Text

June 2003 Exam 50-369/03-301 & 50-370/03-301, Draft SRO Written Exam
	ML032370044
Person / Time
Site:	Mcguire, McGuire
Issue date:	06/16/2003
From:	; Operator Licensing and Human Performance Branch
To:	; Duke Energy Corp
References
	50-369/03-301, 50-370/03-301
	Download: ML032370044 (227)
	v • d • e

Draft Submittal - -_ .

MCGUIRE JUNE 2003 EXAM 50-36912003-301 AND 50-37012063-301

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JUNE 16 30,2003 1. Senior Reactor Operator Written Exam

06/11/03 WED 10:14 FAX 7 0 4 8 7 5 5004 HLP EXAM DEV Bank Question: 1073 Answer; A 1Pt(s) The following conditions exist on Unit 1 I Reactorpower is 100%

1A CA Pump is runningwith 1CAdOA (1 A CA Pump Disch to 1A S/G Control) and ICA-56 (1A CA Pump Disch to I S S/GControl) closed for post maintenancc testin Nil2 level in 1B S/Gincreases to 84% due to ICF-23 (1B S/G Control Valve) failing ope Which o m ofthe following statements conxctly describes the response of the CA system to the above conditions? A CA Pump remains running 1B CA pump auto starts ICA-60A and 1CA-56A fail open 1CA-44B (IB CA Pump Wirch to IC S/G Control) and 1CAdOB (1ECA Pump to Disch to 1D S/G Control) do not repositio Depressing the MD CA Modulating Valve Reset Train A pushbutton will cause 1CA-60A and ICA-56A to clos A CA Pump remains runnhg l B CA pump auto starts ICA-60A and 1CA-56A remain close CA-44B and 1CAdOB fail close Depressing the ND CA Modulating Valve Reset Train A pushbutton wiU cause ICA-60A and ICA-56A to open A CA Pump trips 1B CA pamp remains off ICA-60A and 1CA-56A remain close lCA-44B and ICA-40B do not change positio Depressing the MD CA Modulating Valve Reset Train A pushbutton will cause 1CA-60A and 1CA-56A to ope A CA Pump trips 1B CA Pump remains off 1CA-6OA and 1CA-56A fail open 1CA-44B and 1CA-40B fail ope Depressing the MD CA Modulating Valve Reset Train A pushbutton will muse 1CA-60A and 1CA-56A to clos Post-itb Fax Nole 7671 TO RnvlL & i From G & * A co.lOep Phone I Phone #

Ques-1073.doc Fax ff Fax #

06/11/03 WED 10:14 FAX 7 0 4 875 5094 HLP EXAM DEV moo2 Distracter Analysis:

k Correct: . Incorrect:

Plausible:. Incorrect:

Plausible Incorrect:

Plausible:.

Level: SRO KA. APE. 054 AA204 (4.2/4.3)

Lesson Plan Objective: CF-CA Obj. ## 4, OP-MC-ECC-BE Obj. # 13)

Source:New Level of knowledge: comprehension References:

1. OP-MC-CF-CA page 13, 2. OP-MC-ECC-ISE Q q C 33

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06/11/03 WED 10:14 FAX 704 875 5094 HLP EWU DEV 1 w> The following conditions exist on Unit 1:

Reactor power is 100%

1A CA Pump is running with 1CA-60A (1A CA Pump Disch to IA S/G Control) and 1CA-56 (IA CA Pump Disch to 1B S/G Control) closcd for post maintenance testin N/R level in 1B S/G increases to 84% due to 1CF-23 (1B S/G Control Valve) failing ope Which one of the following statementscorrectly describes the response of the CA system to the above conditions? A CA Pump remains running IS CA pump auto starts 1CA-60A and 1CA-56A fail open 1CA-44E (1B CA Pump Disch to 1C S/G Control) and 1CA-4OB (1B CA Pump to Disch to 1D SIG Control) do not repositio Depressing the MD CA Modulating Valve Reset T r a h A pushbutton will. cause 1CA-60A and 1CA-56A to clos A CA Pump remains running 1B CA pump auto starts 1CA-60A and 1CA-56A remain close lCA-44B and 1CA-4OB fail close Deprcssiog the MD CA Modulating Valve Reset Train A pushbutton will cause 1CA-BOA and 1CA-56A to open A CA Pump trips 1B CA pump remains off 1CA-60A and 1CA-56A remain close C A 4 E and 1CA-40E do not change positio Depressing the MD CA Modulating Valve Reset Trah A pushbutton will cause 1CA-60A and 1CA-56A to ope A CA Pump trips 1B CA Pump remains off 1CA-60A and 1CA-S6A fail open 1CA-44B and 1CA-40B fail ope Depressing the MD CA Modukating Valve Reset Train A pushbutton will cause 1CA-60A and 1CA-56A to clos /11/03 WED 10:15 FAX 704 875 5 0 9 4 HLP EXAM DEV @I004 DUjYE PO WER

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MCGUIRE OPERAnONS TRAINING-NLO NLOR LPRO LPSO LOR .0 .0 OBJECTIVES OBJECTIVE State the purpose of the CA Syste Sketch the system drawing (Fig. 7.1) including all major components and valves, show all tie-ins to associated system Describe all CA suction supply sources, including venting requirements and action .

Discuss the auto-start of the motor driven and turbine driven auxiliaty feedwater pumps, including concurrent BO/Ss signals and BO followed by S Describe the CA pump minimum flow and pump runout protectio Describe the function of the Auto Start Defeat Switches:

include permissive Describe the power supplies and steam supplies for the CA pump State the flow rates of the CA pump Describe the sources of make-up to the Auxiliary Feedwater Storage Tank, include destination of overflow from the Auxiliary Feedwater Storage Tan Describe the Interlock between the CA motor driven pump and the associated train RN pump. Include why the interlock is require Describe the interlock between the CA pump suction pressure and the RN assured makeur, valve Describe the interlock between the RN assured makeup valves (CA-15, CA-18) and the DG Hx Inlet Valve. Include why the interlock is require OP-IMC-CF-CA FOR TRAINING PURPOSES ONLY REV. 33 Page 7 of 83

_ I 06/11/03 WED 10:15 FAX 7 0 4 875 5 0 0 4 HLP EXAM DEV DUKEPOWEp F __--

MCGUIRE OPERATIONS TRAINING INTRODUCTION

-\i Purpose Tht3 auxhiary feedwater system is provided as a backup for the main feedwater syste It is, designed as a means to dissipate heat from the Reactor Coolant System when normal systems are not available. The auxiliary feedwater system may also be used in normal plant startupand shutdown, as main feedwater, when the flow is less than 3%

maximum design feedwater flo .2 General Description LDbjective # 2 1 Reier to Figure 7.1,7.2,7.3,7.1 The CA system assures required feedwater flow to the steam generators for reactor coolant thermal energy dissipation when the CF system is not available through loss of power or other malfunctions. The CA system is reqjired to operate until normal feedwater flow is restored or until the reactor coolant temperature is lowered to the point where the ND system can be utilized. The CA system is designed to start automatically for any event requiring emergency feedwate Sin= the CA system is the only source of makeup water to the steam generators for reaztor coolant heat removal when the main feedwater system becomes inoperable, it ha$ been designed with redundancy and diversity. The CA system contains two motor driven pumps and one steam turbine driven pump for each uni .0 COMPONENT DESCRIPTION Motor Driven CA Pumps jective # 4,7, a I The motor driven CA pumps are powered from essential power, ETA (pump A) and ETB (pump E). Each motor driven pump has a design flow rate of 450 gpm and is capable of supplying two steam generators. CA pump A supplies steam generators A and B while CA pump B supplies steam generators Cand Refar to Figure 7.12. The auto-start signals for the CA Motor Driven pumps are:

,Y4 detectors low-low level in any one SG (17%)

Trip of both Main Feedwater pumps AMSAC Both Feedwater pumps tripped s LOSS of flow to 314 SGs

Sssignal 13lackout signal

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OP-IMC-CF-CA FOR TRAINING PURPOSES ONLY REV. 33 Page 13 of 83

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06/11/03 WED 10:15 FAX 704 875 5 0 9 4 HLP EXAM DEV OiJKE POWER MCGUIRE 0

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OBJ ECTlVE

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1:3 List the setpoints, permissives, and logic required to initiat the following:

Containment Spray (NS)Actuation Phase " B Isolation Main Steam Isolation (MSI)

Main Feedwater Isolation (FWI)

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14 Explain the relationship between SSPS Testing and the operability of the Systems and functions actuated from the Engineered Safety Features Actuation Syste It; Discuss the purpose of the ESF Monitor L i e Panel (BOP Panel).

If; Conceming AP/1 or2/A/5500/35, ECCS Actuation During Plant Shutdow State the purpose of the A Recognize the symptoms that would require implementation of the A ' Conceming the Technical Specifications related to the Engineered Safeguards Actuation System:

Given the LCO title, state the LCO (including any COLR values) and applicabilit For any LCO's that have action required within one hour, state the actio Given a set of parameter values or system conditions, determine i f any Tech Spec LCOs is (are) not met and any action(s) required within one hou Given a set of plant parameters or system conditions and the appropriate Tech Specs, determine required action Discuss the bases for a given Tech Spec LCO or Safety Limi * - SRO Only

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OP.MGECCISE FOR TRAINING PURPOSES ONLY REV. 26 Page 7 of T7

06/11/03 WED 10:16 FAX 7 0 4 8 7 5 5 0 9 4 HLP EXAM DEV a007

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.DIIK&POWER , ' : MCGWRE OPERATIONS nZAINlNG

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Main Feedwater Isolation (FWI).is,initiatedby:

Safety Injection (Ss)

Reactor trip and low T-avg ( P 4 and 553°F on ' / 4 channels)

l-hgh High SI0 level 83% on '13 channels on 'IOSIG (P-14)

.* Manually ('ra pushbuttons)

Feedwater Isolation (Wl)Initiating Signal Automatic Actions to CF (Main Feedwater)

S: (Safety Injection)

FWI (Feedwater Isolation)

Turbine trip Both FWPT's trip P-4 and Low T-avg ,

FWI (Feedwater Isolation)

P-4 generates turbine trip WPT's rollback hold p- 14 .:.

- FWI (Feedwater Iso[ation)

Turbine trip Both FWPT's trip Manual FWI (Feedwater Isolation)

FWPT's rollback hold Vaives that close from FWI (Feedwater Isolation) signal S/G CF Control Valves (CF-32,23, 20, 17)

S/G CF Control Valve Bypasses (CF-104, 105, 106,107)

S/G CF Containment Isolations (CF-35,30, 28,26)

CF to CA Nozzle Isolations (CF-126, 127, 128, 129)

-0P.MGECGISE FOR TRAINING PURPOSES ONLY REV. 26 Page 33 of 77

08/11/03 WED 0 7 : 0 5 FAX 70.1 8 7 5 5 0 9 4 HLP EXAY DEV moo1 following conditions exist on Unit 1 RTP i s at 100%

1A CA kunp is running with 1 CA-60A (1A CA Pump Disch to 1A S/G Control)

and 1CA-56A (IA CA Pump Disch to 1A S/GControl) closcd for post maintenance testin NR level in 113 S/G is at 83% due to 1CF-23 (1B S/G Control.Valve) failing ope 'Miich of the following describepe response of the CA Syslem to the above conditions?

x A. A CA Pump remains muling 1I3 CA pump auto start CA-60A (1A CA Pump Disch to 1A S/G Control) and 1CA-56A (1A CA Pmlp nisch to 1A S/G Control) fail opei CA-44B (1B CA Pump Disc11 to 1C S/G Control) and 1CA-40B (1B CA Pump Disch to ID S/G Control) do not change positio Depressing the MD CA Modulating Valvc Rcset A 'I'rain Pushbuttoll Will cause ICA-60A & 1CA-56A to close*

B. 1A CA Pump remains runnin B CA Pump auto start ICAdOA (1A CA Pump Disch to 1A S/G Control) and 1CA-56A ( l h CA Pump Disch lo 1A S/G Control) remain close&

6 ICA-44R (1R CA Pump Disch to 1C S/G Control) and 1CA-40B (IB CA Pump

Disch 10 1TI S/G Control.) do not change position Depressing the MD CA Modulating Valve Resct A 'hhPushbutton will. cause lCA-60A & 1CA-56A to open C. 1A CA Pump .trips 1B CA Pump remains o f f 1CA-GOA (1A CA Pump Disch to IA S/l; control) and Disch to 1A S/G Control) remain clos

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08/11/03 WED 07:05 FAX 704 875 5 0 0 4 HLP EXAM DEV moo2 1CA-44B (1B CA Pump Disch to 1C S/G Control) (1B CA Pump

'4 Disch to 1D S/G Control) do not change position Depwssing the MD CA Modulating Valve Reset A Train Pushbutton will cause lCA-60A & 1CA-56A to ope D. 1A CA Pump trips 1B CA Pump remains off 1CA-GOA (1A CA Pump Disch to 1.4 S/G Control) and 1CA-56A (1A CA Pump Disch 10 1A S/G Control) fail opc CA-44B (1B CA Punip Disch to 1C S/GControl) and lCA-40U ( I S CA Pump Disch to 1D S/G Control) do not change positiom Depressing the MD CA Modulating Valve Reset A Train Pushbutton will causc 1CA-GOA W 1CA-56A to clos Pt(s) During a plant shutdown on Unit 1, the operators have blocked the CA auto-start signal by depressing the 'AUTO-START DEFEAT' switch. A subsequent loss of both main feedwater pumps occurred at 020 Given the following plant conditions at the times listed:

Time Condition 0200 0205 0210 0215 1) NCS temperature ("F) 557 558 558 559 2) NCS pressure (psig) 1903 1956 1976 1991 3) NRSGA(%) 16 15 14 13 4) NR SG B (%) 20 19 18 17 5) NRSGC(%) 19 18 17 16 6) NRSGD(%) 19 18 17 16 What time would the motor driven CA pumps restart automatically? Ques-16.1 .doc

Bank Question: 1 Answer: B 1 Pt(s) During a plant shutdown on Unit 1, the operators have blocked the CA auto-start signal by depressing the AUTO-START DEFEAT switch. A subsequent loss of both main feedwater pumps occurred at 020 Given the following plant conditions at the times listed:

Time Condition 0200 0205 0210 0215 1) NCS temperature (F) 557 558 558 559 2) NCS pressure (psig) 1903 1956 1976 1991 3) NRSGA(%) 16 15 14 13 4) NRSGB(%) 20 19 18 17 5) NRSGC(%) 19 18 17 16 6) NRSG D (%) 19 18 17 16 What time would the motor driven CA pumps restart automatically? Distracter Analysis A. - Incorrect Plausible Correct Plausible Incorrect Plausible Incorrect Plausible LEVEL: SRO Ques-16.1 .doc

SOURCE: Bank REFERENCES: OP-MC-CF-CApage 13 & 15 LESSON PLAN 0BJECTIVE:OP-MC-CF-CA,Obj. 4 & 6 LEVEL OF KNOWLEDGE: Memory W A APE 000054 AA2.04 (4.2/4.3)

Ques-16.1 .doc

APE OS4 Loss of Main Feedwater (MFW)

AA Ability to determine and inteipret the following as they apply to the Loss of Main Feedwater 0 :

(CFR:43.5 / 45.13)

AA2.01 AA2.02

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Occurrence of reactor andlor turbine trip . . . . . . . . . . . .

Differentiation between loss of all MFW and trip of one MFW pump . .1 .4 AA2.03 .

Conditions and reason.. for AFW uumo startuo . . . . . . . . . .. . . . .. . .2

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Statk ofMFW pumps, regulating and stop valves . . . . . . . . . . . . . . . .

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AA2.05 .7 AA2.06 AFW adjustments needed to maintain proper T-ave. and SIG level . . . . . .3 AA2.07 Reactor trip first-out panel indicator . . . . . . . . . . . . . . . . . . . . . . . . . 3.4* 3.9 AA2.08 Steam flow-feed trend recorder . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3*

NUREG-1122. Rev. 2 4.2-36 I .0 .0 OBJECTIVES

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OBJECTIVE S

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State the purpose of the CA Syste X Sketch the system drawing (Fig. 7.1) including all major X components and valves, show all tie-ins to associated svstem Describe all CA suction supply sources, including venting X requirements and action Discuss the auto-start of the motor driven and turbine driven X auxiliary feedwater pumps, including concurrent BO/&

signals and BO followed by S Describe the CA pump minimum flow and pump runout X protectio .Describe the function of the Auto Start Defeat Switches; X include permissive Describe the power supplies and steam supplies for the CA X pump State the flow rates of the CA pump X Describe the sources of make-up to the Auxiliary Feedwater X Storage Tank, include destination of overflow from the Auxiliary Feedwater Storage Tan Describe the interlock between the CA motor driven pump X and the associated train RN pump. Include why the interlock is reauire ~~ ~ ~ ~ ~ ~ ~

Describe the interlock between the CA pump suction pressure and the RN assured makeup valve Describe the interlock between the RN assured makeup X valves (CA-15, CA-18) and the DG Hx Inlet Valve. Include why the interlock is required.

OP-MC-CF-CA FOR TRAINING PURPOSES ONLY REV. 33 Page 7 of 83

- DUKE POWER MCGUIRE OPERATIONS TRAINING

, INTRODUCTION Purpose I Objective # 1 I The auxiliary feedwater system is provided as a backup for the main feedwater syste It is designed as a means to dissipate heat from the Reactor Coolant System when normal systems are not available. The auxiliary feedwater system may also be used in normal plant startup and shutdown, as main feedwater, when the flow is less than 3%

maximum design feedwater flo .2 General Description Objective # 2 1 Refer to Figure 7.1, 7.2, 7.3, 7.13. The CA system assures required feedwater flow to the steam generators for reactor coolant thermal energy dissipation when the CF system is not available through loss of power or other malfunctions. The CA system is required to operate until normal feedwater flow is restored or until the reactor coolant temperature is lowered to the point where the ND system can be utilized. The CA system is designed to start automatically for any event requiring emergency feedwate Since the CA system is the only source of makeup water to the steam generators for reactor coolant heat removal when the main feedwater system becomes inoperable, it has been designed with redundancy and diversity. The CA system contains two motor driven pumps and one steam turbine driven pump for each uni .0 COMPONENT DESCRIPTION Motor Driven CA Pumps Objective # 4, 7, 8 The motor driven CA pumps are powered from essential power, ETA (pump A) and ETB (pump 6). Each motor driven pump has a design flow rate of 450 gpm and is capable of supplying two steam generators, CA pump A supplies steam generators A and B while CA pump 6 supplies steam generators C and Refer to Figure 7.12. The auto-start signals for the CA Motor Driven pumps are:

2/4 detectors low-low level in any one SG (17%)

Trip of both Main Feedwater pumps AMSAC Both Feedwater pumps tripped LOSS of flow to 314 SGs SS signal Blackout signal OP-MC-CF-CA FOR TRAINING PURPOSES ONLY REV. 33 Page 13 of 83

DUKE POWER - MCGUIRE OP~RATIQNSTRAINING

~ An Auto-Start Defeat Switch can be used to defeat

) 2/4 low-low level in any SG Trip of both Main Feedwater pumps AMSAC (Both Main Feedwater Pumps Tripped)

NC System pressure must be below the P-1 1 setpoint (1955 psig) to enable the Auto-Start Defeat feature. The Auto-Start Defeat feature will "auto unblock" when pressure returns above the P-11 setpoin Objective # 10 The train related RN pump will automatically start upon any start (including Manual) of the corresponding CA pump to provide necessary coolin .2 Turbine Driven CA Pump Objective # 7, 8 Each unit has one Steam Turbine Driven CA pump. The turbine receives steam from

" B and "C" steamlines through two redundant valves. The turbine driven pump has a design flow rate of 900 gpm and supplies all four steam generators. Steam is admitted to the turbine through two piston operated isolation valves, SA-48ABC and SA-49A These valves are held closed by control air from normally energized solenoid valve De-energizing any solenoid valve will bleed-off the control air, open its piston operated valve and admit steam to the turbin I Objective # 4 I Refer to Figure 7.12. The auto-start signals for the CA Turbine Driven pump (which open SA-48ABC and SA-49AB) are:

2/4 detectors low-low level in any two SGs (17%)

Blackout (> 8 seconds)

1/1 detector from SSF SG Wide Range Low-Low Level on 2/4 SGs (72%) (only opens SA-48ABC)

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NOTE: If a Blackout occurs first followed by a Safety Injection, the Sequencer will reset the start signal to the Turbine Driven CA Pump. If the Turbine Driven CA Pump is running at the time of the Safety Injection, it will continue to run. If the Safety Injection occurs first or coincident with the Blackout, the Safety Injection will BLOCK the Turbine Driven CA Pump start because the sequencer selects the Priority Mode. (This does not affect the Low-Low SG Level auto start signal or the SSF Low-Low Level start signal.)

NOTE: The turbine driven pump will also start on loss of VI or loss of power to the solenoid valves, due to the fail-open design of the valves (not considered an auto-start)

OP-MC-CF-CA FOR TRAINING PURPOSES ONLY REV. 33 Page 15 of 83

i 1 P Given the following conditions:

0 Unit 1 is in a refueling outag Fuel movement is in progres b A leak has developed which has caused level to drop in the spent fuel poo The Spent Fuel Pool Level Low computer alarm has actuate Pool was initially at normal level and area radiation at 7 mremlh After 20 minutes the pool level has decreased further and area radiation is 18 mremlh Which one (1) of the following describes the operator response to the current conditions?

A Begin makeup to the pool from the Boric Acid Tank. to restore leve Move the fuel transfer cart to the reactor side and close IKF-122 (fuel Transfer Tube Block Valve). Move the fuel transfer cart to the spent fuel (pit) side and close IKF-122 (fuel Transfer Tube Block Valve). Place the weir gate in position and inflate the seal .1 doc

Bank Question: 3 Answer: C 1 P Given the following conditions:

Unit 1 is in a refueling outag Fuel movement is in progres A leak has developed which has caused level to drop in the spent fuel poo The Spent Fuel Pool Level Low computer alarm has actuate Pool was initially at normal level and area radiation at 7 mrem/h After 20 minutes the pool level has decreased further and area radiation is 18 mremlh Which one (1) of the following describes the operator response to the current conditions? Begin makeup to the pool from the Boric Acid Tank. to restore leve Move the fuel transfer cart to the reactor side and close IKF-122 (Fuel Transfer Tube Block Valve). Move the fuel transfer cart to the spent fuel (pit) side and close IKF-122 (Fuel Transfer Tube Block Valve). Place the weir gate in position and inflate the seals.

LEVEL: SRO

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SOURCE: BANK LEVEL OF KNOWLEDGE: Memory REFERENCES: OP-MC-FH-FC pages 19-25 odd only LESSON: OP-MC-FH-FC OBJECTIVE: OP-MC-FH-FC Obj. 6 K I A 036 AA2.02 (3.41411)

33.1 doc

APE 036 Fuel Handling Incidents

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IvaKe KNOWLEDGE . .

AK Knowledge of the operational implications of the following concepts aS they apply to Fuel Handling Incidents :

CFR 41.8 t41.10 145.3)

AK1.O1 Radiation exposure hazards ................................ , AK1.02 SDM ............................................. .8 AK1.03 Indications of approaching criticality ........................ .3 AK Knowledge of the interrelations between the Fuel Handling Incidents and the following:

(CFR 41.7 t 45.7)

AK2.01 Fuel handling equipment ................................ .5 AK2.02 Radiation monitoring equipment (prrable and installed) . . . . . . . . . . . . .9 AK Knowledge of the reasons for the following responses as they apply to the Fuel Handling Incidents:

(CFR 41.5,41.10 t 45.6 J 45.13)

AK3.01 Different inputs that will cause a reactor building evacuation . . . . . . . . . .1 AK3.02 Interlocks associated with fuel handling equipment . . . . . . . . . . . . . . . .6 AK3.03 Guidance contained in EOP for fuel handliing incident . . . . . . . . . . . . . .1 kk Ability to operate and I or monitor the following as they apply to the Fuel Handling Incidents:

(CFR 41.7 145.5 I45.6)

AAl.01 Reactor building containment purge ventilation system . . . . . . . . . . . . . .8 AA1.02 ARMsystem ....................................... .5 AA1.03 Reactor building containment evacuation alarm enable switch . . . . . . . . . .9 AA1.04 Fuel handling equipment during an incident .................... .7 A Ability to determine and interpret the following as they apply to the Fuel Handling Incidents:

(CFR: 43.5 I45.13)

AA2.01 ARM system indications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9 EAA2.02 &&mefi&'of:a'

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fub]:

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3c'.&u ~ .1 I AA2.03 Magnitude of potential radioactive release ..................... 3.1* 4.2*

'.4 NUREG-1122, Rev. 2 4.2-28 I

-_. DUKE POWER MCGUIRE OPERATIONS TRAINING CLASSROOM TIME (Hours)

NLO NLOR LPRO LPSO LOR .5 OBJECTIVE Describe the roles and responsibilities of Control Room Ooerators during Fuel Handling operation _ _ _ ~

Describe the roles and responsibilities of Fuel Handling SRO's during Fuel Handling operation Describe how monitoring of core reactivity is accomplished durina Fuel Handlin Deleted Describe the requirements that must be met before bypassing a Fuel Handling Interloc Zbnceming AP-25, Spent Fuel Damage; AP40, Loss of , ,.

Refueling Canal; and AP-41, Loss of Spent Fuel Coolingor Level:

State the purpose of the AP Given symptoms, state the AP and Case (if applicable)

Concerning the Technical Specifications related to the FC System; Given the LCO title, state the LCO (including any COLR values) and applicabilit For any LCO's that have action required within one hour, state the actio Given a set of parameter values or system conditions, determine if any Tech. Spec. is (are) not met and any action(s) required within one hou Given a set of plant parameters values or system conditions and the appropriate Tech Specs, determine required action(s).

Discuss the basis for a given Tech. Spec. LCO or Safety Limi * SRO only OP-MC-FH-FC FOR TRAINING PURPOSES ONLY REV. f 2 Page 5 of 43

DUKE POWER MCGUIRE OPERATIONS TRAINING The Symptoms include:

EMF36 UNIT VENT GAS HI RAD alarm EMF38 CONTAINMENT PART HI RAD alarm EMF39 CONTAINMENT GAS HI RAD alarm EMF40 CONTAINMENT IODINE HI RAD alarm EMF42 FUEL BLDG VENT HI RAD alarm EMF16 CONTAINMENT REFUELING BRIDGE alarm (2 - EMF3 on Unit 2)

EMF17 SPENT FUEL BLDG REFUEL BRDG alarm (2 - EMF4 on Unit 2)

Gas bubbles originating from the damaged assemblies Visible evidence of damage with the potential of radioactive releases Operator Actions CAUTION Damage to the rubber Reactor Vessel Cavity Seal may occur if an assembly is dropped on or near it.

Announce on page. If in containment, evacuate containment, assemble in contaminated change room and refer to RP/O/A/5700/11, Conducting a Site Assembly, Site Evacuation, or Containment Evacuation. Isolate containment: stop VP fans, ensure VP valves close, stop any VQ release, ensure equipment hatch closed, ensure one airlock door closed, dispatch Operator to move conveyor to Spent Fuel Pool Building, dispatch Operator to close KF-122. If high containment radiation exists, place Aux Carbon Filters in service per OP. Place Refueling Cavity in purification per OP.

If in Spent Fuel Building, evacuate Spent Fuel Pool area, assemble in contaminated change room. Isolate Spent Fuel Pool area: Check if VF EXH BYP DAMPER closed lite lit, and if not, place it's control switch to "CLOSE", and close the doors to the Spent Fuel Pool area. Ensure KF purification loop in service per OP.

Refer to RP/O/A/5700/00, Classification of Emergency.

3.2.2 AP/l/A/5500/40, LOSS OF REFUELING CANAL LEVEL The purpose is to provide actions in the event of loss of water in the refueling canal.

The Symptoms include:

"Spent Fuel Pool Level Low" computer alarm Decreasing level in refueling canal

"lncore lnst Room Sump Hi Level" alarm EMF16 CONTAINMENT REFUELING BRIDGE alarm (2 - EMF3 on Unit 2)

EMF17 SPENT FUEL BLDG REFUEL BRDG alarm (2 - EMF4 on Unit 2)

OP-MC-FH-FC FOR TRAINING PURPOSES ONLY REV. 12 Page 19 of 43

r DUKE POWER MCGUIRE OPERATIONS TRAINING ODerator Actions NOTE Any available core location may be used when lowering a fuel assembly during emergency condition If fuel movement is in progress: lower any assembly in the reactor building crane to fully down in the core, any assembly in the spent fuel crane to fully down, and any assembly in the upender to fully down. If they won't lower otherwise, manually release the brake and hand crank the hoist down. NOTE: The sequence for lowering the hoist manually should be to put the emergency handwheel on the end of the hoist motor, hold it steady, while another person screws in the brake release (star shaped knob on a threaded stud) which when threaded in forces the brake disengaged. Care should be taken to remove the handwheel before electric operation of the hoist motor. The upender is similar. The bridge and trolly brake release is a lever, otherwise similar. Dispatch Operator to locally move fuel transfer cart to the spent fuel (pit) side. Stop FWST Pump and close FW-13, and dispatch Operator to locally close KF-122. If KF-122 cannot be closed, then notify RP to begin surveys, consider installing the weir gate, and isolate the Spent Fuel Building (VF in filter mode and doors closed). Evacuate nonessential personnel from containment and Spent Fuel Buildin Try to identify and correct the cause of decreasing level. Verify seal integrity and air pressure to the Rx Vessel cavity seal and the Rx Vessel nozzle inspection port seals, and if not, reestablish VI to seals. Dispatch an Operator to locally ensure the Refueling Cavity Drains are closed. Check the S/G Nozzle Dams. Refer to AP/19, Loss of ND or ND System Leakage, while continuing with this procedur Makeup to the canal per OP/1/A/6200/13. CAUTION: Makeup to the SFP could dilute NC system boron concentratio Monitor the Spent Fuel Pool level, If it gets to minus two feet, stop the KF Pump and turn off the lights. Initiated makeup per OP. If pool level low enough for radiation hazard, makeup from R Ensure Containment Integrity with equipment hatch and airlock doors closed. If time permits, turn off canal underwater lights before they become uncovered. If necessary due to increasing radiation levels, consider using ND or NS to transfer water from the containment sump to the FWST for additional makeup capabilit Refer to RP/OIN5700/00, Classification of Emergenc OP-MC-FH-FC FOR TRAINING PURPOSES ONLY REV. 12 Page 21 of 43

DUKE POWER MCGUIRE OPERATIONS TRAINING 3.2.3 AP/l/N5500/41, LOSS OF Spent Fuel Pool Coolinq or Level The purpose of this procedure is to provide actions to take in the event of loss of Spent Fuel Cooling for the following cases:

Case 1 Loss of Spent Fuel Pool Cooling Case 2 Loss of Spent Fuel Pool Level Case 1 Loss of Spent Fuel Pool Coolinq The Symptoms include:

"SPENT FUEL POOL TEMP HI" computer alarm Both KF Pumps off Operator Actions If fuel handling is in progress, lower any fuel assembly in the crane, radioactive component, or fuel assembly in the upender to fully down. Release brake and hand crank hoist or upender, if required.

If KC is aligned to either ND Hx, ensure all KC Pumps are running per KC OP (Encl 4.3, Shifting Trains). Maintain KC flow to less than 4000 gpm per pump in subsequent steps.

Ensure at least one train of KC is aligned to the AB nonessential header. If it's not because of a LOCA, then Enclosure 1 is used to restore cooling during an extended LOCA. Throttle open the appropriate KF Hx Outlet Flow valve for the KF Hx to be placed in service. Check the KF Pump is running, and if not start it. Check the Spent Fuel Pool temperature going down.

If KF Pump can't be started or pool temperature is not going down, place VF in filter mode, close the doors, and notify RP. If boiling occurs, begin makeup per KF OP (Encl 4.4, Spent Fuel Pool Level Control).

Refer to RP/O/N5700/00, Classification of Emergenc Case 2 Loss of Spent Fuel Pool Level The Symptoms include:

"SPENT FUEL POOL LEVEL LO" computer alarm Level in Spent Fuel Pool going down EMF17 SPENT FUEL BLDG REFUEL BRDG alarm (2 - EMF4 on Unit 2)

OP-MC-FH-FC FOR TRAINING PURPOSES ONLY REV. 12 Page 23 of 43

DUKE POWER MCGUIRE OPERATIONS TRAINING OPERATOR ACTIONS Announce on Page.

Dispatch an Operator to check the Pool is isolated from the refueling canal (either KF-122 closed, weir gate installed, or transfer tube blind flange installed). If not, go to AP/40, Loss of Refueling Canal Level.

If fuel handling is in progress, lower any fuel assembly in the crane, radioactive component, or fuel assembly in the upender to fully down. Release brake and hand crank hoist or upender, if required.

If the level is lower than minus 2 feet, stop the KF Pump, deenergize the underwater lights, place VF in the filter mode, close the doors, and notify RP. Initiate makeup per KF OP (Encl 4.4, Spent Fuel Pool Level Control).

Dispatch an Operator to locate and isolate the leak. When the leak is isolated and level has been returned to normal, start the KF Pump per the OP.

OP-MC-FH-FC FOR TRAINING PURPOSES ONLY REV. 12 Page 25 of 43

1 P Unit 1 is in mode Given the following conditions:

(1) Surveillance testing has been recently completed on the ice condenser (2) The surveillance test was not satisfactory as described below Which one (1) of the following situations meets the requirements for a one hour tech spec LCO? The ice condenser door position monitoring system was declared inoperable when one door did not indicate in the open position during a surveillance test. The door was left in the open positio The ice bed was declared inoperable when it was determined that it failed a surveillance test based on total ice weight less than 2,099,790 pounds at a 95% level of confidenc The Ice Bed Temperature Monitoring System was declared inoperable

-

when it failed a Tech Spec surveillance test channel check failur The ice condenser intermediate deck door was declared inoperable when it was discovered to be obstructed from opening by ice and debris.

40.1 doc

Bank Question: 4 Answer: D 1 P Unit 1 is in mode Given the following conditions:

(1) Surveillance testing has been recently completed on the ice condenser (2) The surveillancetest was not satisfactory as described below Which one ( I ) of the following situations meets the requirements for a one hour tech spec LCO? The ice condenser door position monitoring system was declared inoperable when one door did not indicate in the open position during a surveillance test. The door was left in the open positio The ice bed was declared inoperable when it was determined that it failed a surveillance test based on total ice weight less than 2,099,790 pounds at a 95% level of confidenc The Ice Bed Temperature Monitoring System was declared inoperable

-

when it failed a Tech Spec surveillance test channel check failur The ice condenser intermediate deck door was declared inoperable when it was discovered to be obstructed from opening by ice and debris.

LEVEL: SRO SOURCE: BANK REFERENCES: Tech Spec 3.6.13

-.

LESSON: OP-MC-CNT-NF OBJECTIVE: OP-MC-CNT-NF Obj. 21 KIA SYS 025 G2.2.22 (3.4/4.1)

LEVEL OF KNOWLEDGE: Memory 40.1 doc

\ Equipment Control (Continued)

2.2.18 Knowledge of the process for managing maintenance activities during shutdown operation (CFR: 43.5 145.13)

IMPORTANCE RO SRO .2.19 Knowledge of maintenance work order requirement (CFR: 43.5 I45.13)

IMPORTANCE RO SRO .2.20 Knowledge of the process for managing troubleshooting activitie (CFR: 43.5 145.13)

IMPORTANCE RO SRO .2.21 Knowledge of pre- and post-maintenance operability requirement (CFR: 43.2)

IMPORTANCE RO SRO .. 1 2.2.23 Ability to track limiting conditions for operation (CFR: 43.2 145.13)

IMPORTANCE RO SRO .2.24 Ability to analyze the affect of maintenance activities on LCO statu (CFR: 43.2 I45.13)

IMPORTANCE RO SRO .2.25 Knowledge of bases in technical specifications for limiting conditions for operations and safety limit (CFR: 43.2)

IMPORTANCE RO SRO .2.26 Knowledge of refueling administrative requirement (CFR: 43.5 I45.13)

IMPORTANCE RO SRO .2.27 Knowledge of the refueling proces (CFR: 43.6 145.13)

IMPORTANCE RO SRO NUREG-1122. Rev. 2 I

DUKE POWER MCGUIRE OPERATIONS TRAINING

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L L OBJECTIVE P 0 R R-0 -

Concerning the Technical Specifications related to the Ice Condenser Given the LCO title, state the LCO (including any X X COLR values) and applicabilit For any LCOs that have action required within one X X hour, state the actio Given a set of parameter values or system conditions, X X determine if any Tech Spec LCOs is(are) not met and any action(s) required within one hou Given a set of parameter values or system conditions X X and the appropriate Tech Spec, determine required action(s).

Discuss the bases for a given Tech. Spec. LCO or *

Safety Limi * SRO ONLY OP-MC-CNT-NF FOR TRAINING PURPOSES ONLY REV. 20 Page 1 1 of 113

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\, 3.6 CONTAINMENT SYSTEMS 3.6.13 Ice Condenser Doors LCO 3.6.13 The ice condenser inlet doors, intermediate deck doors, and top deck doors shall be OPERABLE and close APPLICABILITY: MODES 1,2,3, and ACTIONS Separate Condition entry is allowed for each ice condenser doo CONDITION REQUIREDACTION COMPLETION TIME

~ ~~ ~~ ~~~~ ~~

.* . .. ..

.. . A,, . 'Oneorinore ice . Restore door td 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

. ,

condenser doors OPERABLE statu * . inoper86le due to being *

physically restrained from openin One or more ice Verify maximum ice bed 3nce per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months condenser doors temperature is 5 27° inoperable for reasons other than Condition A AND or not close Restore ice condenser door 14 days to OPERABLE status and closed position (continued)

McGuire Units 1 and 2 3.6.13-1 Amendment Nos. 184/166

Ice Condenser Doors 3.6.13 CONDITION REQUIRED ACTION COMPLETION TIME Required Action and Restore ice condenser door 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months associated Completion to OPERABLE status and Time of Condition B not closed positio Required Action and Be in MODE hours associated Completion Time of Condition A or C AND not me Be in MODE hours SURVEILLANCE REQUIREMENTS

. .

.... ..

;.

.C

... . SURVEl.LLANCE FREQUE~CY

. *.

' ... SR 3.6.13.4,. . Verify all inlet doors indicate closed by the Inlet Door 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

,.

. . Position Monitoring Syste SR 3.6.1 Verify, by visual inspection, each intermediate deck door 7 days is closed and not impaired by ice, frost, or debri ~ ~~ ~ ~~

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SR 3.6.7 Verify, by visual inspection, each top deck door: 92 days Is in place; and Has no condensation, frost, or ice formed on the door that would restrict its openin (continued)

McGuire Units 1 and 2 3.6.13-2 Amendment Nos. 184/166

>4 ice Condenser Doors -4 3.6.13 SURVEILLANCE REQUIREMENTS (continued)

\ SURVEILLANCE FREQUENCY S R 3.6.13.4 Verify, by visual inspection, each inlet door is not 18 months impairedby ice, frost, or debri S R 3.6.13.5 Verify torque required to cause each inlet door to begin 18 months to open is S 675 in-l S R 3.6.1 Perform a torque test on each inlet doo months SR 3.6.1 Verify for each intermediate deck door: 18 months No visual evidence of structural deterioration: Free movement of the vent assemblies; and Free movement of the door.

...

I .

'-.l McGuire Units 1 and 2 3.6.13-3 Amendment Nos. 1841166

06/10/03 TUE 11:32 FAX 7 0 4 875 5 0 9 4 HLP EXAM DEV 001 Bank Question: 1072 Answer; A 1 WS) As an SRO working on a Complex Maintenance Plan you are asked to evaluate four possible work teams who must repair filter housing in a 1500trii7em/hr radiation fiel Which one of the following work tcam5 would maintain station ALARA? A qualified male worker who has previously performed this task. He can complete this job in 20 minutes. This worker has exceeded his Alert level for exposure and will require a dose extensio Two male workers who are qualified to perform the tas Together they can perform the task in 15 minutes. Both workers have already accumulated 325 &em this yea A team of a female worker who i s qualified to perform the task and a male worker who needs to qualify to this task. The female is a dedared pregnant worker. The team &ll need 15 minutes to complete the task, The female has no dose and the male worker has 200 mRem for tbe yea A team of a male and female both are q u a e d to the task but will take 20 minutes to complete the task. Each has less than 100-ern this yea Correct: 500 mR tota Incorrect: 750 per mrem total Plausible Incorrect: Declared prcgnant wo*k Plausible: Incorrect: 1000 mrem total Level: SRO

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mi

06/09/03 $ION 09:OQ FAX 7 0 4 875 5094 HLP EXAM DEV a 003 Bank Question: 1072 Answer: A 1Pt(s) As an SRO working on a Complex Maiiitenmcc Plan you are asked to evaluate four possible work teams who must repair filter housing in a 1500 mremhr radiation fiel Which one o f the following work teams would maintain station ALARA? A qualified male worker who has previously performed this task. He can complete this job h 20 miuutes. This worker has excccded his Alert level for exposure and will require a dose extensio Two male workers who are qualified to perform the task Together they can perform the task in 15 miuute A team of a female worker who is qualified to perform the task and a male worker who needs to qualify to this task. Both workers have low dose, but the female h si declared pregnant worke A team of a male and female both are qualilied to the task but will take 20 minutes to complete the task. Each has less than 100 mrem this yea I _-____-___-_

Distracter Analysis:

-- --------------------- Correct: 500 nift tota B, Incorre& 750permem total PlausibIe Incorrect: Declared pregnant worke Plausible: lncorrcct; 1000 mrem total Level: SRO KA: G2.3.2 (2.512 9)

Lesson Plw Objective: RAD RP Ojb. 135 Source: New Level of knowiedge: comprehension References:

Ques-lO72.doc

06/09/03 HON 0 9 : 0 9 FAX 7 0 4 875 5094 HLP EXAY DEV a004 1.OP-MC-RAD-Rp page 73 Ques-1072.doc

1 Pt(s) A team of workers must repack the seals on a pump in a 1500 mrern/hr extra high radiation are Which one of the following work teams and estimated repair times would maintain ALAFW?

A 10 people working for 20 minutes people working for 30 minutes people working for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months people working for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Ques-124.l.doc

Bank Question: 12 Answer: B 1 Pt(s) A team of workers must repack the seals on a pump in a 1500 mremlhr extra high radiation are Which one of the following work teams and estimated repair times would maintain ALARA? people working for 20 minutes people working for 30 minutes people working for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months people working for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Distracter Analysis: Incorrect: six people can accomplish the job with 4 112 Re Plausible: Each individual would have the least exposur Correct: Incorrect: six people can accomplish the job with 4 1/2 Re Plausible: fewest individuals not exceeding the admin dose limi Incorrect: six people can accomplish the job with 4 1/2 Re Plausible: Exposes the fewest individual Level: SRO KA: G2.3.2 (232.9)

Lesson Plan Objective: RAD-RP Obj. 135 Source: Bank Level of knowledge: comprehension References:

1. OP-MC-RAD-RP page 73 Ques-124.l.doc

--? Radiation Control 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirement (CFR: 41.12 143.4.45.9 145.10)

IMPORTANCE RO SRO .3.3 Knowledge of SRO responsibfities for auxiliary systems that ace outside

, the control room (e.g., waste disposal and handling systems).

(CFR: 43.4 145.10)

IMPORTANCE RO SRO .3.4 Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorize (CFR 43.4 145.10)

IMPORTANCE RO SRO 3.1

/--.

2.3.5 Knowledge of use and function of personnel monitoring equipmen (CFR 41.11 145.9)

IMPORTANCE RO SRO .3.6 Knowledge of the requirements for reviewing and approving release permit (CFR: 43.4 145.10)

IMPORTANCE RO SRO .3.7 Knowledge of the process for preparing a radiation work permi (CFR 41.10 / 45.12)

IMPORTANCE RO SRO .3.8 Knowledge of the process for performing a planned gaseous radioactive releas (CFR 43.4 I45.10)

IMPORTANCE RO SRO .3.9 Knowledge of the process for performing a containment purg (CFR 43.4 I45.10)

IMPORTANCE RO SRO NUREG-1122,Rev. 2 ,

DUKE POWER

.."-.

-. -

N L

c OBJECTIVE 0

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133 Identify the type of exposures ( Le., Emergency Exposure, X Planned Special Exposure, etc. ) that are included as part of a radiation workers total allowable exposur ~ ~~ ~ ~

134 Discuss the way in which exposure to neutron radiation is X X considered for the following:

Duke's administrative policy requirements for neutron dosimetry for entering the RCA How neutron dose accumulation is tracked between the times that the individual's TLD is read I

~ ~~ ~~ ~

135 State the goals and efforts of the AURA Progra lxlx 136 For the following areas: x x Define and describe MNS Unrestricted Area Describe MNS Restricted Area Describe MNS RCA

-~ ~

137 State the doselimitassociated M z with an unrestricted are x x 138 List the requirements for wearing dosimetry devices inside X X and outside the RC State the additional controls placed on entry/access to: x x High Radiation Area Very High Radiation Area

- Extra High Radiation Area 140 Explain when a RWP is require x x 141 Discuss how long a RWP or SRWP is vali x x 142 Correctly interpret the information on the Daily Dose Repor X

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143 State the purpose of protective clothin x x OP-hfC-RAD-RP FOR TRAINING PURPOSES ONLY REV. 01 Page 45 of 93

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DUKE POWER -

MCGUIRE OPERATIONS TRAINING Regulatory Guide 8.14 requires a personnel neutron dosimeter if the neutron dose equivalent is likely to exceed 100 mrem in a quarter. Duke Power has an administrative requirement which requires all personnel entering the RCA to wear a TLD ( which measures neutron dose equivalent ). Estimation of neutron exposure is a method used to temporarily track exposure until the TLD is processed. Estimated neutron exposure tracking for personnel is required if the neutron dose equivalent is likely to exceed 10 mrem per entry or per job if consecutive multiple entries are required. There are two methods used to estimate neutron exposure:

One method is to measure the neutron dose fate and then calculate the exposure based on stay tim The second method is to determine the gamma exposure dose and neutron exposure dose for the given area. If it is determined that the neutron to gamma ratio is essentially constant during the period of personnel exposure, then a gammdneutron ratio can be utilized. The gamma dose received can be ratioed to find the neutron dose receive A U R A is a philosophy aimed at the minimizing exposure thru a management commitment. The goals and efforts of the McGuire Nuclear station Program are simple:

u To maintainthe annual dose to each individualA U R A To maintain the collective dose ( total person-Rem ) ALARA Both points have to be considered simultaneously, as one without the other is not A U R A . Radiation Areas and Access Controls 1 ObjccthreM36 I

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Unrestricted Area an area, access to which is neither limited nor controlled by the licensee. At McGuire this is the area out side the protected area fenc Restricted Area - an area, access to which is limited by the licensee for the purpose of protectingindividualsagainst undue risks from exposure to radiation and radioactive materials. The restrictedarea at McGuire is enclosed by the protected area fenc P-MC.RADRP FOR TRAINING PURPOSES ONLY RV. 01 Page 73 of 93

1 P Unit 1 is preparing for a reactor start up following a refueling outage. Given the following conditions:

Tav,=515"F Plant heatup in progress using NCPs At 0200, a Station Engineer reports that a mistake had been made in analyzing the containment Appendix J Leak Rate Test results that were conducted prior to exceeding 200 OF in mode 5. Reanalysis indicated that the combined containment leak rate (Type A) had exceeded 1.O L Which one of the following actions are required by Tech Specs in response to this situation?

REFERENCES PROVIDED Commence a plant cooldown to reach mode 5 within 42 hour4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months Commence a plant cooldown to reach mode 5 within 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Commence a plant cooldown to reach mode 5 within 37 hour4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months Commence a plant cooldown to reach mode 5 within 43 hour4.976852e-4 days 0.0119 hours 7.109788e-5 weeks 1.63615e-5 months ___._.___.___.______--.---.---.------

Ques-207.l.doc

Bank Question: 207. I Answer: C 1 P Unit 1 is preparing for a reactor start up following a refueling outage. Given the following conditions:

TavQ = 515°F Plant heatup in progress using NCPs At 0200, a Station Engineer reports that a mistake had been made in analyzing the containment Appendix J Leak Rate Test results that were conducted prior to exceeding 200 OF in mode 5. Reanalysis indicated that the combined containment leak rate (Type A) had exceeded 1.O L,.

Which one of the following actions are required by Tech Specs in response to this situation?

REFERENCES PROVIDED Tech Spec 3.6.land Bases Commence a plant cooldown to reach mode 5 within 42 hour4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months Commence a plant cooldown to reach mode 5 within 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Commence a plant cooldown to reach mode 5 within 37 hour4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months Commence a plant cooldown to reach mode 5 within 43 hour4.976852e-4 days 0.0119 hours 7.109788e-5 weeks 1.63615e-5 months ____._______________-----------

Distracter Analysis: Incorrect Plausible: Incorrect:

Plausible: Correct:

Plausible: Incorrect Plausible:

SOURCE: BANK LEVEL: SRO LEVEL OF KNOWLEDGE: ANALYSIS KA: S Y S 103 A2.01 (2.0*/2.6)

Ques-207.l.doc

OBJECTIVES: OP-MC-ADM-TS Obj. 2 REFERENCES: Tech Spec 3.6.1 and Bases OP-MC-ADM-TS pages 29,3 1 Ques-207.l.doc

SYSTEM: 103 Containment System K6 Knowledge of the effect of a loss or malfunction on the following will have on the containment system:

(CFR: 41.7 145.7)

K6.01 Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1' K6.02 Controllers and psitioners . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1'

K6.03 Pumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .6 K6.04 Heat exchangers and condensers ........................... .7 K6.05 Breakers, relays, and disconnects .......................... .7 K6.06 Sensors and detectors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 BBILLTY A1 Ability to predict andlor monitor changes in parameters (to prevent exceeding design limits) m i a t e d with operating the containment system controls including:

(CFR: 41.5 145.5)

A1.O1 Containment pressure, temperature, and humidity ................ .1 A2 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system-and @) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations (CFR: 41.5 143.5 145.3 145.13)

......

Ih7g@iWl I A2.02 Necessarv Dlant conditions for work in containment . . . . . . . . . . . . . . .

I . .2*

A2 03 Phase A and B isolation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5* 3.8*

A2 04 Containment evacuation (including recognition of the alarm) . . . . . . . . . 3.5' 3.6*

A2.05 Emergency containment entry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9 A3 Ability to monitor automatic operation of the contain-ment system, including:

(CFR: 41.7 145.5)

A3.01 Containment isolation .................................. .2 A4 Ability to manually operate andlor monitor in the control room:

(CFR: 41.7 145.5 to 45.8)

A4.01 Flow control, pressure control, and temperature control valves, including pneumatic valve controller . . . . . . . . . . . . . . . . . . . 3.2* A4.02 Excess letdown divert valves to reactor coolant drain tank . . . . . . . . . . . 2.1, 2.2*

A4.03 ESF slave relays . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.7' 2.7*

A4 04 Phase A and phase B resets . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5* 3.5*

3.5-19 NLREG-1122, Rev. 2 1

Containment 3. .6 CONTAINMENT SYSTEMS 3.6.1 Containment LCO 3. Containment shall be OPERABL APPLICABILITY: MODES 1,2, 3, and 4 ,

ACTIONS CONDITION REQUIRED ACTION COMPLETIONTIME Containment inoperable. Restore containment to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months OPERABLE statu . Required Action and Be in MODE hours associated Completion Time not me AND Be in MODE hours

.

McGuire Units 1 and 2 3.6.1-1 Amendment Nos.207 & 188

Containment 3.6.1

-.. SURVEILLANCE REQUIREMENTS

',

SURVEILLANCE FREQUENCY SR 3.6. . The space between each dual ply bellows assembly on penetrationsbetween the containment building and annulus shall be vented to the annulus during Type A test I 2. Following each Type A test, the space between each dual-ply bellows assembly shall be subjected to a low pressure test at 3 to 5 psig to verify no detectable leakage, or the assembly shall be subjectedto a leak test with the pressure on the containment side of the assembly at P . Type C tests on penetrations M372 and M373 may be perforred without draining the glycol-water mixture from the seats of their diaphragm valves if meeting a zero indicated leakage rate (not including instrument error).

________^___________--------------~----------------------------------------

Perform required visual examinations and leakage rate .n accordance with testing except for containment airlock testing, in the Containment accordance with the Containment Leakage Rate Testing Leakage Rate Progra Testing Program I

.I, McGuire Units 1 and 2 3.6.1-2 Amendment Nos. 207 & 1

Containment 0 3.6.1

,-.

B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment BASES BACKGROUND The containment is a free standing steel pressure vessel surrounded by a reinforced concrete reactor building. The containment vessel, including all its penetrations, is a low leakage steel shell designed to contain the radioactive material that may be releasedfrom the reactor core following a design basis Loss of Coolant Accident (LOCA).

Additionally, the containment vessel and reactor building provide shielding from the fission productsthat may be present in the containment atmosphere following accident condition The containment vessel is a vertical cylindricalsteel pressure vessel with hemispherical dome and a flat circular base. It is completely endosed by a reinforced concrete reactor building. An annular space exists between the walls and domes of the steel containment vessel and the concrete reactor building to provide for the collection. mixing, holdup, and controlled release of containment out leakage. Ice condenser containments utilize an outer concrete building for shielding and an inner steel containment for leak tightnes J

...

Containment piping penetrationassemblies provide for the passage of process. service, sampling, and instrumentationpipelines into the containment vessel while maintainingcontainment integrity. The reactor building provides shielding and allows controlled release of the annulus atmosphere under accident conditions, as well as environmental missile protectionlor the containment vessel and Nuclear Steam Supply Syste The inner steel containment and its penetrationsestablish the leakage limiting boundary of the Containment. Maintainingthe containment OPERABLE limits the leakage of fission product radioactivityfrom the containment to the environment. S R 3.6.1.1 leakage rate requirements comply with 10 CFR 50, Appendix J, Option B (Ref. 1). as modified by approved as above exemption I The isolation devices for the penetrations in the containment boundary are a pari of the containment leak tight barrier. To maintain this leak light barrier: All penetrations required to be closed during accident conditions are e i t h e r '

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McGuire Units 1 and 2 8 3 6 1-1 Revision No. 32

mnrainmenr B 3. BASES BACKGROUND (continued) capable of being closed by an OPERABLE automatic containment isolation system, or closed by manual valves. blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LCO 3.6.3, Containment IsolationValves; Each air lock is OPERABLE, except as provided in LCO 3.6.2, Containment Air Locks; All equipment hatches are closed and sealed; and The sealing mechanismassociated with a penetration(e.g., welds, bellows, or O-rings) is OPERABL APPLICABLE The safety design basis for the containment is that the containment must SAFETY ANALYSES withstand the pressures and temperatures of the limiting Design Basis Accident (DBA) without exceeding the design leakage rate The DBAs that result in a challenge to containment OPERABILITYfrom high pressures and temperatures are a loss of coolant accident (LOCA)

-.- J and a steam line break (Ref. 2). In addition, releaseof significant fission product radioactivity within containment can occur from a LOCA. In the DBA analyses, it is assumed that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate of 0.3% of containment air weight per day (Ref. 3). This leakage rate, used in the evaluation of offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J, Option B (Ref. I ) , as La: the I maximum allowable containment leakage rate at the calculatedpeak containment internal pressure (Pa)resulting from the limiting design basis LOCA. The allowable leakage rate represented by L, forms the basis lor the acceptance criteria imposed on all containment leakage rate testing. La is assumed to be 0.3% per day in the safety analysis at Pa= 14.8 psig (Ref. 3). Satisfactory leakage rate test results are a requirement for the establishment of containment OPERABILIT The containment satisfies Criterion 3 of 10 CFR 50.36 (Ref. 4).

.- ~

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McGutrc Units 1 arid 2 B 3.6 1-2 Revision No. 32

umrainment B 3.6.1 LCO Containment OPERABILITY is maintained by limiting leakage to 5 1.0 L excepf prior to the first startup after performinga required Containment Leakage Rate Testing Program leakage test. At this time, the applicable leakage limits must be me I Compliance with this LCO will ensure a containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysi Individualleakage rates specified for the containment air lock (LCO 3.6.2), purge valves with resilient seals, and reactor building bypass leakage (LCO 3.6.3) are not specifically part of the acceptance criteria of 10 CFR 50,Appendix J. Therefore, leakage rates exceeding these individual limits only result in the containment being inoperable when the leakage results in exceeding the overall acceptance aiteria of 1.0 L.

APPLICABILITY In MODES 1.2.3, and 4. a DBA could cause a release of radioactive material into containment. In MODES 5 and 6. the probabilityand consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, Containment is not requiredto be OPERABLE in MODE 5 to prevent leakage of radoactive materialfrom containment. The requirements for containment during MODE 6 are addressed in LCO 3.9.4, 'Containment Penetrations.'

ACTIONS -

A. 1 In the event containment is inoperable, containment must be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining containment OPERABLE during MODES 1, 2, 3, and 4. This time period also ensures that the probability of an accident (requiring containment OPERABILIPI) occurring during periods when containment is inoperable is minima B . l and If containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within

Containment B 3.6.1 ACTIONS (continued)

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditionsin an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.1.1 REQUIREMENTS Maintainingthe containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program. Failureto meet specific I

leakage limits for the air lock, secondary containment bypass leakage path, and purge valve with resilient seals (as specified in LCO 3. and LCQ 3.6.3) does not invalidate the acceptability of the overall contairifiient leakage determinations unless the specific leakage contribution to overall Type A, B, and C leakage causes one of these overall leakage limits to be exceeded. As left leakage prior to the first startup after performing a required Containment Leakage Rate Testing Program leakage test is requiredto be < 0.6 L,for combined Type B and C leakage, and 5 0.75 L. for Option B for overall Type A leakag At all other times between required leakage rate tests, the acceptance I

criteria is based on an overall Type A leakage limit of 5 1.O La. At 2 Lathe offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by the Containment Leakage Rate Testing Program. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysi The Surveillance is modified by three Note I Note 1 requires that the space between each duel-ply bellows assembly on containment penetrationsbetween the containment building and the annulus be vented to the annulus during each Type A tes Note 2 requires that following each Type A test, the space between each dual-ply bellows assembly be subjected to a low pressure leak test with no detectable leakage. Otherwise. the assembly must be tested with the containment side of the bellows assembly pressurizedto Pa and meet the requirements of SR 3.6.3.8 (bypass leakage requirements).

Note 3 allows penetrationsM372 and M373 to be tested without draining the glycol-water mixture from the associated diaphragm valves (NF-288A. NF-233Eand NF-234A)as long as not leakage is indicated. This test may be used in lieu of 10 CFR 50. Appendix J. Option B as defined

Containment B 3.6.1 BASES SURVEILLANCEREQUIREMENTS (continued)

in ANSVANS 56.8-1994 Section 3.3.5 (Test Medium). The requiredtest pressure and interval are not change I All test leakage rates shall be calculated using observed data converted to absolute values. Error analysis shall also be performed to select a balanced integrated leakage measurement system.

REFERENCES CFR 50, Appendix J. Option . UFSAR, Chapter 1 . UFSAR, Section . 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

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McGulre Units 1 and 2 8 3 6 1-5 Revision No. 32

DUKE POWER MCGUIRE OPERATIONS TRAINING CLASSROOM TIME (Hours)

I NLO I NLOR I I

LPRO I I

LPSO I I

LOR OBJECTIVES OBJECTIVE Ixplain the following terms, as they apply to Technical Specifications:

Limiting Condition of Operation (LCO)

Applicability 0 Actions 0 Condition 0 Required Actions 0 Completion Time 0 Surveillance Requirements Table ADMTSOOI 3iven a Technical Specification, apply the rules of Section o determine the appropriate action@).

ADMTS002 f given a copy of Technical Specifications be able to jetermine the meaning of any of the various terms defined herein, or be able to recall from memory, the definition(s) of he following terms:

COLR (Core Operating Limits Report)

Leakage (including Identified, Unidentified, and Pressure Boundary Leakage)

OPERABLE -OPERABILITY ADMTS003 OP-MC-ADM-TS FOR TRAINING PURPOSES ONLY REV. I 5 Page 5 of 73

DUKE POWER MCGUIRE OPERATIONS TRAINING Objective # 2 I d. The phrase unless otherwise stated may require the completion of specific Required Actions even after the affected equipment has been restored to meet the LC Example:

Specification 3.4.3, RCS Pressure and Temperature (PiT) Limit In this LCO, Condition A contains a NOTE which states that Required Action A.2 shall be completed when this Condition is entered. Therefore, Required Action A.2 must be performed even if compliance with the LCO has been restore e. ACTIONS and Completion Times are applicable any time an LCO cannot be met and includes instances when ACTIONS are entered intentionally, such as for surveillance requirement performance. While considered acceptable, the intentional entry into ACTIONS is NOT intended to be used for operational convenience when alternate means are availabl The imposition of these requirements will limit the time equipment or parameters can be out of servic f. When a MODE change or specified condition is required to satisfy a Required Action, Completion Times for newly applicable LCOs apply from the time the Specification becomes applicabl I Objective # 2 I 3.9.3 LCO 3. Whenever any of the following conditions exist, a reduction in MODE is necessary to place the plant in a known condition (within the parameters utilized in our safety analyses):

_ A n LCO is not met and a plant condition exists that is not addressed by-the associated Conditions (Le., an associated ACTION is not provided)

An LCO is not met and the associated Required Actions are not me Or, if specifically directed by the Required Actions to place the Unit in a MODE or condition where the LCO is no longer applicabl LCO 3.0.3 is oniv applicable in MODES 1.2.3. and 4. LCO 3.0.3 is not applicable in MODE 5 because:

MODE 5 is already the lowest operating MODE, and a change from MODE 5 to MODE 6 may not always be the most conservative action to tak LCO 3.0.3 is sometimes referred to as the Motherhood Statement. However, use of LCO 3.0.3 in lieu of the specified CONDITIONS and their associated REQUIRED ACTIONS I COMPLETION TIMES is not allowed. LCO 3.0.3 is intended to provide guidance when no other guidance exist OP-MC-ADM-TS FOR TRAINING PURPOSES ONLY REV. 75 Page 29 of 73

DUKE POWER MCGUIRE OPERATIONS TRAINING

-. I Objective # 2 1 When in an LCO 3.0.3 Condition, the plant may be operating outside the boundary of the Safety Analysis and must be placed in a MODE where the subject equipment or parameter is not critical to plant safety. The times specified to reach each lower MODE are based on time of discovery and allow for a controlled orderly shutdown, within the maximum cooldown rate limitations, and include a 1-hour preparation period for Unit Shutdow If compliance with the original LCO is restored, completion of the LCO 3. Requirements is not require Exceptions to LCO 3.0.3 requirements are stated in the individual Specification Usually, where a reduction in MODE may not be conservative or, would not provide proper remedial measure Examples:

Specification 3.7.13, Spent Fuel Pool Water Level - If the Spent Fuel Pool Water Level is e23 feet over the top of irradiated fuel assemblies, no safety benefit is gained by placing the unit in a shutdown condition. However, immediately suspending movement of irradiated fuel assemblies in the Spent Fuel Pool does provide an added safety benefi Specification 3.7.5, Auxiliary Feedwater System - If all three Auxiliary Feedwater System Trains are INOPERABLE in MODE 1,2, or 3, no safety benefit is gained by placing the Unit in a shutdown condition. Instead, restoration of one Auxiliary Feedwater System Train to OPERABLE status is required prior to initiating a Unit Shutdown.

.I The following is a purpose of a Licensing Research Memorandum to Technical Specification (LCO) 3.0.3. Specifically, should the unit begin an actual reactor power reduction, Le., add negative reactivity, no later than one hour after entering Tech Spec LCO 3.0.3. LCO 3.0.3 Bases states in part Upon entering LCO 3.0.3,l hour is allowed to prepare for an orderly shutdown before initiating a change in unit operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid ...

However, the technical specifications do not specify how allowed outage times (AOTs) or shutdown times are to be used; that is, when or how specific actions may be taken within those periods. Circumstances may arise when plant safety is better served by delaying a shutdown action to provide a safer configuration for a shutdown transient or to avoid an unnecessary shutdown transien licensee resDonsiblv concludes that plant shutdown should be delaved or corrective actions can be accomplished so that an unnecessarv plant transient can be avoided, we believe that such a decision is permitted as long as the shutdown times specified by the Tech Spec are observed, including the default 3.0.3 provision...

_-

OP-MC-ADM-TS FOR TRAINING PURPOSES ONLY REV. 15 Page 31 of 73

1 Pt Which one of the following statements complies with the requirements of OMP 4-3 (RAP lmplementation Guidelines) regardingthe rules of usage for abnormal procedures (APs) when the emergency procedures (EPs)

have been implemented?

k APs may not be implementedwhen EPs have been entere Only one AP at a time may be implementedwhen EPs have been implemented. Concurrent implementation of APs when EPs are in use is not allowe APs may be implemented concurrently with EPs. However, the APs were written assuming that SI has not actuated and operators must be careful when using APs if SI has occurre APs may be implemented concurrently with EPs with the exception of events where SI has actuated. APs were written assuming the SI had not occurred and cannot be used if SI has actuated.

Ques-338.l.doc

Bank Question: 33 Answer: C 1 Pt Which one of the following statements complies with the requirements of OMP 4-3 ( f / A f lmplemenfafion Guidelines) regarding the rules of usage for abnormal procedures (APs) when the emergency procedures (EPs)

have been implemented? APs may not be implemented when EPs have been entere Only one AP at a time may be implementedwhen EPs have been implemented. Concurrent implementation of APs when EPs are in use is not allowe APs may be implementedconcurrently with EPs. However, the APs were written assuming that SI has not actuated and operators must be careful when using APs if SI has occurre APs may be implemented concurrently with EPs with the exception of events where SI has actuated. APs were written assuming the SI had not occurred and cannot be used if SI has actuate l _ l _ - _ l l _ - - __

_ _ - - - l _ l _ --I-Distracter Analysis: Incorrect: APs may be entered after EOPs have been started Plausible: Many plants have this provision - symptomatic EOPs should address all significant safety challenges without requiring APS Incorrect:No limitation on the number of APs Plausible: Makes sense to limit the number of concurrent procedures in use

- Correct answer Incorrect: No explicit prohibition against use of APs when SI has actuated BUT there is a caution and the APs were written for the situation where SI has NOT occurre Plausible: APs were written for the situation where SI has NOT occurre LEVEL: SRO SOURCE: BANK LEVEL OF KNOWLEDGE: Memory KA: G2.4.5 (2.913.6)

Ques-338. Ldoc

REFERENCE: OMP 4-3pages 21 & 22 Ques-338. Ldoc Emergency Procedures /Plan 2.4.1 Knowledge of EOP entry conditions and immediate action step (CFR: 41.10143.5 145.13)

IMPORTANCE RO SRO 4.6 2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry condition (CFR: 41.7I 45.7 I 45.8)

Note: The issue of setpoints and automatic safety features is not specifically covered in the systems sections).

IMPORTANCE RO SRO 4.1 2.4.3 Ability to identify post-accident instrumentatio (CFR: 41.6145.4)

IMPORTANCE RO SRO 3.8 2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedure (CFR 41.10/ 43.2145.6)

IMPORTANCE RO SRO 4.3 2.4.6 Knowledge symptom based EOP mitigation strategie (CFR: 41.10I43.5/ 45.13)

IMPORTANCE RO SRO 4.0 2.4.7 Knowledge of event based EOP mitigation strategie (CFR: 41.10143.5 145.13)

IMPORTANCE RO SRO 3.8 2.4.8 Knowledge of how the event-based emergencylabnormal operating procedures are used in conjunction with the symptom-based EOP (CFR: 41.10 143.5 / 45.13)

IMPORTANCE RO SRO NUREG-1122, Rev. 2 1

OMP4-3 Page 21 of 28 discontinue use of present procedure and stay in the referenced procedure. The referenced procedure is always entered at the first step unless otherwise specifie REFER TO. PER user is directed to a supplemental procedure/enclosure for actions but will remain in the controlling procedur Stable - Maintained steady. E a parameter is being controlled within a desired range, or if a slight trend in either direction is occurring, operator judgment may be used to determine if parameter is considered stabl Evaluate - Appraise the situation. Includes taking action based on evaluatio .17 Toleranaes -

Ranges or tolerances are provided if it is important to maintain a parameter within a given band. WHEN a range or tolerance (e.g.: 515%) is provided, it is understood to mean extra attention should be paid to maintain the parameter within this rang WHEN a single value is given, it is assumed the value is an ideal value. WHEN an ideal value (e.g.: at no load or 350 psig) isprovided, it is understood to mean attention

,

.. ... should be paid to maintain the parameter at the ideal value but NOT be overly concerned

. .... if the exact value is NOT achieve .18 Multiple Use of EPs and AP The Control Room SRO will determine how many procedures can be implemented at a time and their priority based on manpower availability and the particular event in progress. More than one EP shall NOT be run concurrently unless directed by the procedure. Generally the use of APs in conjunction with EPs should be avoided. In some instances it would be proper to use an AP concurrently during a major accident which is being addressed by the EPs. An example of this is upon loss of all Nuclear Service Water in the middle of an accident, the operators would need to utilize the AP for Loss of RN also. E an AP is used during an SI event, USE CAUTION. APs are generally written assuming an SI has NOT occurred (exception - AP/35, ECCS Actuation During Plant Shutdown). Evaluate any AP steps in post SI events to ensure the steps do NOT conflict with any EP in effect. NOT all AP actions would be appropriate if an SI occurred. (Enclosures in EP/G-I (Generic Enclosures) may be used when reference by EPs or APs).

OMP4-3 Page 22 of 28 Use of most APs that have foldouts will likely be terminated when a reactor trip or SI occurs. There are a few APs with foldouts that could potentially be implemented concurrently with an EP though (Loss of VI or Loss of KC for example). Rules of foldout page use as specified in Section 7.13 should be applied in this situation als Although unlikely, it is possible that the crew may have one EP foldout in effect at the same time as one of the AP foldouts. Implementation and priority of the AP foldouts will be evaluated as discussed in paragraph abov ROs may be given procedure responsibilities when AFs and EPs or multiple APs are in effect at the same tim .19 The STNOperations Shift Manager Interface The STA monitors the Critical Safety Functions (CSF) and otherwise ensures Core Safety through monitoring of activities and parameters. E any one CSF is other than green, the STA will check whether the CSF non-green status isvalid or being caused by an invalid inpu the non-green CSF is invalid, the STA will notify the operating .:.

..., .>! crew of the invalidity. For Red or Orange path procedures, the STA will immediately notify the operating $rew that th% conditioqexists and give the assocja!$d functional restoration procedure to the crew to imp7ement as the controlling procedure. For Yellow path procedures, the.STA will pull the functional restoration procedure and evaluate

. . whether to implement the procedure, with the Operations Shift Managers concurrence,

as time allows. This evaluation should consider whether the optimal recovery procedure is properly addressingthe current plant conditions in as timely a manner as the

.~ . t. . functional restoration procedur ..

Once status tree monitoring is initiated, the STA should monitor status trees continuously if an orange or red condition is found to exist. E . n o condition more serious than yellow is found, monitoring frequency may be reduced to 10-20 minutes, unless some significant change in plant status occurs. Status tree monitoring may be performed using OAC SPDS display or EP/1(2)/N5000/F-O (Critical Safety Function Status Trees). E the OAC SPDS display is being used, the STA will validate the OAC SPDS status every 10-20 minutes using control board indication .~

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IF the STA is NOT available, the Operations Shift Manager shall assume the STA responsibilities or delegate the STA responsibilities to another licensed operato .20 Control Room Team Responsibilities During the Use of EP/APs 7.2 Operations Shift Manager - Responsibilities 7.20. Assume role of Emergency Coordinator upon activation of the Emergency Plan until properly relieved by the Station Manage P Unit 2 is operating at 100% power. 21\11-59 (Cold Leg Accumulator Check Valve) begins to leak at 0200. Given the following accumulator indications:

Time -

0200 -

0300 -

0400 -0500 Level (%) 21% 31% 41% 51%

Pressure (psig) 586 613 640 667 Boron (ppm) 2485 2470 2455 2440 When does the accumulator first exceed a limiting condition for operation?

REFERENCES PROVIDED:

k 0200 Ques-697.3.doc

Bank Question: 69 Answer: B 1 P Unit 2 is operating at 100% power. 21\11-59 (ColdLeg Accumulator Check Valve) begins to leak at 0200. Given the following accumulator indications:

Time -

0200 -

0300 0400 -

0500 Level (%) 21 % 31% 41% 51%

Pressure (psig)

- 586 613 640 667 Boron (ppm) 2485 2470 2455 2440 When does the accumulator first exceed a limiting condition for operation?

REFERENCES PROVIDED:

Tech Spec 3. Unit 1 Data Book curve Unit 1 Cycle 16 COLR - page 24 ._-____-.__.-.__.-------.--.-.--.-----.----

Distracter Analysis: Tech Spec values for CLA parameters are:

Volume 2 6870 (12.3%) but I7342 gal (38.7%) -exceeded at 0400 (high)

Pressure 2 585 but _< 639 psig - exceeded at 0400 Boron concentration t 2475 ppm but 52875 ppm - exceeded at 0300 Incorrect Answer: nothing out of spe Correct: first exceeds Boron concentration (<2475 ppm) at 0300 Plausible: Incorrect:the LCO is first exceeded at 0300 on low boron concentratio Plausible: Incorrect: the LCO is first exceeded at 0300 on low boron concentration Plausible:

Level: SRO KA: G2.1.25 (2N3.1)

Lesson Plan Objective: ECC-CLA Obj. 7

Source: Bank Level of knowledge: comprehension References:

1. OP-MC-ECC-CLA page 23 2. Tech Spec 3.5.1 -PROVIDED 3. Unit 1 Data Book Curve 7.4 - PROVIDED 4. Unit 1 Cycle 15 COLR - PROVIDED

- -/- Conduct of Operations (continued)

2.1.18 Ability to make accurate, clear and concise logs, records,status boards, and report (CFR: 45.12 I 45.13)

IMPORTANCE RO SRO .1.19 Ability to use plant computer to obtain and evaluate parametric information on system or component statu (CFR 45.12)

IMPORTANCE RO SRO .1.20 Ability to execute procedure step (CFR: 41.10143.5 145.12)

IMPORTANCE RO SRO .1.21 Ability to obtain and verify controlled procedure cop (CFR 45.10 145.13)

IMPORTANCE RO SRO .1.22 Ability to determine Mode of Operatio (CFR.43.5 i 45.13)

IMPORTANCE RO SRO .1.23 Ability to perform specific system and integrated plant procedurecJ during all modes of plant operatio (CFR: 45.2 145.6)

IMPORTANCE RO SRO to obtain and interpret station electrical and mechanical drawing .1.24 Ability (CFR 45.12 145.13)

IMPORTANCE RO SRO .1.26 Knowledge of non-nuclear safety procedures (e.g. rotating equipment, ektrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

(CFR: 41.10 I45.12)

IMPORTANCE RO SRO NUREG-1122.Rev. 2 1

DUKE POWER MCGUIRE OPERATIONS TRAlNlNG OBJECTIVES

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N L L L

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L P P O OBJECTIVE O R S R R O 0 7 Concerning the Technical Specifications related to the CIA System:

Given the LCO title, state the LCO (including any COLR X x x values) and applicabilit For any LCO's that have action required within one hour, X x x state the actio Given a set of parameter values or system conditions, X x x determine if any Tech Spec LCO's is (are) not met and any action(s) required within one hou Discuss the suweillance requirement for tank level change X x x for unknown reasons.

.c Given a set of plant parameters or system conditions and X x x

/--

the appropriate Tech Specs, determine the required actio Discuss the bases for a given Tech Spec LCO or Safety X

Limi * SRO ONLY

- I OP-MC-ECC-CLA FOR TRAINING PURPOSES ONLY REV. 18 Page 7 of 35

DUKE POWER MCGUIRE OPERATIONS TRAINING 3.2.2. CLA "A" and " B nitrogen backup to NC PORV's c

Normally all three PORV's are operated by air supplied from VI, however under certain circumstances the nitrogen overpressure from CLA " A and "B" will provide the motive force for all three PORV's, Anytime the low pressure mode is selected on the NC-32B key switch and WR Loop "C" Tmldis < 320 O F , NC431B will open allowing nitrogen from CLA "6"to provide the motive force for NC-32B and NC36B if VI pressure fails. Note that only the NC-32B setpoint is reduced to I 4 0 0 psig and operates off of WR Loop "C" pressur Anytime the low pressure mode is selected on the NC-34A key switch and WR Loop

"D"That is less than 320 O F , NI-430A will open allowing nitrogen from CLA "A"to provide the motive force for NC-34A if VI, pressure fails (setpoint I 400 psig from WR Loop 'D" pressure). NI-430A and Nl-4318 can also be opened at anytime with the control room manual operato NOTE: Minimum CIA Accumulator to maintain LTOP operability is 200 psl .0 TECHNICAL SPECIFICATIONS Objective #7 Tech Spec 3.5.1 ECCS Accumulators P Allowed outage times are variable based on boron concentration to ensure that the reactor is shut down following a LOCA and that any problems are corrected in a timely manner. The minimum required to ensure post-LOCA subcriticality is based on nominal accumulator volume conditions and allows additional time since subcriticality is assured when the boron concentration is above this valu The lower limit for 3.5.1 .c is based on worst case liquid mass, boron concentration and measurement error The accumulator isolation valves are considered to be operating bypasse Bypasses of a protective function are required to be automatically removed whenever permissive conditions are not met. In addition, as these 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 minimizes the time exposure of the plant to a LOCA event occurring concurrent with the failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be opened immediately, 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 require .2 Tech Spec 3.4.14, RCS Pressure Isolation Value (PIV) Leakage OP-MC-ECC-CLA FOR TRAINING PURPOSES ONLY REV. 18 Page 23 of 35

Accumulators 3. .

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

3.5.1 Accumulators LCO 3. Four ECCS accumulators shall be OPERABL APPLICABILITY: MODES 1 and 2, MODE 3 with RCS pressure 1000 psi ACTIONS i CONDITION REQUIRED ACTION COMPLETION TIME One accumulator Restore boron 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable due to boron concentrationto within concentrationnot within limit limit I ,I One accumulator Restore accumulator to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months e- inoperable for reasons OPERABLE statu other than Condition A.

..

.. Required Action and associatedCompletion Be in MODE ' 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Time of Condition A or B AND not me Reduce RCS pressure to 1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 5 lo00 psi Twoormore Enter LCO 3. Immediately accumulators inoperabl I

2 McGuire Units 1 and 2 3.5.1-1 Amendment Nos. 184/166

Accumulators 3.5.1 A SURVEILLANCE REQUIREMENTS - -

d SURVEILLANCE FREQUENCY SR 3.5. Verify each accumulator isolation valve is fully ope hours SR 3.5.1.2 Verify borated water vofume in each accumulator is 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s-5 6870 gallons and 5 7342 gallon SR 3.5. Verify nitrogen cover pressure in each accumulator is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

-> 585 psig and 639 psi SR 3.5.1.4 Verify boron concentration in each accumulator is within 31 days the limits specified in the COL AND

.---NOTE-------

Only required to be performedfor affected ,.- .

accumulators

...

-__..----..---- 4-

....

Once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of 2 1%

of tank volume that is not the result of addition from the refueling water storage tank SR 3.5. Verify power is removed from each accumulator isolation 31 days valve operator when RCS pressure is > 1OOO psi . -

McGuire Units t and 2 3.5.1 -2 Amendment Nos. 1W166

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Page 24 of 29 Revision 24 McCuirc 1 Cyclc 16 Core Operatin- Limit Report

-?

/

2.10 Accumulators (TS 3.5.1)

2.10.1 Boron concentration limits during modes 1 and 2, and mode 3 with RCS pressure

>I000 psi:

Parameter Limit Cold Leg Accumulator minimum boron concentratio ,475 ppm Cold Leg Accumulator maximum boron concentratio ,875 ppm 2.1 1 Refueling Water Storage Tank - RWST (TS 3.5.4)

2.1 1.1 Boron concentration limits during modes I , 2. 3. and 4:

Parameter Limit Refueling Water Storage Tank minimuni boron 2.675 ppm concentratio Refueling Watcr Storage Tank niaxiniuni boron 2.875 ppni concentratio i 1 Pt. Unit 1 was operating at 100%when a large break LOCA with loss of offsite power occurs. One diesel generator fails to start. The operators are entering E-I (Lossof Reactor or Secondaly Coolant).

Given the following critical safety function status indications:

Core Cooling - RED Subcriticality - GREEN Containment - RED Inventory - GREEN D Heat Sink - RED Integrity- RED Which one of the following describes the highest priority problem, and the appropriate operator action?

A Integrity; Transition to FR-P.l, (Response to Imminent Pressurized Thermal Shock). Core cooling; Transition to FRG.1, (Response to Inadequate Core Cooling). Heat Sink; Transition to FR-H.l, (Response to Loss of Secondary Heaf Sink). Containment;Transition to FR-2.1, (Response to High Containment Pressure).

Bank Question: 776. I Answer: B 1 P Unit 1 was operating at 100% when a large break LOCA with loss of offsite power occurs. One diesel generator fails to start. The operators are entering E-I (Loss of Reactor or Secondary Coolant).

Given the following critical safety function status indications:

Core Cooling - RED Subcriticality - GREEN Containment - RED Inventory - GREEN Heat Sink - RED Integrity - RED Which one of the following describes the highest priority problem, and the appropriate operator action? Integrity; Transition to FR-P.l, (Response to lmminenf Pressurized ThermalShock). Core cooling; Transition to FR-C.l, (Response io Inadequate Core Cooling). Heat Sink; Transition to FR-H.l, (Response to Loss of Secondary Heat Sink). Containment; Transition to FR-2.1, (Response to High Containment Pressure).

Distracter Analysis:

- Incorrect: Core Cooling is the highest priority RED Plausible: Correct:

Plausible: Incorrect:core cooling is the highest priority RED Incorrect:core cooling is the highest priority RED Plausible:.

Level: SRO KA: WE 07 EA 2.1 (3.2/4.0)

Lesson Plan Objective:EP-FO SEQ 2 , 3 Ques-776.l.doc

Source: Bank Level of knowledge: Memory References:

1.OP-MC-EP-FOpages13,15 2. OMP 4-3 pages 15-16

-3 EPE: 0 7 Saturated Core Cooling (Continued)

WA N KNOWLEDGE

- - operating EK3.2 Normal. abnormal and emergency - -. procedures

- associated with (Saturated Core Cooling).

IMPORTANCE RO SRO EK3.3 Manipulation of controls required to obtain desired operating results during abnormal, and emergency situation IMPORTANCE RO SRO EK3.4 RO or SRO function within the control room team as appropriate to the assigned I position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violate IMPORTANCE RO SRO ABILITY EA1. Ability to operate and I or monitor the following as they apply to the (Saturated Core Cooling)

(CFR: 41.7 145.5 145.6)

EA1.l Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual feature IMPORTANCE RO SRO EA1.2 Operating behavior characteristics of the facilit IMPORTANCE RO SRO EA1.3 Desired operating results during abnormal and emergency situation IMPORTANCE RO SRO EA2. A$ilitv to determine and interpret the following as they apply to the (Saturated Core Cooling)

(CFR 43.5 / 45.13)

EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility's license and amendment IMPORTANCE RO SRO .5-19 NUREG-1122, Rev. 2 I

DUKE POWER

-

MCGUIRE OPERATIONS TRAINING CLASSROOM TIME (Hours)

I NLO I NLOR 1 LPRO 1 LPSO I I

LOR OBJECTIVES N

L OBJECTIVE 0

' .. X X

OP-MC-EP-FO FOR TRAINING PURPOSES ONLY REV. 06 Page 5 of 83

DUKE POWER MCGUIRE OPERATIONS TRAINING 7 PROCEDURE SERIES BACKGROUND (continued)

Once the Status Trees are being monitored, the following rules of usage apply: The Status Trees should be continuously monitored in order of Critical Safety Function priority. CSF procedures are not to be implemented prior to transition from E-0, Reactor Trip or Safety Injection. If a CSF path is red or orange while the operating crew is in E-0, but has turned to green upon transition from E-0, the CSF procedure, which was in alarm, shall not be implemented. If the CSF path is yellow, it shall be handled as any other yellow path procedure. If there are any valid red or orange path CSFs on transition from E-0 (unless the transition is to ECA-0 (Loss of All AC Power), the associated CSF procedure shall be implemented. If a valid red or orange path flickers into alarm on SPDS but returns to green prior to the crew validating the condition and implementing the procedure (implementation of procedure being that the SRO either hands out fold-out pages or starts reading from the procedure), the CSF procedure shall not be implemented. If the CSF path is yellow, it shall be handled as any other yellow path procedure. Likewise, if a valid red path or orange path goes into alarm during performance of a higher priority CSF procedure, but returns to green prior to transition from the higher priority CSF path procedure to the lower priority CSF procedure, the associated CSF procedure shall not be implemented. If the CSF path is yellow, it shall be handled as any other yellow path procedure. If a CSF procedure directs the operator to return to the procedure and step in effect, AND the corresponding status tree continues to display the off-normal conditions, THEN the corresponding CSF procedure doesnt have to be implemented again, since all recovely actions have been completed. However, if the same status tree subsequently changes to a valid higher priority condition, (OR if it changes to lower condition and returns to higher priority condition again),

THEN the corresponding CSF procedure shall be implemented as required by its priority. Once status tree monitoring is initiated, the STA should monitor status tree continuously if an orange or red path condition exists. If no condition more serious than yellow is found, monitoring frequency may be reduced to 10 - 20 minutes unless some significant change in plant status occurs. Status tree monitoring may be performed using the OAC SPDS display or F-0 (Critical Safety Function Status Trees). If the OAC SPDS display is being used, the STA will validate the OAC SPDS status every 10 - 20 minutes using control board indications. If the STA is not available, the OSM shall assume the STA responsibilities or delegate the STA responsibilities to another licensed operato OP-MC-EP-FO FOR TRAINING PURPOSES ONLY REV. 06 Page 13 of 83

DUKE POWER MCGUIRE OPERATIONS TRAINING Red Path

, If any valid red path is encountered during monitoring, the operator is required to immediately implement the corresponding EP. Any recovety EP previously in progress shall be discontinued. If during the performance of any red path procedure, a valid red condition of higher priority arises, the higher priority condition should be addressed first, and the lower priority red path procedure suspende . Orange Path If any valid orange path is encountered, the operator is expected to scan all of the remaining trees, and then, if no valid red is encountered, promptly implement the corresponding EP. If during the performance of an orange path procedure, any valid red condition or higher priority valid orange condition arises, then the red or higher priority orange condition is to be addressed first, and the original orange path procedure suspende Once a procedure is entered due to a red or orange condition, that procedure should be performed to complgtion, unless preempted by some higher priority condition. It is expected that the actions in the procedure will clear the red or orange condition before all the operator actions are complete. However, these procedures should be performed to the point of the defined transition to a specific procedure or to the procedure and step in effect to ensure the condition remains clear. At this point any lower priority red or orange paths currently indicating or previously started but not completed shall be addressed.

i OP-MGEP-FO FOR TRAINING PURPOSES ONLY REV. 06 Page 15 of 83

OMP4-3 Page 15 of 28

, The configuration control cards filled out in Steps 7.14.1 and 7.14.2 shall be 7.1 handled per the following two situations:

Without operational support center activation The configurationcontrol card will be handled by Ops shift per OMP 7-1 (Removal and Restoration (R&R) Requirements).

With operational support center activation WHEN the OSC is activated, Operations will report to the OSC and shall bring with them all configuration control cards that have been filled ou The cards taken to the OSC shall be given to the OPS SRO in the OS For handling cards in the OSC, refer to RP/O/A/5700/020 (Activation of the Operations Support Center).

7.15 Usage of Status Trees There are six different trees, each one evaluating a separate Critical Safety Function of the plant. Color-coding of the status tree end points will be either red, orange, yellow, or

. green, witti green representing a satisfied.safety status. Each nongreen color represents an action level that should be addressed according to the Rules of Priority as discussed belo *-IThe six Status Trees are always evaluated in the sequence:

Subcriticality Core Cooling HeatSink Integrity Containment Inventory IF identical color priorities are found on different trees during monitoring, the required

.--

action priority is determined by this sequenc Initial monitoring of the status trees should begin on either of the following conditions:

As directed by an action step in EP/l,UA/5000/E-O (Reactor Trip or Safety Injection).

WHEN a transfer is made out of the Safety Injection procedure to another E OMP4-3 Page 16 of 28 An exception to this is that no status tree monitoring is required during the Loss of All AC Power EP since none of the electrically powered safeguards equipment can be used. WHEN power is subsequently restored, EP/1,2/A/5000ECA-O.l or 0.2 (Loss of All AC Power Recovery procedures) will direct the operator when monitoring of status trees is required.

7.15.1 . Implementing CSF Path Procedures 7.15. CSF procedures are NOT to be implemented prior to transition from EP/1,2/A/5000/E-O (Reactor Trip or Safety Injection). a CSF path is red or orange while the operating crew is in EP/I,YA/5000/E-O, but has turned to green upon transition from E-O, the CSF procedure which was in alarm shall NOT be implemented. E the CSF path is yellow, it shall be handled as any other yellow path procedure per Section 7.15. there are any valid red or orange path CSF's on transition from E-0 (unless transition is to EP/1,2/A/5000/ECA-O (Loss of All AC Power), the associated CSF procedure shall be implemente .15. IF a valid red or orange path flickers into a l m on SPDS but returns to green prior to the crew validating the condition and implementing the procedure (implementation of procedure being that the SRO either hands out fold-out pages or s t m reading from the procedure), the CSF procedure shall NOT be Implemente the CSF path is yellow, it shall be handled as any other yellow path procedure pet Section 7.15.1.7. Likewise, if a valid red path or orange path goes into alarm during performance of a higher priority CSF procedure, but returns to green prior to transition from the higher priority CSF path procedure to the lower priority CSF procedure, the associated CSF procedure shall NOT be implemente the CSF path is yellow, it shall be handled as any other yellow path procedure per Section 7.15. .15. IF a CSF procedure directs the operator to return to the procedure and step in effect, AND the corresponding status tree continues to display the offnormal conditions, the corresponding CSF procedure doesn't have to be implemented again, since all recovery actions have been completed. However, if the same status tree subsequently changes to a valid higher priority condition, (OR if it changes to lower condition and returns to higher priority condition again), the corresponding CSF procedure shall be implemented as required by its priorit OMP 4-3 Page 17 of 28 7.15.1.4 Red Path

-

IF any valid red path is encountered during monitoring, the operator is required to immediately implement the corresponding EP. Any recovery EP previously in progress shall be discontinued. E during the performance of any red path procedure, a valid red condition of higher priority arises, the higher priority condition should be addressed first, and the lower priority red path procedure suspended.

7.15.1.5 Orange Path

-

IF any valid orange path is encountered, the operator is expected to scan all of the remaining trees, and then, if no valid red is encountered, promptly implement the corresponding E during the performance of an orange path procedure, any valid red condition or higher priority valid,orange condition arises, the red or higher priority orange condition is to be addressed first, and the original orange path procedure suspended.

7.15.1.6 Completion of red or orange path procedure Once procedure is entered due to a red or orange condition, that procedure should be performed to completion, unless preempted by some higher priority condition. It is expected that the actions in the procedure will clear the red or orange condition before all the operator actions are complete. However, these procedures should be performed to the point of the defined transition to a specific procedure or to the procedure and step in effect to ensure the condition remains clear. At this point any lower priority red or orange paths currently indicating or previously started but NOT completed shall be addresse FF-S.1,P. 1 and Z. 1 can be entered from either an orange or red path status. @the color changes from orange to red while you are in one of these EPs, the crew should continue and complete the EP from where they are. Crew does NOT have to backup and restart the EP. E you exit the orange path, and it subsequently turns red, the EP must be reentered at Step P Unit 1 is operating at 100% power when the OAC registers a low spent fuel pool level alarm. Given the following events and conditions:

The operators read -2.1 ft SFP level and stable on the main control boar The operating KF pump has trippe , An NLO reports a large leak in the auxiliary building has stoppe Normal SFP makeup is not availabl Which one of the following statements correctly describes the corrective action for this event? Implement AP/I/A/5500/41 (Loss of Spent Fuel Cooling or Level), find and isolate the leak on the KF discharge pipin Implement AP/I/A/5500/41 (Loss o f Spent Fuel Coolingof Level) Find and isolate the leak on the KF suction pipin Implement AP/l/A/5500/40 (Loss o f Refueling Canal Level), and initiate assured makeup due a leak on the,discharge pipin Implement AP/l/A/5500/40 (Loss of Refueling Canal Level), and initiate assured makeup due to a leak on the suction pipin ____________________-------

Bank Question: 89 Answer: A 1 Pt. Unit 1 is operating at 100% power when the OAC registers a low spent fuel pool level alarm. Given the following events and conditions:

The operators read -2.1 ft SFP level and stable on the main control boar The operating KF pump has trippe An NLO reports a large leak in the auxiliary building has stoppe Normal SFP makeup is not availabl Which one of the following statements correctly describes the corrective action for this event? Implement AP/I/A/5500/41 (Loss o f Spent Fuel Cooling or Level), find and isolate the leak on the KF discharge pipin Implement AP/I/A/5500/41 (Loss of Spent Fuel Cooling or Level) Find and isolate the leak on the KF suction pipin Implement AP/I/A/5500/40 (Loss of Refueling Canal Level), and initiate assured makeup due a leak on the discharge pipin Implement AP/I/A/5500/40 (Loss of Refueling Canal Level), and initiate assured makeup due to a leak on the suction pipin Distracter Analysis: Correct: Incorrect:The leak is on the discharge pipin Plausible: If the candidate confuses the piping immersion depth with the suction pipes, which are at 4 fee Incorrect:

Plausible:. Incorrect: Do not use the assured source, and the leak is on the discharge pipin Plausible:.

Level: SRO KA: SYS 033 A2.02(2.7/3.0)

Lesson Plan Objective: OP-MC-FH-KF Obj. 4/5/14 Source: BANK

Level of knowledge: Comprehension References:

1. OP-MC-FH-KF pages 19,23,49

.. .,. . . .

-

SYSTEM 033 Spent Fuel Pool Cooling System (SFPCS)

K5 Knowledge of the operational implication of the following concepts as they apply to the Spent Fuel Pool Cooling System:

(CFR: 41.5 145.7)

w.01 Pumptheor y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9 w.02 Heattransfer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9 W.03 . . .. .

DIP detector theory of OPS . . . . . . . . . . . . . . . . . . .. . . .. . .6 w.04 K-eff . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3*

K5.05 . . . . . . . ..

Decay heat . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . .3 W.06 Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .5 K6 Knowledge of the effect of a loss or malfunction on the following will have on the Spent Fuel Pool Cooling System:

(CFR: 41.7 145.7)

K6.01 Pumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9 K6.02 . . . . ..

Heat exchangers . . . . . . . . . . , . . . . . . . . . . . . . . . . . . .

Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

. .7 .7 K6.03 K6.04 Motors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7 K6.05 .. ..

Pressure and pressure detectors , , . . . . , . . . . . . . . . . .

Temperaturesensors ..................................

. .. .. . .8 .8 K6.06 K6.07 . . . . . . . ..

Filters and demineralizers . . . . . . . .. .

.. . . . .. . . . . . .8

,4Em A1 Ability to predict andlor monitor changes in panuneten (to prevent exceeding design Limits) associated with Spent Fuel Pool Cooling System operating the controls including:

(CFR: 41.5 145.5)

A1.01 . . .. ... . . . . . . .. .. .. . . . . .. .

Spent fuel pool water level . . . . . .3 A1.02 Radiation monitoring systems . . . . . . .. . . . .. . . . . . . . . . . . .. . . . .3 A1.03 SFPCS controls and sensors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7 A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System ;and @) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 / 43.5 145.3 / 45.13)

A3 Ability to monitor automatic operation of the Spent Fuel Pool Cooling System including:

(CFR: 41.7 145.5)

A3.01 Temperature control valves ..,.. ........ .... . ........ .... 2.5* 2.7*

A3.02 Spent fuel leak or rupture . . . .'. . . . . . . . . . . . . . . . . . . . . . . . . . . .1 NUREG-1122, Rev. 2 3.8-10

DUKE POWER MCGUIRE OPERATIONS TRAINING CLASSROOM TIME (Hours)

2 2 2 2 2 OBJECTIVE State the purpose of the Spent Fuel Pool Cooling Syste Draw a simplified diagram of the Spent Fuel Pool Cooling System (including all major components) per Training Drawing 7.1, Spent Fuel Pool Cooling System - Simplifie State the flowrates through each of the following flowpaths:

Spent Fuel Pool Cooling Loop Spent Fuel Pool Purification Loop SDent Fuel Pool Skimmer Loop List the sources of makeup to the Spent Fuel Pool Cooling System; including the source grade (Le., borated, non-borated demineralized, and non-borated lake water).

Explain the conditions which would require assured makeup, from the Nuclear Service Water System, to the Spent Fuel Pool Cooling Syste List the power supply for the following Spent Fuel Pool Cooling System Pumps (Unit 1 and Unit 2):

KF Pump(s)

KF Skimmer Pump(s)

Describe the controls, indications, and/or alarms, associated with Spent Fuel Pool Cooling System operation, located within the Control Roo Describe how the KF Pump motor(s) is cooled during system ooeratio ~~ ~ ~

State the cooling medium for the Spent Fuel Pool Cooling System Heat Exchanger(s).

Describe the controls, indications, and/or alarms, associated with Spent Fuel Pool Cooling System operation, located outside the Control Room.

OP-MC-FH-KF FOR TRAINING PURPOSES ONLY REV. 22 Page 5 of 85

DUKE POWER MCGUIRE OPERATIONS TRAINING

-

L P

OBJECTIVE S

-

Given a Limit and/or Precaution, associated with operation of X the Spent Fuel Pool Cooling System, discuss it's basis and applicabilit Describe the Spent Fuel Pool Cooling System response to a X Safety Injection Signal and/or a Blackout Signa Concerning AP/I(2)/A/5500/25, Spent Fuel Damage: X State the purpose of AP/5500/2 Recognize the symptoms that would require implementation of the A Concerning AP/l(2)/A/5500/41, Loss of Spent Fuel Pool X Cpoling or Level:

t '

State the purpose of AP/5500/4 Given a set of symptoms determine which case of AP/5500/41 should be implemente Concerning the Technical Specifications related to the Spent Fuel Pool Cooling System:

Given the LCO title, state the LCO (including any COLR X values) and applicabilit For any LCO's that have action required within one hour, X state the actio Given a set of parameter values or system conditions, X determine if any Tech. Spec. LCO's is (are) not met and any action(s) required within one hou Given a set of plant parameters or system conditions and the appropriate Tech. Specs, determine the required X action(s).

Discuss the bases for a given Tech. Spec. LCO or Safety X Limi * SRO ONLY

-

OP-MC-FH-KF FOR TRAINING PURPOSES ONLY RV. 22 Page 7 of 85

DUKE POWER MCGUIRE OPERATIONS TRAINING A table within OP/1(2)/N6200/05, Spent Fuel Cooling System provides a conservative addition for makeup to the Spent Fuel Pool using Demineralized Water based upon the last boron sample:

Borated water, from the Refueling Water Storage Tank (FWST), can be used for makeup and should be considered the preferred source of makeup unless the last SFP boron sample indicated >2775 pp Objective # 4 & 5 Non-borated lake water (Assured Makeup), from the Nuclear Service Water System (RN), can be used for makeup. The Assured Makeup should only be used if borated and demineralized water are not available for makeup and the Spent Fuel Pool Level is low enough to cause a radiation hazard to employees or the publi Electrical Power Supply I Objective # 6 Each SpentFuel Pool Cooling Pump receives power from its respective Essential Bus, l(2)ETA (4160V) or l(2)ETB (41 60V) (the same buses that can be powered by the Emergency Diesel Generators).

Each KF AHU receives power from its respective Essential Motor Control Center, 7(2) EMXA (SOOV) or l(2) EMXB (SOOV).

The Fuel Pool Skimmer Pump receives power from one of the normal station buses l(2) MXK(600V) and can be operated anytime the bus is energized. However, this pump is not required during emergency operation Motor-operated valve 1(2)KF-12, Purification Loop Isolation Valve, is also powered from a normal station bus l(2) MXJ (SOOV) and can be operated remotely anytime the bus is energized. Manual operation of this valve can be performed if power is unavailabl OP-MC-FH-KF FOR TRAINING PURPOSES ONLY RV. 22 Page 19 of 85

DUKE POWER MCGUIRE OPERATIONS TRAINING The Spent Fuel Pool stores fuel assemblies approximately 33 feet 4 inches below the fuel pool operating deck with approximately 25 feet of borated water above the top of each fuel assembly.

I Objective # 7 I Control Room Indication is provided for Spent Fuel Level and Temperature. (Refer to Training Drawing 7.3, Spent Fuel Pool Control Room Indication.) Spent Fuel Pool Cooling Pumps I Obiective # 7 i Two Spent Fuel Pool Cooling Pumps (KF Pumps) are provided for each Unit. The controls and indications, associated with Spent Fuel Pool Cooling Pump operation, located on the Main Control Board (MC-1l), consist of the following:

START / STOP Control Switch These momentary START / STOP pushbuttons allow the operator to START and STOP the pump, as desire During a Station Blackout the KF Pump(s) will initially lose power (load shed)

. '*

- but receive a manual start permissive when Load Group 9 is loaded onto the bu During a Safety Injection Signal, !h6 KF Pump(s) running prior to SI will

. .. . continue to run. The KF Pump(s) not running, prior to SI, will receive a manual start permissive when Load Group 9 is loaded onto'the bu Any KF Pump(s) running or manually started, while the SI Signal is present, cannot be stopped until the SI Signal is RESE ON / OFF (Red / Green) Indicating Lights These ON / O F F (Red / Green) indicating lights are mounted on the START /

STOP Control Switch and provide indication when the KF Pump breaker is CLOSED (ON) or OPEN (OFF).

Each pumpis designed for 2310 gpm flow and each takes suction from the Spent Fuel Pool, four feet below pool level, and discharge back into the Spent Fuel Pool, six feet above the fuel assemblies. Holes drilled into the Spent Fuel Pool Discharge Header act as a vacuum breaker and limit siphon draining to two feet below normal Spent Fuel Pool leve In addition, each KF Pump is designed to circulate water through the cooling and purification loop at the same time. Under normal operating conditions only one pump is utilized to supply flow through only one cooling loop and the purification loop. During two pump operation each pump will supply flow through a cooling loop but one pump will be selected to provide flow for the purification loop. One pump can also be aligned to both heat exchangers, with or without supplying flow to the purification loo Each pump has mechanical seals provided with leakoff, vent, and drain connection The internal wetted surfaces of these pumps are made of stainless stee OP-MC-FH-KF FOR TRAINING PURPOSES ONLY REV. 22 Page 23 of a5

DUKE POWER MCGUIRE OPERATIONS TRAINING Objective # 13 Abnormal Operating Procedure APll(2)/A/5500/25, Spent Fuel Damage, is provided to identify operator actions required during a spent fuel damage event. Actions are defined for spent fuel damage inside Containment or within the Spent Fuel Pool. This procedure has only a single Case and the Symptoms are:

EMF-36, Unit Vent High Gas Radiation Alarm (Process Monitor)

EMF-38, Containment High Particulate Radiation Alarm (Process Monitor)

EMF-39, Containment High Gas Radiation Alarm (Process Monitor)

EMF-40, Containment High Iodine Radiation Alarm (Process Monitor)

EMF-42, Fuel Handling High Gas Radiation Alarm (Process Monitor)

EMF-16, Containment Refueling Bridge Alarm (Area Monitor)

EMF-17, Spent Fuel Building Bridge Alarm (Area Monitor)

Gas bubbles originating from the damaged assembly(ies).

Visual evidence of damage with potential of radioactive release(s).

Subsequent operator action(s) will first determine the damaged fuel location. The area affected (Containment or the Spent Fuel Pool) must be evacuated and isolated. Those personnel evacuated must be assembled for accountability while remote action(s) are performed to further secure the event to ON-SITE. In addition, the event must be classified and implementation of the Emergency Plan initiated, if require Objective # 14 Abnormal Operating Procedure AP/1(2)/A/550O/4lI Loss of Spent Fuel Cooling or Level, is provided to identify operator actions required during a loss of cooling or level event. This procedure has two Cases; Loss of Spent Fuel Cooling and Loss of Spent Fuel Level.

Symptoms for Case I, Loss of Spent Fuel Cooling are:

Spent Fuel Pool Temperature High Alarm on the OAC (Operator Aid Computer)

Both KF Pumps OFF Subsequeni operator action(s) first determine if fuel movement or movement of any radioactive component, within the Spent Fuel Pool, is taking place. If either of the above is true, then the fuel or component must be placed into a safe position, such as; lowering any fuel assembly within the Spent Fuel Manipulator Crane to the fully down position lowering any radioactive component to fully down O R lowering any fuel assembly within the Upender to the fu//ydownposition. Then a systematic check of KC flow (through the KF Heat Exchanger) is performed, including KC alignment and KC Pump operatio IF BOILING should occur, then makeup in accordance with OP/1(2)/N6200/05, Spent Fuel Cooling System, Enclosure 4.4:

Demineralized Water Borated Water from the FWST Assured Makeup from the RN System OP-MC-FH-KF FOR TRAINING PURPOSES ONLY REV. 22 Page 49 of 85

---_.--

1 P Unit 2 is operating at 100 % power. Given the following events and conditions:

"B" essential train is in servic A RN train is in operation for testin The RN trains are split with 2RN-41B (TRAIN 6 ro NON-SSHDR/SOL)

close Which one of the following statements correctly describes the potential consequence if 2RN-190B (RN To B KC Hx CONTROL) failed to perform its automatic function?

k Overheating 2B RN pum Flashing in the 2B KC heat exchange Overheating the running B train KC pump RN-41B will open to restore flow to the heat exchange __..__________._____----...-------

Ques-894.l.doc

Bank Question: 894 .I Answer: A 1 P Unit 2 is operating at 100 % power. Given the following events and conditions:

" B essential train is in servic A RN train is in operation for testin The RN trains are split with 2RN-41B (TRAIN 5 To NON-ESS HDR /SOL)

close Which one of the following statements correctly describes the potential consequence if 2RN-190B [RN TO 5 KC HxCONTROL) failed to perform its automatic function? Overheating 2B RN pum Flashing in the 28 KC heat exchange Overheating the running B train KC pump RN-416 will open to restore flow to the heat exchange ____________________---------.-

Distracter Analysis: Correct: Lose mini-flow protection for RN pump 2 Incorrect: no flashing should occur, pressure is not changin Plausible: candidate believes that like the letdown regen heat exchanger, flashing on loss of cooling could occur Incorrect: B train pumps cooled by separate suppl Plausible: candidate believes heat exchanger and pump cooling come from the same plac Incorrect:no auto open signal for RN41 Plausible: candidate feels there is some reason for the stated position of 41B in the setup and guesses it can auto open. Valve closes on blackout signa Level: SRO KA: APE 062 AA2.0(2.9/3.6)

Lesson Plan Objective: PSS-RN Obj 7 Source: Bank Level of knowledge: comprehension Ques-894.l.doc

References:

1. OP-MC-PSS-RN pages 23,41,73,85 2. OP-MC-PSS-KC page 39

APE: 062 Loss of Nuclear Service Water AK Knowledge of the operational implications of the following concepts as they apply to Loss of Nuclear Service Water:

(CFR 41.8 I41.10 145.3)

None AK Knowledge of the interrelations between the Loss of Nuclear Service Water and the following:

(CFR 41.7 145.7)

None AK Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water:

(CFR 41.4, 41.8 I45.7)

AK3.01 The conditions that will initiate the automatic opening and closing of the . 3.2' 3.5*

SWS isolation valves to the nuclear service water coolers AK3.02 The automatic actions (alignments) within the nuclear service water . . . . .9 resulting from the actuation of the ESFAS AK3.03 Guidance actions contained in EOP for Loss of nuclear service water . . . .2 AK3.04 Effect on the nuclear service water discharge flow header of a loss . . . . .1 of ccw AA Ability to operate and 1 or monitor the following as they apply to the Loss of Nuclear Service Water (SWS):

(CFR41.7 145.5 1 45.6)

AA1.O1 Nuclear service water temperature indications . . . . . . . . . . . . . . . . . . . I AA1.02 Loads on the SWS in the control room . . . . . . . . . . . . . . . . . . . . . . . .3 AA1.03 SWS as a backup to the CCWS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6' 3.6 AA1.04 CRDM high-temperature alarm system ....................... 2.7+ 2.8 AA1.05 The CCWS surge tank, including level control and level alarms, and radiation alarm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 AA1.06 Control of flow rates to components cooled by the SWS . . . . . . . . . . . . .9 AA1.07 Flow rates to the components and systems that are serviced by the SWS; interactions among the components . . . . . . . . . . . . . . . . . . . . . . . . . .0 AA Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:

(CFR:43.5 145.13)

AA2.01 Location of a leak in the SWS ............................ .5 of the system causing the abnormal condition . . . . . . . . . . . . . . . . . . . .9 AA2.04 The normal values and upper limits for the temperatures of the components cooled by SWS . . . . . . . . . . . . . . . . . . . . . . . . . . .9'

AA2.05 The normal values for SWS-header flow rate and the flow rates to the components cooled by the SWS . . . . . . . . . . . . . . . . . . . . 2.4* 2.5'

AA2.06 The length of time after the loss of SWS flow to a component before that component may be damaged . . . . . . . . . . . . . . . . . . . . . . 2.8* 3.1*

4.2-49 NUREG-1122, Rev. 2 I

DUKE POWER MCGUIRE OPERATIONS TRAINING I

CLASSROOM TIME (Hours)

I NLO I NLOR 1 LPRO I LPSO I LOR I 1 2 1 2 1 2 1 2 1 1 1 OBJECTIVES OBJECTIVE State the purpose of the Nuclear Service Water Syste State the Ultimate Heat Sink requirements and how these requirements are satisfied at McGuire Nuclear Statio List the power supplies for the RN pumps Relating to the RN pumps:

0 List the automatic start signals 0 State the actions required to manually stop the pump(s) following an automatic star Describe the operation of the RN Pump Base Drain, including the sump name to which it is pumpe Concerning the RN Pump Strainers:

State the conditions which will initiate an automatic strainer backwas Describe the actions required if a strainer high delta P

-. .

signal were to occur while a SSsignal was present including the reason for this design featur Concerning the RN pump mini-flow protection:

Describe how mini-flow protection is accomplishe Discuss the operation of the RN mini-flow manual loader and how flow is controlle ."

..-

,

OF-MC-PSS-RN FOR TRAINING PURPOSES ONLY REV. 29 Page 7 of 99

DUKE POWER

- MCGUIRE OPERATICNVS TRAINING 1 Objective #6 i h Each RN pump is provided with a strainer which will remove any trash and/or debris greater than or equal to 3/16 inch in diameter from its suction line. ( Refer to Drawing 7.2 for strainer configuration in the system ). Each strainer has a strainer motor which is safety related. The RN strainer motors have no manual separate or individual controls. Each strainer motor is interlockedto run when its corresponding RN pump is running. The strainer is also provided with an automatic timer which will backwash the strainer once every 120 hours0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months . If the strainer delta P reaches 1.86 psi (51.47 in.wg), an automatic backwash on Hi Strainer Delta P will be initiated. If the auto backwash actuates on the 1.86 psi signal, the 120 hour0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months timer will be reset. The strainer automatic backwash valves receive a close signal on a Sssignal. If a high Delta P signal were to occur while the SSsignal is present, backwash would have to be performed manually (manual valve operation). This feature is designed to prevent unnecessary loss of water from the system. Water for the backwash is provided by the respective RN pump and discharges to the opposite train RC crossover discharge. Manual backwash can also be performed by an operator by using a local pushbutton. When performed manually, the strainers should be backwashed for 3 to 5 minutes. Manual backwash (electric push button operation) is performedwhen the automatic backwash is not available and periodically on "some" frequency. The rounds sheet value calls for checking Delta P below 50 in w Obptive # 7 I Mini-flow protection for the RN pumps is provided by flow through the KC heat

,--. Exchanger ( Refer to Drawing 7.4 ). When an RN pump starts, the train related RN to KC inlet isolation valve (RN86A, RN187B) will open ( providedtheir auto/manual selector switch is in auto ). These valves also open on a train related SSor Blackout signal and can be operated by open/close pushbutton on the RN section of MC11. The train related RN to KC heat exchanger flow is controlled by outlet control valve (

RN89A, RNlSOB ) manual loaders located on the RN section of MC11. If RN flow falls below 2700 gpm, the auto control feature will override the manual loader and open the valve propo.rtiona1 to flow between 2700 gpm and 0 gpm. Valves RN89A and RNlSOB

.

will fail open with the aid of springs to open the actuator on loss of air or SS Meter indication for the RN to KC A(B)HX flow ( 0 to 10,000 gpm) is provided on MC The following alarms on AD12 are providedforthe RN pump and strainer

"A(B) RN PUMP LO SUCTION PRESS

. Setpoint : 2 psig Origin: Comes off LP side of strainer D/P instrumentation Probable cause:

1. fouled strainer 2. valve misalignment 3. low level from suction 4. leak 5. excessive flow on A or B RN header Automatic actions: - None OP-MC-PSS-RN FOR TRAINING PURPOSS ONLY REV. 23 Page 23 of 99

-

DUKE PO WR -

MCGUIRE OPERATIONS TRAINING

,_ The following are the Train A modulating valves: Safe Position

/? RN-89A ( RN to A KC HX Control ) Open*

RN-21A (RN Strainer A Backflush Automatic Supply Isolation) Close RN-22A ( RN Strainer A Backflush Automatic Drain Isolation ) Close ND-29 ( A ND HX Outlet ) Open KC-57A ( A ND HX Return ) Open The following are the Train B modulating valves: Safe Position a RN-190B ( RN to B KC HX Control ) Open*

RN-258 ( RN Strainer B Backflush Automatic Supply Isolation ) Close RN-26B ( RN Strainer B Backflush Automatic Drain lsolation ) Close ND-14 ( B ND HX Outlet ) Open KC-82B ( B ND HX Retum ) Open Theses valves open to their travel stop position ( intermediate) Instrumentation Objective # 13 The following parameter indications associated with the RN System are located on MC09:

A RN Pump Discharge pressure (0-15Opsig)

B RN Pump Discharge pressure (0-15Opsig)

F -

A RN Pump Flow -

( 0 20,000 gpm )

B RN Pump Flow -

( 0 20,000 gpm )

RN Non-essential Header Pressure -

( 0 135 psig )

RN to A KC HX Flow (0-10,000gpm)

RN to B KC HX Flow ( 0 - 10,000 gpm )

RN to A NS HX Flow -

( 0 5,000 gpm )

RN to B NS HX Flow ( 0 - 5,000 gpm )

RN to A D/G HX Flow (0-15,000gpm)

RN to B D/G HX Flow -

( 0 15,000 gpm )

Standby NSW Pond Temp ( 30 - 100 OF Standby NSW Pond Level ( 7381 1" - 741' elev.)

.:. .

-..

OP-MC-PSSRN FOR TRAINING PURPOSES ONLY REV. 29 Page 41 of 99

DUKE POWER MCGUIRE OPERATIONS TRAINING 7.5, RN System Both Units Ss, BO and SpValve Logic (7/15/97)

OP-MC-PSS-RN FOR TRAINING PURPOSES ONLY REV. 29 Page 73 of 99

~

DUKE POWER MCGUiRE OPERATIONS TRAINING

...^. 7.11, RN System Unit Blackout Loads and Valve Logic (7/2/98)

./-

$2

=

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e!

u u $ % K

K a

%

OP-MC-PSSRN FOR TRAINING PURPOSES ONLY REV. 29 Page 85 of 99

DUKE POWER MCGUIRE OPERATIONS TRAINING DRAWINGS KC System (2/12/98) v-4 I I OP-MC-PSS-KC FOR TRAINING PURPOSES ONLY REV. 15 Page 39 of 63

7 -

DUKE POWER MCGUIRE OPERATIONS TRAINING 7.4, RN System Component Loads (1/14/00)

L?

v)

D-P v)

Lo p; 3 OP-MC-PSS-RN FOR TRAINING PURPOSES ONLY REV. 29 Page 71 of 99

06/11/03 WED 0 9 : 2 5 FAX 704 8 7 5 5094 HLP EXAM DEV k9 001 Bank Question: 1071 Answer: C 1Pt(s) After C l ~ i ~ n1e 7300 l Process Control Cabhct Channel Operability Test was completed the Unit 1 Pressurizer level master malfunctions causing it to dcmand full output while in automati which one of the following statements correctly describes the basi,sfor the McGuL-e limit on flow? Letdown flow rates in excess of 135 gpm are limited to ensure proper demineralizer operation and adhere to the design limits of the letdown pipin Letdown flow ratesin excess of 120 gpm exceed the design limits of the letdown orifice valves and induce resonance vibratio Charging flow rates in excess of 100 gpm during normal operation can induce vibration in the regenerative heat exchanger tube Char!$ng flow rates between 65 gpm and 100 gpm total charging flow will cause flashing in the regenerative heat exchange ______-_-__._----------- I - _

Distracter Analysis:

- - Inorrect:

-_. -

Post-IP Fax Note 7R71 rDde lfof I Incorrect Plausible Correct: Phone 11 Plansible Phwe #

FRX R Fax # Incorrect Plausible:

Level SRO KA: APE 028 AA2.09 (2.913.2)

Lesson Plan Objective: OP-MC-PS-NV Obj. 13 Source: New Lwcl of knowledge: comprehension References:

1. OP-MC-PS-W pages 81 & 83 Ques-iO7l.doc

06/11/03 WED 0 9 : 2 5 FAX 704 875 5094 ALP EXAM DEV moo2 DUKEPOWER

=

MCGUIRE QPERAVONS TRAINING Letdown Orifice Isolation Valves must have a full open indication prior to releasing the valve "Open" switch to prevent closure of the associated valve. (PIP M95-1541)

Basis: Self-explanator WHEN the NV Pump shaft driven oil pump is supplying oil pressure, maintain associated NV Lube Oil Pump in the "AUTO' position. (PIP M96-1270) .

Basis: The shaft driven pump provides adequate operating oil pressur When in ATUO, the auxiliary electric lube oil pump will start if the shaft driven pump pressure drops t o 8# to protect the NV Pump bearing When unit shutdown and no NV Pump operating, the NV Lube Oil Pump should be operated in manual for any NV Pump aligned such that a flow path exists through the NV Pump(s).,(PIP M96-1270)

Basis: Provides lubrication when potential for windmilling exist all letdown paths are to be isolated, the gravity drainage paths from the FWST and VCT should be isolated als Basis: This prevents an inadvertent transfer of water to an undesired locatio The actions of Enclosure 4.7 (Operator Action With NV Aux Spray In Service) (of the NV Letdown procedure) should be performed immediately if a Safety Injection occurs and shall not interfere with the actions of EP/l/N5000/E-O (Reactor Trip or Safety Injection), (PIP-O-M96-2396)

Basis: Satisfies FSAR requirements for Aux Spray isolatio IF letdown diverted to the RHT AND unable to maintain VCT level, letdown flow should be reduce Basis: Minimize charging demand and VCT level los = Maximum letdown flows are as follows:

120 gprn through Normal Letdown 150 gpm through ND Aux Letdown with single Mixed Bed Demin in service 185 gpm through ND Aux Letdown with parallel Mixed Bed Demin in servic Basis: Flow is limited to ensure effective demineralizer operation and to adhere to design limits of letdown pipin Maximum letdown header pressure is 255 psig to avoid lifting 1NV-156 (255 psig setpoint)

Basis: Self-explanator OP-MC-PS-NV FOR TRAINING PURPOSES ONLY REV. 42 Page 81 of 145

06/11/03 WED 0 9 : 2 5 FAX 704 875 5094 HLP EXAM DEI' @I003 DUKEPOWER

-_6 .. ..

If NC pressure is less than IO0 psig Q& seal injection water is NOT supplied, Seal Bypass Return Isolation Valve and Seal Retum Isolation Valves shall be close This prevents unfiltered water being supplied to the seal Basis: Self-explanator WHN NC System pressure greater than 7 00 psig temperature greater than 150"F,Seal Injection should be maintained to NC Pump Basis: Ensures pure seal injection water through seals when pressure is high enough to aUow seal leakoff (100 psig) or provide cooling t o seals and bearing when required (150°F).

Ifnormal letdown is lost with PZR water temperature z 250 "F, NV Aux Spray is prohibited (prevents thermal shocking spray nozzle).

Basis: Self-explanator When NV Aux Spray is used, maximum spray D/T is 320 O Basis: Avoids thermal shocking spray nozzl Maximum charging flows are as follows: (PIP M96-2513)

1 144 gpm through Regen HX during transient / shutdown operation 100 gpm through Regen HX during normal operation Basis: Excessive flow can induce vibrations in the regenerative heat exchanger tube WHEN idle NV System components placed in service, boron concentration differences can affect reactivit Basis: Reminder to the operator of the potential impact on core reactivit Design flow rate of # 1 PD Pump is 98 gp Basis: Operator informatio Placement of PD Pump Speed Control in "AUTO" is prohibite Basis: Auto operation i s not maintained by IA Maintain 1NV-159 locked closed and 1NV-157 locked open due to the potential "gas stripping" of the mini-flow orifice. This alignment routes gases to the VCT instead of the charging pump suction. (PIP 98-0036)

Basis: This prevents potential gas binding of the ECCS high head pump Maximum NCP seal injection flow is 12 gpm per pum i

_ Basis; Minimize concern with overpressurizing the seal are OFI-MGPS-NV FOR TRAININGPURPOSES ONLY REV. 42 Page 83 of 145

06/09/03 HOY 0 0 0 0 FAY 7 0 4 875 5094 HLP EXAM DEV d

After Channel 1 7300 Roc Mccrujre limit on flow?

k Letdown flow rates ess of 135 gpm are limited to ensure operation and adhere to the design limits Letdown flow alves and induce resonance vibratio KA: APE 028 AA2.09 (2.9/3.2)

1 Lesson Plan Objective:

Source: New Level of knowledge: comprehension References:

Ques-lO7l.doc

06/09/03 NOi$ 09:OO FAX io4 8 7 5 5004 HLP EXAM DEV moo2 1. OP-MC-

i 1 Pt. Unit 2 has just begun to shutdown (decreasing 2MWe/min) for refuelin Given the following events and conditions:

Pressurizer level is at program level and in automati The controlling pressurizer level transmitter fails at its current outpu No operator action is take Which one of the following statements correctly describes the system response as plant load is reduced? Charging flow decreases Letdown isolates Pressurizer heaters turn off Charging flow increases Pressurizer heaters energize Pressurizer level increase to the trip setpoint Charging flow decreases Letdown will not isolate Pressurizer level decreases until the pressurizer is empty Charging flow increases Pressurizer heaters will not energize Pressurizer level increases to the trip setpoin Bank Question: 90 Answer: A 1 P Unit 2 has just begun to shutdown (decreasing 2MWelmin) for refueling Given the following events and conditions:

Pressurizer level is at program level and in 'automatic'.

The controlling pressurizer level transmitter fails at its current outpu No operator action is take Which one of the following statements correctly describes the system response as plant load is reduced? Charging flow decreases Letdown isolates Pressurizer heaters turn off Charging flow increases Pressurizer heaters energize Pressurizer level increase to the trip setpoint Charging flow decreases Letdown will not isolate Pressurizer level decreases until the pressurizer is empty Charging flow increases Pressurizer heaters will not energize Pressurizer level increases to the trip setpoin ____..__________._._-.-------------

Distracter Analysis: As load is reduced, Tave will decrease, Program Pressurizer level will decrease. The system will see the controlling channel

- maintaining a high level and decrease charging in an effort to reduce leve Actual level will decrease. The backup channel will decrease and at 17%,

letdown will isolate and heaters will de energiz Correct: Incorrect: charging flow will decreas Plausible: candidate believes charging flow will increase in an effort to maintain the higher level Incorrect: letdown will isolate from the bakup channe Plausible: if the candidate believes the low level interlock will not be satisfied only from the controlling channe Incorrect: charging flow will decreas Plausible: candidate believes charging flow will increase to maintain the higher level Level: SRO Ques-902.2.doc

KA: APE 028 AA2.09 (2.913.2)

Lesson Plan Objective: PS-ILE Obj. 12 Source: BANK Level of knowledge: Comprehension References:

1. OP-MC-PS-ILE page 33 (Figure 7.2)

APE ot8 Presnrizm (pzR) Level Control Malfunction BBnrrY A b i i to operate and I or monitor the lokwfng as they apply to the Pressurizer Level Control Malbcth-~ns:

(CFR41.7 145.5 145.6)

AAl.01 PZR le-vel reactor protection bistables . . . . . . . . . . . . . . . . . . . . . . , . 3.8* AAl.02 cvcs ............................................ .4 AA1.03 RCPamlsealwatersystem .............................. .9 AA1.04 Regenerative heat exchanger and temperatun Iimii . . . . . . . . . . . . . . . .8 AA1.05 Initiation of excess letdown per the CVCS . . . . . . . . . . . . . . . . . . . . . .9 AA1.06 C h e d d u g o f R C S l d s ................................. .6 AA1.07 Charging pumps maintenance of PZR level (including manual backup) . . . .3 AAl.08 Selection of an alternate PZR level channel if one has failed .. . . . . . . . .6 AA Ability to determine and Wupret the folkwing as they apply to the Pressurizer Lepel Control Malfunctions:

(CFR: 43.5 I 45.13)

AA2.01 . . .. .

PZR level indicators and alarms . . . . . . . . . . . . . . . . . . . . . . .6 AA2.02 PZR level as a function of power level or T-ave. includiag

. . . . .

intemntatim of malfunction . . . . , .. . . . . . . . . . . . . . . . . . .8 AA2.03 C h & g subsystem flow indicator and controller . . . . . . . . . . . . . . , . .3 AA2.04 Ammcten and running indicators for CVCS charging pumps . , . . . . . , . .1 AA2.05 .

Flow control valve isolation valve indicator . . . . . . . . . . . . . . . . . . . .7 AA2.06 Letdownflowiadicator ................................. .8 M .0 7 Sed water now indicator for RCP . . . . . . . . . . . . . . . . . . . . . . . . . . .9

................

Y AA2.10 Whether tbc automatic mode for PZR level control is fmctioaing improperly. necessity of shift to manual modes .. .. . . . . . . .4 AA2.11 WinPZR ....................................... .6 AA2.12 Cause for PZR level deviation dam: controller mal-function or other instrumentationmalfunction . . . . . . . . . . . . . ... . .

. .5 AA2.13 The actual PZR level, given uncompcnsatcd level with an appropriategraph .................................... .2 AA2.14 The effect on indicatal PZR levels, given a change in ambient pressure and temperature of reflux boiling . .. . .. .. . . . . . .

. .8 4.2-23 NUREG-1122,Rev. 2 I

DUKE POWER MCGUIRE OPERATIONS TRAINING OBJECTIVES OBJECTIVE Describe the protection (signals, setpoints, permissives)

associated with Pressurizer level (logic not required).

Describe the actions the operator must take to restore Pressurizer heater operation following a low level heater cutof ~ ~

For any Pressurizer Level Control System input signal failure, determine the effect and evaluate operator action to be take Determine program Pressurizer level for interim power levels between 0% and 100%.

Concerning the Technical Specifications related to the Pressurizer Level Control System:

Given the LCO title, state the LCO (including any COLR values) and applicabilit For any LCO's that have action required within one hour, state the actio Given a set of parameter values or system conditions, determine if any Tech Spec LCO's is (are) not met and any action(s) required within one hou Given a set of parameters or system conditions and the appropriate Tech Specs, determine required action(s).

Discuss the bases for a given Tech Spec LCO or Safety Limit

SRO ONLY OP-MC-PS-ILE FOR TRAINING PURPOSES ONLY REV. 14 Page 7 of 45

DUKE POWER MCGUIRE OPERATIONS TRAINING 7.2. Pressurizer Level Control Circuit (simplified) (8/14/97)

Objective #2 Z

F

ia2 Y)

m u)

c Y)

N Y)

OP-MC-PS-ILE FOR TRAINING PURPOSES ONLY REV. 14 Page 33 of 45

1 P Which one of the following is a correct list of SAFETY LIMITS? Thermal Power, RCS Highest Loop Tave and Pressurizer Pressur Thermal Power, AFD, Pressurizer Pressur AFD, QPTR and Reactor Powe Linear Heat Generation Rate, Thermal Power and QPT __._____.________.____l_______ _---Y_.--.___--.___-___.

Ques-991. Ldoc

Bank Quesfion: 991. I Answer: A 1 P Which one of the following is a correct list of SAFETY LIMITS? Thermal Power, RCS Highest Loop Tave and Pressurizer Pressur Thermal Power, AFD, Pressurizer Pressur AFD, QPTR and Reactor Powe Linear Heat Generation Rate, Thermal Power and QPT _____..____.______..___________I________-----------.--------.-------.----

Distracter Analysis: Correct:.

Plausible: Incorrect:

Plausible: Incorrect: Incorrect:.

Plausible:

Level: SRO Only KA: G2.1.10 (2.7/3.9)

Lesson Plan Objective: (None)

Source: New Level of knowledge: memory

-.

References:

l.TechSpec2.1.1 Ques-991.l.doc Conduct of Operations (continued)

2.1.9 Ability to direct personnel activities inside the control roo (CFR: 45.5 145.12 145.13)

IMPORTANCE RO SRO 4.0 2.1.11 Knowledge of less than one hour technical specification action statements for system (CFR: 43.2 145.13)

IMPORTANCE RO SRO 3.8 2.1.12 Ability to apply technical specifications for a syste (CFR: 43.2 143.5 145.3)

IMPORTANCE RO SRO 4.0 2.1.13 Knowledge of facility requirements for controlling vital I controlled acces (CFR: 41.10 143.5 145.9 145.10)

IMPORTANCE RO SRO 2.9 2.1.14 Knowledge of system status criteria which require the notification of plant personne (CFR: 43.5 145.12)

IMPORTANCE RO SRO 3.3 2.1.15 Ability to manage short-term information such as night and standing order (CFR: 45.12)

IMPORTANCE RO SRO 3.0 2.1.16 Ability to operate plant phone, paging system, and two-way radi (CFR: 41.10 145.12)

IMPORTANCE RO SRO 2.8 2. I. 17 Ability to make accurate, clear and concise verbal report (CFR: 45.12 145.13)

IMPORTANCE RO SRO J NUREG-I 122, Rev. 2 2-2 I

SLS !

i SLs 2. Reactor Core SLs In MODES Iand 2, the combination of THERMAL POWER, Reactor Coolant System (RGS) highest loop average temperature, and pressurizer pressure shall not exceed the SLs specified in Figure 2.1.1-1 for four loop operatio . RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained 5 2735 psi .2 S L Violations 2. If SL 2.1 .I is violated, restore compliance and be in MODE 3 within 1 hou . If SL 2.1.2 is violated:

2.2. In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hou .2. In MODE 3, 4, or 5, restore compliance within 5 minute McGuire Units 1 and 2 2.0-1 Amendment Nos. 184/166

, . , .,. : . : .... '.

.

.I .

SLS ) 67 DO NOT OPERATE IN THIS A R B 66(

65C 640 h

LL L 630 0)

>

r-"

620

E 610 600

-

590 ACCEPTABLE OPERATION 580 .2 .6 .o ? .2 Fraction of Rated Thermal Power Figure 2.1.1-1 Reactor Core Safety Limits -

Four Loops in Operation

-

McGuire Units 1 and 2 2.0-2 Amendment Nos. 191 Unit 1 172 [ U n i t 2 1

T 1 P Unit 1 has experienced a 50% load rejection which resulted in Control Bank D Group 1 being greater than 12 steps misaligned from its associated step counter. Tech Spec 3.1.4 Rod Control Group Alignment Limits states:

All shutdown and controlrods shall be OPERABLE; with all individual indicafedrodpositions within12 steps of their group step counter demand position:

Which one of the following is the bases for this Tech Spec? Ensure SDM limits are maintained and QPTR is maintained within limit Ensure power distribution and SDM limits are preserve Ensure QPTR is maintained within limits and rod alignments are correc Ensure AFD is maintained and limit power distributio ._._______.__.____.------..-----..---

Ques-1004.l.doc

Bank Question: 100 Answer: B 1 P Unit 1 has experienced a 50% load rejection which resulted in Control Bank

" D Group 1 being greater than 12 steps misaligned from its associated step counter. Tech Spec 3.1.4 Rod Control Group Alignment Limits states:

"All shutdown and control rods shall be OPERABLE; with all individual indicated rodpositions within12 steps of their group step counter demand position'!

Which one of the following is the bases for this Tech Spec? Ensure SDM limits are maintained and QPTR is maintained within limit . Ensure power distribution and SDM limits are preserve Ensure QPTR is maintainedwithin limits and rod alignments are correc Ensure AFD is maintained and limit power distributio _____.____________._-------.--.

Distracter Analysis: Incorrect:.

Plausible: Correct: Incorrect:

Plausible: Incorrect:.

Plausible:

Level: SRO KA: G2.2.25 (2.5/3.7)

Lesson Plan 0bjective:OP-MC-IC-IRX Obj. 14 Source: BANK Level of knowledge: Memory References:

1.T.S. 3.1.4 Bases Ques-1004.l.doc I

-) Equipment Control (Continued)

2.2.18 Knowledge of the process for managing maintenance activities during shutdown operation (CFR: 43.5 145.13)

IMPORTANCE RO SRO .2.19 Knowledge of maintenance work order requirement (CFR: 43.5 145.13)

IMPORTANCE RO SRO .2.20 Knowledge of the process for managing troubleshooting activitie (CFR: 43.5 145.13)

IMPORTANCE RO SRO .2.21 Knowledge of pre- and post-maintenance operability requirement (CFR: 43.2)

IMPORTANCE RO SRO .2.22 Knowledge of limiting conditions for operations and safety limit (CFR: 43.2 145.2)

! IMPORTANCE RO SRO .2.23 Ability to track limiting conditions for operation (CFR: 43.2 145.13)

IMPORTANCE RO SRO .2.24 Ability to analyze the affect of maintenance activities on LCO statu (CFR: 43.2 145.13)

IMPORTANCE RO SRO .2.26 Knowledge of refueling administrative requirement (CFR: 43.5 145.13)

IMPORTANCE RO SRO .2.27 Knowledge of the refueling proces (CFR: 43.6 145.13)

IMPORTANCE RO SRO NUREG-1122,Rev. 2 I

DUKE POWER MCGUIRE OPERATIONS TRAINING OBJECTIVES

--

L L L P P O OBJECTIVE R S R

-

0 0 Describe the Control Room controls, indications and alarms, X x x including alarm setpoint __

Given a limit or precaution associated with the Reactor X x x Control System, discuss its basis and applicabilit __

I F,? .I.:I , -..,.".",,,,ii.:,.-l. ..

..>,I

~ ~ B W i i i g , ~ t ~ ~ ~ ~ e ~ h ~ i ~ a l ~ S p etoc the i~~ati~ris~r~I~ted I

Reactor Control System:

Given the LCO title, state the LCO (including any COLR X x x values) and applicabilit For any LCO's that have action required within one hour, X x x state the actio Given a set of parameter values or system conditions, X x x determine if any Tech Spec LCO's is (are) not met and any action(s) required within one hou s Given a set of plant parameters or system conditions and X x x the appropriate Tech Specs, determine required action(s).

Discuss the bases for a given Tech Spec LCO or Safety x *

Limit

SRO Only

~

OP-MC-/C-/RX FOR TRAINING PURPOSES ONLY REV. 18 Page 7 of 65

I Rod Group Alignment Limits B 3.1.4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Rod Group Alignment Limits BASES BACKGROUND The OPERABILITY (e.g., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SD The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design,"

GDC 26, "Reactivity Control System Redundancy and Protection" (Ref. 1). and 10 CFR 50.46, "Acceptance Criteria for Emergency Core

'

Cooling Systems for Light Water Nuclear Power Plants" (Ref. 2).

Mechanical or electrical failures may cause a control rod to become inoperable or to become misaligned from its group. Control rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SD Limits on control rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserve Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDMs). Each CRDM moves its RCCA one step (approximately 5/8 inch) at a time, but at varying rates (steps per minute) depending on the signal output from the Rod Control Syste The RCCAs are divided among control banks and shutdown bank Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. A bank of RCCAs consists of two groups that are moved in a staggered fashion, but always within one step of each other. The unit has four control banks and five shutdown banks.

McGuire Units 1 and 2 B 3.1.4-1 Revision No. 0

Rod Group Aligpment Limits B 3. BASES BACKGROUND (continued)

.;

The shutdown banks are maintained either in the fully inserted or fully withdrawn position. The control banks are moved in an overlap pattern as described in the Bases for LCO 3.1.6, "Control Bank Insertion Limits."

The control rods are arranged in a radially symmetric pattern, so that control bank motion does not introduce radial asymmetries in the core power distribution The axial position of shutdown rods and control rods is indicated by two separate and independent systems, which are the Bank Demand Position Indication System (commonly called group step counters) and the Digital Rod Position Indication (DRPI) Syste The Bank Demand Position Indication System counts the pulses from the rod control system that moves the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (* 1 step or f 5/8 inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the ro The DRPl System provides a highly accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is six steps. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one system fails, the DRPl will go on half-accuracywith an effective coil spacing of 7.5 inches, which is 12 steps. Therefore, the normal indication accuracy of the DRPl System is i 6 steps (i'3.75 inches), and the maximum uncertainty is * 12 steps (i7.5 inches). With an indicated deviation of 12 steps between the group step counter and DRPI, the maximum deviation between actual rod position and the demand position could be 24 steps, or 15 inches.

,

McGuire Units 1 and 2 B 3.1.4-2 Revision No. 0

- Rod Group Aligyment Limits 0 3. APPLICABLE Control rod misalignment accidents are analyzed in the safety analysis SAFETY ANALYSES (Ref. 3). The acceptance criteria for addressing control rod inoperability or misalignmentare that: There be no violations of: specified acceptable fuel design limits, or Reactor Coolant System (RCS) pressure boundary integrity; and The core remains subcritical after accident transient Two types of misalignment are distinguished. During movement of a control rod group, one rod may stop moving, while the other rods in the group continue. This condition may cause excessive power peakin The second type of misalignmentoccurs if one rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the control rods to meet the SDM requirement, with the maximum worth rod stuck fully withdraw Analyses are performed in regard to static rod misalignment, single rod withdrawal, dropped rod, and dropped group of rods (Ref. 4). With control banks at their insertion limits, one type of analysis considers the case when any one rod is completely inserted into the core. The second type of analysis considers the case of a completely withdrawn single rod from a bank inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio in both of these cases bounds the situation when a rod is misaligned from its group by 12 steps. Another type of misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition is assumed in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 5).

The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserve Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor (Fo(X.Y,Z)) and the nuclear enthalpy hot channel factor (FNd~(X,Y)) are verified to be within their limits in the COLR and the safety analysis is verified to remain vali When a control rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the McGuire Units 1 and 2 0 3.1.4-3 Revision No. 0

,

Rod Group Aligvment Limits B 3. BASES ..

.' APPLICABLE SAFETY ANALYSES (continued)

design peaking factors, and Fa(X,Y,Z) and FNm(X,Y) must be verified directly by incore mapping. Bases Section 3.2 (Power Distribution Limits)

contains more complete discussions of the relation of Fo(X,Y,Z) and FNw(X,Y) to the operating limit Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in the safety analyses. Therefore they satisfy Criterion 2 of 10 CFR 50.36 (Ref. 6).

LCO The requirements on rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The limits on shutdown and control rod alignments ensure that the assumptions in the safety analysis will remain valid, and that the RCCAs and banks maintain the correct power distribution and rod alignment The requirement to maintain the alignment of any one rod to within plus or minus 12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed. Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysi APPLICABILITY The requirementson RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (Le., trippability)

and alignment of rods have the potential to affect the safety of the plan In MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are normally bottomed and the reactor is shut down and not producing fission power. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM. but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5 and LCO 3.9.1, 'Boron Concentration," for boron concentration requirementsduring refuelin ACTIONS A.l.l and A. When one or more rods are untrippable, there is a possibility that the required SDM may be adversely affected. Under these conditions, it is McGuire Units 1 and 2 B 3.1.4-4 Revision No. 0

1 Pt. Unit 1 is operating at 100% power when the following occurs:

LOCA outside cpntainment FWST level is 1 8 0 Containment Sump level is 2.75 feet FWST has not ruptured ES 1.3 Transfer to Cold Leg Recirc being implemented REFERENCEPROVIDED Which of the following describes the correct procedure flowpath?

A. Go to ECA 1.1 (Loss of Emergency Coolant Recirc)

0 Go to ECA 1.2 (LOCA Outside Containment)

C Go to ES 1.2 (Post LOCA Cooldown and Depressurization)

D. Continue in body of ES 1.3 (Transfer to Cold Leg Recirc)

,

Bank Question: 1033 Answer: A 1 P Unit 1 is operating at 100% power when the following occurs:

LOCA outside containment FWST level is 1 8 0 Containment Sump level is 2.75 feet FWST has not ruptured ES 1.3 Transfer to Cold Leg Recirc being implemented REFERENCE PROVlDED ES-1.3 AND Enclosure 2 Which of the following describes the correct procedure flowpath?

A. Go to ECA 1.1 (Loss of Emergency Coolant Recirc)

B Go to ECA 1.2 (LOCA Outside containment)

C Go to ES 1.2 (Post LOCA Cooldown and Depressurization)

D. Continue in body of ES 1.3 (Transfer to Cold Leg Recirc)

...........................................................................................

Distracter Analysis: Correct- Incorrec Incorrect Incorrect Level: SRO KA: EPE W l E l l EA2.1 (3.4/4.2)

Lesson Plan Objective: OP-MC-EP-E1 Obj. 6 Source: New Level of knowledge: Analysis Author: CWS References:

1. OP-MC-EP-E1 page 157 Ques-1033.doc

2. ES 1.3 Transfer to Cold Leg Recirc. Provided Ques-1033.doc

EPE 1 I Loss of Emergency Coolant Recirculation (Continued)

I KIA N KNOWLEDGE EK3.2 Normal, abnormal and emergency operating procedures associated with (Loss of Emergency Coolant Recirculation).

IMPORTANCE RO SRO EK3.3 Manipulation of controls required to obtain desired operating results during abnormal, and emergency situation IMPORTANCE RO SRO EK3.4 RO or SRO function within the control room team as appropriate to the assigned I position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violate IMPORTANCE RO SRO ABILITY EA1. Ability to operate and I or monitor the following as they apply to the (Loss of Emergency Coolant Recirculation)

(CFR: 41.7 145.5 145.6)

EA1.l Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual feature IMPORTANCE RO SRO EA1.2 Operating behavior characteristics of the facilit IMPORTANCE RO SRO EA1.3 Desired operating results during abnormal and emergency situation IMPORTANCE RO SRO EA2. Ability to determine and interpret the following as they apply to the (Loss of Emergencv Coolant Recirculation)

4.5-29 NUREG-1122. Rev. 2 I

DUKE POWER MCGUIRE OPERATIONS TRAINING CLASSROOM TIME (Hours)

I NLO I NLOR I LPRO I LPSO I LOR I 11 NIA 1 I

N/A 1 I I I I I

OBJECTIVES S

E OBJECTIVE 1 Explain the purpose for each procedure in the E-I serie EPElOOi 2 Discuss the entry and exit guidance for each procedure in the E-1 serie FPEinn:

3 Discuss the mitigating strategy (major actions) of each procedure in the E - I serie EPE100:

4 Discuss the basis for any note, caution or step for each procedure in the E-I serie EPElOOd

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5 Given the Foldout page discuss the actions included and the basis for these action EPE1005 6 Given the appropriate procedure, evaluate a given scenario describing accident events and plant conditions to determine li any required action and its basi EPE100f Discuss the time critical task(s) associated with the E-1 series procedures including the time requirements and the basis for these requirement EPElOOi OP-MC-EP-E1 FOR TRAINING PURPOSES ONLY REV.12 Page 5 of 267

DUKE POWER MCGUIRE OPERATIONS TRAINING ES-1.3 Transfer to Cold Leg Recirculation STEP 3 Check containment sump level - GREATER THAN 3 FT.

PURPOSE: To ensure there is sufficient level in the sump to support the transfer to Cold Leg Recirc.

BASIS: This step sends the operator to the proper procedure for dealing with a loss of coolant recirculation before the FWST becomes empty if the transfer cannot be performed in order to prevent loss of suction flow to the pumps and potential pump damage. There are three possible causes of inadequate sump level addressed by the RNO to this step.

First, Small LOCAs of certain size and location have been evaluated to cause sump level indication to be less than the required value by the time the FWST reaches the Lo level setpoint. Evaluation of these conditions shows that the proper action would be to transfer to CLR. The evaluation indicated that sufficient sump level would be achieved by the time the FWST reached the Lo Lo level and that vortexing would not be a concern during the period of the realignment prior to reaching Lo Lo level in the tank.

Sump level buildup is assured since safety injection has occurred, FWST Lo Level setpoint has been reached and containment Spray has actuated due to high pressure in containment. These conditions indicate that reactor coolant and FWST inventory is being released into containment and will eventually find its way to the sump.

LOCA Outside Containment or FWST Rupture (Due to a Tornado) will result in FWST deoletion without the corresoondina SumD Builduo. In this case actions-arespecified in of Outside Containment. In the case anFWST rupture the operator.is.direc?te&h3%?turn to the procedure in effect. A Tornado is not'postulated to occur concurrently with a high-energy line break inside containmen Note that if an Orange or Red path Procedure is in effect upon transition out of ES-1.3, it takes priority over any other procedure including ECA- STEP 4 Check KC flow to ND heat exchangers - GREATER THAN 5000 GP PURPOSE: To ensure KC flow to the ND heat exchanger BASIS: This step assumes that the ND heat exchangers are used for heat removal during the post accident recirculation phase and that either KC flow has been automatically provided to the heat exchangers or the operator has manually established KC flow prior to the switchover. If KC flow had not previously been established, then it should be established at this tim OP-MC-EP-E1 FOR TRAINING PURPOSES ONLY REV.12 Page 157 of 267

1 Pt. Given the following conditions:

Pressurizer Level Channel 1 is at 28% level Pressurizer Level Channel 2 associated bistables are in the tripped condition due to surveillance testing Pressurizer Level Channel 3 fails hig N-41 is 8%

N-42 is 10%

N-43 is 9%

N-44 is 9%

Impulse pressure channel 1 is 11%

Impulse pressure channel 2 is 9%

No reactor trip has occurred Which of the following describes the proper , rator re onse?

A Trip the reactor and enter E-0 (Reactor Trip or Safety Injection)

B. Trip the reactor and enter FR-S.l (Response NuclearPower GenerafiodAMIS)

C. Do not trip the reactor because thermal power is less than P-7 D. Do not trip the reactor. Initiate unit shutdow Bank Question: 1035 Answer: D 1 Pt. Given the following conditions:

Pressurizer Level Channel 1 is at 28% level Pressurizer Level Channel 2 associated bistables are in the tripped condition due to surveillance testing Pressurizer Level Channel 3 fails hig N-41 is8%

N-42 is 10%

N-43is9%

N-44is9%

Impulse pressure channel 1 is 11%

Impulse pressure channel 2 is 9%

No reactor trip has occurred Which of the following describes the proper operator response?

A. Trip the reactor and enter E-0 (Reactor Trip or Safety Injection)

B. Trip the reactor and enter FR-S.l (Response Nuclear Power Generation/ATWS)

C. Do not trip the reactor because thermal power is less than P-7 D. Do not trip the reactor. Initiate unit shutdow ______.__.______._______.___

Distracter Analysis: Incorrect: -This example is not at valid ATWS event. This is a loss of reactor protection. Therefore you do not trip the reacto Incorrect: This example is not at valid ATWS event. This is a loss of reactor protection. Therefore you do not trip the reacto Incorrect: Greater than P-7. Pzr Hi level trip is in effect Correct Level: SRO KA: APE 029 EA2.02 (4.Z4.4)

Lesson Plan Objective: OP-MC-IC-IPE Obj. 10 & 11

Source: New Level of knowledge: Analysis Author: CWS References:

1. OP-MC-IC-IPE pages 47,79 & 81 2. OMP 4-3 page 9 Ques-1035.doc

EPE: 029 Anticipated Transient Without Scram (ATWS)

EA1.09 Manual rod control ................................... .6 EA1.10 Rod control function switch .............................. .2 EA1.ll Manual opening of the CRDS breakers ....................... 3.9* 4.1 EA1.12 MIG set power supply and reactor trip breakers . . . . . . . . . . . . . . . . . .0 EA1.13 Manual trip of main turbine .............................. .9 EA1.14 Driving of control rods into the core ........................ .9 EA1.15 AFWsystem ....................................... .9 EA2 Ability to determine or interpret the following 8s they apply to a ATWS (CFR 43.5 I45.13)

EA2.01 Reactor nuclear instrumentation ........................... .1 EA2.04 CVCS centrifugal charging pump operating indication . . . . . . . . . . . . . 3.2* 3.3; EA2.05 System component valve position indications . . . . . . . . . . . . . . . . . . . 3.4* 3.4*

EA2.06 Main Nrbine trip switch position indication .................... .9 EA2.07 Reactor trip breaker indicating lights ........................ .3 EA2.08 Rod bank step counters and RPI ........................... .5 EA2.09 Occurrence of a main Nrbinelreactor trip ..................... .5 EA2.10 Positive displacement charging pumps ....................... 3.1; 3.4*

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NUREG.1122. Rev 2 4.1-10

DUKE POWER MCGUIRE OPERATIONS TRAINING

- - -

N L L ( S I L P P I E l 0BJ E CTlVE 0 IQI R

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8 Describe the function of the First-Out annunciator pane X ICIPEOOL procedure, discuss its basis and applicabilit ICIPEOOZ e input parameter(s), logic and function. For W

13 Briefly describe the incident that occurred at Salem Nuclear x x x x Plant and how this event affected McGuire Reactor Trip Breaker operatio ICIPE013 OP-MC-IC-IPE FOR TRAINING PURPOSES ONLY REV. 22 Page 1 1 of 147

DUKE POWER MCGUIRE OPERATIONS TRAINING I Objective#iO 1 NC Pump Bus Under Frequency (2/4 busses = 56 Hz) - this anticipatory loss of coolant flow trip protects against DNB. The trip also trips open all four NC pump breakers to prevent electrical braking of the pump motors during frequency decay. A reduction in pump speed would reduce fly wheel inertia and pump coast down flow capability. This at-power trip protection i s auto-blocked e 10%

power (P-7) and is automatically reinstated > P- SG Lo-Lo Level (2/4 channels on 1/4 SGs = 17%) protects against a loss of heat sink. This protection also causes an auto-start of the CA motor driven pumps (2/4 channels on 1/4 SGs) and the CA turbine driven Pump (2/4 channels on 2/4 SGs).

-

Sinqle Loop Loss of Flow (2/3channels in 1/4 loops = 88%) protects against DNB. This protection is auto-blocked < 48% (P-8) and automatically reinstated P- Two Loop Loss of .Flow (2/3channels in 2/4 loops = 88%) protects against DN This protection i s auto-blocked c 10% (P-7) and automatically reinstated > P- Safety Injection (anv SI sisnal 1/2 Trains) initiates a reactor trip during LOCA event Turbine Trip (2/3channels AS0 < 45psiq. 4/4 stop valves closed) protects against loss of integrity by preventing Pressurizer PORVs from opening on turbine trip at high power.

e Objective # 4, 10

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General Warninq /2/2 Trains) protects against a loss of both protection train Anytime a General Warning i s present on both SSPS trains a reactor trip will occur. General Warning i s caused by: loose circuit board card; loss of voltage (AC or DC); SSPS train in Test; a Reactor Trip By-pass breaker in the Connected position and Closed; a Logic Ground Return fuse blow .1.3 Protection Permissive Interlocks I Obiective # 11 I

-

P-4 (Reactor Trip Breaker and Bypass Breaker Open for a W e n train) initiates:

Turbine Trip; Feedwater Isolation (coincident with low Tavg of 553 OF); Allows reset of SI signal after one minute time-out; Inputs t o Steam Dump Control System for plant trip mod P-6 (1/2 IR instruments > 10 amps) allows Manual Block of SR reactor trip. O n a power reduction, provides automatic reinstatement of SR high voltage and SR reactor trip when 212 IR channels < lo- amp P-7 (2/4 PR instruments > 10% or 112 Turbine Impulse Pressures > 10%) Enables (unblocks) the at power reactor trips: Pzr Hi-Level, Pzr Lo-Pressure, 2 Loop Loss of Flow, NCP UV, and NCP UF. The above trips are automatically blocked when below P-7,3/4 PR < 10% gnJ 2/2 Impulse Pressure < 10%.

OP-MC-IC-IPE FOR TRAINING PURPOSES ONLY REV. 21 Page 47 of 145

DUKE POWER MCGUIRE OPERATIONS TRAINING

\ Reactor Trips (3/27/01)

/ REACTOR SETPOINT LOGIC 'ERMISSIVES TRIP MANUAL Sw. turned 45" 112 s operator judgment S.R. NI HIGH lo5CPS I12 c '6, P10 uncontrolled rod withdrawall startup accidents 1.R. NI HIGH mps-25% power 112 c '10 uncontrolled rod withdrawal1 startup accidents P.R. NI LOW 25% power Y4 c '10 reactivity excursion from low powers P.R. NI HIGH 109% power Y4 c reactivity excursion from all powers DNB P.R. POS t5%/2 sec Y4 c DNB (rod ejection)

RATE PZR HIGH 2385 psig Y4 c coolant system integrity PRESS PZR LOW 1945 psig 2/4 c '7 DNB PRESS PZR HIGH 32% Y3 c '7 water through safeties (system LEVEL integrity)

OTAT AT > OTATsp V41c DNB OPAT AT> OPATsp 2/4 c ~~

KWIFT NCP BUS 74% of normal U4 c '7 DNB (anticipatory loss of flow)

LOW VOLT NCP BUS 56 Hz U4 c '7 DNB (anticipatory loss of flow)

LOW FREQ SIG LO-LO 17% a 4 in loss of heat sink LVL 114 slg 1 LOOP 88% 2/3 in '8 DNB LOSS OF 114 loops FLOW 2 LOOP 88% 2/3 in '7 DNB LOSS OF 2/4 loops FLOW -

SAFETY any SI1 signal 112 SI1 trip reactor if trip not INJ ECTlON actuated trains generated by trip instrumentation GENERAL loose card, loss of 2/2 alarms loss of protection WARNING voltage, train in ALARM test, by-pass bkr connectedlclosed, logic ground return fuse blown TURBINE low Auto-stop oil 2/3 AS0 '8 trip reactor on turbine trip TRIP press <45 psig or Press all 4 stop valves switches closed 414 valves

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OP-MC-IC-IPE FOR TRAINING PURPOSES ONLY REV. 27 Page 79 of 145

DUKE POWER MCGUIRE OPERATIONS TRAINING

- Protection Permissive Interlocks (06/15/98)

INTERLOCKS * LOGIC FUNCTION Turbine Trip Feedwater Isolation< Low Tave Arms condenser dumps Allows reset of Safety Injection Signal after time delav

~~

P-6 1/2 I.R. > 10-10 amps Uows manual block of S.R. Reactor Tri le-energizes high voltage to the Source 3ange detectors. On decreasing power, Source Range Level trips are automatically

.eactivated and high voltage restore P-7 2/4 P.R. > 10% FP (P-10) or On increasing power P-7 automatically 1/2 impulse pressure> 10% enables the following trips:

(P-13) Pzr High Level Pzr Low Pressure LOWNC Flow 2/4 Loops NCP Undetvoltage

.. NCP Underfrequency On decreasing power the above listed trim are automaticallv blocke /4 P.R. > 48% FP On increasing power P-8 enables the 114 loop loss of flow Reactor Trip and Reactor Trip on Turbine Trip. On decreasing power, P-8 automatically blocks the above listed tri T.S. REFERENCE OF-MC-IC-IPE FOR TRAINING PURPOSES ONLY REV. 21 Page 81 of 145

.: I o m 4-3 Page 9 of 28 The following is a list of automatic safety signals that can be "blocked" from the main control board:

Feedwater Isolation on Reactor Trip and Lo Tave Pressurizer Lo press. SI (Below P-1 1)

Low Pressure Steamline Isolation (Below P-1 I).

CA Auto Start on Lo Lo S/G Level or both CF pumps tripped (Below P- 11)

S/R Hi Flux Reactor Trip I/R Hi Flux Lo Setpoint Reactor Trip P/R Hi Flux Lo Setpoint Reactor Trip St Jnterlock Bypass on NC Sample valves: NM-22, NM-25, NM-26 These signals should NOT be "blocked" except under the direction of an approved station procedure or to better protect the health and safety of the public or to protect the lives of plant personne .8 A.T. An A.T.W.S. (Anticipated Transient Without Scram) is defined in 10CFRS0.62 as an anticipated operational occurrence followed by the failure of the reactor trip portion of the protective system. An anticipated operational occurrence is defined in IOCFRSO, Appendix A, as those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are NOT limited to loss of power to all NC pumps, tripping of the turbine generator, isolation of the main condenser and loss of all offsite power. Clearly, to have an A.T.W.S. there must be transient followed by a failure of the reactor trip breaker Instrument failures, by themselves, are NOT necessarily transients. For example, if one channel of Pressurizer Pressure was out of service for preventive maintenance (bistable in tripped condition) and another Pressurizer Pressure channel failed (NOT the controlling channel), a reactor trip signal would be generated. E the reactor failed to trip, this would be a failure of the reactor trip breakers and the automatic trip features of-the reactor protection system and NOT an A.T.W.S. event. Obviously, the control operators would have to recognize and check that the channel failure was indeed a channel failure by checking the other two Pressurizer Pressure channels in this exampl This would, however, force Operations to shutdown the affected unit to at least Hot Standby per Tech Spec .9 Adverse Containment Setpoints Many setpoints in the EP's are presented in a dual format with a second setpoint enclosed in parentheses. This second setpoint is used to account for the additional error in the setpoint due to the containment environment following a high-energy line break. The setpoint in parentheses will be used whenever containment pressure has exceeded 3 psi Pt. Given the following conditions on Unit 1:

A steam leak has occurred on the main steam header Unit 1 reactor has been tripped and safety injection has actuated The MSlVs will not close 20 minutes into the event lowest loop NC Tcold is 305 degrees Based on the above conditions which one of the following is the correct procedure flowpath? From E-0 Reactor Trip or Safefylnjection GO TO FR- (Response to Anticipated Pressurized Thermal Shock) From E-0 go directly to ECA 2.1, (Uncontrolled Depressurization of all Steam Generators) From E-0 GO TO FR-P.l, (Response to Imminent Pressurized Thermal Shock) From E-0 GO TO E-2, (Faulted Steam Generator Isolation) and then to ECA \

Bank Question: 1037 Answer: D 1 Pt. Given the following conditions on Unit 1:

A steam leak has occurred on the main steam header Unit 1 reactor has been tripped and safety injection has actuated The MSlVs will not close 20 minutes into the event lowest loop NC Tcold is 305 degrees Based on the above conditions which one of the following is the correct procedure flowpath? From E-0 Reactor Trip or Safefy Injection GO TO FR- (Response to Anticipated Pressurized Thermal Shock) From E-0 go directly to ECA2.1, (Unconfrolled Depressurization of all Steam Generators) From E-0 GO TO FR-P.l, (Response to Imminent Pressurized Thermal Shock) From E-0 GO TO E-2, (Faulted Steam GeneratorIsolation) and then to ECA ..______________.___------.------

Distracter Analysis: Incorrect:

Plausible: Incorrect:

Plausible: Correct Plausible:

LEVEL: SRO KA: W/E12 EA2.1 (324.0)

SOURCE: NEW LEVEL OF KNOWLEDGE: Comprehension AUTHOR CWS LESSON: OP-MC-EP-E2

OBJECTIVES: OP-MC-EP-E2 Obj 2,6 REFERENCES: OP-MC-EP-E2 pages 9, 15 & 23 EP/l/A/500O/F-Opage 7 EP/l/A/5000/E.2 page 2 Ques-1037.doc

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/

EPE: I > Uncontrolled Depressurization of all Steam Generators (Continued)

KIA N KNOWLEDGE EA2. Abilitv to determine and intermet the following as they- apply . to the (Zjncontrolled Depresswhition of all Steam Generators)

(CFR:43.5 145.13)

EA2.2 Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendment IMPORTANCE RO SRO .5-33 NUREG-1122,Rev. 2 I

DUKE POWER MCGUIRE OPERATIONS TRAINING CLASSROOM TIME (Hours)

1 NLO I NLOR I LPRO I LPSO I LOR I I NIA I NIA I .75 I .75 I .75 1 OBJECTIVES OBJECTIVE basis for these actions.

OP-MC-EP-E2 FOR TRAINING PURPOSES ONLY REV. 07 Page 5 of 97

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MNS CRITICAL SAFETY FUNCTION STATUS TREES PAGE N PI1/A/5OOO/F-O 70f11 Reactor Coolant Integrity - Page 1 of 1 Rev. 3 I

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', I '1 GO TO FR-P . I L

ALL NC T-COLDS THAN 280'F GFGATERTHAN GFGATER CSF SAT

-. W T E M P E R A N R E DECREAS ALL NC T-COLDS GREATER l H A N 2 W F i

!

I

!

iI I

ALL NC T.COLDS CSF SAT G F S A T E R l l i A N 300'F I

! I I n

11 CSF SAT

MNS FAULTED STEAM GENERATOR ISOLATION PAGE N O j EPI 1/A/5000/E-2 2 0 f 13 3 UNIT 1 Rev. 7 1 4 Y

I I

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I ACTION/EXPCCTEO RESPONSE 1 R E S P O N S E 1.101 OBTAINED C. Operator Actions

- Monitor foldout pag . Maintain at least one SIG available for NC System cooldown in subsequent step . Maintain any faulted SIG or secondary break isolated during subsequent recovery actions unless needed for NC System cooldow .., ' . Check the following CLOSED: - Close valve All MSlVs

- All MSlV bypass valves

- Check at least one SIG pressure - --IF all S/Gs faulted, THEN GO TO STABLE OR GOING U EP/l/N5000/ECA-2.1 (Uncontrolled Depressurization Of All Steam Generators). Identify faulted SIG(s): Perform the following:

-.

- Any S/G pressure - GOING DOWN IN a. Dispatch operators to search for AN UNCONTROLLED MANNER initiating break:

OR - Main stearnlines

- Any SIG - DEPRESSURIZE Main feedlines

- Other secondacy pipin GO TO Step 1 J

..

DUKE POWER MCGUIRE OPERATIONS TRAINING INTRODUCTION 1 E-2, Faulted Steam Generator Isolation, provides actions to identify and isolate a faulted steam generator (S/G). The procedure is entered from E-0, Reactor Trip or Safety Injection, or E-1, Loss of Reactor or Secondary Coolant, when any S/G pressure goes down in an uncontrolled manner or any S/G is completely depressurizes. Other procedures have a transition to E-2 whenever a faulted S/G is identified and faulted SIG isolation is not confirme After taking the required actions in this procedure, the operator is directed to either E-1, Loss of Reactor or Secondary Coolant, or E-3, Steam Generator Tube Rupture, depending on whether a steam generator tube rupture (SGTR) is identifie ECA-2.1, Uncontrolled Depressurization of All Steam Generators, provides procedural guidance to recover from an event where all S/Gs are depressurizing in an uncontrolled manner. This procedure is entered from E-2 when an uncontrolled depressurization of all SIGs occurs. Potential initiating events for this contingency could include steamline breaks, stuck open relief or safety valves, or any combination of conditions that would affect all S/G ECA-2.1 is exited whenever any S/G pressure boundary is reestablished as indicated by a rise in the associated SIG pressure indication. In this case, the operator transfers to E-2 for further recovery action .l. Emergency Procedures in this Series E-2 Faulted Steam Generator Isolation ECA-2.1 Uncontrolled Depressurization of All Steam Generators OP-MC-EP-E2 FOR TRAINING PURPOSES ONLY REV. 07 Page 9 of 97

DUKE POWER MCGUIRE OPERATIONS TRAINING ECAQ.1, Uncontrolled Depressurization of All Steam Generators, provides procedural guidance to recover from an event where all S/Gs are depressurizing in an uncontrolled manne An uncontrolled depressurization of all S/Gs initiates from a failure/break in a main steamline, main feedwater line, andlor in any piping system that interconnects with the secondary side pressure boundary. This event results in an extensive cooldown and pressure transient. The consequences vary considerably depending upon system parameters: Size(@and location(s) of the break(s) Operational safety systems Operational control systems Initial power level Failures which may occu It should be noted that this event (with an extensive cooldown and subsequent repressurization) might result in a challenge to the Integrity Critical Safety Function (CSF). In this case the Integrity CSF Status Tree may direct the operator to FR-P.1, Response To Imminent Pressurized Thermal Shock Condition, for further action The following describe the purpose of the two emergency procedures in the E-2 series.

'

E-2, Faulted Steam Generator Isolation, provides actions to identify and isolate a faulted S/ ECA-2.1, Uncontrolled Depressurization of All Steam Generators, provides actions for a loss of secondary coolant that affects all S/Gs.

.)

OP-MC-EP-E2 FOR TRAINING PURPOSES ONLY REV. 07 Page 15 of 97

DUKE POWER MCGUIRE OPERATIONS TRAINING STEP 5 -

Check at least one S/G pressure STABLE OR GOING UP:

-\ . PURPOSE: To ensure there is at least one non-faulted S/ BASIS: Any cooldown operations that are performed as subsequent recovery actions will require at least one non-faulted S/G. If all S/G pressures are going down in an uncontrolled manner, this indicates a failure affecting all SIGs. Recovery actions, in this case, should be performed using ECA-2.1, Uncontrolled Depressurization of All Steam Generators, since feedwater flow will be necessary to a faulted S/G and normal level control should not be use One faulted S/G may cause the intact S/G pressures to drop just due to the cooldow In this case, the operator should continue in E- STEP 6 Identify faulted S/G(s)

PURPOSE: To identify any faulted S/ BASIS: An uncontrolled S/G pressure drop (following MSlV closure and feedwater isolation) or a completely depressurized S/G indicates an unisolable failure of the secondary pressure boundary. The operator is directed to search for the initiating break in main steamlines, feedlines, or other secondary piping such as blowdown lines, sample lines, etc. The operator should also check for stuck open atmospheric steam dump valves andlor safety valve V-a OP-MC-EP-E2 FOR TRAINING PURPOSES ONLY REV. 07 Page 23 of 97

.

1 Pt. Given the following conditions on Unit 1:

Chemistry had confirmed two leaking fuel rods A large break LOCA occurs E-0 Reactor Trip or Safety lnjection is complete ES-1.3 Transfer fo Cold Leg Recirc is complete E-I Loss of Reactor or Secondary Coolant is complete ES-1.2 Post LOCA Cooldown and Depressurizationis in effec All Red and Orange Paths have been addressed IEMF 51A is reading 39WHR Pressurizer level is 0%

The SRO is currently considering implementing Yellow Path procedures. Which one of the following describes proper procedure implementation? Go to FR-1.3, (Response to Voids in the Reactor Vessel) and exit ES- . Stay in ES-1.2 and implement FR-1.3 concurrently Go to FR-Z.3, (Response to High Containment Radiation Level)

and exit ES- Stay in ES-1.2 and implement FR-2.3 concurrently

Bank Question: 1042 Answer: D 1 P Given the following conditions on Unit 1:

Chemistry had confirmed two leaking fuel rods A large break LOCA occurs E-0 Reactor Trip or Safety Injection is complete ES-1.3 Transfer to Cold Leg Recirc is complete E-I Loss of Reactor or Secondary Coolant is complete ES-1.2 Post LOCA Cooldown and Depressurizationis in effec All Red and Orange Paths have been addressed IEMF 51A is reading 39WHR Pressurizer level is 0%

The SRO is currently considering implementing Yellow Path procedures. Which one of the following describes proper procedure implementation? Go to FR-1.3, (Response to Voids in the Reactor Vessel) and exit ES- Stay in ES-1.2 and implement FR-1.3 concurrently Go to FR-2.3, (Response to High ContainmentRadiation Level)

and exit ES- Stay in ES-1.2 and implement FR-2.3 concurrently

____________________~------

Distracter Analysis:.

- Incorrect: ESI .2 is the controlling procedure and not to be exited Plausible: Incorrect: FR-1-3 is a lower priority than FR-Z-3 Plausible: Incorrect: ES1.2 is controlling procedure Plausible: Correct Plausible:

LEVEL: SRO KA: WlE16 EA2.1 (4.314.4)

SOURCE: NEW LEVEL OF KNOWLEDGE: Analysis Ques-lO42.doc

AUTHOR CWS LESSON: OP-MC-EP-FO OBJECTIVES: OP-MC-EP-FO Obj 3 REFERENCES: OP-MC-EP-FO page 17 OMP 4-3 page 19 Ques-lO42.doc

EPE: I High Containment Radiation (Continued)

KIA N KNOWLEDGE EK3.2 Normal, abnormal and emergency operating procedures associated with (High Containment Radiation).

IMPORTANCE RO SRO EK3.3 Manipulation of controls required to obtain desired operating results during abnormal, and emergency situation IMPORTANCE RO SRO I EK3.4 RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violate IMPORTANCE RO SRO ABILITY

.. .,.,?l, EA1 ...,M i t y to operate and I or monitor the following as they apply to

&e (High Containment Radiation)

2'. ,. '.

. ...(CFR: 41.7 145.5 / 45.6)

.. .. . .

.. .I

.. . ' EA1.l Components, and functions of $ntrol and safety systems, including

. instrumentation, signals, interlocks, failure modes, and automatic and .2 - manual feature IMPORTANCE RO SRO EA1.2 Operating behavior characteristics of the facilit IMPORTANCE RO SRO EAl.3 Desired operating results during abnormal and emergency situation IMPORTANCE RO SRO NUREG-1122. Rev. 2 4.5-44

DUKE POWER MCGUIRE OPERATIONS TRAINING NLO NLOR LPRO LPSO LOR 2 2 2 OBJECTIVES L

P OBJECTIVE State the purpose of each of the six CSF Status Tree EPFOOOI Explain the bases for all blocks in the six Status Tree EPF0004 I

OP-MC-EP-FO FOR TRAINING PURPOSES ONLY REV. 06 Page 5 of 83

OMP4-3 Page 19 of 28 7.15.1:7 Yellow Path A yellow path does NOT require immediate operator attentio Frequently, it is indicative of an off-noma1 and/or temporary condition which will be restored to normal status by actions already in progress. In other cases, the yellow status might provide an early indication of a developing red or orange condition. The operator is allowed to decide whether or NOT to implement any yellow path procedur Implementation of a yellow path function restoration guideline is based on operator judgment when it is determined that adequate time exists to implement it. In other words, the operator does

-

NOT have to implement a yellow path guideline if a judgment has been made that it is inappropriate based on available time or current plant state; and if an event of higher priority is in progress, the operator should attend to the more important matters prior to implementing a yellow path function restoration guideline. In the prioriti

..

.. .. . -.

.'.

recovery procedure in effect are still applicable and should be monitored by the operator. This concurrent procedure usage should NOT cause the operator any difficulties since yellow path procedures are only performed when adequate time exist For example, if the operator is in Es-1.1 (SI Termination) and decides to implement FR-H.5 because of low SG level and NC subcooling is lost while in FR-H.5, the operator should t e k n a t e FR-H.5and implement the action of the ES-1.1 foldout page to reinitiate SI flo DUKE POWER MCGUIRE OPERATIONS TRAINING Yellow Path

-.\, A yellow path does not require immediate operator attention. Frequently, it is indicative of an off-normal andlor temporary condition, which will be restored to normal status by actions already in progress. In other cases, the yellow status might provide an early indication of a developing red or orange condition. The operator is allowed to decide whether or not to implement any yellow path procedur Yellow path implementation is a judgement call based on current plant conditions and time available. If a higher priority event is in progress the operators are expected to attend to the most important event. In the prioritization scheme in the EPs, the optimal recovery procedures (including applicable foldout pages) have priority over the yellow path function restoration procedures. The yellow path procedure can be considered as a supplementary set of actions that were provided to address one parameter being in an off-normal state. The controlling procedure in effect is the optimal recovery procedure that the operator is in when helshe decides there is sufficient time to perform the yellow path procedural actions. While performinu the actions of the yellow path, continuous actions, or foldout paae items of the optimal recovew procedures in effect are still applicable and should be monitored by the operator. This concurrent procedure usage should not cause the operator any difficulties since yellow path procedures are

, . only performed when adequate time exist . . SPDS Normally, the condition of the status trees is continuously monitored and displayed by

.'

the bA,C,:.The OAC Ean be used to validate any off normal alarm and to determine .

which EP to implement. The entire control room crewjs responsi.blefor monitoring the

'. . 'ei SPDS: .: '.' ' e 1 How Long to Monitor Status Trees Monitoring of status trees may be stopped when any of the following are met:

Cold shutdown or TSC concurrenc OR Transition to normal recovery procedure (OP).

.-- OR Transition to Severe Accident Management Guideline (SAMG).

OP-MC-EP-FO FOR TRAINING PURPOSES ONLY REV. 06 Page 17 of 83

1 Pt(s) Given the following conditions on Unit 1:

Mode3 NC System is at 1700 psig and 450 degrees In process of cooling down and depressurizing the NC System Safety Injection has occurred NC Pressure going down in an uncontrolled manner Containment pressure going up in uncontrolled manner Which one of the following describes the proper procedures to mitigate the above? Enter AP/35 (ECCS Actuation DuringPlant Shutdown) and then go to E-0 (Reactor Trip or Safety Injection). Enter E-O and then go to AP/35 Enter AP/35 and then go to AP/34 (Shutdown LOCA) Enter E-0 and then go to E-1 (Loss ofReactoror Secondary Coolant).

Bank Question: 1053 Answer: A 1 Pt(s) Given the following conditions on Unit 1:

Mode3 NC System is at 1700 psig and 450 degrees In process of cooling down and depressurizingthe NC System Safety Injection has occurred NC Pressure going down in an uncontrolled manner Containment pressure going up in uncontrolled manner Which one of the following describes the proper procedures to mitigate the above? Enter AP/35 (ECCSAcfuation During Plant Shutdown) and then go to E-0 (Reactor Trip or Safety Injection). Enter E-0 and then go to AP/35 Enter AP/35 and then go to AP/34 (Shutdown LOCA) Enter E-0 and then go to E-I (Lossof Reactor or Secondary Coolant).

Distracter Analysis:. Correct:

Plausible: Incorrect:

Plausible:

- Incorrect:

Plausible: Incorrect Plausible:

LEVEL: SRO KA: 006 G 2.4.4 (4.0/4.3)

SOURCE: NEW LEVEL OF KNOWLEDGE: ANALYSIS AUTHOR: CWS LESSON: - AP/1/5500/35 Background Document

OBJECTIVES: OP-MC-AP-35 Obj. 1 MC-AP-34 Obj. 1 REFERENCES: AP/1/5500/35 Background Document pages 2-4 AP/l/N5500/35 page 3 AP134 Backdground Document page 2 Ques-1053.doc

1 Emergency Procedures /Plan 2.4.1 Knowledge of EOP entry conditions and immediate action step (CFR: 41.10 143.5 / 45.13)

IMPORTANCE RO SRO .4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry condition (CFR: 41.7 145.7 145.8)

Note: The issue of setpoints and automatic safety features is not specifically covered in the systems sections).

IMPORTANCE RO SRO I 2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolution , (CFR: 41.10 143.5 145.13)

IMPORTANCE RO SRO .4.6 Knowledge symptom based EOP mitigation strategie (CFR: 41.10 143.5 I 45.13)

IMPORTANCE RO SRO .4.7 Knowledge of event based EOP mitigation strategie (CFR: 41.10 143.5 145.13)

IMPORTANCE RO SRO .4.8 Knowledge of how the event-based emergencylabnormal operating procedures are used in conjunction with the symptom-based EOP (CFR: 41.10 I 43.5 I 45.13)

IMPORTANCE RO SRO NUREG-1122, Rev. 2 I

DUKE POWER MCGUIRE OPERATIONS TRAINING CLASSROOM TIME (Hours)

NLO I I

NLOR I LPRO I LPSO I LOR I .5 OBJECTIVES

L:

N L

OBJECTlVE 0 Concerning AP/1(2)/5500/35 (ECCSActuation During Plant x x Shutdown):

1 State the purpose of the AP 1 Recognize the symptoms that would require implementation of the A AP35001 3ven scenarios describing accident events and plant

onditions, evaluate the basis for any caution, note, or ste AP35002 I

OP-MC-AP-35 FOR TRAINING PURPOSES ONLY REV. 00 Page 5 of 11

APll and 2/A15500/035(ECCS Actuation During Plant Shutdown)

INTRODUCTION

AP135 covers operator actions for an ECCS actuation from initial plant conditions below P-I 1 (Safety Injection Block Permissive, less than 1955 psig).

Summay This procedure provides guidance to the operator in responding to the above abnormal conditions. The actions do not defeat any safety functions or prevent the required operational features of any safety system from performing as required. This AP is for use during inadvertent ECCS actuations while shutdown (below P-I 1). Scenarios for inadvertent ECCS actuation above P-I 1 are addressed by direct entry into E- VALID S/I EVENTS: (Since this AP is not for events that require SI flow, early kickouts provide direction to the appropriate procedure)

This AP is not for use involving events that lead to a required ECCS Actuation. There should be no required ECCS actuations in Mode 5. For S/G tube ruptures or steamline breaks from Mode 4 up to P-I 1, direction is given to go to E-0 since AP134 is &written for LOCA's on the NC System. For LOCA's from P-I 1 down to CLA isolation, direction is given to go to E-0. For LOCA's from CLA isolation down through Mode 4, direction is given to go to AP/34. For LOCA's in Mode 5, this AP shouldn't even be entered. These Mode 5 scenarios should use AP/10 and/or AP/19, as appropriate. Note it's likely Mode 3 or 4 LOCA scenarios have symptoms that would lead the crew to AP/34 prior to the ECCS actuation (manual or Containment SI). Even though both APl34 and AP135 would be in effect if an actuation subsequently happened, the crew should stay in AP134 a s a higher priority. However, even if they went to AP135, this would not be a concern since AP/35 would quickly kick out to AP13 Note "SI" is not manually initiated in this AP. If there is a need for "SI",you should not be in this AP because this AP is not for events requiring ECCS actuation. This AP assumes the event is inadvertent actuation. With the plant shutdown and depressurized, refill of the Pzr will occur much quicker, challenging Pzr PORVs, Pzr Safeties, and ND suction reliefs. This point is emphasized to support a course of action contrary to most operator training. IF ONE TRAIN OF

"SI" ACTUATES, DO NOT ACTUATE THE SECOND TRAIN. This would needlessly challenge the mitigating strategy of this A Finally, this AP is not written assuming an inadvertent Phase "6"or NS actuation has occurre Note the title of the AP is specifically ECCS actuation, not ESF actuations in genera ENTRY CONDITIONS Entry to this procedure will occur if there is actuation of ECCS equipment. This includes valid or inadvertent actuations. For valid actuations, this AP will quickly provide direction to go to E-0or AP/34, in the unlikely event that procedure was not already in effec Page 2 of 32 Rev 0

APll and 21A/55001035 (ECCS Actuation During Plant Shutdown)

STEP DESCRIPTION FOR ECCS ACTUATION DURING PLANT SHUTDOWN STEP 1:

PURPOSE:

Cue the operator to monitor the foldout pag DISCUSSION:

The use of a foldout page is unusual for an AP. One was chosen for this AP as a human-factors' consideration. Maintaining critical items on a separate page ensures they are performed in a timely manner. The foldout page contains actions that apply throughout the AP as described in item below:

"NC pump trip criteria". Tripping the NC pumps on low seal D/P or low leakoff ensure the operator is alerted that minimum # I seal operating conditions may be lost during depressurization or Phase " A isolation. Phase "A" isolation closes the NCP seal return valves, reducing seal D/P by approximately 100 PSID. The 200 PSID and 0.2 gpm leakoff setpoints are based on the operating limits stated in the NC Pump Tech Manual (MCM 1201-01-193, Controlled Leakage Seal Reactor Coolant Pump). These limits ensure sufficient seal and pump radial bearing lubrication and cooling. If associated with a spray valve, the spray valve is closed to facilitate pressure control now or later. Even though the pump is secured, seal injection flow is needed to supply cool filtered water to the seals.

REFERENCES:

MCM 1201-01-193, Controlled Leakage Seal Reactor Coolant Pump STEP 2:

PURPOSE:

Direct the operator to the appropriate procedure for real events, since this AP is only for inadvertent ECCS actuation scenarios.

DISCUSSION:

If temperature is below 2OO0F, the ECCS actuation is very likely to be inadvertent, and so the operator would continue on in the AP to recover from the inadvertent actuation.

If temperature is above 200"F, the actuation could have been caused by a event requiring S/I flow (SGTR, steam leak, or LOCA). There is no procedure guidance in AP134 for SGTRs or steam leak, so if one of these events has occurred, then no matter what temperature or mode, direction is given to go to E-0. If the event is a LOCA, then direction is given to go to the appropriate procedure, AP/34 if CLAs isolated, E-0 if unisolate Page 3 of 32 Rev 0

APll and 21N55001035 (ECCSActuation During Plant Shutdown)

If above 200°F,but n ~ n of e the listed events has occurred, then it is an inadvertent actuation, and the operator would continue on with the next ste The actuation isolates cooling water to the letdown Hx, so letdown from ND is isolated since it doesn't isolate automatically on the actuation. This should prevent the possibility of flashing/water hammer concern REFERENCES:

STEP 3:

PURPOSE:

Ensure NV pumps have a suction sourc DISCUSSION:

The ECCS actuation may be partial. Either VCT outlet closing would isolate the VCT, and even without it isolating, if the NV pump flowrate is high, it wouldn't take long to empty the VCT. It is prudent to ensure the FWST suction is aligned, early in the procedure. The other ECCS pump suctions are not checked since their suctions are normally aligned from the FWST.

REFERENCES:

STEP 4:

PURPOSE:

Ensure proper RN System alignment, especially for partiallone train actuation scenarios.

DISCUSSION:

One possible scenario could be "B" RN Train in operation prior to the event. If only an "A" Train actuation occurred, "B" RN Pump would lose its' suction from LLI and discharge path to RC via the "A" Train valves. Without a "B" Train actuation, it wouldn't align to the SNSWP leaving it without a suction or discharge. Therefor, direction is provided to align "B" Train to the pond.

Another possible scenario is just a " B Train actuation, in which case the RN non-essential header could lose flow. Therefor, direction is given to ensure both RN pumps are on, so that

"A" RN Pump will supply the non-essential headers.

An actuation on Unit 1 will isolate Unit 2's "B" Train from the non-essential header on Unit 2.

Direction is given to have Unit 2 start it's " ARN Pump if running NCPs so they don't lose cooling water.

REFERENCES:

Page 4 of 32 Rev 0

L l l l " " L"""r\~,Ir\,l"l" Y"I\II"U I W,", ~ , I " I Y " Y Y I " r t w t N APll/AI5500/35 3 O f 38 Rev. 14 UNIT 1 A C T I O N / E X P E C T E D RESPONSE RESPONSE NOT O B T A I N E D C. Operator Actions

- Monitor foldout pag . Check NC temperature - LESS THAN Perform the following:

200" a. Ensure INV-121 (ND Letdown Control)

is close S/I was initiated due to known S/G tube rupture or steamline break, THEN:

L

- 1) E CLAs have been isolated, THEN leave them isolated in EP ) GO TO EP/I/A/5000/-0 (Reactor Trip or Safety Injection).

. c. Check for symptoms of LOCA:

- Pzr level -GOING DOWN IN UNCONTROLLED MANNER

- NC pressure - GOING DOWN IN UNCONTROLLED MANNER

- Containment pressure - GOING UP IN UNCONTROLLED MANNE LOCA has occurred, THEN:

- 1) prior to isolation of CLAs. THEN

_ -TO EP/l/A/5000/E-0 (Reactor GO Trip or Safety Injection).

-2) in mode 3 after CLAs isolated or in mode 4, THEN GO TO AP/1/A/5500/34 (Shutdown LOCA). Ensure the following valves - OPEN:

- INV-221A (NV Pumps Suct From FWST)

- . INV-222B (NV Pumps Suct From FWST).

DUKE POWER MCGUIRE OPERATIONS TRAINING CLASSROOM TIME (Hours)

-

NLO NLOR LPRO LPSO LOR .5 OBJECTIVES i:1:

OBJECTIVE 0 0 1 Concerning AP/1(2)/5500/34 (Shutdown LOCA): x x State the purpose of the AP-Recognize the symptoms that would require implementation of the A AP34001 2 Given scenarios describing accident events and plant conditions, evaluate the basis for any caution, note, or ste AP34002 OP-MC-AP-34 FOR TRAINING PURPOSES ONLY REV. 00 Page 5 of 11

API1 and 2lAl55OOlO34 (Shutdown LOCA)

INTRODUCTION APl34 provides the actions for protecting the reactor core in the event of a LOCA that occurs during either Mode 3 after the Cold Leg Accumulators are isolated or Mode Summary This procedure provides guidance to the operator in responding to the above abnormal conditions. The actions do not defeat any safety functions or prevent the required operational features of any safety system from performing as required After entering AP/34, the operator will increase charging flow in an attempt to maintain Pzr leve If both Pzr level and RCS subcooling can be restored with normal charging flow, the operator can terminate AP134 and return to the appropriate procedure. If normal charging can't maintain both subcooling and Pzr level, the operator will continue with AP134 to respond to the LOC If the switchover setpoint in the FWST is reached, which could happen at any time during the AP depending on the size of the LOCA, the operator will align the ECCS for cold leg recirculation to maintain coolant flow to the core using Enclosure After reaching and maintaining cold shutdown conditions (NC less than 200°F), the final step of AP/34 instructs the operators and plant engineering staff to evaluate the long term plant statu At this time, the NC System will be cooled by the ND System or cold leg recirculation. If Pzr level could not be restored and maintained and boron precipitation is a concern, a decision to transfer to hot leg recirculation could be made. Other long term recovery actions can also be determined at this tim For a LOCA that occurs during shutdown operation, the two largest concerns are core heat-up and cold overpressurization. An evaluation has been performed to determine that establishing safety injection from one high head SI (NV or NI) pump within ten minutes and flow from a secondhigh head SI pump within 30 minutes WILL successfully mitigate (core heat-up)

for a small break LOCA (less than 6 in diameter). Hence, the CLA's are not required for a shutdown LOCA event. To successfully mitigate cold overpressurization AP134 only starts SI pumps as required.

.ENTRY CONDITIONS Entry to this procedure will occur if there is an uncontrolled decrease in Pzr level or NC subcooling, or Containment Floor and Equipment Sump level increase, while in Mode 3 after the CLA's are isolated or in Mode Page 2 of 62 Rev 0

1 Pt Given the following conditions on Unit 1:

Unit 1 is at 100% powe A, B, and CVL AHU are running Aand CVL AHUs have tripped and will not restart Attempts to start D VL AHU were unsuccessful Average temperature in lower containment for past 365 days has been 105 degree Maintenance indicated it will take two days to repair the VL AHU Containment lower compartment temperature is 126 degrees and stead Which one (1) of the following describes the required Technical Specification actions to address the high containment temperature?

Reference Provided Restore temperature to within limits in 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Reduce temperature to 4 2 5 degrees in 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months No action is required to address high containment temperatur Be in Mode 3 in 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months .

Ques-1054.doc

\

Bank Question: 1054 Answer: C 1 Pt Given the following conditions on Unit 1:

Unit 1 is at 100% powe A, B, and C VL AHU are running A and C VL AHUs have tripped and will not restart Attempts to start D VL AHU were unsuccessful Average temperature in lower containment for past 365 days has been 105 degree Maintenance indicated it will take two days to repair the VL AHU Containment lower compartment temperature is 126 degrees and stead Which one (1) of the following describes the required Technical Specification actions to address the high containment temperature?

Reference Provided Tech Spec 3. Restore temperature to within limits in 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Reduce temperature to 4 2 5 degrees in 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months No action is required to address high containment temperatur . Be in Mode 3 in 14 hour1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months Incorrect:

Plausible: Incorrect:

Plausible: Correct:

Plausible: Incorrect Plausible:

LEVEL: SRO G: SYS 022 G2.1.12 (2.9/4.0)

SOURCE: NEW LEVEL OF KNOWLEDGE Analysis Ques-1054.doc

AUTHOR CWS LESSON: OP-MC-CNT-VUL OBJECTIVES: OP-MC-CNT-VUL Obj. 11 REFERENCES: OP-MC-CNT-VUL pages 15 & 17 Tech Spec 3.6.5 Ques-1054.doc Conduct of Operations (continued)

2.1.9 Ability to direct personnel activities inside the control roo (CFR: 45.5 I 45.12 145.13)

IMPORTANCE RO SRO .1.10 Knowledge of conditions and limitations in the facility licens (CFR: 43.1 145.13)

IMPORTANCE RO SRO .1.11 Knowledge of less than one hour technical specification action statements for system (CFR: 43.2 I45.13)

IMPORTANCE RO SRO w~s*&,.3 (@R: 43.2 143.5 145.3)

IMpQR.TANcE-. ,..Ro;.,z:9

. ~

. . 4:O

,.-SR , ,

'.

i

. 2.1.13 Knowledge acces .

of facility requirements for controlling vital / controlled

... (CFR: 41.10 / 43.5 145.9 / 45.10)

' IMPOR?ANCE RO SRO "' 2.1.14 Knowledge of system stat; criteria which require the notification of plant

' ' 'personne (CFR: 43.5 145.12)

IMPORTANCE RO SRO .1.15 Ability to manage short-term information such as night and standing ofder (CFR: 45.12)

IMPORTANCE RO SRO .1.16 Ability to operate plant phone, paging system, and two-way radi (CFR: 41.10 I 45.12)

IMPORTANCE RO SRO .1.17 Ability to make accurate, clear and concise verbal report (CFR: 45.12 / 45.13)

IMPORTANCE RO SRO NUREG-1122.Rev. 2 2-2

DUKE POWER MCGUIRE OPERATIONS TRAINING OBJECTIVES

-

L

OBJECTIVE R

-

Given a limit and/or precaution associated with an operating X procedure, discuss its basis and applicabilit Describe the available indications of containment air X temperatur i ~ ~ ~ m g ~ ~ e he ~ ~ ~ ~ ~

Containment Ventilation System:

Given the LCO Title, state the LCO (including any COLR X values) and applicabilit For any LCOs that have action required within one hour, X state the actio Given a set of parameter values or system conditions, X determine if any Tech Spec LCO(s) is (are) not met and any action(s) requif6d within one hou .,.

.

.?, Given 3 setof plant parameters or system conditions and

.i X

.r .

- ..

.... b the appropriate Tech Specs, determine required action Discuss the basis for a given Tech Spec LCO or Safety *

Limi * SRO ONLY -

OP-MC-CNT-VUL FOR TRAINING PURPOSES ONLY REV. 23 Page 7 of 65

. . . .', . >, 1 . :

. .. . .

Containment Air Temperature i

..

- ..

3.6.5 1 3.6 CONTAINMENTSYSTEMS 3.6.5 Containment Air Temperature LCO 3. Containment average air temperature shall be: OF and 2 100°F for the containment upper compartment, and °F and s 120°F for the containment lower compartmen . The minimum containment average air temperature in MODES 2, 3, and 4 may be reduced to 60° , Containment lower compartment temperature may be between 120°F and 125°F for up to 90 cumulative days per calendar year provided lower compartment temperature average over the previous 365 days is less than 120°F. Within this 90 cumulative day period, lower compartment temperature may be between 125°F and 135°F for 72 cumulative hour i .

APPLICABILITY: MODES 1.2. . . 3, and ... ,

ACTIONS CONDITION REQUIRED ACTION I COMPLETIONTIME Containment average air Restore containment 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months temperature not within average air temperature to limit within limit . Required Action and Be in MODE hours associatedCompletion Time not me .2 Be in MODE hours

~ ~

McGuire Units 1 and 2 3.6.5-1 Amendment Nos. 184/166

Containment Air Temperature 3. \

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY S R 3.6. Verify containment upper compartment average air 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months temperature is within limit SR 3.6.5.2 Verify containment lower compartment average air 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months temperature is within limits.

_*

McGuire Units I and 2 3.6.5-2 Amendment Nos. 184/166

DUKE POWER MCGUIRE OPERATIONS TRAINING These units are shunt-tripped from the essential power system upon receipt of an SS

\' '

signal. Once the shunt trip occurs, the fan's HVAC panel "ON- OFF" indication and control power is lost. The motors are overload protecte The purpose of the upper containment return air fans is to remove air from the dome area of the Containment to prevent stratification. The fans draw the air from the dome and discharge near the suction of an associated upper containment ventilation uni There are four (4) fans associated with this purpose, two are normally operating. Each fan is interlocked with a corresponding upper containment ventilation unit so that the fans operate in conjunction with their associated upper containment ventilation uni These are driven by 1HP, 3 phase, non-nuclear safety related motors and perform no emergency functions. The motors are overload protecte Objective #8 Each Return Air Fan is provided with a selector switch ("AUTO-START-STOP" pushbutton) on the HVAC panel. Each may be manually started or placed in the auto mode. If in auto, the return air fan will start when its corresponding air handling unit is started. Status indication is provided on the HVAC pane The normal power supply for these units is a 6OOVAC Essential Motor Control Cente Following a Blackout, all units will be sequenced on regardless of switch positio Under these conditions, these fans can not be stopped until the sequencer is rese ..

These units are shunt-tripped-from the'essential power system upon receipt of an SS

.d:: signal and the "ON-OFF" indkation om the HVAC pinel is los . .

.: Lower Containment Ventilation System (VL)

...:. .;,. '+'. The.VL system is regulatoty-r&wired per Technical Specification'surveillance requirements arid performs no'safety-related functions. The VL system is designed to maintain a maximum temperature (120°F) inside the lower Containment compartment during normal operation and a minimum temperature (60°F) during shutdow Per Tech Spec 3.6.5, the average air temperature for Lower Containment shall be

> 100°F and 5 120°F in Modes 1 - NOTES: The minimum containment average air temperature in MODES 2, 3 and 4 may be reduced to 60° . Containment lower compartment temperature may be between 120°F and 125°F for up to 90 cumulative days per calendar year provided lower compartment temperature average over the previous 365 days is less than 120°F. Within this 90 cumulative day period, lower compartment temperature may be between 125°F and 135°F for 72 cumulative hour This temperature range was determined by incorporating the following temperature limits: (1) the lower Containment compartment temperature assumed in the Containment accident analyses, (2) the equipment qualification temperatures, and (3)

temperature requirements for personnel acces OP-MC-CNT-VUL FOR TRAINING PURPOSES ONLY REV. 23 Page 15 of 65

- DUKE POWER MCGUIRE OPERATIONS TRAINING The Lower Containment Weighted Average Temperature (LCWAT) is used by the operator to determine optimum VURV/RN operations. The LCWAT program calculates the LCWAT using only the operating VL units inputs (temperatures associated with idle fans are not used).

It is desirable for the VL system to operate during events such as a small isolable LOCA, small main steam break inside Containment, blackout and LOOP to avoid a rise in Containment pre.ssure such that Containment Spray is unnecessarily actuate Provisions in the design were made such that selected equipment from this system is capable of receiving safety-related 1E powe The VL system consists of four (4) recirculating ventilation units and their associated cooling coils, fans, and associated ductwork. This equipment is located in the annular concrete chambers around the periphery of the lower Containment compartment (Fan Rooms). The temperature in the annulus between the reactor vessel and the primary shield may exceed the maximum average temperature of lower Containment (This temperature may be allowed to reach 135°F without detrimental effects to the installed instrumentation.)

Objective #E Each V L AHU has an "OFF-LOW-HIGH" selector switch on the HVAC panel. The V L fan motors are overload protected and status indication is provided on the HVAC.p,%iiel.Annunclators are provided to.indicate mixed speed operation,

. . ' tj'ansfei'th

. ' ....

emergencgpower; high speed start and high vibration. Bearing temperatures'are monitored by the OAC. The 2A, 2B & 2C V L fan motors are a

. ' ' '$herent

. .. .design.motaPwhich isdesigned.t6:operate at a higher temperature than m e others on Unit 1 &' 2. Therefore, these game V L fan motors have higher

.... bearing temperature alarm setpoints on the'OA(: (see MM-10562). Discharge

.check damper position is provided on the HVAC panel. Each V L AHU has a suction damper control switch ("AUTO-OPEN-CLOSE" pushbutton) on the HVAC panel. Each VL ventilation unit fan has two-speed capability. At high speed the associated fan operates at 1800 rpm and at low speed the fan operates at 900 rp Objective #2 The cooling water supply is from the Containment Ventilation Cooling Water (RV)

system. Nuclear Service Water (RN) through the RV System is the preferred source of cooling water in Modes 1 through 5. In Mode 6 or No Mode, cooling water is not required. Cooling water flow is maintained to the VL ventilation units until the "Phase B" signal is received. Since the cooling water for the VL ventilation units is raw water, fouling of these tubes is a problem. As the fouling of the heat transfer area increases, the efficiency of the cooling coil is decreased, thus increasing the temperature of lower Containmen Each ventilation unit contains an automatic on-line tube cleaning system. This system incorporates individual brushes that are periodically backwashed through the tubes to remove fouling and silt deposits. The periodic backwash is accomplished with a 4-way reversing valve (1RV433, 1 RV434, 1RV435, and 1RV436) and associated control Backwashing occurs based on a predetermined frequency and duration. The cleaning cycle is initiated automatically by a cycle time OP-MC-CNT-VUL FOR TRAINING PURPOSES ONLY REV. 23 Page 17 of 65

1 Pt. Given the following conditions on Unit 1:

In Mode 5 cooling down for a refueling outag The 1BND pump tripped due to an electrical faul The IA ND pump has been started per AP/l/N5500/19 (Loss of ND or ND System Leakage) End. 14 (Startup of ND Pumps)

NC temperature before the pump trip was 150 degrees NC temperature has increased to 207 degree AP/l/N5500/19 (Loss of ND) is in effect The SRO instructs the RO to cooldown to the pre-event temperatur Which one (1) of the following describes the maximum cooldown rate and minimumflow rate allowed to cooldown?

REFERENCES PRO VlDED Maximum cooldown rate of 50 degreedhr and minimum flow rate of 1500 gp . Maximum cooldown rate of 75 degrees/hr and minimum flow rate of 1000 gpm Maximum cooldown rate of 50 degreeslhr and minimum flow rate of 2000 gp Maximum cooldown rate of 75 degreeslhr and minimum flow rate of 1500 gp ,

Bank Question: 7055 Answer: D 1 P Given the following conditions on Unit 1:

In Mode 5 cooling down for a refueling outag The '1B' ND pump tripped due to an electrical faul The '1A ND pump has been started per AP/I/N5500/19 (Loss of ND or ND System Leakage) Encl. 14 (Startup of ND Pumps)

NC temperature before the pump trip was 150 degrees NC temperature has increased to 207 degree AP/I/N5500/19 (Loss of ND) is in effect The SRO instructs the RO to cooldown to the pre-event temperatur Which one (1) of the following describes the maximum cooldown rate and minimum flow rate allowed to cooldown?

REFERENCES PROVlDED AP/I/N5500/19 End 14 DATA BOOKEncl. 4.3, cutve 1.6b Maximum cooldown rate of 50 degreeslhr and minimum flow rate of 1500 gp Maximum cooldown rate of 75 degreeslhr and minimum flow rate of 1000 gpm Maximum cooldown rate of 50 degreeslhr and minimum flow rate of 2000 gp Maximum cooldown rate of 75 degreeslhr and minimum flow rate of 1500 gp ___________._______________

Distracter Analysis:. Incorrect:

Plausible: Incorrect:

Plausible: Correct Plausible:

LEVEL: SRO KA: 000025 AA2.05 (3.1*/3.5*)

Ques-1055.doc

c SOURCE: NEW LEVEL OF KNOWLEDGE: Analysis AUTHOR CWS LESSON: OP-MC-AP-19 OBJECTIVES: OP-MC-AP-19 Obj 2 REFERENCES: AP-19 Background Document Enclosure 14 AP/1N5500019 Enclosure 14 Provided DATA Book Enclosure 4.3 Provided Ques-1055.doc

SUMMARY FOR ENCLOSURE 14, STARTUP OF ND PUMPS This enclosure attempts to get a ND Pump started under various potential plant condition If a loss of VI has occurred, then numerous compensatory actions are needed and rather than complicate this enclosure, a kickout is provided to a separate enclosure for that plant conditio A couple of system checks are performed prior to starting an ND train. ND-35 is check closed to prevent an inadvertent inventory loss (may have been opened as a makeup option). If open, an operator is dispatched to stand by so it can be closed prior to pump start.

A step is provided to leave ND UD in service if the NC System is solid. The setpoint for checking if NC solid is 96% Pzr level, which includes 4% instrument error. Note there still may be some volume above just full indicated level (dome of Pzr), but that amount can't be assumed to be availabl If SI has occurred then control of RN modulating valves is reestablished. Then direction is provided to go the section of the enclosure to start the desired ND Pum In preparation for starting a ND Pump, the local pump discharge is setup 2 turns open to prevent water hammer concerns. Pump support conditions are established (RN & KC)

and ND-35 is closed at this time, if required. ND suction from the loop is aligned, and ND flow bypassing the ND Hx is aligned. The ND recirc valve is de-energized prior to starting pump to prevent any air that may be in the ND Hx from returning to the pump suction.

Several precautions are taken on ND Pump startup addressing voiding concerns. If air entrainment or voiding has occurred, a cue is provided to continue makeup as required, considering void collapse may occur after pump start. Also, a check is made for subcooling. If subcooling can't be restored, FW-27A is aligned open in conjunction with the loop suction valves until after pump start. Once NC System subcooling restored (should happen quickly with the cool FWST water mixed in), FW-27A is closed.

After the ND Pump is started, ND flow is carefully established using ND Pump discharge valve and ND-34, and then flow through the ND Hx is carefully established to maintain NC temperature (considering NC System cooldown limits). A cue is also provided to secure

"feed & bleed" when less than 200°F, if it had been established. Finally, a cue is provided to flush the idle ND train if air entrainment may have occurred on it.

Ques-1055.doc

SUMMARY FOR ENCLOSURE 14, STARTUP OF ND PUMPS This enclosure attempts to get a ND Pump started under various potential plant condition If a loss of VI has occurred, then numerous compensatory actions are needed and rather than complicate this enclosure, a kickout is provided to a separate enclosure for that plant condition.

A couple of system checks are performed prior to starting an ND train. ND-35 is check closed to prevent an inadvertent inventory loss (may have been opened as a makeup option). If open, an operator is dispatched to stand by so it can be closed prior to pump start.

A step is provided to leave ND UD in service if the NC System is solid. The setpoint for checking if NC solid is 96% Pzr level, which includes 4% instrument error. Note there still may be some volume above just full indicated level (dome of Pzr), but that amount can't be assumed to be available.

If SI has occurred then control of RN modulating valves is reestablished. Then direction is provided to go the section of the enclosure to start the desired ND Pump.

In preparation for starting a ND Pump, the local pump discharge is setup 2 turns open to prevent water hammer concerns. Pump support conditions are established (RN & KC)

and ND-35 is closed at this time, if required. ND suction from the loop is aligned, and ND flow bypassing the ND Hx is aligned. The ND recirc valve is de-energized prior to starting pump to prevent any air that may be in the ND Hx from returning to the pump suction.

Several precautions are taken on ND Pump startup addressing voiding concerns. If air entrainment or voiding has occurred, a cue is provided to continue makeup as required, considering void collapse may occur after pump start. Also, a check is made for subcooling. If subcooling can't be restored, FW-27A is aligned open in conjunction with the loop suction valves until after pump start. Once NC System subcooling restored (should happen quickly with the cool FWST water mixed in), FW-27A is closed.

After the ND Pump is started, ND flow is carefully established using ND Pump discharge valve arid ND-34, and then flow through the ND Hx is carefully established to maintain NC temperature (considering NC System cooldown limits). A cue is also provided to secure

"feed & bleed" when less than 2OO0F, if it had been established. Finally, a cue is provided to flush the idle ND train if air entrainment may have occurred on i APE: 025 Loss of Residual Heat Removal System (RHRS)

BBlllTy

.

AA Ability to operate and I or monitor the following as they apply to the Loss of Residual Heat Removal System:

(CFR41.7 145.5 145.6)

AAl.01 RCSIRHRS cooldown rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7 AA1.02 RCS inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9 AA1.03 LPIpumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3 AA1.04 Closed cooling water pumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.8' AA1.05 Raw water or sea water pumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . .6 AA1.06 AA1.07 AA1.08 NotUsed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

NotUsed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

RHR cooler inlet and outlet temperature indicators . . . . . . . . . . . . . . .

NIA NIA 2.9*

NIA NIA I AA1.09 LPI pump switches, ammeter, discharge pressure gauge, flow meter, and indicators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 AAl.10 LPI pump suction valve and discharge valve indicators . . . . . . . . . . . . . 3.1' AAl.11 Reactor building sump level indicators . . . . . . . . . . . . . . . . . . . . . . . .0 AAl.12 RCS temperature indicators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .5 AA1.13 SWS radiation monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .6 AA1 14 Waste tank radiation monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1* AA1.15 Waste tank level gauges and recorders ....................... .1 AA1.16 Service water pump manual switch, flow gauge, running lights, and ammeters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .2 AA1.17 Service water block valve indicators and flow valve controllers . . . . . . . .0'

LPI header cross-connect valve conuoller and indicators . . . . . . . . . . . .

AA1.18 AA1.19 AA1.20 Block orifice bypass valve controller and indicators . . . . . . . . . . . . . . .

HPI pump control switch, indicators, ammeter running 2.6*

2.6'

2.8'

2.4* I lights, and flow meter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.6' 2.5* I AA1.21 Letdown flow indicator . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .5 AA1.22 Obtaining of water from BWST for LPI system . . . . . . . . . . . . . . . . . 2.9' AA1.23 RHR heat exchangers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9 A Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System:

(CFR:43.5 145.13)

-

AA2.01 . Proper amperage of running LPIldecay heat removal/RHR pump(s) . . . . .9 AA2.02 Leakage of reactor coolant from RHR into closed cooling water system or into reactor building atmosphere . . . . . . . . . . . . . . . . .8 AA2.03 Increasing reactor building sump level . . . . . . . . . . . . . . . . . . . . . . . .8 AA2.04

... .~. .Location and isolabilitv of leaks . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3' kmli 9'

AA2.06 Existence of proper RHR overpressure protec[ion . . . . . . . . . . . . . . . . .4'

AA2.07 Pump cavitation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 4.2-17 NUREG-1 122, Rev. 2 I

DUKE POWER CGUIRE OPER TIONS TRAINING

~

NLO NLOR LPRO LPSO LOR .o OBJECTIVES OBJECTIVE 1 Concerning AP/1(2)/5500/19 (Loss of ND OR ND SYSTEM LEAKAGE):

State the purpose of the AP Recognize the symptoms that would require implementation of the A f ,e conditions, evaluate the basis for any caution, note, or step. APi9002 I I OP-MC-AP-19 FOR TRAINING PURPOSES ONLY REV. 00 Page 5 of 1 1

SUMMARY FOR ENCLOSURE 14, STARTUP OF ND PUMPS This enclosure attempts to get a ND Pump started under various potential plant conditions. If a loss of VI has occurred, then numerous compensatory actions are needed and rather than complicate this enclosure, a kickout is provided to a separate enclosure for that plant condition.

A couple of system checks are performed prior to starting an ND train. ND-35 is check closed to prevent an inadvertent inventory loss (may have been opened as a makeup option). If open, an operator is dispatched to stand by so it can be closed prior to pump start.

A step is provided to leave ND L/D in service if the NC System is solid. The setpoint for checking if NC solid is 96% Pzr level, which includes 4% instrument error. Note there still may be some volume above just full indicated level (dome of Pzr), but that amount can't be assumed to be available.

If SI has occurred then control of RN modulating valves is reestablished. Then direction is provided to go the section of the enclosure to start the desired ND Pump.

In preparation for starting a ND Pump, the local pump discharge is setup 2 turns open to prevent water hammer concerns. Pump support conditions are established (RN & KC) and ND-35 is closed at this time, if required. ND suction from the loop is aligned, and ND flow bypassing the ND Hx is aligned. The ND recirc valve is de-energized prior to starting pump to prevent any air that may be in the ND Hx from returning to the pump suction.

Several precautions are taken on ND Pump startup addressing voiding concerns.

If air entrainment or voiding has occurred, a cue is provided to continue makeup as required, considering void collapse may occur after pump start. Also, a check is made for subcooling. If subcooling can't be restored, FW-27A is aligned open in conjunction with the loop suction valves until after pump start. Once NC System subcooling restored (should happen quickly with the cool FWST water mixed in),

FW-27A is closed.

After the ND Pump is started, ND flow is carefully established using ND Pump discharge valve and ND-34, and then flow through the ND Hx is carefully established to maintain NC temperature (considering NC System cooldown limits).

A cue is also provided to secure "feed & bleed" when less than 200"F, if it had been established. Finally, a cue is provided to flush the idle ND train if air entrainment may have occurred on it.

Ques-1055.doc

.--

MNS LOSS OF ND OR ND SYSTEM LEAKAGE PAGE NO AP11/A/5500/19 UNIT 1 - Enclosure 14 - Page 11 of 23 Startup of ND Pumps ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED

- 37. Have operators open, backseat, and lock IND-24 (A ND Pump Discharge Isol).

3 Throttle the following as necessary to maintain stable NC temperature:

- 1ND-29 (A ND Hx Outlet)

- 1ND-34 (A & B ND Hx Bypass)

.<-

to control NC temperatur " I J5""

p .,;i?j

..

.ir r

- 4 IF AT ANY TIME cooldown is requtib;:!Q+*:.,;

THEN REFER TO Unit 1 Data Book Curve"".."

1.6 b (Heatup and Cooldown Limits for LTOP).

MNS APil/A/5500/19 UNIT 1 LOSS OF ND OR ND SYSTEM LEAKAGE Enclosure 14 - Page 12 of 23 StartuD of ND Pumos PAGE N of 152 Rev. 15 I ACTION/EXPECTED RESPONSE

- a. Initiate cooldownto 200' F on core exit \

Tic' t b. WHEN NC temperature is less than k?\

ZOOo F on core exit Tic's. THEN perform the following:

1$

CAUTION Failure to stop makeup to NC Systemlprior to closing Pzr PORVs may cause low temperature over pre sure concer ) Stop or reduce makeup to NC Syste ib'

2) E PORV (s) not required open as i vent path, AND makeupto NC i

System stopped, THEN: F

.I I

- a) Close PZR PORV i

- b) Place closed Pzr PORVs in

"AUTO

- 3) Control ND flow to maintain NC ure ND flow greater than 1500 GPM.

- 44. Dispatch operator to reclose breaker 1EMXA - F12B (1A ND Pump & Hx Miniflow lsol Motor (1ND-68A)).

- 45. Ensure 1ND-68A(A ND Pump & A Hx Miniflow) remains close UNIT 1 OP/l/A/6100/22 Enclosure Curve 1.6 b Heatup and Cooldown Limits for LTOP Valid Thru Cycle 16 and with a Maximum of One NI or One NV Pump Capable o f Injection Heatup and cooldown instantaneous rate (P1246 - P1249) and rate of change over the last hour of operation should be reviewed for compliance prior to changing temperature ranges, and at least every 30 minutes during heatup and cooldown.

Table 1: With All NCPs OFF (Note 1)

Temperature Range

ND Hx Outlet Temp

= Lowest WR T-cold Table 2: With One or Two NCPs Running (Note 2)

Temperature Range Indicating Cooldown Rate ("Flhr) Heatup Rate ("F/hr)

("F) Temperature 40 50

-.

8 9 - 119 Lowest of:

119-149 ND Hx Outlet Temp 60 50

> 149 Lowest WR T-cold 75 50 Note 2: Minimum temperature required to operate RCPs is 89°F. If indicating temperatures fall below 89"F, restore to > 89°F within 15 minutes or immediately depressurize by adjustment of charging and letdown and open 1 PORV.

Table 3: With Three or Four NCPs Running (Note 3)

Temperature Range Indicating Cooldown Rate ("F/hr) Heatup Rate ("F/hr)

("F) Temperature 91 - 114 20 50 114- 139 . Lowest WR T-cold 40 50 139 - 164 60 50

> 164 75 50 Note 3: Minimum temperature required to operate 3rd RCP is 91°F. Minimum temperature required to operate 4th RCP is 140°F. If lowest WR Tcold falls below these temperature limits, stop at least one pump and compJ with next lower tier of requirement UNIT 1

UNIT 1 OPI1I N 6 100/22 Enclosure Curve 1.6 b Heatup and Cooldown Limits for LTOP Valid Thru Cycle 16 and with a Maximum of One NI or One NV Pump Capable of Injection Heatup and cooldown instantaneous rate (P1246 - P1249) and rate of change over the last hour of operation should be reviewed for compliance prior to changing temperature ranges, and at least every 30 minutes during heatup and cooldow Table 1: With All NCPs OFF (Note I)

1 Temper?? Range Indicating Tempefature Cooldown Rate ("F/hr) Heatup Rate ("Fhr)

less than 89 Lowest of: > 2.75 sq. inch vent 89-119 - ND Hx Outlet Temp 40 50 119-149 . Lowest WR T-cold 60 50

> 149 75 50 Note 1: Minimum temperature to operate pressurized is 89°F. If indicating temperatures fall below 89"F, restore to > 89°F within 15 minutes or immediately depressurize by adjustment of charging and letdown and open 1 POR Table 2: With One or Two NCPs Running (Note 2)

ND Hx Outlet Temp ing emperaures a Table 3: With Three or Four NCPs Running (Note 3)

Lowest WR T-cold UNIT 1

Bank Question: 1057 Answer: C 1 P As a result of thunderstorms Unit 2 has experienced a Loss of Offsite Power and Reactor trip. E-0 (Reactor Trip or Safety Injection) was implemented and the crew has transitioned to ES-0.1 (Reactor Trip Response).

The SRO asks the RO to check NC temperature Which one (1) of the following would the RO use to describe the response of the NC system? NC Tave STABLE or trending to 557 degrees NC T hots STABLE or trending to 553 degrees NC T colds STABLE or trending to 557 degrees NC Tave STABLE or trending to 553 degree _____________---.___------.------

Distracter Analysis:. The reactor coolant pumps have tripped in this scenario. Tave is only checked if the NC pumps are on. T hot should be increasing initially on the establishment of natural circulation. T colds will go to 557. 553 degrees is a commonly used number for steam dump P-1 Incorrect:

Plausible: Incorrect:

Plausible: Correct:

.~

Plausible: Incorrect Plausible:

LEVEL: SRO KA: 00056AA2.32 31 3)

SOURCE: NEW LEVEL OF KNOWLEDGE: Comprehension AUTHOR CWS LESSON: AP/09 Background Document

OBJECTIVES:

REFERENCES: AP/09 Background Document pages 3 & 4 EP/1/N5000/ES-0.1 page 3 Ques-1057.doc

1 P As a result of thunderstorms Unit 2 has experienced a Loss of Offsite Power and Reactor trip. E-0 (Reactor Trip or Safety Injection)was implemented and the crew has transitioned to ES-0.1 (Reactor Trip Response).

The SRO asks the RO to check NC temperature Which one (1) of the following would the RO use to describe the response of the NC system? NC Tave STABLE or trending to 557 degrees NC T hots STABLE or trending to 553 degrees NC T colds STABLE or trending to 557 degrees NC Tave STABLE or trending to 553 degrees.

Ques-1057.doc

APE 056 Loss of Offsite Power

- AA2.30 AA2.31 Switch gear room cooling unit run indicator ....................

Ventilation supply fan and run indicators for the ser- .2 AA2.33 ESF channels. A and B breaker-trip alarms. indicators and bus voltage indicators ............................... 3.6? 3.7? I AA2.34 Rodbottomlights .................................... .2 AA2.35 Reactor trip alarm .................................... .1 AA2.36 Turbine stop valve indicator .............................. .1 AA2.37 EDlG indicators for the following: voltage. frequenc load. 1oad.status. and closure of bus tie breakers ................. 3.7* AA2.38 Load sequencer status lights .............................. 3.7' AA2.39 Safety injection pump ammeter and flowmeter . . . . . . . . . . . . . . . . . . 3.5* AA2.40 Service water pump ammeter and flowmeter . . . . . . . . . . . . . . . . . . . .4 AA2.41 HVAC chill water pump run and alarm indicators . . . . . . . . . . . . . . . . 2.3' 2.3* I AA2.42 Occurrence of a reactor trip .............................. .1 AA2.43 Occurrence Of a turbine trip .............................. .1 AA2.44 Indications of loss of offsite power ......................... .5 AA2.45 Indicators to assess status of ESF breakers (tripped/

not-tripped) and validity of alarms (falselnot-false) . . . . . . . . . . . . . . . .9 AA2.46 That the EDlGs have started automatically and that the bus tie breakers are closed ............................... .4 AA2.47 Proper operation of the EDlG load sequencer ................... .9 AA2.48 Reactor coolant temperature. pressure. and PZR level following a power outage transient .......................... .4 AA2.49 Nonessential equipment to be secured to avoid overload of EDIGs . . . . . .4 AA2.50 That load and VAR limits. alarm setpoints. frequency and voltage limits for ED/Gs are not being exceeded . . . . . . . . . . . . . . .1

.- AA2.51 -T. (core. heat exchanger. etc.) ........................... 3.3* 3.4*

AA2.52 PZR level required for a given power level .................... 2.6* 2.8*

AA2.53 Status of emergency bus under voltage relays . . . . . . . . . . . . . . . . . . . .2 AA2.54 Breaker position (remote and local) ......................... .0 AA2.55 Subcooled margin monitors .............................. .9 AA2.56 RCST-ave . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6, AA2.57 RCS hot-leg and cold-leg temperatures ....................... .1 AA2.58 Air compressors (indicating lights) ......................... .6'

AA2.59 Gland seal pressure gauge ............................... .6

.

AA2.60 . MSIVopen . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.7* 2.9, AA2.61 Condensatepump . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7 AA2.62 Breaker for feedwater pumps ............................. .9'

AA2.63 Feedwater heater drain pump breaker trip ..................... .5 AA2.64 .Circulating water pump switch ............................. .7 AA2.65 Screen wash pump . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7 AA2.66 CVCS charging flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .4 AA2.67 Seal injection flow (for the RCPs) .......................... .1 AA2.68 CVCSletdownflow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9 AA2.69

..

Valve position . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3* 2.5*

AA2.70 Reactor building CCW temperature ......................... .2 AA2.71 Turbine service water heat exchange ........................ .7 AA2.72 Auxiliary feed flow ................................... .3 AA2.73 PZR heater on/off . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .6 AA2.74 PORV position . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7 AA2.75 CVCSmakeup . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .2 AA2.76 Reactor makeup water pump (running) ....................... .6

...

AA2.77 Auxiliary feed pump (running) . . . . . . . . . . . . . . . . . . . . . . . . . . . . .4 AA2.78 Bus voltmeters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .0 4.2-39 .

NUREG-1 122. Rev 2 I

DUKE POWER MCGUIRE OPERATIONS TRAINING CLASSROOM TIME (Hours)

NLO NLOR LPRO LPSO LOR .5 OBJECTIVES N

L bOW OBJECTIVE 0 0 0 Concerning AP/l(2)/5500/09 (Natural Circulation):

State the purpose of the AP Recognize the symptoms that would require implementation of the A AP09001 I I

Given scenarios describing accident events and plant conditions, evaluate the basis for any caution, note, or ste AP09002 OP-MC-AP-09 FOR TRAINING PURPOSES ONLY REV. 00 Page 5 of 11

L -

MNS EPIl/AI5000/ES-0 1 UNIT 1 REACTOR TRIP RESPONSE

- PAGE NO 3 Of 41 Rev 20 ACTION/EXPECTED RESPONSE I RESPONSE NOT O B T A I N E D Check NC temperatures: Perform the following based on plant condition ..-. .. -. .-.

- E any NC pump on, THEN check NC T-Ave - STABLE OR TRENDING TO E temperature less than 557" F AND 557' going down, THEN:

OR - . 1) Ensure all steam dump valves close E all NC pumps off, THEN check NC T-Colds - STABLE OR TRENDING TO 2) E MSR "RESET" light is dark, 557" T":

- a) Depress "SYSTEM MANUAL"

- b) Depress "RESET"

- 3) Ensure all SM PORVs close ) IF any SM PORV can not be closed, T":

- a) Close its isolation valve

- b) SM PORV isolation valve can not be closed, THEN dispatch operator to close SM PORV isolation valv ) Ensure SIG blowdown is isolated 6) E cooldown continues, THEN control feed flow as follows:

a) E SIG NIR level is less than 11% in all SIGs, THEN throttle feed flow to achieve the following:

- Minimize cooldown

- Maintain total feed flow greater than 450 GP b) WHEN N/R level is greater than 11% in at least one SIG, THEN throttle feed flow further to:

- Minimize cooldown

- Maintain at least one SIG NIR level greater than 11O (RNO continued on next page)

AP/l82/A!5500/09 (Natural Circulation)

STEP DESCRIPTION FOR AP109 STEP 1:

PURPOSE:

Ensure adequate secondary heat sink.

DISCUSSION:

One of the design requirements for natural circulation flow in the NC System is to have a heat sink.

To ensure the SIGs are maintained as a heat sink, feed flow to &the SIGs is ensured. This can be from the CMICF system or the CA System. If feed flow is not present, direction is given to use CA, since this system can typically be established much quicker than CMKF.

SIG NR level greater than 11% is used to indicate the water level just in the narrow range (MCC-1552.08-00-0208, EP setpoint file). Water level in the NR ensures the tube bundles are covered. If NR indication is not met, direction is given to maintain feed flow greater than 450 GPM until greater than 11%. The setpoint of 450 GPM ensures decay heat removal via steam release and a net inventory gain in the SIGs. Note: 450 GPM is conservative since this amount of flow covers the maximum amount needed decay heat removal and NC pump heat, and of course there is no NC pump heat during natural circulation.

REFERENCES:

MCC-1552.08-00-0208, EP setpoint file STEP 2:

PURPOSE:

Provide the parameter values necessary to indicate natural circulation is occurring. If it's not, direction is given to increase dumping steam in an attempt to establish it.

DISCUSSION:

The parameters used to verify natural circulation are:

NC Subcoolina > 0-F Although two-phase and reflux boiling are also forms of natural circulation, the preferred form is subcooled natural circulation. It is the most efficient and results in the lowest core temperatures given everything else equa Page 3 of 7 Rev 0

AP/l&21A/5500/09 (Natural Circulation)

SIG Dress - stable or qoinu down Natural circulation cooling is established when the heat generation rate in the core equals the heat transfer rate from the core to the NC System, to the SIGs, and out of the SIGs. If SIG pressures are INCREASING, then the heat transfer rate out of the SIGs is not sufficient, and core temperatures will be increasin NC T-Hots - Stable or Goinq Down When forced NC flow is lost, it takes a few minutes (5 - 10 minutes) for natural circulation flow to set up. During this time, T-Hots are increasing, as expected. T-Colds are relatively stable since they are tied to SIG pressure, and the assumption is S/G pressure is stable unless a cooldown is taking place. Once the delta-T develops (T-Hot increase) sufficiently for driving head for natural circulation flow, T-Hot should no longer be increasing. From this point forward, T-Hots should be going down as decay heat drops off, or be stable as the decay heat curve levels off.

After the initial time period for natural circulation to develop, if T-Hots are going up this means the heat removal from natural circulation is less than the decay heat generation rate, or in another words, inadequate nat. circ for whatever reason.

Core exit T/Cs - Stable or Goinq Down When forced NC flow is lost, it takes a few minutes (5 - 10 minutes) for natural circulation flow to set up. During this time, Core Exit TICS are increasing, as expected. T-Colds are relatively stable since they are tied to SIG pressure, and the assumption is SIG pressure is stable unless a cooldown is taking place. Once the delta-T develops (Core Exit T/Cs increase) sufficiently for driving head for natural circulation flow, Core Exit T/Cs should no longer be increasing. From this point forward, Core Exit TICS should be going down as decay heat drops off, or be stable as the decay heat curve levels off.

After the initial time period for natural circulation to develop, if Core Exit TlCs are going up this means the heat removal from natural circulation is less than the decay heat generation rate, or in another words, inadequate nat. circ for whatever reason.

NC T-Colds - A t Saturation Temperature For SIG Pressure The following is an excerpt from the ERGS Generic Issues section concerning T-Colds: "The cold leg temperature readings can be used as additional verification that heat removal through the steam generators is occurring. The loop T-Cold readings in active loops are quite sensitive to changes in heat transfer rates from the reactor to the secondary sides of the steam generators. Actual test have shown that loop T-Colds follow almost exactly the steam generator pressure with minimal time lag".

In another words, if T-Colds are not following steam generator pressure, it is likely heat removal from the SIGs is not occurring. If this is the case, the core is not being adequately cooled. To facilitate determining whether NC T-Colds are at saturation temperature for SIG pressures, a graph is provided to correlate the two, with a band provided to allow for instrument inaccuracie Page 4 of 7 Rev 0

1 P Radwaste is in the process of releasing WGDT A.1EMF -36 L is inoperable due to PM. Trip 2 is received on OEMF-50 (Waste Gas Discharge). The gaseous waste release is secured as a result of IWG-160 closing. Radwaste MIIS the control room SRO and reports OEMF-50 has been purged and is ready to reinitiate the releas Which one (1) of the following describes the actions of the control room SRO? The SRO can authorize up to two (2) restarts without re-samplin The SRO has Radwasteterminate existing GWR paperwork, and generate new paperwor The SRO can authorize one (1) restart without re-samplin The SRO can authorize Radwaste to jumper control actions of OEMF-50, restart release and take grab samples once per four (4) hours during release.

Ques-1058.doc

Bank Question: 1058 Answer: B 1 P Radwaste is in the process of releasing WGDT '4'. 1EMF -36 L is inoperable due to PM. Trip 2 is received on OEMF-50 (Waste Gas Discharge). The gaseous waste release is secured as a result of IWG-160 closing. Radwaste calls the control room SRO and reports OEMFdO has been purged and is ready to reinitiate the releas Which one (1) of the following describes the actions of the control room SRO? The SRO can authorize up to two (2) restarts without re-samplin The SRO has Radwaste terminate existing GWR paperwork, and generate new paperwor The SRO can authorize one (1) restart without re-samplin The SRO can authorize Radwaste to jumper control actions of OEMF-50, restart release and take grab samples once per four (4) hours during releas _._____________.___________

Distracter Analysis:. Incorrect: Would be correct if both EMF 36 and 50 were operable Plausible: Correct:

Plausible: Incorrect:

Plausible: Incorrect Plausible:

LEVEL: SRO KA: 00060 G 2.3.8 (2.3/3.2)

SOURCE: NEW LEVEL OF KNOWLEDGE: Comprehension AUTHOR CWS LESSON: OP-MC-WE-RGR OBJECTIVES: OP-MC-WE-RGR Obj 5 Ques-1058.doc

REFERENCES: OP-MC-WE-RGR page 13 Radiation Control 2.3.1 Knowledge of 10 C m 20 and related facility radiation control requirement (CFR: 41.12 143.4. 45.9 145.10)

IMPORTANCE RO SRO .3.2 Knowledge of facility ALARA progra (CFR: 41.12 143.4 145.9 145.10)

IMPORTANCE RO SRO .3.3 Knowledge of SRO responsibilities for auxiliary systems that are outside the control room (e&, waste disposal and handling systems).

(CFR: 43.4 145.10)

IMPORTANCE RO SRO .3.4 Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorize (CFR: 43.4 145.10)

IMPORTANCE RO SRO .3.5 Knowledge of use and function of personnel monitoring equipmen (CFR 41.11 145.9)

IMPORTANCE RO SRO .3.6 Knowledge of the requirements for reviewing and approving release permit (CFR: 43.4 145.10)

IMPORTANCE RO SRO .3.7 Knowledge of the process for preparing a radiation work permi (CFR: 41.10 I45.12)

IMPORTANCE RO SRO 3.3 I?.@gg 2.3.9 Knowledge of the process for performing a containment purg (CFR: 43.4 145.10)

IMPORTANCE RO SRO NUREG-1122,Rev. 2 . I

NLO NLOR LPRO LPSO LOR NIA NIA .0 OBJECTIVES OBJECTIVE State the purpose of the Radiological Gaseous Release _ _ _ ~ ~

Given a completed GWR, state the recommended release rat ~ ~~

State what responsibility the Control Room SRO is accepting when he signs to authorize a releas ~

Given a completed GWR, state the proper EMF to be used for the releas Evaluate plant parameters to determine any abnormal system conditions that may exist Concerning the Selected Licensee Commitments (SLC)

related to Gaseous Waste Releases; Given the SLC Manual, discuss any commitments and their applicabilit For any commitments that have action required within one hour, state the actio Given a set of parameter values or system conditions, determine if any commitment is (are) not met and any action(s) required within one hou Given the SLC Manual, discuss the basis for a given commitmen * SRO only OP-MC-WE-RGR FOR TRAINING PURPOSES ONLY REV. 6 Page 5 of 25

- DUKE POWER

-

MCGUIRE OPERATIONS TRAINING 3.0 GASEOUS RELEASES The three types of releases discussed in the section are:

Waste Gas Decay Tank (WG)

Containment Air Release (VQ)

Containment Purge (VP) Waste Gas Decay Tank Release 3.1.1. Limits and Precautions If a WGDT will be released, then ensure that bank is not in-service. If an in-service tank must be removed from service in order for a release to be made, then the change must be made per OP/O/N6200/18, Waste Gas Operatio Neither Shutdown (SID) Tank can be in service when making a release, due to the release flowpath which must be use For the release of a S/D Tank, it must first be transferred to a WGDT (A, B, C, D, E, F)

per OP/O/N6200/18 (Waste Gas Operation). The release can then be made following this procedur The Unit 1 Auxiliary Building Ventilation System or Unit 1 Fuel Building Ventilation System should be in service during a release to ensure that the gas leaves the unit vent completel No release will be made without proper verification of flow rate. OWGLP6140 (WG Disch Flow Loop) is the normal instrument for verifying flow. If inoperable, flow can be monitored using the decay tank pressure (for the tank being released) vs. release tim Bulk hydrogen or nitrogen cannot be added to the waste gas system while releasing a tan Based onmutual agreement between MNS Radiation Protection Manager, General Office Protection Manager, and MNS Radwaste, if EMF-36 and EMF-50 are operable, then releases interrupted by EMF Hi Rad Discharge Trips may be reinitiated up to a maximum of two times (total of 3 attempts) without resampling before terminating release procedure. If EMF-36 or EMF-50 are inoperable only one release attempt shall be mad If EMF-36 and EMF-50 are inoperable, RP management approval is required to make a WG releas Immediately after each release purge EMF-50. Immediately subsequent to purge, reposition and lock appropriate valves per applicable enclosur OP-MC- WE-RGR FOR TRAINING PURPOSES ONLY REV. 6 Page 13 of 25

Y

/

1 Pt. Given the following conditions on Unit 1:

SGTR in the 1AS/G E-0(Reactor Trip or Safety lnjecfion) complete E-3 (Steam Generator Tube Rupture) implemente Cooldown is secured due to operator exceeding Main Steam Isolation set poin Which one (1) of the following describes how the operator continues to cooldown? Go to Bypass Interlock on steam dumps and continue cooldown with steam dump Reset Main Steam Isolation, open MSlVs and continue cooldown with steam dump Reset Main Steam Isolation, and PORVs and continue cooldown using PORVs in manua Reset Main Steam Isolation, and PORVs and continue cooldown using PORVs in automati Bank Question: 1060 Answer: C 1 Pt. Given the following conditions on Unit 1:

SGTR in the '?A' S/G E-0 (Reactor Trip or Safety Injection) complete E-3 (Steam Generator Tube Rupture) implemente Cooldown is secured due to operator exceeding Main Steam Isolation set poin Which one (1) of the following describes how the operator continues to cooldown? Go to Bypass Interlock on steam dumps and continue cooldown with steam dump Reset Main Steam Isolation, open MSlVs and continue cooldown with steam dump Reset Main Steam Isolation, and PORVs and continue cooldown using PORVs in manua Reset Main Steam Isolation, and PORVs and continue cooldown using PORVs in automati ____________________------.-----

Distracter Analysis:. Incorrect:

Plausible: Incorrect:

Plausible: Correct:

Plausible: Incorrect Plausible:

LEVEL: SRO KA: 0041 G2.4.20 (3.3/4.0)

SOURCE: NEW LEVEL OF KNOWLEDGE: Memory

AUTHOR: CWS LESSON: OP-MC-EP-E3 OBJECTIVES: OP-MC-EP-E3 Obj 4 REFERENCES: OP-MC-EP-E3 pages 75,77,79 EP/IA/5500/E-3 page 19-21 Ques-1060.doc

\ Emergency Procedures /Plan (Continued)

2.4.18 Knowledge of the specific bases for EOP (CFR: 41.10 / 45.13)

IMPORTANCE RO SRO .4.19 Knowledge of EOP layout, symbols, and icon (CFR. 41.10 I45.13)

IMPORTANCE RO SRO 3 . 1 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions including:

1. Reactivity control 2. Core cooling and heat removal 3. Reactor coolant system integrity 4. Containment conditions 5. Radioactivity release control.

. ~ .

(CFR: 43.5 / 45.12)

IMPORTANCE RO SRO .4.22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operation (CFR: 43.5 / 45.12)

IMPORTANCE RO SRO .4.23 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operation (CFR: 41.10 / 45.13)

IMPORTANCE RO SRO .4.24 Knowledge of loss of cooling water procedure (CFR: 41.10 / 45.13)

IMPORTANCE RO SRO .4.25 Knowledge of fire protection procedure (CFR: 41.10 / 45.13)

IMPORTANCE RO SRO NUREG-I 122, Rev. 2 I

DUKE POWER MCGUIRE OPERATIONS TRAINING NLO NLOR LPRO LPSO LOR .0 N N L L L L L P P O OBJECTIVE O O R S R R O 0 Explain the purpose for each procedure in the E-3 serie X X EPE3001 3iscuss the entry and exit guidance for each procedure in the X X E-3 serie EPE3002 Discuss the mitigating strategy (major actions) of each X x x xocedure in the E-3 serie EPE3003 Diskuss the basis for anv note: caution or step for each lx x x

~ o c .e d..,u r e : i n ~ t ~i ees~. . ~ ~ ~ s e

~:.,... :

EPE3004 Given the Foldout page, discuss the actions included and the X x x basis for these action EPE3005 Siven the appropriate procedure, evaluate a given scenario X x x describing accident events and plant conditions to determine any rkquired action and its basi EPE3006 Discuss the time critical task@) associated with the E-3 series X x x xocedures including the time requirements and the basis for these requirement EPE3007 OP-MC-EP-E3 FOR TRAINING PURPOSES ONLY REV. 05 Page 7 of 403

DUKE POWER MCGUIRE OPERATIONS TRAINING STEP 16 WHEN P-11 PRESSURIZER SI1 BLOCK PERMISSIVE status rl light (19-18) lit, THEN perform the following:

PURPOSE: To prevent MSlV closure on low steamline pressure during controlled NC cooldow BASIS: An automatic protection feature is provided to close the MSlVs when steam pressure approaches the Low Pressure Steamline Isolation setpoint. In Step 17, the operator is instructed to dump steam from the intact S/Gs, which is expected to reduce their pressure below this setpoint. If automatic isolation occurred, steam flow to the condenser would be terminated requiring the operator to continue the cooldown by dumping steam to the atmosphere. In addition to delaying recovery, this would raise the radiological releases and reduce feedwater suppl NOTE 1 NC pump trip criteria on subcooling does not apply after starting a controlled cooldow NOTE 2 After the Lo Pressure Steamline Isolation signal is blocked, maintaining steam pressure negative rate less than 2 psig per second will prevent Main Steam Isolatio PURPOSE: To alert the operator of the followin . This particular NC pump trip criteria does not apply after starting a controlled cooldown. All other trip criteria remain in effec . The potential for inadvertent steamline isolation during the subsequent S/G depressurizatio BASIS: The NC pump trip criteria is based on subcooled conditions not applicable during the controlled cooldow An automatic protection feature is provided to close the MSlVs when the steam pressure rate signal is exceeded. In the following step, the operator is instructed to dump steam from the intact S/Gs, which may result in exceeding the rate setpoin OP-MC-EP-E3 FOR TRAINING PURPOSES ONLY REV. 05 Page 75 of 403

DUKE POWER MCGUIRE OPERATIONS TRAINING STEP 17 Initiate NC System cooldown:

n PURPOSE: To establish sufficient subcooling in the NC so the primary system will remain subcooled after pressure is decreased to stop primary-to-secondary leakag BASIS: The principal goal of E-3 is to stop primary-to-secondary leakage and to establish and maintain sufficient indications of adequate coolant inventory. These indications include the following: A Pzr level indication used to trend coolant inventor . NC subcooling used to ensure the indicated Pzr level is reliabl This step is designed to establish sufficient subcooling in the NC so the primary system will remain subcooled after NC pressure is lowered in subsequent steps to stop primary-to-secondary leakag The pressure of the intact S/Gs must be maintained less than the pressure of the ruptured S/Gs in order to maintain NC subcooling. Since flow from the ruptured S/G should be isolated, this pressure differential is established by dumping steam only from the intact S/Gs. Steam dump to the condenser is preferred to minimize radiological releases and conserve feedwater supply. However, the PORVs on the intact S/Gs provide an alternative steam release pat If no intact S/G is available, then an exit transition is provided to ECA-3.1, SGTR With Loss Of Reactor Coolant - Subcooled Recovery Desire NC cooldown should proceed as quickly as possible and should not be limited by the 1OO°F/hr Technical Specification limit. Integrity limits should not be exceeded since the final temperature will remain above 350° The table provides for 20°F subcooling for each pressure range. Core exit T/Cs are used because they provide input for S/I termination and reinitiation. The 20°F subcooling is provided as operating margin to accommodate fluctuation in NC temperature, perturbations in ruptured S/G pressure, interpolation between listed ruptured S/G pressures, and overshoot during NC depressurizatio OP-MC-EP-E3 FOR TRAINING PURPOSES ONLY REV. 05 Page 77 of 403

DUKE POWER MCGUIRE OPERATIONS TRAINING STEP 17 (CONTINUED)

c1 The preferred cooldown method is to dump steam to the condenser at max rate while attempting to avoid main steam isolation. If the condenser is not available or steam dump to the condenser is not possible, the SG PORVs should be used per the RNO for step 17e. Any delay in initiating cooldown can lead to ruptured S/G overfill. If main steam isolation occurs, while using S/G PORVs, it will have little impact on dumping of steam. The PORVs can be quickly reset and cooldown reinitiated. Although depressurizing at rate specified (2 psig/sec) is very close to rate that would give isolation, it is justified to push rate as hard as possible (when using S/G PORVs).

Ensuring steamline Isolation is blocked using NC PORV will prevent any delay in dumping steam with S/G PORV If any intact SG PORV cannot be opened from the Control Room, local operation at the valve is directed. The PORVs should be opened fully using the valve handwheel not the manual loader in the doghouse.

.d OP-MC-EP-E3 FOR TRAINING PURPOSES ONLY REV. 05 Page 79 of 403

MNS STEAMGENERATORTUBERUPTURE PAGE N EP/1/A/5000/E-3 19 of 57 Y

UNIT 1 Rev. 12

-

A C T I O N / E X P E C T E D RESPONSE RESPONSE NOT O B T A I N E D NOTE NC pump trip criteria based on subcooling does not apply after starting a controlled cooldow After the Low Pressure Steamline Isolation signal is blocked, maintaining steam pressure negative rate less than 2 PSlG per second will orevent a Main Steam Isolatio . initiate NC System cooldown as follows:

- a. Determine required core exit temperature based on lowest ruptured S/G pressure:

LOWEST RUPTURED S / G PRESSURE CORE E X I T T/Cs (PSIG) ( O F )

G R E A T E R THAN 1099 520 (519 ACC)

1000 - 1099 508 (507 ACC)

900 - 999 494 (493 ACC)

800 - 899 480 (479 A C C )

100 . 799 463 (462 A C C )

GOO - 699 444 ( 4 4 4 A C C )

500 . 599 423 (422 A C C )

400 . 499 3'36 ( 3 9 5 : ~ l . c )

'ill1 . 309 ?ij2 , 3 5 ; :, :;

280 ~ 219 353 (35 3 X t : )

b. Check condenser available: - T O RNO for Step 1 "C-9 COND AVAILABLE FOR STEAM DUMP" status light (lSl-18)-

LIT

- MSlV on intact S/G(s)- OPEN

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EP/lIA/5000/E-3 20 of 57 UNIT 1 Rev. 12 ACTION/EXPECTEO RESPONSE RESPONSE NOT O B T A I N E D 17. (Continued)

c. Perform the following to place steam dumps in steam pressure mode:

- 1) Place "STM PRESS CONTROLLER" in manua ) Adjust "STM PRESS CONTROLLER" output to equal

"STEAM DUMP DEMAND" signa ) Place "STEAM DUMP SELECT" in steam pressure mod WHEN "P-12 LO-LO TAVG" status light (1 SI-18) lit, THEN place stearn dumps in bypass interloc e. Dump steam from intact S/G(s) to e. Perform the following:

condenser at maximum rate while attempting to avoid a Main Steam - 1) E Pzr pressure is greater than Isolatio PSIG. THEN depressurize to 1900 PSIG using Pzr POR ) Depress "BLOCK on Low Pressure Stearnline Isolation block switche ) Maintain NC pressure less than 1955 PSI ) Ensure Main Steam Isolation rese ) Ensure S/G PORVs rese ) IF any intact SG PORV isolation valve is,closed. AND associated PORV is operable, THEN perform the following:

- a) Open SIG PORV isolation valve(s).

- b) IF isolation valve will not open, IHEN dispatch operator to open isolation valv (RNO continued on next page) MNS STEAM GENERATOR TUBE RUPTURE PAGE N EP/1/N5000/E-3 21 Of 57 Rev. 12 ACTION/EXPECTED RESPONSE RESPONSE NOT O B T A I N E D 8) any intact S/G PORV closed, I _

THEN dump steam as follows, at maximum rate:

a) Dispatch operators to:

- Immediately fully open intact S/G(s) PORVs (at valves).

- Establish communication with control roo b ) Monitor pressures in all S/G(s) to ensure the correct S/G PORVs are locally operated.

I c) IF any intact S/G PORV is unavailable, T m evaluate i using the following to dump steam:

- Run TD CA pum Use steam drains pER EP/liN5000/G-l (Generic Enclosures), Enclosure 19 (S/G Deoressurization Usina

~

I ~"

I .+J Steam drains).

(RNO conlinued on next page)

1 Pt. Given the following:

Both Units operating at 100% powe Atrain RN is operating on both unit Operations Test Group is performing B train RN valve stroke timin SRO is instructed to evaluate the consequences of stroking ORN-2848 (Train 16 and 26 Discharge to RC)

Which one of the following describes the consequences of allowing the technician to test this valve? No consequences due to A Train RN running on both unit . Closing ORN-284B will isolate the RN nonessential header return from Unit Closing ORN-2846 will isolate the RN nonessential header return from Unit Closing ORN-2846 will isolate RV pump discharg Bank Question: 1065 Answer: B 1 P Given the following:

Both Units operating at 100% powe Atrain RN is operating on both unit Operations Test Group is performing Btrain RN valve stroke timin SRO is instructed to evaluate the consequences of stroking ORN-2848 (Train 1B and 2B Discharge to RC)

Which one of the following describes the consequences of allowing the technician to test this valve? No consequences due to A Train RN running on both unit . Closing ORN-2846 will isolate the RN nonessential header return from Unit Closing ORN-2846 will isolate the RN nonessential header return from Unit Closing ORN-2846 will isolate RV pump discharg _______._________________I Distracter Analysis:. Incorrect:

Plausible: Correct:

Plausible: Incorrect:

Plausible: Incorrect Plausible:

LEVEL: SRO KA: G 2.2.3 (3.1/3.3)

SOURCE: NEW LEVEL OF KNOWLEDGE: Comprehension AUTHOR: CWS LESSON: OP-MC- PSS-RN OBJECTIVES: OP-MC-PSS.RN Obj. 8 Ques-1065.doc

REFERENCES: OP-MC-PSS-RN page 67 Ques-1065.doc

...\ I Equipment Control 2.2.1 Ability to perform ere-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivit (CFR: 45.1)

IMPORTANCE RO SRO .2.2 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power level (CFR: 45.2)

IMPORTANCE RO SRO mi6pIsti@Qal:,?

2.2.4 (multi-unit) Ability to explain the variations in control board layouts,

.. .

systems, instrqnentation and procedural actions between units at a facilit ... (CFR 45.1'/ 45.13)

..

. $ IMPORTANCE RO SRO 3.0*

/

.. . 2.2.5 Knowledge of the process for making changes in the hcility as described in the safety analysis repor (CFR: 43.3 I45.13)

IMPORTANCE RO SRO .2.6 Knowledge of the process for making changes in procedures as described in the safety analysis repor (CFR: 43.3 145.13)

IMPORTANCE RO SRO . Knowledge of the process for conducting tests or experiments not described in the safety analysis repor (CFR: 43.3 145.13)

IMPORTANCE RO SRO .2.8 Knowledge of the process for determining if the proposed change, test, or experiment involves an unreviewed safety questio (CFR 43.3 145.13)

IMPORTANCE RO SRO NUREG-1122. Rev. 2 I

DUKE POWER MCGUIRE OPERATIONS TRAINING OBJECTIVES OBJECTIVE

. ..

D e & r i 6 & h e R ~ 'System Flow path ( suction sburce,.essenial and non-essential header alignment and discharge point ) for the following:

Normal operation Operation following a Blackout Operation following a Safety Injection Explain the reason for taking a suction on the low level intak Zoncerning the RN essential and non-essential headers:

List the loads supplied by each header Identify which loads are automatically supplied on a Blackout, Safety injection andlor Phase Explain the reason for NOT isolating the auxiliary building ion-essential header on a Blackout signa Iescribe the operation including any interlocks for the

'ollowing valves:

RN42A ( A B Non Ess Supply Is01 )

RNdOA ( 1718 ) ( A(B) D/GSupply Is01 )

1RNI ( Low Level Intake Isolation )

Engineering Safeguards Modulating Control Valves and Reset Circuitry Siven a parameter associated with the RN system, describe the indications for that paramete v e n a Limit and Precaution associated with the RN System, jiscuss its basis and when it applies.

OF-MC-PSS-RN FOR TRAINING PURPOSES ONLY REV. 31 Page 9 of 99

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DUKE POWER MCGUIRE OPERATIONS TRAINING RN System Basic Layout (5/01/01)

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.I A

OP-MCPSS-RN FOR TRAINING PURPOSES ONLY REV. 31 Page 67 of 99

1 Pt Which one of the following describes the bases for prioritizing Critical Safety Functions (CSF)?

k The CSFs are prioritized to address challenges to the boundaries that protect the general public from exposure to radiatio . The CSFs are prioritized to address design bases accidents that are described in the USFA The CSFs are prioritized to ensure the proper optimal response procedure is implemente The CSFs are prioritized to address challenges to parameters that would affect operation of Engineered Safeguard Features equipment.

.-,

Ques-1067.doc

Bank Question: 1067 Answer: A 1 Pt Which one of the following describes the bases for prioritizing Critical Safety Functions (CSF)?

k The CSFs are prioritized to address challenges to the boundaries that protect the general public from exposure to radiatio The CSFs are prioritized to address design bases accidents that are described in the USFA The CSFs are prioritized to ensure the proper optimal response procedure is implemente The CSFs are prioritized to address challenges to parameters that would affect operation of Engineered Safeguard Features equipmen Distracter Analysis:. Correct:

Plausible: Incorrect:

Plausible: Incorrect Plausible:

LEVEL: SRO KA: G 2.4.22 (3/0/4.0)

SOURCE: NEW LEVEL OF KNOWLEDGE: Memory AUTHOR CWS LESSON: OP-MC- EP-INTRO OBJECTIVES: OP-MC-EP-E-1 Obj. 1 & 3 REFERENCES: OP-MC-EP-INTRO pages 21,27,29 Ques-lO67.doc

- Emergency Procedures /Plan (Continued)

2.4.18 Knowledge of the specific bases for EOP (CFR: 41.10 / 45.13)

IMPORTANCE RO SRO .4.19 Knowledge of EOP layout, symbols, and icon (CFR: 41.10/45.13)

IMPORTANCE RO SRO .4.20 Knowledge of operational implications of EOP warnings, cautions, and note (CFR: 41.10 / 45.13)

IMPORTANCE RO SRO .4.21 Knowledge of the parameters and logic used to assem the status of safety functions including:

1. Reactivity control 2. Core cooling and heat removal 3. Reactor coolant system integrity 4. Containment conditions 5. Radioactivity release contro (CFR: 43.5 I45.12)

IMPORTANCE RO SRO . ..

2.4.23 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operation (CFR: 41.10 / 45.13)

IMPORTANCE KO SRO .4.24 Knowledge of loss of cooling water procedure (CFR: 41.10 / 45.13)

IMPORTANCE RO SRO 3 . 1 2.4.25 Knowledge of fire protection procedure (CFR: 41.10 / 45.13)

IMPORTANCE RO SRO NUREG-1122, Rev. 2 I

DUKE POWER MCGUIRE OPERATIONS TRAINING CLASSROOM TIME (Hours)

NIA NIA .0 I I I I I OBJECTIVES N N L L L P OBJECTIVE O O R IQI R O h - 4 ~ i s t t h e s i xCritical Safety Functions in order of importanc X EPINTROOOI 2 List the two EP's which provide the entry points into the EP X se EPINTR0002

.3 Explain when and howthe CSF Status Trees are evaluate X EPINTR0003 4 Apply the EP Rules of Usage to determine required actions X for a step in an EP that is not satisfied when no contingency action (no RNO column) is provide EPINTR0004 5 Apply the EP Rules of Usage to determine required actions X while performing an EP contingency action when the action cannot be performed or is not successfu EPINTR0005 6 State when Foldout Page actions or transitions are X applicabl EPINTR0006 7 Describe how to determine if sequence is important when X performing subtasks within a step of an E EPINTR0007 8 Discuss the purpose and applicability of Notes and Caution X EPINTROOOS 9 Define the "Constrained Language" terms listed in OMP 4-3, X Use of Abnormal and Emergency Procedure EPINTR0009 OP-MC-EP-INTRO FOR TRAINING PURPOSES ONLY REV. 02 Page 5 of 63

DUKE POWER MCGUIRE OPERATIONS TRAINING Critical Safety Functions The concept of Critical Safefy Functions (CSF's) came about afler the TMI accident, and has been implemented in the WOG ERG's. The CSF's define parameters that if maintained within specific limits will assure that radioactive materials will not be released from the plant.

I Objective # 1 I The six CSF's, in the order of their priority are: Achievement of Subcriticality (S) Maintenance of Core Cooling (C) Maintenance of the Heat Sink (H) Maintenance of the Reactor Coolant System Integrity (P) Protection of the Containment Boundary Integrity (2) Maintenance of the Reactor Coolant System Inventory (I)

The ERG's address these concerns first, and only afler challenges to the CSF's are handled is it appropriate to turn the attention of the operating crew to the event caus The CSF's address the secureness of the four boundaries that protect the general public from exposure to radioactive materials that could be released during an acciden These boundaries are the 0 Fuel matrixlcladding, 0 NC system pressure boundary, 0 Containment barriers, and 0 Site boundar The ERG's implemented as the Emergency Procedures (EP's), address only the first three of these. The Site Emergency Plan addresses the fourt OP-MC-EP-INTRO FOR TRAINING PURPOSES ONLY REK 02 Page 21 of 63

DUKE POWER MCGUIRE OPERATIONS TRAINING 2.3.3. Function Restoration Procedures I Objective # 3 I Challenges to the Critical Safety Functions (CSF's) are addressed by the Function Restoration Procedures. Each CSF is monitored by Status Tree in order of priority. Monitoring of Status Trees begins either when directed by E-0, or upon any transition from E- The six Status Trees, one for each CSF, are found in procedure F- . F-0.1, SUBCRITICALITY (Procedure series "S") F-0.2,CORE COOLING (Procedure series " C ) F-0.3,HEAT SINK (Procedure series "H") F-0.4, INTEGRITY (Procedure series "F"') F-0.5, CONTAINMENT (Procedure series "2) F-0.6,INVENTORY (Procedure series "I")

1 Objective # 3 I Each Status Tree includes four color-coded challenges to the CSF being monitored. The color coding represents the severity of the challenge, and thus the priority of the required response. The table lists the colors in their order of priority, defines the challenge, and shows the symbols that are used in the CSF Status Tree Color prioritization is important. A red path is addressed before any orange pat Any orange path is addressed before any yellow path. If more than one Status Tree indicates the same color, then priority is addressed by the monitoring order

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of S C H P 2 OP-MC-EP-INTRO FOR TRAINING PURPOSES ONLY REV. 02 Page 27 of 63

DUKE POWER MCGUlRE OPERATIONS TRAINING STATUS TREE PRIORITY IDENTIFICATION

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Color Severity of Line Code Symbol Challenge The CSF is under extreme challeng Red Immediate operator action is require The CSF is under severe challeng Orange Prompt operator action is require vvvvvv The CSF condition is off-normal or E t Yellow satisfie Operator action may be take The CSF is NOT challenqe Green No operator action is neede C There is only one entry point to each Status Tree in F-0. However here are multiple exit points, but only one exit point is possible. The path depends on the plant parameters which are symptomatic of the particular CSF. The exit is always to one of the several procedures associated with the specific CSF Status Tree, or, if the path is green, to remain in the E series procedure being executed. If the exit is to an FR procedure, then the E series procedure is suspended. The FRPs are designed to restore the condition of the CSF and not the plant. When the FRP is exited, the operator is directed back to the appropriate ORP to continue plant recovery.

OP-MC-EP-INTRO FOR TRAINING PURPOSES ONLY REV. 02 Page 29 of 63

IR 05000369/2003301

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2003 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (1) Assessments Mid Cycle End of Cycle 20(0) 7(0) Security Operator Exams 301(0) 302(0) 302(0) Emergency Planning Other

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