ML081370218

Unit 1

Panel 1-9-8

1-XA-55-8B

Senso rlTrip Point:

1-ARP-9-8B

Rev. 0009

Page 42 of 42

UNIT PFD

SUPPLY

ABNORMAL

(Page 1 of 1)

Relay SE - loss of normal DC power source .

Relay TS - DC Xfer switch transfers to Emergency DC Power Source.

Regulating Transformer Common Alarm.

1-INV-252-001 , INVT-1 System Common Alarm .

Sensor

Location:

Probable

Cause:

EL 593' 250V DC Battery Board 2

A.

Loss of normal DC power source

B. DC power transfer.

C. Relay failure

D. INVT-1 System Common Alarms

1.

Fan Failure Rectifier

2.

Over temperature Rectifier

3.

AC Power Failure to Rectifier

4.

Low DC Voltage

5.

High DC Voltage

6.

Low DC Disconnect

7.

Fan Failure Inverter

8.

Alternate Source Failure

9.

Low AC Output Voltage

10. High Output Voltage

11. Inverter Fuse Blown

12. Static Switch Fuse Blown

13. Over Temperature Inverter

E. PFD Regulating XFMR Common Alarms

1.

Transformer Over temperature

2.

Fan Failure

3.

CB1 Breaker Trip

4.

CB2 Breaker Trip

Auto transfer to DC Power Source on Rectifier failure .

Auto transfer to Alternate AC supply (Regulated Transformer) on Inverter failure.

Automatic

A.

Action:

B.

Operator

A.

Action:

B.

IF 120V AC Unit Preferred is lost, THEN

REFER TO 1-AOI-57-4, Loss of Unit Preferred .

REFER TO appropriate portion of 0-OI-57C, 208V/120V AC

Electrical System.

o

References:

0-45E641-2

10-100467

1-45E620-11

0-20-100756

1-3300D15A4585-1

20-110437

(

b.

(d)

Another Unit's MMG set

The second alternate is from

another unit's MMG set

output. Unit 2 MMG is the

second alternate for either

Unit 1 or Unit 3; Unit 3 is the

second alternate for Unit 2.

Transfers to this source are

done manually at Battery

Board 2 panel 11.

MMG Sets (Unit 2&3)

(1)

The MMG is normally driven By the

AC motor, powered from 480V

Shutdown Board A. Should this

supply fail, the AC motor is

automatically disconnected and the

DC motor starts, powered from

250V Battery Board. The DC

motor has an alternate power

supply from another 250V Battery

Board. Transfer to the alternate

DC source is manual.

Underfrequency on the generator

output will trip the DC motor.

Transfer of the MMG set back to

the AC motor is manual.

(2)

The 1001 and 1003 breakers from

an MMG set will trip on overvoltage

or underfrequency at the output of

the MMG. Also Unit 2 MMG

Breakers are interlocked to prevent

alternate power to unit 1 and 3 at

the same time.

OPL171.102

Revision 6

Page 20 of 69

Obj. V.B.2.b

TP-11

Obj'v.D.2.c

Obj.V.D.2.d/j

Obj V.E.2.c

Obj'v.E.2.d/i

Obj V.B.2.h

Obj'v.C.3.e

Obj'v.D.2.j

Obj'v.E.2.i

(3)

When an under frequency or

overvoltage condition exists at the

Generator Output the following

occurs

(a)

BB panel 10 breakers from

the MMG Set trip.

OPL171.102

Revision 6

Page 21 of 69

Obj. V.B.2.h

Obj. V.C.3.e

Obj. V.D.2.j

Obj. V.E.2.i

U2

U3

1001 (U2)

1001 (U3)

1003 (U1&3)

1003 (U2)

(b)

Excitation is lost and the

MMG Set continues to run.

(The Hold to build up

voltage switch must be

depressed to restore

voltage.)

(

21. RO 263000KI .02 00I/MEMlT2G I1250VDC/3/263000KI .02//RO/SROI

Wh ich ONE of the following statements describes the operation of 250 VDC Battery Charger 2B?

A.

The normal power supply to Battery Charger 2B is 480V Common Board 1.

8.

Battery Charger 2B can supply . directly from unit 2 Battery Board room, any of the six Unit & Plant

250VDC battery boards.

C.

Battery Charger 2B is capable of supplying two Battery Boards simultaneously.

0 .01

Load shedding of the battery charger can be bypassed by placing the Emergency ON select

switch in the Emergency ON Position.

KIA Statement:

263000 DC Electrical Distribution

K1.02 - Knowledge of the physical connections and/or cause- effect relationships between D.C.

ELECTRICAL DISTRIBUTION and the following: Battery charger and battery

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

knowledge of battery charger operation.

References:

OPL171.037

Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall

or recognize discrete bits of information.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following:

1. Normal and Alternate power to Battery Charger 2B.

2. Loads capable of being supplied by Battery Charger 2B.

3. Load Shedding logic and bypass capability.

A is incorrect. This is plausible because 480V Common Board 1 is the Alternate supply to Battery

Charger 2B.

B is incorrect. This is plausible because Battery Charger 2B is capable of supplying any of the six 250V

Battery Boards, but NOT directly from Unit 2 Battery Board Room.

C is incorrect. This is plausible because Battery Charger 2B is sufficiently large enough to support the

loads, but mechanical interlocks prevent closing more than one output feeder breaker.

D is correct.

(

(2)

The Plant/Station Batteries (4, 5, and 6) are

Class Non-1E and are utilized primarily for U-2,

U-1, and U-3 respectively --for normal loads

OPL 171.037

Revision 10

Page 11 of 70

Obj V.B.1

Obj. V.C.1

Obj. V.D.1

(3)

Battery (4) Room is located on Unit 3 in the

Turbine Building on Elev. 586

(4)

Battery (5 & 6) Rooms are located on the

Turbine Floor, Elev. 617

(5)

The boards and chargers for the Unit Batteries

are located in Battery Board Rooms adjacent

to the batteries they serve, with the spare

charger being in the Unit 2 Battery Board

room. (Battery Boards 5 & 6 and their

associated chargers are located adjacent to

the batteries, but are in the open space of the

turbine floor.)

c.

250V Plant DC components

(1)

Battery charger

(a)

The battery chargers are of the solid state

rectifier type. They normally supply loads

on the 250V Plant DC Distribution

System. Upon loss of power to the

charger, the battery supplies the loads.

(b)

The main bank chargers only provide

float and equalize charge when tied to

their loads. The chargers are not placed

on fast charge (high voltage equalize)

with any loads attached.

(c)

They can recharge a fully discharged

battery in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while supplying

normal loads.

(d)

Battery charger power supplies are

Follow Procedure

manual transfer only.

(

250V Battery

Normal Source

Alternate Source

Charaer

(Charger Service bus)

1

480V SD Bd 1A

480V Common Bd 1

Comp 6D

Comp 3A

2A

480V SD Bd 2A

480V Common Bd 1

Comp6D

Comp 3A

2B

480V SD Bd 2B

480V Common Bd 1

Comp6D

Comp 3A

3

480V SD Bd 3A

480V Common Bd 1

Comp 6D

Comp3A

Obj. V.B.2

Obj. V.C.2

Obj V.D.2

(

4

5

480V SO Bd 3B

Com

60

480V Com Bd 1

Com

5C

480V Common Bd 1

Com

3A

(no alternate)

OPL171.037

Revision 10

Page 12 of 70

6

480~o~or;gd 3

(no alternate)

2B spare charger DC output can be directed to any of four

feeders. Three DC outputs can be connected to battery board 1,

2, or 3. The fourth output is connected to a new output transfer

switch (located in battery board room 4) which charges batteries

4, 5, or 6 plant batteries. A meclianical interlocKpermits closing

only: one output feeaer at a time. (A slide bar is utilized in battery

board room 2 and a Kirk key interlock is used in battery board

room 4

TP-2 & TP-7

Attention to Detail

(

XI.

Summary

We have discussed in detail the DC Power Systems at BFN.

The electrical design and operation which makes these

systems so reliable has been explained. The various systems

have been described with reference to function, components,

locations, and electrical loads. Power sources have been

identified, and instrumentation has been noted. Significant

control and alarm aspects have also been pointed out.

OPL171.037

Revision 10

Page 31 of 70

250V Battery Charger

Normal Source

Alternate Source

(Charger Service bus)

1

480V SO Bd 1A, Comp 60

480V Common Bd 1, Comp 3A

2A

480V SO Bd 2A Comp 60

480V Common Bd 1, Comp 3A

2B

480V SO Bd 2B, Comp 60

480V Common Bd 1, Comp 3A

3

480V SO Bd 3A, Comp 60

480V Common Bd 1, Comp 3A

4

480V SO Bd 3B, Comp 60

480V Common Bd 1, Comp 3A

5

480V Com Bd 1 Comp 5C

(no alternate)

6

480V Com Bd 3 Comp 3D

(no alternate)

The 2B spare charger DC output can be directed to any of four feeders. Three DC outputs

can be connected to battery board 1, 2, or 3. The fourth DC output is connected to output

transfer switch (BBR 4) to batteries 4, 5, or 6. Mechanical interlock permits closing only one

output feeder at a time. (A slide bar is utilized in battery board room 2 and a Kirk key interlock

is used in battery board room 4.)

250V DC battery chargers 1, 2A and 2B will load shed upon receipt of a Unit 1 or Unit 2

accident signal and any Unit 1/2 shutdown board being supplied by its respective diesel

generator or cross tied to a Unit 3 shutdown board and a unit three Diesel Generator. 250

VDC Battery Charger 3 will load shed on a unit 3 load shed signal.

e oad shedding feature

can be b~ssed by. placing the "Emergency" switCii on the charger. to tfie "EMERG" P.Qsition.

Station Battery charger 4 does not have load shed logic; however, battery charger 4 will

deenergize when 3B 480 SID Board deenergizes and will return when the 480V SID Board

voltage returns.

They also supply alternate control power for Units 1 and 2 4kV Shutdown Boards; however, on

Unit 3, the A, C, and 0 4kV Shutdown Boards receive both normal and alternate control power

from the 250V DC Unit Systems. (3EB receives alternate control power only.) The 250V DC

RMOV Boards are supplied from the Unit Battery Board as follows:

BB-1 supplies 250V RMOV Boards 1A, 2C, 3B.

BB-2 supplies 250V RMOV Bds 2A, 1C, 3C.

OPL171.037

Revision 10

Page 47 of70

(

-

=

..=.

-

480vSO BO 1A

NOR

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

BATTERY

CHARGER

No.1

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

480v SO B02A

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

BATTERY

CHARGER

No.2A

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

480v SO BO 2B

NOR

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

BATTERY

~

CHARGER

en

No.2B

0:w

u..

enz

1************-

~I-

480v SO B03A

~

0..

I-

NOR

)

,.-------.---i

0

I

aJ

BATTERY

N

CHARGER

0

I-

No.3

............;

480v SO BO 3B

NOR

BATTERY

CHARGER t--------+-----+--+----i--+---;--i----+---+-____

NO.4

1-----' ALT

BATT

BO 1

BATT

B02

BATT

B03

BATT

B04

480v

COMMON

BO 1

..............._..

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

TP-2

250V DC Power Distribution

(

22. RO 264000K5.06 00 l/C/A/T2Gl/82 - DG/9/264000K5.06//RO/SRO/

Given the following plant conditions:

Unit 2 is operating at Full Power.

No Equipment is Out of Service.

A large leak occurs in the drywell and the following conditions exist:

- Drywell Pressure peaked at 28 psig and is currently at 20 psig.

- Reactor Pressure is at 110 psig.

- Reactor Water Level is at -120 inches

- Offsite power is available.

Which ONE of the following describes the proper loading sequence and associated equipment?

A. II 28 RHR and 28 Core Spray pumps start at 7 seconds after the accident signal is received.

B.

RHRSW pumps lined up for EECW start at 14 seconds after the accident signal is received.

c.

Core Spray pumps (2A, 28, 2C, 2D) start immediately when voltage is available on the respective

shutdown board.

D.

2C RHR and 2C Core Spray pumps start at 7 seconds after the accident signal is received.

KIA Statement:

264000 EDGs

K5.06 - Knowledge of the operational implications of the following concepts as they apply to

EMERGENCY GENERATORS (DIESEUJET): Load sequencing

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions and times to correctly determine the effect of.load sequencing on plant equipment

supplied by the Emergency Generators.

References:

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

(

In order to answer this question correctly the candidate must determine the following:

1. Load Sequencing is NVA (Normal Voltage Available) and NOT DGVA (DIG Voltage Available).

2. Based on Item 1 above, theproper load sequencing with a Common Accident Signal (CAS) on Unit-2

alone and NOT in addition to a CAS on Unit 1.

A is correct.

B is incorrect. This is plusible because RHRSW pumps all start at 14 seconds if load sequencing is

DGVA.

C is incorrect. This is plausible based on Load Sequencing logic prior to a modification for Unit 1 restart

activities.

D is incorrect. This is plausible because 2-01-74 P&L 3.2.B defines the start time as 7 second

"intervals".

(

b.

(2)

Opens diesel output breakers if shut.

If normal voltage is available, load will

sequence on as follows: (NVA)

OPL171.038

Revision 16

Page 38 of63

INSTRUCTOR NOTES

ou.v.s.s

ou.v.c.e

Obj.v.D.15

oejv.s. 15

Time After Accident

SID Board

A

C

B

D

, 0-

RHR/GS-A_ l

7

RHR/CS B

14

RHR/CS C

21

RHR/CS D

28

RHRSW

RHRSW*

RHRSW

RHRSW pumps assigned for. EECW automatic start

c.

If

ormal voltage is NeT-available: (DGVA)

(1)

After 5-second time delay, all4kV

Shutdown Board loads except

4160/480V transformer breakers are

automatically tripped.

(2)

Diesel generator output breaker closes

when diesel is at speed.

ouv.e.s

ouv.c.e

c.

(3)

Loads sequence as indicated below

Time After Accident

SID Board

A

B

C

D

0

RHR A

RHR C

RHR B

RHR D

7

CSA

CS C

CS B

CS D

14

RHRSW*

RHRSW pumps assigned for EECW automatic start

d.

Certain 480V loads are shed whenever an

accident signal is received in conjunction with

the diesel generator tied to the board. (see

OPL171.072)

(

Residual Heat Removal System

2-01-74

Unit 2

Rev. 0133

Page 17 of 367

3.2

LPCI (continued)

B.

Upon an automatic LPCI initiation with normal power available, RFiR P-ump 2~

starts immeaiately. and 2B, 2C, 2D sequentially start at 7 second intervals.

Otherwise, all RHR pumps start immediately once diesel power is available

(and normal power unavailable).

C.

Manually stopping an RHR pump after LPCI initiation disables automatic restart

of that pump until the initiation signal is reset. The affected RHR pump can still

be started manually.

3.3

Shutdown Cooling

A.

Prior to initiating Shutdown Cooling, RHR should be flushed to Radwaste until

conductivity is less than 2.0 micromho/cm with less than 0.1 ppm chlorides

(unless directed otherwise by 2-AOI-74-1, Loss of Shutdown Cooling). If CS&S

has been aligned as the keep fill source for two days or more a chemistry

sample should be requested and results analyzed to determine if flushing is

required.

B.

When in Shutdown Cooling, reactor temperature should be maintained greater

than 72°F and only be controlled by throttling RHRSW flow. This is to assure

adequate mixing of reactor water.

1.

[NER/C] Reactor vessel water temperatures below 68°F exceed the

temperature reactivity assumed in the criticality analysis.

[INPO SER 90-017]

2.

[NER/C] Maintaining water temperature below 100°F minimizes the release of

soluble activity.

[GE SIL 541]

C.

Shutdown Cooling operation at saturated conditions (212°F) with 2 RHR pumps

operating at or near combined maximum flow (20,000 gpm) could cause Jet

Pump Cavitation. Indications of Jet Pump Cavitation are as follows:

1.

Rise in RHR System flow without a corresponding rise in Jet Pump flow.

2.

Fluctuation of Jet Pump flow.

3.

Louder "Rumbling" noise heard when vessel head is off.

Corrective action for any of these symptoms would be to reduce RHR flow until

the symptom is corrected.

(

23. RO 300000K2.02 001/MEM/T2Gl/CAI1300000K2.02/2.8/2.8/RO/SR0/1l/16/07 RMS

Which ONE of the follow ing describes the power supplies to the Control and Service Air Compressor

motors?

A.

"A" and "8" are fed from the 480V Common 8d. #1

"C" and "0" from 480V SID 8d. 18 & 28 , respectively

"G" from 4KV SID 8d. 8 and 480 SO 8d. 2A

"E" from the 480V Common 8d. #1

B.

"A" and "0" from 480V Common 8d . 1

"8" and "C" from 480V SID 8d. 18 & 28, respectively

"G" from 4KV SID 8d. 8 and 480V RMOV 8d. 2A

"F" from 480V Common 8d. #3

C.

"A" from 480V SID 8d. 18

"8" and "F" from 480V Common 8d. #3

"C" from 480V SID 8d. 1A

"0" from 480V SID 8d. 2A

"G" from 4KV Common 8d.#2

0. 01 "A" from 480V SID 8d. 18

"8" and "C" from 480V Common 8d . #1

"0" from 480V SID 8d. 2A

"G" from 4KV SID 8d. 8 and 480V RMOV 8d. 2A

"E" from 480V Common 8d. #3

KJA Statement:

300000 Instrument Air

.

K2.02 - Knowledge of electrical power supplies to the following: Emergency air compressor

KJA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

knowledge of the power supplies of ALL air compressors.

References:

Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall

or recognize discrete bits of information.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

(

In order to answer this question correctly the candidate must determine the following:

1. Power supplies to six air compressors.

NOTE: Regarding plausibility, all the power supplies listed in the distractors are capable of supplying

power to each air compressor.

A is incorrect. B, G & E are correct. A, C & D are incorrect.

B is incorrect. F & G are correct. A, B, C, & D are incorrect.

C is incorrect. A, D & F are correct. B, C &G are incorrect

D is correct.

(

X. Lesson Body

A. Control Air System

1. **The purpose of the Control Air System is to process

and distribute oil-free control air, dried to a low dew point

and free of foreign materials. This high-quality air is

required throughout the plant and yard to ensure the

proper functioning of pneumatically operated

instruments, valves, and final operators.

2. Basic Description of Flow Path

a. The station control air system has 5 air compressors,

each designed for continuous operation.

b. Common header (fed by air compressors A-D and G)

(1) The control air system is normally aligned with the

G air compressor running and loaded. The

existing A-D air compressors are aligned with one

in second lead , one in third lead, and at least one

compressor in standby.

(2) 3 control air receivers

(3) 4 dual dryers One for each unit's control air

header (units 1, 2 & 3 through their 4-inch

headers) and One standby dryer supplies the

standby, 3- inch common control air header for all

three units

(4) Outlet from large service air receiver is connected

to the control air receivers through a pressure

control valve 0-FCV-33-1, which will automatically

open to supply service air to the control air

header if control air pressure falls to 85 psig.

c. 4-inch control air header (1 per unit) is supplied from

each unit dryer and backed up by a common, 3-inch

standby header.

3. Control Air System Component Description

a. Four Reciprocating Air Compressors A-D (2-stage,

double acting, V-type) are located EI 565, U-1

Turbine Building.

(1) Supply air to the control air receivers at 610 scfm

each at a normal operating pressure of 90 - 101

psig.

(2) 480V, 60 Hz, 3-phase, drive motors

(3) Power supplies

A from 480V Shutdown Board 1B

OPL171.054

Revision 12

Page 9 of 72

- SOER 88-1

Obj. V.E.1

TP-1

Obj. V.E.3

Obj. V.D.1

The G air compressor

will be discussed later in

this section of the lesson

plan.

normally aligned to all

three units

TP-1

(

o from 480V Shutdown Board 2A

B from 480V Common Board 1

C from 480V Common Board 1

(a) Control air compressors which are powered

from the 480 VAC shutdown boards are

tripped automatically due to:

i.

under voltage on the shutdown board.

ii.

load shed logic during an accident signal

concurrent with a loss of offsite power.

NOTE: The compressors must be

restarted manually after power is restored

to the board.

(b) Units powered from common boards also trip

due to under voltage.

(4) Lubrication provided from attached oil system via

gear-type oil pump

(a) Compressor trips on

lube oil pressure < 10 psig

or

lube oil temperature >180 of

(b) Compressor cylinder is a non lubricated type

(5) Cooling water is from the Raw Cooling Water

system with backup from EECW

(a) Compressor oil cooler, compressor inter-

cooler, after cooler and cylinder water jackets

(b) Compressor inter-cooler and after cooler

moisture traps drain moisture to the Unit 1

station sump .

NOTE: Cooling water flows to the compressors are regulated

such that the RCW outlet temperature is maintained

between 70° F and 100° F. Outlet temperatures

should be adjusted low in the band (high flow rates)

during warm seasons (river temps. ~ 70°F). Outlet

temperatures should be adjusted high in the band

during the cooler seasons (river temps ~ 70°F) to

reduce condensation in the cylinders.

(c) Compressor auto trips if discharge

temperature of air> 310° F.

b. Unloaders

OPL171 .054

Revision 12

Page 10 of 72

Obj. V.B.1.

Obj. V.C.1.

Obj. V.B.2.

Obj. V.C.2.

Obj. V.E.12

Obj. V.D.10

Obj. V.B.2.

Obj. V.C.2.

Obj. V.E.12

Obj. V.D .10

(

(b) Should both the primary and the backup

controllers fail, all four compressors will come

on line at full load until these pressure

switches cause the compressors to unload at

112 psig.

(c) When air pressure drops below the high

pressure cutoff setpoint (110.8 psig), the

compressors will again come on line at full

load until the high pressure cutoff switches

cause the compressors to unload.

d. Relief valves on the compressors discharge set at

120 psig protects the compressor and piping.

e. G Air Compressor - centrifugal type, two stage

(1) Located 565' EL Turbine Bldg. , Unit 1 end.

Control Air Compressor G is the primary control

air compressor and provides most of the control

air needed for normal plant operation.

(2) Rated at 1440 SCFM @ 105 psig.

(3) Power Supply

(a) 4 kV Shutdown Board B supplies power to

the compressor motor.

(b) 480 V RMOV Bd. 2A Supplies the following :

Pre lube pump

Oil reservoir heater

Cooling water pumps

Panel(s) control power

Auto Restart circuit

(c) Except for short power interruptions on the

480v RMOV Bd, Loss of either of these two

power supplies will result in a shutdown of the

G air compressor.

(4) A complete description of the G Air compressor

controls and indications can be found in 0-01-32.

(The G and the F air compressor indications and

Microcontrollers are similar).

(a) UNLOAD MODULATE AUTO DUAL

handswitch is used to select the mode of

operation for the compressor

OPL171.054

Revision 12

Page 14 of 72

Cutout switch setpoints

are set at 112 psig to

prevent spurious

operation when G air

compressor running

Cover 01 illustrations

TP-8

3. Component Description

a. Compressors E and F (EL 565, U-3 Turbine Building)

are designated for service air.

b. The F air compressor is rated for approximately 630

SCFM @ 105 psig, centrifugal type, 2 stages

c. The power supply for both compressors is 480VAC

Common Board 3.

d. FIG air compressor comparison

(1) Controls are similar to that of the G air

compressor. There is no 4KV breaker control on

the F air compressor control panel.

(2) Control system modulates discharge air pressure

in the same manner as is done on the G air

compressor.

(3) Air system is similar to the G air compressor. A

difference is that the 2 stages of compression are

driven by one shaft for the F air compressor. On

the G air compressor, there is a separate drives;

one for each of 3 compression stages.

(4) Oil system similar to that on the G air compressor

with exception of location of components and

capacity. E compressor has an electric oil pump

that runs whenever control power is on.

(5) Cooling system is similar to that on the G air

compressor with exception of flow rate, location,

and capacity of components.

(6) Loss of power will result in F air compressor trip,

loss of the pre lube pump, and the cooling water

pumps .

(7) Restart of the compressor can be accomplished

once the compressor has come to a full stop and

any trip conditions cleared and reset.

e. AlarmslTrips

(1) The Alert and Shutdown setpoints for the Fair

compressor are listed in 0-01-33.

OPL171.054

Revision 12

Page 30 of 72

Obj. V.E.6

Obj. V.DA

TP-16

ouv.s.r

Obj. V.D.5

Set to control at approx.

95 psig - Relief Valve is

set to lift at.~ 115 psig.

TP-17

TP-18

TP-19

See for latest setpoints

(

24. RO 300000K3.01 00 lIelA/T2G lISGT/B 1OB/300000K3.0113.2/3A/RO/SRO/l l/l 6/07 RMS

A LOCA has occurred on Unit 1 and the drywell is being vented to SBGT, when a loss of the Control Air

system occurs.

Which ONE of the following describes the operation of vent valves 1-FCV-64-29, DRYWELL VENT INBD

ISOL VALVE and 1-FCV-84-19, PATH B VENT FLOW CONT?

A.

Both vent valves 1-FCV-64-29 & 1-FCV-84-19 will fail close and can not be operated .

8.

Both vent valves 1-FCV-64-29 & 1-FCV-84-19 will auto swap to control from the CAD supply line

with no operator action required.

C.oI Both vent valves 1-FCV-64-29 & 1-FCV-84-19 will auto swap to control from the CAD supply line,

however CAD supply must be manually aligned from the control room.

D.

The CAD system must be manually initiated and then vent valves 1-FCV-64-29 & 1-FCV-84-19 may

be realigned to the CAD supply.

KIA Statement:

300000 Instrument Air

K3.01 - Knowledge of the effect that a loss or malfunction of the

(INSTRUMENT AIR SYSTEM) will have

on the following: Containment air system

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions to determine the effect on the containment air system due to a loss of Control Air.

References: 1-EOI Appendicies 8G and 12, 1-AOI-32-2

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following :

1. Whether the vent valves automatically swap to be supplied by CAD or must be manually aligned.

2. Whether CAD supply to DW Control Air automatically swaps or must be manually aligned.

A is incorrect. This is plausible because the vent valves DO fail closed, however, they can be operated

with manual alignment of the CAD Tanks.

B is incorrect. This is plausible because the vent valves will auto swap to control from the CAD supply

line, however the CAD tanks must be manually aligned.

C is correct.

D is incorrect. This is plausible becase the CAD system must be manually initiated, however once this is

accomplished, no further alignment is necessary.

(

1*EOI APPENDIX*12

UNIT 1

PRIMARY CONTAINMENT VENTING

Rev. 0

Page 4 ofa

f.

VERIFY 1-FIC-84-20, PATH A VENT FLOW CONT, is indicating

approximately 100 scfm.

g.

CONTINUE in this procedure at step 12.

10.

VENT the Drywell using 1-FIC-84-19, PATH B VENT FLOW CONT, as

follows:

a.

VERIFY CLOSED 1-FCV-64-141 , DRYWELL DP COMP

BYPASS VALVE (Panel 1-9-3).

b.

PLACE keylock switch 1-HS-84-36, SUPPR CHBR/DW VENT

ISOL BYP SELECT, to DRYWELL position (Panel 1-9-54).

c.

VERIFY OPEN 1-FCV-64-29, DRYWELL VENT INBD ISOL

VALVE (Panel 1-9-54).

d.

PLACE 1-FIC-84-19, PATH B VENT FLOW CONT, in AUTO

with setpoint at 100 scfm (Panel 1-9-55).

e.

PLACE keylock switch 1-HS-84-19, 1-FCV-84-19 CONTROL, in

OPEN (Panel 1-9-55).

f.

VERIFY 1-FIC-84-19, PATH B VENT FLOW CONT, is indicating

approximately 100 scfm.

g.

CONTINUE in this procedure at step 12.

11.

VENT the Drywell using 1-FIC-84-20, PATH A VENT FLOW CONT, as

follows:

a.

VERIFY CLOSED 1-FCV-64-141, DRYWELL DP COMP

BYPASS VALVE (Panel 1-9-3).

b.

PLACE keylock switch 1-HS-84-35, SUPPR CHBR I DWVENT

ISOL BYP SELECT, to DRYWELL position (Panel 1-9-54).

c.

VERIFY OPEN 1-FCV-64-31, DRYWELL INBD ISOL VALVE

(Panel 1-9-54).

d.

VERIFY 1-FIC-84-20, PATH A VENT FLOW CONT, in AUTO

with setpoint at 100 scfm (Panel 1-9-55).

e.

PLACE keylock switch 1-HS-84-20, 1-FCV-84-20 ISOLATION

BYPASS, in BYPASS (Panel 1-9-55).

f.

VERIFY 1-FIC-84-20, PATH A VENT FLOW CONT, is indicating

approximately 100 scfm.

(

1-EOI APPENDIX-12

Rev. 0

PRIMARY CONTAINMENT VENTING

Page 7 of 8

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CROSSTIE CAD TO

1-EOI APPENDIX-8G

UNIT 1

DRYWELL CONTROL AIR

Rev. 0

Page 1 of 2

LOCATION:

Unit 1 Control Room

ATTACHMENTS:

None

1.

OPEN the following valves:

0-FCV-84-5, CAD A TANK N2 OUTLET VALVE

(Unit 1, Panel 1-9-54)

0-FCV-84-16, CAD B TANK N2 OUTLET VALVE

(Unit 1, Panel 1-9-55).

2.

VERIFY 0-PI-84-6, VAPOR A OUTLET PRESS, and 0-PI-84-17,

VAPOR B OUTLET PRESS, indicate approximately 100 psig

Panel 1-9-54 and Panel 1-9-55).

3.

PLACE keylock switch 1-HS-84-48, CAD A CROSS TIE TO DW

CONTROL AIR, in OPEN (Panel 1-9-54).

4.

CHECK OPEN 1-FSV-84-48, CAD A CROSS TIE TO DW CONTROL

AIR, (Panel 1-9-54).

5.

PLACE keylock switch 1-HS-84-49, CAD B CROSS TIE TO DW

CONTROL AIR, in OPEN (Panel 1-9-55).

6.

CHECK OPEN 1-FSV-84-49, CAD B CROSS TIE TO DW CONTROL

AIR (Panel 1-9-55).

7.

CHECK MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW,

1-PA-32-31, alarm cleared (1-XA-55-3D, Window 18).

8.

IF

MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW,

1-PA-32-31, annunciator is or remains in alarm

(1-XA-55-3D, Window 18),

THEN

DETERMINE which Drywell Control Air header is

depressurized as follows:

a.

DISPATCH personnel to Unit 1, RB, EI 565 ft, to MONITOR the

following indications for low pressure:

1-PI-084-0051, DW CONT AIR N2 SUPPLY PRESS

indicator, for CAD A (RB, EI. 565, by Drywell Access

Door),

1-PI-084-0050, DW CONT AIR N2 SUPPLY PRESS

indicator, for CAD B (RB, EI. 565, left side of 480V RB

Vent Board 1B).

(~

(

Loss Of Control Air

1-AOI-32-2

Unit 1

Rev. 0001

Page 5 of 27

2.0

SYMPTOMS (continued)

REACTOR CHANNEL A(B) AUTO SCRAM annunciator, (1-XA-55-5B,

Window 1(2)) in alarm.

MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW annunciator,

(1-XA-55-3D, Window 18) in alarm.

3.0

AUTOMATIC ACTIONS

A.

U-1 TO U-2 CONT AIR CROSSTIE, 1-PCV-032-3901, will CLOSE to separate

Units 1 & 2 when control Air Header Control Air Header pressure reaches

65 psig lowering at the valve.

B.

UNIT 2 TO UNIT 3 CONTROL AIR CROSSTIE, 2-PCV-032-3901, will CLOSE

to separate Units 2 and 3 when Control Air Header pressure reaches 65 psig

lowering at the valve.

C.

CAD SUPPLY PRESS REGULATOR, 1-PCV-084-0706, will select nitrogen

from CAD Tank A at s 75 psig Control Air pressure to supply the following:

1.

SUPPR CHBR VAC RELIEF VALVE, 1-FSV-064-0020

2.

SUPPR CHBR VAC RELIEF VALVE, 1-FSV-064-0021

D.

INST GAS SELECTOR VALVE, 1-PCV-084-0033, will select nitrogen from CAD

Tank A to supply the following:

1.

DRYWELL OR SUPPRESS CHMBR EXHAUST TO SGTS,

1-FSV-084-0019

2.

DRYWELL VENT INBD ISOL VALVE, 1-FSV-064-0029

3.

SUPPR CHMBR VENT INBD ISOL VALVE, 1-FSV-064-0032

E.

INST GAS SELECTOR VALVE, 1-PCV-084-0034, will select nitrogen from CAD

Tank B to supply the following:

1.

DRYWELL OR SUPPRESS CHMBR EXHAUST TO SGTS,

1-FSV-084-0020

2.

DRYWELL INBD ISOLATION VLV, 1-FSV-064-0031

3.

SUPPR CHBR INBD ISOLATION VLV, 1-FSV-064-0034.

(

Loss Of Control Air

1-AOI-32-2

Unit 1

Rev. 0001

Page 7 of 27

4.2

Subsequent Actions (continued)

NOTE

CNDS BSTR PMPS DISCH BYPASS TO COND 1C, 1-FCV-002-0029A and CNDS BSTR

PMPS DISCH BYPASS TO COND 1B, 1-FCV-002-0029B both fail CLOSED on a loss of

control air.

[3]

IF there is NOT a flow path for Condensate system, THEN

STOP the Condensate Pumps and Condensate Booster

Pumps. REFER TO 1-01-2.

[4]

IF any Outboard MSIV closes, THEN

PLACE the associated handswitch on Panel 1-9-3 in the

CLOSE position.

NOTE

RSW STRG TNK ISOLATION, 0-FCV-25-32, fails CLOSED on loss of control air.

o

[5]

START a High Pressure Fire Pump. REFER TO 0-01-26.

0

[6]

OPEN CAD SYSTEM A N2 SHUTOFF VALVE, 0-FCV-84-5, at

Panel 1-9-54.

0

[7]

OPEN CAD SYSTEM B N2 SHUTOFF VALVE, 0-FCV-84-16,

at Panel 1-9-55.

0

[8]

CHECK RCW pump motor amps and PERFORM Steps

4.2[8.1] through 4.2[8.5]to reduce RCW flow:

(

25. RO 400000A2.02 OO l/C/A/T2G I/RBCCW//400000A2 .02/3.8/4.I/RO/SRO/ll/l6/07 RMS

With Unit 2 operating at power, the following changes are observed:

- RBCCW Temperature lower than normal.

- Annunciator 2-XA-55-4C-6 RBCCW Surge Tank High Level is in alarm.

Which ONE of the following describes a cause for these indications and the corrective action required?

A.

Reactor Recirculation Pump seal cooler leak into RBCCW. Trip and isolate the Recirculation Pump.

B.oI

RCW leak in the RBCCW heat exchanger(s). Remove RBCCW from service following unit

shutdown.

C.

RWCU leak into RBCCW via non-regenerative heat exchanger. Isolate RWCU.

D.

Drywell equipment drain sump heat exchanger leak into RBCCW. Isolate DW Equipment Drain

Sump heat exchanger.

KIA Statement:

400000 Component Cooling Water

A2.02 - Ability to (a) predict the impacts of the following on the CCWS and (b) based on those

predictions, use procedures to correct, control, or mitigate the consequences of those abnormal

operation: High/low surge tank level

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions to determine the effect of a leak into the RBCCW system and determine which procedure

addresses this condition .

References:

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following:

1. Which leak path would provide the indications given in the question stem.

2. What actions would be required to mitigate the problem .

NOTE: All distractors are plausible leak paths into RBCCW but would indicate higher temperatures.

A is incorrect. A Reactor Recirculation Pump seal cooler leak would cause RBCCW temperature to rise.

B is Correct.

C is incorrect. A RWCU leak would cause RBCCW temperature to rise.

D is incorrect. A DW Equipment Drain Sump HX leak would cause RBCCW temperature to rise.

(

Unit 1

SURGE TANK

LEVEL HIGH

1-LA-70-2A

(Page 1 of 2)

1-XA-55-4C

SensorlTrip Point:

1-LS-070-0002A

1-ARP-9-4C

Rev. 0015

Page 12 of 43

4 Inches Above Center Line of Tank

c.

Sensor

Location:

Probable

Cause:

Automatic

Action:

Operator

Action:

RBCCW surge tank on the fourth floor in the M-G set room .

A.

Makeup valve 1-FCV-70-1 open.

B. Bypass valve 1-2-1369 leaking.

<'S. Leak into the system.

None

A. VERIFY make-up valve 1-FCV-70-1 closed, using RBCCW SYS

SURGE TANK FILL VALVE, 1-HS-70-1 , on Panel 1-9-4.

B. CHECK RBCCW PUMP SUCTION HDR TEMP, 1-TIS-70-3,

indicates water temperature is 100°F or less, on Panel 1-9-4.

C. DISPATCH personnel to verify high level, ensure bypass valve,

1-2-1369, is closed and observe sight glass level.

D. OPEN surge tank drain valve, 1-70-609, then CLOSE valve when

desired level is obtained.

E. REQUEST Chemistry to pull and analyze a sample for total gamma

activity and attempt to qualify source of leak.

F.

CHECK activity reading on RM-90-131D.

Continued on Next Page

o

(

Unit 1

1-XA-55-4C

1-ARP-9-4C

Rev. 0015

Page 13 of 43

RBCCW SURGE TANK LEVEL HIGH 1-LA-70-2A, Window 6

(Page 2 of 2)

Operator

Action:

(Continued)

NOTE

[NERlC] Reactor Recirculation Pump seal cooler leakage may be indicated by a rise in 1-RM-90-131

(Panel 1-9-10) activity (1-RR-90-131/132 Panel 1-9-2) or 1-TE-68-54 or 67 temperature

(Panel 1-9-21) or lowering of any Recirc pump seal pressure.

G. IF it is suspected that the Reactor Recirculation Pump seal cooler is

leaking, THEN

PERFORM the following:

DETERMINE which Reactor Recirculation loop is leaking and at

the discretion of the Unit Supervisor, ISOLATE. REFER TO

1-01-68 Section 7.1 or 8.2 as applicable. COOLDOWN is

required to prevent hanger or shock suppressors from exceeding

their maximum travel range.

0

WHEN primary system pressure is below 125 psig and at the

discretion of the Unit Supervisor, THEN

ISOLATE the RBCCW System to preclude damage to the

RBCCW PIPING.[IEN 89-054, GE SIL-459)

0

H. START selective valving to determine in-leakage source, if present.

0

(

References:

1-45E620-4

1-47E610-70-1

FSAR Section 10.6.4 and 13.6.2

26. RO 400000G2.4.31 00 lICfA/T2G1IRBCCWff4000002.4.3Of/ROfSRO/NO

Unit 3 is at 100% rated power with the following indications :

RECIRC PUMP MTR B TEMP HIGH (3-ARP-9-4B W13) in alarm.

RBCCW EFFLUENT RADIATION HIGH (3-ARP-9-3A W17) in alarm.

RBCCW SURGE TANK LEVEL HIGH (3-ARP-9-4C W6) in alarm.

RX BLDG AREA RADIATION HIGH (3-ARP-9-3A W22) in alarm.

RECIRC PMP MTR 3B WINDING AND BRG TEMP recorder 3-TR-68-84 is reading 170 of and

rising.

RBCCW PUMP SUCTION HDR TEMP 3-TIS-70-3 is reading 140 of and rising.

RWCU NON-REGENERATIVE HX DISCH TEMP HIGH in alarm.

AREA RADIATION MONITOR RE-90-13 and RE-90-14 are in alarm reading 55 mrlhr and rising.

Which ONE of the following describes the action(s) that should be taken?

REFERENCE PROVIDED

A. 01

Enter 3-EOI-3, Secondary Containment Control. Trip and isolate 3B Recirc Pump. Commence a

normal shutdown and cooldown in accordance with 3-GOI-100-12A, Unit Shutdown .

B.

Enter 3-EOI-3, Secondary Containment Control. Trip and isolate 3B Recirc Pump. Enter 3-EOI-1,

RPV Control at Step RC-1.

C.

Trip RWCU pumps and isolate RWCU system. Close RBCCW Sectionalizing Valve 3-FCV-70-48

to isolate non-essential loads and maximize cooling to 3B Recirc. Pump. EOI entry is not required.

D.

Enter 3-EOI-3 , Secondary Containment Control. Trip RWCU pumps and isolate RWCU system.

Commence a normal shutdown in accordance with 3-GOI-100-12A, Unit Shutdown .

KIA Statement:

400000 Component Cooling Water

2.4.31 - Emergency Procedures I Plan Knowledge of annunciators alarms and indications, and use of the

response instructions.

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions to determine the corrective actions required due to an emergency involving RBCCW

based on annunciators and indications.

References:

3-EOI-3 flowchart, 3-ARP 9-3 and 3-ARP-9-4

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble ,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

(

0610 NRC Exam

REFERENCE PROVIDED: 3-EOI-3 flowchart

Plausibility Analysis:

(

In order to answer this question correctly the candidate must determine the following:

1. EOI Entry is required solely based on ARM alarms.

2. Location of the leak is from the 3B Recic Pump.

3. RWCU temperature indications are due to insufficient cooling by RBCCW, not a RWCU leak.

4. Appropriate actions per 3-EOI-3 are to isolate the leak and monitor radiation levels.

5. Justification for Unit Shudwon and Cooldown are due to the Recirc Loop being isolated at rated

temperature and pressure (pipe hanger and support issue), and NOT Directed by 3-EOI-3.

A is correct.

B is incorrect. Entering 3-EOI-1 to initiate a scram is NOT required until radiation levels approach 1000

mr/hr in any area. This is plausible becuase the location of the leak and required isolation are correct.

C is incorrect. This is plausible if the candidate incorrectly determines that RWCU is causing the

temperature issues with 3B Recirc Pump and not vice versa. If RWCU was the leak location, the

RBCCW temperature would not be high enough to provide the given indications. The leak would have to

have occurred in the NRHX which is below the indicated RBCCW temperature.

D is incorrect. This is plausible if the candidate incorrectly determines that RWCU is causing the

temperature issues with 3B Recirc Pump and not vice versa. In addition to the justification above,

commencing a shutdown in accordance with 3-EOI-3 is not appropriate until ARMs indicate greater than

1000 mr/hr.

(

OPL171.047

Revision 12

Appendix C

Page 35 of 41

DEMIN

WATER ----.,r-I~>l<lh

MAKEUP

DRW

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

RCW

t-_........U2

TCV'S

RCW

.""",,~n TCV'S

RCW

,II1II""**"" TCV'S

RCW

I&lfiI~~**~~f:J---+-"OUTLET

626

623

0-70-607

601

U2-11.....-1

RETURN",--====-__J

HEADER

CHEMICAL

FEED

633

SUPPLY

HEADER

70

69

638

U3

67

68

'--........ U3

U2

TP-1: RBCCW SYSTEM FLOW DIAGRAM

(

8FN

Unit 3

3-XA-55-48

3-ARP-9-48

Rev. 0036

Page 17 of 45

RECIRC

PUMP MTR B

TEMP HIGH

3-TA-68-84

(Page 1 of 1)

SensorlTrip Point:

Alarm is from 3-TR-68-84, Panel 3-9-2

3-TE-68-73A RECIRC PMP MTR 3B-THR BRG UPPER FACE (190°F)

3-TE-68-73C RECIRC PMP MTR 3B-THR BRG LOWER FACE (190°F)

3-TE-68-73E RECIRC PMP MTR 3B-UPPER GUIDE BRG (190°F)

3-TE-68-73N RECIRC PMP MTR 3B-LOWER GUIDE BRG (190°F)

3-TE-68-73G RECIRC PMP MTR 3B-MOTOR WINDING A (216°F)

3-TE-68-73J RECIRC PMP MTR 3B-MOTOR WINDING B (216°F)

3-TE-68-73L RECIRC PMP MTR 3B-MOTOR WINDING C (216°F)

3-TE-68-73T RECIRC PMP MTR 3B-SEAL NO.2 CAVITY(180°F)

3-TE-68-73U RECIRC PMP MTR 3B-SEAL NO.1 CAVITY(180°F)

3-TE-68-67 RECIRC PMP MTR 3B-CLG WTR FROM SEAL CLG (140°F)

3-TE-68-70 RECIRC PMPMTR 3B-CLG WTR FROM BRG (140°F)

Sensor

Location:

Probable

Cause:

Automatic

Action:

Temperature elements are located on recirculation pump motor, Elevation 563.12,

Unit 3 drywell.

A. Possible bearing failure.

B. Possible motor overload.

C. Insufficient cooling water.

D. Possible seal failure.

E. High drywell temperature.

None

Operator

Action:

A. . CHECK following on Panel 3-9-4:

RBCCW PUMP SUCTION HDR TEMP temperature indicating

switch, 3-TIS-70-3 normal (summer 70-95°F, winter 60-80°F).

RBCCW PRI CTMT OUTLET handswitch, 3-HS-70-47A

(3-FCV-70-47) OPEN.

o

B. CHECK the temperature of the cooling water leaving the seal and

bearing coolers < 140°F on RECIRC PMP MTR 3B WINDING AND

BRG TEMP temperature recorder, 3-TR-68-84 on Panel 3-9-21.

0

C. LOWER recire pump speed until Bearing and/or Winding

temperatures are below the alarm setpoint.

0

D. CONTACT Site Engineering to PERFORM a complete assessment

and monitoring of all seal conditions particularly seal leakage,

temperature, and pressure of all stages for Recirc Pump seal

temperatures in excess of 180°F.

0

References:

3-45E620-5

GE 731E320RE

3-47E610-68-1

3-SIMI-68B

Tech Spec 3.4.1

FSAR Section 13.6.2

(

Unit3

RBCCW EFFLUENT

RADIATION

HIGH

3-RA-90-131 A

3-XA-55-3A

SensorlTrip Point:

RE-90-131D

ill

(NOTE 2)

3-ARP-9-3A

Rev. 0036

Page 25 of 51

HI-HI

(NOTE 2)

(Page 1 of 2)

Hi alarm from recorder

Hi-Hi alarm from drawer

(2)

Chemlab should be contacted for current setpoints per 0-TI-45.

Sensor

Location:

Probable

Cause:

Automatic

Action:

RE-90-131A RBCCW HX Rx Bldg, EI593, R-20 S-L1NE

HX tube leak into RBCCW system.

None

Operator

Action:

A.

DETERMINE cause of alarm by observing following:

1.

RBCCWand RCW EFFLUENT RADIATION recorder,

3-RR-90-131/132 Red pen on Panel 3-9-2.

2.

RBCCW EFFLUENT OFFLINE RAD MON, 3-RM-90-131D on

Panel 3-9-10.

o

B. NOTIFY Chemistry to sample RBCCW for total gamma activity to

verify condition.

0

C. START an immediate investigation to determine if source of leak is

RWCU Non-regenerative, Fuel Pool Cooling, Reactor Water Sample

or RWCU Recirc Pump 3A or 3B Seal Water heat exchanger(s).

0

D.

(NERlC] CHECK Following for indication of Reactor Recirculation

Pump Seal Heat Exchanger leak:

1.

LOWERING in reactor Recirculation pump 3A(3B) NO.1 or 2

SEAL, 3-PI-68-64A or 3-PI-68-63A (3-PI-68-76A or 3-PI-68-75A)

on Panel 3-9-4.

0

2.

Temperature rise on CLG WTR FROM SEAL CLG TE-68-54, on

RECIRC PMP MTR 3A WINDING AND BRG TEMP temperature

recorder, 3-TR-68-58, on Panel 3-9-21.

0

3.

Temperature rise on CLG WTR FROM SEAL CLG TE-68-67, on

RECIRC PMP MTR 3B WINDING AND BRG TEMP temperature

recorder, 3-TR-68-84, on Panel 3-9-21.

0

Continued on Next Page

(

Unit 3

3-XA-55-3A

3-ARP-9-3A

Rev. 0036

Page 26 of 51

RBCCW EFFLUENT RADIATION HIGH 3-RA-90-131A, Window 17

(Page 2 of 2)

Operator

Action: (Continued)

E. IF it is determ ined the source of leakage is from Reactor Recirc

Pump A(B), THEN

1.

ISOLATE Reactor Recirculation Loop A(B) per 3-01-68, as

applicable.

0

NOTE

Cooldown is required to prevent hangers or shock suppressors from exceeding their maximum travel

range.

2.

WHEN primary system pressure is less than 125 psig, THEN

ISOLATE RBCCW System to preclude damage to RBCCW

piping.

[lEN 89-054 , GE SIL-459)

0

References:

3-45E620-3

3-47E610-90-3

GE 3-729E814-3

Unit3

RX BLDG AREA

RADIATION

HIGH

3-RA-90-1D

(Page 1 of 2)

3-XA-55-3A

SensorlTrip Point:

RI-90-4A

RI-90-8A

RI-90-9A

RI-90-13A

RI-90-14A

RI-90-20A

RI-90-21A

RI-90-22A

RI-90-23A

RI-90-24A

RI-90-25A

RI-90-26A

RI-90-27A

RI-90-28A

RI-90-29A

3-ARP-9-3A

Rev. 0036

Page 32 of 51

For setpoints REFER TO

3-SIMI-90B.

Sensor

RE-90-4

MG set area

Rx Bldg EI. 639

R-17 Q-L1NE

Location:

RE-90-8

Main Control Room

Rx Bldg EI. 617

R-16 R-L1NE

RE-90-9

Clean-up System

Rx Bldg EI. 621

R-16 T-L1NE

RE-90-13

North Clean-up Sys.

Rx Bldg EI. 593

R-16 P-L1NE

RE-90-14

South Clean-up Sys.

Rx Bldg EI. 593

R-16 S-L1NE

RE-90-20

CRD-HCU West

Rx Bldg EI. 565

R-16 R-L1NE

RE-90-21

CRD-HCU East

Rx Bldg EI. 565

R-20 R-L1NE

RE-90-22

Tip Room

Rx Bldg EI. 565

R-19 P-L1NE

RE-90-23

Tip Drive

Rx Bldg EI. 565

R-19 P-L1NE

RE-90-24

HPCI Room*

Rx Bldg EI. 519

R-21 U-L1NE

RE-90-25

RHR West

Rx Bldg EI. 519

R-16 U-L1NE

RE-90-26

Core Spray-RCIC

Rx Bldg EI. 519

R-16 N-L1NE

RE-90-27

Core Spray

Rx Bldg EI. 519

R-20 N-L1NE

RE-90-28

RHR East

Rx Bldg EI. 519

R-20 U-L1NE

RE-90-29

Suppression Pool .

Rx Bldg EI. 519

R-19 U-L1NE

Due to the location of the Rad Monitor in relation to the Test line in the HPCI

Quad, the HPCI Room Rad Alarm may be received when the HPCI Flow test

is in progress.

Probable

Cause:

Automatic

Action:

Radiation levels have risen above alarm set point. HPCI Flow Rate Surveillance in

Progress.

None

Continued on Next Page

(

Unit3

3-XA-55-3A

3-ARP-9-3A

Rev. 0036 *

Page 33 of 51

Operator

Action:

RX BLDG AREA RADIATION HIGH 3-RA-90-1D, Window 22

(Page 2 of 2)

A. DETERMINE area with high radiation level on Panel 3-9-11. (Alarm

on Panel 3-9-11 will automatically reset if radiation level lowers

below setpoint.)

B. IF the alarm is from the HPCI Room while Flow testing is being

performed, THEN

REQUEST personnel at the HPCI Quad to validate conditions.

C. NOTIFY RADCON.

D. IF the TSC is NOT manned and a "VALID" radiological condition

exists., THEN

USE public address system to evacuate area where high airborne

conditions exist

E. IF the TSC is manned and a "VALID" radiological condition exists,

THEN

REQUEST the TSC to evacuate non-essential personnel from

affected areas.

F.

MONITOR other parameters providing input to this annunciator

frequently as these parameters will be masked from alarming while

this alarm is sealed in.

G. IF a CREV initiation is received, THEN

1.

VERIFY CREV A(B) Flow is ~ 2700 CFM, and ~ 3300 CFM as

indicated on 0-FI-031-7214(7213) within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of the CREV

initiation. [BFPER 03-017922]

2.

IF CREV A(B) Flow is NOT ~ 2700 CFM, and s 3300 CFM as

indicated on 0-FI-031-7214(7213) THEN

PERFORM the following: (Otherwise N/A)

[BFPER 03-017922]

a.

STOP the operating CREV per 0-01-31.

b.

START the standby CREV per 0-01-31.

H. IF alarm is due to malfunction, THEN

REFER TO 0-01-55.

I.

ENTER 3-EOI-3 Flowchart.

J.

REFER TO 3-AOI-79-1 or 3-A01-79-2 if applicable.

o

References:

3-45E620-3

3-45E610-90-1

GE 730E356-1

(

Unit 3

SURGE TANK

LEVEL HIGH

3-LA-70-2A

(Page 1 of 2)

3-XA-55-4C

SensorlTrip Point:

3-LS-070-0002A

3-ARP-9-4C

Rev. 0028

Page 12 of 44

4 inches above center line of tank

Sensor

Location:

Probable

Cause:

Automatic

Action:

Operator

Action:

RBCCW surge tank in the MG set room EI 639'.

A. Makeup valve, 3-FCV-70-1, open.

B. Bypass valve 3-BYV-002-1369 leaking.

C. Leak into the system.

None

A. CHECK make-up valve 3-FCV-70-1, 3-HS-70-1, CLOSED on

Panel 3-9-4.

B. CHECK RBCCW system water leaving the RBCCW system heat

exchangers is 100°F or less on 3-TI-70-3, Panel 3-9-4.

C. DISPATCH personnel to verify high level and to ensure

3-BYV-002-1369, FCV-70-1 BYPASS VALVE is CLOSED.

OBSERVE sight glass level.

D. OPEN surge tank drain valve, 3-DRV-070-0609. CLOSE valve

when desired level is obtained.

E. REQUEST Chemistry to pull and analyze a sample for total gamma

activity and attempt to qualify source of leak.

F. CHECK activity reading on 3-RM-90-131 Band 3-RM-90-131 D.

Continued on Next Page

o

n

o

oo

(

Unit 3

3-XA-55-4C

3-ARP-9-4C

Rev. 0028

Page 13 of 44

RBCCW SURGE TANK LEVEL HIGH 3-LA-70-2A, Window 6

(Page 2"of 2)

Operator

Action: (Continued)

NOTE

[NER/C) Reactor Recirculation Pump seal cooler leakage may be indicated by a rise in 3-RM-90-131

(Panel 3-9-10) activity (3-RR-90-131 /132, Panel 3-9-2 or 3-TE-68-54 or 67 temperature,

Panel 3-9-21) or a lowering in any Recirc pump seal pressure.

G. IF it is suspected that the Reactor Recirculation Pump seal cooler is

leaking, THEN

PERFORM the following:

DETERMINE which Reactor Recirculation loop is leaking and

ISOLATE. REFER TO 3-01-68 Section 7.1 or 8.2 as applicable.

Cooldown is required to prevent hangers or shock suppressors

from exceeding their maximum travel range.

0

WHEN primary system pressure is below 125 psig, THEN

ISOLATE the RBCCW System to preclude damage to the

RBCCW piping.

[IEN89-054, GE SIL-459)

0

H. START selective valving to determine in-leakage source , if present.

References:

3-45N620-4

3-47E610-70-1

FSAR Sections 10.6.4 and 13.6.2

3-47E822-1

(

EOI - 3

OPL171.034

Revision 11

Appendix C

Page 30 of 30

TABLE 4

SECONDARY CONTAINMENT AREA RADIATION

APPLICABLE

MAX NORMAL

MAX SAFE

POTENTIAL

AREA

RADIATION

VALUE

ISOLATION

INDICATORS

MRIHR

MR/HR

SOURCES

RHR SYS I PUMPS90-25A

ALARMED

1000

FCV-74-47, 48

RHR SYS II PUMPS

90-2BA

ALARMED

1000

FCV-74-47,48

HPCI ROOM

90-24A

A LARMED

1000

FCV -73 -2, 3, 81

FCV-73-44

CS SYS I PUMPS90-26A

ALARMED

1000

RCIC ROOM

FCV-71 -2, 3, 39

CS SYS II PUMPS90-27A

ALAR MED

1000

NO'l E

TORUS

FCV-73 -2, 3, 81

90-29A

ALAR MED

1000

FCV-74 -47, 48

GENERAL AREA

FCV-71 -2, 3

RB EL 565 W

90-20A

ALARMED

1000

FCV-69-1, 2, 12

SDV VENTS & DRAI NS

RB EL 565 E

90-21A

ALARMED

1000

SDV VENTS & DRAINS

RB EL 565 NE

90-23A

ALARM ED

1000

NO'l E

TIP ROOM

90-22A

ALAR MED

100 ,000

TI P BAL L VALVE

RB EL 593

90-13A, 14A

A LARMED

1000

FCV-74 -47 ,48

RB EL 621

90-9A

ALARMED

1000

FCV-43-13, 14

RECIRC MG SETS

90-4A

ALARMED

1000

NO'lE

REFUEL FLOOR

90-1A, 2A, 3A

ALARMED

1000

NO'lE

TP -7 EOI-3 TABLE 4

E

MINATION

REFERENCE

.PROVIDED TO

CANDIDATE

(

~-oau

C")*-ow

~

il,H-t1UIIrrrn

I

SlH"ttrr-r<lI I I I

~

1!l1 !!

!!

I

-! I

i ,I: .

iiiI III! iii!

II 1 II

I

orII

I iI iiiI 1111 I

I

r

It ..

I I I!!

I I I'"III!

I' IIi I I I

I

C")*-ow

(

27. RO 201003K3.03 OOl/MEM/TIG2/85-3/Bl1/201003K3.03/3.6/3.7/RO/SR0/11/l6/07 RMS

Given the following plant conditions:

AOI 85-3, CRD System Failure, directs a manual scram based on low reactor pressure.

Which ONE of the following PROCEDURAL reactor pressure limits should be adhered to in this case and

WHY?

A.

980 psig reactor pressure, because this would be the lowest pressure a scram can be ensured due

to the loss of accumulators.

B.oI

900 psig reactor pressure, because this would be the lowest pressure a scram can be ensured due

to the loss of accumulators.

C.

445 psig reactor pressure, because this would be the lowest pressure required to lift a control rod

blade.

D.

800 psig reactor pressure, because this is the Technical Specification pressure for scramming

control rods for scram time testing .

KIA Statement:

201003 Control Rod and Drive Mechanism

K3.03 - Knowledge of the effect that a loss or malfunction of the CONTROL ROD AND DRIVE

MECHANISM will have on following : Shutdown margin

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

knowledge of CRD mechanism limitations and the basis for that limitation related to the ability to effect

and maintain shutdown margin.

References:

1/2/3-AOI-85-3, OPL 171.005, OPL171.006

Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall

or recognize discrete bits of information.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

(

In order to answer this question correctly the candidate must determine the following:

1. The minimum pressure allowed by 1/2/3-AOI 85-3, CRD System Failure.

2. The basis for that minimum pressure.

A is incorrect. This is plausible because 980 psig is the setpoint for the Low Accumulator Pressure

alarm.

B is correct.

C is incorrect. This is plausible because the entire statement is accurate, but is not the pressure

specified by 1/2/3-AOI 85-3, CRD System Failure.

D is incorrect. This is plausible because the entire statement is accurate, but is not the pressure

specified by 1/2/3-AOI 85-3, CRD System Failure.

OPL171.006

Revision 9

Page 17 of 60

C

(a)

A specific pattern of control rod

withdrawal or insertion

(b)

Written step-by-step path used by

the operator in establishing the

expected rod pattern and flux

shape at rated power

(c)

Deviation from the established

path could result in potentially

high control rod worths

(9) Shutdown margin

OBJ. V.B.15.c

(a)

Technical specifications of the

plant require knowing whether the

plant can be shutdown to a safe

level

(b)

Without the insertion capability of

Obj. V.B.20.g

all control rods, shutdown margin

will not be as great, thus closer to

an inadvertent criticality

(10)

Control Rod Worth variables

(a)

Moderator temperature

OBJ. V.8.20.e

i.

As temperature rises,

SER 3-05

slowing down length and

thermal diffusion length

increase

ii.

Rod worth increases with

as moderator temperature

increases

(b)

Void effects on rod worth

i.

As voids increase, average

neutron flux energy

increases

ii.

U238 and Pu240 will

(

capture more epithermal

neutrons through

resonance

(

CRD System Failure

1-AOI-85-3

Unit 1

Rev. 0003

Page 7 of 11

4.1

Immediate Actions (continued)

[2]

IF operating CRD PUMP has tripped AND backup CRD PUMP

is NOT available, THEN (Otherwise N/A)

PERFORM the following at Panel 1-9-5:

[2.1 ]

PLACE CRD SYSTEM FLOW CONTROL, 1-FIC-85-11 ,

in MAN at minimum setting.

D

[2.2]

ATTEMPT TO RESTART tripped CRD Pump using one

of the following:

CRD PUMP 1B, using 1-HS-85-2A

CRD Pump 1A, using 1-HS-85-1A

D

[2.3]

ADJUST CRD SYSTEM FLOW CONTROL,

1-FIC-85-11, to establish the following conditions:

CRD CLG WTR HDR DP, 1-PDI-85-18A,

approximately 20 psid.

D

CRD SYSTEM FLOW CONTROL, 1-FIC-85-11,

between 40 and 65 gpm.

D

[2.4]

BALANCE CRD SYSTEM FLOW CONTROL,

1-FIC-85-11 , and PLACE in AUTO or BALANCE.

D

[3]

IF Reactor Pressure is less than 900 psig AND either of the

following conditions exists:

In-service CRD Pump tripped and neither CRD Pump can

be started , OR

Charging Water Pressure can NOT be restored and

maintained above 940 psig, THEN

PERFORM the following: (Otherwise N/A)

[3.1]

[3.2]

MANUALLY SCRAM Reactor and IMMEDIATELY

PLACE the Reactor Mode Switch in the SHUTDOWN

position.

REFER TO 1-AOI-100-1. [Item 020]

D

OPL 171.006

Revision 9

Page 30 of 60

(

(6)

The withdraw motion is terminated prior

to reaching the desired position and the

rod is settled as discussed earlier.

d.

Cooling water is continuously supplied via the

P-under port and insert header.

(1)

Flow from plug type orifice in flange

follows passage between outer tube and

thermal sleeve to outer screen.

(2)

Cooling water is required to protect the

OBJ. V.B.18

graphitar seals from high reactor

temperatures.

(3)

Long exposures at high temperatures will

result in brittle, fast- wearing seals.

(4)

Drive temperature should be maintained

at <350°F and the cause should be

investigated if it exceeds this value.

(5)

Concern is that the high temperature

may be caused by a leaking scram

discharge valve.

(6)

This problem should be corrected as

soon as possible to prevent damage to

the valve.

e.

Scram function

(1)

There are two sources of water that can

OBJ. V.B/E.11,

be used to scram a drive: reactor water

V.D.10

and accumulator water.

(2)

Reactor water scram feature

(a)

Reactor water, if at high enough

pressure, is capable of scramming

More on required

the drive without any accumulator

amount of

assistance.

pressure to lift

drive and control

(b)

The over-piston area is opened to

rod later in LP.

the scram discharge header.

(

(2)

The primary effect is reduced 10 of the

inner tube just below the bottom of the

collet piston.

(a)

In serious overpressure situations,

this squeezes the inner tube

against the circumference of the

index tube.

(b)

The index tube is then held in the

insert overtravel position and often

cannot be withdrawn.

OPL171 .006

Revision 9

Page 35 of 60

(3)

Bulging of the index tube as described

above also occurs.

b.

Extensive procedural controls are specified to

prevent improper valving of the hydraulic

module.

c.

Particular caution should be observed during

the startup test program.

3.

Scram Capability

a.

Piston areas

(1)

Under-piston area equals 4.0 in2.

(2)

Over-piston area equals 2.8 in2.

b.

Normal scram forces

(1)

During a normal scram condition, the

over-piston area is opened to the scram

discharge volume which is initially at

atmospheric pressure.

(2)

Accumulator and/or reactor pressure is

simultaneously applied to the under-

piston area. The net initial force applied

to the drive (taking no credit for the

accumulator) can be calculated as

follows.

Fnet =(Forces Up) - (Forces Down)

(

Fnet = (Rx Pressure x Under-Piston Area) -

(Rx Pressure x Area of Index Tube

+ Weight of Blade + Friction)

Fnet =(1000 psig x 4.0 in2) - [1000 psig

x (4.0 in2 - 1.2 in2)] - 255 Ibs -

- 500 Ibs

Fnet = 4000 - 2800 - 255 - 500

OPL171.006

Revision 9

Page 36 of 60

Note: 4 in2

upward force -

1.2 in2

downward force

= 2.8 in2

Fnet = 445 Ibs

(Upward)

c.

Single failure proof - There is no single-mode

failure to the hydraulic system which would

prevent the drive from scramming .

d.

Accumulator versus reactor vessel pressure

scrams

(1 )

TP-9 represents a plot of 90 percent

scram times versus reactor pressure.

(a)

Reactor pressure only

(b)

Accumulator pressure only

(c)

Combined reactor and

accumulator pressure

TP-9

(2)

Scram times are measured for only the

first 90% of the rod insertion since the

buffer holes at the top end of the stroke

slow the drive.

(3)

Reactor-pressure-only scram

(a)

As can be seen from TP-9, the

drive cannot be scrammed with

reactor pressure ~ 400 psig.

(b)

The net initial upward force

available to scram the drive can

be calculated as follows.

OPL171.006

Revision 9

Page 38 of 60

(

e.

Average scram times (normal drive)

TP-9

(1)

Technical Specifications state that scram

times are to be obtained without reliance

on the CRD pumps.

(2)

Consequently, the charging water must

be valved out on the drive to be tested.

(3)

Maximum scram time for a typical drive

occurs at 800 psig reactor pressure.

(4)

This is why Technical Specifications

specify that scram times are to be taken

at 800 psig or greater reactor pressure.

f.

Abnormal scram conditions

(1)

Scram outlet valve failure to open

(2)

Drive will slowly scram on seal leakage

as long as accumulator charging water

pressure stays greater than reactor

pressure.

(3)

If the accumulator is not available, the

drive will not scram (this is a double

failure).

g.

Control Rods failure to Insert After Scram

Obj. V.D.11

(1)

This condition could be due to hydraulic

lock.

(2)

Procedure has operator close the

See 2-01-85 &2-

Withdraw Riser Isolation valve. Connect

EOI App-1 E for

drain hose to Withdraw Riser Vent Test

detailed

Connection on the affected HCU. Slowly

operations

open Withdraw Riser Vent. When inward

motion has stopped, close Withdraw

Self Check

Riser Vent.

Peer Check

(

28. RO 201006K4.09 OOl/MEM/T2G2/RWM//201006K4.09/3.2/3.2/RO/SR0/11/l6/07 RMS

The Rod Worth Minimizer must be INITIALIZED to properly determine rod position and sequence.

Which ONE of the following describes how RWM System INITIALIZATION is accomplished?

A.

INITIALIZATION occurs automatically when the RWM is unbypassed.

B.

INITIALIZATION occurs automatically every 5 seconds while in the transition zone.

C.oI INITIALIZATION must be performed manually using the INITIALIZATION push-button when the

RWM is unbypassed.

D.

INITIALIZATION must be performed manually using the INITIALIZATION push-button when power

drops below the LPSP.

KIA Statement:

201006 RWM

K4.09 - Knowledge of ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIF IC) design feature(s)

and/or interlocks which provide for the following : System initialization : P-Spec(Not-BWR6)

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific of

which plant condition would INITIALIZE the RWM.

References:

1/2/3-01-85, OPL 171.024

Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall

or recognize discrete bits of information.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following :

1. When RWM INITIALIZATION is required .

2. How RWM INITIALIZATION is accomplished.

A is incorrect. This is plausible because initialization is required when the RWM is unbypassed, but this

must be done manually.

B is incorrect. This is plausible because the RWM automatically initiates a "scanllatch" to determine the

correct latched rod group, but this is not the same as INITIALIZATION.

C is correct.

D is incorrect. This is plausible because the RWM must be manually INITIALIZED, but the RWM does

not require initialization because the LPSP is reached. THe RWM will automatically perform a

"scanllatch" at that point.

OPL171.024

Revision 13

Page 19 of 53

(

INSTRUCTOR NOTES

(2)

The MANUAL indicator light will then be Obj. V.B.6

lit and all error and alarm indications

that were on prior to bypass will be

blanked out on the RWM system

displays.

(3)

A manual bypass will also light the

RWM and PROGR indicator on the

RWM-COMP-PROGR-BUFF

pushbutton.

f.

SYSTEM INITIALIZE pushbutton

switch/indicator

(1)

The SYSTEM INITIALIZE switch is

depressed to initialize the RWM

system.

(2)

Initialization must be performed

whenever the RWM has been taken off

line, as occurs whenever the RWM

program is aborted or manually

bypassed.

(3)

Therefore, following any program abort

or bypass, the SYSTEM INITIALIZE

switch must be depressed before the

program can be run again.

(4)

The SYSTEM INITIALIZE window

lights white while the switch is held

down.

g.

SYSTEM DIAGNOSTIC switch/indicator

(1)

This switch can be pressed at any time

after the system has been initialized to

request that the system diagnostic

routine be performed.

(2)

The RWM program will thereupon be

initiated and will perform the routine,

which consists of applying and then

removing in sequence the insert and

withdraw blocks (nominal 10 second

frequency).

(3)

The operator can verify the operability

NOTE: Rod insert

of the rod block circuits by observing

and withdrawal

(

that the INSERT BLOCK and

permit lights will go

WITHDRAW BLOCK alarm lights come

off when block is

on and then go off as the blocks are

applied.

(

Control Rod Drive System

1-01-85

Unit 1

Rev. 0005

Paue 136 of 179

8.18

Reinitialization of the Rod Worth Minimizer

[1 ]

VERIFY the following initial conditions are satisfied:

The Rod Worth Minimizer is available to be placed in

operation

D

Integrated Computer System (ICS) is available

D

The Shift Manager/Reactor Engineer has directed

reinitialization of the Rod Worth Minimizer

D

[2]

REVIEW all Precautions and Limitations in Section 3.3.

D

[3]

VERIFY RWM SWITCH PANEL, 1-XS-85-9025 in NORMAL.

D

[4]

CHECK the Manual/Auto Bypass lights are extinguished.

D

[5]

DEPRESS AND HOLD INOP/RESET pushbutton.

D

[6]

CHECK all four lights (RWM/COMP/PROG/BUFF) are

illuminated.

D

[7]

RELEASE INOP/RESET pushbutton and CHECK all four

lights extinguished.

D

[8]

SIMULTANEOUSLY DEPRESS OUT OF

SEQUENCE/SYSTEM INITIALIZE pushbutton and

INOP/RESET pushbutton to place the Rod Worth Minimizer in

service.

D

[9]

IF Rod Worth Minimizer will NOT initialize, THEN

DETERMINE alarms on RWM Display Screen and CORRECT

problems.

D

[10]

IF unable to correct problems and initialize RWM, THEN

NOTIFY Reactor Engineer.

D

(

Control Rod Drive System

1-01-85

Unit 1

Rev. 0005

Page 19 of 179

3.3

Rod Worth Minimizer (RWM) (continued)

N.

For group limits only, RWM recognizes the Nominal Limits only. The Nominal

Limit is the insert or withdraw limit for the group assigned by RWM. The

Alternate Limit is no longer recognized by the RWM as an Acceptable

Group Limit.

O.

During RWM latching, the latched group will be the highest numbered

group with 2 or less insert errors and having at least 1 rod withdrawn past its

insert limits.

1.

With Sequence Control ON, latching occurs as follows: (Normally, startups

will be performed with Sequence Control ON)

a.

RWM will latch down when all rods in the presently latched

group have been inserted to the group insert limit and a rod in the next

lower group is selected.

b.

RWM will latch up when a rod within the next higher group is selected,

provided that no more than two insert errors result.

2.

With Sequence Control OFF, latching occurs as follows:

a.

For non-repeating groups, latching occurs as described above, OR

b.

For repeating groups, latching occurs to the next setup or set down

based on rod movement as opposed to rod selection.

P.

Latching occurs at the following times:

1.

System initialization.

2.

Following a "System Diagnostic" request.

3.

When operator demands entry or termination of "Rod Test."

4.

When power drops below LPAP.

5.

When power drops below LPSP.

6.

Every five seconds in the transition zone.

7.

Following any full control rod scan when power is below LPAP.

8.

Upon demand by the Operator (Scan/Latch Request function).

9.

Following correction of insert or withdraw errors.

(

29. RO 202001K6.09 OOl/C/A/T2G2/68 - RECIRC/24/202001 K6.09//RO/SROI

Given the following plant conditions:

Unit 3 is operating at 55% power with Reactor Feed Pump (RFP) "A" & "C" running and RFP "B"

idling.

Both Recirculation Pump speeds are 53%.

The "A" RFP trips, resulting in the following conditions:

Reactor Water level Abnormal alarm sealed in

Reactor Vessel Wtr Level Low Half Scram alarm sealed in

Indicated Reactor Water Level drops to _10" before RFP "B" is brought on line to reverse the level

trend and level is stabilized at 33".

Which ONE of the following describes the steady state condition of both Recirculation Pumps?

A.

Running at 53% speed

B.

Running at 45% speed

c.Y' Running at 28% speed

D.

Tripped on ATWS/RPT signal.

KIA Statement:

202001 Recirculation

K6.09 - Knowledge of the effect that a loss or malfunction of the following will have on the

RECIRCULATION SYSTEM: Reactor water level

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions and times to determine the effect of a change in reactor water level on the Recirculation

System.

References: 3-01-68, OPL 171.007, OPL171.012

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble ,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

0610 NRC Exam

(

l

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following:

1. Did plant conditions exceed the Recirc Runback setpoint.

2. Which Runback is appropriate for the given conditions.

A is incorrect. Total Feedflow would drop below 19% with only one RFP running at 55% rated power,

thus initiating a Recirc Runback to 28%. This is plausible based on the initial power level being close

enough to create doubt on total feedflow resulting from the trip of one RFP.

B is incorrect. This is plausible because a Recirc Runback DID occur, but the 45% speed given in the

distractor is the typical speed the Recirc Pumps run at during startup , not following a RFP trip.

C is correct.

D is incorrect. This is plausible because ATWS/RPT signals are associated with low RPV level, however

the setpoint is -45 inches and level only lowered to -10 inches.

(

Reactor Recirculation System

3-01-68

Unit 3

Rev. 0066

Page 13 of 179

3.0

PRECAUTIONS AND LIMITATIONS (continued)

10. The out of service pump may NOT be started unless the temperature of the

coolant between the operating and idle Recirc loops are within 50°F of

each other. This 50°F delta T limit is based on stress analysis for reactor

nozzles, stress analysis for reactor recirculation components and piping,

and fuel thermal limits.

[GE Sll517 Supplement 1]

11. The out of service pump may NOT be started unless the reactor is verified

outside of regions 1, 2 and 3 of the Unit 3 Power to Flow Map (ICS or

Station Reactor Engineering, 0-TI-248).

12. The temperature of the coolant between the dome and the idle Recirc loop

should be maintained within 75°F of each other. If this limit cannot be

maintained a plant cooldown should be initiated. Failure to maintain this

limit and NOT cooldown could result in hangers and/or shock suppressers

exceeding their maximum travel range.

[GE SIl251, 430 and 517]

M.

Recirc Pump controller limits are as follows:

1.

When any individual RFP flow is less than 19% and reactor water level is

below 27 inches, speed limit is set to 75%(-1130 RPM speed) and if speed

is greater than 75%(-1130 RPM speed), Recirc speed will run back to

75%(-1130 RPM speed).

2.

When total feed water flow is less than 19% (15 sec TD) or Recirc Pump

discharge valve is less than 90% open, speed limit is set to 28%

(-480 RPM speed) and if speed is greater than 28%(-480 RPM speed),

Recirc speed will run back to 28%(-480 RPM speed).

(

Reactor Recirculation System

3-01-68

Unit 3

Rev. 0066

Page 15 of 179

3.0

PRECAUTIONS AND LIMITATIONS (continued)

R.

The power supplies to the MMR and DFR relays are listed below.

VFD3A

I&C BUS A (BKR 215)

ICS PNL 532 (BKR 30)

UNIT PFD (BKR 615)

VFD3B

I&C BUS B (BKR 315)

ICS PNL 532 (BKR 26)

UNIT PFD (BKR 616)

3-RLY-068-MMR3/A & DFR3/A

3-RLY-068-MMR2/A & DFR2/A

3-RLY-068-MMR1/A & DFR1/A

3-RLY-068-MMR3/B & DFR3/B

3-RLY-068-MMR2/B & DFR2/B

3-RLY-068-MMR1/B & DFR1/B

(

S.

A complete list of Recirc System trip functions is provided in Illustration 4. The

RPT breakers between the recirc drives and pump motors will open on any of

the following:

1.

Reactor dome Pressure ~ 1148 psig (ATWS/RPT). (Both pressure

switches in Logic A or both pressure switches in Logic B will cause RPT

breakers to trip both pumps.) (2 out of 2 taken once logic)

2.

Reactor Water Level s -45" (ATWS/RPT) . (Both level switches in Logic A

or both level switches in Level B will cause RPT breakers to trip both

pumps.) (2 out of 2 taken once logic)

3.

Turbine trip or load reject condition, when ~ 30% power by turbine first

stage pressure (EOC/RPT) .

1.

The ATWS/RPT A(B) logic to trip the RPT breakers is defeated if the

ATWS/RPT/ARI A(B) manual logic is armed using the arming collar on

Panel 3-9-5. B(A) logic would still be functional and trip the RPT breakers if the

setpoints are reached. If both manual push-buttons on 3-9-5 are armed,

ATWS/RPT automatic logic is totally defeated (no RPT breaker trip will occur if

the ATWS/RPT trip setpoints are reached). EOC/RPT logic and ATWS/ARI

logic will function without regard to the position of the arming collars.

ATWS/R PT/ARI logic can be reset 30 seconds after setpoints are reset.

(

30. RO 215001Al.Ol OOlIMEMlTIG2/TIPI121500IAl.Ol//RO/SROI

Which ONE of the following describes the procedural requirements in accordance with 2-01-94,

Traversing In-Core Probe System while running TIP traces?

A.

The TIP detector shall be withdrawn to the In-Shield position and the ball valve closed following

each TIP trace.

8.

Running a TIP trace while personnel are working inside the Drywell is prohibited.

C." The Radiation Protection Shift Supervisor is required to be notified prior to TIP System operation.

D.

The TIP Machine will automatically withdraw to the in-shield position, then the ball valve will

automatically close following a PCIS Group 6 isolation.

KIA Statement:

215001 Traversing In-core Probe

A1.01 - Ability to predict and/or monitor changes in parameters associated with operating the

TRAVERSING IN-CORE PROBE controls including: Radiation levels: (Not-BWR1)

KIA Justification: This question satisfies the KIA statement by requiring the candidate to determine the

operating limitations of the TIP system with respect to high radiation .

References:

2-01-94 Precautions & Limitations

Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall

or recognize discrete bits of information.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following :

1. Limitations for running TIP traces with personnel in the Drywell.

2. Notification requirements prior to running TIPs.

3. Which PCIS Group will cause a TIP retraction and isolation.

4. Requirements for running multiple simultaneous TIP traces.

A is incorrect. This is plausible because that limitation is placed on TIP operation, but only when TIP

operation is no longer required. The TIP detector can be stored in the Indexer in-between traces using

the same TIP Machine for ALARA concerns.

8 is incorrect. This is plausible because specific permission and controls are required to allow this

condition, but it is allowable.

C is correct.

D is incorrect. This is plausible because the TIP response to a PCIS isolation is correct, but it is not a

Group 6 isolation.

(

Traversing Incore Probe System

2-01-94

Unit2

Rev. 0029

Page 7 of 26

3.0

PRECAUTIONS AND LIMITATIONS

A.

[NER/C] Verification of a digit in CORE LIMIT and DETECTOR POSITION

windows prior to or during TIP insertion ensures TIPs retain the ability to

determine its proper position. This will prevent malfunctions which could

damage the TIP detector.

[GE SIL-166]

B.

To prevent accidental exposure to personnel , immediately evacuate the area if

the TIP drive area radiation monitor alarms.

C.

[NER/C] Always observe READY light illuminated prior to inserting detector.

[GE

SIL-166]

D.

(NERlC] DO NOT move CHANNEL SELECT switch with detector inserted past

Indexer position (0001). The common channel interlock can be defeated in this

manner resulting in detector and equipment damage.

[GE SIL-092]

E.

(NERlC] Should detector fail to shift to slow speed when it enters the core, the

LOW switch should be turned on, switched to manual mode, and the detector

withdrawn.

[GE SIL-166]

F.

[NER/C] Length of time detector is left in core should be minimized to limit

activation of detector and cable.

[GE SIL-166]

G.

(NERlC] When TIP System operation is not desired, detectors should be retracted

and stored in chamber shield with ball valves closed .

[GE SIL-166] Storage of

detector in Indexer (0001) is allowed only for ALARA concerns and to prevent

unnecessary masking of multiple inputs to annunciator RX BLDG AREA

RADIATION HIGH 2-RA-90-1 D (2-XA-55-3A, Window 22).

. H.

[NER/C] Upon receipt of a PCIS signal (low reactor water level or high drywell

pressure), any detector inserted beyond its shield chamber should be verified to

automatically shift to reverse mode and begin withdrawal. Once in shield, ball

and purge valves close.

[GE SIL-166] Ball valve cannot be reopened until PCIS is

reset on Panel 2-9-4 and manual reset of TIP ISOLATION RESET pushbutton

2-HS-94-7D/S2 located on Panel 2-9-13.

I.

A detector should not be abruptly stopped from fast speed to off without first

switching to slow speed.

J.

[NER/C] Drive Control Units (DCU) should be monitored during withdrawal to

prevent any chamber shield withdrawal limit from being overrun. Detectors

should be stopped manually at shield limit if auto stop limit switch should fail

and verify ball valve closes.

[GE SIL-166]

K.

Only one TIP at a time should be operated when maintenance is being

performed in TIP drive area.

(

l

Traversing Incore Probe System

2-01-94

Unit2

Rev. 0029

Page 8 of 26

3.0

PRECAUTIONS AND LIMITATIONS (continued)

L.

[NRC/CJ DO NOT operate TIPswith personnel inside TIP Room or in vicinity of

TIP tubing and Indexers in Drywell. Requirement may be waived with approval

of Shift Manager and site RADCON manager or designee. In this instance,

RADCON is required to establish such controls as are necessary to prevent

access to TIP tubing and Indexer areas to preclude unnecessary exposure to

personnel working in Drywell. RADCON Field Operations Shift Supervisor is

required to be notified prior to operation of TIP System.

[NRC InformationNotice88-063,

Supplement2J

M.

No channel should be indexed to common channel 10 unless all other channels

are not indexed to channel 10 and all their READY lights are illuminated.

N.

[NERlC] DO NOT turn MODE switch to OFF on Drive Control Unit if detector is

outside shield chamber unless personnel safety requires it. [GE SIL-166J This

removes power preventing automatic withdrawal on PCIS signal and causing

ball valves to close on cable or detector. Tip Ball Valves CANNOT fully close

and shear valves may have to be actuated.

O.

CHANNEL SELECT switches on Drive Control Units should always be rotated

in clockwise direction when selecting channels.

P.

Connector on shear valve indicator circuit should not be removed while testing

shear valve explosive charges or performing shear valve maintenance with

detector inserted. This will cause an automatic detector withdrawal.

Q .

Continuous voice communication should be maintained between TIP operator

or maintenance personnel in control room and drive mechanism area while

maintenance is being performed and TIP detector driving is necessary.

R.

Each applicable ball valve should be opened prior to operating that TIP

machine.

S.

TIP Drive Mechanisms and Indexers should have continuous purge supply

unless required to be removed from service for maintenance.

T.

During outages when containment is deinerted for personnel access, TIP

Indexer purge supply should be transferred from nitrogen to Control Air for

personnel safety.

U.

Detector damage is possible if TIP ball valve is left open, or is opened during

DRYWELL PRESSURE TEST. (GE SIL-166)

(

31. RO 216000Kl.l O00l/MEM/T2G2/PR.INSTRJ9/216000Kl.lO//RO/SRO/

Which ONE of the following indicates how raising recirculation flow affects the Emergency System Range

indicators (3-58A -58B) and Narrow Range Indicators (e.g., L1-3-53) on Panel 9-5?

A.

No effect on Emergency System Range; Narrow Range will indicate higher.

B.

Emergency System Range will indicate higher; Narrow Range will not be affected.

C.

Both Emergency System Range and Narrow Range will indicate lower.

D.oI Emergency System Range will indicate lower and Narrow Range will not be affected.

KIA Statement:

216000 Nuclear Boiler Inst

K1.10 - Knowledge of the physical connections and/or cause- effect relationships between NUCLEAR

BOILER INSTRUMENTATION and the following : Recirculation flow control system

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

knowledge of the effect of changes in Recirculation flow on reactor water level instrumentation.

References:

OPL171.003

Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall

or recognize discrete bits of information.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the effect of raising Recirc flow on

Normal Range and Emergency Systems Range level instrumentation.

A is incorrect. This is plausible because Narrow Range instruments may read slightly higher at colder

conditions, but this does NOT apply to Recirc flow changes.

B is incorrect. This is plausible because Narrow Range instruments are not effected by Recirc Flow

changes, but Emergency System Range isntruments will read lower.

C is incorrect. This is plausible because Emergency System Range instruments will read lower, but the

Narrow Range instruments will not.

D is correct.

(

d.

Four ranges of level indication

OPL 171.003

Revision 17

Page 20 of 54

INSTRUCTOR NOTES

Normal Control Range (Narrow Range)

(1)

(a)

oto +60 inch range covering the

normal operating range (analog) with

+60" up to +70" digital and 0" down to

- 10" digital readings.

Obj. V.B.5

Obj. V.B.6

TP-3 shows only

analog scale

(b)

Referenced to instrument zero

(c)

Four of these instruments are

used by Feedwater Level Control

System (FWLCS). The level

signal utilized by the FWLCS is

not directed through the Analog

Trip System.

i.

Temperature

compensated by a

pressure signal

Obj. V.B.11.

Obj. V.B.13.

(

ii.

Most accurate level

indication available to the

operator

iii.

Calibrated for normal

operating pressure and

temperature

(d)

These indicators and a recorder

point (average of the four) are

located on Panel 9-5.

NOTE: An air bubble or leak in

the reference leg can cause

inaccurate readings in a non-

conservative direction resulting in

a mismatch between level

indicators.

This problem is particularly

prevalent after extended outages

when starting up from cold

shutdown conditions and at low

reactor pressures.

LER 85-006-02

(See LP Folder)

(Section X.C.1.j.

provides more

detail)

(

(e)

Four other narrow range

instruments are located in the

control room, two above the

FWLCS level indicators on panel 9-5 (3-208A & D), one above

HPCI (3-208B)and one above

RCIC (3-208C)on panel 9-3.

OPL171 .003

Revision 17

Page 21 of 54

INSTRUCTOR NOTES

Associated with

RFPT/Main Turbine

and HPCIIRCIC trip

instruments

(2)

Emergency Systems Range (Wide Range) 2 Analog meters

and 2 Digital meters .

(a)

-155 to +60 inches range

covering normal operating range

and down to the lower instrument

nozzle return

(b)

Referenced to instrument zero

(c)

Four MCR indicators on Panel 9-

5 monitor this range of level

indication.

(d)

Calibrated for normal operating

pressure and temperature

(e)

The level signal utilized by the

Wide Range instruments have

safety related functions and are

directed through the Analog Trip

System.

(f)

Level indication for this range is

Obj. V.B.12.

also provided on the Backup

Control Panel (25-32).

(3)

Shutdown Vessel Flood Range (Flood-up

Range)

(a)

oto +400 inches range covering

upper portion of reactor vessel

(b)

Referenced to instrument zero

Calibrated for cold conditions

<<212°F, 0 psig)

(c)

Provides level indication during

vessel flooding or cool down.

(

Transient flashing effects can cause

indicated level to oscillate or be

erratic. As the reference leg refills,

the indicated level approaches a

more accurate water level indication .

The RVLlS mod decreases the time

necessary for this refill to occur

j.

Normal Control Range (Narrow Range) and

Emergency Systems Range (Wide Range) Level

Discrepancies

(1)

Narrow Range level instrumentation is

calibrated to be most accurate at rated

temperature and pressure (particularly

the instruments for FWLCS, since they

are temperature compensated). At cold

conditions the non-FWLCS instruments

read high (not temperature

compensated).

(2)

Wide Range instruments are also

calibrated for rated temperature and

pressure

OPL171.003

Revision 17

Page 32 of 54

INSTRUCTOR NOTES

(a)

The indicated level on the Wide

Range (9-5) is also affected by

changes in the subcooling of

recirculation water and the

amount of flow at the lower

(variable leg) tap.

Obj. V.B.15

(b)

At rated conditions with

minimum recirculation flow the

Wide Range instruments are

accurate. As recirculation flow is

increased past the lower tap it

has a significant velocity head

and some friction loss which

reduces the pressure on the

variable leg to the differential

pressure instrument, resulting in

an indicated level lower than

actual. This could be as much

as 10-15 inches error when at

rated flow and power.

(c)

Due to calibration for rated

conditions and no density

compensation at cold conditions

these instruments read high.

(

32. RO 219000K2.02 00l/C/A/T2G2/0I-74//219000K2.02//RO/SRO/NEW 10/16/07

Given the following plant conditions:

Unit-2 is at 100% rated power with RHR Loop II in Suppression Pool Cooling mode to support

a HPCI Full Flow test surveillance.

Unit-1 experiences a LOCA which results in a CAS signal initiation on Unit-1.

Which ONE of the following describes the current status of Unit-2 RHR system and what actions must be

taken to restore Suppression Pool Cooling on Unit-2?

A.

2A and 2C RHR Pumps are tripped. 28 and 2D pumps are unaffected . No additional action is

required.

B. 28 and 2D RHR Pumps are tripped. 2A and 2C pumps are unaffected. Place RHR Loop I in

Suppression Pool Cooling immediately.

c. All four RHR pumps receive a trip signal. Place RHR Loop II in Suppression Pool Cooling

immediately.

D~ All four RHR pumps receive a trip signal. Place RHR Loop II in Suppression Pool Cooling after a 60

second time delay.

KIA Statement:

219000 RHR/LPCI: Torus/Pool Cooling Mode

K2.02 - Knowledge of electrical power supplies to the following: Pumps

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions and times to determine which RHR pumps can be used for Suppression Pool Cooling.

References: 2-01-74, OPL 171.044

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble ,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

(

In order to answer this question correctly the candidate must determine the following:

1. Response of Unit-2 RHR pumps due to a Unit 1 CAS initition.

2. Recognize the difference between a Single Unit CAS and Simultaneous Unit CAS.

3. Recognize that Preferred and Non-preferred ECCS pumps do NOT apply with the given conditions.

A is incorrect. This is plausible based on RHR Loop II being the Preferred pumps for Unit-2.

B is incorrect. This is plausible if taken from the perspective of Unit 1 operation, not Unit 2 operation.

C is incorrect. This is plausible because all four RHR pumps on Unit 2 will trip, but they are locked out

from manual start for 60 seconds based on D/G and/or Shutdown Board loading concerns.

D is correct.

(

Residual Heat Removal System

2-01-74

Unit2

Rev. 0133

Page 331 of 367

Appendix A

(Page 2 of 7)

Unit 1 & 2 Core Spray/RHR Logic Discussion

2.2

ECCS Preferred Pump Logic

Concurrent Accident Signals On Unit 1 and Unit 2

With normal power available, the starting and running of RHR pumps on a 4KV

Shutdown Board already loaded by the opposite unit's Core Spray, RHR pumps, and

RHRSW pumps could overload the affected 4KV Shutdown Boards and trip the

normal feeder breaker. This would result in a temporary loss of power to the

affected 4KV Shutdown Boards while the boards are being transferred to their

diesels. To prevent this undesirable transient, Unit 2 RHR Pumps 2A and 2C are

load shed on a Unit 1 accident signal and Unit 1 Pumps 1Band 10 will be load shed

on a Unit 2 accident signal. Unit 2 Core Spray Pumps 2A and 2C are load shed on a

Unit 1 accident signal and Unit 1 Core Spray Pumps 1Band 10 will be load shed on

a Unit 2 accident signal. This makes the Preferred ECCS pumps Unit 1 Division I

Core Spray and RHR Pumps and Unit 2 Division 2 Core Spray and RHR Pumps.

Conversely, the Non-preferred ECCS pumps are Unit 1 Division 2 Core Spray and

RHR Pumps and Unit 2 Division 1 Core Spray and RHR Pumps.

The preferred and non-preferred ECCS pumps are as follows:

UNIT 1 & 2

PREFERRED ECCS Pumps

CS1A,CS1C,RHR1A,RHR1C

CS 2B, CS 20, RHR 2B, RHR 20

NON-PREFERRED ECCS Pumps

CS 1B, CS 10, RHR 1B, RHR 10

CS2~CS2C,RHR2A,RHR2C

UNIT3

Unit 3 does not have ECCS Preferred/Non-Preferred Pump Logic.

Accident Signal On One Unit

With an accident on one unit, ECCS Preferred pump logic trips all running RHR and

Core Spray pumps on the non-accident unit.

(

OPL171.044

Revision 15

Page 50 of 159

INSTRUCTOR NOTES

Note:

Presently Unit 1 Accident signal will not affect Unit 2 due to DCN H2735A that lifted wires

from relays. Unit 2 will still affect Unit 1.

However, the following represents modifications

to the inter-tie logic as it will be upon Unit 1 recovery.

(

f.

(1)

Unit 1 Preferred RHR pumps are 1A and 1C

(2)

Unit 2 Preferred RHR pumps are 28 and 2D

(3)

Unit 2 initiation logic is as follows:Div 1 RHR

logic initiates Div 1 pumps ( A and C), and Div

2 logic initiates Div 2 pumps (B and D)

Accident Signal

(1)

LOCA signals are divided into two separate

signals, one referred to as a Pre Accident

Signal (PAS) and the other referred to as a

Common Accident Signal (CAS).

PAS

-122" Rx water level (Level 1)

OR

2.45 psig DW pressure

CAS

-122" Rx water level (Level 1)

OR

2.45 psig DW pressure AND <450

psig Rx pressure

(2)

If a unit receives an accident signal, then all

its respective RHR and Core Spray pumps

will sequence on based upon power source to

the SD Boards.

(3)

All RHR and Core Spray pumps on the non-

affected unit will trip (if running) and will be

blocked from manual starting for 60 seconds.

Obj. V.B.13.

Obj. V.C.3

Obj. V.C.7

Obj. V.D.6

Obj. V.E.II

Obj. V.B.13.

Obj. V.C.3

Obj. V.C.7

Obj. V.D.6

Obj. V.E.II

Note:

It should be clear

that the only

difference

between the two

signals is the

inclusion of Rx

pressure in the

CAS signal. The

PAS signal is an

anticipatory signal

that allows the

DG's to start on

rising OW

pressure and be

ready should a

CAS be received.

OPL171.044

Revision 15

Page 51 of 159

(

INSTRUCTOR NOTES

(4)

After 60 seconds all RHR pumps on the non-

Operator diligence

affected unit may be manually started.

required to

(5)

The non-preferred pumps on the non-

prevent

overloading SO

affected unit are also prevented from

boards/DG's

automatically starting until the affected unit's

accident signal is clear.

(6)

The preferred pumps on the non-affected

unit are locked out from automatically starting

until the affected unit accident signal is clear

OR the non-affected unit receives an

accident signal.

g.

4KV Shutdown Board Load Shed

Obj. V.C .B.

(1)

A stripping of motor loads on the 4KV boards

occurs when the board experiences an

undervoltage condition. This is referred to as a

4KV Load Shed. This shed prepares the board

for the DG ensuring the DG will tie on to the

bus unloaded and without faults.

(2)

The Load Shed occurs when an undervoltage

is experienced on the board i.e. or if the Diesel

were tied to the board (only source) and one of

the units experienced an accident signal which

trips the Diesel output breaker.

(3)

Then, when the Diesel output breaker

interlocks are satisfied, the DG output breaker

would close and, if an initiation signal is

present (CAS) the RHR, CS, and RHRSW

pumps would sequence on

(4)

Following an initiation of a Common Accident

Signal (which trips the diesel breaker), if a

subsequent accident signal is received from

another unit, a second diesel breaker trip on a

"unit priority" basis is provided to ensure that

the Shutdown boards are stripped prior to

starting the RHR pumps and other ECCS

loads

(5)

When an accident signal trip of the diesel

Occurs due to

breakers is initiated from one unit (CASA or

actuation of the

(

CASB), subsequent CAS trips of all eight

diesel breaker

diesel breakers are blocked.

TSCRN relay

(

33. RO 226001A4.I2 OOlIMEM/T2G2/PC/P//226001A4.12/3.8/3.9/RO/SRO/

Given the following plant conditions:

A pipe break inside containment results in the below parameters:

- Drywell pressure is 20 psig

- Drywell temperature is 210°F

- Suppression chamber pressure is 18 psig.

- Suppression chamber temperature is 155°F.

- Suppression pool level is +2 inches

- Reactor water level is +30 inches

Which ONE list of parameters below must ALWAYS be addressed to determine when it is appropriate to

spray the drywell?

A.

-Suppression Chamber temperature

-Drywell pressure

-Drywell temperature

B.

-Suppression Chamber pressure

-Drywell temperature

-Suppression Pool level

C." -Drywell pressure

-Drywell temperature

-Reactor water level

D.

-Reactor water level

-Suppression Chamber temperature

-Drywell pressure

KIA Statement:

226001 RHR/LPCI: CTMT Spray Mode

A4.12 - Ability to manually operate and/or monitor in the control room: ContainmenUdrywell pressure

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

knowledge of which containment parameters are used to determine when Containmerit Sprays can be

used.

References: 1/2/3-EOI-2 Flowchart

Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall

or recognize discrete bits of information.

0610 NRC Exam

(

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following:

1. Orywell temperature and pressure are always required to ensure Curve 5 limits are not exceeded.

2. RPV level is always required to verify adequate core cooling is assured prior to diverting RHR flow

for Orywell sprays.

3. Suppression Pool level is always required to verify Suppression Chamber to Orywell vacuum breakers

are uncovered.

4. Suppression Chamber pressure is ONLY required when initiating Orywell Sprays from flowpath PC/Po

5. Suppression Chamber temperature is NOT required to initiate Orywell Sprays.

A is incorrect. This is plausible because OW temp and press are required , but SC temp is not.

B is incorrect. This is plausible because OW temp and SP level are required , but SC press is ONLY

required when initiating OW Sprays using PC/Po

C is correct.

D is incorrect. This is plausible because RPV level and OW press are required, but SC temp is not.

WHEN

SUPPR CHMBR PRESS EXCEEDS 12 PSIG,

THEN

CONnNUE INTHISPROCEDURE

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-_..._....----_.....__.__.._---------_...., ..

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~'.

PClP-7

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SHUT DOWNRECIRC PUfA'PS ANDOWBLOWERS

2

PUMP NPSH AND VORTEX m"TS

INITlAm r:JN SPRAYS USING W:lL:!PUMPSWIREQUJRED

ro ASSUREAIEQUATE OORE COOLING BY CON11NUOUS

INJ(APPX 178)

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!:!

~

"

,p'

0"

..,J~"~

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SHUT DOWN RSCIRC i'IIllWS RJO r:1"BLO'/IB'tS

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34. RO 234000G2.4.50 OO l/C/NTIG2///234000G2.4.50/IRO/SRO/

Given the following plant conditions:

Fuel movement is in progress for channel changeout activities in the Fuel Prep Machine.

Gas bubbles are visible coming from the de-channeled bundle.

An Area Radiation Monitor adjacent to the SFSP begins alarming.

Which ONE of the following describes the action (s) to take?

Immediately STOP fuel handling, then

_

A.

notify RADCON to monitor & evaluate radiation levels.

B."

evacuate non-essential personnel from the RFF.

C.

evacuate ALL personnel from the RFF.

D.

obtain Reactor Engineering Supervisor's recommendation for movement and sipping of the

damaged fuel assembly.

KIA Statement:

234000 Fuel Handling Equipment

2.4.50 - Emergency Procedures / Plan Ability to verify system alarm setpoints and operate controls

identified in the alarm response manual

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions to determine the corrective actions involving Fuel Handling equipment under emergency

conditions.

References:

1/2/3-AOI-79-1 & 79-2, 1/2/3-ARP-9-3A (W1)

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble ,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

0610 NRC Exam

(

REFERENCE PROVIDED: None

Plausibility Analvsis:

In order to answer this question correctly the candidate must determine the following :

1. Whether indications are consistent with fuel damage or inadvertant criticality.

2. Based on the answer to Item 1 above, enter the appropriate AOI.

3. Immediate Operator Actions for the selected procedure, AOI-70-1.

A is incorrect. This is plausible because RADCON notification is a subsequent action in AOI-70-1,

however non-essential personnel evacuation is an IMMEDIATE action.

B is correct.

C is incorrect. This is plausible because evacuation of ALL personnel is an IMMEDIATE action in

AOI-70-2 , however non-essential personnel evacuation is an IMMEDIATE action in the appropriate AOI.

D is incorrect. This is plausible because RE recommendations are a subsequent action in AOI-70-1,

however non-essential personnel evacuation is an IMMEDIATE action.

2-ARP-9-3A

(

Unit2

2-XA-55-3A

Rev. 0036

Page 4 of 50

FUEL POOL

SensorlTrip Point:

FLOOR AREA

RADIATION HIGH

RI-90-1B

RI-90-2B

For setpoints

2-RA-90-1A

RI-90-3B

REFER TO 2-SIMI-90B.

11

(Page 1 of 1)

Sensor

RE-90-1B

EI664'

R-11 P-L1NE

Location:

RE-90-2B

E1664'

R-10 U-L1NE

RE-90-3B

E1639'

R-10 Q-L1NE

Probable

Cause:

Automatic

Action:

Operator

Action:

References:

A. Change in general radiation levels.

B. Refueling accident.

C. Sensor malfunction.

None

A.

CHECK 2-RI-90-1A, 2-RI-90-2A and 2-RI-90-3A on Panel 2-9-11.

B. NOTIFY refuel floor personnel.

C. IF Dry Cask loading/unloading activities are in progress, THEN

NOTIFY Cask Supervisor.

D. IF airborne levels rise by 100 DAC AND RADCON confirms, THEN

REFER TO EPIP-1.

E. REFER TO 2-AOI-79-1 or 2-AOI-79-2 as applicable.

F. IF this alarm is not valid, THEN REFER TO 0-01-55.

G. IF this alarm is valid, THEN

MONITOR the other parameters that input to it frequently. These

other parameters will be masked from alarming while this alarm is

sealed in.

H. ENTER 2-EOI-3 Flowchart.

0-47E600-13

2-47E610-90-1

2-45E620-3

GE 730E356 Series, TVA Calc NDQ00902005001/EDC63693

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Fuel Damage During Refueling

2-AOI-79-1

Unit 2

Rev. 0017

Page 3 of7

1.0

PURPOSE

This instruction provides the symptoms, automatic actions and operator actions for a

fuel damage accident.

2.0

SYMPTOMS

A.

Possible annunciators in alarm:

1.

FUEL POOL FLOOR AREA RADIATION HIGH (2-XA-55-3A, window 1).

2.

AIR PARTICULATE MONITOR RADIATION HIGH (2-XA-55-3A,

window 2).

3.

RX BLDG, TURB BLDG, RF ZONE EXH RADIATION HIGH (2-XA-55-3A,

window 4).

4.

REACTOR ZONE EXHAUST RADIATION HIGH (2-XA-55-3A, window 21).

5.

RX BLDG AREA RADIATION HIGH (2-XA-55-3A, window 22).

6.

REFUELING ZONE EXHAUST RADIATION HIGH (2-XA-55-3A,

window 34).

B.

Gas bubbles visible, in the Spent Fuel Storage Pool and/or Reactor Cavity,

attributed to physical fuel damage.

C.

Known dropped or physically damaged fuel bundle.

D.

Portable CAM in alarm.

E.

Radiation level on the Refuel Floor is greater than 25 mr/hr and cause is

unknown.

Fuel Damage During Refueling

2-AOI-79-1

Unit2

Rev. 0017

Page 5 of 7

4.0

OPERATOR ACTIONS

4.1

Immediate Actions

[1]

STOP all fuel handling.

[2]

EVACUATE all non-essential personnel from Refuel Floor.

4.2

Subsequent Actions

CAUTION

o

The release of iodine is of major concern. If gas bubbles are identified at any time, Iodine

release should be assumed until RADCON determines otherwise.

[1]

VERIFY secondary containment is intact.

(REFER TO Tech Spec 3.6.4.1)

[2]

IF any EOI entry condition is met, THEN

ENTER the appropriate EOI(s).

[3]

VERIFY automatic actions.

[4]

NOTIFY RADCON to perform the following:

n

o

EVALUATE the radiation levels.

0

MAKE recommendation for personnel access.

0

MONITOR around the Reactor Building Equipment Hatch,

at levels below the Refuel Floor, for possible spread of the

release.

0

[5]

REFER TO EPIP-1 for proper notification.

o

(

Fuel Damage During Refueling

2-AOI-79-1

Unit 2

Rev. 0017

Page 6 of 7

4.2

Subsequent Actions (continued)

[6]

MONITOR radiation levels, for the affected areas, using the

following radiation recorders and indicators:

A.

2-RR-90-1 (points 1 and 2), 2-MON-90-50 (Address 11),

2-RR-90-142 and 2-RR-90-140 (Panel 2-9-2) .

0

B.

2-RM-90-142, 2-RM-90-140, 2-RM-90-143

and 2-RM-90-141 Detectors A and B (Panel 2-9-10).

0

C.

2-RI-90-1A and 2-RI-90-2A (Panel 2-9-11).

0

D.

0-CONS-90-362A (Address 09, 10, 08) for Unit 1, 2,

3-RM-90-250, respectively (Panel 1-9-44).

0

[7]

IF possible, MONITOR portable CAMs &ARMs.

[8]

REQUEST Chemistry to perform 0-SI-4.8.8.2-1 to determine if

iodine concentration has risen.

0

[9]

NOTIFY Reactor Engineering Supervisor, or his designee, and

OBTAIN recommendation for movement and sipping of the

damaged fuel assembly.

0

[10]

OBTAIN Plant Managers approval prior to resuming any fuel

transfer operations.

0

[11]

WHEN condition has cleared AND if required, THEN

RETURN ventilation systems, including SGTS, to normal.

REFER TO 2-01-30A, 2-01-30B, 0-01-30F, 0-01-31,

and 0-01-65.

0

(