ML080940427

Draft - RO & SRO Written Exam (Folder 2)

ML080940427
Person / Time
Site:	Nine Mile Point
Issue date:	01/28/2008
From:	Constellation Energy Group
To:	NRC Region 1
Hansell S
Shared Package
ML073040288	List: IR 05000410/2008301 ML073100864 ML080670184 ML080930186 ML080940413 ML080940427 ML080940442 ML080940454 ML080940462 ML081010150 ML081010159 ML081010170 ML081010219 ML081010232
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Download: ML080940427 (276)

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Y Constellation Energy Nine Mile Point Nuclear Station NINE MILE POINT UNIT 2 NRC EXAMINATION Written Examination Draft S u January 28,2008

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 1 2 203000 A4.01 4.3 B NMP2 Bank SYSID: 12925 LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents F 1 N2-OP-31 RHWLPCI: Injection Mode - Ability to manually operate and/or monitor in the control room: Pumps QUESTION 1 Following a plant transient, the following conditions exist:

0 Drywell Pressure is 2.5 psig, rising.

0 RPV Water Level is 90 inches, lowering at 1 inch per minute.

0 Reactor Pressure is 500 psig, lowering at 5 psig per minute.

0 THEN annunciator 601447, RHR A INJ VLV 24A OPEN PERMISSIVE alarms.

Which one of the following is the indicated flow on the 2RHS*F114A Division 1 RHS Loop Total Flow Indicator five minutes after annunciator 601447 is received?

A. Ogpm

6. 1,000gpm C. 3,500gpm D. 7,000gpm Correct Answer: A When 601447 alarms, Reactor Pressure is 130 psid above RHR Pressure, which was already operating at minimum flow due to High Drywell Pressure.

5 minutes later, Reactor Pressure is 105 psid above RHR Pressure, resulting in Shutoff Head conditions. Minimum Flow Valve, 4A taps off BEFORE the Flow Element, which results in 0 gpm indicated flow.

Plausible Distractors:

B is plausible; would be true if Minimum Flow path produced flow indication.

C is plausible; would be true if Reactor Pressure were -250 psig.

D is plausible; would be true if Reactor Pressure were -150 psig.

Objective Link: 02-OPS-001-205-2-00 EO 1.6 Page 1 of 88

ATTACHMENT 4 i C o r i t )

2CEC*PNL601 S E R I E S 4 0 0 A&ARM RESPONSE FROCEDURES Xeflash: NO RHR A IMJ VLir 240, OPEN PERWI SC IVE 445 CCQWTJTER POINT PR I N T N T S(.,URCE

.- SETPOIMT RHSBC07 RHR A I N J VLV 2RHS

• PDT3lk c 13'J ps1ci 21.3. PERMIS Ai.1 t c!nia t ic! Re sgonee I F an RWS LGCA s i y l i a l is present, RHS*M07J;:4A, L P C I A INJECTION VLV, opens

( D ~ - y w e l lpl'erssure 1.68 p i g r;.R R P V Level 1.7.8 inches)

Q x r a t o i - Ac!tir>ns NOTE: T h i s aniiunciator will be nnrmally enercjiztd WHEN t h e RPV i s depressurized .

1. TF a n RH$ LOCA s i g n a l is present., v e r i f y 2KH.S*MOVZ.IA has an open s i g n a l a t 2CEC*PNL601.

2. I F LEY1 iriiect.iori is r e q u i r e d per N2-EOPs, v f y ZRHS*M(?Vi4A is r p m at.

~CIEC*PNL~O~.

Pcssitlw Causas Reactor pressure lowered u n t i l it is h s s than 130 p s i above Residual H e a t Rernclval System Pi-References r u -OF- 3 1 Page 2?(;

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 2 2 205000 A3.03 3.5 B NMP2 Bank SYSID: 4832 LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents H 1 ARP 603105 Shutdown Cooling - Ability to monitor automatic operations of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) including: Lights and alarms QUESTION 2 The following conditions exist:

0 The reactor is in Mode 3.

0 RHR Loop "A" is in Shutdown Cooling.,

0 RHR Loop "B" is in Suppression Pool Cooling.

0 DIVISION I RHR/LPCS REAC WTR LVL LOW alarms.

0 DIVISION II RHR REAC WTR LVL LOW alarms.

Which one of the following describes the automatic RHR system response?

A. RHR Loop A continues to operate to shutdown cooling and RHR B pumps trips.

B. RHR A and B pumps trip, RHR Loop B realigns to the injection mode and the RHR B pump restarts.

C. RHR pump A trips and RHR Loop E3 realigns to the injection mode and the RHR B pump continues to run.

D. RHR Loop A continues to operate in shutdown cooling and RHR Loop B.

Correct Answer: C An RPV Level signal will cause a Group 5 isolation. (This is satisfied by RPV level 1). Since the RHS-A loop is in Shutdown cooling, the isolation signal will cause closure of MOVlI2, which will result in a pump trip. RHS-B will realign and inject per design Plausible Distractors:

A is plausible; would be true for RHR B Pump overload condition.

B is plausible; identifies potential misconception on Suppression Pool Cooling / LPCl automatic reaIignment .

D is plausible; identifies potential misconception on Shutdown Cooling Isolation.

Objective Link: 02-OPS-001-205-2-00 EO-4d Page 2 of 88

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Interlock Setpoint Function Suppression Pool The suppression pool cooling valve Prevents diverting water Cooling Valve cannot be opened during a LOCA unless from LPCI injection 2RHS*FV38A its respective LPCI injection valve 2RHS*FV38B 2RHS*MOV24A (8)is shut If open, it will automatically shut on a LPCI initiation ESK-06RHSO19, 034 807E1707Y Sheet 3, 8, 10 Group 4 Containment The Group 4 (RHS sampling and Prevent uncontrolled Isolation Radwaste discharge) valves shut on the primary coolant release following signals: from primary Group 4 Valves 2RHS*MOV142 Low RPV level, 159.3 inches (possible containment 2RHS*MOV149 indication of leak) 2RHS*SOV35A & B High drywell pressure, 1.68 psig (possible 2RHS*SOV36A & B indication of leak)

Group 5 Containment The Group 5 (shutdown cooling and Prevent uncontrolled Isolation head spray) valves shut on the p rima ry coo la nt release Group 5 Valves following signals: from primary 2RHS*MOV112 High reactor pressure, 128 psig (protect containment 2RHS*MOV113 piping and HX) 2RHS*MOV104 Low RPV level, 159.3 inches (possible 2RHS*MOV40A & B indication of leak) 2RHS*MOV67A & B High RHS area temperature, 135OF (indicates a leak)

High Reactor Building (RB) temperature, 130.2OF (indicates a leak)

High RB pipe chase temperature, 135OF (indicates a leak)

Steam Condensing I n the steam condensing mode, the Prevents exceeding RHS High Pressure pressure control valve bypass valve design pressure Isolation closes if reactor pressure is above 465 2RHS*MOV23A Psig 2RHS*MOV23B ESK-06RHS009, 010 Student Guide (~2101205000C01) 66 0 1 224 Printed: 04/10/2007

4. System / Component Interlocks Interlock Setpoint Function

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RHS Pump Suction RHS pump controls are interlocked with Prevent running RHS Trip the valves which can1 isolate the suction pump without suction path for each pump. flowpath ZRHS*PlA ZRHS*PlB Pump starting is prohibited unless one of the following com'binations of valves ESK-OSRHSOOl, 002 are fully open:

807E1 70TY Sheet 3, 8, 9 Suction from the reactor recirculation loop:

2RHS*MOV112, MOVl 13, and MOV2A (B)

Suction from the suppression pool:

2RHS*MOVlA (B)

I f a RHS pump is running and its suction valve leaves its fully open position, the pump trips.

RPV Draindown The shutdown cooling suction valve Prevent inadvertent 2RHS"MOVZA (B) is interlocked with draining of the reactor the suppression pooll suction valve vessel ESK-06RHS002, 004, 029, 030 2RHS*MOVlA (B), the suppression pool cooling valve 2RHS*FV38A (B), and the suppression pool spray valve 2RHS*MOV33A (B).

I f the suppression pool suction valve, suppression pool cooling valve, and the suppression pool spray valve are not fully closed, the shutdown cooling suction valve to that loop cannot be opened.

Also, in order to open the suppression pool suction valve, suppression pool spray valve, or suppression pool cooling valve, the shutdown cooling suction valve must be fully closed.

LPCI Initiation Each division of LPCI is automatically Automatically initiates a initiated by a one-out-of-two twice logic division of LPCI scheme of triple-low reactor water level 807E170TY Sheet 3, 8, 9 807E171 TY Sheet 2, 4 (Ll, 17.8 inches) and/or high drywell pressure (1.68 psig) signals.

Student Guide ( ~ 2 1 0 1 2 0 5 0 0 0 ~ 0 1 ) 64 of: 224 Printed : 04/10/2007

4. Control a n d indication f o r those c o m p o n e n t s necessary t o establish operation o f RHS Loop A and/or B in t h e s h u t d o w n cooling and suppression pool cooling m o d e s is provided a t t h e r e m o t e s h u t d o w n panel.

5. Low Pressure Coolant I n j e c t i o n Mode

a. The Low Pressure Coolant I n j e c t i o n (LPCI) m o d e o f RHS is a n e m e r g e n c y m o d e o f operation. I n t h e event o f a loss-of-coolant accident (LOCA), t h e LPCI mode, in conjunction w i t h CSL, High Pressure Core Spray (CSH),

and/or t h e Automatic Depressuri;zation S y s t e m (ADS), restores and m a i n t a i n s t h e desired w a t e r level in t h e reactor vessel. Maintaining this w a t e r level will provide adequate cooling capability t o p r e v e n t fuel overheating.

b. I n t h e LPCI m o d e all t h r e e RHS p u m p s t a k e suction f r o m t h e suppression pool a n d p u m p w a t e r i n t o t h e r e a c t o r vessel via separate lines. The RHS h e a t exchangers are n o t required f o r initial s h o r t t e r m LPCI m o d e operation due t o t h e h i g h h e a t capacity o f t h e suppression pool. For long t e r m cooling t h e h e a t exchangers (Loop A a n d 8) can be placed in service t o reject h e a t t o SWP. Since t h e c o n t a i n m e n t s t r u c t u r e is watertight, a closed loop is f o r m e d using t h e suppression c h a m b e r as a s u m p .

c. The LPCI m o d e initiates automatically o n a LOCA indication of high d r y w e l l pressure (1.68 psig) o r low RPV w a t e r level (Ll, 17.8 inches). To provide overpressure protection #oft h e RHS piping, t h e LPCI injection valve will n o t open until RPV pressure lowers t o w i t h i n 130 psid o f LPCI injection pressure. This m o d e m a y also be r e m o t e manually initiated f r o m t h e Control Room. LPCI m o d e s h u t d o w n is b y operator action f r o m t h e Control Room. I f RHS is operating in t h e Suppression Pool Cooling, Suppression Pool Spray, S t e a m Condensing, o r Testing m o d e a n d a LPCI Student Guide ( N ~ ~ O ~ ~ O ~ O O O C O ~ ) 33 of 224 Printed: 04/10/2007

initiation signal is received, RHS will divert from that mode to the LPCI mode.

6. Suppression Pool Cooling Mode

a. The suppression pool cooling mode of RHS provides a method of cooling the suppression pool water to limit the peak pool temperature in the event of reactor blowdown, safety relief valve testing, I C s operation, or post accident conditions.

b. I n this mode, which is manually initiated and secured by operator action, either RHS Loop A or B (or both, if required) can be used. Water is pumped from the suppression pool through the RHS heat exchanger (where the suppression pool heat is rejected to SWP) and back to the suppression pool.

7. Containment Spray Cooling Mode

a. During post accident conditions, ,the containment spray cooling mode of RHS may be manually initiated to reduce the primary containment pressure. The spray of cooled suppression pool water directly into the containment atmosphere will condense any steam and cool any non-condensable gases that may be present. Either RHS Loop A or B may be used in this mode. Water is pumped from the suppression pool through the RHS heat exchangers (where heat is rejected to SWP) and into the containment atmosphere via the containment spray headers.

b. The containment is provided with three spray headers. Two are located in the drywell and one in the suppression chamber. Each of the drywell spray headers is supplied by different RHS loops. This establishes two 100 percent capacity, independent, drywell spray cooling subsystems.

The spray and condensate collects in the bottom of the drywell until the

~ ~ ~ ~

S t u d e n t Guide ( ~ 2 1 0 1 2 0 5 0 0 0 ~ 0 1 ) 34 01' 224 Printed: 04/10/2007

ATTACHMENT 4 (Cont) 2CECfPNL601 SERIES 400 ALARM RESPONSE PROCEDURES Reflash: YES 2CEC*PNL601 601403 DIVISION I RHR/LPCS REAC WTR LVL LOW COMPUTER POINT COMPUTER PRINTOUT SOURCE SETPOINT cs LLCO1 Dl RHR/LPCS RX 2ISC*LT9A t17.8 i n .

WTR LVL CSLLC02 D1 RHR/LPCS RX 21SCfLT9C t17.8 i n .

WTR LVL Automatic ResDonse 0 I F BOTH channels sense l o w Reactor water l e v e l , t h e f o l l o w i n g occurs:

- D i v i s i on 1 Emergency D i e s e l Generator, 2EGS*EGl s t a r t s .

- LPCS system a l i g n s i n t h e i n j e c t i o n mode.

- RHR Loop A a l i g n s i n t h e LPCI mode.

- RHSfMOV8A, HEAT EXCHANGER 1A INLET BYPASS VLV opens AND s e a l s i n f o r 10 minutes.

- RHR Steam Condensing mode i s o l a t e s .

ODerator A c t i o n s

1. V e r i f y by m u l t i p l e i n d i c a t i o n s alarming c o n d i t i o n .

2. R e f e r t o Emergency Operating Procedures.

3. V e r i f y automatic response, a t P601/P852.

Page 238 N2-ARP-01 Rev 00

ATTACHMENT 6 (Cont) 2CEC*PNL601 SERIES 600 ALARM RESPONSE PROCEDURES Reflash: YES zCEC"PNL60 1 601603 DIVISION I1 RHR REAC WTR LVL LOW COMPUTER POINT PR I NTOUT SOURCE SETPO I NT ISCBC03 RHR/ADS WATER Trip Unit < 17.8 inches LEVEL CH B U22-N691B (21SCfLT9B)

ISCBCO4 RHR/ADS WATER Trip Unit < 17.8 inches LEVEL CH F B22-N691 F (21SCfLT9D)

Automatic Response I F BOTH Channels sense low Reactor l e v e l t h e f o l l o w i n g occurs:

0 2EGS*EG3, D i v i s i o n 2 Emergency Diesel Generator, s t a r t s .

0 RHS Loops B & C a l i g n i n t h e LPCI mode.

0 A Containment BOP LOCA i s o l a t i o n occurs.

0 RHR Steam Condensing mode i s o l a t e s .

Operator A c t i o n s

1. Confirm t h e alarming c o n d i t i o n by m u l t i p l e i n d i c a t i o n s .

2. Refer t o Emergency Operating Procedures (EOPs) .

3. A t ECEC"PNL601, v e r i f y RHS B AND C LPCI auto i n i t i a t i o n p e r N2-OP-31, Subsection F.2.0.

Page 425 N2-ARP-0 1 Rev 00

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier KIA Statement IR Origin Source Question 3 2 205000 K5.02 2.8 N NA LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents H 1 N2-OP-31 6.24 through 6.26 Rev 15 Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) : Valve operation QUESTION 3 RHS B Loop is operating in Shutdown Cooling Mode, with conditions as follows:

0 The plant is in MODE 4, Cold Shutdown.

0 NO Reactor Recirculation Pumps are in operation.

0 RHR B Loop Flow is 5000 gpm.

0 Cooldown Rate is 100°F/hr.

RHS*MOV8B, HEAT EXCHANGER 1B-INLET BYPASS VLV THROTTLE is 50% open.

Which one of the following valve throttling manipulations is required to operate at rated system flow conditions?

A. RHS*MOV40B, SDC B RETURN THROTTLE in the OPEN direction and throttling RHS*MOV8B, HEAT EXCHANGER 1B INLET BYPASS VLV in the OPEN direction.

B. RHS*MOV40B, SDC B RETURN THROTTLE in the OPEN direction and throttling RHS*MOV8B, HEAT EXCHANGER IB INLET BYPASS VLV in the CLOSED direction.

C. RHS*MOV40B, SDC B RETURN THROTTLE in the CLOSED direction and throttling RHS*MOV8B, HEAT EXCHANGER 1B INLET BYPASS VLV in the OPEN direction.

D. RHS*MOV40B, SDC B RETURN THROTTLE in the CLOSED direction and throttling RHS*MOV8B, HEAT EXCHANGER 1B INLET BYPASS VLV in the CLOSED direction.

Correct Answer: A With Loop Flow LOW, it is required to throttle RHS*MOV40B, SDC B RETURN THROTTLE in the OPEN direction. This will RAISE Cooldown rate. To LOWER Cooldown rate, it is required to throttle RHS*MOV8B, HEAT EXCHANGER 1B INLET BYPASS VLV in the OPEN direction.

Page 3 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal Plausible Distractors:

B is plausible; identifies misconception on RHR Loop Flow acceptable value (7450 gpm) and action to reduce Cooldown Rate.

C is plausible; identifies misconception on RHR Loop Flow acceptable value (7450 gpm).

D is plausible; identifies misconception om action to reduce Cooldown Rate.

Page 4 of 88

D. PRECAUTIONS AND LIMITATIONS 1.O Do not exceed 130°F RHR HX service water outlet temperature. This (C8) temperature limit is due to pipe design limits of the Service Water system and is applicable even when SWP is isolated. Do not isolate SWP to Ht Ex when Rx coolant is > 130°F during operation.

2.0 If a pump temperature problem is suspected, observe the following with a pyrometer:

Alarm Shutoff 0 Maximum winding temperatures, 311°F 338°F 0 Maximum bearing temperatures, 194°F 212°F 3.0 Observe the following RHR pump motor start limitations:

4.0 Observe the following RHR Pump limitations:

0 Maximum full load current, 126 amps 0 Pump runout flow, 8400 gpm 0 Maximum continuous ambient temperature, 148°F 8 Pump rated flow, 7450 gpm 5.0 With both reactor recirculation pumps idle, reducing RHR Shutdown Cooling flow can result in RPV thermal stratification. This can cause water surface temperature to rise above boiling, resulting in pressurization of the vessel or the venting of steam. This can be prevented by the following:

5.1 Do not throttle RHR Shutdown Cooling injection flow below rated flow unless there is a recirculation pump running in the opposite loop in Mode 3 and 4.

5.2 Control RHR Shutdown Cooling injection temperature by throttling the amount of RHR flow through the heat exchanger. If the tIX bypass valve is full open and less cooling is required, throttle RHR HX service water flow by throttling the service water outlet valve (2SWP*MOV33A or 6).

5.3 Refer to N2-SOP-31, LOSS OF SHUTDOWN COOLING OR N2-SOP-31R, REFUELING ALTERNATE SHUTDOWN COOLING, as applicable.

Page 12 N2-OP-31 Rev 16

F. NORMAL OPERATION (Cont)

NOTES: 1. 5 100"FIHr cooldown rate and 5 130°F RHR Service Water outlet temperature is maintained by throttling 2RHS*MOV40B(A).

2. Steps F.6.22 through F.6.27 may have to be repeated or performed simultaneously to meet all of the temperature restrictions discussed above.

CAUTIONS

1. During a refueling outage, total drive flow through the jet pumps should be less than 5700 gpm when incore instrumentation is not fully surrounded (all four corners) by fuel andlor blade guides to preclude incore instrumentation from damage due to flow induced vibration. This includes RHR Shutdown Cooling and Recirculation Drive Flow.

2. The RHR pump is without minimum flow protection, Minimum flow of 2 1000 gpm must be established within 40 seconds of pump start. Use of a stopwatch is recommended to ensure the pump is tripped within the required time if minimum flow is not achieved. Do not allow pump to run for > 15 seconds deadheaded.

NOTE: To assist in controlling reactor water level, prior to the start of an RHR pump, the Condensate Booster Pump should be secured per N2-OP-3, Section G, 1.16. Unless otherwise directed by the EOP's.

6.21 Start 2RHS*PlB(A) at 2CEVPNL6Ol.

6.22 Throttle open RHS*MOV40B(A), SDC B(A) RETURN THROTTLE to 2 1000 gpm.

6.23 IF RHSMOV40B(A) does not begin to open in 15 seconds system flow is NOT 2 1000 gpm 40 seconds after pump start. Place RHS*PlB(A) control switch to STOP, THEN release to Normal-After-Stop.

6.24 Slowly throttle open 2RHS*MOV40D(A), NOT to exceed the following:

For normal operation, 7450 gpm.

In anticipation of commencing fuel off -load in accordance with the applicable FHPs, 5700 gpm.

0 for simultaneous operation of Reactor Recirc AND RHR Shutdown Cooling. The following may be used as a guideline to verify proper operation.

Page 59 N2-OP-31 Rev 16

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin ' Source Question 4 2 20900 1 K6.05 2.8 N NA LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents H 1 N2-OP-52 Att 6 Rev 7 Knowledge of the effect that a loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM : ECCS room cooler(s)

QUESTION 4 The plant is in Mode 2, with the following:

0 Annunciator 870121, DIVISION I UNIT COOLERS MOTOR OVERLOAD alarms.

0 Computer Point HVRTC08, LPCS PMP RM UC402B MOT, Source 49X-2HVRB12 caused the alarm.

Which one of the following describes the affect, if any, of this condition on the operation of the Low Pressure Core Spray System?

If the Low Pressure Core Spray System is INITIATED due to a Loss of Coolant Accident, excessive motor temperatures:

A. will not occur due to this failure, because of the present Mode.

B. will not occur due to this failure, because one Unit Cooler will meet the cooling requirements of the Low Pressure Core Spray System.

C. will not occur due to this failure, because Reactor Building HVAC Supply Fans provides the required cooling for the Low Pressure Core Spray System.

D. will occur as a result of this failure, because BOTH Unit Coolers are required to meet the cooling requirements of the Low Pressure Core Spray System.

Correct Answer: B ONE of TWO LPCS Room Coolers is required to provide adequate cooling. With HVR*UC402A still available, LPCS Room Cooler requirements will be met.

Page 5 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal Plausible Distractors:

A is plausible; Coolers are required in Made 1, 2, and 3.

C is plausible; Normal RB HVAC Supply Fans do not operate under accident conditions.

D is plausible; ONLY ONE Cooler is required.

Page 6 of 88

5. ECCS room coolers

a. AC powered air coolers, using service water as a cooling medium, remove motor heat and energy released from accident conditions (reactor coolant leaks) from the pump room.

b. Loss of the cooling capability could result in component overheating if accident conditions or prolonged operation exist.

6. Keep fill system

a. Keep fill is provided by a common jockey pump shared between CSL and RHS "A". The keep fill maintains CSL piping full of water.

b. Loss of keep fill will render the CSL inoperable and possibly non-functional.

7. ECCS room integrity

a. ECCS pump rooms are watertight t o preclude system loss due to flooding that could occur under postulated accident conditions.

b. Loss of room integrity could result in CSL being inoperable and could lead t o loss of function if adverse environmental conditions were t o occur.

8. Automatic Depressurization System (ADS)

a. Under certain accident conditions, ADS opens dedicated SRVs t o lower reactor pressure below the shutoff head of the CSL pump.

b. I f the ADS is not available on small loss of coolant accidents, then the CSL may not be able t o inject due t o reactor pressure remaining above the shutoff head of the CSL pump.

Student Guide (~21oi209ooicoi) 59 of 102 Printed: 04/10/2007

UNIT COOLER AREA DIV ACTION LCO HVR*UC401A/D RHS Pump Room A 1 lo 3.6.2.3/3.6.2.4/3.6.1.6/

3.4.9/3.6.4.3/3.5.1 HVR*UC401B/E RHS Pump Room C II 1 3.6.4.3/3.5.1 HVR*UC401CIF RHS Pump Room B II 3.6.2.3/3.6.2.4/3.6.1.6/

I @ 3.4.9/3.6.4.3/3.5.1 HVR*UC402A/B LPCS Pump Room I 1 3.6.4.313.5.1 HVR*UC403A/B HPCS Pump Room Ill 7 3.6.4.313.5.1 HVR*UC405 RHS HX Room A I 2 3.6.2.3/3.6.2.4/3.6.1.6/

3.4.9/3.5.1/3.6.4.3 HVR*UC406 RHS HX Room B II 2 3.6.2.313.6.2.413.6.1 -61 3.4.913.5.1/3.6.4.3 HVR*UC412AIB RClC Pump Room 3.6.4 313.5.3 111I 3@

I HVR*UC413A/B Recirc Unit Cooler 1/11 3.6.4.3 4@

IHVR*UC415AIB GTS Fi!ter Roo8 f Ill! Ir 13 13.6.4.3 I HVR*UC408NB N Aux Bav, Elec MCC II 16 13.8.8 IHVR*UC409A/B I S Aux Bay, Elec MCC II I 16 13.8.8 I

@Refer to N2-OP-78 Attachment 11 for potential compensatory actions prior to making 2HVR*UC401A (401C, 412A, 413A) inoperable.

NOTES: 1. The following actions in regards to GTS operability are applicable for drawdown concerns durinq Operational Condition 1, 2, 3.

Action 5.c is also applicable during Modes 4 AND 5.

2. For actions requiring declaration of an "RHS System" inoperable in Modes 1, 2 OR 3, the inoperable modes to consider are:

LPCI, Suppression Pool Cooling, Suppression Chamber Spray, Drywell Spray AND Shutdown Cooling.

3. Both unit coolers in RHS Pump Rooms A & B should be in standby. This will assure that during the Shutdown Cooling Mode of operation, the RHR Pump Room temperature remains below the leak detection trip setpoints.

4. During accident conditions with only one unit cooler in service, RHR Pump Rooms A & 6 leak detection system may be activated, however this will NOT impact the ECCS function, Page 70 N2-OP-52 Rev 08

ATTACHMENT 6 (Cont)

ACTIONS: WHEN a HVR unit cooler is out of service OR to be removed from service for any reason, observe the requirements of Attachments 6 AND 7 (on accompanying pages),

1.a One unit cooler may be removed from service in any RHSlLPCS pump room with no compensatory action.

.b With BOTH unit coolers in any RHSlLPCS pump room inoperable, declare the associated RHSlLPCS system inoperable, AND in addition, declare the associated train of GTS inoperable. Perform actions as described in the applicable LCOs.

2. With the unit cooler inoperable in any RHS heat exchanger room, declare the respective train of GTS inoperable, declare the associated RHS system inoperable, AND follow the actions described in the applicable LCOs.

3.a With one of the RClC pump room unit coolers inoperable declare the associated GTS train inoperable. In addition, restore the inoperable unit cooler to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR declare RClC inoperable

.b With BOTH unit coolers in the RClC pump room inoperable declare RClC AND a GTS train inoperable.

4.a With one recirculation unit cooler inoperable declare one train of GTS inoperable AND take actions as described by the applicable LCO.

.b With BOTH recirculation unit coolers inoperable declare both trains of GTS inoperable AND take actions as described by the applicable LCO.

5.a With one GTS filter room unit cooler inoperable, declare the associated train of GTS inoperable AND take the actions as described in the applicable LCO.

.b With BOTH GTS filter room unit coolers inoperable, declare BOTH trains of GTS inoperable AND take the actions as described in the applicable K O .

.c During Modes 4 AND 5, GTS Train(s) operability will be maintained with 2HVR*UC415A(B) out of service IF the following requirements are met:

0 The fan motors associated with Unit Coolers 2HVR*UC415A(B) are disabled WHEN the unit coolers are rendered inoperable.

0 Non-emergency lights in the GTS room are secured within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after GTS initiation, OR temperature monitoring is initiated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of GTS operation AND the lights secured WHEN the temperature reaches 95°F.

0 Temporary constructionlmaintenance related heat loads (Le. welders, heaters, light banks, etc.) in the GTS AND adjacent rooms are secured within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of GTS initiation, OR WHEN the temperature in the affected space reaches 85°F.

Page 71 N2-OP-52 Rev 08

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 5 2 209002 K5.04 3.8 N NA LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents H 1 NER-2M-039, REV. 5 4-45 Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE CORE SPRAY SYSTEM (HPCS): Adequate core cooling: BWR-5,6 QUESTION 5 Following a Loss of Coolant Accident, the following conditions exist:

0 ALL Control Rods are inserted.

0 Reactor Pressure is 950 psig.

ONLY High Pressure Core Spray is injecting.

NO OTHER Injection Systems are operating.

0 When RPV Water Level LOWERED to -30 (actual) inches, Automatic Depressurization System Safety Relief Valves were MANUALLY OPENED.

0 RPV Water Level LOWERED and stabilized at -60 (actual) inches, with High Pressure Core Spray injecting 6500 gpm.

Which one of the following describes the status of adequate core cooling during this event sequence?

A. NOT been maintained, because RFV Water Level LOWERED below the Minimum Zero Injection RPV Water Level.

B. NOT been maintained, because RPV Water Level LOWERED below the Minimum Steam Cooling RPV Water Level.

C. IS maintained because RPV Water Level has continuously provided adequate cooling to the fuel by SUBMERGENCE.

D. IS maintained, because RPV Water Level and Core Spray Flow provided adequate cooling by CORE SPRAY COOLING.

Correct Answer: D Adequate Core Cooling has been maintained, because Core Spray exceeds 6350 gpm with RPV Water Level above -62 (minus 62) inches.

Page 7 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal Plausible Distractors:

A is plausible; would be true with NO Injection, because RPV Water Level LOWERED below

-55 (minus 55) inches. (MZIRWL)

B is plausible; would be true if RPV Water Level LOWERED to -39 (minus 39) inches prior to Blowdown. (MSCRWL)

C is plausible; would be true if RPV Water Level remained above -14 (minus 14) inches.

Page 8 of 88

NMP-2 EOP-RPV Basis Discussion (continued)

Following a large recirculation line break, it may not be possible to restore and maintain RPV water level above -39 inches if the available injection capacity is insufficient to overcome flow through the break. The core may still be considered adequately cooled, however, if design basis core spray flow requirements are satisfied:

Following a design basis recirculation line break in B W 3 through B W 6 designs (NMP-2 is a - B W 5 ) , RPV water level is expected to stabilize at the elevation of the top of the jet pumps, below the MSCRWL of -39 inches. The covered portion of the core is then cooled by submergence while the uncovered portion is cooled by spray flow.

If RPV water level can be restored and maintained above -39 inches. or "

conditions can be established, the core will remain adequately cooled and no further action need be taken. Efforts to restore RPV water level above -14 inches, the top of the active fuel, should continue, but at some point it may become necessary to throttle ECCS flow to observe NPSH restrictions or divert flow to containment cooling functions in accordance with design basis assumptions. The system operating details and cautions listed in Step L-6 and Detail E l continue to apply; NPSH and vortex limits should be observed if possible, but may be exceeded if thle situation warrants.

If RPV water level cannot be restored and maintained above the -39 inches using only preferred injection sources and core spray cooling cannot be established, alternate injection systems must be used, if not already in service. While earlier steps permitted use of the alternate injection systems, injection may not have been previously required if preferred systems or subsystems were available.

If RPV water level cannot be restored and maintained above -39 inches and core spray cooling cannot be established, drywell (primary containment) flooding is required. If a primary system break exists, flooding the primary containment will backfill the RPV through the break. While flooding will be possible only if a primary containment fill source of sufficient capacity is available, immediate transfer to the S A P S is prescribed in anticipation of possible core geometry changes and so that instructions appropriate to the condition will be in effect when the necessary injection or fill systems are available.

NER-2M-039, REV. 5 4-45

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 6 2 21 1000 K2.02 3.1 B NMP2 Bank SYSID: 12961 LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents H 1 SG N2101211000C01 page 105 Knowledge of electrical power supplies to the folloi. ing: Explosive valves QUESTION 6 The plant is operating at 100% power with the Standby Liquid Control System (SLS) in a normal standby configuration.

Which one of the following abnormal conditions will result in a loss of continuity to (SLS) Explosive Valve 2SLS*VEX3A?

A. Trip of SLS Pump A motor thermal overload device.

B. Trip of SLS Pump A suction valve power supply breaker.

C. Blown power supply fuse for Div I Redundant Reactivity Control logic.

D. Blown control power fuse for SLS Pump A motor power supply breaker.

Correct Answer: D 2EHS*MCC102 provides motor operating power (600 VAC) and motor control power for Explosive Squibs for 2SLS*VEX3A.

Plausible Distractors:

A is plausible; would be true if Motor Breaker power were supplied to Explosive Squibs.

B is plausible; Suction Valve power is from a separate breaker on 2EHS*MCC102.

C is plausible; RRCS provides an SLS Initiation signal, does not provide power to Explosive Squibs.

Objective Link: 02-OPS-001-211-2-00 Page 9 of 88

Power Supply Control Power Source Component Breaker or Fuse Breaker or Fuse

~~ ~

SLCS STORAGE TANK 2 EHS

• MCC30 2-. 16B 115VAC, 3 AMP breaker OUTLET MOV control fuse (2SLS*MOVlB)

STANDBY LIQUID PUMP A 115 VAC, 10 AMP (2SLS*PlA) Breaker control fuse Squib P-619 panel F2/ F3 STANDBY LIQUID PUMP B 115 VAC, 10 AMP (2SLS*PlB) Breaker control fuse Squib P-619 panel F2/F3 SLCS OUTBOARD ISOL 115VAC, 3 AMP Breaker VALVE (2SLS*MOV5A) control fuse SLCS OUTBOARD ISOL 2EHS* MCC302-4D 115VAC, 3 AMP Breaker VALVE (2SLS*MOV5B) control fuse Heat Tracing from 2HTS* PNLOO 1- 1 l2OVAC power to Heat 2SLS*MOVlA to Trace panel from 2SLS EJ1A 2HTS* PNLOO 1-3 2EJS*PNL103A (2SLS-003-3-2)

Heat Tracing from 2 HTS* PNLOO 1- 1 4 2SLS*V63 to Air Sparger (2SLS-002-96-2) 2 HTS* PNLO0 1- 34 From *V66 to 2SLS*V67 2HTS*PN LOO1-3 1 (2sLs-002-100-2) 2HTS*PNLOO 1-33 Heat Tracing from 2 HTS* PNLOO3 - 1 120VAC power t o Heat 2SLS*MOVlB to Trace panel from 2SLS* EJ 16 2HTS *PN L003-3 2EJS* PNL3026 (2SLS-003-1-2)

Student Guide ( ~ 2 1 0 1 2 1 1 0 0 0 ~ 0 1 ) 45 of 158 Printed: 04/10/2007

3. 600 VAC Power Distribution (N2-OP-71C)

a. 2NHS-MCCOll supplies power t o the SLS Storage Tank operating and mixing heaters

b. A loss of 2NHS-MCCO11 results in the inability t o properly prepare a Sodium Pentaborate solution, and the inability t o maintain storage tank temperature a t the 7OoF minimum required by Technical Specifications

4. Standby & Emergency AC Distributiori (N2-OP-72)

a. Supports SLS components as follows:

1) 2EHS*MCClO2 provides motor operating power (600 VAC) and motor control power (transformed down t o 115 VAC) t o t h e following Division I components a) SLS Pump P I A b) SLS STORAGE TK OUTLET VLV, SLS*MOVlA c) OUTBOARD ISOL STOP CHECK VLV, 2SLS*MOV5A d) Squibs for 2SLS*VEX3A

2) 2EHS*MCC302 provides motor operating power (600 VAC) and motor control power (transformed down t o 115 VAC) t o the following Division I1 components:

a) SLS Pump P l B b) SLS STORAGE TK OUTLET VLV, SLS*MOVlB c) OUTBOARD ISOL STOP HECK VLV, 2SLS*MOV5B d) Squibs for 2SLS*VEX3B This distribution system also powers the SLS heat tracing system through panels 2HTS*PNL001 and 003.

~~

oooco1)

S t u d e n t Guide (~2101211 105 of 3.58 Printed: 04/10/2007

I. (

STOP K6A 14A

<42 Closeon Tank LOW Level OVERLOAD 0

I 7

SQUIB VALVE READY METER

/READY 3

SQUIB

~

T SQUIB PUMP LOSS OF CONTINUITY TO SQUIB VALVE OPENLS I I SWITCH OVER PUMP LOAD

- - - - SWITCH OPEN LIMIT

-I 420 K49 K74 42c MOV-1A-1B ANNUNCIATOR Figure 2 02-OPS-001-211-2-00 TITLE Standby Liquid Control System Handout

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 7 2 212000 K4.06 3.0 N NA LOK Grp 10 CFR 55.41 (b) 7 LOD (1-5) Reference Documents H 1 SG N2101212000C01 Knowledge of REACTOR PROTECTION SYSTEM design feature(s) and/or interlocks which provide for the following: Select rod insertion: Plant-Specific QUESTION 7 The plant is operating at full power. Both NORM-TEST-SRI Toggle Switches are placed in TEST on the Hydraulic Control Unit of a withdrawn control rod.

Which one of the following is the control rod response, if any?

A. Remains at the same position.

B. Inserts at SCRAM speed. Displaced water will drain from the Scram Discharge Volume.

C. Inserts at SCRAM speed. Displaced water will accumulate in the Scram Discharge Volume resulting in a Control Rod Block.

D. Inserts at SCRAM speed. Displaced water will completely fill the Scram Discharge Volume resulting in a Reactor Scram.

Correct Answer: B is correct, Scram Valves remain OPEN. The control rod will insert a SCRAM speed. Displaced water will drain from the Scram Discharge Volume. SDV Vents and Drains REMAIN OPEN.

Plausible Distractors:

A is plausible; would be true for ONE SRI Switches for ONE HCU.

C is plausible; would be true for BOTH SRI Switches at several HCUs simultaneously.

D is plausible; would be true for BOTH SRI Switches at several HCUs simultaneously.

Page 10 of 88

Control Device Function Location

~~

Selected Rod Insertion Two, three position (NORMAL, Located on each control (SRI) Switches TEST, SRI) toggle switches used t o rod drive Hydraulic test the individual coils of the Control Unit scram pilot solenoid valve, and to perform surveillance testing of the individual control rod scram times.

Placing both toggle switches in TEST or SRI position will de-energize both solenoids of the rod being tested, causing that particular rod t o scram into the core.

b. (RO) Control Room Controls r ~~ ~

Control De I Function I Location RPS Power Source Select Three position (NORMAL, ALT A, 2CEC* PNL6 10 Switch ALT 8) switch NORMAL: RPS MG sets supplying ALT A: 2LAT-PNI-100 is supplying ALT B: 2LAS-PNt400 is supplying Student Guide (~2101212000C01) 53 of 210 Printed: 04/04/2007

j. The Bypass Selector Switches for- the Neutron Monitoring System (1for SRM, 2 for IRM, 1 for APRM) allow bypassing a neutron monitor channel in RPS Trip System A or B.

Picture 21, Hydraulic Control Unit Selected Rod Insertion Switches

k. Located on each control rod drive Hydraulic Control Unit are two, three position toggle switches. The Selected Rod Insertion (SRI) switches are used t o test the individual coils of the scram pilot solenoid valve, and t o perform surveillance testing of the individual control rod scram times.

One switch will de-energize the coil for RPS Trip System A (A solenoid),

the other Trip System B. Placing both toggle switches in TEST or SRI will de-energize both solenoids of the rod being tested, causing that particular rod t o scram into the core.

~~

Student Guide ( ~ 2 1 0 1 2 1 2 0 0 0 ~ 0 1 ) 111 of 210 Printed: 04/04/2007

13) The HCU components making up the scram section consist of the following :

a) Inlet Scram Valve AOV126 and Outlet Scram Valve AOV127 b) Scram Pilot Air Valve AOIV139 and t h e associated Scram Pilot Air Isolation Valve V116 c) Scram Accumulator, N2 Cylinder and associated valves and monitoring instrumentation.

14) The Scram Valves (AOV126 and 127) are globe valves with soft Teflon seats t o minimize seat leakage, and are held closed during reactor plant operation by 70-75 psig Instrument Air pressure supplied through Scram Pilot Air Valve AOV139 t o the top of their diaphragm actuators.

a) A single air line from valve AOV139 supplies control pressure t o both Scram Valves.

b) The Scram Valves open by internal spring force plus reactor vessel water pressure, or Charging Water pressure when control air pressure is vented after repositioning the Scram Pilot Air Valve t o vent t o the atmasphere.

c) The internal spring preload in the Outlet Scram Valve AOV127 is slightly greater than in the Inlet Scram Valve AOV126. This permits AOV127 (outlet) t o open slightly before AOV126 (inlet) opens t o prevent a backpressure buildup above the CRDM Drive Piston upon application of scram pressure below the Drive Piston, Student Guide ( ~ 2 1 0 1 2 0 1 0 0 1 ~ 0 1 ) 39 of 2:10 Printed: 03/09/2007

d) The Outlet Scram Valve, when opened, connects the Withdrawal riser t o the Scram Discharge riser. Water from above the CRDM Drive Piston is routed into the SDV.

e) Valve AOV126 opens the. Insert riser t o the Charging Water header and the Scram Accumulator, so t h a t maximum differential pressure is applied across the CRDM Drive Piston.

IS) During normal reactor plant operation, each of the two RPS logic systems (A and B) energizes one of the solenoids on the 3-way scram pilot valve (AOV139) for each HCU.

a) When one or both of the solenoids is energized the pilot valve supplies 70-75 psig Instrument Air pressure t o the diaphragm actuators of the Inlet and Outlet Scram Valves, maintaining the scram valves closed.

b) To reduce the probability of a spurious scram, AOV139 is pneumatically configured so both solenoids must be de-energized before control pressure is vented from the AOV126 and 127 valves. De-energizing either one of t h e solenoids will n o t result in a scram, since control pressure will continue t o be applied t o the scram valve actuators. Upon initiation of a scram, both RPS logic systems are de-energized, venting control pressure from the scram valve actuators and permitting the valves to open.

16) The Scram Accumulator is a piston-type water accumulator pressurized by a volume of N2 gas a t a precisely controlled pressure a t known temperature in the I12 cylinder.

a) The accumulators and their instrumentation occupy the lower portion of the HCU.

Student Guide (~2101201001~01) 40 of 210 Printed: 03/09/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal

.............................. .... " ..... ..... ..... . ............................................................... ......,......,.. ..i .,.. ..... ............................................................ ...... ..... ... ............. ..........

RO I Tier I WA Number j Statement j IR i Origin Source Question 8  ; 2 , 215003 j ~1.07 3.0 B 2001 Quad Cities NRC Exam LOK G r p 10 CFR 55.41(b) 7 . L.OD (1-5) Reference Documents F $ 1 SG N2101215002C01 Knowledge of the physical connections and/or cause- effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following: Reactor vessel QUESTION 8 Which one of the following describes the AXIAL LOCATION within the Reactor Vessel of a FULLY INSERTED Intermediate Range Monitor (IRM) Detector AND a FULLY WITHDRAWN IRM Detector?

FULLY INSERTED FULLY WITHDRAWN IRM Detector IRM Detector A. Top of Active Fuel 2.5 feet BELOW the Bottom of Active Fuel

6. 1.5 feet ABOVE Core Bottom of Active Fuel Centerline C. Top of Active Fuel 1.5 feet BELOW Core Centerline D. 1.5 feet ABOVE Core 2.5 feet BELOW the Centerline Bottom of Active Fuel Correct Answer: D When an IRM Detector is FULLY INSERTED, it is axially located at 1.5 feet ABOVE Core Centerline and when an IRM Detector is FULLY WITHDRAWN, it is axially located at 2.5 feet BELOW Bottom of Active Fuel.

Plausible Distractors:

A is plausible; identifies misconception concerning FULLY INSERTED location.

B is plausible; identifies misconception concerning FULLY WITHDRAWN location.

C is plausible; identifies misconception concerning BOTH axial locations.

Objective Link:

Page 11 of 88

Student Guide (N2 101215002CO1):

Compoiient Description / Design Drive Mechanism Used to position IRM detector in 4 SRM A till-tt a its dry tube, which extends from

• IRM A thru H the bottom of the reactor vessel up into the reactor core

'jtrcke: F; 10 feet I wortion 1 Withdt-swiil Speed: 3 Wnr:rr

.I i:ktector Position: $2 1.5 feet above I:OE ce ritei'li ne (iirlly iiiserteil)

ore t?citt.ttulr>(fLl!ly witi?i';rawli)

+ Drive Mcrtoi Input Power-: 120,QOH

?'AC, 3 pl~i%>+?, 50 H2 c _

Student Guide (N2 101215002C01):

Setpoint - Any one of the following will initiate a block:

(1) SRM counts < 3 cps (a) Bypassed if:

(1) Associated logic channel IRMs are on Range 3 or above; (2) Mode Switch in RUN with APRM channels not downscale; or (3) SRM channel bypassed using bypass joystick (2) SRM counts < 100 cps when detector is not fully inserted (a) Bypassed if (1) Associated logic channel IRMs are on Range 3 or above; (2) Mode Switch in RUN; or (3) SRM channel bypassed using bypass joystick Setpoint - Any one of the following will initiate a block:

(a) Bypassed if:

(1) Mode Switch in RUN; or (2) IRM channel bypassed using bypass joystick

( 2 ) IRM reading < 5 / 125 of scale (a) Bypassed if (1) IRM is on Range 1 (2) Mode Switch in RUN; or (3) IRM channel bypassed using bypass joystick (3) IRM reading > 108 / 125 of scale (a) Bypassed if:

(1) Mode Switch in RUN; or (2) IRM channel bypassed using bypass joystick

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 9 2 215003 K5.01 2.6 N NA LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents F 1 SG N2101215002C01 Knowledge of the operational implications of the following concepts as they apply to INTERMEDIATE RANGE MONITOR (IRM) SYSTEM : Detector operation QUESTION 9 Which one of the following conditions will cause the IRM B Detector Retract Permit green indicating lamp to be illuminated?

A. The Reactor Mode Switch in RUN.

B. ALL IRMs are on Range 3 or above.

C. IRM Channel B Detector is fully withdrawn.

D. IRM Channel B indicates greater than 108/125.

Correct Answer: A The Reactor Mode Switch in RUN will illuminate the IRM B Detector Retract Permit green indicating lamp.

Plausible Distract0rs:

B is plausible; ALL IRMs are on Range 3 or above bypasses SRM Detector Retract function.

C is plausible; IRM Channel B Detector fully withdrawn results in white OUT indicating lamp illuminated.

D is plausible; IRM Channel B indicates greater than 108/125 results in UPSCALE indicat ing lamp illuminated.

Obiective Link: 02-OPS-001-215-2-02 Page 12 of 88

Instrument Range / Function Location IRM Indicating Lights Light on / off 2CEC*PNL603 (one set for each IRM)

Used t o indicate I R M status Detector Position Retract Permit (green) Detector Position Lights I n (white) Retract Permit - on if any one of the Out (white) following exist: (1) IRM ISbypassed; or (2)

Mode Switch is in RUN Indicating Lights I n - on if detector is fully inserted Upscale Trip or INOP Out - on if detector is fully withdrawn from (red) the core Upscale Alarm (amber)

Downscale (white) Indicating Lights Bypass (white) Upscale Trip or INOF' - on if IRM reading >

120/125 of scale or any one of the following exist: (1) high voltage power supply failure; (2) any module unplugged; or (3) mode switch out of OPERATE Upscale Alarm - on IF IRM reading >

108/125 of scale Downscale - on if IRM reading < 5/125 of scale Bypass - on if IRM has been bypassed using the BYPASS IRM joystick at 2CEC*PNL603

~

IRM Recorder 0 t o 40 Or 0 t o 125 (IRM Range Switch 2CEC*PN L603 position)

R603A-ZNMPNOl RED Pen (IRM A, B, E, F)

(C51-R603A, IRM A & C) Used t o monitor / record IRM flux BLUE Pen (IRM C, D, G, H)

R603B-2NMPN01 (C51-R603B, IRM B & D)

R603C-ZNMPN01 NOTE: These recorders are also used to monitor (C51-R603C, IRM E & G) / record Average Power Range Monitor (APRM) or R603D-ZNMPN01 Rod Block Monitor (RBM) readings (C51-R603D, IRM F & H)

Student Guide ( ~ 2 i o i 2 i 5 0 0 2 c o i ) 59 of 154 Printed: 03/28/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 10 2 2 15004 A I .06 3.1 B NMP2 Bank SYSID: 13601 LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents H 1 Ability to predict and/or monitor changes in parameters associated with operating the SOURCE RANGE MONITOR (SRM) SYSTEM controls including: Lights and alarms QUESTION 10 The following conditions exist for SRM A, during a plant startup:

0 SRMs are fully inserted.

0 SRM Channel A indicates IXI O5 cps.

0 Annunciator 603203, SRM UPSC/INOPERABLE alarms.

0 Annunciator 603442, CONTROL ROD OUT BLOCK alarms.

0 Amber SRM "UPSC ALARM OR INOP" light on P603 is LIT.

0 Red SRM "UPSC TRIP" lights on P603 are OFF.

0 White "INOP" light at the drawer is OFF.

Which one of the following actions is required, by procedure, to clear the Control Rod Out Block being generated by SRM A?

A. Retract SRM Detector to clear the UPSC condition.

B. Bypass SRM A Channel to clear the INOP condition.

C. Wait until ALL IRMs are on Range 8, THEN retract SRM Detector to clear the UPSC condition.

D. Wait until ALL IRMs are on Range 3, THEN verify the Control Rod Out Block automatically clears.

Correct Answer: A Upscale Alarm condition is indicated by light status. With a startup in progress, SRMs are required to be maintained on scale by retracting detectors.

Plausible Distractors:

B is plausible; would be true with the white INOP Light on the drawer LIT.

C is plausible; but does not maintain SRM indication on scale, as required.

D is plausible; SRM <IO0 cps is BYPASSED when ALL IRMs are on Range 3.

Page 13 of 88

ATTACHMENT 1 2 :Cont )

2CEC*PNL60? SERIES 2 0 0 ALXiRM RESPONSE PROCEDURES Reflash: No 6 0 1207 SRM UPSCALE/

INOPERAELE C o n i p u t a 1- Po i lit Computer P r i n t o u t Source NI*iSECC!3 SRM Upscaln,'INOP a . Great.er t h a n 1 x Iris cps (Ch. A , B, c or U)

b. SRM mode s w i t c h not i n o p e r a t e ICH. A , E ,

c or U!

c . Module unplugged (Ch.

A , E , C or D J

d. H.V. power s u p p l y f a i l e d IC!H. A, E , C or I?)

e . Greater t.i1an I09 "PS

! ? 0 - 1 ? 9 CPS) d u r i n g R2 re l c a d Aut omat i c Re spon Y 2

a. R o d w i t t i d r a w l block i n i t iatarl.

Cc.rract.ive A c t ion

a. Determine a f f e c t e d r:hannal by otssarvirig which amher "Upscale Alarm/Inop" l i g h t ( s ! &re i l l u m i r . a t e d .

b. Chock a s x c i a t e d t::cunt rate iitet.EL^ [xi P 6 0 3 t o d e t e r m i n e upscale - h i g h i n d i m t i o n , o r Inop - l o w i11di~::ation.

I f withdrawilig rods c o i n c i d e w i t h i.ipscalt alariu, retract. detector t o b r i n y i n d i c a t i o n within range.

d. Consult with t h e S . S . S . aiid bypass f a u l t y SRM c h a r m e l ,

45403,46536 Page 9 2 0 N2-ARP- 01 R e v 00

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier KIA Number Statement IR Origin Source Question 11 2 215005 K2.02 2.6 N NA LOK ' Grp ' 10 CFR 55.41(b) 7 ' LOD (1-5) Reference Documents H 1 SG N2101215003C01 Knowledge of electrical power supplies to the following: APRM channels QUESTION 11 With the plant operating at full power, a loss of power from 2VBB-UPS3A occurs.

Which one of the following describes the impact of this power loss of the Average Power Range Monitors?

A. APRM Channels 1 and 2 are de-energized and produces an RPS A Half Scram.

B. APRM Channels 3 and 4 are de-energized and produces an RPS B Half Scram.

C. APRM Channels 1 and 2 lose a power source, but remain energized.

Loss of power to a 214 Logic Module produces an RPS A Half Scram.

D. APRM Channels 3 and 4 lose a power source, but remain energized.

Loss of power to a 214 Logic Module produces an RPS B Half Scram.

Correct Answer: C APRMs are powered from Quadruple Low Voltage Power Supplies with redundant supply sources. Loss of 2VBB-UPS3A removes one of two inputs of power to QLVPS, and a loss of power to the 214 Logic Module which produces an RPS A Half Scram.

Plausible Dktractors:

A is plausible; identifies misconception about QLVPS power redundancy.

B is plausible; identifies misconception about QLVPS power redundancy.

D is plausible; would be correct for a Loss of 2VBB-UPS3B.

Page 14 of 88

b. The RBMs will initiate a rod block upon loss of power.

c. Depending on the specific power supply failure various results will occur:

Loss of 2VBB-UPS3A would result in a loss of both 2VBSXPNLA103 and A104 with respect to the 4PRMs. Loss of this UPS would not affect the APRM and LPRM chassis, since the APRM and LPRM chassis receive power through the quad low voltage power supplies. These power supplies receive power from both ZVBB-UPS3A and 3 8 and have auctioneering circuits theit will continue to provide power to the respective chassis.

a> A half scram will occur however, due to loss of power t o ZVBSYPNLA103 and A104 which supply power to 214 Module 1 and 3 which will send a trip signal to RPS Channel A 1 and A2 respectively.

Loss of 2VBB-UPS38 would result in a loss of both 2VBS*PNLB103 and 8104 with respect to the APRMs. Loss of this UPS would not affect the APRM and LPRM chassis, since the APRM and LPRM chassis receive power through the quad low voltage power supplies. These power supplies receive power from both 2VBB-UPS3A and 3 8 and have auctioneering circuits that wit1 continue to provide power to the respective chassis.

a) A half scram will occur however, due t o loss of power t o ZVBS*PML6103 and 8104 which supply power to 2/4 Module 2 and 4 which will send a trip signal to RPS Channel 81 and 82 respectivef y .

A loss of power to 2VBB-UPSJA and 3 8 would result in a full reactor scram due t o loss of RPS trip power and all APRM power.

Printed: 04/04/2007

Figure 21, PRNMS Power Distribution Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 12 2 2 17000 K6.04 3.5 N NA LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents H  ; I ARP 601343 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC): Condensate storage and transfer systern QUESTION 12 Following a reactor scram, the following conditions exist:

0 Reactor Core Isolation Cooling was manually initiated and is injecting.

Annunciator 601343, RClC CNDS TK 1A WATER LEVEL LOW alarms.

Which one of the following will occur as a result?

A. ICS"MOV129, PMP 1 SUCT FROM CONDENSATE STOR TK will shut, AND 601318 RClC PUMP 1 DISCH FLOW LOW will alarm.

B. ICS*MOV129, PMP 1 SUCT FROM CONDENSATE STOR TK will shut, AND 601327 RClC PUMP 1 SUCTION PRESS LOW will alarm.

C. ICS*MOV129, PMP 1 SUCT FROM CONDENSATE STOR TK will shut, then ICS*MOV136, PMP SUCT FROM SUPPRESSION POOL will open.

D. ICS*MOV136, PMP SUCT FROM SUPPRESSION POOL will open, then ICS*MOV129, PMP 1 SUCT FROM CONDENSATE STOR TK will shut.

Correct Answer: D With a CST Low Level condition, ICS*MOV136, PMP SUCT FROM SUPPRESSION POOL will open, then ICS*MOV129, PMP 1 SUCT FROM CONDENSATE STOR TK will shut.

Plausible Distractors:

A is plausible; would be true if RClC Injection Valve closure occurred.

B is plausible; would be true if Suppression Pool Suction valve remained shut.

C is plausible; identifies potential misconception on sequence of valve reposition events.

Page 15 of 88

ATTACHMENT 3 ( C o n t 1 2CEC*PMLGO1 SEP,IES 3 0 0 ALARM RESPONSE PROCEDURES K?flash: NO 2CEC

• PNLG 0 1 601343 RCTC CXDS TK IA XATER LEVEL LOW 343

- c r i n ou __

Source I IS e t I I I I I I I I

?OlIkt fcsLcos fl"IC CNST T K l A 2IC'S*LTIA AND 145'.5" a t Loot ATP, LVL 2ICS*LT3C v i a valve connecticn Relay E51A-NOTE: Relay ET;lA-I(80 is located i n 2 C E C

• P N L 6 2 1 Automatic' R e s ~ c m s e Opens ICS*MC)'ii13 6 , PPIP SUCT FROM S U P P R E S S I O N POOL.

4 A f t e r ICS*MOV136 is f u l l open, closer; ICS*HOVl:9, PMP 1 SUCT FRClM CONDENSATE STOR TK .

Ope r R tc r A c t i o n s

1. V e r i f y t l r e autcinatic response.

2. A t 2rEC-PNL851, moniVor Condensate StGraqe T k i C S T f 1A 1E levels.

3. I F CST l e v e l s a r e normal, monitor t r i p u n i t s E 5 1 - M 6 3 5 A & NJ E51-N635E, CST VXR LVL LO on 2 C E C

• P N L 6 2 9 for p o s s i b l e f a i l u r e .

4. I F alarm w a s caused by t r i p u n i t f a i l u r e , i n i t i a t e P I D AND contact I & C Department.

5. I F required, r e f i l l Cundcr:sata S t o r a g e Tanks p e r N2-OP-4.

NOTE : The amount of time t h a t R C I C i n j e c t s i n t o t.he FA. rvith s u c t i o n from the Suppression P o o l should be m i n i m i z e d .

• * * * + * * * * * * * * *

• f t * * * *

• t f *

• t *

• t C *

• t

• CAUTION S h i f t i r i q R C I C suctiori f r o m the Suppression Pool to the CST with R C I C operating i e NOT allowed.

• t * *

• t *

• f f *

• t t

• t * * *

• f * * * * * ~ * * * * * * *

5. I F required AND w i t h R C I C shutdown, s h i f t RCIC ~ U N Is u ~ ction by opening 21CSfMOV129 c l o s i n g 2ICS*MOVL3h.

7. Dispat.ch an ciperator t o check for l e a k s from CSTs kl.JD R C I C System.

45482,46536 Page 178 M2 -ARP- 0 1 Rev 00

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 13 2 218000 A3.01 4.2 N NA LOK Grp 10 CFR 55.41(b) 7 1-OD (1-5) Reference Documents H 1 ADS Logic Diagram Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including: ADS valve operation QUESTION 13 Following a MANUAL initiation of the Automatic Depressurization System (ADS),

the following conditions exist:

0 Reactor Pressure LOWERED to 80 p i g .

0 ALL Low Pressure ECCS Pumps have degraded, due to Suction Strainer clogging and have been TRIPPED.

0 ONLY the High Pressure ECCS Pump is STILL injecting.

0 RPV Water Level reached a MlNlMLlM of 15 inches and is 30 inches, RISING.

What is the status of the ADS Safety Relief Valves (SRVs)?

A. ADS SRVs remain OPEN, due to a logic seal-in.

B. ADS SRVs will SHUT, due to loss of logic seal-in.

C. ADS SRVs will SHUT, due to recovering RPV Water Level.

D. ADS SRVs will remain OPEN, because an ECCS Pump is providing a permissive.

Correct Answer: A Initially, LP ECCS Pumps were RUNNING. Manual Initiation switches energize K6A Initiation Relay AND K8A Seal In Relay. When K8A is energized, it seals itself in by closing a contact above the Seal In Reset pushbutton in the diagram. ADS Valves REMAIN OPEN, even if LP ECCS Pump permissives are subsequently LOST.

Plausible Distractors:

B is plausible; the seal in circuit IS still made up after the pumps trip, so the valves remain open.

C is plausible; RPV Water Level has restored above the actuation setpoint (RPV Water Level I ) , 17.8 inches and valves would close without seal in function on RPV Water Level, Manual Initiation, or LP ECCS Pump Discharge Pressure permissive.

D is plausible; High Pressure ECCS (HPCS) Pump does NOT provide a Discharge Pressure permissive signal to ADS.

Page 16 of 88

CHANNELS A AN D E I I (SEALIN) (LEVEL 1)

K12

-r

-L I-;

I

---I 513A , S13A S%JA SEAL IN SEALIN 5EhL 1N SEAL IN RESET RESET RESET RESET 5% AUTO $34 AUTO

~ ~ ~ T ~ A T l ~ ~

INHIBIT Student Guide (~2101218000~01) 2 1 of 148 Printed: 04/18/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 14 2 223002 A2.09 3.6 N NA LOK 1 Grp

~ . . ..........................................................

10 CFR 55.41(b) 7 i LOD (1-5) i Reference Documents H I I /  ! ARP602228 Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: System initiation QUESTION 14 The plant is operating at 1OO%, when the following occurs:

0 Annunciator 602228, MN STEAM LINE PIPE TUNNEL TEMP HI-HVDIFF TEMP HI, alarms.

0 Trip Units E31-N604A and B, MSL TUNNEL EL 263 FT indicate 170°F.

Which one of the following describes the PClS response to these annunciators, and what action is required?

A. ONLY the MSlVs will isolate. It is required to enter N2-SOP-101C, REACTOR SCRAM.

B. The MSlVs AND the INBOARD MSL Drains will isolate. It is required to enter N2-SOP-101C, REACTOR SCRAM.

C. The MSlVs AND the OUTBOARD MSL Drains will isolate. It is required to enter N2-SOP-101C, REACTOR SCRAM.

D. NO automatic actions will occur. It is required to Dispatch an operator to 2CES-IPNL202, TURBINE BLDG HVAC panel, to confirm proper HVAC operation.

Correct Answer: A With Logic inputs A and B tripped, ONLY the MSlVs will isolate. It is required to enter N2-SOP-IOIC REACTOR SCRAM.

Plausible Distractors:

B is plausible; would be true if B and C tripped.

C is plausible; would be true if A and D tripped.

D is plausible; would be true if temperature provided were below the isolation setpoint of 163°F.

Page 17 of 88

ATTACHVENT 4 I Cor1t,1 2CEC*PNL602 S E R I E S 2 0 0 ALARM RESWNSE PROCEDURES

1. Using the Procnas Computer OR rmsd i n d i c a t i n g liqhts 011 t l i e t r i p u n i t s a t 2CEC*FNLS09, 6 1 1 , G 3 2 6.12, c o n f i r m w I i i c h - d e t e c : t o r s have alarmed.

&. A s a p p l i c a b l e , v e r i f y t h e a u t o s a t i e zrrspoose

3. I F appliccible, m f e r tx, tAa E r n e r g m c y 0 p e ~ : a t i n gPro*r:edures !EOFJ'u) .

4. IF d Reector Scratrr GCCUL-O, e n t e r M%-SOP-lOlC:, REACTOR SCRAM.

5. D i s p a t c h an operator t.3 2CES-IPpTL202, TURP81NE &LUG HVAC panel, tr; coofirnl proper cperat.ioa af the f o l l o w i n g equipsent.:

ZKJT-UCZ1OA AEJO ITCZlOP, E T M I TTJbREL UNIT C!OCLEF:' Y.

• 5KJT-TUC?11, b1N 6:T LT, ENC'LOSIJRE UNIT COOLER.

21WI - FW1 1, STEX4 Tr.WNEL T R W S F E R FM4 ~

P d q e '57%

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO I Tier KIA Number Statement IR Origin Source Question 15 I 2 239002 K1.04 3.6 N NA 1

LOK Grp 10 CFR 55.41(b) 5 LOD (1-5) Reference Documents H ' 1 SG N2101239001C01 Knowledge of the physical connections and/or cause- effect relationships between RELIEFEAFETY VALVES and the following: Main steam QUESTION 15 The plant is operating at full power, when the following indications occurred:

Reactor Pressure LOWERED from 1020 psig to 1015 psig and stabilized.

0 Main Steam Line A Flow is 3.7 Mlbmlhr.

Main Steam Line B Flow is 2.9 Mlbm/hr.

Main Steam Line C Flow is 3.7 Mlbm/hr.

Main Steam Line D Flow is 3.7 Mlbmlhr.

Which one of the following failures is indicated?

A. OPEN Safety Relief Valve.

B. SHUT Turbine Control Valve.

C. OPEN Turbine Bypass Valve.

D. SHUT Main Steam Isolation Valve.

Correct Answer: A With LOW MSL Flow and LOWERING Reactor Pressure, an OPEN Safety Relief Valve is indicated.

Plausible Distractors:

B is plausible; Reactor Pressure would RISE and MSL Flows would be equal.

C is plausible; Reactor Pressure would remain the same and MSL Flows would remain balanced.

D is plausible; Reactor Pressure would RISE and MSL Flow would be ZERO in one MSL.

Page 18 of 88

IS} Discharge from each of the SRV's is directed to the suppression pool with the line terminating a t a T-quencher connection at the bottom of the suppression pool, The SRV discharge lines are arranged to provide an evenly distributed heat load in the suppression pool when a group of SRV's lift.

c, Safety Relief Vacuum Breakers

1) Two vacuum relief valves on each 5 R V discharge line 5erve to admit drywell atmosphere to the SRV discharge line, The relief valves prevent siphoning water int3 the SRV discharge pipe as it cools off after an opening cycle.

d. Reactor Head Vent

1) The vent from the reactor pressure vessel head is routed to main steam line "A" between the vessel and the first SRV during Power Operations. The head vent Is also routed to the Suppression Pool during Shutdown and Startup conditions. The connection for the steam supply to the Reactor Core Isolation Cooling system (ICs) turbine is on main steam line "B" upstream of the safety relief valves.

2) Main Steam Line Flow Restrictors a) A venturi-type flow nozzle is located in each main steam line in the vertical run of pipe immediately downstream of the last SRV I ,and upstream of the inboard M5IV. The purposes of these flow restrictors are to limit the steam flow and depressurization rate under the assumed conditions of a main steam llne rupture outside containment and t o provide a steam line flow signal for the following:

(1) Indication and control

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 16 2 239002 2.4.18 3.3 N NA LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents F 1 NER-2M-039 Rev 5 I I_ __ __

Safety Relief Valves - Emergency Procedures / Plan: Knowledge of the specific bases for EOPs.

QUESTION 16 The LEVEL Leg of N2-EOP-C5, Failure to Scram, contains the following override statement condition:

An SRV is open OR Drywell Pressure is above 1.68 psig Per the EOP Basis, which one of the following does this condition indicate?

A. Loss of Coolant Accident has occurred or an RPV Blowdown has been performed.

B. Heat is being added to the Suppression Pool by either the Downcomers or the SRV Tailpipes.

C. Reactor Coolant System is breached and injected Boron will NOT remain in the core.

D. Conditions exist which jeopardize the continued availability of Reactor Water Level Instruments.

Correct Answer: B These conditions indicate that heat is being added to the Suppression Pool by either the Downcomers or the SRV Tailpipes.

Plausible Distractors:

A is plausible; Drywell Pressure > 1.68 psig is a LOCA signal, and SRVs are open following an RPV Blowdown.

C is plausible; Drywell Pressure > I.68 psig indicates the RCS barrier is breached.

D is plausible; if SRVs are opened and Reactor Pressure lowered with conditions of High Drywell Temperature, Reactor Water Level Instruments may become unavailable.

Page 19 of 88

" W P - 2 EOP-C5 Basis EOP-CS Step (Ovemde L-5 secorid line and L-9)

Discussion The combination of high reactor power (above 4%, the APRM downscale trip), high suppression pool temperature (above 110°F,the Boron Injection Initiation Temperarure), and an open SKV or high drywell pressure (above 1.68 psig), are symptomatic of heat being rejected to the suppression pool at a rate in excess of that which can be removed by the suppression pool cooling system. Unless mitigated, these conditions ultimately result in loss oFNPSH for ECCS pumps taking suction on the suppression pool, containment overpressurization, and (ultimately) loss of primary containrnent integrity-which in turn could lead to a loss of adequate core cooling and uncontrolled release of radioactivity 10 the, environment.

The conditions listed in Step L-5, combined with the inability to shut down the reactor through control rod insertion, dictate a requirement to promptly reduce reactor power since, as long as these conditions exist,.suppression pool heatup will continue. Reactor power must be reduced to limit the heatup rate of the suppression pool.

NEN-2M-039. REV. 5 12-21

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 17 2 259002 A I .04 3.6 B NMP2 Bank SYSID: 22750 LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents H 1 SG N2101259002C01 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR WATER LEVEL CONTROL SYSTEM controls including: Reactor water level control controller indications QUESTION 17 The plant is operating at 100% with the following with FWS-PIA and FWS-PI B in service.

At 12:00:00, FWS-LVIOA has a loss of control power At 12:00:30 the following results:

0 Reactor scrams 0 Power LOWERS to 40%

0 SRVs are cycling to control Reactor Pressure Which one of the following describes the condition of FWS-LVIOB and FWS-LV1OA at 1201:OO, with NO operator actions?

FWS-LVIOA FWS-LV1OB A. Auto and Open Auto and Shut B. Auto and Open Auto and Open C. Manual and Open Manual and Shut D. Manual and Shut Manual and Shut Correct Answer: D 25 seconds after a High Reactor Pressure condition ( indicated by SRVs controlling pressure) with APRMs NOT Downscale, RRCS CLOSE signal will override the Loss of control signal and shut all FWS control valves.

Plausible Distractors:

A, B, and C are plausible; identify potential misconception on Loss of Control Power compounded by RRCS actuation. RRCS signal will shift all FWS valves to Remote Manual and close them for 25 seconds regardless of a Loss of Control Signal.

Page 20 of 88

B. SYSTEM DESCRIPTION The Redundant Reactivity Control (RRCS) System is designed to mitigate the potential consequences of an Anticipated Transient Without Scram (ATWS) event.

The RRCS System consists of Reactor Vessel pressure and level sensors, solid state logic, Control Room cabinets and indications, and interfaces with several systems which may be actuated to mitigate an ATWS event. The solid state logic is divided into Divisions 1 and 2, each of which is subdivided into two channels. The logic is energized to trip, and both channels of either division must be tripped in order to initiate the RRCS System protective actions. The system can be manually initiated by depressing two pushbuttons (tripping both channels) in the same division. This manual initiation function is designed so that no single operator action can result in an inadvertent initiation. The pushbutton's collar is rotated to arm the switch and depressing will trip the logic. The manual initiation pushbuttons are located on the vertical section of 2CEC*PNL603. There are four manual initiation pushbuttons for RRCS.

The RRCS System logic monitors Reactor dome pressure and water level. High pressure (1065 psig), low water level (Level 2 108.8 inches) or RRCS manual initiation will cause the Alternate Rod Insertion (ARI) valves to scram the Reactor independently of the Reactor Protection System I Low water level alone will, in addition to an ARI scram, cause an immediate Recirculation Pump Trip (RPT) by tripping the 60 Hz circuit breakers (in the normal supply lines to the recirculation pump motors), and the 15 Hz circuit breakers (in the LFMG supply to the pump motors). After 98 seconds of continued low water level and if the APRM channels are not downscale or are inoperative (INOP), the RRCS System initiates the Standby Liquid Control (SLS) System which isolates the Reactor Water Cleanup (WCS) System.

High pressure alone will, in addition to an ARI scram, immediately trip the 60 Hz circuit breakers and initiate transfer of the Recirculation Pumps to LFMG (low speed) operation. After 25 seconds, if the APRM channels are not downscale or are INOP, the RRCS System trips the 15 Hz circuit breakers to complete the RPT. It also initiates a feedwater runback and the feedwater min flow valves fail open which are both sealed in for 25 seconds. After an additional 73 seconds with the APRM channels still not downscale or are INOP, the RRCS System initiates the SLS System which isolates the WCS System.

Manual initiation alone causes an immediate ARI scram. After 98 seconds, if the APRM channels are not downscale or are INOP, the RRCS System initiates the SLS System which isolates the WCS System. Manual initiation does not cause an RPT or feedwater runback.

Page 3 N2-OP-36B Rev 03

0. SYSTEM DESCRIPTION The Redundant Reactivity Control (RRCS) System is designed to mitigate the potential consequences of an Anticipated Transient Without Scram (ATWS) event.

The RRCS System consists of Reactor Vessel pressure and level sensors, solid state logic, Control Room cabinets and indications, and interfaces with several systems which may be actuated to mitigate an ATWS event. The solid state logic is divided into Divisions 1 and 2, each of which is subdivided into two channels. The logic is energized to trip, and both channels of either division must be tripped in order to initiate the RRCS System protective actions. The system can be manually initiated by depressing two pushbuttons (tripping both channels) in the same division. This manual initiation function is designed so that no single operator action can result in an inadvertent initiation. The pushbutton's collar is rotated to arm the switch and depressing will trip the logic. The manual initiation pushbuttons are located on the vertical section of 2CEC*PNL603. There are four manual initiation pushbuttons for RRCS.

The RRCS System logic monitors Reactor dome pressure and water level. High pressure (1065 psig), low water level (Level 2 =: 108.8 inches) or RRCS manual initiation will cause the Alternate Rod Insertion (ARI) valves to scram the Reactor independently of the Reactor Protection System.

Low water level alone will, in addition to an ARI scram, cause an immediate Recirculation Pump Trip (RPT) by tripping the 60 Hz circuit breakers (in the normal supply lines to the recirculation pump motors), and the 15 Hz circuit breakers (in the LFMG supply to the pump motors). After 98 seconds of continued low water level and if the APRM channels are not downscale or are inoperative (INOP), the RRCS System initiates the Standby Liquid Control (SLS) System which isolates the Reactor Water Cleanup (WCS) System.

High pressure alone will, in addition to an ARI scram, immediately trip the 60 Hz circuit breakers and initiate transfer of the Recirculation Pumps to LFMG (low speed) operation. After 25 seconds, if the APRM channels are not downscale or are INOP, the RRCS System trips the 15 Hz circuit breakers to complete the RPT. It also initiates a feedwater runback and the feedwater min flow valves fail open which are both sealed in for 25 seconds. After an additional 73 seconds with the APRM channels still not downscale or are INOP, the RRCS System initiates the SLS System which isolates the WCS System.

Manual initiation alone causes an immediate ARI scram. After 98 seconds, if the APRM channels are not downscale or are INOP, the RRCS System initiates the SLS System which isolates the WCS System. Manual initiation does not cause an RPT or feedwater runback.

Page 3 N2-OP-366 Rev 03

b. High pressure/low flow control valves 2FWS-LV55A and B - only Reactor Feed Pumps 1A and 18 have an associated high pressure/low flow control valve and, therefore, would be used during startup.

c. High pressurefhigh flow control valves ZFWS-LVlOA, B and C - used during normal plant operation.

5. ZFWS-LVlOA (B,C) uses a variable frequency Limitorque motor.

a. When the valve position deviation error is 5% or greater, the motor operates at 10Oo/o speed (60 HZ and 18 seconds nominal full stroke time).

b. When the valve position deviation error is 2.5% or less, the motor operates at 10% motor speed ( 6 HZ and 180 seconds nominal full stroke time).

c. Motor speed and valve stroke time are variable between these values of deviation error.

d. The position feedback mechanism is driven from the actuator motor, NOT the valve stem.

e. ZFWS-LVlOA (B,C) will lock-up, As-Is, on a loss of control signal. The interlock must be manually reset after the alarm condition has cleared.

6. During an Anticipated Transient Without Scram (ATWS), the Redundant

+ Reactivity Control System will shift the FWLC System to manual and close the flow control valves.

a, The close signals will apply to all control valve actuators and will override the "loss of control signals".

b. Manual control of the feedwater control valves is available 25 seconds after the RRCS feedwater runback signal initiated, Student Guide : N Z : 0 1 3 9 ~ O Z C 0 1 ) 20 of 100 Printed: OS/l0/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier I WA Number I Statement

~ IR Origin Source Question 18 1 2 1 26 1000 j 2.2.42 3.9 B NMP2 Bank SYSID 13219 LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents F 1 TS 3.6.4.3 Amendment 101 SGTS - Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

QUESTION 18 The plant is in COLD SHUTDOWN with one train of Standby Gas Treatment (GTS) inoperable.

Which one of the following plant changes results in the requirement for BOTH trains of GTS to be operable?

A. Reactor Water Temperature unexpectedly rises above 200°F.

6. Loss of Secondary Containment Differential Pressure.

C. New Fuel is being placed in the Spent Fuel Pool.

D. Operational Mode is changed to MODE 5.

Correct Answer: A TWO GTS trains are required to be OPERABLE in MODES 1,2, and 3 (MODE 3 is RPV temperature above 200°F).

Plausible Distractors:

C is plausible; OPDRVs and irradiated fuel moves in secondary containment requires 2 GTS trains OPERABLE in MODES 4 or 5.

D is plausible; due to applicability of T.S. 3.6.4.3.

B is plausible; because new fuel is not "irradiated fuel".

Page 21 of 88

SGT System 3.6.4.3 3.6 C O ~ A I N M SYSTEMS E~

3.6.4.3 Standby Gas Treabnent (SGT) System LCO 3.6.4.3 Two SGT subsystems shJ be OPERABLE.

I,APPLlo4f31LITY: MODES 1,2, and 3, During movement ofrecentiy kradiatedfuel assemtdies in Ute I secondary containment During -idions with a potentialfor draining the reactor I vessel (OPDRVs).

ACTIONS C ~ D I ~ ~ N REQUIRED ACTION A. One SGT subsystem A. 1 R e s h SGT subsystem 7 days inoperable. to OPERABLE status.

B. RequiredAction and 8.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Conditkrn A natmetinMODE1,2, or 3. 8.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> I

C. RequiredAction and - II_ NOTE*-------

associatedCompletion LCO 3.0.3 is not applicable, Time of Condition A -*. *"*".*--------

not met during movement of recently irradiated C.1 Place OPERABLE SGT lmmediatdy fuel assemblies in We subsystemin secondary Gontainment operation.

or during DPDRVs.

OR (continued)

NMP2 3.6.4.3-4 ~ n d m e nW, t 101

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 19 2 26200 1 A4.02 3.4 M NMP2 Bank SYSID 13256 LOK

• Grp ' 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents F 1 N2-OP-1OOA Rev 9 Ability to manually operate and/or monitor in the control room: Synchroscope, including understanding of running and incoming voltages QUESTION 19 The plant is at 100% power with the following:

Division I Diesel Generator has been started and is being set up to perform a load test in parallel with 2ENS*SWG101 En accordance with N2-OP-IOOA. The operator is ready to parallel the engine with 2ENS*SWGIOl.

Which one of the following conditions will ensure the Division I Diesel Generator does NOT trip due to excessive Reactive Load?

A. SYNCHROSCOPE (EMERGENCY POWER-DIV I) is rotating slowly in the clockwise direction.

B. 4.16KV 2ENS*SWG101 INCOMING VOLTS indicates 5.20 kV AND 4KV RTX-XSRl A/ABSXl/ EGS*EGI RUNNING VOLTS indicates 4.20 kV.

C. 4.16KV 2ENS*SWG101 INCOMING VOLTS indicates 4.20 kV AND 4KV RTX-XSRlA/ABSXl/ EGS*EGI RUNNING VOLTS indicates 4.20 kV.

D. BREAKER 2ENS*SWG101-1 is closed when SYNCHROSCOPE (EMERGENCY POWER-DIV I) is at 5 minutes before the 12 o'clock position.

Correct Answer: C Voltage match prevents excessive Reactive Load during synchronization. Running Voltage is 2ENS*SWGIOl, Incoming Voltage is Division I Diesel Generator.

Plausible Distractors:

A is plausible; prevents overload trip condition from excessive REAL Load.

B is plausible; prevents REVERSE POWER trip condition.

D is plausible; prevents overcurrent trip condition.

Page 22 of 88

IF the Primary Feeder Breaker Synchroscope Circuit fails or ts inoperable, parallel bus transfer should be made across the Alternate Feeder Breaker. Attempts to match Bus and Generator speed with an inoperable synchroscope circuit may result in Diesel Generator Speed oscillations.

f

• f i

• X * *

• X *

• f

• Z i . * * * ~ * ~ * ~ ~ ~ * * *

• r ) * * ~ ~ ~ ~ * ~ ~ ~ ~ i * ~ ~ ~ ~ * ~ i ~ ~ * * ~ * ~

• 4.1 1 Place SYNCHRONIZE TO BUS 101 (103) switch to ON.

4.12 Using EMERGENCY DSL GEN 1 (3) VOLTAGE REGULATOR switch, verify voltage control by varying 4.16KV BUS 2ENSSWG101 (103) INCOMING VOLTS.

4.13 Using EMERGENCY DSL GEN 1 (3) GOVERNOR sv~itch.verify governor control by varying SYNCHROSCOPE indication.

Using VOLTAGE REGULATOR switch, match voltages on 4 16KV ZENS'SWGIOI (103)

INCOMING VOLTS meter AND 4KV RTX-XSRIA (Bj12ABS-XIIZEGS*EGl (3) RUNNING VOLTS meter 4.15 Adjust GOVERNOR switch to establish slow clockwise rotation on SYNCHROSCOPE (slow in fast directionj, as indicated by:

Meter movement between 112 to 1 inch per second

• 12 to 24 seconds for 360 degree rneter sweep 5: The following steps are for hading the Diesel Generator to run at any desired load, not to exceed rates given in Precaution 0.4.0.

4 16 IF 2ENSSWG101-1 (103-14). OUTPUT BREAKER 101-1 1103-141, is in PULL-TO-LOCK.

THEN WHEN SYNCHROSCOPE indicates 5 minutes before 12 o'clock (11 o'clock position) place Breaker Control Swrtch in Yormal-After-TRIP Page 40 N2-OP-1ODA Rev 09

Source Question: NMP2 Bank SYSID 13256 The plant is at 100% power with the following:

Division I Diesel Generator has been started and is being set up to perform a load test in parallel with 2ENS*SWG101 in accordance with N2-OP-100A.

The operator is ready to parallel the engine with 2ENS*SWGlOl Which one of the following conditions will ensure the Division I Diesel Generator picks up sufficient load to prevent a reverse power trip to the engine? MODIFICATION A. Synchroscope is rotating slowly in a counter-clockwise direction. OLD CORRECT B. Output breaker is closed when synchroscope is at 5 till noon position.

C. Synchroscope is rotating slowly in a clockwise direction.

D. Incoming and Running frequencies are equal. NEW CORRECT

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier KIA Number Statement IR Origin Source Question 20 2 262002 K3.07 2.6 N NA LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Document s F 1 N2-SOP-71 Rev 4 Knowledge of the effect that a loss or malfunction c, the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following: Movement of control rods: Plant-Specific QUESTION 20 With a LOSS of power from UPS I A , which one of the following prevents Control Rod movement?

Control Rod movement is prevented by a loss of power to the:

A. PMS Process Computer B. Intermediate Range Monitors C. Reactor Manual Control System D. Scram Discharge Volume Level Instrumentation.

Correct Answer: C The Reactor Manual Control System loses power, which prevents Control Rod movement.

Plausible Distractors:

A is plausible; The PMS Process Computer does not lose power from UPS 1A loss. It is powered from UPS G B is plausible; IRMs are powered from 24/48 VDC.

D is plausible; SDV Level Instruments do riot lose power from UPS 1A loss. The SDV Level Instruments are powered from UPS 3A/3B (GE Drawing 105E1503T4)

Page 23 of 88

5.0 DISCUSSION 51 The reason f o tripping

~ Divcsion I [livisron It Containment Purge Manual Isolation IS to satisfy ITS 3 3 6.1 Condition B Group 9 isolatron on SGTS Exhaust Radiation-High will not occur when 2VBB-UPSIA is de-energized The action to trip Division I AND Divmon I1 Containment Purge Manuel Isolations wtthin one hour conservatively meets the requirements af ITS 3 3.6 1 Condition B If Condition B

~

cannot be completed within the hour then ITS Table 3 3 6 1-1 (Function 2 c) directs you to Gondilron F. with the Required Action to rsolate the affected penetration flow pith@)within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the loss of the UPS 52 Impact of Losing UPSIA Loss of FWS!ERFEPDS Computer Loss of RMS CAB780 Vent GEMs Loss of RFtIS-PNL20O Stack and Vent GEMs Comtn Link to PGCC Loss of Stack and Vent GEP& "rinters Loss of !dul:iplexers for GETARs Shut down of "A" Recirc HPlls on loss of power supplies in PNL634 and "A* RCS Modicon Loss of Isolation Signal to CPS'AOVs 104-111; 1 19-122 from GTS-RUlOS Loss of Drywell Unit Coolers I A , IC, 2A, 2C, 3A (LOCA Override inhibited)

Loss of SRV Tail Piece Temp FtCDR Loss of Digital Displays Panel 503,873 Lass of Scram Time Panel 610 Loss 0%RWtEM, RPIS, RSCS Loss of CRD Tern;, RCDR and Alarm a i RX Manual Control I , LOSS Loss of Recorder Powx Loss of Power to Panel 619 Jet Pump Flows, Core D!P, Total IP Flow Page .5 N2-SOP-7 1 Rev #4

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement IR Origin Source Question 21 2 262002 K6.02 2.8 N NA LOK Grp 10 CFR 55.41 (b) 7 LOD (1-5) Reference Documents F 1 N2-SOP-04 Rev 2 Knowledge of the effect that a loss or malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) : D.C. electrical power QUESTION 21 2VBB-UPSIB has been placed on battery power with 2VBB-TRSI out of service for repair. DC Bus 2BYS-SWG001C is LOST.

Which one of the following describes the condition of Uninterruptible Power Supply (UPS) 2VBB-UPS1B loads following the DC Power Loss?

A. DE-ENERGIZED, due to loss of ALL redundant power sources.

6. ENERGIZED, because 2VBB-UPS1B Inverter Output is UNAFFECTED by this DC Power failure.

C. ENERGIZED, due to MAINTENANCE power AUTOMATICALLY aligned through the Static Switch.

D. DE-ENERGIZED until MAINTENANCE power is MANUALLY aligned through the 2VBB-UPS1B Maintenance Switch.

Correct Answer: C With NORMAL and UC (Inverter Input) Power lost, the Static Switch will automatically align Maintenance Power to UPS Loads. Loads will remain ENERGIZED.

Plausible Distractors:

A is plausible; would be true if MAINTENANCE Power Source was not available.

B is plausible; would be true if a DIFFERENT DC Power Source were lost. Inverter Output will be lost following a loss of BOTH Normal AC and DC Input Power.

D is plausible; would be true if the Static Switch was not available.

Page 24 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal

2. Specific Flow Path

a. 2VBB-UPSlA, 2VBB-UPS18, and 2VBB-UPSlG

1) The normal feed for 2VBB-lJPS1A (ZVBB-UPSlB, and 2VBB-UPSlG) is supplied through an automatic transfer switch 2VBB-TRS1. On loss of the normal supply (2N3S-US3) to 2VBB-TRS1, the transfer switch S t u d e n t Guide ( N ~ ~ ~ Z ~ ~ O O Z C O I ) 29 of 134 Printed: 02/12/2007 Page 26 of 97

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal will automatically transfer t o the alternate supply (2NJS-US4). When the normal supply is re-energized, the load will automatically transfer back after a 10-30 second time delay.

2) The static switch on 2VBB-UPSlA (2VBB-UPSlB, and ZVBB-UPSlG) is NOT a full time switch. The static switch uses two circuit breakers in the transfer sequence. Circuit breaker 3 (CB3) is located between the inverter and the loads, and circuit breaker 4 (CB4) is connected in parallel with the static switch to supply power from the maintenance supply to the loads.

a) With 2VBB-UPSlA (2VBB-UPS1B1 and 2VBB-UPSlG) in operation and supplying power to loads, CB3 will be closed and CB4 will be open.

b) I f the UPS inverter trips, the static switch will gate ON allowing CB3 and CB4 to change states, transferring the load to the maintenance supply. Since circuit breakers are electro-mechanical devices, and require a finite amount of time to open or close, the static switch is designed to pass power t o the UPS loads until CB3 can open isnd CB4 can close.

e) The static switch turns off once the breaker transfer takes place.

3) Loads a) 2VBB-UPSlA (1) Rod Block Monitor (2) APRMs (3) NMS recorders (P603')

(4) Recirc flow recorder (P6O2)

(5) RPIS power supply S t u d e n t Guide ( ~ ~ i u i ~ 2 0 n ~ t : r i i ) 30 of 134 Printed: 02/12/2007 Page 27 of 97

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 22 2 263000 K4.02 3.1 N NA LOK ' Grp 10 CFR 55.41(b) 7

• LOD (1-5) ' Reference Documents H 1 N2-SOP-04 Rev 1 Knowledge of D.C. ELECTRICAL DISTRIBUTION design feature(s) and/or interlocks which provide for the following: Breaker interlocks, permissives, bypasses and cross ties: Plant-Specific QUESTION 22 The plant is operating at full power, when DC Power from 2BYS*SWG002B is LOST.

Which one of the following describes the affect, if any, of this failure on Reactor Recirculation Pump Breakers?

A. ONLY B Reactor Recirculation Pump FAST Speed Breakers TRIP.

B. BOTH Reactor Recirculation Pumps FAST Speed Breakers TRIP.

C. ONLY B Reactor Recirculation Pump SLOW Speed Breakers have lost control power.

D. BOTH Reactor Recirculation Pumps FAST and SLOW Speed Breakers are unaffected.

Correct Answer: B Plant at full power implies FAST Speed Recirculation Pump operation. With a Loss of DC power from 2BYS*SWG002B, BOTH Reactor Recirculation Pumps FAST Speed Breakers TRIP.

Plausible Distractors:

A is plausible; would be true for loss of 2BYS-SWGOOIB (listed in N2-SOP-04 page 7)

C is plausible; identifies misconception about Reactor Recirculation I Loss of DC power breaker interlocks.

D is plausible; would be true for loss of 2CES-IPNL-414 Page 25 of 88

AUTOMATIC RESPONSES o IF running in high speed, BOTH Recirc Pumps trip to zero speed Trip of 2WCSPlA AMD FIB, CLEANUP PUklPS 0 Isolation of Group 8 AND 9 Outboard {Inboard) i V s c Loss of Divwon 1 (11) load shedding AND had s ~ ~ u ~ n ~ ~ ~ g

• Loss of ~ i v ~I (11)~ Electrical i o ~ DisBribuiron iodicatron, wntrol, interlocks AND protection Loss of 2EGS"EGl (EG3)

Loss of GTS Tram A (B)

• Lass of Df;p o ~tor 2 ~ B A ~ L ~(2Bf P~Z~

o toss of Divmoil I ill)A n n u n ~ a ~ s ~ s Lass of Drvrsion ill Electrical D i ~ t r i ~ u ~~i onn~ i c a ~control, i o ~ , interlocks AND p r o t e c t i ~

o Loss of 2EGYEG2, D i ~ i ~ i o111i lDiesel Generator

~ ~ n g111 Diesel Generator Pumps Loss of the f ~ l l ~ Division c 2EGWP3. DC T u r k Charger Lirb Oil Pump

• ISC Fuel P ~ ~Pump ~ t n ~

DC ~ i ~ ~ uOillpump ~ t ~ n ~

IF running in high speed trips "A" Rearc Pump to zero speed 0 IF running in Test mnde. trips off 2EGS"EGl AND EG3 e IF Main Turbine s p e d is L 1300 rpm, trips the &lain'Turbine 0 Trrp HVR Supply Fans which muses a Reactor Building isolation 0 TRP Z D ~ S ~ tC, ~ ~2 4i 2C, A AND 3A, DRYWELL UNIT COOLERS 0 Loss of ~TIvIL-PS,Emerge~cyBearing 01Pump LOSSof Df,ptlvger to 2~5B"lJPSlAAND UPSIC Loss of control power to 2 N ~ $ ~ supply ~ ~tlreakers

~ GAND 0 of~bus~protection

~ 109s ~ ~ ~

Loss of control power to 2 Y ~ ~ - ~ IMDS3, ~ S lL1DS5

, and FrtDSlO e Loss of ~ N P S - SAND ~ ~ l auto transfer ~ ~ p a b i l i ~

~ ~$'&GO03 Loss of Reactor Level Narrw Range "6" c Loss sf 811 225VDC Uain Turbine trips 28'/SSWG0018 -Attachment 6 Load List Trrp 2 D ~ S - U ~ID, l ~2R.

, 20, AND 38,DRYWELL I.1NlT COOLERS o Loss of 2GUO-P2, Emergency Seal Oil Pump e Lass of Of;power to 2VBB-IJPSID AND UPS36 o Lass of Reactor Level Narrokr Range '8" 0 Lass of 'Be F W. Header flat# indicationand input to FKLC System 0 Loss of control pouter to 2 N ~ S . ~ load ~ breakers

~ ~ ~ ~ ~ ~ 2 ~ 3 o Loss of mfiirol pavrer to 2YtJC-MDS2, !,IDS and MI3520 0 Loss d control power to SYXC-FdDSI 2BYS.SWG001C - Attachment 7 Load List Loss of DC power to ZVBB-UPSIB UPSlG AND UF'S3A Page 3 N2-SOP-04 Rev 01

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier KIA Number Statement IR Origin Source Question 23 2 264000 K3.01 4.2 B NMP2 Bank I SYSID 13261 LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents F 1 SG N2101264000C01 Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: Emergency core cooling systems QUESTION 23 Given the following conditions:

0 A Loss of Offsite Power and LOCA have occurred.

0 2EGS*EG3 TRIPS on overspeed during start and CANNOT be started.

0 2EGS*EG1 & 2 operate as designed and power their respective buses.

Which one of the following lists includes ALL the ECCS Pumps that have power available?

A. CSH*Pl, CSL*Pl , and RHS*PIA ONLY.

B. CSH*PI, RHS*PIB, and RHS*PIC ONLY.

C. CSH*PI, CSL*Pl, RHS*Pl B, and RHS*Pl C ONLY.

D. CSL*PI, RHS*PIA, RHS*Pl B, and RHS"P1C ONLY.

Correct Answer: A 2EGS*EG3 powers Division II ECCS Pumps, which are RHS*PI B, and RHS*PlC.

Plausible Distractors:

B is plausible; would be true for 2EGS-EG1 failure.

C is plausible; would be true for RHS*PIA TRIP with no EDG failures.

D is plausible; would be true for 2EGS-EG2 failure.

Obiective Link: 02-OPS-001-264-2-01 Page 26 of 88

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separate fusing identical to the normal control power circuit. (The normal circuits are defeated to prevent spurious operation during a fire and remote operation from outside Control Room provided in case Control Room is NOT habitable.)

3) Fuses are housed in non-conducting fuse blocks and plug in to a fuse block base installed inside the breaker cubicle. To the extent possible fuse blocks should be separated by pulling from straight on since torque applied a t angles stresses the fuse block and can cause damage t o non-conductive material.

2. Automatic Operation

a. The On-site Emergency AC Electrical Distribution System is automatic, and self diagnostic after start-up, and operator action is not required for normal operation,

b. Each of the three 4160 VAC emergency buses has a diesel generator to carry its loads in case of a loss of off-site power (LOOP) or in case of a sustained degraded voltage condition on the off-site source. Each emergency diesel generator is separate from and independent of the other diesel generators so that failure of one diesel generator will not impede the operation of the other diesel generators.

1) 2EGS"EGl supplies Division I bus 2ENS*SWGlO1

2) 2EGS*EG2 supplies Division 111 bus 2ENS*SWG102 II) 3) 2EGSXEG3 supplies Division I1 bus 2ENSXSWG103 c, Two levels of under voltage protection are provided at the 4160 VAC emergency buses; one t o detect loss of off-site power and one t o detect degraded voltage conditions.

Printed: OZllZf2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement IR Origin Source Question 24 2 264000 A4.04 3.7 N NA LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents H 1 N2-OP-I OOA Rev 9 Ability to manually operate andlor monitor in the control room: Manual start, loading, and stopping of emergency generator: Plant-Specific QUESTION 24 The plant is operating at 100% power with the following:

0 EMERGENCY DSL GEN 1 LOCA SIGNAL BYPASS switch is ON.

0 RPV Water Level LOWERS to 15 inches.

Which one of the following describes the affect of these conditions on the Division 1 EDG?

Division 1 EDG will:

A. NOT START when DIVISION IZEGS*EGI START switch is placed in START.

6. MANUALLY START when DIVISION 1 2EGS*EG1 START switch is placed in START.

C. AUTOMATICALLY START with NO Emergency Diesel Generator TRIP signals BYPASSED.

D. AUTOMATICALLY START with SOME Emergency Diesel Generator TRIP signals BYPASSED.

Correct Answer: B With the EMERGENCY DSL GEN 1 LOCA SIGNAL BYPASS switch ON, LOCA start signals are bypassed.

Plausible Distractors:

A is plausible; would be true for MAINTENANCE Mode.

C is plausible; identifies misconception about emergency trip bypass function and LOCA bypass function.

D is plausible; would be true with EMERGENCY DSL GEN 1 LOCA SIGNAL BYPASS OFF.

Page 27 of 88

D. PRECAUTIONS AND LIMITATIONS (Cont) 4.0 To avoid potential damage to Diesel Generator components, the following continuous and 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> short-term power output ratings must be strictly adhered to (refer to Dispositionfor DER 2 0633):

Continuous rating of 4400 KW should not be exceeded.

Continuous rating of 4520 KW shall not be exceeded.

9 Operation of Diesel Generator at loads of 4400-4520 KW shall be limited to 16000 hours between Engine overhauls (normally done every 18 months i24 months}).

• 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> short-term rating of 4840 KW should not be exceeded 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> short-term rating of 4950 KW shall not be exceeded.

Operation of Diesel Generator at loads of 4840-4950 KW shall be limited to 1630 hours0.0189 days <br />0.453 hours <br />0.0027 weeks <br />6.20215e-4 months <br /> between Engine overhauls (normally done every 18 months (24 monthsfj 5.0 If the Diesel Generator is shutdown during conditions where it is a major load for the Service Water System, Service Water Pumps may experience low flow.

60 Diesel Generator 2EGS"EGl and 2EGS*EG3 LOCA bypass switches shall be in OFF position unless a specific task requires the switches be placed in ON Placing the switches in the ON position will disable the following with or without Offsite Power available:

Automatic start of applicable Diesel Generator on LOCA signal Automatic start of applicable Emergency Core Cooling System components on a LOCA signal 7.0 A Diesel Generator becomes administrativelyinoperable when its associated Crane is out of its stored position. (2MHS-CRN3 for Division I and 2MHS-CRN2 for Division 11.)

8.0 Do not store sounding rod (dipstickj on the steps outside the Diesel Generator Rooms (south side).

Store sounding rod (dipstick) on the right side (east) of the Diesel Generator Building.

Page 16 N2-OP-IOOA Rev 09

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 25 2 300000 A2.01 2.9 B NMP2 Bank SYSID 13377 LOK ' Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents H 1 N2-SOP-19,4.2.5 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences

- of those abnormal operation: Air dryer and filter malfunctions QUESTION 25 The plant is operating at 100% power, when the following occurs:

Annunciator 603306, CRD SCRAM VALVE PILOT AIR HDR PRESS HIGH/LOW, alarmed.

0 Instrument Air (IAS) header pressure is 120 psig.

Auxiliary Operator reports the Scram Air Header pressure is steady at 62 psig.

0 NO control rods are drifting.

Which one of the following statements describes the action that is to be attempted to restore Scram Air Header pressure, per N2-SOP-19, Loss of Instrument Air?

A. Verify all IAS Compressors are loaded and bypass IAS Dryers.

B. Swap Scram Air Header Supply Filters and Pressure Control Valves.

C. Bypass Scram Air Header Supply Filters and Pressure Control Valves.

D. Verify all IAS Compressors are loaded and isolate Service Air Header.

Correct Answer: B Per N2-SOP-19, it is required to swap Scram Air Header Supply Filters and Pressure Control Valves.

Plausible Distractors:

A is plausible; would be true for Dryer Valve malfunction.

C is plausible, not procedural. Scram air header is required to be filtered and regulated at 70 -75 psig.

D is plausible; would be true if IA Header Pressure was below 85 psig, due to a SA Header rupture.

Objective Link: 02-OPS-001-279-2-00 Page 28 of 88

ATTAGHMENT 3 RESPONSE TO CRD SCRAM VALVE PILOT AIR HDR PRESS HlGHlLOW I0 IF Annunaator 603306, CRD SCRAM VALVE PILOT AIR HDR PRESS HIGHILOW, ES~n alarm, pertom the fallomng NiA Annunciator 603306 NOT in alarm c-1 11 Dfspatch an Operator to munrtor 2ROS-Pl133 scram air header pressure (RB el 261 "r (-1 12 IF pressure on 2RDS-PI133 lowers to 5 60 psg, SCRAM the Reactor per N2-SOP-101C L-1 13 Determrne pressure on 2lAS-PI191 AND 2RDS-PI133 NOTE: Annuncrator 603306 alarms at scram header pressure of 65 psiy and fowering

+' IF 21AS-P119-l~> 70 psig AND 2RDS-PI133 is < 65 psig, THEN swap the in-service scram air header supply filter AND pressure control vatvve as fallows MA, 21AS-Pi194 ISC: 70 psig OR 2RDS PI133 IS> 65 (-1 1.4.1 Open 2RDS-V2012f2014),PCL'-19A(B) OUTLET ISOL L) 1.4.2 Open 2RDS-V2011(20f3), PCV-IYA(3) INLET ISOF (-1 1.4.3 Sl~wlyC I O S ~ 2RDS-V2013{201l ) , PCV-ISB(A) INLET ISOL (-1 1.4.4 V e ~ f yZRDS-PCV1YAfB), AIR I'RESSURE CONTROLLER TO AOV123, 121, 130, 1312,mainta ns outlet pressure at 70-75 paig as indicated on 2RDS-PII33 L-)

145 Close 2RDS-V2014{2012], PCG\I-ISB(A)OUTLET 1SOL t-1 Page 1 2 N2-SOP-I9 Rev 01

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 26 2 400000 K1.02 3.2 B NMP2 2002 NRC Exam LOK Grp 10 CFR 55.41(b) 4 LOD (1-5) Reference Documents F 1 N2-SOP-13 Rev 2 Knowledge of the physical connections and / or cause-effect relationships between CCWS and the following: Loads cooled by CCWS QUESTION 26 Piping failures resulted in a loss of Reactor Building Closed Loop Cooling (CCP) flow. Which one of the following components can still be cooled following the loss of CCP?

A. Drywell Unit Coolers B. RDS Pump Seal Coolers C. RHS Pump Seal Coolers D. Drywell Equipment Drain Coolers Correct Answer: C RHS Pump Seal Coolers can also be supplied with Service Water.

Plausible Distractors:

A, B, and D are loads cooled by CCP only.

Page 29 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 27 2 201001 A4.06 2.8 N NA LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents F 2 SG N2101201001CO1 Ability to manually operate and/or monitor in the control room: SDV isolation valve test switch QUESTION 27 N2-OSP-RDS-Q001, SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVE OPERABILITY TEST is in progress. The two Scram Discharge Volume Vent and Drain Pilot Valve TEST pushbuttons are DEPRESSED.

Which one of the following describes system response?

Air is:

A. ALIGNED to the valve actuators causing the SDV Vent and Drain Valves to SHUT.

B. ALIGNED to the valve actuators causing the SDV Vent and Drain Valves to OPEN.

C. ISOLATED and VENTED from the valve actuators causing the SDV Vent and Drain Valves to SHUT.

D. ISOLATED and VENTED from the valve actuators causing the SDV Vent and Drain Valves to OPEN.

Correct Answer: C When the two Scram Discharge Volume Vent and Drain Pilot Valve TEST pushbuttons are DEPRESSED, air is ISOLATED and VENTED from the valve actuators causing the SDV Vent and Drain Valves to SHUT.

Plausible Distractors:

A, B, and D are plausible; and identify misconceptions regarding system response to TEST function.

Page 30 of 88

c) The CRD Pump 1A(Bf switch, C126-53A(B), is a four-position, spring-return t o NORMAL control switch. STOP and START positions control electrical power to the pump.

d) The two Scram Discharge Volume Vent and Drain Pilot Valve 4 pushbuttons are used for test purposes, Depressing both the TEST push-type switch will interrupt the RPS System A and B signals t o the valve(s). This opens the SDV Isolation valve (SOV-154 or 155), venting control air from t h e vent and drain valve operators and permitting these valves t o close.

2 ) The RMC and RPS systems have controls which affect the RDS system. These are covered in the RMC and RPS Lessons.

c. Interlocks

1) The RDS pumps will trip on low suction pressure a t 25 Hg absolute.

2) Level switches activate when water in the SDV exceeds three inches to provide an alarm, switches also activate if SDV level exceeds 16.5 inches to provide a Rod Withdrawal block.

3 ) Other level switches activate when water level in the SDV exceeds 43.4 inches by transmitter or 48.5 inches by level switch t o initiate a Reactor Scram. The scram is generated in anticipation of the SDV problem t o ensure sufficient volume t o allow for a complete Reactor Scram.

4) Four SDV high level trip bypass switches are located on P603. This allows for resetting of the Reactor Scram t o allow for draining of the SDV. In order for this bypass t o be effective the Reactor Mode Switch (2CEC*PNL603) must be in SHUTDOWN or REFUEL.

Student Guide ~ ~ 2 1 0 1 2 0 ~ 0 0 i c 0 1 ) 132 of 210 Printed: 03/09/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier KIA Number Statement IR Origin Source Question 28 2 215002 K6.04 2.8 B NMP2 Bank SYSID 13842 LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents H 2 SG N2101215003C01 Knowledge of the effect that a loss or malfunction of the following will have on the ROD BLOCK MONITOR SYSTEM : APRM reference channel: BWR-3,4,5 QUESTION 28 The plant is operating at 98% power, with the following:

0 Control Rod 34-19 (a non-peripheral rod) is selected for movement 0 APRM #2 fails DOWNSCALE.

0 NO operator actions are taken.

Which one of the following identifies the ability of Rod Block Monitor (RBM)

Channels A and B to generate a rod block on high local power?

RBM Channel A RBM Channel B A. Can generate blocks Can generate blocks B. Can generate blocks Cannot generate blocks C. Cannot generate blocks Can generate blocks D. Cannot generate blocks Cannot generate blocks Correct Answer: B APRM #2 is the Reference APRM for RBM Channel 6. If APRM #2 fails downscale, it would provide a signal to the RBM B that is below 30% power. This would result in RBM B being automatically bypassed, so no blocks will be generated.

RBM A is not affected Plausible Distractors:

A is plausible; would be true if a non-reference APRM failed DOWNSCALE.

C is plausible; would be true if APRM 1 failed DOWNSCALE.

D is plausible; would be true if a peripheral rod were SELECTED.

Objective Link: 02-OPS-001-215-2-06 Page 31 of 88

the OPRM trip are voted independently such that an APRM trip in one channel and a concurrent OPRM trip in another channel will not produce a reactor scram.

3. USAR Section 7.7.1.7, Neutron Monitoring System (TIP, SRM, RBM)

a. Each of the two RBM channels uses input signals from a number of LPRM channels. A trip signal from either channel initiates a rod block. One RBM channel can be bypassed without loss of subsystem function.

b. The RBM signal is generated by averaging a set of LPRM signals. One RBM channel averages the signals from LPRM detectors at the A and C positions in the assigned LPRM assemblies. The second RBM channel averages the signals from the LPRM detectors at the B and D positions.

Assignment of assemblies used in RBM averaging is controlled by the selection of control rods, I f a peripheral rod is selected, the RBM is automatically bypassed and the RBM output is set to zero. I f any LPRM detector assigned to a RBM is bypassed, the computed average signal is adjusted automatically to compensate for the number of LPRM input signals.

c. The RBM instruments receive APRM flux level from each APRM via fiber optic cables. The RBM uses the APRM flux from its reference APRM channel as the reference APRM flux for RBM calculations. When a control rod is selected, the gain of each RBI4 channel output is normalized to the reference APRM flux. The gain setting is held constant during the movement of that particular control rod to provide an indication of the change in the relative local power level. If the reference APRM flux used t o normalize the RBM reading is indicating less than 30-percent power, the RBM is zeroed and the RBM outputs are bypassed. The RBM Student Guide ( N Z I O I ~ O O ~ C O ~ ) 145 of 206 Printed: 04/04/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 29 2 204000 K1. I 6 2.8 N NA

................................. .i ........ .................................................... ............................................. ,................................ .................................... .................................................. ".............................................................

LOK 1 Grp I 10 CFR 55.41(b) 10 I LOD (1-5) Reference Documents i F i 2 ; i N2-SOP-30 Rev3 1 ................................ ..................................... ................................................................................................................................................................................................................................................................A i Knowledge of the physical connections and/or cause- effect relationships between

..A i REACTOR WATER CLEANUP SYSTEM and the following: CRD system: Plant-1 SDecific QUESTION 29 The plant is operating at full power when the following alarms occur:

0 Annunciator 603318, CRD PUMPS SUCTION FLTR DlFF PRESSURE HIGH 0 Annunciator 603309, CRD PUMP 1A SUCTION PRESS LOW.

0 Annunciator 603308, CRD PUMP IA l l B AUTO TRIP.

0 Annunciator 602324, RWCU PUMP CLG WTR TEMP HIGH.

Which one of the following actions is required?

A. ISOLATE the WCS System

6. START the standby CRD Pump.

C. PLACE the Reactor Mode Switch in SHUTDOWN.

D. THROTTLE WCS*MOV200 until in-service Filter Demins are in HOLD and TRIP the WCS Pump.

Correct Answer: D With NO CRD Pump operating and RWCU PUMP CLG WTR TEMP HIGH alarming, it is required to THROTTLE WCS*MOV200 until in-service Filter Demins are in HOLD.

Plausible Dist ractors:

A is plausible; Group 6 Isolation occurs due to a different High (NRHX Outlet)

Temperature condition.

B is plausible; Low Suction Pressure condition precludes a start of the standby CRD Pump until filters are swapped.

C is plausible; and required if Reactor Pressure were LOWER and a withdrawn Control Rod had an Accumulator Trouble.

Page 32 of 88

TEMP HIGH, is received.

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement IR Origin Source Question 30 2 214000 A3.01 3.4 N NA LOK

• Grp 10 CFR 55.41(b) 7 ' LOD (1-5) Reference Documents F 2 ARP 603443 Ability to monitor automatic operations of the ROD POSITION INFORMATION SYSTEM including: Full core display QUESTION 30 The plant is operating at 100% power, with the following:

0 Control rod 30-31 is withdrawn to 12 0 Rod 30-31 is selected at the Rod Select Module (P603) 0 RPS Trip System "B" is tripped due to a failed sensor 0 Scram Pilot Valve "A" solenoid fuse at the HCU for the rod blows Which one of the following describes the resulting 4 Rod Display rod position indication?

A. Double X B. Double Blank C. Double Dash D. Double Zero Correct Answer: B The rod is scrammed and inserted past the 00 reed switch and displays blanks, if selected on the RSM.

Plausible Distractors:

A is plausible; XX indicates lost RPlS due to bad reed switch C is plausible; - - indicates rod is passing through odd notch position.

D is plausible; 00 indicates rod is settled at full in notch position zero.

Page 33 of 88

f) SCRAM - a blue indicating light that is lit if both scram valves on the HCU associated with the rod have moved to the open position.

g) LPRM Display XX YY A(-D) UPSC - an amber indicating light that is lit if an LPRM reads upscale.

h) LPRM Display XX YY A(-D) DNSC - a white indicating light that is lit if an LPRM reads downscale.

8) Four-Rod Display Picture 7, Four-Rod Display a) Whenever any rod in a Four-rod group is selected on the RSM, that group is displayed on the four-rod display panel. A four-rod group consists of 4 control rods surrounded by a maximum of 4 LPRM detector strings.

b) Some four-rod groups located near the edge of the core may contain only 2 or 3 control rods and have less than 4 LPRM Student Guide ( ~ 2 1 0 1 2 0 1 0 0 2 ~ 0 1 ) 20 of 150 Printed: 03/09/2007

detector strings monitoring them. The rod t h a t has been selected will have a lit background. The vertical position of a rod is shown as a two digit number ("00" when rod is full in, "48" when fully withdrawn, or may be blank if still under insert (i.e.

Scram)). All notch positions show up as even numbers. Odd numbers are indicated by dashes (--). A double X (XX) indicates that the RPIS is receiving abnormal data. I f the selected rod is a member of a group with less than four rods the display window corresponding t o a rod that does not exist remains blank.

C. Operation Modes

1) I n the Operator Follow Mode, only five controls are used in maneuvering control rods. When a rod is a t rest, all directional control valves on a HCU are closed. I n order t o withdraw a rod one notch (6"),the operator momentarily depresses the WITHDRAW pushbutton.

Student Guide ( ~ 2 1 0 1 2 10 0 0 2 ~ 0 1 ) 2 1 of 150 Printed: 03/09/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 31 2 256000 K2.01 2.7 B NMP2 Bank SYSID 13196 LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents H 2 SG N2101256000C01

............. .. .i ................................................. . . ........................................... ........... .......... ..........., ........................................ ..... .. ...... ... ........... .......,.............................. . . . . . .................................. ........... ....................... ..........

Reactor Condensate : Knowledge of e'lectrical power supplies to the following: System pumps QUESTION 31 Given these conditions:

0 The reactor scrammed from 100%.

0 All plant systems have functioned as designed for the scram.

0 Condensate pumps " A and "B" are running.

Which one of the following describes the electrical power supply lineup to the Condensate Pumps 5 minutes after the reactor scram?

A. 2CNM-PIA from 2NPS-SWGOOI , 2CNM-PI B from 2NPS-SWG003.

These switchgears are powered by associated Reserve Transformers.

B. 2CNM-PIA from 2NNS-SWGOI 1 2CNM-PI B from 2NNS-SWG013.

These switchgears are powered by associated Reserve Transformers.

C. 2CNM-PIA from 2NPS-SWGOOI , 2CNM-PI B from 2NPS-SWG003.

These switchgears are powered by the Normal Station Service Transformer.

D. 2CNM-PIA from 2NNS-SWGOI 1 2CNM-PI B from 2NNS-SWGOI3.

These switchgears are powered by the Normal Station Service Transformer.

Correct Answer: B Within five minutes, bus transfers result in Reserve Transformers powering auxiliary loads. 2CNM-PIA from 2NNS-SWGOI 1,2CNM-P1 B from 2NNS-SWGOl3.

Plausible Distractors:

A is plausible; would be true for Feedwater Pumps.

C is plausible; would be true for Feedwater Pumps with NO scram.

D is plausible; would be true with NO scram.

Page 34 of 88

aker o r Fuse ST PKG EXHAUSTER 2N H S-MCCO 10-32 0 SUCTION VALVE, 2CNM-MOV666

~

COND PUMP B DISCH V, ZNHS-MCC010-33A 2CNM-MOV3B COND BOOSTER PUMP 2NHS-MCC010-338 SUCTION VALVE, 2CNM-MOV73 HTR STRING B OUTLET ZNHS-MCCO10-33C BLOCK V, 2CNM-WOV32B HTR STRING 6 INLET 2NHS-MCC010-33D BLOCK V, 2CNM-MOV33B A Condensate Pump Motor ZSCA-PNL103- 1.

Heater, ZCNM-H1A C Condensate Pump Motor ZSCA-PNL103-2 Heater, ZCNM-HlC B Condensate Booster Pump Motor Heater, 2CNM-H2B B Condensate Punip Motor 2SCA-PNL103-6 Heater, ZCNM-H1B A Condensate Booster ZSCA-PNL103-7 Punip Motor Heater, 2CNM-H2A C Condensate Booster ZSCA-PNL103-8 Pump Motor Heater, 2CNM-H2C CONDENSATE PUMP A, 2NNS-SWGOll-7 2CNM-P1A CONDENSATE PUMP C, ZNNS-SWGO11-3 ZCNM-P1C Student Guide (N2lolz56aoncCll) 41 of 158 Printed: 05/04/2007

CONDENSATE PUMP C, ZNNS-SWG013-2 2CNM-PlC CONDENSATE PUMP B, ZNNS-SWG013-3 2CNM-P 1%

CONDENSATE BOOSTER ZMPS-SWGD01-7 PUMP, 2CNM-P2A CONDENSATE BOOSTER ZNPS-SWG001- 12 PUMP C, 2CNM-P2C CONDEDJSATE BOOSTER ZNPS-SWG003- 5 PUMP B, 2CNM-P2B TO 2NPS-SWG001 2NPS-SWG003-11 CONDENSATE BOOSTER PUMP, 2CNN-P2C COND BOOSTER PUMP ZNHS-MCCOlO-:L 1A AUX O I L PUMP, 2CNO-P2A COND BOOSTER PUMP 2NHS-MCC010-28E AUX OIL PUMP, 2CNO-P2B COND BOOSTER PUMP ZNHS-MCCODJ-ZA AUX OIL PUMP, 2CNO-P2C Low Energy Feedwater ZSCI-PNLB101-24 Cleanup Control Valve, ZFWS-HVXll3 Low Energy Feedwater ZSCI-PNLB101-24 Cleanup Control Valve, ZFWS-HVYI 13 High Energy Feedwater 2SCI-PNL6101-24 Cleanup Control Valve, 2ZFWS-HVX111 High Energy Feedwater 2 x 1 - P N Lr3101-24 Cleanup Control Valve, 2FWS-HVY 111 Student Guide {~15'10125600(:1r301) 4 2 of 158 Printed: 05/04/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 32 2 2 16000 A2.04 2.9 N NA LOK Grp 10 CFR 55.41(b) 5 LOD (1-5) Reference Documents H 2 N2-SOP-06 Rev 4 Nuclear Boiler Instrumentation - Ability to (a) predict the impacts of the following on the NUCLEAR BOILER INSTRUMENTATION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Detector diaphragm failure or leakage QUESTION 32 Which one of the following indicates that the normally SELECTED Narrow Range RPV Water Level detector has experienced a diaphragm failure, and what action is required?

A. Narrow Range A indicates 186 inches, RISING.

Narrow Range B indicates 179 inches, LOWERING.

Narrow Range C indicates 187 inches, RISING.

It is required to select Narrow Range A for Feedwater Level Control.

6. Narrow Range A indicates 186 inches, RISING.

Narrow Range B indicates 179 inches, LOWERING.

Narrow Range C indicates 179 inches, LOWERING.

It is required to select Narrow Range B for Feedwater Level Control.

C. Narrow Range A indicates 186 inches, RISING.

Narrow Range B indicates 186 inches, RISING.

Narrow Range C indicates 186 inches, RISING.

It is required to scram the reactor and trip the Main Turbine.

D. Narrow Range A indicates 179 inches, LOWERING.

Narrow Range B indicates 179 inches, LOWERING.

Narrow Range C indicates 179 inches, LOWERING.

It is required to scram the reactor.

Correct Answer: B with NR A rising, and redundant NR B and C lowering, NR A diaphragm failure is indicated, this can be mitigated by swapping to NR B.

Page 35 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal Plausible Distractors:

A is plausible; variable leg leak on NR B (with B SELECTED) is indicated.

C is plausible; feedwater flow instrument failure indicated with three element control.

D is plausible; steam flow instrument failure indicated with three element control.

Page 36 of 88

5. Nuclear Boiler Instrumentation (ISC) (N2-OP-34)

a. The Feedwater Control System receives indicated water level signals from the Nuclear Boiler Instrumentation System.

b. I f the selected narrow range level indication fails, the Feedwater Control System will be adversely affected in any automatic mode of operation (both single- and three-element control).

c. I f the indicated level signal fails high, the Feedwater Control System will shut the flow control valves in an attempt t o restore level. Without intervention, the low level scram setpoint will be reached.

d. I f the indicated level signal fails low, the Feedwater Control System will open the flow control valves in an attempt t o restore level. Without intervention, the high level turbine trip setpoint will be reached.

D. Practice

1. Describe the affect of a loss of Instrument Air t o the FWLCS.

Student Guide ( ~ 2 1 0 1 2 5 9 0 0 2 ~ 0 1 ) 75 of 100 Printed: 05/10/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 33 2 22600 1 K5.02 2.6 N NA LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents F 2 N2-OP-31 Rev 15 Knowledge of the operational implications of the following concepts as they apply to RHWLPCI: CONTAINMENT SPRAY SYSTEM MODE : Water hammer QUESTION 33 Given the following:

0 LOCA signal is present.

0 RHS Loop A is aligned for Suppression Pool Spray and Drywell Sprays.

0 A LOSS of Reserve Transformer 1A occurs.

0 Emergency Diesel Generators start and re-energize their respective busses.

Which one of the following describes the affect of these conditions on RHS*PIA and the operational hazard associated with this sequence of events per N2-OP-31, Residual Heat Removal System?

A. DOES NOT AUTOMATICALLY RESTART. Primary Containment may be DAMAGED by the lack of Suppression Pool and Drywell Sprays.

B. AUTOMATICALLY RESTARTS. RHS Pump may be DAMAGED by excessive starting current from receiving an automatic start signal with Suppression Pool and Drywell Spray Valves OPEN.

C. AUTOMATICALLY RESTARTS. KHS piping may be DAMAGED by water hammer caused by RHS Pump starting after piping drained through the OPEN Suppression Pool and Drywell Spray Valves.

D. DOES NOT AUTOMATICALLY RESTART. Primary Containment may be DAMAGED by steam bypassing of Suppression Pool Downcomers through the OPEN Suppression Pool and Drywell Spray Valves.

Correct Answer: C per P&L 31 .O of N2-0P-31, A potential for water hammer exists when the RHR pumps are restarted after a pump trip while operating in the Suppression Pool Cooling, Suppression Pool Spray, OR Flow Test modes. Following a loss of power, AND prior to pump restart, water can drain down from the piping into the Suppression Pool creating voids. A restart of the pump without refilling the drained piping can result in severe water hammer.

Page 37 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal Plausible Distractors:

A is plausible; the RHS Pump will AUTOMATICALLY start. Suppression Pool and Drywell Sprays are not required to maintain Primary Containment integrity following a Loss of Coolant Accident.

B is plausible; RHS Pump Breakers have overcurrent trip relays designed to protect the pump motor from overcurrent conditions, irrespective of the cause.

D is plausible; the RHS Pump will AUTOMATICALLY start. Piping with open valves will be connected to both Drywell and Suppression Pool, there is NO such procedural precaution.

Page 38 of 88

D. PRECAUTIONS AND LLMITATLOMS (Ccrrit j 28.0 I f r a d i a t i o n m o n i t o r S;*:P13k !E>, becomes u n a v a i l a b l e or 1s INOF, e I 1 S U r e r a d i a t i c n monitor SKG146A i B : is on line and c q e r a t l n g properly.

29.0 Reactor C o o l a n t and Spent F u r 1 P o o l temleiature s h a l l be maintained

> 7 0 O F a t all times.

30.0 To minimize t h p o t e n t i a l f o r watar h a m m e r i n mcre than one RHK loop c o i r r c i d z n t with a LOOP,LCC'A event, clo NOT operata two loap.3 concur1 Pritl:; i n Supprebsiori I-'ool Cooling, Suppression P c a l Spray, OR Flow T e s t mode \ E x c e p t as J i r e c t e d by EOPs, .

31.0 A potential for water hammer e x i s t s wlzen t h e RHR pumps a r e r e s t a t e d a f t e x a pimp trip w h i l ~o p e r a t i n g i n the $kippressic.in P ~ o lCocling, Suppression Pacl Spray, OR FLow T e s t modes. F o l l c w i n y x l o s s of II, p c ' e z , mJD pricjr to p n i p restdrt, water c a n d r a i n dcwn from the p i p i n g i n t o the SupFxrssion Pool c*rsatiriq v o i d s . A r e s t a r t o f the pump withc-rut refilling t h e dra1114 p i p i r i g an r e s u l t i n c'P"sre wat el' L F _ J-.

tiammer .

32. 0 A p o t e r i t i a l f c r w a t e ~t i a m m e i exists when, a f t e r warming uil shutdown cooling, Reactor c.ooldowxi/dzprrssurizatiori coritinues, prior to cut t i n 3 i n shutdown c m l i n g . This r e s u l t s in shutdown i.onling being at R 1iigheL temperature t h a n R e a c t a r , arid water hammer could r?milt when c u t t i n y i n sliutdcxn i ~ o o ing l ,

33.0 Khen t h e Emercpncy Diesel Gwierator is power iiig t h e E m e r y e n c y Bus, r e f e r t o N ~ - S O P - ( J Sf o r staitirrc of an RHR-pump c r i the ED<.

34.0 Any o p e r a t i o n of this systenr diiring Mode 3 { ie . S t a L t u p caE Shiltdowi (C2) C o o 11ng 1 shall re qu i ri c onei ir r en t t e rf oL nian of NZ -OF- 1G1C ,

Attachment I . In addition, manipulations i n WxIe 2 are r e q u i r e d t o have Tndependent V r r r t i c a t i s performed. These 1-eqiiir%wnta are only in e f f e c t I F RDS Eackf 1 I n J e c t l o l l 1 8 nut o f h e L v i C e to One or t n o r e RPV Level refrrrnze Le

?5.0 Do riot close 2RHS*7432A!B) , S t e a m Cmidensing Manual I,-;olat ion '?alve, w h e n r ? a r t o r p r e ~ s ~ i ri-s q r + 3 t e r thaii 5 O Q PSIG and the RCIC Steam Supply Ccntaiimient I s o l a t irln Valves d i e opari, u n l e s s a drain path 2 I" i s c p e n p r i o r t 3 dnd d u r i n g the p e r i o d t h a t 2 R H S

• V 4 3 2 A i B ) is C l ~ S e d . CltheYdise, the ~ ~ c ~ t - t r lei:lsts ti~l to O V ~ T ~ Y ~ S S U1111% ~ ' ~2RHS-Z ~

0 0 8 - 5 4 - 2 ! A - 1 01' LKHS-038-57-2(B-).

A f i r e i n Zona 1 3 8 SW, R.eactor B u i l c l i r q E l . 2 4 0 ' ~oi.11~1 render 2RHS*MOV113, Shutdown C c r o l i i i g S u e t i o n outside i s o l a t i o n valve electrically inoperable. Mmudl operation of the v 2 l - e cr the u s e cf Al t ernd t e $::hutdown C"cm1ing indy be u e c p i red .

3'7.0 A f i r p i i i Zona 255 SX, R e a c t o 1 Pdlding Sciitfi El. 2 8 9 ' c o u l d ~ - e r i d r ~

(C5) 2RHS*MG1JLO4, F?HR B tc Ecactcr Head .Spla;~ e l e c t r i c a l l p incpt?rak.la.

Manrial clieraticn of t h e valw m a y be ~ c y u i r e d .

38.0 I f a l l control rods are not f u l l y i n s e r t e d , Steam Condensing i s r i o t to be placed i n opeiaticixi, u n l c s x directed by ECjP' s.

Page 1 5 N2 -OP- 3 1 R e v 15

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 34 2 234000 2.1.23 4.3 B NMP2 Bank SYSlD 4809 LOK Grp 10 CFR 55.41(b) 10 LQD (1-5) Reference Documents H 2 N2-OP-39 section 5.3 Fuel Handling Equipment - Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

QUESTION 34 A CORE RELOAD is in progress with the following:

A fuel assembly is to be removed from the Spent Fuel Pool and transferred to the Reactor Cavity.

The fuel assembly is grappled and raised to Normal-Up (digital readout =

000.00).

Refueling bridge is moved in the reverse direction. Proximity switch (LSI) has actuated indicating the refueling bridge is near the reactor.

The following indications are received on the INTERLOCK STATUS DISPLAY panel:

0 ROD BLOCK INTERLOCK # I 0 BRIDGE REVERSE STOP # I 0 FUEL HOIST INTERLOCK Which one of the following caused the INTERLOCK STATUS DISPLAY panel indications?

A. At least one control rod is NOT fully inserted into the reactor core.

B. Grapple ENGAGE/RELEASE switch was momentarily placed in the RELEASE position.

C. Refueling Bridge (mast) position is outside the safe region of the Boundary Zone Controller.

D. Grapple RAISE/LOWER control was positioned to LOWER while Bridge REVERSE motion was still commanded.

Correct Answer: A These conditions indicate that at least one control rod is NOT fully inserted into the reactor core.

Page 39 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal Plausible Distractors:

B is plausible; causes a RAISE BLOCK.

C is plausible; causes a SAFETY TRAVEL INTERLOCK.

D is plausible; causes a SAFETY TRAVEL INTERLOCK.

Objective Link: 02-OPS-001-234-2-01 Page 40 of 88

6. SYSTEM DESCRIPTION (Cont) 5.3 (Cont)

MONO AUX. HOIST INTERLOCK - When this light is lit, the Monorail Aux Hoist receives a Raise Block and a Lower Block and can not be moved. This light is lit when:

Any Control Rod I:, withdrawn. and Monorail Aux Horst Load is >400 pounds, and The Bridge approaches the Reactor Vessel (Limit Switch 2FNR-ZS1 (LSI) is picked up)

TROLLEY AUX HOIST INTERLOCK - When this tight is lit the Frame Mounted Aux Hoist receives a Raise Block and a Lower Block and can not be moved This light IS lit when Any Control Rod is withdrawn.

Frame Mounted Aux Hoist Load IS>400pounds, The Bridge approaches the Reactor Vessel (Limit Switch 2FNR-ZS1 (LSI) is picked up)

ROD BLOCK INTERLOCK $1 - When this light is lit a Control Rod Withdrawal Block signal is sent to the Reactor Manual Control System in the Control Room This light is lit when Any Hoist is loaded (Main Hoist >700 pounds Aux Hoists >400 pounds).

and The Bridge approaches the Reactor Vessel [Lmilt Switch 2FNR-ZS1 (LSI) is picked up)

ROD BLOCK INTERLOCK #2 - When this light is lit a Control Rod Withdrawal Block signal is sent to the Reactor Manual Control System in the Control Room This light 1s lit when.

Any Hoist is loaded (Main Hoist ~ 7 0 0pounds Aux Hoists >400 pounds) and The Bridge is over the Reactor Vessel (Limit Switch 2FNR-ZS2 (LS2) is picked up)

BRIDGE REV STOP $1 - When this light is lit Bridge movement towards the Reactor Vessel is blocked This light will be lit when Any Control Rod ISwithdrawn, Any Hoist is loaded (Main Hoist >700 pounds Aux Hoists >400 pounds) and Page 26 N2-OP-39 Rev 08

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier KIA Number Statement IR Origin Source Question 35 2 239001 A I .01 3.6 N NA LOK Grp 10 CFR 55.41 (b) 5 LOD (1-5) Reference Documents H 2 N2-SOP-23 Rev 4 Ability to predict andlor monitor changes in parameters associated with operating the MAIN AND REHEAT STEAM SYSTEM controls including: Main steam pressure QUESTION 35 The plant is operating at 85% power, with the following:

Turbine Throttle Pressure Transmitter for the in-service EHC regulator fails HIGH.

Which one of the following describes the affect of this failure on initial reactor pressure and pressure control?

A. RISES and the reactor continues to operate at power. Pressure will be controlled by operation of the Control Valves.

B. RISES until an automatic Reactor Scram occurs. Pressure will then be controlled by operation of the Bypass Valves.

C. LOWERS until an automatic Reactor Scram occurs. Pressure will then be controlled by operation of the Bypass Valves.

D. LOWERS until an automatic Reactor Scram occurs. Pressure will then be controlled by operation of the Safety Relief Valves.

Correct Answer: D When the Turbine Throttle Pressure Transmitter fails HIGH, Turbine Control Valves will OPEN. This causes Reactor Pressure and Power to LOWER. The automatic scram is accompanied by MSlV closure. With MSlVs closed, Reactor Pressure is controlled by SRVs.

................................................................................................................................................................................................................................................................t Plausible Distractors:

A is plausible; would be true for a LOW failure of the Turbine Throttle Pressure Transmitter. TCV closure causes void collapse which raise Reactor Pressure and Power.

B is plausible; would be true for closure of ONE MSlV or similar steam flow reduction.

Void collapse will raise Reactor Pressure and Power, after the scram, Bypass Valves will control Reactor Pressure.

C is plausible; identifies misconception about concurrent MSlV closure.

Page 41 of 88

D. Explanation

1. Detectors and Monitoring Instrumentation

a. Pressure Picture 63, PNL 8 5 1 Pressure Indicators

1) Two pressure transmitters (PT 143/144) sense the pressure averaging manifold pressure for the pressure control unit, and provide pressure indication on P851 EHC panel.

2) A narrow range pressure indicator PT 238 (880 - 1050 psig) is also provided on Panel 8 5 1 which also senses pressure averaging manifold pressure.

3) Pressure indicated shows the effect of the A and B EHC pressure regulator setpoints is indicated on the 8 5 1 Panel S t u d e n t Guide (N2101248000C01) 160 of 274 Printed: 04/24/2007

b) I f a stop valve failed t o close during a turbine trip, the control valves closing on the scram signal would isolate steam flow from the turbine.

5) Steam Flow a) Pressure Transmitter Fails HighlLow:

(1) I f the 'A" pressure transmitter PT143 output fails "high" the "A" pressure regulator will attempt t o reduce the pressure signal by generating a large pressure error signal. This large pressure error signal will pass through the High Value Gate and be converted t o a large steam flow demand signal.

This will result in the Control Valves opening t o the 100°/~

open position and one bypass valve partially opening t o reach the combined flow limit of 115% o f normal turbine steam flow. The large increase in steam flow will cause RPV water level t o swell and RPV pressure t o drop. The reactor will scram when either the MSIVs close a t 766 psig or when the main turbine trips a t level 8. The closure of the MSIVs will isolate steam t o the bypass valves and require the use of other systems (i.e. SRVs) t o regulate and control RPV pressure.

(2) Should the 'A" pressure transmitter fail "low" the High Valve Gate connecting the "A" and "B" regulator outputs will block the 'A" regulator signal and pass the "6" regulator signal.

Due t o the + 3 psi bias on the "B" regulator the steam valves will be positioned t o stabilize pressure a t a value 3 psi higher than the original value.

6) Loss of condenser vacuum Student Guide (~zioi248000~01) 216 of 274 Printed: 04/24/2007

5.0 DtSCUSStON 5.1 If a miatfunchon occurs cauang a failure of the in-service Pfessuse Regulator in a valves dose direction resutting xn rasmg reactor pressure, the Backup Pressure Regulator will assume control of pressure at a slightly higher setpornt (5 to 20 psig higher depending on bias setting). The operator should be alerted to this event by Annunciator 851 148. B PRESS REGULATOR IN CONTROL This annunciator rviH alarm if the transfer of control 18fTotn A Regulator to B Regulator, and wilt clear rf transfer is from !3 to A If Reactor Pressure contrnues to rise, the Reactor ivtll scram at 1052 psg 5.2 If a malfunction occurs caustng a failure of either Pressure Regulator tn the valves open direction resultrng rn lowering reactor prwitire, the operator should be initially alerted 40 this event by the TURBINE BYPASS VALVE OPEN and TURBINE TROllBLE annunciators Observatioxnof valve position should confirm that Turbine Control and Bypass valves are open as much as allowed by the Maximum Conibrned Flow Limrter Reactor pre3sure will continue to lower until either the Operator closes the MSIWs or low Reactor pressure in the RUN mode a w e s MSIV closure 5.3 If the Pressure Regulator fails in a condition which muses Reactor pressure to lower, Turbine Control and Bypass valves will be open as much as allowed by the Maximum Conibined Flow Limiter. The operator will have nu othe. m a n s to control Reactor pressure and is therefore II, directed to Scram the Reactor and close the PJSIVs to prevent further depressurization of the Reactor This would occur rn any event without operator action since Turbine trip wrlf occur at Level % or the MSlUs will close when pressure lowers to 766 pig while the Idode smtch IS in the RUN position, fotlowed by Reacfor !$cram due to Turbrne Stop Valve or F,SSIV closure. If the Reactor is in the STARTUP mode Turbine trip or MSW isolatron wilh resultant Reactor Scrani will not occur automatically therefore the operator is directed under all conditrons to place !he Mode sbvitch to SHUTDOWN and close the k1SIVs 5.4 When directed to verify closed the Outhd MSlVs, the operator wilt place the following b&31\1 control swtches to the CLOSE position at panel 2CECPNE602

• liilSIV TiEiS*AOV7A MSlV EivlSS'AOV7B

• M5IV M S S'AOVTC

• M51V MSS"AOV70 If Reactor pressure continues to h e r , the operator should then close the lnbd F,lSlVs by placing the folfowing MSIV control switches to the CLOSE position at panel 2CECPNL602.

5.5 Computer PT MSSPA16 may be used to sliondor Turbine Throttle pressure Page 4 N2-SOP-23 Rev 04

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 36 2 259001 K4.11 3.5 B NMP2 Bank SYSID 12921 LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents H 2 N2-0P-29 Reactor Feedwater - Knowledge of REACTOR FEEDWATER SYSTEM design feature(s) and/or interlocks which provide for the following: Recirculation runbacks QUESTION 36 Power ascension is in progress with power at 50%. Preparations are being made to start the second Feed Pump. The following conditions exist:

0 Feed Pump A is in service.

0 FWLC Narrow Range "A" is selected.

0 FWLC Master Level Controller is in MANUAL.

0 Both Recirc Pumps (RRPs) are on High Speed.

0 Both RCS Flow Control Valves (FCVs) are at 22% open.

0 THEN FWLC Narrow Range "A" fails DOWNSCALE.

0 Reactor Operator maintains Reactor Water Level 180 to 185 inches.

Which one of the following describes the effect on the Reactor Recirculation System Flow Control Valves and Pumps?

A. Motion Inhibit is generated and RRPs remain on high speed.

B. Motion Inhibit is generated and RRPs downshift to low speed.

C. FCV Runback signal generated and RRPs remain on high speed.

D. FCV Runback signal generated and RRPs downshift to low speed.

Correct Answer: D FCVs runback to minimum due to L4 with only one feedpump operating; Recirc pumps shift to slow speed due to L3. Both functions come from failure 1 of the SELECTED FWLC NR transmitter.

................................................................................................................................................................................................................................................................/ .................................................................................... :

Plausible Distractors:

A is plausible; Motion Inhibit will not occur. Recirc Pumps downshift to slow speed.

B is plausible; Motion Inhibit will not occur.

C is plausible; Recirc Pumps downshift to slow speed.

Objective Link: 02-OPS-001-202-2-02 Page 42 of 88

D. PRECAUTIONS AND LIMITATIONS (Cont) 8.0 The following conditions will result in the transfer OX a reactor recirculation pump from 60 Hz to 15 Hz:

a. High to low speed transfer (BRKR 5A and 58 positioned to LFMG)

b. During a low speed start when pump speed reaches 95'/0.

c. Main turbine stop valves closure (less that- 95% open) or control valve closure (less than 530 psi EHC pressure) when the reactor power is greater than 30%.

d. A differential temperature between the reactor recirculation pump suction and the steam dome less than 10.7"F for 45 seconds.

e. Redundant Reactivity Control System (RRCS) high reactor pressure 1065 psig

+ f.

g.

Low water level 159.3" (level 3).

Total feed flow less than 3.35 million lbihr IApprox. 22.3% power) for 15 seconds.

90 To prevent exceeding reactor vessel AND recirculation system thermal shock interlocks when starting the recirculation pump (thermal shock interlocks for both pumps must be met to satisfy start logic), verify the following temperatures AND tlow rates within 15 minutes prior to starting the idle recirculation pump AND enter the applicable temperature dtfferencesAND time they were checked in the CSO log 9.1 The following Technical Specificattons (TS) requirements apply per TS SR 3 4 11 3 and 3 4 11 4 for D 9 1 a and b and USAR Section 7 7 1 2 for D 9 1 c

a. Bottom head coolant temperature AND the Reactor Pressure Vessel (RPV) coolant temperature AT IS I145°F.

2. Bottom head coolant temperature can be determined by computer point RCSTAIOI (2212°F) W W C S bottom head drain temperature at 2CEPPNL602 2 RPV coolant temperature can be determined either by reactor pressure (2212°F) using steam tables. computer point RCSTA102 OR reactor shell temperatures

(< 212'F) at 2CEC-PNL614, RHR Heat Exchanger Inlet Temperature (if RHR is in SDC).

Page 14 N2-OP-29 Rev 09

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 37 2 268000 2.1.32 3.8 N NA LOK ' Grp 10 CFR 55.41(b) 10 ' LOD (1-5) " Reference Documents H 2 N2-0P-31 Rev15 Radwaste - Conduct of Operations: Ability to explain and apply system limits and precautions.

KA Tie -Waste Collector Tank is part of Radwaste System, and has operational limitations with RHR.

QUESTION 37 Reactor is in Hot Shutdown, with the following conditions:

0 RHS B Loop is operating in Shutdown Cooling Mode at 6000 gpm.

0 RPV Water level is 200 inches.

Reactor Pressure is 20 psig.

0 RHR Heat Exchanger Outlet Temperature is 235°F.

0 THEN 2RHS*MOVl42 RHR DISCH TO RADWASTE MOV is throttled open to reduce RPV Water Level by 20 inches.

Which one of the following LIMITATIONS will be exceeded by these conditions?

A. Minimum flow through the RHR Pump.

B. Maximum pressure for the piping t0 Radwaste.

C. Maximum temperature for the Waste Collector Tanks.

D. Minimum RPV Water Level for Shutdown Cooling operation.

Correct Answer: C The maximum Temperature for the Waste Collector Tank will be exceeded.

Plausible Distractors:

B is plausible; RHR has High Reactor Pressure Isolation, which will protect Radwaste interface piping.

A is plausible; RHR Pump flow will increase when 2RHS*MOV142 is opened..

D is plausible; 180 inches is not below the minimum RPV Water Level, 159 inches.

Page 43 of 88

D. PRECAUTIONS AND LIE.1ITATTCNG iC'ont 1 19.13 D o n o t s e c u r e 01- p l a c e LPCI Lystem i n MANUAL made u n l e s s , by a t least two independent indicatic,ms, (1) misoperation i n AUTOMATIC mode is confirmed, o r ! 2 ) adequate rore c'ooling is a s s u r e d .

t'Misoperatio~it'includes both i n a p p r o p r i a t e i n i t i a t i o n of a LPCI system and coirtiriued operat ion of d LPCI system beyond autotnatic t r i p setpoints. If LPCI system i.s placed i n MANUAL mode! i t w i l l not i n i t l a t e a u t o m a t i c a l l y . Make frequent checks of the i n i t i a t i n g o r c o n t r o l l i n g parameter. Vhen t4P.NUAL inode i s no longer r e q u i r e d ,

r e s t o r e t h e LPCI system t o AIJTIMATICISTANDBY p e r Subsections G . 2 . 0 atid E.1.0.

20.0 Suppression chamber p r e s s u r e st.ould be maintained betwyen 1 4 . 2 and 15.45 p x i a . I f n o t , follow Technical S p e c i f i r a t i o n 3 . 6 . 1 . 4 a6 applicable.

21.0 F r i a r t o opening 2RHS*MVV112 ar.d 2RHS*MOVlI3, v e r i f y 2 R H S

• M W l A ( 6 ) is (C9) closed aiid 2RHS*MOV2k:Ei i s open. This w i l l preclude d r a i n i n g the RPV t o t h e suppression p c o l exc<et-dingt h e D / P r a t i n g o f 2FHS*lrlC1'LT2A (33;.

22.0 When it? aliutdotiti c o o l i n g , do fitlt exceed l t l O ° F / H R RPV c o ~ l d o w nr a t e .

23.0 Rue t o the fact that Radwaste I'ollector Tank Temperature should not a c e e d L50DF, the following r e n t r i c t i o n s apply. During f l u s h i n g e v o l u t i o n s t h e RHS pump m u s t bc w i t h i n lOcI*P of RPV coolant temperature b e f e r e s t a r t i n g , t l t e r r f a r e , RHS eEfluetitb should not exceed 1 B O O F . This higher temperature allows f c x 1 r a p e r system warm u p , and is low enough for Radwsste QPS tO maintain 150°F i n t h e X a s t e C o l l e c t o r T a n k s . During l e v e l c o n t r o l e v o l u t i o n s , because o f t h e l a r g e volutntts o f water, E H S e f r l u e n t s shcsuld m t exceed 1 5 0

• F . Note, e f f l u e n t s of .-l'i03F C d l l Le discharged t G Radwaste kecause KadWa.+te tan control Ifaste C o l l e c t o r Ta1.k Temperatures p e r N2-XHP-2Q Limits may be exceeded f o r s h o r t periods of t ime, during 3r immedidtely followltig Reactor shutdown, i f Radwaste can s u r p o r t Waste C'ollbccor Tank Temperacure Control p r o c e t u r e s .

24.0 A l t e r n a t e shutdown c c o l i n y ~ h a i i l donly be used i f nornial shutdown c o o l i n g cannot be e s t a b l i s h e d and i t is determined t h a t a l t e r n a t e shutdown cooling i s r e q u i r e d , IAW 112-SOP-31 or N2-SOP-31R.

25.0 T h i s sybtern s h a l l be kept f u l l of water arid p r e s s u r i z e d t o prevent ijiping damage due t o wa t ex- hamnisr .

26.0 If LPCI o p e r a t i o n is required c u r i n g suppression pool pumpdown (A QT C loops:, t h e applicable s t e p s of t h e putupdown procedure slicsiild be completed a s quickly as p ~ s s i l r l t . (LPCI A o r C is i m p Nue t o manual valve oysratioris, f u l l i t i > w c t i m flow would he hindered. 1 27.0 Suppressisn Pool pumpdown shall not he done when maiiitaining Rsactor water l e v e l w i t h the PHR Syster'l per Subsection H . 1 1 . 0 .

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 38 2 271000 K3.01 3.5 N NA LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents F 2 N2-SOP-17 Rev 3 Knowledge of the effect that a loss or malfunction of the OFFGAS SYSTEM will have on following: Condenser vacuum QUESTION 38 The plant operating at full power with the following:

0 Off Gas Radiation Monitors 20FG-RE13A and 20FG-RE13B go to HIGH ALARM.

0 20FG-AOV103, OFFGAS EXHAUST TO MAIN STACK is CLOSED.

0 Main Condenser Vacuum is 26 inches Hg (vacuum) and LOWERING.

Which one of the following actions will be required as a result of this condition?

A. SCRAM the Reactor B. START a Mechanical Vacuum Pump.

C. SWAP operating Steam Jet Air Ejectors.

D. TRIP Hydrogen Water Chemistry Injection.

Correct Answer: A With OG Isolated, it is required to SCRAM the reactor, due to lowering Main Condenser Vacuum.

Plausible Distractors:

B is plausible; MVP operation is prohibited > 5% power or during high radiation cond itions.

C is plausible; would be true for a SJAE malfunction resulting in Vacuum LOWERING.

D is plausible; High Hydrogen Flow can raise MSL radiation levels due to N-16. N-16 has decayed prior to reaching these radiation monitors.

Page 44 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 39 1 295001 AKI .01 3.5 N NA LOK ' Grp 10 CFR 55.41(b) 10

• LOD (1-5) ' Reference Documents F 1 N2-SOP-29 Rev 03 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : Natural circulation QUESTION 39 The plant is operating at 100% power, with the following:

0 Rod Line is 105%

0 THEN, BOTH Reactor Recirc Pumps trip.

Which one of the following identifies the actions required by N2-SOP-29, Sudden Reduction In Core Flow?

A. RESTART one Recirc Pump to raise flow and exit the EXIT region.

B. INSERT the first four Cram Rods to reduce Rod Line below 100%.

C. INSERT a Manual Reactor Scram, only if operating to the left of the natural circ line.

D. INSERT a Manual Reactor Scram, regardless of the operating point relative to the natural circ line.

Correct Answer: D A manual scram is required by N2-SOP-29 immediate actions, if no Recirc pumps are operating (natural circ) and the mode switch is in RUN or STARTUP. This action is required regardless of whether or not operation is to the left of the natural circ line.

Plausible Distractors:

A is plausible; because it recovery actions allow raising flow to exit the EXIT region, but it is not allowed by restarting a pump, only by raising with FCVs.

B is plausible; would be true for one pump trip with pre-transient rodline above 100%

and operating in an EXIT region.

C is plausible; would be true for one pump trip. A determination must be made and a manual scram is required if to the left of the natural circ line.

Page 45 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 40 1 295003 2.4.1 1 4.0 B NMP2 2002 NRC Exam LOK Grp 10 CFR 55.41(b) 10 LQD (1-5) Reference Documents F 1 N2-SOP-03 Rev 7 Partial or Complete Loss of AC - Emergency Procedures / Plan: Knowledge of abnormal condition procedures QUESTION 40 The plant is operating at 100% power with the following:

0 A complete loss of Off-site Power occurs 0 ALL three divisional diesels (EDGs) are powering their respective switchgear 0 N2-SOP-03, Loss of AC Power is being implemented 0 An additional Division I Service Water Pump is to be started Which one of the following actions is required to be taken prior to starting the additional Service Water Pump?

A. START RHS Pump "A" and CSL Pump.

B. RE-OPEN non-essential Service Water MOVs.

C. VERIFY RHS "A" and CSL Pumps are secured D. VERIFY Service Water crosstie MOVs are closed.

Correct Answer: A RHS and CSL pumps need to be running prior to starting an additional Service Water Pump due to starting current considerations.

Plausible Distractors:

B is plausible; but incorrect - if non-essentials opened prior to pump start, pump runout will occur.

C is plausible; but incorrect - wrong sequence.

D is plausible; but incorrect - Service Water crosstie MOVs are required to be open when the additional pump is started.

Page 46 of 88

ATTACHMENT 12: EDG LOADING I,ECCS PRIOR to starbng more than one Service Water Pump on Diu tiDiv II EDG, all available pumps for that Division must be runnrng

• X * * * * . *

• t f *

• t + t * * . * *

• f *

• l f * . I I I 1 X r t t l * * * * * ~ ~ * * ~ * ~ . * ~ ~ * , ~ .

12 1 Raising EDC, Load 12 1 I Verify all available low pressure ECCS pui-rrps in the affected division are running 12.1.2 IF . . Either ECCS pump trips while bemg powered from If3 EDG D ..It 1s desired to restart the tripped ECCS pump THEN a Reduce the number :if Service Water Puinps powered from that EDG te ONE b Revtart the tripped ESCS pump c Start additional loads as required in accordance with this secteon 12 1.3 Oeterrnrne load to be started [see table l(2) for loading guidance) 12 I.4 Obtain SMutiCRS permiss~on 12.1.5 Sfart load in accordance wrth applicable procedures 12.1.6 Venb EDG KW loading IS Iesv than 3400 K W (4840 KW 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Irmi!) for Div 1/11 8R 2500 KW 12850 K W 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> lmit):for Cliv !$I 12.2 Lowerine EDG Load 12 2.1 Use Table l(2j to determine which load is to be removed.

12 2 2 Obtain SWCRS permission.

12 2 3 Secure lead m accordance with applicable procedures Page 73 N2-SOP-03 Rev 07

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO 1 Tier j KIA Number i Statement IR 1 Origin ! Source Question 41 1 1 I 295004 ' ~ ~ 3 . 0 1 2.6 ~ I N NA

........LOK .Grp ........ I.. . ....CFR .55.'41 ( , b ) 7.....LOD , .Doc~uments ...................................................

j H \ N2-SOP-04 Rev 2 1 Knowledge of the reasons for the following responses as they apply to PARTIAL OR i COMPLETE LOSS OF D.C. POWER : Load shedding: Plant-Specific j ................................................................................................................................................................................................................................................................X QUESTION 41 The plant is operating at full power, when the following occurs:

0 Annunciator 852208 DIV II EMER BUS BYS 0028 125VDC SYSTEM TROUBLE alarms.

BATTERY BUS BYS 002B D.C. VOLTS indicates 0 VDC.

0 THEN a plant transient results in the initiation of load shedding signal to the AC switchgear.

Which one of the following describes the affect on the operation of 2ENS*SWG103 load feeder breakers?

A. WILL TRIP and CAN be remotely CLOSED.

B. WILL TRIP and CANNOT be remotely CLOSED.

C. WILL NOT TRIP and CAN be remotely TRIPPED.

D. WILL NOT TRIP and CANNOT be remotely TRIPPED.

Correct Answer: D A complete Loss of Div II DC Power is indicated. IF a Load Shed is initiated with these conditions, the 2ENS*SWG103 feeder breakers WILL NOT TRIP and CANNOT be remotely (from the control room) TRIPPED, because no power is available to breaker trip or close circuits.

Plausible Distractors:

A is plausible; would be true with a Loss of DIV I DC Power.

B is plausible; would be true with a 2ENS*SWG103 Bus Lockout.

C is plausible; would be true for a Load Shed Logic circuit failure.

Page 47 of 88

1.0 EVENT DESCRIPTION Annunciator 852108. DIV 1 EMER BUS BYS 002A 125VDC SYSTEM TROUBLE. is in alarin due to loss of DC power to ZBYS'SWG002A 0 Annunciator 852208, DIV ll EMER BUS BYS 0028 125VDC SYSTEM TROUBLE, is in alarm due to loss of DC power to 2BYS*SWG002B Annunciator 852308, DIVISION 111 BUS BYSOO2C 125 VDC SYSTEM TROUBLE, is in alarm due to loss of DC p w e r to 2CES11PNL414.

Annunciator 852501, STATION BAT IAIIBIIC 125 VDC SYSTEM TROUBLE, is in alarm due to loss of DC power to 2BYS-SWGGOlA (B,C) 0 Annunciator 852542,24148VDC DISTRIBUTION PANEL 300A UNDERVOLTAGE, is in alarm due to loss of DC power to 2BWS-PNL300A 0 Annunciator 852552,24148VDC DISTRIBUTION PANEL 3003 UNDERVOLTAGE, is in alarm due to loss of DC power to 2BWS.PNL300B 2.0 AUTOMATIC RESPONSES 21 IF loss of DC power to 26WS-PNL300A(B) the fcrllowing automatic responses occur NOTE Attachment 1, LOADS AFFECTED BY LOSS OF 2BWS-PNL300A(B). contains a detailed listing of the affected loads AND the actions on a loss of power 0 IF the Reactor Mode Switch ISNOT in RLIN. THEN a trip of RPS Channel A iB) occurs 22 IF loss of DC power to 2BYS*SWG002A(B), the following automatic responses occur NOTE Attachment 2(3) LOADS AFFECTED BY LOSS OF 2BYSXSWG002A(B),contains a detailed listing of the affected loads AND the actions on a loss of power

• IF running m high speed, BOTH Recirc Pumps trip to zero speed Trip of 2WCS-PIA AND PIE, CLEANUP PUMPS Isolation of Group 8 AND 9 Outboard (Inboard) I V s 0 Loss of Division I (11) load shedding AND load sequencing 0 Loss of Division I (11) Electrical Distribution tndication control, interlocks AND protection Loss of 2EGS*EGI (EG3)

• Loss of GTS Train A (B) 0 Loss of DC power to 2VBA*UPS2A (28) 0 Loss of Division I (11) Annunciators Page 1 N2-SOP-04 Rev 02

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier KIA Number Statement IR Origin Source Question 42 1 295005 M2.06 2.6 N NA LOK

• Grp ' 10 CFR 55.41(b) 10 ' LQD (1-5) ReferenceDocuments H 1 Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP : Feedwater temperature QUESTION 42 The plant is operating at 24% power when the Turbine Generator TRIPS.

Which one of the following describes the affect of this transient on Reactor Power, as indicated on the APRMs and why?

A. RISES due to Reactor Pressure increase.

B. LOWERS due to an automatic Reactor Scram.

C. RISES due to Feedwater Temperature reduction.

D. LOWERS due to Recirculation Pump Speed Downshift.

Correct Answer: C When the Turbine Generator trips, Turbine Bypass Valves will OPEN. Reduced Extraction Steam flow to Feedwater Heaters will LOWER Feedwater Temperature, which will cause Reactor Power to RISE.

Plausible Distractors:

A is plausible; would be true for a LOW failure of the Pressure Regulator.

B is plausible; would be true for initial Reactor Power at > 30%.

D is plausible; Recirculation Pump Dowrtshifi occurs due to EOC RPT. With initial power below 26% (Low FW Flow Cavitation Interlock), initial Recirculation Pump Speed must be SLOW. Downshift is excluded by initial power.

Page 48 of 88

5.0 DISCUSSION 51 When directed to reduce power, the expectation is a power reduction below 90% The target power level 13approximately 85% This target vias established to be consrstent wrth other operations procedures to ensure power 13reduced to and rernains less than 90°h 52 Loss of Feedwater Heating 5.2. f A Loss of Feedwater Heating ci.W occur in two ways 0 Loss of heating medwn (Extraction Steam, Cascading Drain Water)

Heater String Bypass Valve, 2CNWAOV101 or PFWS-RIOVf02. IS opened 522 2CNM-AOV10.1 LOW PRESS HTR STRtNG BYPASS VLV. opening ISconsidered a loss of all three Low Pressure Heater Strings until 2CNRl-AOV101 can be closed 523 2FWS-MOV102,fifH POINT HEATERS BYPASS VLV opening IS considered a loss of all three High Pressure Neater Strings until 2FWS-MQV102 can be closed 524 Frgure 1, Feedwater TemperatureiThermaI Power Limit, provides the u?per and lower limits of Feedwater Temperature for a given Thermal Power (when Thermal Power is >3Oo,oj. Thermal Power must be obtained from OD-3 ar manual heat balance Steady stale operation beyond the limits of Figure 1 has not been analyzed and should be avoided (Cl) 5.25 Rx power shall be lkmrfed to .-" (30% while opera!ing with o w High Pressure Feedwater Heater completely isolated with MSR's in service.

53 Rearc FCV Failure 531 Recirc FCV movetnent open or closed could be caused by either a hydraulic OR controller farlure 532 Failure of the Master or Flux Controllers, when in Flux Auto or Flux Manual modes, could result in both Recirc FCVs opening or closing at a maximum rate of I1 Sbisecond as limrted by the Loop Controller liniiter The aciual rate of movement may be much stower 5.3 3 Failure of a Loop Cnntrofler could resuli in the affected Recirc FCV d03lng at a maximum rate of SOKisecund 01opening at a maximum rate of 309bisecond as lrmited by valve hydraulics The actual rate of movement may be much slower Page 5 N2-SOP-08 Rev 03

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement IR Origin Source Question 43 1 295006 AKl.01 3.7 B LaSalle 2003 NRC Exam LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents H 1 ARP 603127 Knowledge of the operational implications of the following concepts as they apply to SCRAM : Decay heat generation and removal QUESTION 43 A power ascension is in progress with power at 17%.

A failure causes ALL Turbine Control Valves to fail OPEN, resulting in an automatic scram and the following indications:

RPV Water Level reached a minimum of 167 inches, and stabilized at 185 inches.

0 Reactor Pressure reached a minimum of 700 psig and is rising.

With NO operator action, which one of the following lists the decay heat removal methods which are immediately available?

A. Safety Relief Valves and Reactor Water Cleanup ONLY.

B. Safety Relief Valves, Main Steam Line Drain Valves, and Turbine Bypass Valves ONLY.

C. Safety Relief Valves, Reactor Water Cleanup, and Main Steam Line Drain Valves ONLY.

D. Safety Relief Valves, Reactor Water Cleanup, Main Steam Line Drain Valves, and Turbine Bypass Valves.

Correct Answer: A; At an initial power of 17% the Reactor Mode Switch is in RUN.

Reactor Pressure lowered below 766 psig with the Reactor Mode Switch in RUN, resulting in a Group 1 Isolation. MSlVs and MSL Drains are SHUT. RPV Water Level remained above L2 (108.8 inches), so RWCU remains available.

Plausible Distractors:

B is plausible; Main Steam Line Drain Valves and Turbine Bypass Valves are NOT available due to Group 1 Isolation.

C is plausible; Main Steam Line Drain Valves are NOT available due to Group 1 Isolation.

D is plausible; Main Steam Line Drain Valves and Turbine Bypass Valves are NOT available due to Group 1 Isolation.

Page 49 of 88

S e t point R e l ~ yR 2 % H - K 4 A I 766 PSIG

< ?.m.C*PNLG 0 91 NOTES: 1. The following table identi l e a the equiptoent i n c l u d e d i t 1 the yeueration of the sLyrtal tq> t i t + &mve r e l a y s .

2. A l l ot t h e helaw trip u n i t s kl;lD relays are located 1n 2CEC*PWL609.

Paye 7 7 g

ATTACHMENT 11 ( Cont I 2 C E C f P N L 6 0 3 SERIES 1 0 0 ALaRPl RESPONSE PROCEDURES Automatic Response (Cant)

NOTE: The MSIV i s o l a t i o n s i g n a l clue t.ci Main St.eam Line Icw p r e s s u r e iS bypassed when t h e Reactor mode s w i t c h i s NOT i n RUN.

0 IF coincident. w i t h A l a r i i i 6 0 3 4 2 7 ('C?iv.2 Channel B E D) b e i n g e n e r g i z e d , TEEN t h e MSI'J's w i l l i s o l a t e .

0 I F Channel A ( D i v . 1 ) AND Channel .E ( t t i v . 2 ) a r e t r i p p e d , THEN t h e MSIV' s AND t h e Outhoard MSL D r a i n s !2lvlSS*MDV112 @ JMOV208) will J

isolate.

0 I F Channel (1 ( D i v . 1 ) AND Channel B ( D i v . 2 ) are t r i p p e d , THEN t h e MSIV' s AND t h e Inboard MSL Drain !2MSS*MnV111) w i l l i s o l a t e .

Operator Act i o n s NOTE: T h i s i s an infoi-niation alarm d u r i n g Reactor s t a r t u p AbJD Reactor shutdowi.

1. Using t h e red T r i p l i g h t s on t h e t r i p u n i t s , confirm which d e t e c t o r s have alarmed MID rnonit.01- f o r proper operat.ioi.1.

2. A s a p p l i c a b l e , i i e r i f y t h e automatic rsapanse.

3. I F a p p l i c a b l e , refer t.o t h e Emerqency Operat.iny Procedures (EOP' s ) .

4. I F a Reactor Serarn o c c u r s , e n t e r N2-SOP-101C, REYiCTOR SCRAM.

5. Determine AND e c r r e c t cause of a l a r m .

5. WEN Phe isolation s i g n a l clears ANND w i t h SSS y e i m i s s i o n , r e s e t t h e i s o l a t i o n by d e p r e s s i n g E22H-S32 AND/OR B 2 2 H - S S 3 RESET p u s h h i t t o n s on 2YECfPNLii02.

P o s s i b l e Causes 0 Bad p r e s s u r e t r a n s n i i t t a r , t r i p U t i l t OR r e l a y 0 Steam l e a k OR itiipinper v a l v e l i n e u p e Inproper valve 1int=up 0 Plant s t a r t u p OR shutdown R t f erelice s N2 -EOP' s 0 N2 -SOP- 10 1C N2-OP- 1 45482,46536 Page 7 8 0 N2 - ARP - 0 1 Rev 0 0

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal I RO 1 Tier ~ WA Number i1 Statement IR i Origin ! Source Question 1

4 4 1 1 : 2950 16  ! .AK2.01 4.4 B i NMP2 Bank

. ............................................... ........................................................................................................................................... ............................................. ............... ... ............... ....................................... ......I........SYSID

........... ..... ..13064 I

~

LOK Grp 10 CFR 55.41(b) 7 ! LOD (1-5) Reference Documents

' H 1 1 1

~

1 N2-SOP-78 Rev 4 Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following: Remote shutdown panel: Plant-Specific QUESTION 44 Following a fire in the Control Room, the following conditions exist:

AUTO DEPRESSURIZATION TRANSFER SWITCHES in the EMERGENCY position.

0 APPENDIX R DISCONNECT SWITCHES are in ACTUATE position.

Which one of the following describes the automatic operation of the Safety Relief Valves under these conditions?

ADS SRVs can operate automatically in the:

A. RELIEF mode only.

B. SAFETY mode only.

C. SAFETY and ADS modes only.

D. SAFETY and RELIEF modes only.

Correct Answer: B With the APPENDIX R DISCONNECT SWITCHES in the ACTUATE position, Relief and ADS Modes of SRV operation are defeated.

Plausible Distractors:

A is plausible; RELIEF function is defeated.

C is plausible; ADS function is defeated.

D is plausible; RELIEF function is defeated.

Objective Link: 02-OPS-001-218-2-01 Page 50 of 88

5.0 DISCUSSION 6'

Y. s crirsider~unrnhabitabie If a 5@&4 IS the iom It wodd be irapprcpriate$0 trj ic run tile stabon Upera*tossare SCBA hired and SCBh a*e siaged r w the Ccrtrwl Roo?: to $c?l4ate Control Rocm evazuatim ana response t3 pant enersencies NOT to extend or maintain opesationIR the Maifi Contra1 R m n .

and personnei saetty are not an i v s d i a t e wncern. the 534 (or GFiS) r the inipad osthe fre on skitior, ec;rlipinent Gcrtrol ROOTevacualon ma$ sti I be appropriate c a d on

• The locahon 3rd bundaries of the fire The exten! ard cha The npad or p o s ~ hre ow the control cf equipment requ red for safe &-&own

• The npad or possible impact o'ary eteclncai nalfunctrcr 01 the cuntrcl of eq2ipi;ent rewired 'cr safe s 5.2 OIY "9"mode of cpera: on, the aLrto closwe of 2 I C 8

i. has been defeated R e r e k , on Readar h gh *eevel,Lh s mud he nafiuarlf accnqdished

+53 nwe htCIrn2hZ Man sv721 ncids m Panel 54 Ciese oJtput breakers IQI-1 and 103-14 and feeder trea the assmated b r e a w 5.5 plap Other wr11urim'on arrargemenb ShaAd be made as mon as pssible 5.6 Cue t3 GontrA %on1 n:*amatian ard ,Apper h opera:irtn the C esel Roar; start and rhn wheher the C sin9 or not 57 s transfwred b the KSS parel :fa sititches 2 3, 4 cn or isclation 3 gnal 8s pwsen: the nrtiaticn or iso 5.8 Tc determine Rods 'e NOT scramined wi I Page Ec

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier . WA Number I Statement ; IR Origin ' Source Question 4 5 1 1 ; 295018 i 14A2.01 ~ 3.3 1 N NA LOK Grp I 10 CFR 55.41(b) 7 1 1

............................................................. i....................................................................................................... ............,. .................

LOD (1-5)

~ Reference Documents H 1 1 : i ARP 602319 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Component temperatures QUESTION 45 The plant is operating at 100% power with the following:

0 Loss of Reactor Building Closed Loop Cooling Water occurs 0 Annunciator 602319, RWCU FILTER DEMIN INLET TEMP HI-HI alarms Which one of the following describes the affect of this condition, if any, on the operation of the Reactor Water Cleanup Pumps?

A. CONTINUE to operate since no trips are received.

B. TRIP immediately due to isolation valve position.

C. TRIP directly due to the high temperature signal.

D. TRIP after a low flow condition exists for 15 minutes.

Correct Answer: B When 602319 alarms, at WCS*MOVI 12, CLEANUP SUCT OUTBOARD ISOL VLV closes. When the valve is NOT FULLY OPEN, the WCS Pumps TRIP immediately.

Plausible Distractors:

A is plausible; would be true for RWCU FJD Inlet (NRHX Outlet) temperature below 140°F.

C is plausible; identifies misconception about RWCU Pump Trip directly from High Temperature signal. High Temperature initiates a PClS Isolation. When the Isolation Valve is NOT FULLY OPEN, the RWCU Pump TRIPS.

D is plausible; would be true for valve closures OTHER THAN WCS*MOVI 12. A Low Flow condition would develop which initiates a time delayed RWCU Pump TRIP.

Page 51 of 88

Reflash: No RNCU FILTER IIEt*TII?

INLET TEblP HI-HI 319

6. Deterrniiia the ilau.9~of tlis h i g h temparature carzdition AND c o r m & ,

b. Predict the effect the interrelating system has on the system under study during all modes of plant operation

c. (SRO) Prioritize crew response t o the loss of multiple interrelated systems and assess the effect on system operability of the loss

3. The evaluation methods for this objective are comprehensive written examination (multiple choice or short-answer) and simulator or in-plant evaluation.

C. Main Idea

1. Reactor Building Closed Loop Cooling Water (CCP)

a. The Reactor Building Closed Loop Cooling Water (CCP) system provides cooling water for the Non-Regenerative Heat Exchanger and a means of cooling the backup source of seal water for the WCS Pump Seals.

1) CCP also provides a source of cooling water t o the WCS Pump bearings and the WCS pump pedestal.

b. A loss of CCP cooling t o the Non-Regenerative Heat Exchanger will result in a high temperature alarm on the Heat Exchanger outlet a t 130°F and a high temperature isolation of 2WCS*MOV112 a t 140°F.

1) A loss of CCP would also result in overheating of the WCS pump bearings and a loss of cooling t o the WCS pump pedestal.

2) I f t h e backup source of seal water were in service, a loss of CCP t o the seal water cooler would result in overheating of the mechanical seals on the WCS pumps.

3) A high temperature alarm would be received a t 185°F and the pump would be required t o be removed from service if temperature continued t o rise or if temperature could NOT be lowered less than

~ ~~

Student Guide (~2101204000~01) 120 of 202 Printed: 03/23/2007

a. The purpose of the Non-Regenerative Heat Exchanger is t o further reduce the Reactor Vessel water temperature by transferring the heat t o the Reactor Building Closed Loop Cooling Water System (CCP).

1) The Non-Regenerative Heat Exchanger is a single pass, counter flow, U-Tube heat exchanger

2) Non-Regenerative Heat Exchanger Design parameters; a) Shell pressure - 150 psig b) Shell temperature - 370 OF c) Tube pressure - 1,410 psig d) Tube temperature - 575 OF

3) The water from the Reactor Vessel passes through the tube side of the heat exchanger; while Reactor Building Closed Loop Cooling water is on the shell side.

a) During normal operation, the Reactor Vessel water enters the heat exchanger a t approximately 233"F, and is reduced t o less than 100°F a t the discharge of the heat exchanger.

b) A loss of CCP will result in a high temperature alarm a t 13OoF and an automatic closure of 2WCS*MOVll2 and subsequent trip of the running WCS pump a t 14OOF.

4) Maintaining a low discharge temperature is critical since higher temperatures could cause damage t o the resin, resulting in the release of soluble and insoluble impurities into the Reactor Coolant System, resulting in possible crud buildup and/or fuel element defects.

4. Filter/Demineralizer Units Student Guide (~~ioi204000~0i) 28 of 202 Printed: 03/23/2007

4. System/Component Interlocks Interlock Function Setpoint Automatic closure of Primary Containment Isolation 1) Low-Low Reactor 2WCS*MOV102 and Water Level a t 108.8 ,

2WCS*MOV112 2) High Pump Room and Group 6 and 7 Heat Exchanger Room containment isolations Ambient Temperature of 135OF for "A" room, 15OOF for "6" room,

3) High Differential Flow of 150.5 gpm with a 45 sec. time delay,

4) High Non-Regen Heat Exchanger Outlet Temp era t u re of 14 0O F (2W CS* MOV 112 only) ,

5) Reactor Building Radioactive Pipe Chase High Temperature of 135OF, and

6) Standby Liquid control or Redunda nt Reactivity Control Initiation.

Student Guide ( ~ 2 1 0 1 2 0 4 0 0 0 ~ 0 1 ) 82 of 202 Printed: 03/23/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 46 1 295019 AK3.02 3.5 B NMP2 Bank SYSID 13114 LOK Grp 10 CFR 55.41(b) 10 LQD (1-5) Reference Documents F 1 N2-SOP-19 Rev 3 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR : Standby air compressor operation QUESTION 46 A plant startup is in progress. The plant is at 85% power. The following conditions exist:

Compressors are configured with 3B in Lead, 3C in Lag, and 3A in Backup.

0 Instrument Air Pressure is 120 psig.

0 Oil Pressure on the B Compressor LOWERS to 10 psig.

Which one of the following describes the affect, if any, on the air system?

A. NO affect, the B Compressor will continue to operate.

B. B compressor will TRIP and ONLY A will START when Instrument Air Header Pressure is 95 psig.

C. B compressor will TRIP and ONLY C will START when Instrument Air Header Pressure is I 0 0 psig.

D. B compressor will TRIP and BOTH A and C will START when Instrument Air Header Pressure is 100 psig.

Correct Answer: C When Oil Pressure lowers to 10 psig, 3B will TRIP. ONLY C will START when Instrument Air Pressure is 100 psig.

Plausible Distractors:

A is plausible; 3B TRIPS.

B is plausible; 3A starts at 85 psig.

D is plausible; 3A starts at 85 psig.

Page 52 of 88

2.0 A U T ~ ~ A T RESPONSES IC The rate of Occurrence of following Automatic Responseswill be dependant on the rate of the loss of Instrument air pressure:

105 nsia &

Annunciator 851229, INSTR AIR SYSTEM TROUBLE Computer Point IASPCO6, lNSTR & SVCE AIR PRESS Lag Air Compressor auto starts 90 psig 3.

~nnunciator851218, INSTAIR RCVR TK 3 PRESS LOW Computer Point IASPC39, lNST AIR RCVR TK 3 PR LO 85 psig -b Annunciator 851208, INST AIR RCVR TK 2 PRESS LOW Computer Point IASPC05, INST AIR RCVR TK 2 PR LO Backup Air Compressor auto starts 21A.S-AOVI71 auto closes Annunciator 851239. SER AIR SYS 2IAS-AOV171 CLOSED Computer Point IASZCOI. SER AIR SYS 21AS-AOV171 74 osin 3.

Loss of Instrument Air to MSlV Pilot Valves causes MSlVs to close 65 psig 3.

Annunciator 603306. CRD SCRAM VALVE PILOT AIR HDR PRESS HlGHiLOW 60 psig 4 Feedwater Pump Minimum Flow Valves fail full open. results in Loss of Feedwater flow to Reactor Vessel Loss of air to Scram Air Header, results in indivtdual Control Rods inserting into core Annunciator 603443, CONTROL ROD DRIFT 70-0 psig 3.

Various air operated components will fait as ln~trumen~ Air pressure decays. See step 5.3 for a more coniplete list of components and their possible affects Page 2 N2-SOP-19 Rev 03

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO I Tier K/A Number I Statement IR Origin Source Question 4 7 1 1 ; 29502 1 ~ AA2.01 3.5 N NA i.............

LOK j Grp I 10CFR55.41(b)7 LOD (1-5) Reference Documents H SG N2101223002C01 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING : Reactor water heatup/cooIdown rate QUESTION 47 The plant is in Hot Shutdown, with the following conditions:

0 Reactor Pressure is 100 psig.

0 Suppression Pool Temperature is 80°F.

0 RHS Loop B is operating in the Shutdown Cooling Mode.

0 Cooldown Rate is 40°F / hr.

Which one of the following describes the affect of a leak resulting in RPV Water Level LOWERING to 165 inches and RHS Area Temperature RISING to 140°F?

A. Cooldown Rate will RISE due to ECCS actuation.

B. Cooldown Rate will LOWER due to SDC Isolation.

C. RPV Water Level will RISE due to ECCS actuation.

D. RPV Water Level will LOWER due to SDC operation.

Correct Answer: B When RHS Area Temperature exceeds 135OF, SDC will isolate and Cooldown Rate will LOWER.

Plausible Dist ractors:

A is plausible; If RPV Water Level LOWERED to L2 (108.8 inches), CSH would inject.

C is plausible; If RPV Water Level LOWERED to L2 (108.8 inches), CSH would inject.

D is plausible; If no actuations occurred, SDC operation and cooldown would lower RPV Water Level.

Page 53 of 88

e, Group 5, RHS Shutdown Cooling (SDC) and Head Spray Isolation Function

1) Associated Valves a) 2RHS*MOV104, RHS Reactor Head Spray Outboard I V b) 2RHS*MOV4OA &, B, Shutdown Cooling Return Outboard IVs c) 2RHS*MOV67A & B, SDC Inboard I V Bypass Valves d) 2RHS*MOV112, SDC Supply Inboard I V e) 2RHS*MOV113, SDC Supply Outboard I V

2) Group 5 isolation occurs on the following:

a) Low RPV Water Level (Low, Level 3) b) High RHS Pump Room A (8)Temperature I, High Reactor Building (RB) General Area Temperature c) d) High RB Pipe Chase Temperature e) High RPV Pressure { 123 psig) f) Manual 3 ) The RHS SDC Isolation receives input signals from six functions. The Low RPV Water Level (Level 3) and High Reactor Vessel Pressure functions each have four channels. The outputs from the reactor vessel water level channels are connected into two two-out-of-two trip systems. The reactor vessel pressure is arranged into two one-out-of-two trip systems. High RHS Equipment Room Area Temperature, High RB Pipe Chase Area Temperature, and High RB General Area Temperature functions each have one channel in each trip system in each area (one-out-of-one logic for each area) for a total of four, eight, and ten channels per function, respectively. The Student Guide { ~ 2 1 0 1 2 2 3 . 1 1 ~ 0 ~ ~ 0 1 ) 96 of 182 Printed: 03/14/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 48 1 295023 2.1.28 4.1 N NA LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents H 1 N2-OP-38 Rev 10 Refueling Accidents - Conduct of Operations: Knowledge of the purpose and function of major system components and controls.

QUESTION 48 The plant is in MODE 4, COLD SHUTDOWN, with irradiated fuel movement being performed in the Spent Fuel Pool.

What is the purpose for the MINIMUM Spent Fuel Pool Water Level limitation of 352 feet 3 inches?

To provide sufficient water level above:

A. the RPV Flange to accommodate a Loss of Decay Heat Removal.

B. fuel seated in racks to accommodate a Loss of Spent Fuel Pool Cooling.

C. the RPV Flange to absorb fission products released during a Fuel Handling Accident.

D. fuel seated in racks to absorb fission products released during a Fuel Handling Accident.

CorrectAnswer~DWlththe p.l.ant""inMODE4,CoLD...sHuTD.oWNI"the RPV Heads...

installed. The only applicable reason for Spent Fuel Pool Water Level limitation is to absorb fission product released during a fuel handling accident. Reduction in radiatior levels is a functional

.............................................. .............. .........equivalent

. .... ........................... to shielding.

Plausible Distractors:

A is plausible; would be true in MODE 5 (RPV Head removed) with ONE Shutdown Cooling system.

B is plausible; Decay Heat is removed by Spent Fuel Pool Cooling, but this is not the reason for the procedural minimum.

C is plausible; would be true in MODE 5 during Core Alterations.

Page 54 of 88

Spent Fuel Storage Pool Water Level B 3.7.6 B 3.7 PLANT SYSTEMS B 3.7.6 S p e n t Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the spent fuel storage pool meets the assumptions of iodine decontamination f a c t o r s following a fuel handling accident.

A general description of t h e spent fuel storage pool design i s found i n the USAR, Section 9.1.2 (Ref. 1 ) . The assumptions of the fuel handling accident a r e found in the USAR, Section 15.7.4 (Ref. 2 ) .

APPLICABLE The water level above the i r r a d i a t e d fuel assemblies i s an SAFETY ANALYSES e x p l i c i t assumption of the fuel handling accident (Ref. 2 ) .

A fuel handling accident i s evaluated t o ensure t h a t t h e radi ol ogi cal consequences (cal cul ated who1 e body and thyroid doses a t the exclusion area and low population zone boundaries) a r e 5 25% (NUREG-0800, Section 15.7.4, Ref. 3 )

of the 10 CFR 100 (Ref. 4 ) exposure guidelines. A fuel handling accident could r e l e a s e a f r a c t i o n of the f i s s i o n product inventory by breaching the fuel rod cladding as discussed in the Regulatory Guide 1 . 2 5 (Ref. 5 ) .

The fuel handling accident i s evaluated f o r t h e dropping o f an i r r a d i a t e d fuel assembly o n t o the r e a c t o r core. The consequences of a fuel handling accident over the spent fuel storage pool are l e s s severe than those of the fuel handling accident over the reactor core (Ref. 2 ) . The water level i n the spent fuel storage pool provides f o r absorption of water soluble f i s s i o n product gases and t r a n s p o r t delays o f soluble and insoluble gases t h a t must pass t h r o u g h the water before being released t o the secondary containment atmosphere. This absorption and transport delay reduces the potential radioactivity of t h e release d u r i n g a fuel handling accident.

The spent fuel storage pool water level s a t i s f i e s Criterion 2 of Reference 6 .

LCO The specified water level preserves t h e assumption of t h e fuel handling accident analysis (Ref. 2 ) . As such, i t i s the minimum required f o r fuel movement within t h e spent fuel storage pool.

(continued)

NMP2 B 3.7.6-1 Revision 0

D. PRECAUTIONS AND LIMITATIONS 1.O Radiation Protection Precautions 1.I Applicable radiological precautions shall be observed. Radiation Protection shall be contacted for guidance, as required.

1.2 ALARA practices shall be observed to minimize personnel exposure and spread of contamination.

1.3 All effluents from the SFC System are to be treated as contaminated. Necessary provisions to contain leakage shall be made when breaking connections. venting. or draining lines.

14 Radiahon Protection must be notified whenever Fuel Pool System draining or venting is to be performed Draining and venting can cause localized hot spots and significant changes in radiological conditions 20 Sufficient Fuel Pool Cooling components shall be operable to maintain adequate cooling when there is irradiated Fuel in the Spent Fuel Pool Normal Spent Fuel Pool Level is 352' 1 0 (352' 8 to 353') and shall be maintained r352' 3" at all times to provide adequate shielding from irradiated fuel and ensure Tech Specs niinimum level requirements are met.

4.0 An Operator should be assigned to visually monitor Spent Fuel Pool and Reactor Cavity Levels during all evolutions which could affect stable level conditions such as filling or draining. Pump starts. and system ffowpath changes.

5.0 If Spent Fuel Pool or Surge Tank Level Instrumentation becomes unavailable. continuously monitor Spent Fuel Pool Level locally during evolutions that affect Spent Fuel Pool or Reactor Cavity Water Level.

6.0 The FILTER SELECT switch shall be maintained in the MANUAL position except as directed in this procedure.

7.0 Maintain 4 5 0 F in the tube and shell sections of the SFC Heat Exchangers.

80 Prior to Plant heatup, the Section of piping between valves 2SFC*V203 and 2SFC'V204 INNER REFUEL SEAL LEAK DETECTION ISOLATIONS must be verified drained to prevent damage due to expansion of trapped water. See N2-OP-IOIA. Attachment 2.

9.0 Fuel Transfer Gates must be installed prior to draining the Reactor Cavity or the Internal Storage Pit.

Page 8 N2-OP-38 Rev 10

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 49 1 295024 2.4.50 4.2 N NA LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents H 1 ARP 603140 Rev 00 High Drywell Pressure - Emergency Procedures / Plan: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

QUESTION 49 The plant is operating at full power, when the following occurs:

0 Alarm 603140 DRYWELL PRESS HIGH / LOW alarms.

0 Drywell Pressure is 0.75 psig.

0 Drywell Temperature is 140°F.

NO Barometric Pressure changes have occurred.

Which one of the following statements reflects the status of Drywell Pressure and the action required?

A. LOW. RAISE Drywell Temperature.

B. LOW. INITIATE Nitrogen Addition.

C. HIGH. VENT Nitrogen from the Drywell.

D. HIGH. ENTER N2-EOP-RPV and N2-EOP-PC.

Correct Answer: C 0.75 psig is the HIGH alarm setpoint. With HIGH Drywell Pressure it is required to VENT Nitrogen from the Drywell.

Plausible Distractors:

A is plausible; would be true if Drywell Pressure indicated -0.50 psig and Drywell Temperature were below normal.

B is plausible; would be true if Drywell Pressure indicated -0.50 psig.

D is plausible; would be true if Drywell Pressure exceeded 1.68 psig.

Page 55 of 88

ATTACHE.IENT 11 (Coiit )

2CEC*PNL603 S E R I E S 1 O C ALARM RESPONSE PROCEDLJRES Reflash: NO ~-

2CEC* PPJLI; 0 3 603140 Cornputxr Point Con1yuts.r Printout RFSPCO 1 RES DW PRESS 21SC*PS1651 Lc): - 0 . 5 pslg 2ISC*PS165 3 HI: 0 . 7 5 p s l g Aut nmat i c Response NONE Operator A c t i o n s 1.0 Check D r y w e 11 pressure 1 ntli ca t 01s on 2 CEC* PNL6 09 AND 2CEC*PNL&1 1 to determine whether d r y w e l l presstire is h i g h OR lcw.

Drywell P r e s s u r e i n P S I A ( Z C M S - P I l 7 R o r Cornpu%erPoint OlSPAO5)

D r p 7 e l l Temperatures D r y w e l l Leak Rates Radiation L e v e l s 3.0 IF Drywell p r e s s u r e change i E i NClT due t o Barometric: change, OR as d i r e c t e d by S.SS/ASSS, perform t h e following:

3.1 3.2 I F p r e s s u r e is h i g h , perform N Z - O P - r ; l A ,

I F presslure is low, ~ e r f o r mN2-I)F-61A, 0

S u b s e c t i o n H.I.O.

Subsection F. 7 . 0 .

45482,46536 Page 801 N2 -ARP- 01 Rev 00

1. Performance of this subsection may be required by the EOPs Changes (including renumbering) are required to be reviewed by the EOP Coordinator. (N2-EOP-PCj

2. Sample results prior to venting are not required if venting is directed by EOPs.

3. Chemistry sampling takes a minimum of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to complete.

4 Only one Standby Gas Treatment subsystem is required to be OPERABLE to allow opening the 12" and 14" containment purge valves in MODES 1 , 2 and 3 provided 2GTS*AOV101 is closed (SR 3 6 1 3 1) 11 IF required, as determined by Chemistry Supervision, notify Chemistry to sample the containment for the pre-vent acceptance criteria of ODCM Table D3 2 1-1 per Precaution and Limitation D.13 0 guidance If no venffpurge is in progress, sample both the Drywell and the Suppression Chamber.

Suppression Chamber Venting may be performed with a pre-vent evaluation sample being performed ONLY on the Suppression Chamber. Drywell Venting requires a pre-vent evaluation sample on BOTH the Suppression Chamber and Drywell.

Commencing DW venting does NOT require an additional SC sample if the SC has been sampled within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1. I . 1 Verifj Chemistry has a current sample PrjoF to venting and it is permissible to vent based on the results (ODCM Table 03.2.1-I),

1.I .2 IF the Nitrogen low flow makeup to Primary Containment is in service, secure I

makeup per Subsection F.8.0, Securing Nitrogen Addition to Drywell OR f.10.0, Securing Nitrogen Addition to Suppression Chamber.

Page 89 N2-OP-61A Rev 11

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 50 1 295025 EA1.03 4.4 M Shown Below LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents F 1 SG N2101239001CO1 Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: Safetyhelief valves: Plant-Specific QUESTION 50 Following a Group 1 Isolation, Reactor Pressure is 1145 psig and STABLE.

Which one of the following identifies the status of the Control Room P601 and P628/P631 RED Indicating Lights for Safety Relief Valve, 2MSS*PSV137 at this time?

P60 1 P628/P631 A. ON ON B. ON OFF C. OFF ON D. OFF OFF Correct Answer: B 1145 exceeds the RELIEF setpoint of 2MSS*PSV137, so the P601 RED Indicating Lights will be ON. NO ADS setpoint has been reached (STABLE Pressure) , so the P628/P631 Lights are OFF.

Plausible Distractors:

A is plausible; would be true with a concurrent ADS actuation.

C is plausible; would be true BELOW the RELIEF setpoint and with a concurrent ADS actuation.

D is plausible; would be true BELOW the RELIEF setpoint

! Source Question: Modified from 2002 NMP-2 NRC Exam A Loss of High Pressure injection occurs, with the following: MODIFICATION Automatic initiation of ADS MODIFICATION RPV Pressure lowers to 80 psig MODIFICATION RPV level is 180 inches Page 56 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal Which one of the following identifies the expected status of the Control Room P601 and P628/P631 red indicating lights for SRV 137 at this time?

P601 P628lP631 A ON ON B ON OFF NEW CORRECT C OFF ON OLD CORRECT D OFF OFF Page 57 of 88

130, 134 and 137) are also supplied with 175 psig nitrogen from the gaseous nitrogen system. This additional capacity provided t o the ADS SRVs assures that they can sufficiently depressurize the RPV to allow the lowest pressure Emergency Core Cooling Systems t o inject with a loss of nitrogen supply t o the ADS pneumatic system. During power plant operations all SRV operating cylinders are supplied with nitrogen,

5) There are three modes of SRV operation (Relief, Safety, and ADS) and two distinct pressure setpoints for each SRV (Relief, Safety).

The Relief mode utilizes pressure sensing transmitters to actuate the "C" solenoid, opening each S W at set pressure. The SRV's are divided into five groups, each with a different relief set pressure.

Two valves are set t o open at 1103 psig; four valves are set at 1113 psig; four valves are set a t 1123 psig; four valves are set a t 1133 psig; and the remaining four valves are set a t 1143 psig. The "C" solenoid on each SRV can also be actuated by the individual keylock switches on panel 601 in the control room.

The ADS mode utilizes the "A" and "6" solenoids of the ADS designated SRV's. They are actuated upon receiving an ADS initiation signal.

The "A" and "C"solenoids foI all SRVs are powered from 2BYS5WG002A (Division I DC). The "B" solenoids are powered from ;ZBYS*SWGOOZB (Division I1 DC).

The Safety mode does not utilize solenoids for actuation. This mode serves as a backup to the relief mode. Direct steam pressure in opposition t o spring force causes the SRV t o open. The SRV's are divided into five groups according t o the safety set pressure. Two Student Guide ( N Z I O I ~ ~ ~ O O ~I )C C I 16 of 156 Printed: 04/18/2007

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement IR Origin Source Question 51 1 295026 EA1.02 3.6 N NA LOK Grp 10 CFR 55.41(b) 10 LOD (I -5) Reference Documents H 1 N2-EOP-PC Rev 12 Ability to operate and/or monitor the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool spray: Plant-Specific QUESTION 51 Following a plant transient, the following conditions exist:

0 Drywell Pressure is 2.8 psig 0 Drywell Temperature is 200°F.

0 Suppression Chamber Pressure is 2.6 psig 0 Suppression Pool Temperature is 92°F.

0 Suppression Pool Level is 200 feet.

iM

.a. !

. l 9 YX s

3 25u rh b

Y@

l!x Which one of the following actions is required?

It is required to INITIATE:

A. Suppression Chamber Sprays ONLY.

B. Drywell Sprays AND Suppression Pool Cooling.

C. Suppression Chamber Sprays AND Drywell Sprays.

D. Suppression Chamber Sprays AND Suppression Pool Cooling.

Correct Answer: D With Suppression Pool Temperature above 90°F and before Suppression Chamber Pressure reaches 10 psig; it is required to initiate BOTH Suppression Chamber Sprays AND Suppression Pool Cooling.

Page 58 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal 1 Plausible Distractors:

A is plausible; would be true with Suppression Pool Temperature below 90°F.

i B is plausible; would be true below the Drywell Spray Initiation Limit.

1 C is plausible; would be true with Suppression Pool Temperature below 90°F.

Page 59 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement IR Origin Source Question 52 1 295028 EK2.03 3.8 N NA LOK Grp 10 CFR 55.41 (b) 10 LOD (1-5) Reference Documents H 1 N2-EOP-RPV Rev 10 PROVIDE Steam Tables Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following: Reactor water level indication QUESTION 52 Following a steam leak in the Drywell, the following conditions exist:

0 RPV Water Level is 190 inches.

0 High Pressure Core Spray is injecting and has been THROTTLED.

0 Reactor Pressure is 900 psig.

0 Drywell Temperature is 340°F.

0 Drywell Sprays are NOT available.

0 RPV Blowdown has been initiated.

Which one of the following challenges regarding Reactor Water Level will be presented by these conditions?

A. RPV Water Level Transmitter electronics will FAIL due to excessive temperature.

6. ACTUAL RPV Water Level will LOWER below Top of Active Fuel due to inventory loss from the Safety Relief Valves.

C. INDICATED RPV Water Level will trend HIGHER than ACTUAL due to reference leg boiling.

D. INDICATED RPV Water Level will trend LOWER than ACTUAL due to dissolved gases coming out of solution.

Correct Answer: C At about 110 psig, Saturation Limits will be exceeded. INDICATED RPV Water Level will trend HIGHER than ACTUAL due to reference leg boiling.

Plausible Distractors:

A is plausible; Transmitter electronics are located in the Reactor Building. 340°F is based on NOT exceeding Equipment Qualification (EQ) temperatures.

B is plausible; a steam leak is given. HPCS, LPCS and LPCl can maintain RPV Water Level above TAF.

D is plausible; Reference Leg Keep Fill will have prevented dissolved gas accumulation in reference legs, none will evolve during RPV Blowdown.

Page 60 of 88

llslRPV ~ a t ~ Temperature r a ~ ~ ~ ~

250 200 0

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 53 1 295030 EK3.04 3.5 N NA LOK 1 Grp I 10 CFR 55.41(b) 10 LQD (1-5) 1 Reference Documents I F I I I 1 ARP 601720

....................................................................................................................................................................... i.............................................................. i . ...........................................................................................................................................

~ EK3.04 - Knowledge of the reasons for the following responses as they apply to LOW j SUPPRESSION POOL WATER LEVEL: HPCS operation: Plant-Specific QUESTION 53 Per N2-EOP-PC, Primary Containment Control, High Pressure Core Spray should be operated with caution during conditions of Low Suppression Pool Water Level.

Which one of the following describes the hazard to the High Pressure Core Spray Pump, if this caution is NOT complied with?

A. TRIP on Low Suction Pressure, if aligned to its NORMAL source.

B. TRIP on Low Suction Pressure, if aligned to its ALTERNATE source.

C. RISK DAMAGE due to cavitation, if aligned to its NORMAL source.

D. RISK DAMAGE due to cavitation, if aligned to its ALTERNATE source.

Correct Answer: D With Low Suppression Pool Water Level, HPCS will RISK DAMAGE due to cavitation, if aligned to its ALTERNATE source.

Plausible Distractors:

A is plausible; RClC has a Low Suction Pressure TRIP, HPCS does NOT. The Suppression Pool is ALTERNATE source for HPCS.

B is plausible; RClC has a Low Suction Pressure TRIP, HPCS does NOT.

C is plausible; The Suppression Pool is ALTERNATE source for HPCS.

Page 61 of 88

ATTACHMENT 7 < Cont )

2 C E C

• P N L 6 0 1 S E R I E S 7 C 0 !&ARM RESPONSE PROCEDURES Reflash: NG I HPCS PIslclF 1 SUCTION PRESS HIGH /LCN 720 COMPTJTEK P O I N T C:OMPTJTER PRINTOUT  ;:GURCE SETPOINT CSHPC05 HPCS PMPl SUCTION iXH*PT102 <4 p i g OR PRESS  :.SO p s i g A ~ t o m a t i cResponse None Ope rd t o r A c t . i cns

1. V e r i f y l i r c a l l y a t HPCS r a c k H 2 2 - P O 2 4 , C S H

• F I 1 0 3 , PUMP SUCTION PRESS

2. Verify E22-N652 pressure readinq i*t 2CECfPNL625.

3. I F CSH punir> s u c t i o n pressure is low, perform th3 f o l l m d i n g :

V e r i f y C'SH pinip s u c t i o n valve linniup per Att.achnient 1.

I F CEH*MOV101, PUMP SUCT FR, CNDS TK i s open, v e r i f y 2CNS-TK1B l e v e l i s i 1 2 . 5 feet, a s read on w - L r m , at ~ ~ 5 1 .

I F C'SH*MOVllEI, PMEJ Sl!CT FRO!( SUPPRESSION FOOL is op'n, verify Suppression Pool level i s z 199.5 f e e t , a t 2 C E C

• P N L 6 0 1 .

S h i f t CSH pump suc:tir,xi, a s r e q u i r e d

4. I F CSH p i m p s u c t i o n p r e s s u r e is h i g h , perform t h e f o l l o w i n y :

m: Taniporary change packaye N 2 - 0 2 - 0 8 7 disaoilnect+d tlie a c t u a t o r a r m on 2CSH*AOVlOY such t.hat t.he c o n t r o l s w i t c h on 2 C E C

• P N L 6 0 1 i s not, f u n c t i o n a l AND valve d i s c i s f r e e t o swing. The p o e i t i o n i n d i at i on or1 2 CEC* DNL:; 1?1 renia i n s f unc t i m a 1.

V e r i f y CZH*MCPJli"7, FPIP 1 I N J E C T I 9 N VLV, CSH*AOVlOR, TESTABLE CHEClK VLV, are i:losed I F NCI:: rerpiixed for RPV i n j ectinn.

V e r i f y ZCSH*V10, El1 I ? I S C H CHECK BYF is closed.

45482,46536 Page 5 4 1 NZ-ARP-01 Rev 0 0

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement I IR 1 Origin ; Source Question 54 1 295031 2.2.44 I 4.2  ; B I NMP2 Bank

................................................... i ........................... i .......................... I: .............................................................................

SYSlD22723 LOK Grp 10 CFR 55.41 (b)'lO LOD (1-5) i Reference Documents H 1 j N2-0P-82 Rev 6 Reactor Low Water Level - Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives effect plant and system conditions. I _

QUESTION 54 The plant is experiencing a LOCA and a concurrent loss of Line 6, with the following:

0 Affected Emergency Diesel Generators start and load 0 RPV level is 100 inches and being maintained by RClC and CSH.

0 Drywell Pressure is currently 1.3 psig and dropping.

Which one of the following describes the status of the H2/02 monitors AND the action required to determine containment Hydrogen and Oxygen concentrations?

A. IN SERVICE. MONITOR P601 Hydrogen and Oxygen indications B. IN STANDBY. START the sample pumps CMS*P2A/B, then MONITOR P601Hydrogen and Oxygen indications.

C. ISOLATED due to power loss. OPEN SOVs and START the sample pumps CMS*P2A/B, then MONITOR P601Hydrogen and Oxygen indications.

D. ISOLATED due to a LOCA signal. OVERRIDE the LOCA signal and PLACE the H2/02 monitors in service, then MONITOR P601Hydrogen and Oxygen indications.

Correct Answer: D With L2 LOCA signal active (RPV Water Level < 108.8 inches), it is required to OVERRIDE the LOCA signal, PLACE the H202 monitors in service, then MONITOR P601Hydrogen and Oxygen indications.

Plausible Distractors:

A is plausible; would be true if L2 auto started H202 System.

B is plausible; would be true with NO LOCA signal.

C is plausible; would be true for power interruption with NO LOCA signal.

Page 62 of 88

OFF-NORMAL OPERATIONS (Cont)

Post Accident SamDling NOTES: 1 This Subsection is written for both divisions with Division II components In parentheses

2. This Subsection is used to support the following:

0 N2-EOP-PC. PRIMARY CONTAINMENT CONTROL "HYDROGEN' N2-EOP-PCH, HYDROGEN CONTROL Chemistry procedure N2-CSP-13for Post Accident Sampling.

3 Performance of this Subsection may be required by the EOPs Changes to this Subsection (inctudingrenurnbenng) are required to be reviewed by the EOP Coordinator

4. Keys PA235. PA1235 and PA2235 are interchangeable.

5 E only one division is being placed in service. THEN only that division is required to be placed in "OVERRIDE in Step 2 1.

+

2.1 IF a LOCA isolation signal is present, using a PA235 KEY place the following keylock switches to OVERRIDE' e ISOL VLV OVERRIDE on 2CECPNL873 (875) 2.2 At 2CEC*PNL873 (875). place the Division l(11)SAMPLE PATH SELECTOR switch in the RESET position NOTE: The CMS vafves and the selector switch will be positioned as directed by Chemistry in accordancewith N2-CSP-13.

2.3 H2102 Analyzers may be placed rn service for post accident sampling per Section F.l .O.

30 Loss of H202 Heat Trace 12HTSPNL001 &2HTS"PNL003)

(C7)

NOTE: 1. Heat Tracing is required for operahility of the assocrated H202 analyzer. to prevent excessive condensationof moisture during accident conditions

2. The readout assemblies in the heat trace panel are number by card file location. The card file number is labeled on front of the readout assembly. (Ex: C2-4). The heat trace circuit numbers are associated with breaker number that supplies that channel (Ex: 2HTS-001-18) 3.1 If heat tracing is lost to 2CMS*PNL66A(B) evaluate the lost circuit per the table below and declare the associated H202 panel inoperable if required.

Page 26 N2-OP-82 Rev 06

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement IR Origin Source Question 55 1 295037 EKI .07 3.4 N NA LOK ' Grp 10 CFR 55.41(b) 10 ' LOD (1-5) Reference Documents H 1 EOP Basis Document Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Shutdown margin QUESTION 55 Following a reactor scram, the following conditions exist:

0 Control Rod 26-39 remains at position 48.

0 Control Rod 06-15 remains at position 04.

0 Control Rod 18-31 remains at position 02.

Which one of the following describes the status of the reactor and action required, if any, to ensure that the reactor is Shutdown WITHOUT Boron?

A. Shutdown WITHOUT Boron. No actions are required.

9. NOT Shutdown WITHOUT Boron.

Condition achieved by INSERTING rod 06-15 to position 02.

C. NOT Shutdown WITHOUT Boron.

Condition achieved by INSERTING rod 26-39 to position 02.

D. NOT Shutdown WITHOUT Boron.

Condition achieved by INSERTING rods 06-15 and 18-31 to position 00.

Correct Answer: D Demonstration of Shutdown Margin is one method of determination authorized by EOPs. One Control Rod at position 48, with others fully inserted. Since all rods are not at least to 02 and SDM configuration is not established, the reactor is not shutdown under all conditions, without reliance on boron.

................................................................................................................................................................................................................................................................F Plausible Distractors:

A is plausible; would be true with ALL INSERTED to at least 02 OR ONE Control Rod at position 48.

B is plausible; action does not produce Shutdown Under All Conditions WITHOUT Boron.

C is plausible; action does not produce Shutdown Under All Conditions WITHOUT Boron.

Page 63 of 88

EOP-RPV Step (Ovemde Step 2)

While in this procedure:

IF I THEN The reactor will EdBT stay shhuldcwn witnoul boron FAILURE TO SCRAM:

OR It is u n k w n IItho reactor will s h y shuldown wi:houl bamn Ex8 thi3 ufwadurs + Enrar EOP C5 Discussion Positive confirmation that the reactor will remain shutdown u d c r all conditions is best obtained by detemning that no control rod IS withdrawn beyond notch position 02 (the Maximum Subcritical Bankcd Withdrawal Position (MSBWP)). The MSBWP is the greatest bankcd rod position at which the reactor will remain shutdown under all conditions. Refer to Section 14 for a detailed discussion of the MSBWP.

Other critena can also be used to demonstrate that the reactor will remain shutdown.

Possibilities include the existence of the core design basis shutdown margin with the single strongest control rod full-out and 311 other control rods full-in and compliance with TcchnicaI Specification requirements goveining control rod positioii and the allowable number of inoperable control rods. In cases such as thcse, the control room operating crew is able to make the deterrnkarion themselves. In other cases, however, it is expected that a reactor engineer or other technically qualified member of the statf will make the determination. Note that the instruction requires a positive determinalion, not only that the reactor I S shutdown, but that it will remain shutdown, without reliance upon boron, under aont-case cold shutdown conditions. The phrase "without boron" does not imply that the condition cannot he met if boron has been injected, hut that credit cannot he taken for the negative reactivity contnhutcd by the boron. Control rod insertion alone must provide the necessary shutdown margin.

NER-2M-039, REV. 5 4- 7

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal 1

I i

RO 1 Tier K/A Number 1 Statement IR Origin Source Question 5 6 1 1 1 295038 ] EK2.05 3.7 I B 2002 Clinton

....................... i.......................  :...............................................................................................

I NRCExam LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents F / I / i N2-EOP-SC/RR Rev 10

................................................................................................................................................................................................................................................................Z Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: Site emergency plan QUESTION 56 Which one of the following is the LOWEST Emergency Action Level for an Offsite release rate which REQUIRES entry into the Radioactivity Release Control Leg of N2-EOP-SC/RR, Secondary Containment Control and Radioactivity Release Control?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency I

Correct Answer: B ALERT is the LOWEST Emergency Action Level which REQUIRES entry into the Radioactivity Release Control Leg of N2-EOP-SC/RR, Secondary Containment Control and Radioactivity Release Control Plausible Distractors:

A is plausible; DOES NOT REQUIRE entry into the Radioactivity Release Control Leg C is plausible; NOT the LOWEST EAL.

D is plausible; NOT the LOWEST EAL.

Page 64 af 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement IR Origin Source Question 57 1 600000 AKI .02 2.9 N NA LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents F 1 EPIP-EPP-28 Rev 13 Knowledge of the operational applications of the following concepts as they apply to Plant Fire On Site: Fire Fighting QUESTION 57 After receiving indications of a plant fire INSIDE the Protected Area the following occurs:

0 Fire Protection Water (FPW) system actuates 0 Control Room Operators sound the Plant Fire Alarm 0 Operations Support Center is ACTIVATED Which one of the following is the first action required of the Fire Brigade Leader per EPIP-EPP-28 Fire Fighting?

A. Contact the Operations Support Center and request a briefing.

B. Proceed to the vicinity of the fire and establish a command post.

C. Proceed to the scene of the fire and provide confirmation of the fire to the CRO.

D. Proceed to the scene of the fire and direct the Fire Brigade in extinguishment of the fire.

Correct Answer: A with the OSC activated, it is required to contact the OSC and obtain a briefing. Reactor Operators are qualified Fire Brigade Leader. Fire Brigade Leader actions are unique to the Licensed Operator level.

Plausible Distractors:

B is plausible; would be true if the OSC were NOT activated.

C is plausible; would be true for an unconfirmed fire with the OSC NOT activated. The FPW System actuations serve as confirmation, with no Fire Brigade Leader dispatch required.

D is plausible; would be true AFTER obtaining an OSC Briefing and establishing command post in the vicinity of the fire.

Page 65 of 88

3.2.4 Fire Brigade Leader Actions

a. When the Station fire alarm is sounded AND the OSC is activated, then:

1. Acknowledge receipt of the alarm to the CSO

2. Based upon the location of the fire, report to and assess the fire scene.

Establish a command post in a Iocation away from the fire scene, from which fire fighting activities can be safely directed.

NOTE: Consideration should be given to establishing the command post at the fire panel closest to the event. However, the location of the command post will be dictated by actual conditions and is at the discretion of ?he Brigade Leader.

3. Provide direction to fi+ebrigade members as appropriate.

4. Inform the CSO of actuat conditions at the scene, confirm the fire condition, if appropriate, and report location of command post.

5. Request off-site assistance, if needed. I

b. \n-'hen the Station fire alarm is sounded & NJ the OSC is activated. then:

1. Acknowledge receipt of the alarm to the CSO

2. Contact the OSC Cornmunicator at ext. 2282 or via Gaitronics or fire radio.

3. Request briefing.

4. Report to and assess the fire scene.. briefing fire brigade members concerning safety considerations received from the OSC (personal and radiological safety)

5. Establish a command post in a location away from the fire scene, from which fire fighting activities can be safely directed.

Page 4 EPIP-EPP-28 Rev 13

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier 1 KIA Number Statement i""" IR 1 Origin 1 Source Question 1 5 8 : l I 700000 i AA1.03 3.8 N 1 NA LOK i Grp

• .......................... I 10 CFR 55.41(b) 5 i LOD (1-5) Reference Documents

..................................................................................................................... ~

F 1 1 ' i N2-SOP-68 Rev 2 PROVlDE CAPAB ILlTY CURVE Ability to operate and/or monitor the following as they- apply . . - to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Voltage regulator controls QUESTION 58 Following a Grid Disturbance, conditions are as follows:

0 Generator Power is 1200 Mwe.

0 Reactive Power is 550 MVAR (LAG).

0 Generator Hydrogen Pressure is 75 psig.

The System Dispatcher has requested additional reactive load support to maintain grid voltage.

Considering the attached Capability Curve, which one of the following actions is required?

A. RAISE Recirculation Flow to increase the Reactive Load on the Generator.

6. LOWER Recirculation Flow, because Generator Load limits have been EXCEEDED.

C. MANUALLY RAISE the Voltage Regulator setting to increase the Reactive Load on the Generator.

D. MANUALLY LOWER the Voltage Regulator setting, because Reactive Load limits have been EXCEEDED.

Correct Answer: D MANUALLY LOWER the Voltage Regulator setting to decrease Reactive Load on the Generator because the generator is currently operating above the maximum VARS rating of 500m MVAR..

Plausible Distractors:

A is plausible; but will raise REAL Load. This does not pick up maximum Reactive Load.

B is plausible; but will lower REAL Load and Reactive Load is what is out of specification..

C is plausible; would be true with Reactive Loadingc500MVAR.

Page 66 of 88

ATTACHMENT 1: ESTIMATED CAPABILITY CURVE i AIJTICN DO N l Y EXCEED 6017 FIELD AMPS LRR 580 k1MR TO E I I J S r n a ? TO bEN ATP 4 POLE 1 3% 4%KVA iHOrJ RFFJ 250011 V R T S 20 PF 58 SCR 75 PSlG HYDROGEI?;FF(ES5IJhE 5% V(3L TS EXCITATION 1600 r w E l 1200 1000 800 (9

2 400 tvr 200 a

a

-0x

-l 00 8

2

.c -200 Q

tf: -400

-600

-800

- 1680 CIJRVE AB LIUITECIBY FIELD HEATING CURVE Br LIktITEC~3Y ARhlATJRE HEATIN(;

CURVE L D LlfbllTED BY ARMATURE 3RE END HEATING Page 7 N2-SOP-68 Rev 02

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement IR Origin Source Question 59 1 295008 AA2.05 2.9 N NA LOK ' Grp I O CFR 55.41(b) 10 LQD (1-5) ' Reference Documents H 2 N2-EOP-RPV Rev 10 Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL : Swell QUESTION 59 The plant was operating at full power, when the following occurred:

0 BOTH Feedwater Pumps TRIPPPED 0 The reactor automatically scrammed.

0 ONLY one Control Rod is at position 48.

0 ALL OTHER Control Rods are FULLY INSERTED.

0 HPCS initiation RAISED RPV Water L.evelfrom 100 inches.

0 HPCS injection stopped as RPV Water Level reached 200 inches.

Plant conditions are currently:

0 Reactor pressure 700 psig, rising at 10 psig per minute.

0 MSlVs are OPEN.

0 The OPERATING CRD Pump TRIPPED.

Which one of the following is the RPV Water Level response over the next TEN MINUTES, and what action will be required?

A. RISE due to SWELL. ALLOW steam off to lower RPV Water Level BELOW 202.3 inches.

B. LOWER due to SHRINK. USE HPCS to maintain RPV Water Level ABOVE 159.3 inches.

C. LOWER due to SHRINK. USE ONLY RClC to maintain RPV Water Level ABOVE -14 (minus 14) inches.

D. RISE due to SWELL. TERMINATE AND PREVENT Injection Systems to lower RPV Water Level BELOW 100 inches.

Correct Answer: A HPCS injected (I00 inches x 200 gal per inch=) 20,000 gallons of cold CST water. As this water is heated, SWELL occurs. It is required to maintain RPV Water Level below Level 8 (202.3 inches). Steaming with no injection will lower RPV Water Level. Shrink cannot occur because heatup and pressurization of saturated Page 67 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal system is in progress with NO steam voids. ALL SRVs and Bypass Valves are shut for the next ten minutes because Reactor Pressure will be below 800 psig.

Plausible Distractors:

B is plausible; with Reactor Pressure rising, and if a substantial void fraction were assumed - SHRINK would identify a misconception.

C is plausible; with Reactor Pressure rising, and if a substantial void fraction were assumed - SHRINK would identify a misconception. The minimum ATWS Level control band limit is -14 inches, which would identify a misconception regarding the withdrawn rod.

D is plausible; The highest ATWS Level control band limit is 100 inches, which would identify a misconception regarding the withdrawn rod.

Page 68 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 60 1 295015 2.4.2 4.5 N NA LOK Grp 10 CFR 55.41(b) 7 LOD (1-5) Reference Documents H 2 or 10 N2-0P-36B Rev 2 Incomplete Scram - Emergency Procedures / Plan: Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

QUESTION 60 The Turbine Generator TRIPPED at full power.

0 NO Control Rod motion occurred.

0 NO RPS actuations occurred.

0 Reactor Pressure is 1055 psig, RISING.

With these conditions:

(1) Which one of the following lists the FIRST EOP ENTRY required?

AND (2) With NO operator action, what automatic actuation will terminate this transient?

A. (1) N2-EOP-RPV, RPV Control entry is required.

(2) Backup Scram Valve actuation will terminate this transient.

B. (1) N2-EOP-C5 Failure To Scram entry is required.

(2) Backup Scram Valve actuation will terminate this transient.

C. (1) N2-EOP-RPV, RPV Control entry is required.

(2) Redundant Reactivity Control System actuation will terminate this transient .

D. (1) N2-EOP-C5 Failure To Scram entry is required.

(2) Redundant Reactivity Control System actuation will terminate this transient.

Correct Answer: C With Reactor Pressure exceeding 1052 psig and power above 4%

when a scram is required - it is required to enter N2-EOP-RPV, RPV Control. When Reactor Pressure exceeds 1065 psig, RRCS actuation will occur.

- -I Page 69 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal Plausible Distractors:

A is plausible; Backup Scram valves require RPS actuation.

B is plausible; N2-EOP-C5, Failure To Scram requires transition from N2-EOP-RPV, RPV Control. Backup Scram valves require RPS actuation.

D is plausible; N2-EOP-C5, Failure To Scram requires transition from N2-EOP-RPV, RPV Control.

Page 70 of 88

3. OFF- NORPlAL QPERWTIONS 1.0 Reactor Hiqh P r e s s u r e A u t o I n i l i a t i o n e: The high pressure i n i t i a t i o n s i g n a l seals i n .

1.l Clonfiriti Reactor pressure i s 2 1065 pig.

1.2 %xify the following actione i n d i c a t i o n s immediately occi.ir u p n receipt. of the i n i t . i a t . i u n s i g n a l :

F.eacto~ Scram (RRI function)

Reactor Fiecirculritiori Pumps transfer %A LFt4G power supp1y Division I I1 RRCS ARI INITIATE amber light-s are lit at. 2 C E P P N L 6 3 3 1.3 Vexify tlie fo:l~wing w t . i o n s occiir 25 t;econds AFTER receipt w i t h APP.Ms INOP E NOT of i n i t i a t i c m s i g n a l AND c t m c ~ r r e i i %

dowrrscale :

• Reactor R e c i r e u l a t i u n Primps LFMG power supply t r i p s Feedwater f l o w c:on%.rolrunback i n i t i a t e s AI.lD t h e fee8wat.er min J Q W valves f a i l open {Sealed in f u r 21; seconds )

1.4 I F a Feedwater runhack has c3cci.i B a AFTER the 2 5 second s o a l in p e r i o d , t a k e manual crrntrol of Feetiwater a s d i r e c t e d by the sss.

1 , 5, V e ri fy t h e followirig act.ions occur 96 seconds AFTER r e c e i p t o f i n i t i a t i o n signal iiHD coIimr1'ent wit.li APRMs INQP MOT downscale:

. ACS System i s o l a t e s 7 . 7 0 SLS Gysteii~i r i i t i a t e s i ~ 1 Di n j e c t s 1.6  ::HEN directed by t h e S S S , reset t,he RRCS i n i t i a t i o n s i g n a l i n accordance w i t h S:ibscction H . 3 . 0 .

t

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement IR Origin Source Question 61 1 295020 AKI .01 3.7 B 2000 NMP-2 NRC Exam LOK Grp ' 10 CFR 55.41(b) 5 LOD (1-5) Reference Documents H 2 N2-EOP-RPV Control Rev 10 Knowledge of the operational implications of the following concepts as they apply to INADVERTENT CONTAINMENT ISOLATION : Loss of normal heat sink QUESTION 61 The EOPs have been entered following a plant trip due to an inadvertent containment isolation. The following conditions exist:

Group 1 isolation signal has occurred.

Group 8 isolation signal has occurred.

Drywell pressure is 1.71 psig.

RPV pressure is 1050 psig and rising.

Which one of the following systems is used for Reactor Pressure control?

A. Turbine Bypass Valves

9. Main Steam Line Drains C. Reactor Core Isolation Cooling D. Steam Condensing Mode of RHR Correct Answer: C With Group 1 and 8 Isolations present, ONLY RClC may be used for Pressure Control Plausible Distractors:

A is plausible; listed as the Preferred Pressure Control method in N2-EOP-RPV Control, cannot be used due to MSIV / MSL Drain Isolation (Group 1).

B is plausible; listed as an Alternate Pressure Control method in N2-EOP-RPV Control, cannot be used due to MSlV / MSL Drain Isolation (Group 1).

D is plausible; listed as an Alternate Pressure Control method in N2-EOP-RPV Control, cannot be used due to High Drywell Pressure.

Page 71 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal I:

i  :

RO I Tier K/A Number Statement I IR 1 Origin 1 Source Question I 6 2 1 1 1 295029 ~ ~ ~ 2 . 0 I 3 3.3 N NA

.............. i .................................................................................................... ....................... 2 ............................................. / ............................................................................ "..........................................................................................

LOK Grp 10 CFR 55.41(b) 10 LQD (1-5) j Reference Documents H 1 2 1 I N2-EOP-PC, Primary Containment j Control Rev 12 I PROVIDE SRV TAILPIPE LEVEL

......................................................................................................................................................................... i LIMIT Curve

............................................................... N in question stem.

i Knowledge of the interrelations between HIGH SUPPRESSION POOL WATER LEVEL and the following: HPCS: Plant-Specific QUESTION 62 The plant has experienced a transient, with the following conditions:

0 Reactor Pressure is 750 psig, lowering.

0 RPV Water Level is 110 inches, stable.

0 Suppression Pool Water Level is 21 1 feet and rising.

S R V T ~Pipe ~ I tevei Limit Which one of the following injection systems is ALLOWED to be used to maintain RPV Water Level per the Emergency Operating Procedures?

A. Reactor Feedwater Pump B. Control Rod Drive Hydraulic Pump C. HPCS with suction from the Suppression Pool D. RClC with suction from the Condensate Storage Tank Page 72 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal Correct Answer: C With Suppression Pool Water Level approaching the SRV Tailpipe Level Limit, N2, EOP-PC, Primary Containment Control prohibits injection sources from outside the Containment.

Plausible Distractors:

A is plausible; may NOT be used because the Hotwell is a source outside Containment.

B is plausible; may NOT be used because the Hotwell AND the CST are sources outside Containment.

D is plausible; CST is a source outside Containment.

Page 73 of 88

r f Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 63 1 295033 EKI .02 3.9 N NA LOK Grp 10 CFR 55.41(b) 12 LOD (1-5) Reference Documents F 2 GAP-RPP-08 Rev 15 Knowledge of the operational implications of the following concepts as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS : Personnel protection QUESTION 63 Due to an EMERGENCY condition, access is required to a Locked High Radiation Area to open a valve. It is expected that this task will result in 500 mRem TEDE.

Which one of the following describes how the FASTEST access will be obtained in accordance with Radiation Protection procedures?

A. It is required to be authorized on a specific Radiation Work Permit for the Locked High Radiation Area PRIOR to access.

B. It is required to be authorized on a specific Radiation Work Permit and attend an ALARA brief PRIOR to accessing the Locked High Radiation Area.

C. It is required to be accompanied by a Radiation Protection Technician WHILE accessing the Locked High Radiation Area. Master Keys stored in the "break-to-enter" key box in the Control Room will provide access.

D. It is required to be a volunteer, attend a brief on potential biological consequences of emergency exposure, and be accompanied by a Radiation Protection Technician WHILE accessing the Locked High Radiation Area.

Correct Answer: C It is required to be accompanied by a Radiation Protection Technician WHILE accessing the Locked High Radiation Area. Master Keys stored in the "break-to-enter" key box in the Control Room will provide access.

Plausible Distractors:

A is plausible; would be true for a non-emergency.

B is plausible; would be true for a non-emergency.

D is plausible; would be true for emergency exposures exceeding 5 rem. (EPIP-EPP-15)

Page 74 of 88

3.4 Access Requirements for Very High Radiation Areas In addition to the controls of Sections 3 2, 3.3-NOTE:With the exception of Radiation Protection ( ANSI-qualified technicians and management staff) site personnel shall not be in control of posted Locked and Very High Radiation Area keys.

3.4.1 To the extent possible entry should be forbidden unless there IS a sound operational or safety reason for entering 3.4.2 Entry into Very High Radiation Areas shall be approved by RP Manager AND the SSS 3.4.3 A specific RWP is required for entry into Very High Radiation Area 3.4.4 An RP Technician should accompany the person entering the Very High Radiation Area to the entryway to determine radiation conditions at the time of entry and render assistance if necessary 3.5 Ernerqency Access (C3) 3 5 1 The SSS may use or authorize use of the HRA master keys stored in the "break-to-enter" key box located in the Control Room. The box contains keys to access high, locked high and very high radiation areas. The SSS shall utilize an RP Technician to ensure L compliance with Technical Specifications NOTE: RP Supervisory approval for Locked or RP Manager approval for Very High Radiation Area key issuance is not required in emergency sttuations or urgent operational conditions as determined by the SSS The RP branch shall initiate a Deviation Event Report to document such occurtences.

3.5.2 When notified by the SSS that HRA master keys from the "break-to-enter" box have been used, the RP branch shall initiate a Deviation Event Report.

3.6 Kev Holder Responsibilities NOTE: With the exception of Radiation Protection ( ANSI-qualified technicians and management staff), site personnel shall not be in control of posted Locked and Very High Radiation Area keys.

3 6.1 Personnel issued a key for access to a High. Locked High or Very High Radiation Area shall maintain positive access control to the area. Control should include:

a. Ensuring individuals accessing area meet the applicable minimum exposure requirements for that area.

Page 5 GAP-RPP-08 Rev 15

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal i RO 1 Tier 1 K/A Number I Statement ' IR I Origin 1 Source Question I

6 4 j 1 ! 295035  ! EA1.02 3.8 B Susquehanna

.......................................... ~ . . ............... 1 2002NRCExam  !

i LOK . Grp ' 10 CFR 55.41(b) 7

~... ..................................................................................................................................... .......,.. .....................................................

LOD (1-5) Reference Documents i F  : 2 ! j SG N2101261000C01 Ability to operate and/or monitor the following as they apply to SECONDARY I CONTAINMENT HIGH DIFFERENTIAL PRESSURE: SBGT/FRVS I QUESTION 64 Which one of the following identifies the Standby Gas Treatment System (SGTS) response to a HIGH Reactor Building pressure (positive RB pressure) condition WHILE SGTS Train A is operating in response to a LOCA signal start?

A. Recirculation Throttle Valve, 2GTS*PV5A, will modulate in the OPEN direction.

B. Recirculation Throttle Valve, 2GTS*PV5AI will modulate in the CLOSED direction.

C. Filter Train Inlet Valve, 2GTS*AOV2AI will modulate in the OPEN direction.

D. Filter Train Inlet Valve, 2GTS*AOV2A, will modulate in the CLOSED direction .

Correct Answer: B With HIGH RB Pressure, more flow through SGTS Train A will be obtained by Recirculation Throttle Valve, 2GTS*PVSA, modulating in the CLOSED direction.

Plausible Distractors:

A is plausible; would be true for Reactor Building Pressure too low (or negative).

C is plausible; Filter Train Inlet Valve strokes OPEN when the SGTS Train is started. In this case, it was already OPEN due to SGTS Train initiation due to a LOCA signal.

D is plausible; Filter Train Inlet Valve strokes CLOSED when the SGTS Train is secured.

Page 75 of 88

Component Function Location Cross Connect Varve When closed, isolates the filter Standby Gas Treatment

• 2GTS+AOV28A trains from one another Building

• 2GTSIAOV28B When both valves are open, allows AOVZHA. .. EL 261 operating fan to pull air through AOV28E. ... EL 261 offline filter train for decay heat removal

• Nor nially upen Recirculation Tlirottle Throttled to vary the amount of fan Standby Gas Treatment Valve discharge air recirculated back t o Building 2GTS+PV5A the filter train inlet to maintain FN1A.. .. . .EL 775 2GTS*PVSB proper system flow for efficient FNlB . .. .. EL 275 filter train operation and a negative differential pressure of I 0.25 inch WG in the Reactor 8uilding Radiation Monitor Monitors the radioactivity level of Main Stack

• 2GTS-RE105 the effluent discharge and provides RE105 .'. ,.,EL 261 that signal to the Digital Radiation Monitoring System (DRMS).

Upon receipt o f a high radiation signal will cause the Containment Purge Valves to close

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 65 1 295036 EK3.04 3.1 B NMP2 Bank SYSlD 15931 LOK Grp 10 CFR 55.41(b) 10 ' LOD (1-5) Reference Documents H 2 N2-EOP-SC/RR Rev 10 Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL : Pumping secondary containment sumps QUESTION 65 The plant is operating at 100% power with the following:

A fire has occurred in the RClC room.

Fire system has actuated to suppress the fire.

Later, RClC room sump High-High level alarm is reached due to fire suppression system operation to control several reflash fires.

Which one of the following action(s) is (are) required by N2-EOP-SC, Secondary Containment Control. for these conditions?

A. COMMENCE a controlled reactor shutdown.

B. RUNBACK Reactor Recirc and initiate a manual scram.

C. CONTINUE to control reactor building sump levels using available sump pumps.

D. ISOLATE the fire suppression system for RClC area and attempt to control the fire by other means.

Correct Answer: C Secondary Containment Control EOP authorizescontinued use of fire protection systems. Rising water levels are mitigated by operating available sump pumps.

The fire poses a greater threat than the water level.

Plausible Distractors:

A is plausible; would be true if TWO or more areas EXCEED Max Safe Water Level.

B is plausible; would be true if a primary system were discharging and ONE area EXCEEDS Max Safe Water Level.

D is plausible; EOPs require primary systems discharging into the Secondary Containment IS0LATED.

Page 76 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin 1 Source Question 66 3 Generic 2.1.4 3.3 I N j NA LOK Grp 10 CFR 55.41(b) 10 LQD (1-5) Reference Documents F NA 10 CFR 55.53 (i)

Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 10CFR55. etc.

QUESTION 66 Which one of the following is specified in 10 CFR 55, Operator Licenses as a REQUIREMENT to MAINTAIN an ACTIVE Nuclear Regulatory Commission License?

The licensee SHALL:

A. pass a medical exam every two years.

6. pass one comprehensive requalification written examination AND one operating examination every six years.

C. actively perform the functions of the licensed position on a minimum of seven 8-hour shifts or five 12-hour shifts per calendar year.

D. complete seven 8-hour shifts or five 12-hour shifts under instruction per calendar year, including a plant tour and all turnover tasks.

Correct Answer: A is correct. Biennial medical exams are REQUIRED by 10 CFR 55.53 (i).

Plausible Distractors:

B is plausible; it is required to complete an ANNUAL operating exam and a comprehensive written examination per 10 CFR 55. Licenses are granted for a period of six years.

C is plausible; would be true if statement ended with "per calendar quarter".

D is plausible; and is true for REACTIVATION of an INACTIVE License, NOT REQUIRED to MAINTAIN an ACTIVE License - if "per calendar quarter".

Page 77 of 88

5 55.53 Conditions of licenses.

Each license contains and is subject to the following conditions whether stated in the license or not:

(a) Neither the license nor any right under the license may be assigned or otherwise transferred .

(b) The license is limited to the facility for which it is issued.

(c) The license is limited to those controls of the facility specified in the license.

(d) The license is subject to, and the licensee shall observe, all applicable rules, regulations, and orders of the Commission.

(e) If a licensee has not been actively performing the functions of an operator or senior operator, the licensee may not resume activities authorized by a license issued under this part except as permitted by paragraph (f) of this section. To maintain active status, the licensee shall actively perform the functions of an operator or senior operator on a minimum of seven %hour or five 12-hour shifts per calendar quarter. For test and research reactors, the licensee shall actively perform the functions of an operator or senior operator for a minimum of four hours per calendar quarter.

(f) If paragraph (e) of this section is not met, before resumption of functions authorized by a license issued under this part, an authorized representative of the facility licensee shall certify the following:

(1) That the qualifications and status of the licensee are current and valid; and (2) That the licensee has completed a minimum of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of shift functions under the direction of an operator or senior operator as appropriate and in the position to which the individual will be assigned. The 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> must have included a complete tour of the plant and all required shift turnover procedures.

For senior operators limited to fuel handling under paragraph (c) of this section, one shift must have been completed. For test and research reactors, a minimum of six hours must have been completed.

(g) The licensee shall notify the Commission within 30 days about a conviction for a felony.

(h) The licensee shall complete a requalification program as described by 9 55.59.

II, (i) The licensee shall have a biennial medical examination.

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier . K/A Number Statement IR i Origin 1 Source Question 67 3 : Generic  ; 2.1.41 2.8 1 B I NMP-2 Bank

........................... j ........................................................................................................................ ................................................ .................................... 8..

I SYSID13134 LOK Grp 10 CFR55.41(b)7 i LOD (1-5) Reference Documents F NA \ 02-OPS-00 1-234-2-0 1 page 18; N2-OP-39 H.6 Knowlec ge of the refueling process.

QUESTION 67 Given the following conditions:

Core reload is in progress.

Refuel platform is over the core.

Main Hoist is LOADED and in the NORMAL UP position.

0 The Refueling Platform air compressor head cracks.

The air compressor is manually isolated but not before the air receiver is completely depressurized.

Which one of the following describes the failure mode of the fuel grapple and its impact on refueling operations?

A. CLOSED and can ONLY be manipulated manually.

B. OPEN and can be manually supplied from Service Air.

C. OPEN and refueling operations must be SUSPENDED.

D. CLOSED and can be manually supplied from Service Air.

Correct Answer: D The fuel grapple fails CLOSED and can be manually supplied from Service Air.

Plausible Distractors:

A is plausible; identifies misconception regarding backup from Service Air.

B is plausible; identifies misconception regarding grapple failure mode.

C is plausible; identifies misconception regarding grapple failure mode.

Page 78 of aa

H OFF-NORMAL PROCEDURES (Cont) 55 Press back ball in air line on Aux Hoist to allow moisture AND air to bleed off.

NOTE: When performing the next step the tool will only actuate tn one direction because only one air line IS connected 56 Actuate Tool using GRAPPLE ENGAGE AND RELEASE push buttons UNTIL ALL moisture is expelled from disconnected port on Tool 5.7 Reconnect air line on Tool 5.8 Disconnect first air line on Tool.

NOTE: When performing the next step the Tool will only actuate in one direction because only one air line is connected 59 Actuate Tool using GRAPPLE ENGAGE AND RELEASE push buttons UNTIL ALL moisture is expelled from disconnected port 5.10 Reconnect air line on Tool.

5 11 Perform an operational test on Tool in accordance with applicable Subsection OR procedure.

I,6.0 SuPplvina Refuel Bridge Air System From Service Air System 61 Perform the following.

Place 2FNR-SWS2C Refuel 6ridge Air Compressor Disconnect Switch, in OFF e

Place 2FNR-SWSZD, Refuel Bridge Air Dryer Disconnect Switch, in OFF Open 2FNR-V4. REFUEL BRIDGE AIR DRYER BYPASS.

Close 2FNR-V6, REFUEL BRIDGE AIR DRYER INLET ISOF.

Close 2FNR-117, REFUEL BRIDGE AIR DRYER OUTLET ISOL.

E:The following Etectricalihlechanical Service Boxes contain Service Air connections 2FNR-ESB07 2FNR-ESB08 2FNR-ESB09 2FNR-ESB10 2FNR-ES61 I 2FNR-ESBI 2 2FNR-ESB13 2FNR-ESB14 6.2 Connect an air hose of suficient length to aflaw freedom of Bridge movement to the Sewice Air connection in one of the Electrical'MechanicaIService Boxes on the Refuel Floor.

6.3 Route air hose such that it will NOT be run aver during Bridge movement.

Page 141 N2-OP-39 Rev 08

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal

................................................................................................................................................................................................................................................................0 RO Tier K/A Number j Statement I IR Origin Source Question 68 3 Generic 2-2-39 3.9 I N NA LOK Grp 10 CFR 55.41(b) 10

! LQD (1-5) 1 Reference Documents F NA  : LCO 3.1.5 C.l and D.l Knowledge of less than or equal to one hour Technical Specification action statements for systems.

QUESTION 68 With a reactor startup in progress, the following conditions exist:

0 Reactor Pressure is 60 psig.

0 The OPERATING Control Rod Drive Pump TRIPS.

0 Charging Water Header Pressure is 0 psig.

0 Control Rod 06-15 is at position 48 and has an Accumulator N2 Pressure of 875 psig.

Which one of the following actions are required to comply with Technical Specifications?

A. IMMEDIATELY SCRAM the reactor.

B. IMMEDIATELY START the standby RDS Pump.

C. WITHIN 20 MINUTES, INSERT Control Rod 06-15.

D. WITHIN 20 MINUTES, DECLARE Control Rod 06-15 INOPERABLE.

Correct Answer: A With an inoperable accumulator, withdrawn control rods must be verified inserted immediately. That cannot be completed. It is required to scram the reactor by placing the RMS in SHUTDOWN.

Plausible Distractors:

B is plausible; does not meet C is plausible; would be true with Reactor Pressure > 940 psig with TWO inoperable accumulators.

D is plausible; would be true with Reactor Pressure > 940 psig with ONE inoperable accumulator.

Page 79 of 88

Control Rod Scram Accumulators 3.1-5 3.1 REACTIVITY CONTROL SYSTENS 3.1.5 Control Rod Scram Accumu I ators LCO 3.1.5 Each control rod scram accumulator shall be OPERABLE.

APPLICABILITY: MODES 1 an ACTIONS CONDITf ON REQUIRED ACTION COMPLETION TIME A. One control rod scram A. 1 --------NOTE---------

accumulator inaperable Only applicable i f w i t h reactor steam the associated dome pressure control rod scram 2 900 pslg. time was within the limits o f Tab1e 3 . 1 . 4 - 1 during the last scram time Surveillance.

--I_--- ---------.- --_--

Declare the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> assoclated control rod scram time "slow."

!3 A.2 Declare the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> associated control rod inoperable.

(cont inued)

NMPE 3.1.5-1 Amendment 91

Control Rod Scram Accurnul ators 3.1.5 ACTIONS (continued)

CONOITION REQUIRED ACTION CMPLETION TIME B. Two or more control 1. I Restore charging  ?.O minutes from rod scram accumulators water header pressure jiscovery of inoperable w i t h t o 2 940 p i g . bndition B reactor steam dome :oncurrent w i t h pressure 2 900 psig. Eharging water Reader pressure c 940 p i g w

3.2.1 --------NOTE---------

Only applicable i t the associated control rod scram time was within the limits o f Table 3.1.4-1 during t h e f a s t scram time Survei 11ante.

-__.#... ...---_*----- I*- --

Declare t h e 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> associated control rod scram time

" SI OH. "

OR 8.2.2 Declare the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> dssucidted controt rod inoperable.

+'. One or more c o n t r o l

~~~~~~~~~~~~~1 ator:

V e r i f y a l l control rods associated with inoperable Immediately upon discovery of charging water reactor steam dome accumul atom are header pressure pressure < 900 psig. I f u l l y Inserted. < 940 p s i y NMP2 3.1.5-2 Amendment 91

I Control Rod Scram Accumulators 3.1.5 ACTIONS CONDITION REQU I RED ACT I ON CMPLFTTON TIME I

C, (continued) C.2 Declare the I hour associated control rod 4 noperabl e.

uired Action 8.1 or 0.1 ---.".*--- NOTE---------

Nut applicable i f a l l T i m not inoperable control met. rod scram accumulators are associated w i t h ful l y inserted control rods.

---

_--LII-------^c_-

Place the reactor mode switch in the shutdown posit'lon.

SURVEI LLANCE REQU IRENENTS SR 3.1.5.1 V e r i f y each control rod scram accumulator pressure i s 2 940 p s i g , I 7 days NMP2 3.1.5-3 Amendment 91

x * ^ t Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal I RO / Tier I K/A Number Statement IR Origin Source Question I 6 9 1 3 Generic 2.2.1 2 3.7 B 2002 NMP-2

"""""LOK . G r p ! . CFR 55.41 ~ . b ) I O NRC Exam LOD (1-5) Reference Documents

F / N A i GAP-SAT-01 need reference revision number Knowledge of surveillance procedures.

QUESTION 69 While performing a step in an Operations Surveillance Procedure, it has been determined that a normally open, motor operated Primary Containment Isolation Valve will not stroke in the closed direction, as required by the procedure.

Which one of the following identifies when the Technical Specification LCO action time is started per GAP-SAT-01, Surveillance Test Program?

A. As soon as the valve failure is recognized.

B. When the surveillance was logged as started.

C. When the surveillance is logged as complete.

D. At the time the surveillance was last satisfactorily completed.

Correct Answer: A When it is discovered that a component does not meet its Surveillance Requirements -the component is considered INOPERABLE at the time of discovery.

Plausible Distractors:

B is plausible; time of discovery is used.

C is plausible; time of discovery is used.

D is plausible; time of discovery is used.

Page 80 of 88

SR 3.0.1 SRs shall be met during the MOOES or other specified conditions in the

+ A ~ p ifor~individual ~ ~ unless otherwise stated in the SR. Failure to

~ i LCOS, meet it Sutvsilkince,whether stlchfailure is experienced during the performance of the Surveillance OT between performancesof the Surveillance, shall be failure to meet the LCO. Failure to p e r f m a Surveillance within the spdfied Frequency Shall be lasure to meet the LCO except as provided in SR 3.0.3. Suntellances do nat have to be performed on inoperable equipment Or variables outside specified limits.

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 70 3 Generic 2.2.2 4.6 B NMP-2 Bank SYSID 22775 LOK Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents F NA N2-OP-IOIA Rev 14 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

QUESTION 70 Plant startup is in progress with the following:

Mode switch is in StarVHot Standby.

0 RSCS Group 2 rods are being withdrawn using Continuous Withdrawal Reactor is Subcritical.

Which one of the following describes the criteria for using SINGLE NOTCH WITHDRAWAL per N2-OP-1OIA, Plant Startup?

A. Starting with RSCS Group 4 until criticality is achieved.

B. Starting with RSCS with Group 5 after the Reactor is critical.

C. When TWO SRMs approach 3 count rate doublings in RSCS group 4.

D. When TWO SRMs approach 3 count rate doublings prior to RSCS group 3.

Correct Answer: D When TWO SRMs approach 3 count rate doublings prior to RSCS group 3, SINGLE NOTCH WITHDRAWAL is required per N2-OP-IOlA, Plant Startup.

Plausible Distractors:

A through C are plausible; and ALL of these answer choices invoke Single Rod Withdrawal requirements too late in the startup process to meet the requirements of N2-OP-101A Page 81 of 88

ATTACHMENT 1 (Cont)

D. Technical Briefing Material I Control Rods Guidance for the conduct of reactivity control is contained in GAP-OPS-05 Reactivity Management (C4) All control rod motion shall be performed in stnct compliance with approved control rod sequence or Reactor Engineering instructions (C5) Conservative actron is required whenever an unexpected situation arises while positioning control rods (C7,19) e The following precautions are applicable when withdrawing control rods.

Critical predictions should be used only as a gross estimate of the critical rod pattern since there are many calculational uncertainties in the prediction process Criticality should be expected whenever control rods are being withdrawn.

Extra caution shall be used when pulling control rods in the region of criticality to avoid short periods.

Single notch control rod movement shall be performed from notch positions 00 to 36:

At the direction of the Reactor Engineer IF Estimated Critical Position (ECP) is prior to Rod Sequence Control System (RSCS) r:,d group 3.

IF 3 count rate doublings are approached on at least 2 SRMs prior to RSCS rod group 3.

Starting with RSCS rod group 3 until criticality is achieved.

For RSCS rod groups 3 and 4 after the reactor is critical, unless othetwise directed by Reactor Engineer.

Page 45 N2-OP-10 1A Rev 14

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal I

/ RO I Tier K/A Number  : Statement j IR Origin Source Question 1 7 1  : 3 ! Generic 2.3.4 j 3.7 . M i Listed Below i

LOK Grp : 10 CFR 55.41(b) 12 [ LOD (1-5) Reference Documents i .................................................................................................................. . . . . . . ........................................................................................................

1 F I N A ' j EPIP-EPP-15 Rev 7

................................................................................................................................................................................................................................................................z i Knowledge of radiation exposure limits under normal or emergency conditions QUESTION 71 A GENERAL EMERGENCY has been declared.

Emergency actions are necessary to perform operations to isolate a radioactive release for the PROTECTION of the DOWNWIND POPULATION.

Which one of the following is the HIGHEST LISTED (TEDE dose) emergency exposure that a VOLUNTEER may receive, which does NOT exceed the limits specified in EPIP-EPP-15, Emergency Health Physics Procedure?

A. 5REM B. 10 REM C. 25REM D. 50 REM Correct Answer: D Exposures in excess of 25 REM are authorized by EPIP-EPP-15 1 protecting large populations.

Plausible Distractors:

A is plausible; maximum for all activities during an emergency.

B is plausible; maximum for protecting property during an emergency.

C is plausible; is the maximum for non-volunteers to save a life or protect large populations.

Source Question:

A plant transient is in progress, with the following:

An emergency at the ALERT level has been declared (MODIFICATION-GEN EMERGENCY)

You are a radiation worker who has been authorized to receive an emergency radiati exposure for the purpose of protecting valuable plant equipment (MODIFICATION-DOWNWIND POPULATION)

Page 82 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal Which one of the following identifies the highest (TEDE dose) emergency exposure that you may receive, (MODIFICATION-VOLUNTEER)without exceeding the limits specified in EPIP-EPP-15, Emergency Health Physics Procedure?

A 5 REM B 10 REM OLD CORRECT C 25 REM D 50 REM NEW CORRECT Page 83 of 88

EMERGENCY EXPOSURE CONTROL ATTACHMENT 1 i

Sheet 1 of 4 1.0 EMERGENCY EXPOSURE LIMIT NOTES: 1 Emergency exposure limits are excluswe of current occupational exposure 2 Personnel shall not receive more than one emergency exposure per lifetime a The SSS,'ED or EDlRtvl (as applicable)shall provide authorization to waive or modify station radiation exposure or respiratory guidelines b The decision on how to address exposure received dunng an emergency situation shall be made by the SSSED or EDlRhl (as applicable). based upon recommendations provided by the RAM andjor ODAM as appropriate 1 I When an Alert or higher emergency classificationIS declared or when re-entryirecovery operations are being planned. emergency response personnel exposures may be addressed in either of the following two ways a If exposures are not expected to result in exceeding normal station radiation exposure guidelines, then the individuats may continue to work kinder those guidelines b If the potential exists to exceed normal station radiationworker exposure guidelines, then the emergency exposure guidelines described below should be assigned 1 2 If it is necessary for personnel to receive an emergency exposure in excess of 5 rem TEDE then

, a. Select the task from the table below and determine emergency exposure limit 1 TEDE Limit (rem) Activity b Select personnel to perform task f . Ensure that personnel who will receive emergency exposure are a) not a declared pregnaiit worker, AND b) have not received a previous emergency exposure, AND c) have not received a planned special exposu+e Page 5 EPIP-EPP-15 Rev 07

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier WA Number Statement IR Origin Source Question 72 3 Generic 2.3.1 3 2.9 N NA LOK

• Grp ' 10 CFR 55.41 (b) 10 ' LOD (1-5) ' Reference Documents F NA ARP 851254 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

QUESTION 72 With core alterations in progress, a fuel assembly contacts the core top guide, resulting in the following indications:

851254, PROCESS AIRBORNE RADN MON ACTIVATED alarmed.

0 Process Radiation Monitors 2HVR*RE14A-I and B-I, HVR ABOVE REFUEL FLOOR are alarming.

Which one of the following lists actions which will minimize the release of radioactivity?

A. OPERATE Turbine Building Ventilation.

B. RESTART Reactor Building Ventilation.

C. VERIFY that the Normal Control Room Air Intake Isolates and Special Filter Trains START.

D. VERIFY that Reactor Building Isolation occurs and the Standby Gas Treatment Systems START.

Correct Answer: D The unfiltered ground level release of radioactivity will be minimized by actions which VERIFY Reactor Building Isolation and initiation of Standby Gas Treatment.

Plausible Distractors:

A is plausible; is a required action of N2-EOP-SC/RR Radioactivity Release Control Leg, which is not required by these symptoms.

B is plausible; is a required action of N2-EOP-SC/RR Secondary Containment Control Leg, which is NOT permitted due to radiation levels.

C is plausible; is a required action when Control Room Intake Radiation Monitors alarm.

Page 84 of 88

ATTACHMENT 2 1 \ CGnt I 2CEC+PNL851 SERIES 200 ALARM RESPONSE PROC'EDURES Reflash. PPS PROCESS AIREORNE MDEI MON ACTIVATED 851254 45482,46536 Page 123'7 H2 -ARP- 0 1 Rev 00

ATTACHMENT 2 1 (Cont) 2CEC*PNL851 S E R I E S 2 0 0 AI,AR.P,M RESPONSE PROCEDURES NOTES: I. 2HVR*CAE14A/E arid 2HYR*C:AE32AiPj a l a r m s are not v a l i d w h e n t h t R*ac!t.or B u i l d i r g i s i s o l a t e d arid n o r m a l H3.m is shut d o w n .

2. 2HVR-CAB229 i s u s e d t o n \ G n i t . o r and t r e n d Reactor B u i l d i n g i7ent.ilation w h e n t h e R e i i c t o r B u i l d i n g is i s o l a t e d and the E m e r g e n c y HVR R e c i r c : 17ni ts i,2HVR*UC413A/Bi arii i n service AUTO C'CFiRECTIVE EQUIP. NO. AREA MONITOREG RESPONSE AC'T I C W 2 H W - CAB 19 5 - 1 FX EQUI P EXZIAUST NONE 4

~HvW - CAE195 - 2 kW EQUIP EXHAUST NONE 4 2WJW-CABl?E-l RW TANK ' \ r E N T NONE (1 2HW-CAB136-2 RK TANK VENT MCNE d 2HV"-CAB19T-1 RW BLUCJ VENT NONE Li 2HVw-CAB147 - 2 RW ELDG VENT NONE d 2H'Jh-CABl95-1 DECON AREA EXHAUST NPME il 2HVd-CAB199-2 DECOW AREA EXJAUST NONE Li A u t om? t i c' Respoils5 Gaseour A x 1 Level High i n i t i a t e s t h e f o l l o w i n g .

W

1. F S F l d g V e n t Emergency *UCT413A(B! starts. Shubs S u c t i o n T e s t rMPR*AOD34A(E) .

2 . Shuts KX B l d g V e i i t . i l a t i o i r Supp A i r Is01 DfvlPK

• kODlA[Hj.

3. S h u t s KX B l d g V e i i t i 1 a t . i c . n E s h A i r I s o l DMPR *A9D9AlB!.

4. Shuts KX E l c l y V e n t i l a t i o n R e f u e l Area E x h A i r Is01 riMPR *AODl@A(E!.

5. I n i t i a t a s St.a~~db:J Gas T r e a t . m e n t Fi1t.e.r T r a i n A S t - a r t Signal.

b. High R a d L e s r e I o r E q u i p m e n t F a i l u r e C o i n c i d e n t Chan. k arid C h a n . C! A u t o S t a r t 2HVC*FN2A ( D I V I.] and close f i l t c r t r a i n bypass valve 2HVC*MOVlA. High Rad Eqi.iipitimt F a i l u r e C o i n c i d e n t Chan. E! aria Chari. LI Auto S t a r t .

IHVC*FN2E i D I V 11) arid c l o s e s p e c i a l f i l t e r t . r a i n bvpas:;

valve 2HVC*MOV1E. &I-'? equipiiient , f a i l u r e a l a r m w i l l - r e s u l t i n d t.i-ip o f t h e a f f e ? t . e c radiation m o n i t o r .

45482,45536 P a g e 12 A N2 -ARP- 0 1 R e v 00

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier KIA Number Statement IR Origin Source Question 73 3 Generic 2.3.15 2.9 N NA LOK Grp 10 CFR 55.41(b) 12 LOD (1-5) Reference Documents F NA ARP 851254 I Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

QUESTION 73 With the Reactor Building Isolated and Standby Gas Treatment in service, which one of the following radiation monitors will STILL detect airborne radiation levels in the Reactor Building?

A. Standby Gas Treatment Monitor, GTS-RE105 B. Emergency Recirc Mode Monitor, HVR-RE229 C. Below Refuel Floor Monitors, HVR-RE32A and B D. Main Stack Gaseous Effluent Monitor, 2RMS-193 Correct Answer: B Emergency Recirc Mode Monitor will still detect airborne radiation in the Reactor Building.

Plausible Distractors:

A is plausible; will indicate SGTS filtered outlet radiation levels, which are NOT representative of the Reactor Building.

C is plausible; has no air flow following isolation and SGTS initiation.

D is plausible; will indicate stack effluent radiation levels, which are NOT representative of the Reactor Building.

Page 85 of 88

ATTACHMENT 2 1 (C'ont 1 2CEC*PNLR51 SERIES 230 ALARM RESPONSE FROCEDURES NOTES: I.  ?HX'R*CkB14A 'B and 2H\'F*CAB32A/E alarms d r ~~i o t v a l i d when t h e Reautcr B u i l d i n y i s i s o l a t e d and normal HVR 1,;

shut do^^.

2. 2H'JK-CAB223 i s used tr:, monitor and t r e n d Reactor B u i l d i r i j V e n t i l a t i o n w h e n t h e k e a c t o r Buildiiiq i s i s o l a t e d and t h e Emergency EWR Recirc U n i t s ( 2 H V R

• U C 4 1 3 A / B j a r e i n s e r v i c e .

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Origin Source Question 74 3 Generic 2.4.14 3.8 N NA LOK

• Grp 10 CFR 55.41(b) 10 LOD (1-5) Reference Documents F NA NMP-2 EOP Basis Document Knowledge of general guidelines for EOP usage.

QUESTION 74 A DIAMOND shaped box in an Emergency Operating Procedure flowchart indicates which one of the following?

A. Hold Point B. Before Step C. Override Step D. Decision Point Correct Answer: D Diamond shaped boxes contain a Decision Point.

Page 86 of 88

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal Plausible Distractors:

A is plausible; Octagonal Shaped Box is a Hold Point.

B is plausible; Pentagonal Shaped Box is a Before Step.

C is plausible; Rounded Rectangular Shaped Box is an Override Step.

Page 87 of 88

Diamonds symbolize sequential decision points leading to alternative branch paths.

Questions inside the d i t ~ ~identify d sdecisions that must be made,

~ ~ ~ the i

Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement 1 IR 1 Origin 1 Source Question 75 3 Generic 2.4.35 I 3.8 I B I 2002 NMP-2 i NRCExam LOK Grp 10 CFR 55.41 (b) 10 LOD (1-5) Reference Documents F NA N2-EOP-6 Rev 8 Knowledge of local auxiliary operator asks during emergency and the resultant operational effects.

QUESTION 75 The plant is at 100% power when a SCRAM occurs, with the following:

0 RPS is TRIPPED ALL BLUE lights are illuminated on the Full Core Display 0 NO Control Rod motion occurred.

0 SLS fails to initiate.

0 Radiation Levels make Reactor Building Elevation 328 inaccessible Which one of the following methods should be used to shut down the reactor?

A. Inject Boron using WCS.

B. Vent the Scram Air Header C. Individually Scram the Control Rods.

D. Inject Boron using the SLS Hydro Pump.

Correct Answer: D Inject Boron using the SLS Hydro PumprCan be achieved without regard to Alternate Rod Insertion procedures and without access to Reactor Building 328.

Plausible Distractors:

A is plausible; but requires Reactor Building 328 access.

B is plausible; would be ineffective with Scram Valves OPEN, indicated by Blue Lights.

C is plausible; would be ineffective with Scram Valves OPEN, indicated by Blue Lights.

Page 88 of 88

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier K/A Number Statement IR Origin Source Question 76 1 295001 AA2.06 3.3 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 1 LCO 3.4.3 Amendment 91 and Basis PROVIDE TS 3.4.1, 3.4.2 & 3.4.3 NO BASIS Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : Nuclear boiler instrumentation QUESTION 76 Reactor Power unexpectedly LOWERED from 85% to 80% power.

The following indications are noted:

0 Loop A Jet Pump Flow LOWERED to 33 Mlbm/hr.

0 Loop B Jet Pump Flow INCREASED to 42 Mlbm/hr.

0 Jet Pump 3 and Jet Pump 4 Flow LOWERED from 3.5 to 0.5 Mlb/hr.

Which one of the following actions is required as a result of these conditions?

A. Declare Jet Pumps 3 and 4 INOPERABLE. Be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> because the ability to reflood the core following a Loss of Coolant Accident is NOT assured.

B. Declare Jet Pumps 3 and 4 INOPERABLE. Be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> because Loop coastdown characteristics assumed in the Loss of Coolant Accident analysis will NOT be preserved.

C. Declare Reactor Recirculation Loop A to be not in operation within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> because the ability to reflood the core following a Loss of Coolant Accident is NOT assured. Plant operation may continue in Single Loop.

D. OPEN Loop A FCV to raise Loop A Jet Pump Flows to re-establish MATCHED Reactor Recirculation Loop Flows within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> because Loop coastdown characteristics assumed in the Loss of Coolant Accident analysis will NOT be met. Plant operation may continue in Two Loop.

Correct Answer: A Jet Pump 3/4 common riser separation is indicated. JP 3 and JP 4 flow deviates from remainder of Loop A by >IO%. With this condition, it is required to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> because the ability to reflood the core following a LOCA is not assured.

Page 1 of 31

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal Plausible Distractors:

B is plausible; Basis is not correct for Jet Pumps INOPERABLE.

C is plausible; Recirculation Loops are mismatched, continued operation in Single Loop is NOT permitted because Jet Pumps 3/4 are INOPERABLE.

D is plausible; Recirculation Loops are mismatched, continued operation in Single Loop is NOT permitted because Jet Pumps 3/4 are INOPERABLE.

Page 2 of 31

Jet Pumps 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Jet Pumps LCO 3.4.3 All j e t pumps sha be P LE.

APPLICABILITY: MODES 1 and 2.

ACT IONS CONDITION I REQUIRED ACTION I COMPLETION TIME A. One or more j e t pumps A.l Be i n MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> i noperabl e.

NMP2 3.4.3-1 Amendment 91

J e t Pumps 3.4.3 SURVEILLANCE REQUIREMENTS SURVE I LLANC E FREQUENCY SR 3 . 4 . 3 . 1 -------__-__--_---_ NOTES-------------------

1. Not required t o be performed until I .

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> a f t e r associated recirculation loop i s in operation.

2. Not required t o be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a f t e r > 25% RTP.

Verify a t least two of the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> c r i t e r i a ( a , b, and c) are s a t i s f i e d f o r each operating recirculation l o o p :

a. J e t pump loop flow versus flow control valve position d i f f e r s by I 10% from established patterns.

b. J e t pump loop flow versus recirculation loop drive flow d i f f e r s by 5 10% from established patterns.

c. Each j e t pump d i f f u s e r t o lower plenum

. differential pressure d i f f e r s by I 20%

from established patterns.

NMP2 3.4.3-2 Amendment 91

J e t Pumps E 3.4.3 BASES APPLICABLE The capability o f reflooding the core t o two-thirds core SAFETY ANALYSES height is dependent upon the structural integrity o f the j e t (continued) pumps. If the structural system, including the beam holding a j e t pump i n place, f a i l s , j e t pump displacement and performance degradation could occur, resulting in an increased flow area through the j e t pump and a l w e r tore flooding elevation. This could adversely affect the water level in the core during the reflood phase o f a LOCA as well as the assumed blowdown f l o w during a LOCA.

J e t pumps s a t i s f y Criterion 3 o f Reference 2 .

The structural failure o f any o f the j e t pumps could cause significant degradation in the a b i l i t y o f the j e t pumps t o allow reflooding t o two thirds core height during a LOCA.

OPERABILITY of a l l j e t pumps i s required t o ensure t h a t operation of the Reactor Recirculation System will be consistent with the assumptions used in t h e 1 icensing basis analysis (Ref. I ) .

APPLICABILITY In MODES 1 and 2 , t h e j e t pumps a r e required t o be OPERABLE since there i s a large amount o f energy i n the reactor core and since the liniiting DBAs are assumed t o occur i n these MODES. T h i s i s consistent with the requirements f o r operation o f the Reactor Recirculation System ( K O 3 . 4 . 1 ) .

I n MODES 3, 4 , and 5 , the Reactor Recirculation System i s not required to be in operation, and when n o t i n operation s u f f i c i e n t flow i s not available to evaluate j e t pump OPEWBILXTY.

ACT IONS A.1 An inoperable j e t pump can increase the blowdown area and reduce the capability t o reflood during a design b a s i s LOCA.

Xf one or more o f the j e t pumps are inoperable, the plant must be brought t o a MODE i n which the LCO does n o t apply.

To achieve t h i s s t a t u s , t h e plant must be brought t o MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Time o f 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> i s reasonable, based on operating experience, t o reach MODE 3 from full power conditions J n an orderly manner and without challenging plant systems.

~-

(continued)

NMP2 E 3.4.3-2 Revision 0

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 77 1 295006 AA2.01 4.6 N NA 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents LCO 3.3.1.1 Action E.l Amendment 92 PROVIDE TS 3.3.1.1 NO BASIS &

TABLE w/ setpoints removed Ability to"detern,,ne and/or interpret the following as they apply to SCRAM : Reactor power QUESTION 77 The plant is operating at 40% power indicated on Average Power Range Monitors.

THREE Turbine Bypass Valves OPENED due to an EHC malfunction.

0 Turbine First Stage Pressure LOWERED to 100 psig.

With these conditions, which one of the following describes the affect on the Turbine Stop Valve / Turbine Control Valve Fast Closure function?

The TSVKCV Fast Closure Scram function is:

A. OPERABLE and will generate a reactor scram if the Turbine Generator TRIPS.

B. INOPERABLE. It is required to LOWER Reactor Power below 30% within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C. INOPERABLE. It is required to LOWER Reactor Power and be in MODE 2 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D. INOPERABLE. It is required to place affected TSVlTCV Closure Channels in TRIP within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Correct Answer: B Turbine First Stage Pressure is used to determine Reactor Power, which enables the TSV/TCV Reactor Scram Function above 30% power. With 40% indicated on APRMs and THREE Turbine Bypass Valves OPEN (15%), First Stage Pressure will correspond to 25% (40%-I 5%) power. This BYPASSES the TSVKCV Fast Closure Scram function when it is required OPERABLE. The LCO is met by LOWERING Power below 30% within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Plausible Distractors:

A is plausible; would be true with Bypass Valves closed.

C is plausible; NOT required to be in MODE 2.

D is plausible; would be true if within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Page 3 of 31

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 (continued)

REQUIREHENTS turbine bypass f l o w can affect t h i s setpoint nonconservatively (MERMAL POWER is derived from turbine first stage pressure, where a f i r s t stage pressure of 136.4 psig is equivalent t o 30% RTP), the main turbine bypass valves must remain closed during an in-service calibration at THERMAL POWER 2 30% RTP t o ensure that the calibration i s v a l i d .

I f any bypass channel setpoint i s nonconservative (4.e., the Functions are bypassed a t 2 30% RTP, e i t h e r due t o open main turbine bypass valve(s) or other reasons), then the affected Turbine S t o p Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions are considered inoperable. Alternatively, the bypass channel can be placed i n the conservative condition (nonbypassf. I f placed in the nonbypass condition, t h i s SR i s met and t h e channel i s considered OPERABLE.

The Frequency o f 24 months i s based on engineering judgment and re1 1abl1 i t y o f the components.

SR 3.3.1.1.16 T h i s SR ensures t h a t scrams i n i t i a t e d from the APRM OPRM-Upscale Function will n o t be inadvertently bypassed when THERMAL POWER i s t 30% RTP and recirculation drive flow is c 60% rated recirculation drive flow.

f f any bypass channel setpoint Is nonconservative ( i . e - , L l t r Function is bypassed a t 2 30% RTP and < 60% r a t e d recirculation drive flow), t h e n the affected channel i s considered inoperable.

The Frequency o f 24 months i s based on Ref. 15.

SR 3.3.1.1.17 I T h i s SR ensures t h a t the individual channel response times a r e less t h a n or equal t o the maximum values assumed i n the accident analysts. The RPS RESPONSE TIME acceptance c r i t e r i a are included i n Reference 12.

I cant i nued WP2 B 3.3.1.1-34 Revision $4, 1

RPS Instrumentation 3.3.1.1 ACT IONS I cont i nued1 CONDITION REQUlRED ACTION C. One or more Functions C.l Restore RPS trip 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> w i t h RPS trip capabil 5 ty.

capabll f t y not mai ntai ned .

D. Required Action and D.1 Enter the Condition Immed-1ately associated Completion referenced i n Time o f Condition A, Table 3.3.1.1-1 for the B, or C not met. channel <,

E. As required by E. 1 Reduce THERMAL POWER t o 0 hours Required Action 0.1 < 30% RTP, and referenced i n Table 3.3.1.1-1.

E. As required by F.l Initiate alternate 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Required Action D.1 method to detect and and referenced 5n suppress thernal-Table 3.3.1.1-1. hydraulic instability oscilf ations.

f .2 Restore required channel 120 days to OPfR&BLE status.

6. Required Action and 6.1 3e i n WDE 2.

associ ated Cornpl et ion T h e o f Condition F not met.

-OR As required by Required Action 0.1 and referenced i n Table 3.3.1.1-1.

(continued)

NMP2 3.3.1.1-2 Amendment $I 92

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier K/A Number Statement IR Origin Source Question 78 1 295019 AA2.01 3.6 B 2003 Fitzpatrick NRC Exam LOK Grp 10 CFR 55.43(b) 5 ~ LQD (1-5) Reference Documents H 1 N2-SOP-08 Rev 3 N2-SOP-19 Rev 3 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR : Instrument air system pressure QUESTION 78 With the plant operating at full power, the following alarms and indications occur:

0 851229, INSTR AIR SYSTEM TROUBLE alarms.

0 851218, INST AIR RCVR TK 3 PRESS LOW alarms.

0 603306, CRD SCRAM VALVE PILOT AIR HDR PRESS HIGH/LOW alarms.

0 603443, CONTROL ROD DRIFT alarms.

0 TWO Control Rod DRIFT Lights are LIT.

603139, REACTOR WATER LEVEL HIGH/LOW alarms.

RPV Water Level is 188 inches.

Which one of the following indicates the Scram Pilot Air Header Pressure, based on these indications AND the appropriate procedure to mitigate these conditions?

Scram Pilot Air Pressure Procedure to Mitigate A. BELOW 65 psig N2-SOP-30, Control Rod Drive Failures B. BELOW 65 psig N2-SOP-19, Loss of Instrument Air C. ABOVE 75 psig N2-SOP-30, Control Rod Drive Failures D. ABOVE 75 psig N2-SOP-19, Loss of Instrument Air Correct Answer: B 603306 Control Rod Drift indicates IA Pressure is below 65 psig. this requires entry into N2-SOP-19, Loss of Instrument Air.

TWO Rod Drifts requires an immediate scram per N2-SOP-08 Unplanned Power Changes.

Plausible Distractors:

A is plausible; does not address the loss of Instrument Air condition.

C is plausible; plant status indicates Instrument Air Header Pressure is below 65 psig , does not address loss of Instrument Air condition.

D is plausible; plant status indicates Instrument Air Header Pressure is below 65 psig Page 4 of 31

The rate of wr3urrence of f5I~o~~ing Automatic Responses will be dependant on the rete sf the bss pressure.

of I ~ s ~ r ~atrr n #~~

100 p i g J A ~ ~ u n ~ i851229 a t u ~ IMSTR AIR SYSTEM TROUBLE

~ ~ m p u Point

~ e r IASPCOG INSTR & SYCE AIR PRESS Lag Air Compressor auto stads 90 psig 3, i ~ t ~ r INST AIR RCVR TK 3 PRESS LOW A n n ~ n ~ 851218.

~ e r1ASPC39 INST AIR RCVR TK 3 PR LO

~ Q ~ p u Point 85 psig

~ ~ ~ u n ~ i a 5t IHST ~ r~ AIR RCVR 2 ~TK 2 PRESS

~ LOW Computer Point IASPCOS INST AIR RCVR TK 2 FIR I S 65 psig 3, ti0 pslg 4 Feeelwater Pump $ . ~ l i n i ~Flow u ~ iValues fail fult open results in Loss of air to Scram Air Header res@lf$in i ~ d i v i Control

~ u ~ ~Rod8 ~ ~ into

~ core

~ r i i ~ ~

~ n ~ u n ~ 603443.

i a t ~ r CONTROL ROD DRIFT 70-0 psig J Various arr uperakd components ivill fail 8s l~struii~ent Air pressure decays See s:ep 5 3 for a more complete 1st of componeats and their possible affects Page 2 N2-SOP- 19 Rev 03

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 79 1 295021 2.4.9 4.2 B NMP-2 Bank SYSID 17829 LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents F 1 LCO 3.9.8. Amendment 1 PROVIDE TS 3.9.8 and 3.9.9 NO BASIS Loss of Shutdown Cooling - Knowledge of low power/shutdown implications in accident (e.g.,

loss of coolant accident or loss of residual heat removal) mitigation strategies.

QUESTION 79 The plant is in MODE 5, with the following:

0 Fuel Movements are in progress.

0 An equipment failure results in the inability to establish Shutdown Cooling flow.

0 Shutdown Cooling Flow has been secured for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Which one of the following ACTIONS is required by plant Technical Specifications?

A. SUSPEND Core Alterations immediately.

6. RESTORE Secondary Containment Integrity immediately.

C. START the Standby Gas Treatment system within one hour.

D. VERIFY an alternate method of decay heat removal available within one hour.

I --

Correct Answer: D is correct - Flow can only be secured for up to two hours, then must be declared inoperable/unavailable per TS 3.9.8 action A (high level.).

Plausible Distractors:

A is plausible; this is follow up action if an alternate method of decay heat removal is determined to be not available within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

B is plausible; actions would be for TS 3.9.9 (low level - Mode 5 not moving fuel).

C is plausible; actions would be for TS 3.9.9 (low level - Mode 5 not moving fuel).

Page 5 of 31

RXR--Hlgh Water Level 3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR)-High Water Level APPLICABILITY: MODE 5 w i t h irradiated Fuel in the reactor pressure vessel (RPV) and with the water level 2 22 f t 3 Inches above the top o f the RPY flange.

ACTIONS e cooling subsystem inoperab?e .

method o f decay heat

8. Required Action and B.1 Suspend loading lmnedi ateiy associated Compl et ion irradiated fuel Time o f Condition A assemblies into the not %et. RPK.

1 (continued)

NMP2 3.9.8-1 Amendment 91

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 80 1 295025 2.1.20 4.6 B 2003 LaSalle NRC Exam LOK Grp , 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 1 ' N2-EOP-C5 Rev 10 High Reactor Pressure - Conduct of Operations: Ability to interpret and execute procedure steps.

QUESTION 80 N2-EOP-C5, Failure To Scram is being executed following a condenser boot rupture, with plant conditions as follows:

RPV Water Level is 150 inches, LOWERING.

0 Reactor Pressure is 1080 psig, STABLE.

0 TWO Safety Relief Valves are OPEN.

0 Suppression Pool Temperature is 108°F.

Which one of the following is the HIGHEST RPV Water Level that may be MAINTAINED?

A. +202 inches

8. +I50 inches C. +IO0 inches D. -14 inches Correct Answer: C interpreting 1080 psig STABLE Reactor Pressure with TWO SRVs open, indicates that Reactor Power is at -10% (or at least above 4%). Per N2-EOP-C5, with Power

>4%, and RPV level above 100 inches, it is required to Terminate and Prevent Injection and let Level LOWER to at least +IO0 inches.

Plausible Distractors:

A is plausible; would be true if executing N2-EOP-RPV.

B is plausible; it's the CURRENT RPV Water Level in this condition.

D is plausible; is the LOWEST allowable RPV Water Level in this condition and the highest if SPT was above 110°F Page 6 of 31

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier K/A Number Statement IR Origin Source Question 81 1 295026 EA2.01 4.2 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 1 N2-OP-34 Rev 8 Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool water temperature QUESTION 81 The plant is operating at full power, with the following:

0 ONE Safety Relief Valve OPENED.

0 BOTH Loops of RHR are started in Suppression Pool Cooling mode.

Which one of the following describes the affect that Suppression Pool Cooling operation will have on Suppression Pool Water Temperature (SPT)?

A. Doesnt maintain SPT above the lowest Technical Specification LCO value.

B. Maintains SPT at a nearly constant value until the Safety Relief Valve is successfully CLOSED.

C. Maintains SPT below the Technical Specification LCO value for performing testing that adds heat to the Suppression Pool.

D. Doesnt maintain SPT below the Technical Specification LCO value which requires a Reactor Shutdown to be performed.

Correct Answer: D SRV -900,000 Ibm/hr Rated Steam Flow or -6% Rated Thermal Power (6% x 3467 Mwt = 208 Mwt = 709 x I O 6 BTU/hr)

RHR Hx = 150x I O 6 BTU/hr x 2 = 300x I O 6 BTU/hr. ONE OPEN SRV substantially exceeds the heat capacity of all Suppression Pool Cooling Plausible Distractors:

A is plausible; would be true if RHR capacity exceeded SRV heat addition.

B is plausible; would be true if RHR capacity matched SRV heat addition.

C is plausible; TS LCO 3.6.2.1.b specifies a higher allowable Suppression Pool Temperature when testing is in progress.

Page 7 of 31

2.1 Safety Function T h e s a f e t y function is an a u t o m a t i c pressure-induced f u n c t i o n isovercome spring tension) t o l i m t reactox pressure t o acceptable levelr;. T h e safoty/relief v a l v e s self -opening c a p a c i t y is s i z e d on t h e basis of a f u l l t u r b i n e t r l p without bypass c a p a c i t y s t a r t i n g from 1135%of r a t e d steam flow. Spring r jet pressures a r e as follows:

PSV' s Sprinq Set Pressui-e

1. 12e and 1 3 3 1165 p s i g 1175 p i g 1185 P P ~ J 1195 p i g

1. 127, 129, 1 3 4 & 1 3 7 1205 p i g T h e P S V ' s reseat from d s a f e t y itlode l i pressure.

Page In_ EJ2 -0P- 3 4 Rev 08

4. Discharge Piping

a. The pump discharge piping for RHS Loop A and B is similar in function and design so they will be described together. RHS Loop C is described separately.

b. RHS Loop A and 5

1) The pump discharge piping is provided with a minimum flow bypass line, which directs flow back to the Suppression Pool via the suppression pool cooling piping. The minimum flow valves (ZRHS"MOV4A & B) open automatically when flow is less than 1400 gpm after an 8 second time delay to provide adequate cooling when pumping against a shutoff head.

a) When the system is in standby the minimum flow valve is open.

2) Next in the discharge piping is the RHS heat exchanger, a single pass shell and U-tube heat exchanger, using SWP on cooling. The heat exchanger is designed to BTU/Hr.

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question ,

82 1 700000 2.4.4 4.7 N NA LOK Grp 10 CFR 55.43(b) 5 ' LOD (1-5) ' Reference Documents F 1 N2-SOP-70 Rev 1 Generator Voltage and Electric Grid Disturbances - Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal ooeratina orocedures.

QUESTION 82 The plant operating at full power, with the following:

0 System Power Control notifies the Control Room of unstable grid conditions Loss of Offsite Power is IMMINENT.

Grid Voltage and Frequency are LOWERING.

Which one of the following actions is required by the SOP'S?

A. MAINTAIN ALL Emergency Diesel Generators in STANDBY B. START Division 1 AND Division 2 Emergency Diesel Generators and operate them LOADED.

C. START Division 1 AND Division 2 Emergency Diesel Generators and operate them UNLOADED.

D. START Division 1, Division 2, AND Division 3 Emergency Diesel Generators and operate them UNLOADED.

Correct Answer: C When System Power Control notifies the Control Room of unstable Grid Conditions with an imminent Loss of Offsite Power, N2-SOP-70, Major Grid Disturbances is entered. N2-SOP-70 requires STARTING and operating (only) Division 1 AND Division 2 EDGs UNLOADED.

Plausible Distractors:

A is plausible; EDGs are normally in STANDBY for unanticipated Loss of Offsite Power Events.

B is plausible; N2-SOP-70 states that damage to buses and equipment may result from parallel operation to a degrading grid.

D is plausible; N2-SOP-70 does NOT required starting Division 3 EDG.

Page 8 of 31

m.E&y into Iha SOP may require clasificatio? of an emergency per EPIP-EPP.02 ( EAL 6 1)

I in the loss of the program(state Estimator) is out of SBN~CRAND h e 5 A N W R Lire 6 vallage IS less that I 1 0 KV or pBS2,

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 83 1 295002 2.4.31 4.1 B NMP-2 Bank SYSID 6939 LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents F i 2 N2-SOP-9 Rev 2 Loss of Main Condenser Vac - Emergency Procedures / Plan: Knowledge of annunciator alarms, indications, or response procedures.

QUESTION 83 A plant startup is in progress with the following conditions:

Generator Load is 300 MWE.

851358, TURBINE CNSR A/B/C VACUUM LOW alarms.

Main Condenser Vacuum LOWERS and STABILIZES at 24 inches Hg, Which one of the following actions is required?

A. TRIP the Main Turbine per N2-SOP-21, Turbine Trip.

B. LOWER Power per N2-SOP-101D, Rapid Power Reduction.

C. START a Mechanical Vacuum Pump per N2-0P-9, Condenser Air Removal.

D. COMMENCE a normal Turbine Shutdown per, N2-OP-21 Main Turbine System.

Correct Answer: A With Generator Load below 363 Mwe and Condenser Vacuum below 24.6 inches Hgvac,it is required to TRIP the Main Turbine per N2-SOP-21, Turbine Trip.

Plausible Distractors:

B is plausible; would be true if Generator Load exceeded 363 Mwe.

C is plausible; would be true if Reactor Power were below 5%.

D is plausible; would be true if TRIP criteria were not exceeded.

Page 9 of 31

N2-SOP-09 LOSS OF VACUUM 1

IF

+-)

[ EVENT DESCRIPTION See Section 1.O.

Vacuum is (24.6 Hg AND turbine is loaded to

\

THEN (30% (363 MWe), 1 TRIP the turbine per N2-SOP-21 I

c SCRAM the Reactor per N2-SOP-101C Pressure 4 9 psia AND Reactor is still critical c

As required, lower power per N2-SOP-IOID to stabilize vacuum.

1 Verify proper operation of:

0 SJAE per N2-OP-9 CI Off-gas per N2-OP-42 0 Circ Water per N2-OP-1OA 1

IF THEN

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 84 1 295009 AA2.02 3.7 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 2 ARP 603139 Rev 0 Low Reactor Water Level - Ability to determine andlor interpret the following as they apply to LOW REACTOR WATER LEVEL : Steam flowlfeed flow mismatch QUESTION 84 The plant is operating at full power. ONE Feedwater Flow input to the Feedwater Level Control System fails LOW.

With NO operator actions, which one of the following will result FIRST, and what actions are required?

A. BOTH Reactor Recirculation Pumps will DOWNSHIFT, and actions will be required per N2-SOP-08, Unplanned Power Change.

B. RPV Water Level will RISE and stabilize below Level 8, and actions will be required per N2-SOP-06, Feedwater Failures.

C. RPV Water Level will RISE until reaching Level 8, and actions will be required per N2-SOP-101C, Reactor Scram.

D. RPV Water Level will LOWER and until reaching Level 3, and actions will be required per N2-EOP-RPV, RPV Control.

Correct Answer: C full power implies Three Element Auto control. When ONE Feedwater Flow Input FAILS LOW, FLCS will increase Feedwater Flow until RPV Water Level 8 is reached. L8 cause a Main Turbine and RFP Trips, which produces a scram requiring N2-SOP-101C entry.

Plausible Distractors:

A is plausible; would be true if BOTH Feedwater Flow Inputs failed LOW. ~ 2 2 . 4 %for 15 seconds downshifts both RR Pumps.

B is plausible; would be true for a single Steam Line Flow Input failed HIGH.

D is plausible; would be true for the selected RPV Water Level Channel failed HIGH.

Page 10 of 31

NOTES feedwater flow signal, the feedwater control system circuits will see low feed flow which then compared t o the actual steam flow signal will result in a signat which will open the flow control valves until level reaches an equilibrium point. For the loss of one of the flow signals a t rated power, reactor level will increase approximately 25" before this equilibrium point is reached. This results tn a Level 8 trip.

c. Loss of Reactor Level Signal - If the reactor level signal input t o the circuit is lost, the crrcuit sees a low level t o which it responds by increasing feed flow. The feed flow will be increased until feed flow is at its maximum value or the signal balances the loss of level signal. Since feed flow is greater than steam flow, the reactor water level will increase to the reactor high level trip as sensed by the two remaining narrow range level transmitters. On 18 High Rx Water Level the LV-137 and LV55AfB valves will transfer to manual. (The LV-IOA, B and C valves do not transfer to manual on L8.) Additionally, when 2 of 3 narrow range transmitters exceed Level 8 a turbine trip and reactor feedpump trip will result.

Instructor Guide jwzioiissonrcoi; 115 of 136 Printed: 05/10/2007

ATTACHMENT 11 (Cont1 2CEC*FNL603 S E R I E S I t 0 ALARM RESPONSE PROCEDURES Reflash: NO ZCEC*PLJLGG 603139 REA2T OR W-TEK LEVEL HIGH,' LOX 1'39 CornputY r Point. Compute r P r i n t .011t Soul-ce Setpoint F'dS LC Li I R.EACTOR XATER LVL 2 ISC-LSliS3 5 HI: 187.3"<L71 HI/Lt3 LO: 1 7 8 . 3 " 1 L 4 )

Automatic Response I F wat-er l e v e l i n c i r e a s e s to l a v e 1 8 ( 2 0 2 . 3 ' ' t h e r e a c t o r fee? pumps AND irtaiii tuxbiiia will t r i p . I F w a t e r l e v e l flesr+.as+sto l e v e l 2 (15:>.?,)a React.or SCmM w i l l o c c u r .

1. I F KPV pressure 2 OR = 9 0 0 p:;ig, enter I.K!-.SClP-Oi;, Feedwater Failures.

2. IF R.EW prtssura c 900 p s i g , ~ ) e r f o n nt.he followiiicj ZIB xequired:

d. HIGH LEVEL

1) Reduce feed rat-e t o the RPV by the fsllo!*.ing as reyuired :

• Closing Feedwarer Level Control ira1va.r FXS-LVlOs, L V l s 5 ~OR CNM-L:'13'?.

• R a i s e reject fl.ow r a t e by t . h r - o t t l i n y open A C . ~ FV135.

• Reduce CRD flow by t l i r o t t l i n g c l o s e d RDS-FC107.

• Shutdown Feedwater pumps, Feedwater Eoost,er purnyjs OR Condensate pumps .

C l o s i n g 2FWS-MCW2ls OR 2FWS-bIOV47s CR 2CNM-ldOV84s.

b. LOX LEV'EL

1) Raise feed r a t e to thi. RPV by ANS of t h e f o l l o w i n g :

• Reduce riiject f l o w b'f c l o s i n g Tr:tS*FV135.

• Raise C:RD i n j e c . t i o n f l o w t o approximately G 3 q m .

• Restore f e e d AND cwndeneate to t h e RP': per N2-OP-2.

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier K/A Number Statement IR Origin Source Question 85 1 500000 2.4.6 4.7 B 2002 NMP-2 NRC Exam LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 2 N2-EOP-PCH Rev 2 Provided: N2-EOP-PCH Rev 2 High CTMT Hydrogen Concentration - Knowledge of EOP mitigation strategies.

QUESTION 85 A LOCA is in progress AND Hydrogen (H2) has been released into the Primary Containment. The following conditions exist:

Drywell H2 is 6.5%.

Drywell O2 is 5.0%.

Suppression Chamber H2 is 6.0%.

0 Suppression Chamber O2 is 5.0%.

0 Suppression Pool Level is 205 feet.

0 Drywell Spray Initiation Limit is satisfied.

Which one of the following identifies the requirement for Drywell AND Suppression Chamber Spray operation?

Drywell Spray Suppression Chamber Spray A. NOT REQUIRED NOT REQUIRED B. REQUIRED NOT REQUIRED C. NOT REQUIRED REQUIRED D. REQUIRED REQUIRED Correct Answer: D DW is at deflagration and SC is deinerted and deflagration Hydrogen exists in DW.

Plausible Distractors:

A and C are plausible; meet conditions to spray DW per N2-EOP-PCH.

B is plausible; meet conditions to spray SC per N2-EOP-PCH.

Page 11 of 31

1. Stop the r e ~ ~ ~ j nfOP-62 e r § Section G.1.0).

2. ~ L ~ W D ~ W ~ :

I . Enter EOP-RPV wtiiie c~ntjriu~ng tiere. JI

2. Enter EUP-C2 while continuing here, J, 2

3. F ...,+.. . s ~ p ~ r ~ spool n level is below Et. 21 7 ft, s i ~water HEN ..operate suppression chamber sprays (EOP-6 A 8 22f IC. CAUTION: Operating RHR with suppression pool water level below El. 195 ft may cause system damage.

• r Operate sprays wen tf core cooling will be lost IC OK to defeat suppresston chamber spray iiiteriocks (EOP-6 Att 22)

(t OK to u s e external spray sources (EOP-6 A& 5 and 6) if yau can restore and maintain suppression chamber pressure below the Pnmary Containment Pressure Limit (Fig 0).

4. IF ........suppression pool water levef is befow El. 217 Zf THEN ..purge the with air OR nitrogen (EOP-6 Att 25).

m- Use whichever method will reduce hydrogen below 6% QE oxygen betow 5% faster u OK to defeat &isolations

• r OK to exceed release rate limits

WELL 4

f Stop recmbiners taking suctim on the drywll (OP-62 Secttofl G 1 0 )

2 BLOWDOWN 1 Enter Em-RPVwhile coniinuing here 4 2 En%r EOP+CZMil@ ~ r ~ ~ i n here uiri~ &

J Purge the dryv.ell with air nitrogen at the maxrmun-1rate (EOF-6Att 25)

• r U m flhichevwr method ~ 1 r e1 ~ u t ohydrgen bdow 6% pBcrxvgcr bekrv, 5% Firstor R OK !O defeat ISoldttDBs

• r OK to ~ X C W GreJIe8w a t e limrts c OK tu pLryo throut;h tk w(yxrua51on pod te wdcce reIur1%erate Ljul g&,.d 1 5 uppression pod weer level is &&,Y El 217 ft AND Tho s~rpprosslonchamber can be verilod you are inside the Drywell Spray Initratran Limit (Fig KI.

suppression pool water Iw& is &&a El 21 7 ft spray the drywdl 1 Trip all recirc pumps 2 Trip all dry,@ll unit coolc~s 3 Operate d r y w l l sprays (EW-6At! 221

• I CAUTION Operatingt RHR with suppression pool water level betow El. I 9 5 R ma) cavse system damage.

R Opurstu sprdtys& core mo rg v.1I1 be lost IC OK to &eat Liryv;*ellspfty crttlrClcks (EweAtt 22)

.c OK to m e external spray sourem (EOP6 Att 3 avd 6)if you c8r rwlixe ard nmrtaln wpprc!;t,ar chamber pressure bo1o.v the Primdry Corlacimrorek i~rew.wbmit (Fig 0)

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier K/A Number Statement IR Origin Source Question 86 2 215003 2.2.40 4.7 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 1 LCO 3.3.1.1 Amendment 92 Provided: TS 3.3.1.1 NO BASIS Intermediate Range Monitor - Equipment Control: Ability to apply Technical Specifications for a system.

QUESTION 86 A reactor startup is in progress with Intermediate Range Monitor (IRM) Channel A INOPERABLE and BYPASSED, when the following occurs:

IRM Channel D indicates upscale at 125/125, irrespective of Range Switch position.

0 IRM Channels B, C, E, F, G, and H indicate 32/40 on Range 7.

0 ALL Average Power Range Monitors (APRMs) are DOWNSCALE.

Which one of the following actions is to be directed?

A. PLACE IRM Channel D in a TRIPPED condition and continue the Reactor Startup.

B. SHUTDOWN per N2-0P-I01C, Plant Shutdown; because REQUIRED Intermediate Range Monitors are INOPERABLE.

C. BYPASS IRM Channel D using the joystick per N2-OP-92, Neutron Monitoring; RESET the Half Scram, and CONTINUE the Reactor Startup.

D. BYPASS IRM Channel D by placing the Reactor Mode Switch in RUN per N2-OP-IOIA, Plant Startup; RESET the Half Scram, and CONTINUE the Reactor Startup.

Correct Answer: C In MODE 2, there are 3 REQUIRED Channels per Trip System. With IRM A and D INOP, the LCO is satisfied. It is permitted to BYPASS IRM D, RESET the Half Scram and continue the startup.

Plausible Distractors:

A is plausible; IRM is in a TRIPPED condition, needs to be BYPASSED to continue startup.

B is plausible; would be true with < 3 IRMs per TRIP SYSTEM OPERABLE.

D is plausible; Reactor Power is too low to place the Reactor Mode Switch in RUN.

Page 12 of 31

RPS Instrumentation 3.3.1.1 table 3.3.1.%-9 (past 1 of 3)

Rwetor Protoctfor) Systaa fwtrurentation COWDlTfONS APPLICABLE REPUIaEO REFERENCE0 WODES OR OTHER CHADIUELS FRMI SPEC1F I D PER TRlP REPUtRED SURMILWCE ALLOUA8I.E FUllCf IOU eOuDfTlDIiS SYSTEM ACTiDl D.1 REWIREMENTS V A M H SR 3.3.1.1.1 I 122f325 I OR 3.3.1.1.4 divieims SR 3.3.1.1.5 of full SR 3.3.1.1.6 scale S t 3.3.1.1-13 SR 3.3.1.1.14 f SR 3.3.1.1.1 I la125 I SR 3.3.1.1.4 divisiow SR 3.3.1.1.13 of f u l l SR 3.3.1.1.14 seal@

b. Imp n SR 3.3.1.1.4 nh i SR 3.2.1.1.14 I SR 3.3.1.1.6 #A I SR 3.3.1.1.14 H SR 3.3.1.1.2 1243% RTP I SR 3.3.1.1.5 SR 3.3.1.1.7 SR 3.3.1.1.10 SR 3.3.1-1.13 G SR 3.3.1.1.2 '< .5& + I SR 3,J.l.l.S bZX RTP afid SR 3.3.1.1.7 i 1152%

SR 3.3,1.1,10 RTPfb)

SR 3.3.1.1.13 c.

FExed #mtron Ftu~U p a ~ l ~

C SR 3.3.1.7.2 5R 3.3.1.1.3 SR 3.3.1.1.7 d 12H RTP 1 SR 3.3.1.1.10 SR 3.3.1.1.13

d. fnap x SR 3.3.1.1.7 YA I SR 3-3.1.1.10

e. OPRW-Upscale F SR 3.3.1.1.2 AS SR 3.3.1.1.7 dpeciffaf SR 3.3.1.1.10 in the IX#R SR 3.3.1.1.13 SR 3.3.1.1.14 H SU 3.3.1.1.2 Nh 1 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3-3.1.1.17 (cant inued) i (a) Uith any cmtrol rod u i t h d r m froa a core cell cwtsiniw me or mDrr fuek asscrablics, (b} -

kltouabte Valw lo .WW 5%) + 62% RTP d m reset for s i n g t r Loop operaticm p r LM 3.4.1,

• Retirwlratim Limps @mating.#

MP2 3.3.1.1-8

NOTES c) The NMS scram logic trip contacts for IRM and APRM/OPRM can be bypassed by selector switches located in t h e main control room. Bypassing either an IRM o r an APRM/OPRM channel will not inhibit the NMS from providing protective action where required.

(1) IRM channels A, C, E, and G bypasses are controlled by one selector switch, and channels B, D, F, and H bypasses are controlled by a second selector switch. Each selector switch will bypass only one IRM channel a t any time.

(2) A single selector switch allows bypass of one of the four APRM/oscillation power range monitor (OPRM) channels. None of t h e four two-out-of-four voter channels may be bypassed.

2) Reactor Vessel High Pressure a) A Reactor vessel pressure rise during Reactor operation compresses the steam voids and results in higher reactivity; this causes higher core Instructor Guide ( ~ 2 1 0 1 2 1 2 0 0 0 ~ 0 1 ) 167 of 274 Printed: 04/04/2007

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 87 2 209001 A2.07 3.6 B Monticello 2003 NRC Exam LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 1 TS 3.5.1.A Provided: TS 3.5.1 NO BASIS Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Core Spray Line Break QUESTION 87 The plant is operating at full power, with the following:

0 601419, LPCS LINE BREAK alarms.

0 LPCS Line Break h o p Status Light is LIT.

0 Report received that the D/P indicating switch reading is oscillating at around +4.0 psig.

0 NO other annunciators alarm.

Which one of the following describes the LOCATION of this piping break AND the Technical Specification implication of this failure?

A. LPCS piping BETWEEN the Reactor Pressure Vessel wall and the Core Shroud.

AND Enter a 7 day LCO and place the LPCS Pump in PULL TO LOCK.

B. LPCS piping INSIDE the Core Shroud.

AND Place the LPCS Pump in PULL TO LOCK and reduce Reactor Coolant Temperature to < 212" F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. LPCS piping INSIDE the Core Shroud.

AND Enter a 7 day LCO, place the LPCS Pump in PULL TO LOCK, and deactivate the Injection Valve.

D. LPCS piping BETWEEN the Reactor Pressure Vessel wall and the Core Shroud.

AND Enter a 14 day LCO, place the LPCS Pump in PULL TO LOCK, and deactivate the Injection Valve.

Correct Answer: A Line Break annunciator detects a line break BETWEEN the Reactor Pressure Vessel and the Core Shroud. With spray function not assured, a 7 day LCO is required by Page 13 of 31

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal Plausible Distractors:

B is plausible; Break INSIDE Core Shroud will NOT be annunciated. Not required to be < 212°F in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C is plausible; Break INSIDE Core Shroud will NOT be annunciated.

D is plausible; 14 day LCO is applicable for HIGH Pressure Core Spray function.

Page 14 of 31

ATTACHMENT 4 ( C o n t )

2CECfPML601 S E R I E S 4 0 0 ALkR.M RESPGNSE PROCEDURES Rt:flash: NO LPCS LINE BREAK 419 (IOMPUTEF. POINT COMPUTER PRINTOUT SOURCE S ETPO I NT CSLBC13 LPCS LIME BREAK 2RHS*PT)TlHk ., 3 . 8 p s i d Automat.i c R e sponse None Operator A c t ions

1. U t i l i z e LPCI R!C I F P.EY injeztAon i s r e q u i r e d , p e r W - E O F s

? Refer t.o Technical S p e c i f i c a t . i c m s .

P o s s i b l e Causes E: Coiidit.ions s u c h as plant. s h u C d o w r l or scram can p l a c e l i n e break

!C15) i n s t r i i m e n t a t i o n o u t s i d e i t s c a l i b r a t i o n cnndit.ioiis. T h i s alarm niay be received i f thnss c:ondi t i o n s ocicui- .

Pipirig break has occurred.

-+*

1nst.rumentation out of c a l i b r a t icn.

References 0 N2-OF- 32 0 Techni cd 1 Spec i f i car i oils 45482,46536 Paye 251 N2 -ARP- 0 1 R e v 00

G SHUTDOWN 1.0. Shntdc.nrn To StaiidLy

\ CCu*f 1I

[EOP) 1.1 Deyxesb: L P C I A / L P C S RESET l u s h b u t t o n AND v e i i f y white s e a l - i n (SOP, light is ofr at 2CECI*PNLGOI.

1.2 Stor1 C$L*P1, PMP 1.

1.3 Close <ZSL*MOVlb-l, FIIP 1 I N J E C T I D N VL'J.

1. 4 Vel i r y CSL*M~l'll(17, PMP 1 l\llNIMLlM FLOW VL;7 cjpslis.

1.5 Restc)re L P C I A / L P C S MATJUAL I N I T I A T I O N FLlshbuttot1 c o l l a l to DISARM a s r e q u i r e d .

1.G V e r i f y Standhy C o n d i t i o n S t a t u s Checks p e r N i a 3 e c T i o n F . 1.

1.7 N o t i f y SSS that 2I:SL*F'J114 is c l o s e d and t h e o p r i b l l i t y conc.ern per DER 2 - 9 8 - 0 5 5 7 110 l ~ n y e re x i s t s .

2.0 S1iutdc.wii T o I n n p e r a b l p 2 .1 Depress LPCS WTUALLY GUT OF SVCE pushbutton a t 2CEC*PETL601 2.2 P1dc.t. C S L

• F 1 , FMP 1 C o n t r o l S w i t c h i n PTJLL-TCJ-LWK.

NOTE: D o NUT zontiiiup i f C S L

• P l , FI4P 1 is to L e ncaintdinecl dvailaLle d u r i n g a u n i t outage

2. 3 A t 2EIIS*STr:G101-7, rack out 2 C S L

• P l LPCS Punq~ Motor C i r c u i t BI.eakF'r.

2.4 Verity CSL*M3','104, PMP 1 TPIJECTION 7 L 7 -*lased a t .?CEC+PNLGOl.

2. 5 A t 2EHS*MCClOZC-l5R, p l a c e 2CSL*M3'$104 LPCS Irije~!t1 ~ 1 1valve Motor Circuit B r s a k e r t o OFF.

2.6 I F rwrfcrming F i l l .UJD Vent, clepressuriziny LPCS i s NOT 1q u i r e d .

2.7 I F LPCS i s t i i be dep ULI Z F J , parform cm? of the following:

2.7.1 C l o s , : ~;CSL*V17, 2C:;L*Pt Piiiup Gischdryi I s c l o r ,

NOTE: S h u t t i l l y down 2 C S L

• P 2 LPCS/'RWR A Watel L e g Pun~yi ,1111 d a p k s s s u r i z e RHR A 2.7 2 Shutdown CSL*F:, LITS/RHR A WTR LE2 PMP at 2CEC*PNL601.

Page 12 N 2 -OP- 32 Rev 06

ECCS--Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION CQOtlNG (RCIC) SYSTEM 3.5.1 ECCS -OOpef&itIg LCO 3.5.1 Each ECCS injection/spray subsystem and Uw AutomatlC Depressurization system (ADS) W o n of six safety/reliif valves shall be OPERABLE.

3, except ADS valves are not required to he OPERABLE with mctot stem dome pressure i 150 pig.

ACTIONS CONDITION COMPLETIONTIME A. Om,low pressure ECCS 7 days injectionfspray ECCS injectiotlrspray subsystem inoperable. subsystem to OPERAEL status.

6. illgh Pressure Core B.1 Venfyby Spray QIPCS) System administrative means inoperable. RCJC System is OPERABLE when RCIC is required ta be OPERABLE.

AND 8.2 Restore HPCS System $4 days to OPERABLE status.

{continued)

NMP2 3.5.1 -1 Amendment 91,109

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier K/A Number Statement IR Origin Source Question 88 2 21 1000 2.2.25 4.2 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 1 TS 3.1.7, SRs 3.1.7.2 and 3.1.7.3 PROVIDE FIGURE TS 3.1.7, SRs 3.1.7.2 and 3.1.7.3 NO BASIS Standby Liquid Control - Equipment Control: Knowledqe of bases in technical specifications for limiting conditions for operations and safety limits:

QUESTION 88 During operation at 100% power an operator reports the heat tracing on the suction side of the Standby Liquid Control (SLS) pumps is damaged and inoperable. The following conditions exist:

0 Liquid Poison Tank Concentration is 14%.

0 Liquid Poison Tank Volume 4600 gallons.

0 Reactor Building Ambient Temperature at SLS Pumps is 72°F.

Which one of the following describes the condition of the SLS system and required actions per Technical Specifications?

A. OPERABLE as long as Reactor Building Temperature is above 7OoF,which is based on preventing Boron from precipitating out of solution inside system components.

B. INOPERABLE because the Boron Solution Concentration is BELOW the minimum. A 7 day LCO is required, which is based on injecting 780 ppm of Boron solution into the reactor core.

C. INOPERABLE because the Boron Solution Concentration is BELOW the minimum. An 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> LCO is required, which is based on injecting 780 ppm of Boron solution into the reactor core.

D. INOPERABLE because the Boron Solution Temperature is BELOW the minimum. An 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> LCO is required, which is based on preventing Boron from precipitating out of solution inside system components.

Correct Answer: A SR 3.1.7.2 (verify solution and piping temps >70°F) WILL be met with RB Temperature at 72°F in the vicinity of SLS components. Heat Trace functionality is not specified as a Surveillance Requirement.

Page 15 of 31

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal Plausible Distractors:

B is plausible; with ONE subsystem INOPERABLE, a 7 day LCO would be applicable.

C is plausible; with BOTH subsystems INOPERABLE, an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> LCO would be applicable.

D is plausible; would be true with Level / Concentration outside of LCO values.

Page 16 of 31

SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3 . 1 . 7 Two SLC subsystems shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A. 1 Restore SLC subsystem 7 days inoperable. to OPERABLE status.

B. Two SLC subsystems B. 1 Restore one SLC 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable. subsystem to OPERABLE status.

C. Required Action and c.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume o f sodium 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> pentaborate solution is within the limits o f Figure 3.1.7-1.

NMP2 3.1.7-1 Amendment 91

SLC System 3.1.7 r

SR 3.1.7.2 Verify temperature o f sodium pentaborate solution i s 5 70°F.

SR 3.1.7.3 Verify temperature o f pump suctian piping 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> up to the pump suction valve i s L 70°F.

I SR 3 . 1 . 7 . 4 Verify continuity o f explosive charge. 31 days SR 3 . 1 . 7 . 5 V e r j f y the concentration o f sodium 31 days pentaborate i n solution i s within the limits o f Figure 3.1.7-1. !it.@

Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water o r sodlun pentaborate i s added to sol u t i o n Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature i s restored t o 2 70°F SR 3.1.7.6 Verify each SLC subsystem manual, power 31 days cperated, and automatic v a l v e i n the f l o w path that i s not locked, sealed, or otherwise secured i n position i s in the correct position, or can be aligned to the correct position, (continued)

NMP2 3 , 1 7-2 I ~ e n ~ ~ 91 e n t

SLC System 3.1.7 I

• For &ton-1 0 Isotope Enrichment2 25 Atom Percent I Figure 3.1.7-1 (Page 1 of 1)

Sodium Pentaborate Solution V o l u ~ ~ C o n ~ n tRequirements

~tio~

3.1.7-4 Amendment a, 111

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 89 2 217000 A2.12 3.0 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 1 N2-0P-35 Rev 6 A2.12 - Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve openings QUESTION 89 RClC is initially in STANDBY, when the following occurs 21CS-MOV143, Pump Minimum Flow to the Suppression Pool inadvertently OPENED.

0 Annunciator 601325, RClC WTR LEG PUMP 2 DISCHARGE PRESSURE LOW alarms.

0 Annunciator 601348, RClC HIGH PT VENT LEVEL LOW alarms.

CST Level has NOT changed.

0 Suppression Pool Level has NOT changed.

Which one of the following describes the condition of the RCIC system AND the required actions?

A. OPERABLE, because the CST Level has NOT substantially LOWERED. CLOSE the RClC Minimum Flow Valve.

B. INOPERABLE because the RCIC System CANNOT produce rated flow into the reactor if initiated. Enter a 14 day LCO,and CLOSE the RCIC Minimum Flow Valve.

C. INOPERABLE, because the system may be damaged if initiated. Enter a 7 day LCO, CLOSE the RCIC Minimum Flow Valve and complete a Fill and Vent of the RClC System.

D. INOPERABLE, because the system may be damaged if initiated. Enter a 14 day LCO, CLOSE the RClC Minimum Flow Valve and complete a Fill and Vent of the RClC System.

Correct Answer: D indications that fill has been lost renders RClC INOPERABLE per Page 17 of 31

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal Plausible Distractors:

A is plausible; CST drainage to Suppression Pool could render RClC INOPERABLE due to minimum water inventory consideration.

B is plausible; Minimum Flow Valve will receive a CLOSE signal when RClC is initiated.

C is plausible; 7 day LCOs are applicable for ECCS Systems.

Page 18 of 31

RCIC System 3.5'3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RfACTOR CORE ISOLATION COOLING (RGIC) SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCC System shall be OPERABLE.

APPLICABlLllY: MODE 1, MODES 2 and 3 with reactor steam dome pressure 7 150 pig.

COtSMflON II REQUIRED ACTION COMPLETION TIME A. RGC System A.1 Verifyby Ir n ~ e ~ i a t ~ y inoperama. a d r n ~means i ~ ~ ~ ~

High Pressure Core Spray System is OPERABLE.

A.2 Rsstars RCIC System 14 days to OPERABLE status.

5. Required Action and B.l 8s On MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. &&Q B .2 Reduce reactor stearn 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> dome pressure to s I50 pig.

NMP2 3.5.3-1

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier K/A Number Statement IR Origin Source Question 90 2 239002 2.2.12 4.1 N NA LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents F 1 N2-OP-34 Rev 8 Safety Relief Valves - Knowledge of Surveillance Procedures.

QUESTION 90 As a result of Safety Relief Valve Testing, it is discovered that Safety Relief Valve 2MSS*PSV130 will OPEN when the A or B solenoid was energized and WILL NOT OPEN when the C solenoid energized.

Which one of the following actions is required, if any?

A. Enter a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> LCO, due to the ADS function of a required Safety Relief Valve being INOPERABLE.

B. Enter a 14 day LCO, due to the ADS function of a required Safety Relief Valve being INOPERABLE.

C. Enter a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> LCO, due to the SAFETY function of a required Safety Relief Valve being INOPERABLE.

D. NO ACTIONS are required, because the RELIEF function of Safety Relief Valve 2MSS*PSV130 is NOT REQUIRED.

Correct Answer: D The RELIEF function of 2MSS*PSV130 is indicated by failure of the C solenoid. 2MSS*PSV130 is an ADS Valve, but the RELIEF function is not considered by accident analyses.

Plausible Distractors:

A is plausible; 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would be true for a SAFETY function discovered INOPERABLE.

B is plausible; 14 day would be true for an ADS function (solenoid A or B) discovered INOPERABLE.

C is plausible; would be true for a SAFETY function discovered INOPERABLE.

Page 19 of 31

SYSTEM D E S C R I P T I O N ( C c ~ nj t 2.2 Pressure Relief Function

~ a c hof t h e 1 8 s a f e t y / r e l i e f v a l v e s h i v e a t h r e e p o s i t i o n key-operated switch on panel 501 t o s e l e c t e i t h e r t h e l'OFF1', t'AIJTO'l o r "OPENT1 mode. Open and closed i n d i i ' a t i n g l i g h t s a l s o a r e provided far each valve which are a c t u a t e d from . a i a m u s t i c monitor located 011 t h e PSV d i s c h a r g e p i p . I n t h e "OFF" mode, the v a l v e w i l l riot o p e r a t e as a p r e s s u r e relief (pneumatic operat i o n for o v e r p r e s s u r e p r o t e c t i o n v i a t h e C S o l e n o i d ) but w i l l r e t a i n i t s s a f e t y relief f u n c t i o n { l i f t s a s r e a c t o r p r e s s u r e orercomas s p r i r i y t e n s i o n ) . I n the taAArJTQ't mode, t h e valve is r o n t r o l Led thruugh two p r e s s u r e t r a n s m i t t e i s v i a two a s s o c i a t e d t r i p u n i t s . Trip u n i t set p o i n t s a r b a s follows:

P S V ' s Operated Press. R e l i e f S e t Point T r i p Units

1. I25 & 1 3 3 110.3 p s i g N668P. & N 6 5 8 E

1. l23, 1 2 4 , 131 & 136 1113 p s i g M669A & N569E

1. 1213, 122, 1 2 5 L 132 1123 psig N670A Ei N670E

1. 121, 125, 130 & 1 3 5 1133 psig M671A & N671E

1. 127, 129, 134 & 137 1143 p s i s N677A & N67ZE Each s a f e t y i r e l i e f valve h ? s t h r e e s o l e n o i d v a l v e s { C , A & 8 ) and an SKV n i t r o g e n accumulator t a n k s u p p l i e d from t h e R;r Building n i t r o y e n headex through primaxy r ' o n t a i m e n t a u t o i s o l a t i o n v a l v e s . 6n v a l v e s t h a t do riot have an A D S f u n c t i o n , only t h e 'lC" solenoid i s connected e l e c t r i c a l l y . When the t r i p u n i t s reach their high pressure s?t p o i n t , t h e a s s o c i a t e d llCil s o l e n o i d s are energized t a p o r t n i t r o g e n to t h e cylinder of t h e s a f e t y / r e l icf valve a c t u a t o r s causing t h e val:-es t o open. A s r e a c t o r p r e s s u r e decreases and t h e t r i p u n i t s rsset, t h e llCt' s c l e n o i d de-energizes allowing t h e a c t u a t i n g n i t r o g e n t o vent from the cylinder and spring p ~ e s s u i - ec l o s e s the v a l v e .

ThermGCoUpleS downst ream af each s a f e t y / r a l l e valva a r e provided t u determine r e s e t arid valve leakage. Temperatures dre recorded on a c h a r t recorder on Z C E P P N L 6 1 4 . The keys f o r t h e s a f e t y f r e l i e f valve key-locked c o n t r o l switches can only be removed i n t h e tiAUTO'l p o s i t i o n . The p r e s s u r e r e l i e f mode i s not s a f e t y r e l a t e d s i n c e the accident. a n a l y s i s u s e s o n l y t h e S a f e t y mode of c , p e r a t i o n , but is pi-nvided t o ensure p o s i t i v e s e a t i n g of t h e PSV'S s i n c e the relief set point 1 8 b 2 p s i y l e s s than t h e s a f e t y s e t p o i n t . I t a l s o a l l o w s f o r remote manual operat ~ t . 1 1 of the PSV' ,;.

2.3 ADS Funct.ioii The ADS function is provided t o lower t h e pressurt.- of t h e r e a c t o r coolant t o allow t h e low p r e s s u r e ECCS Systems (LPCS G L P C I ) t u provide eooliriy water to the core. I t Is a back-up t o t h e High Pressure Core Spray Syst.em wh+n, duz t o a small l i n e bleak, a low r e a c t o r water l e v e l O C C U L B without a s i y n i f i c a r i t loss of p r e s s u r e and HPCS is e i t h e r u n a v a i l a b l e o r has i i i s u f f i c i e i i t c a p a c i t y t o maintain v e s s e l water l e v e l . Contxols f o r m S v a l v e s are l o c a t e d on 2CECrPNL628 (Div. I loyicj arid 2CEC*PNL631 ( D i v . I T l o g i c - ) .

Page I1 N2-OP-34 Rev 08

S/RVs 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 Safety/Reli ef Val ves (S/RVs)

LCO 3 . 4 . 4 The safety function of 16 S/RVs shall be OPERABLE, APPLICABILITY: MODES 1, 2, and 3.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.l Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> S/RVs inoperable.

AND I A-2 Be in MODE 4.

I 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify the safety function lift setpoints In accordance o f the required S/RVs are as follows: with the Inservice Number o f Setpoint Testi ng Program S/RVs (psiq) 1165 psig k 35.0 1175 psig k 35.0 1185 psig k 36.0 1195 psig rt 36.0 1205 psig f 36.0 Following testing, lift settings shall be within k 1%.

NMP2 3.4.4-1 Amendment 9 1

ECCS-Operating 3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

~

C. Two ECCS injection c. 1 Restore one ECCS 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> subsystems inoperable. injecti on/spray subsystem to OPERABLE OR status.

One ECCS injection and one ECCS spray subsystem inoperable.

D. Required Action and D. 1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time o f Condition A, -

AND B, or C not met.

D.2 Be in MODE 4 . 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> E. One required ADS valve E. 1 Restore ADS valve to 14 days i noperabl e. OPERABLE status.

F. One required ADS valve F. 1 Restore ADS valve t o 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> i noperabl e. OPERABLE status.

-OR One low pressure ECCS F.2 Restore 1 ow pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> injection/spray ECCS injectionlspray subsystem inoperable. subsystem to OPERABLE status.

(continued)

NMP2 3.5.1-2 Amendment 91

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 91 2 214000 A2.03 3.9 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 2 ARP 603444 N2-0P-30 Rev 9 Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Overtravel/in-out QUESTION 91 The plant is operating at 75% power, with the following:

0 Control Rod 14-27 has been withdrawn to position 48.

0 Control Rod 14-27 is then given a continuous withdraw signal 0 Annunciator 603444, CONTROL ROD OVERTRAVEL alarms.

Which one of the following is indicated by this alarm AND what action is required for Control Rod 14-27 as a result?

A. Collet Finger failure. FULLY INSERT it using the INSERT pushbutton per N2-SOP-08, Unplanned Power Changes.

B. Collet Finger failure. SCRAM it using BOTH SRI Toggle Switches per N2-SOP-08, Unplanned Power Changes.

C. Spud Coupling failure. RECOUPLE it by inserting TWO notches, withdraw to position 48, and perform a coupling check per N2-OP-30, Control Rod Drive.

D. Spud Coupling failure. FULLY INSERT it using the INSERT pushbutton per N2-OP-30, Control Rod Drive.

Correct Answer: C A decoupled Control Rod is indicated by 603444, CONTROL ROD OVERTRAVEL alarm. Recoupling attempt is required by N2-OP-30.

Plausible Distractors:

A is plausible; would be true for Rod Drift in the OUT direction. Collet Failure requires insertion and deactivation. A coupled DRIFTING Control Rod will NOT OVER TRAVEL.

B is plausible; would be true for Rod Drift in the OUT direction. Collet Failure requires insertion and deactivation. A coupled DRIFTING Control Rod will NOT OVER TRAVEL.

D is plausible; it is required to INSERT and DEACTIVATE the Control Rod, after attempts to recouple have failed.

Page 20 of 31

' ATTACHMENT 14 (Cont) 2CECfPNL603 SERIES 400 ALARM RESPONSE PROCEDURES Reflash: No 603444 CONTROL ROD OV ERTRAVE L 603444 Comouter Point Comouter P r i n t o u t Source RDSBC 10 CONTROL ROD Relay K1:

OVERTRAV EL Overtravel reed switch c l osed.

Automatic Resoonse NONE Correct ive Act ion

a. Monitor the LPRM i n d i c a t i o n s next t o t h e a f f e c t e d r o d on t h e 4-Rod display, f o r i n d i c a t i o n s o f a c o n t r o l r o d drop.

b. IF a Control Rod Drop i s indicated, e n t e r N2-SOP-08, Unplanned Power Changes, AND execute concurrently w i t h t h i s procedure.

C. Refer t o N2-OP-30, Section H. I 43551 Page 927 N2-ARP-01 Rev 00

H. OFF NORMAL PROCEDURES (Cont)

Initials 2.0 Uncoupled Control Rod (TS)

NOTES: 1. While performing rod coupling check, any rod which becomes uncoupled from its drive mechanism will energize annunciator 603444, CONTROL ROD OVERTRAVEL.

2. While drive is being moved, rod uncoupling could also be indicated by lack of a noticeable change in neutron monitoring indication.

3. Exact location of rod (vertically) may be determined by a TIP trace.

2.1 IF a control rod is found uncoupled, declare the control rod INOPERABLE AND perform the following within three hours per Technical Specification 3.1.3:

NOTE: IF permitted by Rod Worth Minimizer (RWM may be bypassed as allowed by TIS 3.3.2.1):

2.1 .I Insert control rod two notches in an attempt to recouple rod.

2.1.2 Consult with Reactor Engineering before control rod withdrawal.

2.1.3 Withdraw control rod to position 48 WITH Reactor Engineering concurrence, AND observe nuclear instrumentation response during rod movement.

2.1.4 Apply a continuous withdraw signal at position 48 AND verify 603444, CONTROL ROD OVERTRAVEL, does NOT alarm.

2.1.5 IF the control rod is successfully recoupled on the first attempt, reposition the control rod WITH Reactor Engineering concurrence.

2.1.6 IF recouple is NOT accomplished on the first attempt, refer to TIS 3.1.3.

2.2 IF recoupling attempt NOT permitted by the RWM, fully insert control rod per TIS 3.1.3 Required Action C.1, AND disarm control rod per subsection F.13.0 (T/S 3.1.3, Required Action C.2).

Page 78 N2-OP-30 Rev 11

N2-SOP-08 UNPLANNED POWER CHANGES See Section 1 0 Immediate Actions Shadowed scrammed OR drifted, NZ-SOP-1OlC.

Depress "HYDRAULIC PRESSURE UNIT SHUTDOWN" pushbutton at ZCEC'PNL602.

NO Reduce Reactor power to !I approximately 85% per b 4b NZ-SOP-1OlD.

. Monitor Offgas Main Steam Line Radiation Monitorj for evidence of Fuel Element Failure.

IF THEN Cause due to Control Rod Drift, Continue at IQ Cause due io Loss of Feedwater Heating, Cause due to Recirc FCV Failure, Perform Attachment 2.

Cause due to Control Rod Drop, Perform Attachment 3.

Cause due to Single Control Rod Scram, Perform Attachment 4.

Cause is Unknown, Perform Attachment 5.

NINE MILE POINT NUCLEAR STATION UNIT 2 SPECIAL OPERATING PROCEDURE PROCEDURE NUMBER N2-SOP-08

N2-SOP-08 CONTROL ROD DRIFT -

Iden!!fy which con!rol rod !s drifting and in what d!rec!ton m: Specific recommendations from the on call Rx Eng. may override CR IN I OUT directions for plant power reductions.

1 flowchart.

Using INSERT pushbutton,+J!Y inser! drif!ed control rod.

Contact Rx Eng. for additional Yes 11 control rod identified in the CRC book for the drifted

-l

+

Reduce core flow to approximately 11 65 mlbmlhr per r L

NZ-SOP-IO1D m a s Depress and hold recommendedby Rx Eng.

INSERT pushbutton to maintain control rod fully inserted. I1 Isolate stuck A

When control rod is f NZ-OP-30,Section 7.

If Collet Fingers are stuck. rod may drift as V81Ve.5 are c!Osed. '.

fully inserted, close the following valves at the

' . HCU for the drifting control rod.

No PRDS'V103 11 w SCRAM the drifted control rod by placing BOTH NORM-TEST-SRI toggle switches to TEST, m a s recommended by Rx Eng.

Reduce core flow to NZ-SOP-1O1D m as 7 Restore hydraulics by opening the following . Refer to TS 3.1.3, Control Rod Operability.

valves: b  !&j&J directed by the SM, exit lhis procedure.

ZRDS'V105 Refer to TS 3.1.5, Control Rod !Scram Accumulator Operability ZRDS'V103 I NINE MILE POlNl NUCLEAR STATION UNIT 2 SPECIAL OPERATINGPROCEDURE PROCEDURE NUMBER N2-SOP-08

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier K/A Number Statement IR Origin Source Question 92 2 245000 2.4.47 4.2 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H 2 N2-SOP-68 Rev 2 Main Turbine Gen. / Aux. - Emergency Procedures / Plan: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

QUESTION 92 The plant was operating at 100% power, when the running Stator Cooling Water Pump TRIPPED. Standby Stator Water Cooling Water Pump COULD NOT be started.

The following conditions resulted:

Recirculation Core Flow LOWERED to 55 Mlbm/hr.

0 Reactor Power LOWERED to 68%.

0 Turbine Bypass Valves are OPENING.

Which one of the following actions is required in this condition?

A. TRIP the Main Turbine per N2-SOP-21, Turbine Trip.

B. INSERT CRAM Rods per N2-SOP-101D, Rapid Power Reduction.

C. INSERT Control Rods per N2-SOP-08, Unplanned Power Changes.

D. SCRAM the reactor per N2-SOP-29, Sudden Reduction in Core Flow.

Correct Answer: A When Core Flow has lowered to 55 Mlbm/hr and Bypass Valves are OPENING, it is required to TRIP the Turbine Generator.

Plausible Distractors:

B is plausible; Rapid Power Reduction via Control Rod Insertion is required prior to exceeding SCRAM Criteria.

C is plausible; an Unplanned Power Change has occurred, Control Rod Insertion is required prior to exceeding SCRAM Criteria.

D is plausible; N2-SOP-29 SCRAM criteria involve Thermal Hydraulic Instability. Core Oscillation indications have NOT been provided.

Page 21 of 31

N2-SOP-68 LOSS OF STATOR WATER COOLING Runback may be icldrcated by the followtnq Y

See Soirion i'0.

Load set molor lowering b TB's open C%IlatdlUl Auxfftdlltrs Truubla (851112)

Comp Pt GhEEC07, GEN PROT CKT ENERGIZED Turbine has tripped, Exit ths SOP A 3 rqcover GMC per U-OP-2B c

v

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 93 2 202001 2.4.41 4.0 B NMP-2 Bank SYSID 22782 LOK GrP 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H 2 N2-SOP-08 Rev 4 LCO 3.4.1. B Amendment 97 PROVIDE LCO 3.4.1 NO BASIS Reactor ecirculation - Conduct of operations: Ability to apply technical specifications for a system QUESTION 93 Because of rising flow on the B Recirc Loop, the following actions have been taken:

0 Depressed the B FCV HYDRAULIC PRESSURE UNIT SHUTDOWN pushbutton at 2CEC*PNL602.

Closed the LOOP B HYDR FLUID OUTSIDE ISOL VALVES.

With B Recirc Loop Flow increase stopped, the following indications are observed at 08:OO:

A Recirc Loop Flow is 39 x I O 6 lbslhr (2CEC*PNL602)

B Recirc Loop Flow is 45 x lo6Ibs/hr (2CEC*PNL602)

Reactor Power is STABLE at 85%.

I&C has determined the Recirc FCVs LVDT (position feedback) has failed and it will take four (4) hours to repair the fault.

Which one of the following is the correct response?

A. A Recirc Loop is required to be IMMEDIATELY declared NOT IN OPERATION.

Initiate actions for Single Loop Operation. Raise A Loop Flow to 39.6 x 1O6 Ibs/hr to exit Single Loop Operation.

B. B Recirc Loop is required to be IMMEDIATELY declared NOT IN OPERATION.

Initiate actions for Single Loop Operation. Raise A Loop Flow to 39.6 x I O 6 Ibs/hr to exit Single Loop Operation.

C. Align the alternate RVDT for the B FCV and LOWER B Loop Flow to 44.4 x I O 6 Ibs/hr. If this action is not complete by 10:00, THEN declare the A Recirc Loop NOT IN OPERATION and initiate actions for Single Loop Operation.

D. Align the alternate RVDT to the B FCV and LOWER B Loop Flow to 44.4 x lo6 Ibs/hr. If this action is not complete by 10:00, THEN declare the B Recirc Loop NOT IN OPERATION and initiate actions for Single Loop Operation.

Page 22 of 31

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal Correct Answer: C With Recirculation Loop Mismatch exceeding 5%, it is required to declare the Recirculation Loop with the LOWER Loop Flow NOT IN OPERATION within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (3.4.1.Action B.l). Per SOP-8, OP-29, the alternative RVDT can be aligned to position the B FCV.

Plausible Distractors:

A is plausible; TS allow 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> before declaring the loop with lower flow "not in operation" and this time can be used make repairs and restore the flow mismatch to within limits.

B is plausible; TS allow 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> before declaring the loop with lower flow "not in operation" and this time can be used to restore the flow mismatch to within limits. Also, the loop to be declared "not in operation" is the loop with the lower flow which is the A loop .

D. is plausible; The loop to be declared "not in operation" is the loop with the lower flow which is the A loop .

Page 23 of 31

ATTACHMENT 2. W R C FCU FAILURE NOTE: Recrrc FCV Closure due to Reactor Feedwater Pump-Low Level Interlock 13covered 4n N2-SOP-29 10 IF required, remove affected Recirc Pump from sewice per N2-OF-29, Subsection G 2 0 (--I NiA Affected loop tu remain in service L-1 11 Exit this procedure AND establish Singhi Loop Operation per N2-OP-29, Subsectton H 6 0 (Lj Refer tu ITS 3 4 1 for loop flow mismatch limits 8-)

IF 18C determines the Recirc FGV's RUDT (position feedback) has i&d Alternate LVDT may be selected per procedure N2-OP-29 Section H 11 0 f-1 N,A, RVDJ is working properly (-1 50 IF BOTH Recirc FCVs have been hydraulcaily isolated, perform the following.

N:A, only one R e m FCV has been hydraulically isolated fLj 3 '9 Control Reactor power in accordance with Reactor Engineering reconmendattons [-I 42 Maintaier uperation of the Reactor Recirc: System rn accordance with ITS 3 4 .. (-J 50 WHEN came of Recirc FCU drift has been corrected OR as directed by Shl perfom one of the follorPilng IF Recrrc Hydraulic Power Unit was shut down exit this procedure AND enter N2-QP-29 at Subsection E 1 0 LA 0 IF Recirc Hydraulic Power LInrt was NO r shut down, exrt this procedure .. i-)

Page 20 N2-SOP-OB Rev 04

Ret ireul a t 1on Loops Operating 3.4.1 ACTIONS A. No recirculation loops A.l Be i n MODE 2.

in operation.

AND A. 2 Be f n MODE 3 .

8. Recirculation loop 3.1 Declare the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> I

+ flow mismatch not within 1 imits, recirculation loop w i t h lower f l o w t o bo not in operation.

C. Requirements of the C.l S a t i s f y the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> I LCO not met for requjrements o f the reasons other than LCO.

Conditions A and 8. I D. Required Action and D.l Be i n MODE 3.

associated Completion Tlrae o f Condltirzn C NMPP 3.4.1-2

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 94 3 Generic 2.1.25 4.2 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents H NA N2-EOP-C5 Rev 10 PROVIDE EOP Details Z and J in Question Stem as shown Ability to interpret reference materials, such as graphs, curves, tables, etc.

QUESTION 94 Following an RPV Blowdown under ATWS conditions, the following conditions exist:

0 Injection is Terminated AND Prevented.

0 FIVE Safety Relief Valves are OPEN.

0 Reactor Pressure is 250 psig LOWERING.

0 Indicated RPV Water Level is -65 inches.

FIG Z Fuel Zone Correction Curve Number of RPV Pressure Open SRVs (PSid 7 165 6 195 5 235 4 300 3 4 05 2 610

~

Which one of the following describes the status of Adequate Core Cooling?

A. ASSURED because RPV Water Level is ADEQUATE.

B. NOT ASSURED because RPV Water Level is INADEQUATE.

C. ASSURED because Steam Flow through OPEN Safety Relief Valves is ADEQUATE.

D. NOT ASSURED because Steam Flow through OPEN Safety Relief Valves is INADEQUATE.

Page 24 of 31

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal Correct Answer: C With FIVE SRVs open and Reactor Pressure above 235 psig, Table J value, indicates that Core Cooling is ASSURED because Steam Flow through OPEN Safety Relief Valves is ADEQUATE. Fuel Zone corrects to BELOW -39 inches, Level is NOT providing assured core cooling.

Plausible Distractors:

A is plausible; would be true with Reactor Pressure below 235 psig if indicated water level were above -55 inches.

B is plausible; would be true with Reactor Pressure below 235 psig.

D is plausible; would be true with Reactor Pressure below 235 psig.

Page 25 of 31

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 95 3 Generic 2.1.35 3.9 B NMP-2 Bank SYSID 22803 LOK Grp 10 CFR 55.43(b) 2 LOD (1-5) Reference Documents H NA LCO 3.9.1 Amendment 91, 10CFR20, EOP-RR, TS 3.9 (ALL)

PROVIDE 10CFR50.72 Knowledge of fuel handling responsibilities for SROs QUESTION 95 The plant is in Mode 5 with the following conditions:

0 A fuel bundle is inadvertently loaded into a cell with a partially withdrawn control rod.

0 A Reactor scram occurs due to an inadvertent criticality.

Which one of the following is an operational implication of this event and why?

A. I-hour report is required because the Refuel Bridge personnel will exceed 10CFR20.2202 limits.

B. Tech Spec actions must be entered because the Refueling Interlocks failed to perform their intended function.

C. General Emergency declaration because of the expected damage to this Fuel Assembly as a result of this event.

D. EOP-RR, Radioactivity Release Control, must be entered because a Ground Level Release is expected as a result of this event.

Correct Answer: B Moving the Refuel Bridge over the Reactor AND Lowering a Fuel Assembly should be prevented with a withdrawn Control Rod. Tech Spec Actions are required for a failed Safety Function.

Plausible Distractors:

A is plausible; I-hour report is not required. The criticality was momentary and in the Source Range, therefore the change in dose to the refueling personnel will be small.

C is plausible; Fuel Damage which would produce a General Emergency is not expected since the fuel cladding is designed to sustain the energy of an operating reactor core.

D is plausible; If the radiation level increase was substantial, the Reactor Building would isolate and SBGT would start to ensure that the secondary containment atmosphere is filtered before being released at an elevated release point (not a ground level release). EOP-SC would be entered on the high radiation level.

Page 26 of 31

Refuel ing Equtpment Interlocks 3.9.1 3.9 REFUELING OPERarXONS 3.9.1 Refueling Equipment lnterlocks LCO 3.9.1 The refuel ing equipment inter1ocks associated w l t h the reactor mode switch refuel posftion shall be OPERABLE.

APPLICABILITY: During In-vessel f u e l movement with equiwent associatird w i t h the interlocks when the reactor mode switch i s in the refuel position..

ACTIOHS CONDITION REWIRED ACTION COMPLETXON TIME A. One or more required Suspend in-vessel fremedi atel y T

refuel ing equipment fuel movement w l t h interlocks inoperable. equi w e n t a ssoci ated with the inoperable interlock(s).

B A.2.1 Insert a control rod Immedi atel y withdrawal block.

sfip A.2.2 Verify all control Immediately rods are f u l l y inserted in core cell s containing one or more fuel assemblies.

NMPL 3.9.3-1 Amendment 91

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier K/A Number Statement IR Origin Source Question 96 3 Generic 2.2.43 3.3 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents F NA GAP-OPS-01 Rev 44 Knowledge of the process used to track inoperable alarms.

QUESTION 96 Which one of the following methods is used when one input has been disabled to a MULTIPLE input Control Room Alarm per GAP-OPS-01, Administration of Operations?

A. An entry is made into the Control Room Log ONLY.

B. The status is documented in a Temporary Change ONLY.

C. A transparent Red Sticker is attached to the Alarm Window.

D. A transparent Yellow Sticker is attached to the Alarm Window.

Correct Answer: D When a single input into a multiple input alarm has been defeated, a transparent Yellow Sticker is used to identify this condition.

Plausible Distractors:

A is plausible; may be used - but is incomplete per GAP-OPS-01 Attachment 1, Defeated Annunciator Log, Attachment 1.

B is plausible; may be used - but is incomplete per GAP-OPS-01 Attachment 1, Defeated Annunciator Log, Attachment 1.

C is plausible; may be used when ALL inputs to an alarm have been defeated.

Page 27 of 31

3.11.5 (Cant)

I On shift licensed personnel perrodically should revgew the limiting conditions for uperatron and action statements in effect to ensure that the requtred actions are met.

3 1 I 6 A Clearance File containing outstanding clearances shall be maintained in or near the Control Room 3 11 7 A Defeated Annunciator Log listing outstandingiunresolved defeated annunciators, shall be maintained in or near the Control Room a Annunciator crrcurts or components that malfunction may be taken out of service under a clearance without processit.ig a temporary change b Annunciator crrcuds or components may be removed from ~erviceby a temporary change

c. Annunciator crrcurts 01 components may be removed from Service during performance of a procedure or a Work Order d Operations Branch personnel shall ensure annunciators defeated per 3 11 7 a and 3 11 7 b are entered in the Defeated Annunciator Log (Attachment 1)

e. Annuncrators defeated per 3 1I 7 a, h. and c. shall be evaluated by the Control Room staff for compensatory actions necessary while the annunciator is defeated f Annunciators defeated per 3 1 t .7 c shall be evaluated for inclusion in the Defeated Annunciator Log lf not entered in Defeated Annunciator Log.

Annunciators defeated per 3 1 I 7 c shall have additional administrative controls placed on them feg: Ragged, Control f?ooni Log e Y Defeated annuncrators are identified 21sfollowlnrs A transparent yellow sficker shatt be used to indicate one or more rnullrple

-

• Inputs have been defeated A transparent red sticker shall be used to indicate all inputs have been defeated When the last active input is defeated. the CSO shall replace the yellow sticker with a red sticker.

The document number authorizing the defeated annunciator (such as temporary change or clearance), and the associated computer pointfs) should be ildentified on the strcker, if practical Page 33 GAP-OPS-01 Rev 44

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 97 3 Generic 2.2.23 4.6 N NA LOK GrP 10 CFR 55.43(b)5 LOD (1-5) Reference Documents F NA GAP-OPS-01 Rev 44 Ability to track Technical Specification limiting conditions for operations.

QUESTION 97 Given the following:

0 High Pressure Core Spray Pump has been taken out of service for performance of a scheduled s urveiIlance.

0 The system is declared INOPERABLE.

0 A Clearance has been generated and a Work Request has been approved.

0 Pump is to be returned to OPERABLE before the end of the current shift.

Which one of the following describes the administrative requirement for tracking the Limiting Condition for Operation (LCO), per GAP-OPS-01, Administration of Operations?

A. A note is made in the Clearance.

B. A note is made in the Work Request.

C. An entry is made in the Control Room Log.

D. An entry is made in the Equipment Sta.tus Log.

Correct Answer: C. Since the pump is to be declared operable before the end of shift, an ESL entry is NOT required. The entry is made in the Control Room Log Plausible Distractors:

A is plausible; the pump will have a Clearance and the LCO may be noted, but this does NOT meet the requirements of GAP-OPS-01.

A is plausible; the pump will have a Work Request and the LCO may be noted, but this does NOT meet the requirements of GAP-OPS-01.

D is plausible; LCOs are required to be tracked, but the FORMAL tracking of the LCO is made in the Equipment Status Log. Since the pump is to be declared operable before the end of shift, an ESL entry is NOT required Page 28 of 31

3.1 1.3 The following controls are applicable to entries made by the SMlCRS or designee:

a. When a separate SM Log is used, it shall be considered a legal record subject to being entered as evidence in a court of law.

b. The SM shall record an overall summary of station operations during the respective shift, including:

e Names of the supervisors, ROs, and auxiliary operators on-duty e Date and time of entries e Surveillance tests conducted and deviations from acceptance criteria e Returning Technical Specification, TRM, ODCM (Unit 2) OR related equipment to operable status e Implementation and clearance of temporary changes 0 Reportable occurrences Entering or leaving a Tech Spec, TRM OR ODCM (Unit 2) action statement e Implementation of the Site Emergency Plan or Emergency Operating Procedures e Significant changes in radiological conditions e A narrative of significant events, including notifications to the Plant General Manager, Manager Operations, On-Call Operations Supervision, and the NRC

c. With the approval of the SM and provided the log is reviewed and approved by the SM, log entries may be written by the CRS or an CRSlSM in training, or other designee.

d. The log should remain in the Control Room until completed and reviewed.

e. Other data and format guidance may be specified by the General Supervisor Operations in separate instructions.

f. When electronic logs are used, the Control Room Log and SM Log may be combined.

Page 31 GAP-OPS-01 Rev 43

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 98 3 Generic 2.3.11 4.3 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents F NA EOP-6 Attachment 21 ODCM D 3.2.6 Ability to control radiation releases.

QUESTION 98 Following an accident, the following conditions exist:

0 Drywell Pressure is 5.0 psig.

0 Suppression Chamber Pressure is 4.5 psig.

0 It is necessary to VENT the Primary Containment.

Which one of the following Primary Containment Vent Paths will provide the LOWEST radioactivity release rate?

A. Drywell THROUGH the GTS Train.

B. Drywell with the GTS Train BYPASSED.

C. Suppression Chamber THROUGH the GTS Train.

D. Suppression Chamber with the GTS Train BYPASSED.

Correct Answer: C The Design Pressure of the GTS Train is 0.36 psig. With Drywell Pressure at 5.0 psig and Suppression Chamber Pressure at 4.5 psig, it is still possible to vent the primary containment through the GTS train using the 2 Line because the 2 line has a pressure reducing valve. Venting the Suppression Chamber provides water scrubbing to produce the LOWEST radioactivity release rate.

Plausible Distractors:

A is plausible; provides no water scrubbing.

B is plausible; provides no water scrubbing.

D is plausible; a lower release rate can be achieved by venting through the GTS train.

Page 29 of 31

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier WA Number Statement IR Origin Source Question 99 3 Generic 2.4.21 4.6 N NA LOK GrP 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents F NA PROVIDE EPIP-EPP-02 Attachment 1 Rev12 Knowlec le of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

QUESTION 99 Following a LOCA in the Drywell, the following conditions exist:

0 ALL Control Rods are fully inserted.

-:I->* RPV Water Level is UNKNOWN.

ing"is-TnirogEss.

ae=

> Radiation is 50 R/hr.

lu L-sv Which one of the following Emergency Action Levels applies to this condition?

A. Unusual Event

6. Alert C. Site Area Emergency D. General Emergency Correct Answer: C Site Area Emergency 2.1.2 is required when RPV Flooding is in effect.

Plausible Distractors:

A is plausible; Leakage in the Orywell exceeding 25 gpm warrants Unusual Event per 2.1.I B is plausible; Drywell Radiation warrants an Alert per 1.3.1.

D is plausible; Drywell Radiation above 5.2 x I O 6 Whr would warrant a General Emergency per 1.3.3.

Page 30 of 31

ATTACHMENT 1 (Cont) 2.1.2 SITE AREA EMERGENCY RPV water level CANNOT BE RESTORED AND MAINTAINED > top of active fuel OR RPV Flooding is required FPB LOSSlPotential LOSS:

Fuel clad potential LOSS, RCS LOSS Mode Applicability:

Power Operation, StartuplHot Standby, Hot SHUTDOWN, Cold SHUTDOWN, Refuel Basis:

The RPV water level used in this EAL is the top of active fuel (TAF). This value corresponds to the level which is used to INDICATE challenge to core cooling and LOSS of the fuel clad barrier.

Uncovery of the fuel irrespective of the event that causes fuel uncovery is justification alone for declaring a SITE AREA EMERGENCY. This includes events that could lead to fuel uncovery in any plant operating mode including cold SHUTDOWN and refuel. Escalation to a GENERAL EMERGENCY occurs through radiological effluent addressed in EAL 1.3.3 for DRYWELL radiation and in the EALs defined for Category 5.0, Radioactivity Release.

The terminology of "CANNOT BE RESTORED AND MAINTAINED" is intended to be consistent with the interpretation that is used in the EOPs for "restored and maintained". Momentary drops below the level limit would not REQUIRE classification at this level.

"The value of the identified parameter@)islis not able to be kept abovelbelow specified limits. This determination includes making an evaluation that considers both current and future systems performance in relation to the current value and trend of the parameter@). Neither implies that the parameter must actually exceed the limit before the classification is made nor that the classification must be made before the limit is reached."

This definition would REQUIRE the emergency classification be made prior to water level dropping below TAF IF, based on an evaluation of the current trend of RPV water level and in consideration of current and future INJECTION system performance, that RPV water level will not likely be maintained above TAF.

The EOPs REQUIRE RPV Flooding under conditions where RPV water level CANNOT BE DETERMINED. The operator is directed to ESTABLISH RPV Flooding conditions to assure ADEQUATE CORE COOLING while attempting to RESTORE RPV water level indication. Because actual RPV water level is not known under these Conditions, it must be assumed that RPV water level is below the TAF thus warranting declaration of a SITE AREA EMERGENCY.

Page 28 EPMP-EPP-0102 Rev 10

Nine Mile Point Unit 2 Senior Reactor Operator Written Examination Draft Submittal SRO Tier K/A Number Statement IR Origin Source Question 100 3 Generic 2.4.30 4.1 N NA LOK Grp 10 CFR 55.43(b) 5 LOD (1-5) Reference Documents:

H NA 10CFR50.72(b)(2)(iv)(B) 10CFR50.73 (a)(2)(iv)(B)(I)

PROVIDE 10CFR50.72.10CFR50.73 Knowledge of events related to system operation/statuz :hat must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

IOCFR 55.43(b)(5) Note: Selection of correct procedure is implied rather than stated. To select the correct answer, the appropriate procedure or knowledge of that procedure was required.

QUESTION 100 During the performance of a planned shutdown, the following occurs:

0 Reactor Mode Switch is placed in STARTUP.

0 An automatic actuation of the Reactor Protection System produced a scram, because Reactor Power was excessive.

Reactor Water Level remained stable at 185 inches.

Which one of the following is the correct reporting requirement for these conditions?

A. 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report. NO LER is required.

B. 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report. LER is required within 60 days.

C. 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report. NO LER is required.

D. 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report. LER is required within 60 days.

Correct Answer: B 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report. LER is required within 60 days. IOCFR50.72(b)(2)(iv)(B)Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. 10CFR50.73 (b)(3)(iv)(B)(I) Reactor protection system (RPS) including: reactor scram or reactor trip.

Plausible Distractors:

A is plausible; but incorrect because an LER is required.

C is plausible; but incorrect because a 4-hour report is required due to a valid actuation of RPS and an LER is required.

D is plausible; but incorrect because a 4-hour report is required due to a valid actuation of RPS.

Page31 of31

50.72 Immediate notification requirements for operating nuclear power reactors.

(a) General requirements. (1) Each nuclear power reactor licensee licensed under !j!j 50.21(b) or 50.22 holding an operating license under this part or a combined license under part 52 of this chapter after the Commission makes the finding under !j 52.103(g), shall notify the NRC Operations Center via the Emergency Notification System of:

(i) The declaration of any of the Emergency Classes specified in the licensees approved Emergency Plan; or (ii) Those non-emergency events specified in paragraph (b) of this section that occurred within three years of the date of discovery.

(2) I f the Emergency Notification System is inoperative, the licensee shall make the required notifications via commercial telephone service, other dedicated telephone system, or any other method which will ensure that a report is made as soon as practical to the NRC Operations Center.3 (3) The licensee shall notify the NRC immediately after notification of the appropriate State or local agencies and not later than one hour after the time the licensee declares one of the Emergency Classes.

(4) The licensee shall activate the Emergency Response Data System (ERDS) as soon as possible but not later than one hour after declaring an Emergency Class of alert, site area emergency, or general emergency. The ERDS may also be activated by the licensee during emergency drills or exercises if the licensees computer system has the capability t o transmit the exercise data.

-) When making a report under paragraph ( a ) ( l ) of this section, the licensee shall identify:

(i) The Emergency Class declared; or (ii) Paragraph (b)(l), One-hour reports, paragraph (b)(2), Four-hour reports, or paragraph (b)(3), Eight-hour reports, as the paragraph of this section requiring notification of the non-emergency event.

(b) Non-emergency events--(l) One-hour reports. I f not reported as a declaration of an Emergency Class under paragraph (a) of this section, the licensee shall notify the NRC as soon as practical and in all cases within one hour of the occurrence of any deviation from the plants Technical Specifications authorized pursuant t o Sec. 50.54(x) of this part.

(2) Four-hour reports. I f not reported under paragraphs (a) or ( b ) ( l ) of this section, the licensee shall notify the NRC as ,

soon as practical and in all cases, within four hours of the occurrence of any of the following:

(i) The initiation of any nuclear plant shutdown required by the plants Technical Specifications.

(ii)-( iii) [Reserved]

(iv)(A) Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actiiation results from and is part of a pre-planned sequence during testing or reactor operation.

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

\O-(x) [Reserved]

,,,I) Any event or situation, related t o the health and safety of the public or onsite personnel, or protection of the

iironment, for which a news release is planned or notification t o other government agencies has been or will be made.

2h an event may include an onsite fatality or inadvertent release of radioactively contaminated materials.

(3) Eight-hour reports. I f not reported under paragraphs (a), ( b ) ( l ) or (b)(2) of this section, the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any of the following:

(i) [Reserved]

(ii) Any event or condition that results in:

(A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or (8) The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.

(iii) [Reserved]

(iv)(A) Any event or condition that results i n valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

(8) The systems t o which the requirements of paragraph (b)(3)(iv)(A) of this section apply are:

(1) Reactor protection system (RPS) including: Reactor scram and reactor trip.

(2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main

<team isolation valves (MSIVs).

Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including: High-headl intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

(4) ECCS for boiling water reactors (BWRs) including: High-pressure and low-pressure core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.

(5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system.

(6) PWR auxiliary or emergency feedwater system.

(7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.

(8) Emergency ac electrical power systems, including: Emergency diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs.

(v) Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or 1 Mitigate the consequences of an accident.

\ Events covered in paragraph (b)(3)(v) of this section may include one or more procedural errors, equipment failures,

.d/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant t o paragraph (b)(3)(v) of this section if redundant equipment in the same system was operable and available t o perform the required safety function.

(vii)-(xi) [Reserved]

(xii) Any event requiring the transport of a radioactively contaminated person t o an offsite medical facility for treatment.

(xiii) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).

(c) Followup notification. With respect t o the telephone notifications made under paragraphs (a) and (b) of this section, in addition t o making the required initial notification, each licensee, shall during the course of the event:

( 1 ) Immediately report (i) any further degradation in the level of safety of the plant or other worsening plant conditions, including those that require the declaration of any of the Emergency Classes, if such a declaration has not been previously made, or (ii) any change from one Emergency Class t o another,, or (iii) a termination of the Emergency Class.

(2) Immediately report (i) the results of ensuing evaluations or assessments of plant conditions, (ii) the effectiveness of response or protective measures taken, and (iii) information related t o plant behavior that is not understood.

(3) Maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC.

-E3 FR 39046, Aug. 29, 1983; 48 FR 40882, Sept. 12, 1983; 55 FR 29194, July 18, 1990, as amended at 56 FR 944, Jan.

. 1991; 56 FR 23473, May 21, 1991; 56 FR 40184, Aug. 13, 1991; 57 FR 41381, Sept. 10, 1992; 58 FR 67661, Dec. 22,

-393; 59 FR 14087, Mar. 25, 1994; 65 FR 63786, Oct. 25, 2000; 72 FR 49502, Aug. 28, 20071

1. Other requirements for immediate notification of the NRC by licensed operating nuclear power rectors are contained elsewhere in this chapter, in particular Secs. 20.1906, 20.2202, 50.36, 72.216, and 73.71.

2. These Emergency Classes are addressed in Appendix E of this part.

3. Commercial telephone number of the NRC Operations Center is (301) 816-5100.

4. Requirements for ERDS are addressed in Appendix E,Section VI.

5. Actuation of the RPS when the reactor is critical is reportable under paragraph (b)(Z)(iv)(B) of this section.

50.73 Licensee event report system.

(a) Reportable events. (1) The holder of an operating license under this part or a combined license under part 52 of this chapter (after the Commission has made the finding under 5 52.103(g) of this chapter) for a nuclear power plant (licensee) shall submit a Licensee Event Report (LER) for any event of the type described in this paragraph within 60 days after the discovery of the event. I n the case of an invalid actuation reported under 5 50.73(a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification t o the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. Unless otherwise specified in this section, the licensee shall report an event if it occurred within 3 years of the date of discovery regardless of the plant mode or power level, and regardless of the significance of the structure, system, or component that initiated the event.

(2) The licensee shall report:

(i)(A) The completion of any nuclear plant shutdown required by the plant's Technical Specifications.

(B) Any operation or condition which was prohibited by the plant's Technical Specifications except when:

(1) The Technical Specification is administrative in nature; (2) The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found t o be capable of performing its specified safety functions; or (3) The Technical Specification was revised prior t o discovery of'the event such that the operation or condition was no longer prohibited at the time of discovery of the event.

Any deviation from the plant's Technical Specifications authorized pursuant t o Sec. 50.54(x) of this part.

(ii) Any event or condition that resulted in:

(A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or (B) The nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.

(iii) Any natural phenomenon or other external condition that posed an actual threat t o the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.

(iv)(A) Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section, except when:

(1) The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or (2) The actuation was invalid and; (i) Occurred while the system was properly removed from service; or (ii) Occurred after the safety function had been already completed.

(B) The systems t o which the requirements of paragraph (a)(2)(iv)(A) of this section apply are:

Reactor protection system (RPS) including: reactor scram or reactor trip.

(2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).

(3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including: high-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

( 4 ) ECCS for boiling water reactors (BWRs) including: high-pressure and low-pressure core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.

(5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system.

(6) PWR auxiliary or emergency feedwater system.

(7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.

(8) Emergency ac electrical power systems, including : emergency diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs.

( 9 ) Emergency service water systems that do not normally run and that serve as ultimate heat sinks.

(v) Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to:

\ Shut down the reactor and maintain it in a safe shutdown condition; (8) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

(vi) Events covered in paragraph (a)(2)(v) of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant t o paragraph (a)(2)(v) of this section if redundant equipment in the same system was operable and available t o perform the required safety function.

(vii) Any event where a single cause or condition caused at least one independent train or channel t o become inoperable in multiple systems or two independent trains or channels t o become inoperable in a single system designed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition;

( 8 ) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

(viii)(A) Any airborne radioactive release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, resulted in airborne dionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in

Yendix B t o part 20, table 2, column 1.

(B) Any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, exceeds 20 times the applicable concentrations specified in appendix B t o part 20, table 2, colurnn 2, at the point of entry into the receiving waters (i.e.,

unrestricted area) for all radionuclides except tritium and dissolved noble gases.

(ix)(A) Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

(1) Shut down the reactor and maintain it in a safe shutdown condition; (2) Remove residual heat; (3) Control the release of radioactive material; or (4) Mitigate the consequences of an accident.

(B) Events covered in paragraph (a)(2)(ix)(A) of this section may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required t o report an event pursuant t o paragraph (a)(2)(ix)(A) of this section if the event results from:

(1) A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or (2) Normal and expected wear or degradation.

Any event that posed an actual threat t o the safety of the riuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.

(b) Contents. The Licensee Event Report shall contain:

(1) A brief abstract describing the major occurrences during the event, including all component or system failures that contributed t o the event and significant corrective action taken or planned t o prevent recurrence.

(2)(i) A clear, specific, narrative description of what occurred so that knowledgeable readers conversant with the design of commercial nuclear power plants, but not familiar with the details of a particular plant, can understand the complete event.

(ii) The narrative description must include the following specific: information as appropriate for the particular event:

(A) Plant operating conditions before the event.

(B) Status of structures, components, or systems that were inoperable at the start of the event and that contributed t o the event.

(C) Dates and approximate times of occurrences.

(D) The cause of each component or system failure or personnel error, if known.

(E) The failure mode, mechanism, and effect of each failed component, if known.

The Energy Industry Identification System component function identifier and system name of each component or

tern referred t o in the LER.

(1) The Energy Industry Identification System is defined in: IEEE Std 803-1983 (May 16, 1983) Recommended Practice for Unique Identification in Power Plants and Related Facilities--Principles and Definitions.

(2) IEEE Std 803-1983 has been approved for incorporation by reference by the Director of the Federal Register in accordance with 5 U.S.C. 552(a) and 1CFR part 51.

(3) A notice of any changes made t o the material incorporated by reference will be published in the Federal Register. Copies may be obtained from the Institute of Electrical and Electronics Engineers, 445 Hoes Lane, P.O. Box 1331, Piscataway, NJ 08855-1331. IEEE Std 803-1983 is available for inspection at the NRC's Technical Library, which is located in the Two White Flint North Building, 11545 Rockville Pike, Rockville, Maryland 20852-2738; or at the National Archives and Records Administration (NARA). For information on the availability of this material at NARA, call 202-741-6030, or go to:

http://www. archives. gov/federaI_register/code-of_federaI_reg~ilations/ibr-locations. html.

(G) For failures of components with multiple functions, include a list of systems or secondary functions that were also affected.

(H) For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from the discovery of the failure until the train was returned t o service.

(I) The method of discovery of each component or system failure or procedural error.

( J ) For each human performance related root cause, the licensee shall discuss the cause(s) and circumstances.

'Y) Automatically and manually initiated safety system responses.

(L) The manufacturer and model number (or other identification) of each component that failed during the event.

(3) An assessment of the safety consequences and implications of the event. This assessment must include:

(i) The availability of systems or components that could have performed the same function as the components and systems that failed during the event, and (ii) For events that occurred when the reactor was shutdown, the availability of systems or components that are needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.

(4) A description of any corrective actions planned as a result of the event, including those t o reduce the probability of similar events occurring in the future.

(5) Reference t o any previous similar events at the same plant that are known t o the licensee.

(6) The name and telephone number of a person within the licensee's organization who is knowledgeable about the event and can provide additional information concerning the event and the plant's characteristics.

(c) Supplemental information. The Commission may require the licensee t o submit specific additional information beyond that required by paragraph (b) of this section if the Commission finds that supplemental material is necessary for complete understanding of an unusually complex or significant event. These requests for supplemental information will be made in writing and the licensee shall submit, as specified in 5 50.4, the requested information as a supplement to the initial LER.

' Submission of reports. Licensee Event Reports must be prepared on Form NRC 366 and submitted t o the U.S. Nuclear

iulatory Commission, as specified in !j 50.4.

(e) Report legibility. The reports and copies that licensees are required t o submit t o the Commission under the provisions of this section must be of sufficient quality t o permit legible reproduction and micrographic processing.

(f) [Reserved]

(9) Reportable occurrences. The requirements contained in this section replace all existing requirements for licensees to report "Reportable Occurrences" as defined in individual plant Technical Specifications.

[48 FR 33858, July 26, 1983, as amended at 49 FR 47824, Dec. 7, 1984; 51 FR 40310, Nov. 6, 1986; 56 FR 23473, May 21, 1991; 56 FR 61352, Dec. 3, 1991; 57 FR 41381, Sept. 10, 1992; 58 FR 67661, Dec. 22, 1993; 59 FR 50689, Oct. 5, 1994; 63 FR 50480, Sept. 22, 1998; 65 FR 63787, Oct. 25, 2000; 69 FR 18803, Apr. 9, 2004; 72 FR 49502, Aug. 28, 20071