This procedure contains no immediate actions.

01. CHECK Turbine -IN OPERATION 01. GO TO Step 5.

02. CHECK Condenser pressure in both 2. PERFORM the following:

Zones less than:

7.5 inches Hg absolute AND Turbine first stage pressure is o a. IF Reactor power is greater than P-10 (10%),

greater than 60% TURBINE THEN TRIP Reactor LOAD AND GO TO EOP-Path-1.

              -OR-

5 inches Hg absolute AND 0 b. IF Reactor power is less than Turbine first stage pressure is less P-10 (10%),

than 60% TURBINE LOAD THEN TRIP Turbine AND GO TO Step 5.

3. REDUCE Turbine load as necessary to maintain_ Condenser vaCl!um _u_stng ONE of the following:

o

GP-006, Normal Plant Shutdown from Power Operation to Hot Standby o
AOP-038, Rapid Downpower

04. CONTINUE Turbine load reduction until directed otherwise by Unit SCO based on the following:

Cause of vacuum loss identified and corrected
Vacuum stable or increasing
Plant conditions require Reactor or Turbine trip

05. CHECK Condenser Vacuum Pump - 05. START Standby Condenser Vacuum OPERATING. Pump per OP-133, Main Condenser

_ Air Removal System. AOP-012 I Rev. 18 I Page 4 of 22

Ref for SRO #7, 051AA2.02 PARTIAL LOSS OF CONDENSER VACUUM

 ~                 INSTRUCTIONS                                RESPONSE NOT OBTAINED                L 11'-----_ _ _---'                                                                             ---'I

( L . . . . . - -_ _ _ 3.0 OPERATOR ACTIONS

06. DISPATCH Operator(s) to locally perform actions of Attachment 1, Local Actions for a Loss of Condenser Vacuum.

7. VERIFY the following valves - SHUT:

o

1CE-44 7, Condenser Vac Breaker o
1CE-475, Condenser Vac Breaker

08. CONTACT Radwaste Control Room to determine if recent equipment operations using auxiliary steam or condensate may have caused loss of vacuum.

09. CHECK Circulating Water Pumps - 09. GO TO Step 11.

ANY TRIPPED. 010. VERIFY associated pump discharge valve - SHUT. NOTE If a Circulating Water Pump has tripped, it is not considered available until the cause of the trip has been identified and corrected. 011. CHECK ALL available Circulating 011. START ALL available Circulating Water Pumps - RUNNING. Water Pumps per OP-138.01, Circulating Water System. 012. CHECK at least ONE Condensate 12. OPEN the following valves to Booster Pump - RUNNING. establish a flow path for the Gland SealExhaustCondense~ o

1CE-290, 1CE-293 Inlet Line Isolation o
1CE-294, 1CE-293 Outlet Line Isolation AOP-012 I Rev. 18 I Page 5 of 22

Ref for SRO #7, 051AA2.02 PARTIAL LOSS OF CONDENSER VACUUM

          ---i              INSTRUCTIONS                                                RESPONSE NOT OBTAINED
                                                                                                                          ~

3.0 OPERATOR ACTIONS

13. CHECK BOTH of the following conditions EXIST: (indicates complete failure of a Circulating Water System expansion joint) [A.1]

0 a. CHECK ALB-021-8-5 in ALARM 0 a. GO TO Step 15. due to Condenser Pit High Level.

b. CHECK EITHER of the following 0 b. GO TO Step 15.

conditions EXISTS: 0

A known expansion joint failure 0
CTMP-7-1, COOLING TOWER 1 LEVEL HI/LO alarm due to low level.

14. PERFORM the following:

0 a. CHECK Reactor power is greater 0 a. IF Turbine is in operation, ( than P-10 (10%). THEN TRIP Turbine AND GO TO Step 14.c. 0 b. TRIP Reactor AND GO TO EOP-Path-1. (Perform substeps 14.c -14.f as time allows.) 0 c. TRIP ALL Circulating Water Pumps. 0 d. TRIP Normal SW Pumps. 0 e. REFER TO AOP-022, Loss of Service Water. 0 f. EXIT this procedure. 015. CHECK for major unisolable leak in 015. GO TO Step 18. Circulating Water System - EXISTS.

    - - -                            ~-- .--~--   ------- ---- - - - - - -----       --

-l~ AOP-012 I Rev. 18 I Page 6 of 22

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

8. 2009A NRC SRO 0081NEW1H/31PEP-11ON12009A NRC SRO/061AG2.4.411 Given the following plant conditions:

              - The plant is operating at 100% power
              - The crew has noted indications that an RCS leak is in progress and has entered AOP-016, Excessive Primary Plant Leakage The following alarms and indications are observed:
              - Plant Vent Stack #1 WRGM Effluent is alarming at 4.8E4 uCi/sec and rising
              - Charging Pump 1B Room Area Radiation monitor alarming at 1200 times normal and rising
              - GFFD is in alarm and has increased by 80,000 CPM over the last 20 minutes
              - 1RM-1 CR-3589-SA, CNMT HI Range Accident Monitor, is alarming at 14.7 R/hr and rising
              - 1RM-1 CR-3590-SB, CNMT HI Range Accident Monitor, is alarming at 18.4 R/hr and rising Which ONE of the following is the EAL classification to be declared for this event?

(Reference provided) A. EAL 2-1-1 B. EAL 2-1-2 C!' EAL 2-1-3 ( D. EAL2-1-4 Plausibility and Answer Analysis A Incorrect. Plausible since the conditions for this classification have been met by the increase in GFFD, but this is not the most limiting EAL. B Incorrect. Plausible since the conditions for this classification have been met due to the Table 2 area rad monitor reading >1200 times normal, but this is not the most limiting EAL. C Correct. Multiple radiation alarms require evaluating Cnmt Hi Range Accident Monitors. With one above the setpoint the correct classification is to identify two fission product barriers breached. D Incorrect. Plausible since it is possible to reach a General Emergency based solely on Plant Vent Stack WRGM, but the threshold for this classification has not yet been reached. Friday. December 26. 2008 1:06:53 PM 16

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - ARM System Alarms - Knowledge of the emergency action level thresholds and classifications. ( Importance Rating: 3.9 4.1 Technical

Reference:

Side 1 of EAL Flow Path Rev. 05-1 References to be provided: Side 1 of EAL Flow Path Rev. 05-1 Learning Objective: Question origin: NEW Comments: SRO justification: Requires assessing plant conditions and making appropriate determination of EAL classification. This task is only performed by SRO qualified individuals. Origin: NEW Cog Level: H Difficulty: 3

Reference:

PEP-110 Ref. Provided?: Y KeyWords: 2009A NRC SRO KIA 1: 061AG2,4,41 KIA 2: Friday, December 26, 2008 1:06:53 PM 17

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

9. 2009A NRC SRO 009IBANKlH/3IFRP-P.1/N/2009A NRC SROIWE08EG2.4.47/

Given the following plant conditions:

              - The plant is operating at 100% power
              - A LOCA occurs
              - FRP-P.1, Response to Imminent Pressurized Thermal Shock, is in progress Thirty (30) minutes after the initiating event, the following conditions are observed:
              - RCS pressure has lowered to 600 psig and is now stable
              - Tcolds have lowered to 220°F and are now stable
              - Containment pressure is 12.3 psig and slowly rising Which ONE of the following identifies requirements for this event in accordance with FRP-P.1?

Soak Requirement Subsequent Cooldown Limit A'! Soak required B. Soak required C. Soak NOT required D. Soak NOT required (. Plausibility and Answer Analysis A Correct. The trend over the last 30 minutes would warrant implementation of procedural guidance to perform a 1 hr soak and subsquent cooldown limit of 50°Flhr. B Incorrect. Plausible since a one hour soak is required but 1OO°Flhr is not established in FRP-P.1. 1OO°Flhr is the cooldown rate used in other EOPs (EPP-009, EPP-020, EPP-021). C Incorrect. Plausible since a soak is not always required in FRP-P.1 but with the existing cooldown in the last 30 min, a soak is required. Subsquent cooldown limit of 50°Flhr is correct. o Incorrect. Plausible since a soak is not always required in FRP-P.1 but with the existing cooldown in the last 30 min, a soak is required. 1OO°Flhr is not established in FRP-P.1 but this is plausible since 1OO°Flhr is the cooldown rate used in other EOPs (EPP-009, EPP-020, EPP-021). f~--------------------------------------- Friday, December 26, 2008 1 :06:54 PM 18

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - RCS Overcooling - PTS - Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference ( material. Importance Rating: 4.2 4.2 Technical

Reference:

FRP-P.1 Rev. 18, pages 44 & 46 References to be provided: None Learning Objective: EOP-LP- 3.14, Obj 2d Question origin: Bank Comments: SRO justification: Requires assessing plant conditions to recognize conditions warranting implementation of specific strategies from plant procedures. Origin: BANK Cog Level: H Difficulty: 3

Reference:

FRP-P.l Ref Provided?: N KeyWords: 2009A NRC SRO KIA 1: WE08EG2.4.47 KIA 2: {,/~---------------------- Friday. December 26. 2008 1 :06:54 PM 19

Ref for SRO #9, WEOBEG2.4.47 RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK ~ Instructions Response Not Obtained ~ I~------------~ ~------------~.- CAUTION Following an excessive cooldown, reactor vessel stress must be relieved to enhance and maintain vessel integrity. Do NOT perform any actions that increase pressure OR cause an RCS cooldown until the soak is complete.

32. Determine RCS Soak Requirements:

a. RCS cooldown rate - GREATER a. GO TO Step 34.

THAN 100°F IN ANY SIXTY MINUTE PERIOD

b. Perform one hour RCS soak:

o Maintain RCS temperature stable. o Maintain RCS pressure stable. o Perform actions of other procedures that do NOT cause an RCS cooldown OR increase

               .p res sure.

EOP-FRP-P.l I Rev. 18 I Page 44 of 50

Ref for SRO #9, WEOBEG2.4.47 RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK ( rl'--------' Instructions Response Not Obtained t--

33. Establish Subsequent Cooldown Limits:

a. RCS subcooling monitor - a. Maintain RCS pressure AND AVAILABLE temperature within the limits of Figure 2 OR Figure 3 based on CNMT conditions.

GO TO Step 33c.

b. Maintain RCS subcooling between 10°F and 190°F

[40°F and 158°F].

c. Maintain RCS cooldown rate less than 50°F in any sixty minute period.

34. RETURN TO Procedure And Step In Effect.

                                      - END -

( EOP-FRP-P.l I Rev. 18 I Page 46 of 50

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

10. 2009A NRC SRO OIOINEWIHI4/CSFST/N/2009A NRC SROIWE13EA2.1I Given the following plant conditions:

( - The USCO is evaluating FRPs for implementation.

             - Containment pressure is 0.8 psig
             - The following Steam Generator conditions exist:
                - 'A' SG Pressure = 1175 psig
                - 'A' SG Level        = 79%
                - 'B' SG Pressure =1235 psig
                - 'B' SG Level        =65%
                - 'c' SG Pressure = 1100 psig
                - 'c' SG Level        =23%

Based on the CSFST for Heat Sink, which ONE of the following identifies the FRP that should be addressed first? A'! FRP-H.2, Response to Steam Generator Overpressure B. FRP-H.3, Response to Steam Generator High Level C. FRP-H.4, Response to Loss of Normal Steam Release Capability D. FRP-H.5 Response to Steam Generator Low Level Plausibility and Answer Analysis ( A Cqrrect. This yel/ow path procequre would be addressed first by the status trea B Incorrect. Plausible since FRP-H.3 is a yel/ow path that will need to be addressed for the ~' Steam Generator but current conditions will require addressing the overpressure condition first. C Incorrect. Plausible since FRP-H.4 is a yel/ow path that will need to be addressed for the ~' Steam Generator but the current conditions require implementing H.2 o Incorrect. Plausible since FRP-H.5 is a yel/ow path that will need to be addressed for 'c' Steam Generator but the current conditions require implementing H.2 Friday, December 26, 2008 1:06:54 PM 20

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Steam Generator Overpressure - Ability to determine and interpret the following as they apply to (EMERGENCY PLANT EVOLUTION): Facility conditions and ( selection of appropriate procedures during abnormal and emergency operations. Importance Rating: 2.9 3.4 Technical

Reference:

EOP-CSFST Rev. 9, page 2 of 3 (Side 1) References to be provided: None Learning Objective: EOP-LP-3.11, Obj 4a Question origin: NEW Comments: SRO justification: Requires assessing plant conditions and then prescribing the procedure required to mitigate event in progress Origin: NEW Cog Level: H Difficulty: 4

Reference:

CSFST Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: WE 13EA2. 1 KIA 2: (~

-~- -------------

Friday, December 26, 2008 1 :06:54 PM 21

I r--" I II, Ref for SRO #10 I W!.. .1 I: CSFST ~~~fRltICALITY CREA TER TH~ 5,; CSF-1 RED REACTOR COOLANT TEMPERATURE GO TO FRP-S,l AND INTERMEDIATE RANGE SUR POSITIVE ORANGE GO TO PRESSURE LIMITATIONS FRP-S.1 INTERMEDIATE RANGE SUR MORE POSITI'IE mAN _0.2 OPM YELLOW POVt£R RANGI GO TO LESS THAN 5' SOURCE RANGE FRP-S,2 NOT ENERGIZED INTERMEDIATE RANGE SlJR MORE NEGATIVE THAN ~O.2 OPM GREEN CSF-SAT INTERMEDIATE RANGE SUR ZERO OR NEGATIVE SOURCE RANGE SUR POSITIVE YELLOW GO TO SOURCE RANGE FRP-S,2 ENERGIZED SOURCE RANGE SUR ZERO OR NEGATl GREEN CSF-SAT COR NG CSF-2 CORE EXIT Te GREATER THMl12000F RED GO TO FRP-C,1 PRESSURE 1 (PSIG) R'.tJS FULL RANGE LESS THAN 39% RED GO TO CORE EXIT res FRP-C.l GREATER THAN 730°1'" R>uS FULL RANG< I CREA"fER THAN 39%

                                                                                                                                                           ,"\lb,

,)@'11 ORANGE GO TO NO Reps ,,~*MY FRP-C,2 RUNNING R\tlJS FULL RANGE LESS THAN 39%

ORANGE GO TO CORE EXIT res FRP-C,2 LESS 1HAN 730'F

                      ~cs SUBCOOlJNG LESS THAN                                                                           R'.1JS FULL RANGE GREATER THAN 39%

YELLOW 1~~~~~=~ GO TO FRP-C,3 ORANGE GO TO CORE EXH Te AT LEAST ONE RCP RUNNING FRP-C,2 TEMPERATURE LESS THAN 12Po'F YELLOW (. F) GO TO FRP-C,3 GREEN CSF-SAT HEA RED GO TO FRP-H,1 PRESSURE IN ANY Sa GREATER mAN 1230 PSiG YELLOW GO TO FRP-H,2 NARROW RANGE LE~ IN ANY SG GREATER THAN 78:0: YELLOW GO TO FRP-H,3 NARROW RAN PRESSURE IN AU. SG", LEVEL IN AT SG GREATER LESS THAN 1230 PSIG PRESSURE IN ANY SG GREATER THAN 1170 PSiG YELLOW GO TO NARROW RANGE FRP-HA LEva. IN All SG!I NARROW RANGE

                                                                             ~                                           LEVEL IN ANY SG                           YELLOW r

LESS THAN 25:{ 4;):r.i GO TO PRESSURE IN AU. SGs FRP-H.5 LESS THAN 1170 PSIG NARROW RANGE LEVEL IN AlL SG!I GREATER THAN 25,;: r4;)' GREEN CSF-SAT EOP- REV, 9 PAGE 2 of 3

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

11. 2009A NRC SRO OIIINEW/F/2IPATH-I/N/2009A NRC SRO/006A2.13/

An Inadvertent Safety Injection has occurred. The crew is implementing PATH-1 and has reached the point of terminating Safety Injection in PATH-1. The following conditions exist:

            - Conditions are met for terminating Safety Injection
            - 'A' CSIP is running
            - 1CS-214, Normal Miniflow Common Isolation Valve, is SHUT and will not open
            - 1SI-3 and 1SI-4, BIT Outlet Valves are OPEN
            - 1CS-235 and 1CS-238, Charging Line Isolation Valves are SHUT
            - The crew has just shut FCV-122, Charging Flow Control Valve in accordance with Path-1 Which ONE of the following describes, in order, the actions to be taken in accordance with PATH-1?

A. SHUT 1SI-3 & 1SI-4, Fully OPEN FCV-122 to 100%, OPEN 1CS-235 and 1CS-238, Throttle SHUT FCV-122 to maintain less than 60 GPM flow B. SHUT 1SI-3 & 1SI-4, Throttle OPEN FCV-122 to 30%, OPEN 1CS-235, and 1CS-238, ( Throttle OPEN FCV-122 to establish at least 60 GPM flow C~ OPEN 1CS-235, and 1CS-238, Throttle OPEN FCV-122 to 30%, SHUT 1SI-3 & 1SI-4, Throttle OPEN FCV-122 to establish at least 60 GPM flow D. OPEN 1CS-235 & 1CS-238, Fully OPEN FCV-122 to 100%, SHUT 1SI-3 & 1SI-4, Throttle SHUT FCV-122 to maintain less than 60 GPM flow Friday, December 26, 2008 1 :06:54 PM 22

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 Plausibility and Answer Analysis A Incorrect. Actions taken are in the incorrect order, plausible if candidate believes that SI can not be run in parallel with charging. This is normally the case, but failure of a miniflow isolation is an exception to this to ensure the running ECCS equipment is not deadheaded. Final goal is also incorrect, it is to maintain at least 60 gpm, not less than 60 gpm (common mistake). B Incorrect. Actions taken are in the incorrect order, plausible if candidate believes that SI can not be run in parallel with charging. This is normally the case, but failure of a miniflow isolation is an exception to this to ensure the running EGGS equipment is not deadheaded. C Correct. FCV is opened a minimal amount prior to isolating SI flow path in order to prevent deadheading a running CSIP. o Incorrect. FCV is throttled to 30% open first to provide minimal flow and then goal is to establish and maintain AT LEAST 60 GPM flow. KIA statement - Emergency Core Cooling - Ability to (a) predict the impacts of the following on the (SYSTEM) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Inadvertent SIS actuation 1mportance Rating:. 3.9 4.2* Technical

Reference:

EOP-GUIDE-1 Rev. 23, pages 29 & 31 References to be provided: None Learning Objective: Question origin: NEW Comments: SRO justification: Requires assessing plant conditions and recalling what strategy written into plant procedures is required Origin: NEW Cog Level: F Difficulty: 2

Reference:

PATH-l Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: 006A2.13 KIA 2: { ..-.------------------------------------------- Friday, December 26, 2008 1:06:54 PM 23

Ref for SRO #11, 006A2.13 PATH -1 GUIDE ---1 Il~ ____________ Instructions ~ Response Not Obtained L---

19. Reset S1. IF any train of SI will NOT reset at MCB, THEN reset at SSPS using Attachment 12.

20. Manually Realign Safeguards Equipment Following A Loss Of Offsite Power.

(Refer to Attachment 2.)

21. Stop All But One CSIP.

22. Check RCS Pressure - STABLE OR GO TO EPP-009, "POST LOCA INCREASING COOLDOWN AND DEPRESSURIZATION",

Step 1.

23. Isolate High Head SI Flow:

a. Open normal miniflow a. Observe NOTE prior to isolation valves: Step 24 AND GO TO Step 24.

1CS-182 1CS-196 1CS-210 1CS-214

b. Shut BIT outlet valves: b. Locally shut OR isolate valves.

1SI-3 1SI-4 1SI-3 (A-230-FX32-W6-S1) 1SI-4 (A-230-FX32-W3-S2)

c. Verify cold leg AND hot leg c. Locally shut valves.

injection valves - SHUT lSI-52 (A-250-GY38-W2-S6) lSI-52 lSI-86 (A-230-FX25-W4-N3) 1SI-86 1SI-107 (A-245-FV20-W6-N9) 1SI-107

d. Observe CAUTION prior to Step 25 AND GO TO Step 25.

EOP-GUIDE-1 I Rev. 23 I Page 29 of 96

Ref for SRO #11, 006A2.13 (' FOLDOUT A o RCP TRIP CRITERIA IF both of the following occur. THEN stop all RCPs: o SI flow - GREATER THAN 200 GPM oRCS pressure - LESS THAN 1400 PSIG o AFW SUPPLY SWITCHOVER CRITERIA IF CST level decreases to less than 10%. THEN switch the AFW water supply to the ESW system using OP-l37. "AUXILIARY FEEDWATER SYSTEM". Section 8.1. o RHR RESTART CRITERIA IF RCS pressure decreases to less than 230 PSIG in an uncontrolled manner. THEN restart RHR pumps to supply water to the RCS. o ALTERNATE MINI FLOW OPEN/SHUT CRITERIA o IF RCS pressure decreases to less than 1800 PSIG. THEN verify alternate miniflow isolation OR miniflow block valves - SHUT o IF RCS pressure increases to greater than 2200 PSIG. THEN verify alternate miniflow isolation AND miniflow block valves - OPEN EOP-GUIDE-1 Rev.*23 Page 30 of 96

Ref for SRO #11, 006A2.13 PATH-1 GUIDE

             ~I~_ _             Instructions                                                                              ~

( _ _----, ""--_ _ _ _----'r-Response Not Obtained NOTE: The following step contains an SI termination sequence for which CSIP normal miniflow is not available. The charging flow control valve is opened a minimal amount prior to isolating the BIT to ensure the running CSIP is not deadheaded.

24. Establish Minimum Charging Flow AND Isolate BIT Flow:

a. Shut charging flow control valve:

FK-122.1

b. Open charging line isolation valves:

1CS-235 1CS-238

c. Set charging flow controller demand position

( to 30%.

d. Shut BIT outlet valves: d. Locally shut eR isolate valves.

lS1-3 lS1-4 lS1-3 (A-230 FX32-W6-S1) lS1-4 (A-230-FX32-W3-S2)

e. Verify cold leg AND hot leg e. Locally shut valves.

injection valves - SHUT lSI-52 (A-250-GY38-W2-S6) lSI-52 lS1-86 (A-230-FX25-W4-N3) lS1-86 lSI-107 (A-245-FV20-W6-N9) lSI-107

f. Establish and maintain at least 60 GPM flow through CSIP .

g. Observe CAUTION prior to Step 26 AND GO TO Step 26.

..--~-~--.-.- -~-.~---.--~-.~~--~~ .... -~-.-~.-- ..----.. .. ..

                                                           ~-~~ ~ -.-~-~---~.-~---.----~-   ..--.-- ..----.-.-..- .. --~-~-.--4--- ..--.. ~--

EOP - GUIDE-1 J Rev. 23 I Page 31 of 96

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

12. 2009A NRC SRO o12INEWIHI3/0WP-RP/N/2009A NRC SRO/012A2.0l/

Given the following plant conditions:

          - The plant is operating at 100% power
          - The AFD Target value is (-)1.5%
          - Reactor Engineering is performing a flux map
          - PT-456, Pressurizer Pressure Channel II, has failed and OWP-RP-02 is in place
          - Repair of PT-456 will not be completed for another 17 hours The following conditions occur:
          - ALL normally lit bistable lights for N-43 on TSLB-4 are OFF
          - N-43 continues to indicate 100% power
          - I&C reports that N-43 repairs will take 15 hours and the instrument must be deenergized to facilitate repairs Which ONE of the following describes the action that should be completed for these conditions?

A. N-43 should be bypassed while repairs are completed on PT-456 B. Reduce Power to less than 75% within 4 hours C. Trip the bistables for N-43 within 6 hours D~ Place the plant in Mode 3 within 13 hours Plgusibility and Answer Analysis A Incorrect. This is plausible because Action 2.b. says the inoperable channel may be bypassed but this is only for surveillance testing and is only allowed for up to 4 hours. B Incorrect. This is plausible because Action 2.c requires power to be reduced to 75% within 4 hours or perform QPTR (flux map) evety 12 hours. Flux map is in progress and can continue. C Incorrect. This is plausible because Action 2.a requires tripping bistables but precautions of OWP require checking bistables when taking instruments out of service to preclude a plant trip. o Correct. Because taking the instrument out of service will produce a plant trip, Action 2.a can not be performed. 6 hours of Action 2.a and 7 hours of TS. 3.0.3 is a total of 13 hours. Friday, December 26, 2008 1:06:54 PM 24

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Reactor Protection - Ability to (a) predict the impacts of the following on the (SYSTEM) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Faulty bistable operation Importance Rating: 3.1 3.6 Technical

Reference:

OWP-RP Rev. 15, pages 96-99 Tech Spec 3.3.1 pg 3/4 3-2, 3/4 3-6&7 (pages 150,154, and 155) . Tech Spec 3.0.3 pg 3/4 0-1 (page 103) References to be provided: None Learning Objective: RPS Obj 12 Question origin: NEW Comments: SRO justification: Requires evaluation of T.S. Table 3.3-1 Action 2.a and determining that based on plant procedures it can not be complied with, resulting in T.S. 3.0.3 entry. Origin: NEW Cog Level: H Difficulty: 3

Reference:

OWP-RP Ref. Provided?: N KeyWords: 2009A NRC SRO KIA I: 012A2.01 KIA 2: -E._~--------------------- Friday, December 26, 2008 1 :06:54 PM 25

Ref for SRO #12, 012A2.01 OWP-RP-25 Sheet 1 of 4 EIR Number: ( 1. OWP - --'R_P_-_2..::.5_ _ __ w/o Number: Clearance Number:

2. System: Nuclear Instrumentation

3. Component: POWER RANGE N-43

4. Scope: LCO action required due to inoperable Channel 3 Power Range Nuclear Instrumentation

5. Applicable Requirements: 3.3.1 (Modes 1 and 2), 4.2.1.1 and 4.2.4.2 (Mode 1 above 50% RATED THERMAL POWER)

6. Precautions: 1) Ensure only one channel out of service at a time.

2) This procedure does not alter the input to the P-8 or P-10 permissives.

7. Component lineups completed per attached /

sheet (s) . Signature Date

8. Testing required on redundant equipment while the component is inoperable. Perform EST-915 once per 12 hrs if Rx power is greater than 75% with one Channel inoperable. Perform OST-1039 once per 12 hrs if Rx power is greater than 50%.

9. Testing/Action required to restore operability. (N/A if tracked on EIR)

OST-1021, 1022 or 1033 / OST-1004 / OST-1039 (above 50% RTP) / MST-I0046 / Signature Date

10. Component lineups restored per attached /

sheet(s) . Signature Date

11. Remarks:

12. Reviewed By:

Superintendent - Shift Operations Date After receiving the final review signature, this OWP becomes a QA RECORD and should be submitted to Document Services. ~-----~IHO~~~V~P~-R~P~=============+========~R~e~v.~15~========F==========PP~ag~e~9m6~O~f~1TIOK4TI-----

Ref for SRO #12, 012A2.01 OWP-RP-25 Sheet 2 of 4 Bistable/status Light Lineup position for Restored Inoperabili ty position Component ID or Number Initial/verified Initial/Verified NOTE: This OWP must be performed in order to prevent possible spurious rod motion or level control swings. On MAIN CONTROL BOARD: ROD BANK SELECTOR Switch MANUAL - - -I - - - MANUAL _ _1 _ - FW Reg Byp Valve Controllers: FK-479.1 MANUAL _ _I _ - MANUAL _ _1 _ - FK-489.1 MANUAL _ _I _ - MANUAL _ _1 _ - FK-499.1 MANUAL I

                                                  --- ---      MANUAL              I In PIC 3 on Card C3-822:

NOTE: Concurrent verification is preferred while tripping bistables. BS3 (TB/432Cl OT~T) TEST I NORMAL _ _1 _ - BS4 (TB/432C2 OT~T C-3) TEST _ _1__ NORMAL _ _1 _ - On DETECTOR CURRENT COMPARATOR Drawer: UPPER SECTION Switch PRN43 I NORMAL _ _1 _ - LOWER SECTION Switch PRN43 _ _1__ NORMAL _ _1 _ - c On MISCELLANEOUS CONTROL AND INDICATION PANEL: ROD STOP BYPASS Switch . BYPASS PR N43 I OPERATE _ _1 _ - POWER MISMATCH BYPASS BYPASS Switch PR N43 I OPERATE I On COMPARATOR AND RATE Drawer: COMPARATOR CHANNEL DEFEAT Switch N43 _ _1__ NORMAL _ _1 _ - On Power Range Drawer N43A NOTE: The purpose of the sign installed below is to alert personnel of tripped bistables that may not be obvious at the NI drawer. The wording in quotations is the recommended wording, but similar words may also be used. Sign stating "Bistables Tripped - OWP-RP in Affect" Installed I Removed _ _1 _ - ,-=/ . ( ~-f~O~W~P~-R~P~============F========OR~evT.~15~=======T========~P~~9~7~f~1~04rlr----age 0

Ref for SRO #12, 012A2.01 OWP-RP-25 Sheet 3 of 4 Bistable/Status Light Lineup ( Position for Restored Inoperability position Component ID or Number Initial/Verified Initial/Verified In POWER RANGE N43: NOTE: Concurrent verification is preferred in the following Step. At the rear of N43 Drawer A, disconnect P312 from J312 DISCONNECTED / CONNECTED / (On completion of the above lineup, check the following.) On TSLB-3: C TRIP O/TEMP L1T ENERGIZED / DE-ENERGIZED / TB432C1 (Window 9-1) C RUN BK O/TEMP L1T ENERGIZED / DE-ENERGIZED / TB432C2 (Window 9-3) On TSLB-4:

Circle required state as determined by present plant conditions.

PR LO PWR HI FLUX NC 43P *ENERGIZED OR (Window 5-3) ENERGIZED / DE-ENERGIZED / PR HI PWR HI FLUX NC 43R (Window 6 - 3) ENERGIZED / DE-ENERGIZED /

         **May require manual reset of rate trips locally at drawer.

PR HI FLUX RATE NC 43U/K ** (Window 7-3) ENERGIZED / DE-ENERGIZED / On BYPASS PERMISSIVE LIGHTS Panel: PR OVERPWR ROD WTHDRWL BLK BYPASS CHAN III (Window 3-7) ENERGIZED / DE-ENERGIZED / i:----F~O~Wmp~-~R~P~============~========~R~e~V~.T15r=========~========~pCa~g~e~9~8~o~f1~O~4~1-

Ref for SRO #12, 012A2.01 OWP-RP-25 Sheet 4 of 4 Bistable/Status Light Lineup position for Restored Inoperability position Component ID or Number Initial/Verified Initial/Verified On ERFIS Computer: (After status lights have been checked, perform the following using the DR function. ) ANM0122M - PWR RNG DELETED FROM RESTORED TO CHANNEL N43 QI I-MIN AVG PROCESSING / PROCESSING / On MAIN CONTROL BOARD:

         + Circle appropriate position as determined by plant conditions.

ROD BANK SELECTOR Switch MAN/AUTO + / MAN/AUTO + / FW Reg Byp Valve Controllers:

         + Circle appropriate position as determined by plant conditions.

FK-479.I MAN/AUTO + / MAN/AUTO + / FK-489.I MAN/AUTO+ / MAN/AUTO + / FK-499.I MAN/AUTO + / MAN/AUTO + / ( ~~----hlo~v~vre*p~-R~P~============~======~R~e~V~.1~5~=======T========~p~a~g~e~9~9~Ofr.1~Oa411-----

k**, i I f\ .--., I Ref for SRO #12, 01(2A_ J1 II TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM

                                                     ,TOTAL NO. CHANNELS  CHANNELS   APPLICABLE FU TIONAL UNIT                                  OF CHANNELS i

TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 2 1 2 1 2 1 2 I 2 3', 4", 5" 9

2. I i Power Range, Neutron Flux

a. High Setpoint 4 2 3 1, 2 2

b. Low Setpoint 4 2 3 1###, 2 2

3. I Power Range, Neutron Flux 4 2 3 1, 2 2 High Positive Rate

4. I Power Range, Neutron Flux, 4 2 3 1, 2 2 High Negative Rate Intermediate Range, Neutron Flux 2 1 2 1###, 2 3

Ii Source Range, Neutron Flux

a. Startup 2 1 2 2## 4

b. Shutdown 2 1 2 3, 4, 5 5 Overtemperature 8T 3 2 2 1, 2 6 7*1 Overpower 8T

8. 3 2 2 1, 2 6

9. I Pressurizer Pressure--Low (Above P-7) 3 2 2 1 6(1) 10j Pressurizer Pressure--High 3 2 2 1, 2 6

11. Pressurizer Water level--High 3 2 2 1 6 (Above P-7)

HE~RON HARRIS - UNIT 1 3/4 3-2 Amendment No. 84 '

Ref for SRO #12, 012A2.01 TABLE 3.3-1 (Continued) TABLE NOTATIONS ( 'When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.

         "Whenever Reactor Trip Breakers are to be tested.
       ##Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
      ###Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(l)The applicable MODES for these channels noted in Table 3.3-3 are more restrictive and. therefore. applicable. ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement. restore the inoperable channel to OPERABLE status within 48 hours or be . in HOT STANDBY within the next 6 hours. ACTION 2 - With the number of OPERABLE channels one less than the Total (- Number of Channels. STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours.

b. The Minimum Channels OPERABLE requirement is met:

however. the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1. and

c. Either. THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to Jess than or equal to 85% of RATED THERMAL POWER within 4 hours: or. the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2.

--=--- - - - - - - - - - - - -

SHEARON HARRIS - UNIT 1 3/4 3-6 Amendment No. 101

Ref for SRO #12, 012A2.01 TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued) ACTION 3 - Wit.n tne number of channels OPERABLE one less tnan the Minimum Channels OPERABLE requirement and witn the THERMAL POWER level:

a. Below the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint, restore the inoperable cnannel to OPERABLE status prior to increasing THERMAL POWER above tne P-6 Setpoint, and

b. Above tne P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL roWER, restore the inoperable cnannel to OPERABLE status prior to increasing THERMAL roWER above 10% of RATED THERMAL POWER. ACTION 4 - Witn the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes. ACTION 5 - a. With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoper-able channel to OPERABLE statu.s within 48 hours!; or open the Reactor Trip SYltem breakers, and verify compliance with the s,hutdown margin' requirements of Specification 3.1.1.2 within 1 hour and at least once per 12 hours thereafter.

b. With; no channeh OPERABLE, open the Ruetor Trip System breakers within 1 hour and suspend all operations involving positive reactivity changes: Verify compliance with the SHUTDOWN MARCIN requirements of Specification 3.1.1.2 within 1 hour and at least once per 12 hours thereafter.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP andlor POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours, and

b. The Hinimum Channels OPERABLE requirement is met; nowever, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 7 - With less than the Minimum Number of Channels OPERABLE, within 1 hour determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

"~,~------------------------------------------

SHEARON HARRIS - UNIT 1 3/4 3-7 Amendment No. 15

3/4.0 APPLICABILITY Ref for SRO #12, 012A2.01 LIMITING CONDITION FOR OPERATION ( 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met. 3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION reqUirements is not required unless otherwise noted in the ACTION statement. 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:

a. At least HOT STANDBY within the next 6 hours,

b. At least HOT SHUTDOWN within the following 6 hours, and

c. At least COLD SHUTDOWN within the subsequent 24 hours.

Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual ( __0 specificat ions. This specification is not applicable in MODE &orG. 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Conditions for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements or that are part of a shutdown of the unit. Exceptions to these requirements are stated in the individual specifications. 3.0.5 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 3.0.1 above for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. t*~OO ________________________ SHEARON HARRIS - UNIT 1 3/4 0-1 Amendment No. 84

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

13. 2009A NRC SRO o13/NEWIH/3/T.S. 3.7.1.1/N/2009A NRC SRO/039A2.02/

Given the following plant conditions: ( - The plant is operating at 100% power The following indications are received:

            - Pressurizer level is lowering
            - RCS pressure is lowering
            - Charging flow is increasing
            - RCS Tavg is lowering
            - Turbine first stage pressure has lowered 25 psig and is stable
            - Electrical output has lowered 30 MW and is stable
            - A field operator reports that a Safety Valve is lifting on the 'A' SG Which ONE of the following describes the procedure required to mitigate this event and the Power Range High Flux Trip Setpoints required to allow continued operation in accordance with Tech Specs?

Procedure to be entered Power Range High Flux Trip Setpoints A. AOP-015, Secondary Load Rejection 33% B. AOP-015, Secondary Load Rejection 50% C. Enter AOP-038, Rapid Down Power 33% Dt Enter AOP-038, Rapid Down PQwer 50% Plausibility and Answer Analysis A Incorrect. AOP-015 is plausible as it contains actions for a safety valve opening but a load rejection has not occurred as indicated by primary plant parameters. 33% is plausible because this is the action for two inoperable safeties. B Incorrect. AOP-015 is plausible as it contains actions for a safety valve opening but a load rejection has not occurred as indicated by primary plant parameters. 50% is . correct. C Incorrect. AOP-038 is correct. 33% is plausible because this is the action for two inoperable safeties. o Correct. AOP-038 is correct. 50% is correct. Friday, December 26, 2008 1 :06:54 PM 26

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Main and Reheat Steam - Ability to (a) predict the impacts of the following on the (SYSTEM) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Decrease in turbine load as it relates to steam escaping from relief valves Importance Rating: 2.4 2.7 Technical

Reference:

AOP-038 Rev. 18, page 3 Tech Spec 3.7.1.1 pg 3/4 7-1, 3/4 7-2 (pages 303-304) AOP-015 Rev. 17, page 3 References to be provided: None Learning Objective: AOP-LP-038, Obj 3 Question origin: NEW Comments: SRO justification: Requires knowledge of a 4 hour T.S. Origin: NEW Cog Level: H Difficulty: 3

Reference:

T.S.3.7.1.1 Ref Provided?: N KeyWords: 2009A NRC SRO KIA 1: 039A2.02 KIA 2: Friday, December 26, 2008 1:06:54 PM 27

Ref for SRO #13, 039A2.02 RAPID DOWN POWER 1.0 PURPOSE This procedure provides guidance to lower Reactor power and Turbine load in a rapid fashion based on plant conditions or Tech Spec requirements while maintaining all plant parameters within necessary limits. This procedure can be used to lower power to a reduced output or take the plant to Turbine shutdown. This procedure shall not be used for planned down powers. For planned down powers, Reactor Engineering should be contacted for a Reactivity Plan. 2.0 ENTRY CONDITIONS 2.1 Plant Conditions require a rapid reduction in power level to preclude a plant trip or in lieu of a plant trip. 2.2 The following are potential conditions (NOT a complete listing) that may require a rapid power reduction to maintain the plant in operation while recovery efforts are executed or take the plant to Turbine shutdown:

Impending loss of a Main Feedwater Pump
Main Turbine or Generator problems
Impending loss of control fluids

           --   Condenser vacuum abnormalities

Any condition that would require a complete power reduction in less than 60 minutes 2.3 Any condition requiring greater than 5 MW/min load reductions.

AOP-038 I Rev. 18 I Page 3 of 23

Ref for BRO #13, 039A2.02 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All ma in steam 1i ne Code safety va 1ves associ ated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2. APPLICABILITY: MODES 1, 2, and 3. ACTION:

a. With one or more main steam line Code safety valves inoperable, operation may proceed provided. that within 4 hours. either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Tab1e 3.7-1; otherwise. be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDO~JN within the following 6 hours.

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by the Inservice _lesti ng~rogram. -E_---------------------- SHEARON HARRIS UNIT 1 3/4 71 Amendment No. 127

Ref for SRO #13, 039A2.02 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LOOP OPERATION MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) 1 50 2 33 3 16 ~~"------------------------------------- SHEARON HARRIS - UNIT 1 ~/4 I

7-2 Amendment No. 107 I

Ref for SRO #13, 039A2.02 SECONDARY LOAD REJECTION 1.0 PURPOSE Provides actions to respond to and recover from a secondary load rejection, turbine runback or malfunction of the DEH controller. 2.0 ENTRY CONDITIONS This procedure is entered upon a secondary load rejection, turbine runback or malfunction of the DEH controller with the following exception:

AOP-010, Feedwater Malfunctions, should be entered if turbine runback was caused by a loss of Main Feed Pump or Loss of BOTH Heater Drain Pumps.

(

 ~~ ~--~--~--~----~~-.~--~-~~---.---~~-~-~-.-.-..     --------..-.~--.----~-.-~~-~-~.-------.-...--..--.-

AOP-015 I Rev. 17 I Page 3 of 19

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

14. 2009A NRC SRO o14IBANKIHI4IFRP-J.1/N/2009A NRC SROI076G2.4.301 Given the following plant conditions:

( - FRP-J.1, Response to High CNMT Pressure, is in progress

               - 'A' CNMT Spray Pump is under clearance for motor replacement
               - 'A' ESW Booster Pump is running
               - The breaker for 1B2-SB has tripped and cannot be reclosed
               - 1SW-116, AH-2&3 SW Return Orifice Bypass Isol, is OPEN Which ONE of the following describes the action required by FRP-J.1 and the basis for this action?

A. Notify Chemistry to sample ONLY the A-SA ESW Return Header; Activity may have entered the A train of ESW from CNMT B:' Notify Chemistry to sample BOTH A-SA and B-SB ESW Return Headers; Activity may have entered BOTH trains of ESW from CNMT C. Shut the ESW isolation valves to ONLY A-SA CNMT Fan Coolers; To prevent activity from entering the 'A' train of ESW from CNMT D. Shut the ESW isolation valves to BOTH A-SA and B-SB CNMT Fan Coolers; To prevent activity from entering any train of ESW from CNMT (~ , Plausibility and Answer Analysis A Incorrect Since 1SW-116 is not shut sampling the A-SA header is required. Chemistry would sample only the A-SA header if 'B' train ESW booster pump and orifice bypass isolation had operated properly. B-SB booster pump did not start because of fault on 1B2-SB bus. Basis is correct for sampling A-SA train. B Correct. Action and bases are correct due to no running Containment Spray pumps and the failure of both trains of ESW Booster pumps to operate properly. C Incorrect. This action would be correct if one or more Containment Spray pumps were running and the B-SB ESW Booster pumps had operated properly. o Incorrect. This action would be correct if one or more Containment Spray pumps were running. Since none are, the priority is to maintain the cooling to the fan coolers, despite the potential that activity may enter the lower pressure ESW system.

~. ,------------------------

Friday, December 26, 2008 1 :06:55 PM 28

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Service Water - Knowledge of events related to system ('" operations/status that must be reported to internal organizations or outside agencies. Importance Rating: 2.7 4.1 Technical

Reference:

FRP-J.1 Rev. 13, pages 7-9 References to be provided: None Learning Objective: EOP-LP-3.13, Obj 3 Question origin: Slightly modified from bank question Comments: SRO justification: Requires recalling what strategy is written into the body of the procedure to address the plant conditions given. Origin: BANK Cog Level: H Difficulty: 4

Reference:

FRP-J.l Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: 076G2.4.30 KIA 2: ~---------------------------------------------- Friday, December 26, 2008 1:06:55 PM 29

Ref for SRO #14, 07682.4.30 RESPONSE TO HIGH CONTAINMENT PRESSURE ( ~ Instructions Response Not Obtained ~

                                            ~------------~.-

7. Check CNMT Spray Pump Status:

a. Check CNMT spray pumps - a. GO TO Step 7c.

BOTH RUNNING

b. GO TO Step 9.

c. Check CNMT spray pumps - c. GO TO Step 7e.

ONE RUNNING

d. GO TO Step 8.

e. Check all of the following: e. Coordinate with plant operations staff AND 0 ESW booster pump A-SA chemistry to sample "A" ESW

                  - RUNNING                       return header for activity.

0 Orifice bypass isolation valve lSW-116 - SHUT

f. Check all of the following: f. Coordinate with plant operations staff AND 0 ESW booster pump B-SB chemistry to sample "B" ESW

                  - RUNNING                       return header for activity.

0 Orifice bypass isolation valve ISW-1l8 - SHUT

g. Observe NOTE prior to Step 11 AND GO TO Step 11.

EOP-FRP-J.l I Rev. 13 I Page 7 of 12

Ref for SRO #14, 076G2.4.30 RESPONSE TO HIGH CONTAINMENT PRESSURE

il~ ____________ Instructions ~ 1....--_ Response __ Not_ Obtained _ _--.1 I-

8. Check ESW Booster Pump Status:

a. Check all of the following: a. GO TO Step Bc.

o ESW booster pump A-SA

                  - RUNNING o    Orifice bypass isolation valve lSW-1l6 - SHUT o    ESW booster pump B-SB
                  - RUNNING o     Orifice bypass isolation valve lSW-1l8 - SHUT

b. GO TO Step 11.

c. Check any ESW header with c. Perform the follOWing:

both of the follOWing:

1) Shut valves in one of a Associated ESW booster the following ESW fan

( pump - RUNNING cooler trains: o Associated orifice o Train A: bypass isolation valve

                  - SHUT                                     ISW-91 ISW-92 lSW-97 lSW-I09 o    Train B:

lSW-225 lSW-227 lSW-110 ISW-98

2) Coordinate with plant operations staff AND chemistry to sample for activity in the ESW return header with open CNMT fan cooler ESW isolation valves.

3) GO TO Step 10.

EOP-FRP-J.l I Rev. 13 I Page 8 of 12

Ref for SRO #14, 076G2.4.30 RESPONSE TO HIGH CONTAINMENT PRESSURE

  ~
  -a~  ______________

Instructions ~

                                            ~    ______________

Response Not Obtained L---

~r-

9. Check ESW Booster Pumps:

a. Check both of the a. Shut CNMT fan cooler ESW following: isolation valves:

o ESW booster pump A-SA ISW-91

                   - RUNNING                       lSW-92 ISW-97 o     Orifice bypass                  ISW-I09 isolation valve ISW-116 - SHUT

b. Check both of the b. Shut CNMT fan cooler ESW following: isolation valves:

o Check ESW booster pump lSW-225 B-SB - RUNNING ISW-227 lSW-110 o Orifice bypass lSW-98 isolation valve lSW-1l8 - SHUT (~ EOP-FRP-J.l I Rev. 13 I Page 9 of 12

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

15. 2009A NRC SRO 015/NEWIH/3/FRP-J.l/N/2009A NRC SROIl03G2.4.4/

The crew has transitioned to PATH-1, Entry Point C and is presently evaluating the ( RHR System capable of Cold Leg Recirculation. The following conditions exist:

               - Offsite Power has been lost
               - 'B' EDG immediately tripped
               - CNMT Pressure is 17 psig and rising
               - CNMT High Range Rad Monitors are in alarm
               - CNMT Wide Range Sump Level is reading 211 inches
               - ALB-001-2-3, Spray Pump A Autostart Fail/Override, is in alarm
               - ALB-001-2-4, Spray Pump A O/C Trip or Close Circuit Trouble, is in alarm Which ONE of the following is the required procedure transition and when will a transition back to PATH-1 be allowed?

A'I FRP-J.1, Response to High CNMT Pressure; After completion of required actions even if the Orange Path still exists B. FRP-J.1, Response to High CNMT Pressure; ONLY after the condition causing the Orange Path has been corrected C. FRP-J.2, Response to Containment Flooding; After completion of required actions even if the Orange Path still exists (~ D_ FRP-J.2, Response to- Containment Flooding; - ONLY after the condition causing the Orange Path has been corrected Friday, December 26, 2008 1 :06:55 PM 30

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 Plausibility and Answer Analysis ( A Correct. An ORANGE path exists due to containment pressure greater than 10 psig with no CNMT Spray Pump running and FRP-J. 1 does not require the condition to be corrected prior to exit as other FRPs do (FRP-S.1, FRP-H.1). B Incorrect. An ORANGE path exists due to containment pressure greater than 10 psig with no CNMT Spray Pump running and FRP-J. 1 does not require the condition to be corrected prior to exit as other FRPs do (FRP-S.1, FRP-H.1). C Incorrect. Conditions are met for FRP-J.2 but this ORANGE path would only be evaluated if CNMT pressure was less than 10 psig. Additionally, FRP-J.2 also allows exit before the cause of the ORANGE path is corrected. D Incorrect. Conditions are met for FRP-J.2 but this ORANGE path would only be evaluated if CNMT pressure was less than 10 psig. Additionally, FRP-J.2 does not require the condition to be corrected prior to exit as other FRPs do (FRP-S. 1, FRP-H.1). KIA statement - Containment - Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal c operating procedures. Importance Rating: -4.5 4.7-Technical

Reference:

EOP-CSFST Rev. 9 pg 3 of 3 (side 2) FRP-J.1 Rev. 13, page 10 References to be provided: None Learning Objective: EOP-LP-3.13, Obj 1 Question origin: NEW Comments: SRO justification: Requires assessing plant conditions and prescribing correct procedure to mitigate event in progress. Additionally, requires individual to recall the strategy of the FRP in that it will attempt to mitigate the problem but exit is made before condition is corrected. Origin: NEW Cog Level: H Difficulty: 3

Reference:

FRP-J.l Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: l03G2.4.4 KIA 2: Friday, December 26, 2008 1 :06:55 PM 31

Ref for SRO #15, 103G2. 4 . 4 ICSFST RCS INTffGRny CSF-4 Res INTEGRITY

                                    ~y   RCS PRESSURE-COLO LEG TEMPERA.TURE POINT 10 LEFT OF UMIT A                                                                                                         RED                                            PTS LIMITS CURVE GO TO FRP-P.l            3000 TEMPERA.TURE DECREASE IN ~NY COLD LEG GREATER THAN l00-F I THE LAST 60 MINUTES                                                            ANY RCS COLD LEG TEMPERA1l.IRE LESS THAN 240"F ORANGE GO TO FRP-P.l ALL Res PRESSURE-COLD LEG TEMPERATURE POINTS TO RIGHT OF UMIT A ANY RCS COLO LEG IDoIPERA TURE LESS THAN 270"F                              YELLOW GO TO ALL RCS COLD LEG                                                                     FRP-P.2 ID.IPERAllJRES GREATER THAN 240",

AU. RCS COLO LEG TEMPERATURES GREAlER THAN 270" f GREEN CSF-SAT MolY RCS COLO LEG TEMPERATURE LESS THAN 2-40"1' /llib ORANGE PRESSURE GO TO Res PRESSURE GREATER THAN LOW FRP-P.l (PSIG) lDIPERATURE OVERPRESSURE Ut.4IT ANY RCS COLD LEG ALL RCS COlD LEG TEt.lPERA TURES GREA1ER THAN 240" F YELLOW TEMPERATURE LESS THAN 325'F GO TO FRP-P.2 ~~~~~!f~EI~N& COLD Res PRESSURE L£SS THAN !.OW lD.IPERATURE OVERPRESSURE UMIT GREEN LAST 60 MINUTES CSF-SAT ALL RCS COLO LEG TEMPERATURES GREA1ER THAN J~ F GREEN CSF-SAT CONTAIN E COLD LEG TEMPERATURE eF) CONTAINMENT PRESSURE GREATER THAN .f.5 PSiG RED GO TO FRP-J.l I NO CONTAINMENT Sf'ftAY PUMPS RUNNING ~/ GO [ ORANGE TO FRP-J.l CQIITAINMEHT PRESSURE GREAlER THAN 10 PSIG AT LEAST ONE CONTAINMENT SPRAY PUMP RUNNING YELLOW GO TO CONTAINMENT PRESSURE FRP-J.l LESS THAN 45 PSiG WIDE RANGE CONTAINMENT SUMP L.f\o£L GREATER THAN 198.5 INCHES ORANGE GO TO FRP-J.2 CONTAINMENT PRESSURE HIGH RANGE CNMT POST LOCA LESS THAN 10 PSiG RADIATION MONITORS YELLOW WIDE RANGE GREAlER THAN Al..ARM S(lPOINT GO TO LOW TEMPERATURE OVERPRESSURE LIMIT CONTAINMENT SUMP lEVEL. FRP-J.3

                                                                            ! lESS THAN 198.5 INCHES                 IHIGH  RANGE CNMT POST LOCA RADIATION MONITORS LESS THAN AlARM SETPOINT GREEN                   500           : i CSF-SAT RCS INV8NTtDRY CSF-6 RVUS BASED ON RCP STATUS LESS THAN TABLE YELLOW GO TO

_________________ - Ni ___ _ FRP-1.3 PRZ LEVEL GREAlER THAN 92X ~~ g:m~lg ~~ : 1~g.!t 94'; D'Jl.IAMIC RANG( - 3 Ra; 4<l0 RVUS BASED ON ReP STATUS GREA1ER THAN TABLE YELLOW GO TO FRP-I.l PORV SETPOINT PRZ lEVEL. LESS THAN 1],; YELLOW (PSIG) GO TO FRP-1.2 PRZ LEW.. LESS THAN 92% RVUS BASED ON RCP STATUS YELLOW LESS THAN TABLE GO TO 300--~~Ll4-Ll~L+~LL~~~ PRZ LEVEl.. 94X UPPER RANGE - NO RCP FRP-1.3 35X D'Jl.IAMIC RAHG£ - 1 RCP 100 200 300 400 GREA1ER THAN 17%

                                                                                                                          ~~ g~~g ~~gf : ~ ~g:

RVUS BASED ON RCP STATUS GR£AD THAN TABLE GREEN CSF-SAT RCS TEMPERATURE ("F) Eop-1 SFS i REV. 9 PAGE 3 of 3

Ref for SRO #15, 103G2.4.4 RESPONSE TO HIGH CONTAINMENT PRESSURE Instructions Response Not Obtained

10. Monitor Conditions To Restore ESW To Isolated Fan Coolers:

a. Check ESW - ISOLATED TO ANY B. Observe NOTE prior to FAN COOLERS IN STEPS 8 Step 11 AND GO TO Step 11.

OR 9

b. Check for any of the b. WHEN any of the conditions following: occurs. THEN do Step lOco o Check CNMT pressure - Observe NOTE prior to LESS THAN 10 PSIG Step 11 AND Continue with Step 11.

o Check ESW header isolated to fan coolers for both of the following: o Associated ESW booster pump - RUNNING o Associated orifice bypass isolation valve ( SHUT

c. Restore ESW to isolated fan coolers.

The Containment Status Tree may continue to display a "non-satisfied" condition after completion of the procedure. If this is the case. the appropriate Function Restoration Procedure does not need to be implemented again since all necessary actions have already been performed.

11. RETURN TO Procedure And Step In Effect.

                                     - END -

EOP-FRP-J.l Rev. 13 Page 10 of 12

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

16. 2009A NRC SRO 016/NEW/H/4IFRP-H.l/N/2009A NRC SROI002A2.011 Given the following plant conditions:

( - 'B' MDAFW Pump is under clearance for motor replacement

              - Unit tripped from an Inadvertent Safety Injection
              - Offsite Power was subsequently lost
              - 'A' EDG failed to start
              - The TDAFW Pump tripped on overspeed
              - FRP-H.1, Response to Loss of Secondary Heat Sink, is in progress
              - RCS Bleed and Feed has been initiated The following conditions exist:
              - Secondary Heat Sink has been established by restoring the TDAFW Pump
              - RCS Bleed and Feed is being terminated
              - 1RC-118, Pressurizer PORV (PCV-44SA-SA), cannot be closed Which ONE of the following identifies the correct procedure to be implemented?

A. Remain in FRP-H.1 B~ Go to PATH-1 Entry Point C C. Return to Procedure and Step in effect D. Go to EPP-009, Post LOCA Cooldown and Depressurization

          . Plausibility and Answer Analysis A Incorrect. This would be correct if the PORV Block Valve had power to be shut but the ~' EDG tripped.

B Correct. This is the appropriate response for the loss of coolant Inventory in progress. C Incorrect. Plausible since this would be correct if the heat sink had been restored PRIOR to starting bleed and feed. o Incorrect. Plausible as EPP-009 will be the ultimate procedure required to address the SBLOCA in progress, but it will be entered from Path-1. {~-------------------------------------- Friday, December 26, 2008 1 :06:55 PM 32

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Reactor Coolant - Ability to (a) predict the impacts of the following on the (SYSTEM) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Loss of coolant inventory Importance Rating: 4.3 4.4 Technical

Reference:

FRP-H.1 Rev. 23, page 38 References to be provided: None Learning Objective: EOP-LP-3.11, Obj 4e Question origin: NEW Comments: SRO justification: Requires assessing plant conditions and the prescribing the procedure section with which to proceed. Origin: NEW Cog Level: H Difficulty: 4

Reference:

FRP-H.1 Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: 002A2.01 KIA 2: -C.. ~-----------------~ Friday, December 26, 2008 1 :06:55 PM 33

Ref for SRO #16, 002A2.01 RESPONSE TO LOSS OF SECONDARY HEAT SINK (- rl' - - - - - - - - - - - Instructions ___________ Response Not Obtained r-L--- NOTE: After shutting a PRZ PORV, RCS pressure should be allowed to increase to determine if SI can be terminated.

30. Check SI Termination Criteria:

a. Check for both of the a. GO TO Step 31.

following:

1) RCS subcooling -

GREATER THAN 60°F [90°F] - C 70° F [100° F] - M

2) Check RVLIS full range GREATER THAN 63%

b. GO TO Step 32.

31. Check RCS Bleed Path Status:

a. Check PRZ PORVs AND a. GO TO PATH-I, entry associated block valves - point C.

ANY BLEED PATH OPEN

b. Shut one PRZ PORV AND place b. Shut its block valve.

in auto. GO TO Step 31d.

c. Observe NOTE prior to Step 30 AND RETURN TO

              . Step 30.

d. Block valves - SHUT FOR ANY d. GO TO PATH-I, entry STUCK OPEN PRZ PORV point C.

e. Observe NOTE prior to Step 30 AND RETURN TO Step 30.

32. Reduce SI Flow:

a. Both CSIPs - RUNNING a. GO TO Step 33.

b. Stop Qne CSIE'.

EOP-ERP-H.1 I Rev. 23 I Page 38 of 58

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

17. 2009A NRC SRO 0171BANKIHI21PLP-114N12009A NRC SROI034G2.1.231 Given the following plant conditions:

(' - On Feb 1, at 0600, a plant shutdown, for refueling, was initiated from 100% power

              - The Reactor was shutdown at 2000 on the same day
              - CCW heat exchanger outlet temperature is currently 97.4°F Which ONE of the following indicates the MINIMUM number of hours after shutdown before fuel movement in the Reactor Vessel may begin in accordance with PLP-114, Relocated Technical Specifications and Design Basis Requirements? (Reference provided)

A. 104 hours B. 120 hours C~ 152 hours D. 160 hours Plausibility and Answer Analysis A Incorrect. This is number you get if you use 97.4 degrees vice effective CCW temperature of 102.4 and interpolate. B Incorrect. This is number you get if you use 97.4 degrees vice effective CCW temperature of 102.4.and dO not interpolate. C Correct. Effective CCW becomes 102.4 degrees. Interpolation gives 8 hours beyond lower boundary of 144 hours. D Incorrect. This is the most likely number if candidate performs interpolation incorrectly (168-8 vice 144+8). KIA statement - Fuel Handling Equipment - Ability to perform specific system and integrated plant procedures during all modes of plant operation. Importance Rating: 4.3 4.4 Technical

Reference:

PLP-114" Rev. 18, page 8 References to be provided: Provide PLP-114 Rev. 18, page 8 Learning Objective: TS-LP-2.0/3.0/5.0/8.0, Obj 5 Question origin: Modified slightly from bank to improve distractor plausibility Comments: SRG justification: u

                                        ... ---- Knewledgeoffuel handling facilitiesandprocednres

~-------------------- Friday, December 26, 2008 1 :06:55 PM 34

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 Origin: BANK Cog Level: H Difficulty: 2 ( Ref Provided?: Y

Reference:

KeyWords: PLP-114 2009A NRC SRO KIA 1: 034G2.1.23 KIA 2: ~.-------------------------------------- Friday, December 26, 2008 1:06:55 PM 35

Ref for SRO #17, 034G2.1.23 (Ref Provided) ( Attachment 2 \ Sheet 1 of 3 Refueling Operations 1.0 OPERATIONAL REQUIREMENTS - DECAY TIME 1.1 The reactor shall be subcritical for a minimum period of time as determined by Table A. APPLICABILITY: During movement of irradiated fuel in the reactor vessel. ACTION: With the reactor subcritical for a time less than determined by Table A, suspend all operations involving movement of irradiated fuel in the reactor vessel. Fuel movement in the reactor vessel may continue provided the minimum decay time is greater than the time shown on Table A. 2.0 SURVEILLANCE REQUIREMENTS 2.1 The reactor shall be determined to have been subcritical for a minimum period of time as determined using Table A by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor vessel. 2.2 CCW temperature shall be monitored every 12 hours during the movement of fuel in the reactor vessel to ensure the temperature used to determine decay time is not exceeded. Table A ( Time from Reactor Subcritical (Hours) Effective CCW Temperature (oF) 100 96.9 120 99.3 144 101.7 168 103.8 192 105.6 216 107.2 240 108.6 NOTE 1: - Linear interpolation between listed points is acceptable. NOTE 2: - These delay times are applicable to end of cycle full core off-loads only. A mid-cycle core off-load assumes two CCW and Fuel Pool Cooling trains available and does NOT require compliance with these limits. NOTE 3: - Effective CCW temperature refers to actual CCW heat exchanger outlet temperature plus 5 o F. NOTE 4: - The table assumes the core off-load duration is 30 hours or greater. Spent Fuel Pool Cooling analysis assumes full core off-load occurs no sooner than the earliest allowed time to start core off-load after reactor subcritical on PLP-1l4 Rev. 18 Page 8 of 26

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

18. 2009A NRC SRO 018/NEW/H/3/T.S. 3.6.3/N/2009A NRC SRO/068A2.04/

Given the following plant conditions: ( - The plant is in Mode 3

                    - During slave relay testing 1ED-164, RCDT Vent IRC Isolation, failed to shut automatically
                    - 1ED-164 also failed to shut remotely from the MCR
                    - 1ED-161, RCDT Vent ORC Isolation, operated as expected
                    - Maintenance reports that repair of 1ED-164 will take approximately 15 hours Which ONE of the following is required in order to comply with Technical Specifications and what is the limiting operational concern for this failure?

A':' Shut AND then remove fuses for 1ED-161; Potential to damage the #2 RCP seals B. Shut AND then remove fuses for 1ED-161; Potential to damage the #3 RCP seals C. Shut 1ED-161, fuses for 1ED-161 do NOTneed to be removed; Potential to damage the #2 RCP seals D. Shut 1ED-161, fuses for 1ED-161 do NOT need to be removed; Potential to damage the #3 RCP seals ( Plausibility and Answer Analysis A Correct. Technical Specification action is correct. Technical Specifications require isolation valve be deactivated in the shut position. Operational concern is correct. If RCO T pressure increases to > 15 psig damage could occur to the #2 RCP seals. B Incorrect. Technical Specification action is correct. Technical Specifications require isolation valve be deactivated in the shut position but the operational concern is incorrect. Plausible because #3 seals do send -400 cc of their discharge to the RCOT. But even without this flowpath #3 seal would still have flow from the standpipe to the containment sump C Incorrect. Technical Specification action is incorrect. Technical Specifications require isolation valve to be deactivated in the isolation position. This action would be correct if 1EO-164 had failed in the shut position. Operational concern is correct. If RCOT pressure increases to > 15 psig damage could occur to the #2 RCP seals. o Incorrect. Technical Specification action is incorrect. Technical Specifications require isolation valve to be deactivated in the isolation position. This action would be correct if 1EO-164 had failed in the shut position. Operational concern is .~--=- ~_~~._. .~.Jn~o"FectJb..ut playsible-b.f!.f...flli§§#32 (}.a/$ do sflnsl-4Q.Q.cJlJJtJb.e1r disGha[gEUa1be._~~..... RCOT. But even without this flowpath #3 seal would still have flow from the ~ .. n _ . . standpfpe-tcrtmrcontainrrrel'""7thS,...UTrTTvn7p...------------------------ - - Friday, December 26, 2008 1:06:55 PM 36

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Liquid Radwaste - Ability to (a) predict the impacts of the following on the (SYSTEM) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Failure of automatic isolation Importance Rating: 3.3 3.3 Technical

Reference:

PLP-106 Rev. 44, pages 18, 32, 33 T.S. 3.6.3 pg 3/46-14 (page 296) APP-102 Rev. 12 pages 3-6 References to be provided: None Learning Objective: Question origin: NEW Comments: (KIA Match) KA match since the RCDT is a key component of the Liquid Radwaste system SRO justification: Requires knowledge of required actions per Tech Specs

                                               > 1 hour Origin:         NEW                            Cog Level:   H Difficulty:     3

Reference:

T.S.3.6.3 Ref. Provided?: N KeyWords: 2009A NRC SRO (- KIA 1: 068A2.04 KIA 2: ~~- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - Friday, December 26, 2008 1:06:56 PM 37

Ref for SRO #18, 068A2.04 Attachment 5 ('.. Containment Isolation Valves Sheet 2 of 17 VALVE NO. MAXIMUM PENETRATION CP&L ISOLATION APPLICABLE REDUNDANT NO. (EBASCO} FUNCTION TIME (SEQ} NOTES VALVE(S}

1. PHASE A ISOLATION (continued) 73A 1SP-12 HYDROGEN ANALYZER A 60 7,13 1SP-915 (SP-V300) 73A 1SP-915 HYDROGEN ANALYZER A 60 7,13 1SP-12 (SP-V348) 73B 1SP-941 HYDROGEN ANALYZER A 60 7,13 1SP-917 (SP-V301) 73B 1SP-917 HYDROGEN ANALYZER A 60 7,13 1SP-941 (SP-V349) 74 1ED-94 CNMT SUMP PUMP 60 7,13 1ED-95 (MD-V36) DISCH 74 1ED-95 CNMT SUMP PUMP 60 7,13 1ED-94 (MD-V77) DISCH 76A 1SI-179 ACCUMULATOR FILL 10 7,13 1SI-182 (SI-V554) FROM RWST 76B 1SI-263 ACCUMULATOR DRAIN 10 7,13 1SI-264 C 76B (SI-V555) 1SI-264 TO RWST ACCUMULATOR DRAIN 10 7,13 1SI-263 (Sl-V550} TO RWST 77A 1SI-287 NITROGEN SUPPLY 10 7,13 1SI-290 (SI-V530) 77B 1RC-141 PRT NITROGEN 10 7,13 1RC-144 (RC-D528) CONNECTION 77B 1RC-144 PRT NITROGEN 10 7,13 1RC-141 (RC-D529) CONNECTION 77C RCDT HYDROGEN 10 (WG-D590) CONNECTION 77C 1ED-161 RCDT HYDROGEN 10 7,13 1ED-164 (WG-D291) CONNECTION 78A 1SP-948 RCS SAMPLE 60 7,13 1SP-949 (SP-V111) 78A 1SP-949 RCS SAMPLE 60 7,13 1SP-948 (SP-V23) 78B 1SP-40 PRESSURIZER LlQ 60 7,13 1SP-41 (SP-V11) SAMPLE 78B ***~SP-4*1***-- *.R*R~SlJRI.zERLIQ ---------~~------

                                                                                -60        7;13*      1SP~4Q (SP-V12)        SAMPLE

~~. PLP-I06 Rev. 44 Page 18 of 74

Ref for SRO #18, 068A2.04 Attachment 5 Sheet 16 of 17 c Containment Isolation Valves TABLE NOTATIONS For this valve, the valves listed in the REDUNDANT VALVE column are the valves which are used to meet the initial ACTION statement of Specification 3.6.3 which states "maintain at least one isolation valve OPERABLE. Further action under Specification 3.6.3 is still required to isolate the affected penetration or to shut down in accordance with Actions a, b, c, or d. Reopening of an inoperable valve is allowed to permit surveillance testing to demonstrate its operability or the operability of other equipment per Specification 4.6.3.1, or to change to another action statement for the LCO. A change between action statements is permitted for activities directly related to restoring the valve to an operable status. Each deviation constitutes a new LCO entry and the 4-hour action requirement applies. The following guidance is provided for complying with the follow-up action requirement, specified in Actions a, b, or c, to isolate the penetration: A valve having the same safety class and seismic design class in series with the inoperable valve must be closed (and de-activated, if applicable for power-operated valves). If the piping branches, each branch must be isolated. Check valves may not be used to isolate a penetration beyond the four-hour period. (8) For this valve, the closed, water-sealed system outside containment is considered to be an OPERABLE isolation valve for purposes of compliance with the ( initial ACTION statement of Specification 3.6.3 which states "maintain at least one isolation valve OPERABLE . . . " Further action under Specification 3.6.3 is still required to isolate the affected penetration or to shut down in accordance with Actions a, b, c, or d. Reopening of an inoperable valve is allowed to pe~mit surv~illance testing to-demonfrtrateits operability Or the operability of other equipment per Specification 4.6.3.1, or to change to another action statement for'the LCO. A change between action statements is permitted for activities directly related to restoring the valve to an operable status. Each deviation constitutes a new LCO entry and the 4-hour action requirement applies. The following guidance is provided for complying with the follow-up action requirement, specified in Actions a, b, or c, to isolate the penetration: Either the inoperable valve must be closed (and de-activated, if applicable for power-operated valves), OR A valve having the same safety class and seismic design class in series with the inoperable valve must be closed (and de-activated, if applicable for power-operated valves). If the piping branches, each branch must be isolated. Check valves may not be used to isolate a penetration beyond the four-hour period. PLP-106 Rev. 44 Page 32 of 74

Ref for SRO #18, 068A2.04 Attachment 5 Sheet 17 of 17 Containment Isolation Valves TABLE NOTATIONS (9) For relief valves, the "REDUNDANT VALVE(s)" column applies to the action statement whenever the relief is unable to isolate (that is, excessive leakage or failed open). When a relief is unable to open or cannot otherwise adequately relieve design overpressure conditions, the penetration must be isolated, and the penetration and closed system (if Note 6 or 8 is applicable) drained to eliminate the potential for an overpressurization event. (10) For this valve, no redundant valve or closed system is available for purposes of compliance with the initial ACTION statement of Specification 3.6.3 which states "maintain at least one isolation valve OPERABLE. Immediate action to isolate the penetration must be initiated or Technical Specification 3.0.3 should be applied. (11) Deleted. (12) Engineering review required when maximum isolation time exceeds 8 seconds to ensure the corresponding Engineered Safety Features Response Time meets the ten second requirement. If the valve is inoperable, perform Technical Specification Surveillance Requirement 4.6.1.1.a at least once every 31 days. PLP-I06 Rev. 44 Page 33 of 74

Ref for SRO #18, 068A2.04 CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 Each containment isolation valve specified in the Technical Specification Equipment List Program, plant procedure PLP-I06, shall be OPERABLE with isolation times less than or equal to required isolation times. APPLICABILITY: MODES I, 2, 3, and 4. ACTION: With one or more of the containment isolation valve(s) inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and:

a. Restore the inoperable valve(s) to OPERABLE status within 4 hours, or

b. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation pOSition, or

c. Isolate each affected penetration within 4 hours by use of at least one closed manual valve or blind flange, or c d. Be in at least HOT STANDBY within the next SHUTDOWN within the following 30 hours.

6 hours and in COLD SURVEILLANCE REQUIREMENTS 4.6.3.1 Each isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time.

SHEARON HARRIS - UNIT 1 3/4 6-14 Amendment No. 84

Ref for SRO #18, 068A2.04 ALB-102-1-1 Sheet 1 of 4 ALARM UNIT 1 RCDT PRESS HI AUTOMATIC ACTIONS None applicable CAUSE

1. 1ED-164 or 1ED-161 Containment Isolations shut due to "T" signal or improper valve lineup.

2. 1ED-182, RCDT HY Vent Line Ck Vlv, stuck shut.

3. 1ED-153, Hydrogen Inlet Control Valve not operating properly.

4. Waste Gas Compressor is shut down.

5. 1ED-178, Hydrogen Vent Line Control Valve not operating properly.

( OBSERVATIONS

1. CONTACT Main Control Room to verify position of 1ED-164 aFld 1ED~161.

2. PI-1004 on WPCB. Normal RCDT pressure is 0 to 10 psig. Note pressure increases.

3. RCDT level is 20 to 80% at LI-1003.

4. RCDT temperature is 170°F or less at TI-1 058.

5. Computer Graphic 56 and Computer Points PA020, LA020, and TA020.

1-- ---;===================:;====================;:::===========:-- I APP-102 Rev. 12 Page 3 of 48 I

Ref for SRO #18, 068A2.04 ALB-102-1-1 Sheet 2 of 4 ACTIONS CAUTION Operating the RCDT at greater than 15 psig can cause damage to the #2 seal of the RCPs. Efforts should be made as quickly as possible to reduce the RCDT pressure if RCDT PCVs are not operating properly.

1. VERIFY proper valve lineup.

2. CHECK Waste Gas System for proper operation.

3. VERIFY with the MCR the following valves are OPEN:

1ED-164, RCDT VENT IRC ISOLATION
1ED-161, RCDTVENT ORC ISOLATION

4. Locally, CHECK 1ED-178, RCDT Hydrogen Vent Line Control Valve to be open.

5. IF 1 ED-178 is open and pressure is not dropping in the RCDT, THEN PERFORM the following:

( a. MANAULLY AGITATE 1ED-182, RCDT HY Vent Line Ck Vlv.

b. tF check valve 1 ED-182 wilf not open, THEN INITIATE work order on 1ED-182.

6. IF pressure remains high for over 5 minutes OR 1ED-178, Hydrogen Vent Line Pressure Control Valve is not operating properly, THEN PERFORM the following:

a. NOTIFY MCR and RCP System Engineer of annunciator, RCDT pressure, and inform them that this could damage #2 seal of the RCPs.

1APP-102 Rev. 12 Page 4 of 481

Ref for SRO #18, 068A2.04 ALS-102-1-1 Sheet 3 of 4 ( ACTIONS (continued)

b. REDUCE pressure in the RCOT by performing one or all of the following steps:

CAUTION OP-120.08, Section 8.10 contains limits for pumping down RCOT in manual. Oamage can occur to the RCP #2 Seals if level is dropped quickly and RCOT loses pressure. (1) TAKE MANUAL CONTROL of LK-1003 RCOT LCV/IRC ISOLATION 1EO-121 per OP-120.08, Section 8.10 to open valve and lower RCOT level. Level in RCOT should remain above 20%. Return controller to automatic operations. (2) IF Waste Gas System is inservice, THEN TAKE MANUAL CONTROL of 1EO-178, Hydrogen Vent Line Pressure Control Valve as follows: (a) OPEN 1EO-178, Hydrogen Vent Line Pressure Control Valve, by adjusting pressure controller at valve to control pressure in the RCOT below 6 psig. (b) CONTINUE to monitor RCOT pressure to determine if 1EO-178 is operating properly. (3) IF Waste Gas System is shutdown or RCOT pressure cannot be maintained in desired range by manipulating RCOT level and PCV, THEN PERFORM the following steps: (a) CONTACT MCR to determine PRT pressure. If RCOT pressure is higher than PRT, obtain MCR permission to vent the RCOT to the PRT. (b) TAKE MANUAL CONTROL of LK-1003 RCOT LCVIIRC ISOLATION 1EO-121 and open valve to lower RCOT level until RCOT Pump trips. (c) VERIFY RCOT PUMP AlB RECIRC 1EO-143 is OPEN. (d) OPEN 1EO-138, RCOT PUMPS AlB TO PRESSURIZER RELIEF TANK. 1 APP-102 Rev. 12 Page 5 of 481

Ref for SRO #18, 068A2.04 ALB-102-1-1 Sheet 4 of 4 ACTIONS (continued) (e) ALLOW RCDT TO VENT to the PRT until RCDT pressure is at desired pressure or until RCDT pressure quits dropping. (f) To secure the RCDT vent to PRT, SHUT 1ED-138, RCDT PUMPS AlB TO PRESSURIZER RELIEF TANK. (g) NOTIFY MCR that RCDT vent to PRT is secured. (h) RETURN LK-1003 RCDT LCVIIRC ISOLATION 1ED-121 to automatic control. (i) RESTART RCDT Pumps per OP-120.08, Section 5.0.

7. INITIATE a Work order if a control valve is malfunctioning or instrument has failed.

DEVICE/SETPOINTS PS-01WL-1004W 10 psig POSSIBLE PLANT EFFECTS

1. Potential to release fission gases to containment.

2. Possible damage to RCP's #2 seals.

REFERENCES

1. OP-120.08

2. CAR-2166 B-401, Sheet 4386, Control Wiring Diagram

3. CPL-2165 S-1313, Containment Building Waste Processing System

4. AR 60984, 1ED-182 failures

~:_~-"""~"_"_'~_~~~. _. . _~ _._~~_.,,_~. ___ ~~_~_~~:---="~-=-~__ ~~~~-- _=--_ _ _ _ _

                                                                                       .. _ _ _ _ _ ~~ _ _ _-~~--~-"_,~."~u._~~~_~~,*,~-~'~_~_*'_~4_~_>>O~

(-r====~====r========================;===========~ 1APP-102 Rev. 12 Page 6 of 481

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

19. 2009A NRC SRO 019IPREVIOUS NRCIF/3IFHP-020/N/2009A NRC SRO/G2.1.411 Given the following plant conditions:

( - The plant is in Mode 6 with refueling in progress Current plant conditions are:

              - Fuel movement has stopped due to a problem with the gripper tube top limit switch on the Manipulator
              - The Main Control Room has been informed that initial troubleshooting is in progress on the Manipulator
              - The troubleshooting team desires to operate TS-3, Bridge Left Interlock Bypass, in order to move the bridge while the gripper tube is not at the top limit Which ONE of the following describes the approval and concurrence, if any, required for this action in accordance with FHP-020, Refueling Operations?

A. The SSO must approve. NO concurrence is required. B. The SSO must approve with concurrence of Reactor Engineering. C. The SRO-Fuel Handling must approve. NO concurrence is required. D~ The SRO-Fuel Handling must approve with the concurrence of the SSO. Plausibility and Answer Analysis A Incorrect. SRO-Fu~1 Handling does give p~rmjssjon for this evolution, but f~ bypassing interlocks the SSO MUST concur. B Incorrect. The SSO only needs to concur with bypassing the interlock. Reactor Engineering is required to be involved in fuel moves but not in troubleshooting of the equipment doing the moves. C Incorrect. Concurrence is required from SSO per FHP-020, not approval. D Correct. FHP-020 Rev. 37 P & L 26 reads: Bypassing of fuel handling equipment interlocks which are not specified in approved procedures shall require permission of the SRO-Fuel Handling and concurrence of the Superintendent - Shift Operations. Friday, December 26, 2008 1 :06:56 PM 38

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Conduct of operations - Knowledge of the refueling processes Importance Rating: 2.8 3.7 Technical

Reference:

FHP-020 Rev. 37, P&L 26, page 12 References to be provided: None Learning Objective: Question origin: Comments: Slightly modified from 2008 NRC exam item, 034G2.1.14, but evaluated as not significantly modified. SRO justification: Knowledge of fuel handling facilities and procedures Origin: PREVIOUS NRC Cog Level: F Difficulty: 3

Reference:

FHP-020 Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: G2.1.41 KIA 2: Friday, December 26, 2008 1 :06:56 PM 39

Ref for SRO #19, G2.1.41 REFERENCE USE 5.1 Precautions and Limitations for Fuel Movement in the Fuel Handling Building (continued) (

24. Fuel Handling Building integrity is maintained by verifying the Fuel Handling Building operating floor hatch cover is in place.

25. The FHB Operator reports to the SRO-Fuel Handling while performing fuel handling functions for core alterations.

26. Bypassing of fuel handling equipment interlocks which are not specified in approved procedures shall require permission of the SRO-Fuel Handling and concurrence of the Superintendent - Shift Operations.

27. Loads in excess of 2300 pounds are prohibited from travel over fuel assemblies in a storage pool with irradiated fuel in the pool.

28. Due to the possibility of a limit switch failure, the hoist upper limit switch on the Spent Fuel Pool Bridge Crane will not be used to halt tool vertical movement except as required to meet the traverse inhibit interlock. In such cases, the limit switch shall be approached slowly.

R 29. The Fuel Handling equipment operators shall use all available indications, load cells, rail indexing, and long handled tool markings, to anticipate limit switch and/or interlock functions. The equipment shall be stopped if limit switch setting or procedural limit is exceeded or if an interlock fails to function. (Reference 2.7.1) (

30. During fuel movement in the Fuel Handling Building, CQmmunications will be available with the Control Room and Containment. Periodic communications checks are sufficient to meet this requirement.

(at least once per shift)

31. The load cell should be monitored at all times when lifting or lowering a fuel assembly. If greater than 100 pounds above or below the suspended weight is observed, stop fuel movement and refer to Attachment 12. If load continues to increase in excess of 250 pounds, stop fuel movement and refer to Attachment 12 for additional information.

32. During core alterations, all activities involving movement of fuel in the Fuel Handling Building shall be under the direct supervision of the FHB Operator. The FHB Operator reports to the SRO Fuel Handling while performing fuel handling functions.

 ~ FHP-020

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

20. 2009A NRC SRO 020/MODIFIEDIF/3IPLP-702/N/2009A NRC SRO/G2.2.13/

Given the following plant conditions: ( - 'A' ESW Header is under clearance.

               - 1SW-39, Normal SW Supply to Header A, has been determined to have seat leakage and must be manually shut Which ONE of the following identifies the permission level that is required to operate 1SW-39 manually and what must be accomplished to restore the valve to an OPERABLE status?

Permission To restore operability A. SSO restore power B. SSO restore power AND stroke the valve electrically C. USCG restore power D~ USCG restore power AND stroke the valve electrically Plausibility and Answer Analysis OMM-014, section 5. 1 (Operations Clearances) step 17 contains the operability c requirements. PLP-702 contains permission requirements A Incorrect. USCG permissionis requir~d to operate an MOV manually JAW PLP-702 . section 4. 1.8 but SSO permission is required for other things. PLP-702 section 4.1.8.c states manual action should be taken to get the valve off the seat but operability specifically requires the valve be stroked from the control switch (PLP-702 section 4. 1. 8.g. (1)). Had the valve not been manually shut then restoring power is all that would be required to restore operability. . B Incorrect. USCD permission is required to operate an MDV manually lAW PLP-702 section 4.1.8 but SSD permission is required for other things. PLP-702 section 4.1.8.g.(1) requires the valve be stroked from the control switch to declare operable. C Incorrect. USCD permission is required to operate an MDV manually lAW PLP-702 section 4.1.8. PLP-702 section 4.1.8.c states manual action should be taken to get the valve off the seat but operability specifically requires the valve be stroked from the control switch (PLP-702 section 4. 1. 8.g. (1)). Had the valve not been manually shut then restoring power is all that would be required to restore operability. D Correct. USCD permission is required to operate an MDV manually lAW PLP-702 section 4.1.8. PLP-702 section 4. 1. 8.g. (1) requires the valve be stroked from the

             . contro/switch to dee/are operable;"                                      -_....

c.--------------------------------- Friday, December 26, 2008 1 :06:56 PM 40

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Equipment Control - Knowledge of tagging and clearance procedures. Importance Rating: 4.1 4.3 Technical

Reference:

OMM-014 Rev. 58, step 17, page 15 PLP-702 Rev. 23, step 8, pages 6-7 OPS-NGGC-1301 Rev. 16, page 30 References to be provided: None Learning Objective: Question origin: Modified from Bank (ORQ VAL-18.0-R5) Comments: OPS-NGGC-1301 section 9.2.1 step 19 allows manual operation of MOVs for seat leakage and refers you to site procedures. OMM-014, section 5.1 (Operations Clearances) step 17 contains the operability requirements. PLP-702 contains permission requirements. SRO justification: Approving MOV manual operation and making operability calls are the responsibility of the SRO. Origin: MODIFIED Cog Level: F Difficulty: 3

Reference:

PLP-702 Ref. Provided?: N (~ KeyWords: 2009A NRC SRO KIA 1: G2.2.13 KIA 2: l,-- - - - - - - - - - - - - - - : - - - - - - - Friday, December 26, 2008 1 :06:56 PM 41

Ref for SRO #20, G2.2.13 5.1.1 Standard Practices (continued)

b. If configuration control will not be maintained by the Clearance, perform the following:

(1) The Superintendent - Shift Operations shall approve not maintaining configuration control by the clearance. This approval should be noted in the Special Instructions. (2) The OP electrical/valve lineup or a System ElectricalNalve Lineup Checklist from OMM-001 should be used to document the restoration of all components within the Clearan"ce boundary. This should be completed prior to restoring the Clearance. When the restoration is complete, process the lineup per OMM-001.

12. The Tag Hanger or individual directing the removal of the Operations Clearance should have as a minimum the Clearance Checklist, or a copy, present during installation and removal of tags.

13. When boundary valves leak by their seat and a complete draining cannot be accomplished, the Unit SCQ and the Clearance Holder should determine when conditions are safe to perform the required maintenance.

14. If the Clearance involved draining a portion of a system, QMM-001, c Refilling and Venting Systems After Draining section, should be referenced. For filling and venting of an entire train or system, the applicable Operating Procedure should be referenced for instructions. A CM~ (Comment) step wiUbe added to the clearance checklist for fill and vent steps.

15. If the component that was drained has a heater associated with the drained portion, the fill and vent must be performed prior to energizing the heater.

16. If applicable, specific Sections of procedures can also be specified to assist in removing the Operations Clearance in the correct order and/or verifying appropriate plant conditions exist before returning the equipment to service. If only a portion of a procedure is to be performed, the applicable steps should be clearly identified and the partial procedure performance accomplished per PRO-NGGC-0200. The Unit SCO shall be fully cognizant of the procedure Steps to be performed and verify that plant conditions are appropriate.

17. All Limitorque 5MB-OO/SB-OO motor operated valves, if manually operated, are required to be stroked electrically from the control switch to be declared operable. All of the applicable 5MB-OO/SB-OO valves are listed on Attachment 7. (Reference 2.4.8)

~~-{I~O~M~M~-O~1~4============L=======~R~e~V.~5~8========+/-=========!P~ag~e~I~5~o~f6~8QI-----

Ref for SRO #20, G2.2.13 4.1 Generic Information on Valve Manipulation (continued)

6. Pipe caps must be installed properly to ensure a leak tight barrier is established. The following guidance is provided for installing pipe caps:

a. Verify drain valves are fully shut and line is fully drained.

b. The proper sealant should be used. (sealant is required for borated systems) CHE-NGGC-0045, can provide information as to the acceptability of the sealant being used. (Grafoil is an approved pipe thread tape for stainless steel pipes.)

c. A pipe wrench should be used to install pipe caps. Pipe caps should not be left hand tight unless the cap is frequently removed.

Pipe caps on borated systems should be fully tightened. The use of tape is not required for frequently removed caps.

d. If pipe or cap threads are found to be damaged, a work order should be intiated.

NOTE: Stem travel observation is not normally an acceptable method for verifying valve position. Stem lengths vary and valves may not be totally open or shut. Valves with vendor supplied stem position indicators may be checked by observing these indicators. If the operator suspects faulty operation of these indicators, the valve should be manually checked, and corrective action taken to restore the position indicator to an operable status.

7. If a valve cannot be positioned or checked because it appears stuck, inform the Unit SCQ before making further attempts to reposition the valve. In such cases, the position of the valve stem may have to be used to verify position of the valve.

8. All motor-operated valves will normally be operated by the respective motor. Motor-operated valves may be manually shut or back seated only with Unit SCQ permission or under emergency situations. The following guidelines apply to operation of motor operated valves:

a. A visual inspection will be performed before manually operating a motor operated valve. This inspection should look for damage to the motor operator and to the valve body.

b. Power will normally be removed from the motor before local operation of the valve to prevent inadvertent actuation of the motor during manual operation.

Ref for SRO #20, G2.2.13 4.1 Generic Information on Valve Manipulation (continued)

c. Place a caution tag on the valve(s) control switch(es) to denote that manual operation to shut or backseat a valve has occurred, and manual action should be taken to get a valve off its shut/open seat before operating with the motor.

d. If the valve performs a safety related function, declare the valve inoperable unless determined that the manual operation has not affected operability.

e. No motor operated valve should be operated with a cheater bar, valve wrench or other device which increases the mechanical advantage. The use of excessive force can cause catastrophic failure of valve components resulting in injury or death to the Operator.

f. Limitorque valves can not be operated in manual if the clutch key is in direct interference with manual operation lugs. This makes declutching impossible. To eliminate the interference, the hand wheel wi" need to be rotated so that the clutch can be engaged.

This problem is specifically related to 5MB-OO and 'SMB-OOO Limitorque operators but could include a" types of Limitorque valves. (reference NCR 87560)

g. Motor operated valves that have been operated in manual should

( be operated from the control switch when power is restored. (1) Refer to OMM-014 for 5MB-OO/SB-OO valves. The sp6C1fied valves in OMM-014 are to be declared inoperable until stroked from the control switch. (Reference CAP Item 95H0426) CAUTION The RAB Emergency Exhaust Vortex Dampers (Model NH94 Hydramotor) require that power not be removed from the hydramotor when operating manually. The normally shut dump valve must have power to relieve hydraulic pressure.

9. A" hydramotor operated valves will normally be operated by the respective hydramotor. Hydramotor operated valves may be manually operated only with Unit SCO permission or under emergency situations. The following guidelines apply to operation of hydramotor operated valves:

a. Turn off power to the hydramotor before operating the handwheel.

I PLP-702 Re-v.23

Ref for SRO #20, G2.2.13 9.2.1 Administrative (Cont.)

19. Motor operated valves may be used as an isolation boundary point provided, after the valve has been positioned for the clearance, its power supply is isolated and tagged and the handwheel is tagged to indicate the valve position.

The valve should not be manually engaged to check position. This will prevent inadvertent damage to the torque switch and/or valve seat, and prevents the drifting problem associated with some Limitorque operated valves. Since the valve position may not be available after the motor breaker is turned off, concurrent verification may be used to determine valve position before isolating the power supply. If the valve is determined to have seat leakage, it is permissible to manually engage the handwheel and torque the valve shut. Refer to site procedures for positioning and position verification associated with motor operated valves.

20. Conditions may exist such that it is not practical for a single Operations Clearance to cover the scope of planned work.

It is permissible to use another Operations Clearance in conjunction with the original clearance to perform such work. When more than one Operations Clearance is used to allow work, the other clearance numbers should be listed on the Checklist Cross Reference screen for the existing clearance. [RI . 21. If items are added to the Checklist Cross Reference, an ~ entry should be made in the Clearance Order Special Instructions indicating the Checklist has cross references listed. Prior to making boundary changes, the Checklist Cross Reference Screen should be checked. Review of the Checklist Cross References is to ensure no other Clearance Order or Checklist is impacted by the change. The review should be completed by an SRO (CNO for CR3) prior to making any change to an established boundary. I OPS-NGGC-1301 Rev. 16 Page 30 of 75 I

Original for SRO #20, G2.2.13 QUESTIONS REPORT for ORQ 1. ( VAL-18.0-RS 001 5MB-OO Limitorque MOV 1CS-214 has had power removed and it has been closed with the handwheel with approval by the SCQ. The valve was declared inoperable. Work has been completed and the valve is now ready to be restored to service. What action(s) is(are) required to declare the valve operable? A. Restore power. B. Manually move the valve off the closed seat and restore power. C. Manually place the valve in it's required position and restore power. D~ Restore power and position the valve electrically. Reference Operations Clearances section of OMM-014 LOCT05-01A; (TTT comments resolved) 06-01-2 11/07/06 Deleted same question from LOR bank.

Reference:

OPS-NGGC-13 0 1

Reference:

OMM-014 Primary KIA: C~ 2.2.13 KIA Value: 3.6/3,8 Sec. KIA: NA KIA Value: NA ORQYes/NO: YES NRC Q? Yes/NO: NO l<<.~ Friday, December 26,20081:35:19 PM 1

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

21. 2009A NRC SRO 021IMODIFIEDIF/3/T.S. 3.8.2.1/N/2009A NRC SRO/G2.2.40/

Given the following plant conditions: ( - The plant is operating at 100% power

                 - Maintenance reports that 'A' Battery electrolyte level is overflowing in several cells
                - 'A' Station Battery is declared inoperable Which ONE of the following is the MAXIMUM time allowed before the plant must be in Mode 3, and the reason why?

A'I 8 Hours; A subsequent loss of the remaining DC system would hamper mitigation and control of accident conditions within the facility B. 14 Hours; A subsequent loss of the remaining DC system would hamper mitigation and control of accident conditions within the facility C. 8 Hours; Sufficient instrumentation and control capability is no longer available to monitor and maintain the unit status. D. 14 Hours; Sufficient instrumentation and control capability is no longer available to monitor ( and maintain the unit status. ( Friday, December 26, 2008 1 :06:56 PM 42

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 Plausibility and Answer Analysis A Correct. Category B parameter is outside allowable parameter (overflowing), TS action required in 2 hours or be in at least HSB within the next 6 hours. The reason why is correct based on TS bases for A.C. and D.C. power sources in Modes 1-3. B Incorrect. Plausible because this time applies for an AC ESF bus not fully energized, but this is for reserve power to a DC bus. (8 hours to reenergize AC ESF bus or be in HSB within the next 6). The reason why is correct based on TS bases for A.C. and D.C. power sources in Modes 1-3. C Incorrect. Plausible due to Category B parameter is outside allowable parameter (overflowing), TS action required in 2 hours or be in at least HSB within the next 6 hours. Time is correct, but reason given is basis for DC distribution operability during shutdown or refueling conditions but conditions are plant is in Mode 1. D Incorrect. Plausible this time applies for an AC ESF bus not fully energized but this is for reserve power to a DC bus. (8 hours to reenergize AC ESF bus or be in HSB within the next 6). Reason given is basis for DC distribution operability during shutdown or refueling conditions but conditions are plant is in Mode 1. KIA statement - Equipment Control - Ability to apply technical specifications: for a system. Importance Rating: 3.4 4.7 Technical

Reference:

Tech Spec 3.8.2.1 pg 3/4 8-12 (page 342) Tech Spec Bases pg B 3/4 8-1 (page 462) References to be provided: None Learning Objective: ADEL-LP-2.7, Obj 1b Question origin: Modified from Bank OIT Exam Bank DCP (12A) 001 Comments: Revised times for distractors to AC sources times. Original distractor times were not based on either AC or DC Tech Specs SRO justification: Requires knowledge of required tech spec bases Origin: MODIFIED Cog Level: F Difficulty: 3

Reference:

T.S.3.8.2.1 Ref. Provided?: N KeYWords: 2009A NRC SRO K/AI: G2.2.40 K/A2: Friday, December 26, 2008 1:06:56 PM 43

Ref for SRO #21, G2.2.40 ELECTRICAL POWER SYSTEMS ( 3/4.8.2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum. the following D.C. electrical sources shall be OPERABLE:

a. 125-volt Emergency Battery Bank lA-SA and either full capacity charger. lA-SA or lB-SA. and,

b. l2S-volt Emergency Battery Bank lB-SB and either full capacity charger, lA-S8 or 1B-S8.

APPLICABILITY: MODES 1. 2. 3, and 4. ACTION: With one of the required D.C. electrical sources inoperable. restore the inoperable D.C. electrical source to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ( SURVEILLANCE REQUIREMENTS 4.8.2.1 Each l25-volt Emergency 8attery and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1. The parameters in Table 4.8-2 meet the Category A limits, and

2. The total battery terminal voltage is greater than or equal to 129 volts on float charge.

b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above ISO volts, by verifying that:

1. The parameters in Table 4.8-2 meet the Category 8 limits.

2. There is no visible corrosion at either terminals or connectors. or the connection resistance of these items is less than ISO x 10~ ohm. and

~ _____ .. _ _3 . The average electro lyte temperatur~J2fJ~Qee£pnne~t.eJLcelkj5~-~~-~~- eee~._ ~.~--~---~------~--*-atmVF7o--r.----~~-~--*--*~~--~*---~-~ SHEARON HARRIS - UNIT 1 3/4 8-12

Ref for SRO #21, G2.2.40 3/4.8 ELECTRICAL POWER SYSTEMS BASES

   .3/4.8.1 0 3/4.8.2, AND 3/4.8.3 A.C. SOURCES. D.C. SOURCES, AND ONSITE POWER DIS I RIB rION The OPERABILITY of the A.C. and O.C power sources and associated distribution systems during operation ensures that sufficient power will be available to suppLY the safety-related equipment required for: (1) the safe shutdown of the racility, ana (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C.

power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50. The switchyard is designed using a breaker-and-a-half scheme. The switchyard currently has seven connections with the CP&L transmission network: each of these transmission lines is physically independent. The switchyard has one connection with each of the two Startup Auxiliary Transformers and each SAT can be fed directly from an associated offsite transmission line. The Startup Auxiliary Transformers are the p-referred power source for the Class IE ESF buses. The minimum alignment of offsite power sources will be maintained such that at least two physically independent offsite circuits are available. The two physically independent circuits may consist of any two of the incoming transmission lines to the SATs (either through the switchyard or directly) and into the Class IE sy'stem. As long as there are at least two transmission lines in service ana two circuits through the SATs to the Class IE buses, the LCO is met. During MODES 5 and 6, the Class IE buses can be energized from the offsite transmission net work via a combination of the main transformers, and unit auxiliary transformers. This arrangement may be used to satisfy the requirement of one physically independent circuit. (- The ACTION requirements specified for the levels of degradation of the power sources provide restrictioo upon continued facility .operation CQm~ensurate w;th the level of degradatlon. The OPERABILITY or the power sources are consistent with the lnitial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and aSSOCiated distribution systems OPERABLE during aCCident conditions coincident with an assumed loss-of-offsite power and single failure of the other onsite A.C. source. The A.C. and D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, "Availability of Electrical Power Sources," December 1974. There are additional ACTION requirements to verify that all required feature(s) that depend on the remaining OPERABLE A.C. sources as a source of emergency power, are also OPERABLE. These requirements allow a period of time to restore any required feature discovered to be inoperable. e.g. out-of-service for maintenance, to an OPERABLE status. If the required feature(s) cannot be restored to an OPERABLE status, the ACTION statement requires the redundant required feature, i.e. feature receiving power from an inoperable A.C. source. to be declared

  . inoperable. The allowed operating times to restore an inoperable reguired feature to an OPERABLE status is based on the requirements in NUREG 1431. The term "verify". as used in these ACTION statements means to administratively check by examining logs or other information to determine the OPERABILITY of requirea feature(s). It does not mean to perform the Surveillance Requirement needed to demonstrate the OPERABILITY of the required feature(s).

~c~------------------- SHEARON HARRIS - UNIT 1 8 3/4 8-1 Amendment No. 78 I

Original for SRO #21, G2.2.40 QUESTIONS REPORT for OIT Exam Bank

1. DCP (l2A) 001 Given the following conditions:

                - The plant is at 100% RTP
                - Maintenance reports that 'A' Battery electrolyte level is overflowing in several cells
               - 'A' Station Battery is declared inoperable Which ONE of the following is the maximum time allowed before the plant must be in Mode 3, and the reason why?

A. 7 Hours; A subsequent worst case single active failure would result in loss of all DC subsystems with attendant loss of ESF functions. B!'" 8 Hours; A subsequent worst case single active failure would result in loss of all DC sUbsystems with attendant loss of ESF functions. C. 7 Hours; Sufficient control and instrumentation capability is no longer available to monitor and maintain the unit status. D. 8 Hours; Sufficient control and instrumentation capability is no longer available to monitor and maintain the unit status.

             . A. ~neorrect. 2- hours is altowed, but the reason is correct. 1 hour comes from verification of battery cell parameters, but in this case, they are exceeded.

B. Correct. Category B parameter is outside allowable parameter, TS action required in 2 hours due to reasons described. C. Incorrect. Reason given is basis for DC distribution operability during shutdown or refueling conditions. D. Incorrect. Time is correct, but reason is for shutdown or refueling conditions. KIA Statement - Knowledge of the bases in Technical SpeCifications for limiting conditions for operations and safety limits. Importance Rating: RO 3.2/ SRO 4.2 Technical

Reference:

TS 3.8.2.1 References to be provided: None Learning Objective(s): Question origin: New 2/08 (modified from HBR NRC Exam) Comments: Friday, December 26, 2008 1:37:53 PM 1

QUESTIONS REPORT for OIT Exam Bank Program: S Cog Level: H Difficulty:

Reference:

TS 3.8.2.1 Ref Provided?: N KeyWords: KIA 1: 063G2.2.25 KIA 2: ( Friday, December 26, 2008 1:37:53 PM 2

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

22. 2009A NRC SRO 022INEW/H/2/PEP-330IN12009A NRC SRO/G2.3.121 Given the following plant conditions:

          - A General Emergency has been declared
          - All Emergency Response facilities are activated
          - A non-licensed operator must be dispatched from the Operations Support Center to an area with an identified radiation field of 110 Rem/hour in order to isolate the pathway for a large release to the environment
          - The operator will only be in the area for approximately 15 minutes Assuming all of the available operators have volunteered and are fully aware of the risks involved, which ONE of the following operators should be used to perform the task?

A. Worker A is 24 years old and has received an acute dose of 25 Rem TEDE B. Worker B is 52 years old and has received an acute dose of 30 Rem TEDE C. Worker C is 29 years old and has received a cumulative dose of 15 Rem TEDE D~ Worker D is 49 years old and has received a cumulative dose of 20 Rem TEDE Plausibility and Answer Analysis From PEP-330 attachment 1 Limitations for Lifesaving and Emergency Reentry/Repair Actions A Incorrect. Individual has volunteered and is aware of the risk (a and b of 6) but this individual should not be used because of age and already used once before (c and d of 6). B Incorrect. Individual has volunteered and is aware of the risk (a and b of 6) but this individual should not be used because already used once before (d of 6). C Incorrect. Individual has volunteered and is aware of the risk (a and b of 6) but this individual should not be used because of age (c of 6). o Correct. Individual has volunteered and is aware of the risk (a and b of 6) and this individual is the appropriate age and never been used before (c and d of 6). Friday, December 26, 2008 1:06:56 PM 44

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Radiation Control - Knowledge of radiological safety principles c pertaining to licensed operator duties Importance Rating: 3.2 3.7 Technical

Reference:

PEP-330 Rev. 9, page 17 References to be provided: None Learning Objective: Question origin: NEW Comments: PEP-330, Attachment 1, additional criteria for exposures

                                           >25 rem TEDE SRO justification:               The SEC would make the decision as to who would perform the task Origin:         NEW                           Cog Level: H Difficulty:     2

Reference:

PEP-330 Ref. Provided?: N KeyWords: 2009A NRC SRO KIA 1: G2.3.12 KIA 2: Friday. December 26. 2008 1:06:56 PM 45

Ref for SRO #22, G2.3.12 Attachment 1- Limitations for Lifesaving and Emergency Reentry/Repair Actions Sheet 1 of 1

1. A Declared Pregnant Woman shall not take part in these actions.

2. Internal exposure should be minimized by the use of the most appropriate respiratory protection or ALARA practice whenever possible, and contamination should be controlled by the use of protective clothing when practical.

3. Emergency worker exposures during lifesaving and repair/reentry efforts should be limited to the following:

DOSE LIMIT ACTIVITY CONDITION (rem TEDE) 5 All All 10 Protecting valuable property Lower dose not practicable 25 Lifesaving or protection of large Lower dose not practicable populations

            >25             Lifesaving or protection of large    Only on a voluntary basis to persons populations                          fully aware of the risks involved

( 4. Limit dose to the lens of the eye to three (3) times the above values and doses to any other organ (including thyroid, skin and body extremities) to ten (10) times the above values.

5. Entry into radiation fields of greater than 25 Rem/hr or exposure in excess of 5 Rem TEDE shall not be permitted unless specifically authorized by the SEC.

6. In emergency situations where a exposure in excess of 25 rem TEDE would be required, the following additional criteria shall be considered:

a. Rescue personnel must be volunteers.

b. Rescue personnel should have a full awareness of the risks involved (See Attachment 2).

c. Other things being equal, volunteers above the age of 45 should be selected whenever possible for the purpose of avoiding unnecessary genetic effects.

d. Exposure under these conditions should be limited to once in a lifetime, and shall be included when calculating future lifetime permissible exposures.

1 PEP-330 Rev. 9 Page 17 of 571

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

23. 2009A NRC SRO 023INEW/H/2/AOP-0311N12009A NRC SRO/G2.3.141 Given the following plant conditions:

( - The plant is in Mode 6 with fuel movement in progress

                - The manipulator crane is latched on to a fuel assembly and transporting it to the upender The following occur:
               - Offsite Power is lost
               - Safety Buses are reenergized by the EDGs
               - The Fuel Handling SRO reports that Refueling Cavity Level is 18' 6" and rapidly lowering
               - CNMT Ventilation Isolation radiation monitors have increased to 115 mR/hr Which ONE of the following lists the action(s) required in accordance with AOP-031, Loss of Refueling Cavity Integrity for present plant conditions?

A. Place the fuel assembly in the Reactor Vessel AND evacuate unnecessary personnel B. Place the fuel assembly in a Spent Fuel Pool rack AND evacuate unnecessary personnel C~ Evacuate ALL personnel since no method is available for personnel to move the fuel assembly D. Evacuate ALL personnel since radiation levels are tOQ high for personneL to safely

               . store the fuer assembly

( Friday, December 26, 2008 1:06:56 PM 46

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 Plausibility and Answer Analysis ( From AOP-031 and AOP-031-BD A Incorrect. Unnecessary personnel would be evacuated while others stored the fuel assembly if radiation level was less than 150 mRlhr AND offsite power was available, but offsite power has been lost. The reactor vessel is an approved storage location. B Incorrect. Unnecessary personnel would be evacuated while others stored the fuel assembly if both radiation level was less than 150 mRlhr AND offsite power was available, but offsite power has been lost. A SFP rack is an approved storage location. C Correct. Offsite power has been lost and no method is available to store the fuel assembly so all personnel will be evacuated to minimize their exposure to this radiation level while no work can be performed. D Incorrect. All personnel would be evacuated if radiation level exceeded 150 mRlhr due to exposure, but that level has not been exceeded. Other procedures (AOP-013) that are not in progress for this event do evacuate at 100 mRlhr. ( KIA statement - Radiation Control - Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities Importance Rating: 3.4 3.8 Technical

Reference:

AOP-031 Rev. 16, pages 4 & 34 AOP-031-BD Rev. 3, page 45 References to be provided: None Learning Objective: AOP-LP-3.31, Obj 3 Question origin: NEW Comments: (KIA Match) A Radiation Hazard has been created by the lowering refueling cavity level and the candidate must recall the strategy in an abnormal condition to protect individuals in the CNMT and FHB. SRO justification: Requires assessing plant conditions and detailed recollection of the strategy in Attachment 1 of AOP-031 and its background document to come to the correct answer. Origin: NEW Cog Level:. H _..__________

 . ~_..~.~~~----=~~:..:..:~-.--.---*~-:..----*--~*RIrrerence:-~---xOP-(r3T-*-~~ -.--. -~.--~.-~

C Ref Provided?: N KeyWords' 2009ANRC-SR0'---------- KIA 1: G2.3.14 KJA 2: Friday, December 26, 2008 1 :06:57 PM 47

Ref for SRO #23, G2.3.14 LOSS OF REFUELING CAVITY INTEGRITY

11'----_ INSTRUCTIONS RESPONSE NOT OBTAINED L 3.0 OPERATOR ACTIONS NOTE This procedure contains no immediate actions.

01. VERIFY one CSIP RUNNING. 01. GO TO Step 3.

02. ADJUST CSIP flow to a maximum of 150 gpm AND ATTEMPT to maintain Refueling Cavity level.

NOTE Loss of Refueling Cavity integrity may require initiation of the HNP Emergency Plan. [C.1]

03. REFER TO PEP-110, Emergency Classification and Protective Action Recommendations, AND ENTER the EAL Network at

        - entry point X.

04. CHECK ALL fuel assemblies are *04. IMPLEMENT Attachment 1, Placing safely positioned in one of the Fuel in Safe Storage, while continuing following storage locations: with this procedure.

Reactor vessel
Spent Fuel Pool fuel rack
An area isolated from the Refueling Cavity by:
Spent Fuel Pool gate OR
Fuel Transfer Tube Gate Valve

(- AOP-031 I Rev. 16 I Page 4 of 36

Ref for SRO #23, G2.3.14 LOSS OF REFUELING CAVITY INTEGRITY Attachment 1 Sheet 1 of 2 Placing Fuel in Safe Storage

 -1               INSTRUCTIONS                            RESPONSE NOT OBTAINED            ~

01. EVACUATE all unnecessary personnel from CNMT and FHB.

02. DIRECT that each fuel assembly be placed in the nearest safe storage location listed below:

Reactor vessel
A Spent Fuel Pool fuel rack
An area physically isolated from the Refueling Cavity.

03. CHECK offsite power available. 03. EVACUATE all personnel from CNMT and FHB AND DIRECT Security to verify all personnel are clear.

(

4. CONTACT Health Physics AND INFORM them of the f-ollowing (as known):

0

Existence of Refueling Cavity leak 0
Affected area(s) 0
Evacuated area(s) 0
Any elevated radiation levels

05. DIRECT Health Physics to monitor radiation levels in CNMT areas where plant personnel are stationed.

( AOP-031 I Rev. 16 I Page 34 of 36

Ref for SRO #23, G2.3.14 LOSS OF REFUELING CAVITY INTEGRITY-BASIS DOCUMENT Attachment 1-Placing Fuel in Safe Storage Step Description 3 I: Check offsite power available. RNO: Evacuate all personnel from CNMT and FHB, and direct Security to verify all personnel are clear. If offsite power is not available, the Manipulator Crane will lose power. With no method of moving fuel assemblies available and Refueling Cavity level dropping, all personnel are evacuated, including refueling personnel, to minimize personnel exposure as described above. Because radiation levels could rise to levels capable of causing serious illness or death, Security verifies that all personnel are clear. 4 I: Contact Health Physics and inform them of the following (as known):

Existence of Refueling Cavity leak
Affected area(s)
Evacuated area(s)
Any elevated radiation levels Health Physics is responsible for controlling personnel access to CNMT, c\ and for monitoring radiological conditions in order to maintain habitability from a radiological standpoint. The operator is directed to provide them with the information needed to facilitate those functions.

5 I: Direct Health Physics to monitor radiatlor'llevers in CNMT areas where plant personnel are stationed. Radiation levels need to be monitored in order to determine how long personnel can safely remain to perform activities related to event mitigation. Health Physics has the capability of performing this function, and is assigned that responsibility. N6 I: CNMT Ventilation Isolation radiation monitors are set to alarm at less than or equal to 150 mR/hr for fuel movement. This provides the operator with one method for detecting rising radiation levels without Health Physics assistance. The radiation monitor is adjusted to this setpoint for refueling in accordance with Technical Specification 3.3.3.1 Table 3.3-6 Item 1a.

   ",n
  .'-If -VoJ I-DU I              Rev. 3               I                 Page 45 of 49

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

24. 2009A NRC SRO 024INEW/F12IPEP-230/N/2009A NRC SRO/G2.4.301

( Given the following plant conditions: 0815 Site Area Emergency declared 0829 State and County officials notified 0848 SEC upgraded declaration to a General Emergency 0855 A release to offsite is confirmed to be in progress Which ONE of the following is the NEXT notification that is required lAW PEP-230, Control Room Operations? When A. NRC 0915 B. NRC 0929 C~ State and County 0903 D. State and County 0910 Answer and plausibility of distractors A Incorrect. Plausible if candidate does not recognize that an upgrade in classification ( resets the 15 minute clock on notifications to state and county. B Incorrect. Plausible if candidate does not recognize that an upgrade in classification resets the 15 minute clock on notifications and belIeves the 1 hour on NRC notification is based off of 1 hour from State and County notifications vice event time. C Correct. An upgrade in emergency classification resets the 15 minute clock on notifying the State and County officials. o Incorrect. Plausible if candidate believes that a release in progress resets the 15 minute clock on notifying State and County officials. Friday, December 26, 2008 1:06:57 PM 48

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Emergency Procedures/Plans - Knowledge of events related to system ( operations/status that must be reported to internal organizations or outside agencies. Importance Rating: 2.7 4.1 Technical

Reference:

PEP-310 Rev. 23, page 3 References to be provided: None Learning Objective: Question origin: NEW Comments: SRO justification: Requires assessment of follow up notifications and Emergency Action Level Upgrade notification. Expected that the Emergency Communicator would understand 15 minutes for first notification but this question is evaluating follow up and upgrade. Origin: NEW Cog Level: F Difficulty: 2

Reference:

PEP-230 Ref Provided?: N KeyWords: 2009A NRC SRO KIA 1: G2.4.30 KIA 2: Friday, December 26, 2008 1 :06:57 PM 49

Ref for SRO #24, G2.4.30 1.0 PURPOSE The purpose of this procedure is to provide instructions and documentation for:

1. Requesting assistance from offsite support organizations (Immediate Response Organizations).

2. Notifying HNP Emergency Response Organization (ERO) personnel by automated and manual means.

3. Notification of offsite Emergency Response Organizations and authorities.

4. Notifications to the Nuclear Electric Insurance Limited (NEIL), Institute of Nuclear Power Operations (INPO) and American Nuclear Insurers (ANI).

2.0 INITIATING CONDITIONS

1. An emergency has been declared.

2. An event has occurred which requires a response from an offsite support organization (such as fire, medical, local law enforcement or the ERO).

3.0 GENERAL 3.1 Regulations and Other Commitments

1. Alerting of on site personnel via Public Address announcement is required within 15 minutes of event declaration.

2. Federal Regulations state "A licensee shall have the capability to notify responsible state and local governmental agencies within 15 minutes after declaring an emergency." This is satisfiedwhel1 contact is made with the first agency (state or county).

3. Notification of event declaration to the NRC is required "as soon as possible" and no later than 60 minutes after an event declaration.

4. Activation of the NRC ERDS data link is required within 60 minutes of an Alert or higher event declaration.

5. Notification to Institute of Nuclear Power Operations (INPO) and American Nuclear Insurers (ANI) must occur within four (4) hours after declaration of an Alert, Site Area Emergency, or General Emergency.

6. Nuclear Electric Insurance Limited (NEIL) notification is only applicable to events involving equipment damage.

7. The NRC Authentication Code is provided daily to the Main Control Room R5.2.9 (MCR). The NRC will provide this code during the morning plant status update.

The code will be valid at 0800. Until then the MCR will use the previous day's code. ( 1 PEP-310 Rev. 23 Page 3 of 53 1

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1

25. 2009A NRC SRO 025/NEW/F/4/T.S 3.5.2 BASES/N/2009A NRC SRO/G2.4.461 Given the following plant conditions:

                     - The plant is operating at 100% power
                     - The 'c' CSIP is under clearance with its breaker racked out for motor replacement
                     - The 'A' CSIP has suffered a bearing failure and is being placed under clearance
                     - The 'A' CSIP breaker has been racked out to support the clearance Which ONE of the following describes the expected alarm for these conditions and the Tech Spec bases for the ECCS Subsystems Limiting Conditions for Operations?

A'f ALB-001-6-5, ESF SYS TRN A Bypassed OR Inoperable; To supply sufficient core cooling to limit peak cladding temperature to within acceptable limits for all postulated break sizes B. ALB-001-6-5, ESF SYS TRN A Bypassed OR Inoperable; To supply sufficient borated water to keep the recovered core subcritical during the early reflooding phase of a Large Break LOCA C. ALB-006-1-3, CHRG Pumps A Trip OR Close CKT Trouble; To supply sufficient core cooling to limit peak cladding temperature to within acceptable limits for all postulated break sizes D. ALB-006-1-3, CHRG Pumps A Trip OR Close CKT Trouble; To supply sufficient borated water to keep the recovered core subcritical during the ( early reflooding phase of a Large Break LOCA Plausibility and Answer Analysis' A Correct. ALB-01-6-5, ESF SYS TRN A Bypassed OR Inoperable will be received and locked in during these conditions. The Tech Spec Bases listed is correct for ECCS Subsystems. B Incorrect. ALB-01-6-5, ESF SYS TRN A Bypassed OR Inoperable will be received and locked in during these conditions. The Tech Spec Bases listed is incorrect but this is plausible because it is correct for the ECCS Accumulators. C Incorrect. ALB-06-1-3, CHRG Pumps A Trip OR Close CKT Trouble is plausible because this alarm will be received during the racking process when the control power is removed but will clear when the breaker is racked out. The Tech Spec Bases listed is correct for ECCS Subsystems. o Incorrect. ALB-06-1-3, CHRG Pumps A Trip OR Close CKT Trouble is plausible because this alarm will be received during the racking process when the control power is removed but will clear when the breaker is racked out. The Tech Spec Bases listed is incorrect byffJ1ausLb1e_beJ:.ause it is*correGtforthe EGCS .... .__ .__ ~.~-. . .- ....

 - . .-.. _.._. __.._;4~I8tOrs-:--'------"-~-~

Friday, December 26, 2008 1:06:57 PM 50

QUESTIONS REPORT for 2009A NRC SRO ONLY QUESTIONS REV1 KIA statement - Emergency Procedures/Plans - Ability to verify that the alarms are ( consistent with the plant conditions. Importance Rating: 4.2 4.2 Technical

Reference:

APP-ALB-001 Rev. 18, page 23 APP-ALB-006 Rev. 21, page 5 APP-ESF-A Rev. 10, page 7 Tech Spec Bases 3/4.5.2 and 3/4.5.3 pg B 3/4 5-1,2 (pages 446,447) References to be provided: None Learning Objective: SIS Obj 10c Question origin: NEW Comments: SRO justification: Requires assessing plant conditions and prescribing a procedure to mitigate as all actions occur Origin: NEW Cog Level: F Difficulty: 4

Reference:

T.S 3.5.2 BASES Ref. Provided?: N KeYWords: 2009A NRC SRO KIA 1: G2.4.46 KIA 2: ~ .. Friday, December 26, 2008 1:06:57 PM 51

Ref for SRO #25, G2.4.46 6-5 DEVICES: NONE SETPOINT: N/A ESF SYS TRN A BYPASSED OR INOPERABLE REFLASH: NO OPERATOR ACTIONS:

1. CONFIRM alarm by checking for the following:

One or more lighted windows on ESF Bypass Panel A
Abnormal MCB indications for the system indicated in alarm on ESF Bypass Panel A

2. VERIFY Automatic Functions: None

3. PERFORM Corrective Actions:

a. REFER TO APP-ESF-A for the window indicated in alarm on ESF Bypass Panel A.

b. WHILE ALB-001-6-5 is ALARMED, MONITOR ESF Bypass Panel A AND VERIFY that additional alarms received on that panel are noted in a timely manner.

c. IF ALB-001-6-5 is ALARMED AND no windows are illuminated on ESF Bypass Panel A, THEN:

(1) PRESS the test pushbutton on ESF BYPASS Panel A to perform a lamp check, AND CHECK for burned out bulbs. (2) IF a burned out bulb is identified, THEN PERFORM corrective actions a and b above for the alarm with a burned out bulb. (3) CHECK power available to ARP-7 at DP-1A-2-27. (4) IF power is available AND no bulbs are burned out, THEN INITIATE corrective maintenance on the alarm circuitry.

d. IF maintenance is to be performed, THEN REFER TO OWP-ESF.

CAUSES:

1. Alarm condition on Engineered Safeguard Feature Bypass Panel A

2. Loss of power to ARP-7

3. Alarm circuit malfunction

REFERENCES:

1. APP-ESF-A

2. Technical Specification 3.3.2

3. 6-B-401 0585

4. OWP-ESF APP-ALB-001 I Rev. 18 I Page 23 of 36

Ref for SRO #25, G2.4.46 1-3 DEVICES: 62 relay in charging SETPOINT: N/A spring motor circuit CHRG PUMPS A 86/HR lockout relay TRIP OR CLOSE CKTTROUBLE REFLASH: NO OPERATOR ACTIONS:

1. CONFIRM alarm by checking the following:

EI-221 SA or EI-223SA, breaker current indication
Pump status indication

2. VERIFY Automatic Functions:

Charging pump will trip on overcurrent

3. PERFORM Corrective Actions:

a. REFER TO OP-156.02, AC Electrical Distribution, AND INVESTIGATE tripped breaker prior to reclosure.

b. CHECK for breaker malfunction or testing.

c. IF necessary, THEN START another charging pump.

d. REFER TO Technical Specifications 3.1.2.3, 3.1.2.4, 3.5.2, and 3.5.3, AND INITIATE action where appropriate.

( CAUSES:

1. Breaker for Charging Pump 1A-SA or 1C-SAB in test, fails to close, or control power off

2. Instantaneous overcurre-nt or rate of rise relay for Charging Pump 1A-SA or 1C-SAB energized

3. AC time overcurrent relay for Charging Pump 1A-SA or 1C-SAB energized

4. Alarm circuit malfunction

REFERENCES:

1. Technical Specifications 3.1.2.3, 3.1.2.4, 3.5.2, 3.5.3

2. 6-B-401 0225

3. OP-156.02, AC Electrical Distribution

( APP-ALB-006 I Rev. 21 I Page 5 of 23

Ref for SRO #25, G2.4.46 2-1 DEVICES: None applicable SET POINT: N/A SAFETY INJECTION SYSTEM REFLASH: NO OPERATOR ACTIONS:

1. CONFIRM alarm using:

a. Position indication for the following:

(1) 1CS-291 SA, Suction From RWST LCV-115B (2) 1SI-4 SA, Boron Injection Tank Outlet

b. CSIP 1A-SA and 1C-SAB indicating lights

c. RHR Pump 1A-SA indicating lights

2. VERIFY Automatic Functions: None

3. PERFORM Corrective Actions:

a. Locally CHECK status of the following breakers:

(1) 1A-SA-7 and 1A-SA-5 (CSIPs) (2) 1A35-SA-8D (1CS-291 SA) (3) 1A31-SA-4C (1SI-4 SA) (4) 1A2-SA-5A (RHR Pump 1A-SA) ( CAUSES: 1.. Both 1A-SA-5 (for 1A-SACSLP} and 1A-SA-7 (fer 1C-SAB CSt?) tackecf out or in TEST position

2. 1A2-SA5A (for RHR Pump 1A-SA) racked out or in TEST position

3. Motor or control circuit power loss to 1CS-291 SA or 1SI-4 SA

4. Alarm circuit failure

5. Performance of OWP-SW

REFERENCES:

1. Technical Specifications 3.1.2.2, 3.1.2.3, 3.1.2.4, 3.5.2, 3.5.3, 3.9.8.1, 3.9.8.2

2. 2166-B-401 0221,0242,0321,0407,0409,0586,0602

3. OWP-SW, Service Water APP-ESF-A I Rev. 10 I Page 70f26

Ref for SRO #25/ G2.4.46 3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met. The value of 66% indicated level ensures that a minimum of 7440 gallons is maintained in the accumulators. The maximum indicated level of 96% ensures that an adequate volume exists for nitrogen pressurization. The accumulator power operated isolation valves are considered to be "operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required. The limits for operation with an accumulator inoperable for any reason except an isolation valve closed or boron concentration not within limits minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. The boron in the accumulators contributes to the assumption that the combined ECeS water in the partially recovered core during the early refloodingphase of a large break LOeA is sufficient to keep that portion of the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on the available EeCS water and an insignificant effect on core subcriticality during reflood. Boiling of ECCS water in the core during reflood concentrates boron in the saturated liquid that remains in the core. In addition, current analysis demonstrates that the accumulators do not discharge following a large steam line break for HNP. Therefore, 72 hours is permitted to return the boron concentration to within limits. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required. 3/4.5.2 AND 3/4.5.3 fCCS SUBSYSTEMS The OPERABILITY of two independent Eces subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each EeCS subsystem provides long-term core cooling capability in the recirculation mode-- during

                      ------ - - -  ~ - ,- ---

the-- - accident rec()very period. SHEARON HARRIS - UNIT 1 B 3/4 5-1 Amendment No. 86

Ref for SRO #25, G2.4.46 li4.5 EMERGENCY CORE COOLING SYSTEMS ( BASES fees SUBSYSTEMS (Continuedl With the ReS temperature below 350°F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements. The limitation for a maximum of one charging/safety injection pump to be OPERABLE and the Surveillance Requirement to verify one charging/safety injection pump OPERABLE below 325°F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. SHEARON HARRIS - UNIT 1 B 3/4 5-la Amendment No. 86 I}}

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