ML081370205

Residual Heat Removal System

2-01-74

Unit 2

Rev. 0133

Page 23 of 367

3.5

INTERLOCKS (continued)

7.

The RHR spray/cooling valves, 2-FCV-74-57(71), receive an auto closure

signal in the presence of a LPCI initiation signal and they are interlocked to

prevent opening if the in-line torus spray valve, 2-FCV-74-58(72), is not

fully closed. The in-line valve interlock can be by-passed if the following

conditions exist.

(1)

Reactor level is >2/3 core height and a LPCI initiation signal is

present and the select reset switch is in the SELECT position.

The requirements for >2/3 core height and a LPCI initiation signal

may be by-passed using the keylock bypass switch,

2-XS-74-122/30.

8.

If primary containment cooling is desired with reactor level at <2/3 core

height, the keylock bypass switch is required to be placed in BYPASS

before the select reset switch is placed in SELECT to ensure relay logic is

made up.

9.

The RHR torus spray valves, 2-FCV-74-58(72), have the same in-line valve

interlocks as those outlined in Step 3.5A.8 for the torus spray/cooling

valves. Additionally these valves have an interlock preventing opening

unless drywell pressure is ~1.96 psig which cannot be bypassed .

10. The RHR torus cooling/test valves, 2-FCV-74-59(73), receive an auto

closure signal in the presence of a LPCI initiation signal. Auto closure may

be bypassed by the same conditions/actions outlined in Step 3.5A.8.

11. The RHR containment spray valves, 2-FCV-74-60(74) and 61(75), have

in-line valve interlocks similar to these described in Step 3.5A.8

through 3.5A.1afor the RHR torus spray valves 2-FCV-74-57(58)

and 71(72).

12. If 2-FCV-74-59(73) LOCA CLOSURE TIME light (2-IL-74-59Y Loop I;

2-IL-74-73Y Loop II) on Panel 2-9-3 is extinguished due to its associated

valve being opened, that Loop is inoperable for LPCI.

13. If 2-HS-74-148(149) RHR SYSTEM I (II) MIN FLOW INHIBIT switch is in

the INHIBIT position, the pumps on that loop do not have automatic

minimum flow protection.

(

2-EOI APPENDIX-17B

Rev . 10

Page 2 of 13

6.

INITIATE Drywell Sprays as follows:

a.

VERIFY at least one RHRSW pump supplying each EECW

header.

b .

IF . ... . EITHER of the following exists:

LPCI Initiation signal is NOT present,

OR

Directed by SRO,

THEN ... PLACE keylock switch 2-XS-74-122(130),

RHR

SYS 1(11)

LPCI 2/3 CORE HEIGHT OVRD,

in

MANUAL OVERRIDE.

c.

MOMENTARILY PLACE 2-XS-74-121(129),

RHR SYS 1(11)

CTMT SPRAY/CLG VLV SELECT,

switch in SELECT.

d.

IF ..... 2-FCV-74-53(67),

RHR SYS 1(11)

LPCI

INBD

INJECT VALVE, is OPEN,

THEN ... VERIFY CLOSED 2-FCV-74 -52(66),

RHR SYS 1(11)

LPCI OUTBD INJECT VALVE.

e.

VERIFY OPERATING the desired System 1(11)

RHR

pump(s)

for Drywell Spray.

f.

OPEN the following valves:

2-FCV-74-60(74),

RHR SYS 1(11)

DW SPRAY OUTBD VLV

2-FCV-74-61(75),

RHR SYS 1(11)

DW SPRAY INBD VLV.

g .

VERIFY CLOSED 2-FCV-74-7(30),

RHR SYSTEM 1(11)

MIN

FLOW VALVE.

h.

IF

Additional Drywell Spray flow is necessary,

THEN

PLACE the second System 1(11)

RHR Pump in

service.

i.

MONITOR RHR Pump NPSH using Attachment 2.

j.

VERIFY RHRSW pump supplying desired RHR Heat

Exchanger(s) .

(

2. RO 205000K4.02 OOIIMEMISYS/RHR//205000K4.02//RO/SRO/ll/27/07 RMS

Given the following conditions on Unit 2:

Reactor level +20"

Reactor pressure 90 psig

Drywell pressure 1.7 psig

Which ONE of the following describes which modes of RHR are available for use (consider interlocks

only)?

A.

LPCI, Drywell Sprays , Shutdown Cooling

B.

Suppression Pool Sprays, Shutdown Cooling, Suppression Pool Cooling

C. II LPCI, Suppression Pool Cooling, Shutdown Cooling

D.

LPCI, Supplemental Fuel Pool Cooling, Drywell Sprays

KiA Statement:

205000 Shutdown Cooling

K4.02 - Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) design

feature(s) and/or interlocks which provide for the following: High pressure isolation: Plant-Specific

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

plant conditions to determine which interlocks apply to those conditions including the High Pressure

isolation of RHR SDC mode.

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:

Plausibility Analysis:

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

1. Drywell pressure < 1.96 psig prohibit use of containment sprays (DWand SP).

2. RPV Pressure <450 psig allow the use of LPCI Injection.

3. RPV pressure of 90 psig allows the use of Shutdown Cooling.

NOTE: All given answers are plausible since they all contain at least one acceptable lineup with the given

conditions.

A is incorrect. Drywell Sprays will not function < 1.96 psig in the Drywel l.

B is incorrect. Suppression Pool Sprays will not function < 1.96 psig in the Drywell.

C is correct.

D is incorrect. Supplemental Fuel Pool Cooling will not function based on RPV pressure and Drywell

Sprays will not function < 1.96 psig in the Drywell.

(

(3) .: Can be opened when Rx press is ~ 450 psiq.

(4) . Automatically opens on LPCI initiation signal

when reactor pressure is < 450 psiq.

(5)

LPCI Injection Valve Open Signal Bypass

Switch (Keylock switch on 9-3) can be utilized

to bypass the open signal during execution of

EOl's. Allows operator to manually (pnl 9-3)

close the injection valve

(6)

The normally open outboard injection valves

(1-FCV-74-52,74-66;2-FCV-74-52, 74-66; 3-74-

. 66) have added circuitry so that a fire cannot

energize the closing coil and shut the valve

(any close signal with the Control Room

handswitch in NORMAL. Shorts out the

closing coil and blows the control power fuses).

Modifications also disabled the local control

"Close" pushbutton on 1-FCV-74-52/66, 2-

FCV-74-52/66 and 3-FCV-74-66.

(7)

Control Circuit

OPL171.044

Revision 15

Page 40 of 159

INSTRUCTOR NOTES

NOTE: Effect of

Logic failure and

valve operation

NEW!

1OA-S155A(B)

Unit 1 and 2

ONLY at this time

Indicating light

informs operators

when open signal

logic is bypassed.

1-74-52

1-74-66

2-74-52

2-74-66

3-74-52

3-74-66

Operate Interlock

74-53

74-67

74-53

74-67

74-53

74-67

Outgoing interlock

74-53

74-67

74-53

74-67

74-53

74-67

Normal/Emero Sw

X

Local Controls

X

Controls at Bkr

X

Panel 9-3 controls

X

Local Indication

X

Igts

Lights on Bkr

X

LiQhts on Pnl 9-3

X

k.

LPCI inboard injection valves

(1)

Normally closed - non-throttling

(74-53; 74-67)

Obj. V.C.5.

TP-29, 30, and

31

(

(2)

Interlock prevents normal opening unless in-

line valve (74-52; 74-66) is fully closed with

reactor pressure> 450 psig

Operation of the valve at the breaker using the

controls there will bypass the in-line and 450

psig interlock; prevents automatic opening and

closure due to logic; and prevent any operation

except from breaker

(3)

Can be opened when Rx pressure is < 450

psig

(4)

Automatically opens when Rx pressure < 450

psig with an LPCI initiation signal present and

is interlocked open until LPCI initiation signal is

cleared and reset.

(5)

Only Respective Divisional LPCI Initiation logic

will close the valve.

(6)

Automatically close (both valves) if:

FCV 74-47 and 48 (SID Cooling supply valves)

open and a Group 2 isolation signal occurs

Automatic closure signal seals in (light

indication). Can be reset (FCV 74-53/67

Shutdown Cooling isolation reset pushbuttons)

when any of the conditions above are cleared.

Note that this closure signal will prevent

opening if an LPCI signal is received.

(7)

The normally closed inboard injection valves

(2/3-FCV-74-53 and 74-67) have a new App 'R'

Emergency Open Switch on the power supply

board to bypass all interlocks and other

circuitry (except the fully open limit switch) to

open the valve.

OPL171.044

Revision 15

Page 41 of 159

INSTRUCTOR NOTES

1-74-53 only

1-74-67 only

2-74-53 only

3-74-53 only

NOTE: Effect of

Logic failure and

valve operation

NEW!

The Redundant

logic has been

removed.

Sys I-XS-74-126

Sys II-XS-74-132

(

(7)

Separate bypass switch allows bypassing

interlock from Valves 74-2/13 (74-25/36)

OPL171.044

Revision 15

Page 43 of 159

INSTRUCTOR NOTES

(8)

Control Circuit

1-74-57

1-74-71

2-74-57

2-74-71

3-74-57

3-74-71

Operate Interlock

74-58,

74-72,

74-58,

74-72,

74-58,

74-72,74-

74-2/13

74-25/36

74-2/13

74-25/36

74-2/13

25/36

Outgoing interlock

74-2/13

74-25/36

74-2/13

74-25/36

74-2/13

74-25/36

Normal/Emerg Sw

X

Local Controls

X

Controls at Bkr

X

Panel 9-3 controls

X

Local Indication

X

IQts

Lights on Bkr

X

Liqhts on Pnl 9-3

X

Bypass Switch

X

(

m.

RHR Suppression Pool spray valves

(1)

No automatic opening logic

(2)

Interlock prevents normal opening if in-line

valve not full closed (74-57; 74-71)

(3)

Automatically closed and interlocked closed on

LPCI initiation signal.

(4)

The in-line valve interlock and/or the LPCI

closure signal can be bypassed if the following

exist:

(a)

Reactor level ~-183 inches and drywell

pressure ~ 1.96 psig and LPCI initiation

signal and Select-Reset switch to

SELECT position.

(b)

Reactor level interlock and LPCI initiation

signal may be bypassed by use of

keylock bypass switch (XA 74-122/130)

(74-58; 74-72)

Obj. V.C.5.

TP-33, 36 and

37

(

1-74-58

1-74-72

2-74-58

2-74-72

3-74-58

3-74-72

Operate Interlock

74-57

74-71

74-57

74-71

74-57

74-71

Outgoing interlock

Normal/EmerQ Sw

Local Controls

X

Controls at Bkr

Panel 9-3 controls

X

Local Indication Igts

X

Lights on Bkr

Liqhts on Pnl 9-3

X

OPL171.044

Revision 15

Page 46 of 159

(

INSTRUCTOR NOTES

o.

Containment Spray valves

(74-60/61 ;

74-74175)

(1)

No automatic opening logic

Obj. V.C.5

(2)

IN-line valve interlock prevents normal opening

TP-33, 40, 41,

unless other valve fully closed

42,43

(3)

Automatically c1osedlinterlocked closed on

LPCI signal

(4)

Automatic closure signal andlor the in-line

valve interlock may be bypassed if the

following exist:

(a)

Reactor level ~-183 inches and drywell

pressure ~1 .96 psig and LPCI initiation

signal present and Select=Reset switch

placed to SELECT position

(b)

Reactor level interlock and LPCI initiation

signal may be bypassed by use of keylock

bypass switch (XS-74-122/130)

(5)

Amber light above the "SELECT" switch

Obj. V.C.6

indicates:

Switch in "Select or Normal after Select"

AND

DWP is ~1.96 psig

AND

RPV level ~-183" and have LPCI signal

OR

Keylock in Bypass position

(a)

As long as the light remains "on", the

valves may be opened and a LPCI signal

will not close them.

(6)

Drywell pressure interlock prevents drawing

This interlock

vacuum on containment under accident

cannot by

condition.

bypassed.

(

(7)

Emergency position at breaker bypasses both

Obj. V.D.8

of the normal control circuits

U2 & U3-74-60

(opening/closinglinterlocks)

U1-74-74

3. RO 206000K6.09 OOl/C/A/SYS/HPCV4/206000K6.09//RO/SR0/1l/27/07 RMS

Conditions have required entry into EOI-1, RPV Control and EOI-2, Primary Containment

Control .

I

Given the following plant conditions:

I

Unit 2 reactor water level initially lowers to -69 inches.

(

After water level recovery, the HPCI Pump Injection Valve (73-44) is manually closed and

HPCI is placed in pressure control to remove decay heat.

Subsequently, CST level drops below 6800 gallons.

Drywell Pressure is now less than 2.45 psig.

Which ONE of the follow ing describes the status of HPCI, assuming NO operator action has been taken

other than the pressure control lineup?

A.

HPCI would be operating in pressure control with suction from the CST.

B.

The HPCI turbine would trip on overspeed due to loss of suction during the transfer.

C.'; HPCI would be operating at shutoff head with suction from the suppression pool.

D.

HPCI would be pumping to the CST with suction from the suppression pool.

KIA Statement:

206000 HPCI

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

PRESSURE COOLANT INJECTION SYSTEM : Condensate storage and transfer system: BWR-2,3,4

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

plant conditions to determine the effect of low CST level on HPCI operation .

References: OPL171.042 Rev 19 Page 36

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. Recognize that the HPCI initiation signal is reset to allow HPCI to be placed in Pressure Control.

2. Recognize that the HPCI Pressure Control lineup if from the CST and back to the CST.

3. Recognize that the current CST level would initiate a suction swap to the SUppression Pool.

4. Recognize that HPCI would not receive a trip signal as the suction valves re-aligned.

5. Recognize that the CST Test Isolation Valve will auto close on low CST level.

A is incorrect. This assumes the low CST level has not initiated the suction swap. This is plausible since

the specific level is given using both tank capacity and elevation above sea level.

B is incorrect. HPCI will not trip on low suction pressure under this specific condition. The SP suction

valves begin to open before the CST suction valve closes. This is plausible since closure of the suction

path to HPCI typically results in a low suction trip.

C is correct.

D is incorrect. This lineup would occur if the HPCI Test Isolation Valve did NOT receive a close signal

following the suction swap logic initiation. This is plausible since the ONLY auto closure interlock of the

HPCI Test Isolation Valve is under this specific condition .

Obj. V.BA

Obj. V.CA

OPL171.042

Revision 19

Page 36'of 67

INSTRUCTOR NOTES

2 ~'

"

<lf d~ring : HP.GI:<:>RE3ratiq:ri~

*suppre~~ibn ', poo LVI{~ter

level i H'breases) c{ 7.";C$:*2":6b ':Uni( 3Yapo" e:zero'or'if

", "

.

, ::_~,

.

.'

,"

...,

'A""

~".

'

,,,,~. _

.

CST level'drop's'to ;552'6"'abbve',sea leYel (7000

gallons), tlien HJ?,el~pu'nipsuctibn valves (rom;.the

suppressiof(pobl .(73-26 :aI1Cr~3-:27) *()p~h '. '(This:will

' ".

'

~./{

':I"' .'

"

,';.

<. _ ~ . ,

..:'

.-,

"Y..*

"',." . ,

, . _. ._,,"T-

then causeJheCSTs'uctiori,valve]oc!ose once the'

SP.suction valvesgetfullopen).

NOTE: There -are.normally.300,OOO.galiOns

available in ,the CST for HP't~Land*RCIC use.

3 "

... ' A flow switch tappedin parallel with-the HPCI

system flow controller closes .the minimum flow

bypass valve to suppression.pool' (73-30) at 1255

gpm increasinq: andopens it at 900 gpm

decreasing, only iran auto start signal ispresent.

Minimum flow valve closes on a Turbine Trip signal.

.lfeither of the -suppression pool.suction line

isolation val\\(es(7j:3:26~6r 73-27)8re full open then

the HpCltest line to the.CSTvalves (73-:-35 and*73-

36) will close.

4.'

(

8.

5.

If the HPCI turbine isolation valve (73-16) is fully

closed, then gland seal condenser condensate

pump discharge valves to clean radwaste (73-17A

and 7B-178) will open if the gland seal condenser

hotwell has high level.

6.

If the HPCI turbine isolation valve (73-16) is fully

closed, then HPCI turbine steam line drain pot

discharge isolation valves to the main condenser

(73-6A and 73-68) will open.

7.

If 73-16 is full closed, the auxiliary oil pump will not

start from the control room. When 73-16 opens

10% and the control switch is in the start position,

the auxiliary oil pump will run.

If the HPCI turbine steam line drain pot level

reaches the high level setpoint, then the

downstream trap bypass valve (73-5) will open.

Unit 1 73-5 has been replaced with a manual valve.

DCN 51221

Unit difference

(

4. RO 209001K5 .04 00 IIMEMlT2G1IBASISI1209001K5.04//RO/SROI

During EOI execution when injection from low pressure systems is required to restore and maintain RPV

level, Condensate, RHR LPCI Mode and Core Spray are preferred systems if all control rods are inserted.

If all control rods are not inserted , Core Spray is not on the list of preferred systems for low pressure

injection.

Which ONE of the following describes the basis for this difference?

A.

Cold water from Core Spray creates a rapid pressure reduction and cooldown rates cannot be

controlled.

B."

Core Spray injects directly on the fuel bundles inside the shroud which could damage fuel and

cause a power excursion.

c.

Core Spray injection creates a steam blanket at the top of the fuel which inhibits heat transfer via

steam flow past the fuel.

D.

Core Spray flow cannot be throttled for several minutes with an initiation signal present.

KIA Statement:

209001 LPCS

K5.04 - Knowledge of the operational implications of the following concepts as they apply to LOW

PRESSURE CORE SPRAY SYSTEM: Heat removal (transfer) mechanisms

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

unique heat removal mechanisms of Core Spray and recall a condition where that mechanism can result

in unfavorable consequences.

References:

EOIPM Section O-V-K

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 bases behind the restriction of Core Spray injection during an ATWS emergency.

A is incorrect. This is plausible since high volume Core Spray injection at close to the maximum

injection pressure would cause a rapid pressure reduction, however this is of minor consequence.

B is correct.

C is incorrect. This phenomenom, referred to as Counter Current Flow Instability, is plausible but is only

of significant concern with the core completely uncovered and is the basis for removing Spray Cooling

from the EPG definition of Adequate Core Cooling.

D is incorrect. This is plausible since RHR LPCI injection valves cannot be throttled for several minutes

following a CAS signal. However, Core Spray valves CAN be throttled immediately.

(~

CS, LEVEUPOWER CONTROL BASES

I

DISCUSSION: STEP C5-16 (Continued)

EOI PROGRAM MANUAL

SECTION O-Y-K

I

(

In comparison to Minimum Zero-Injection RPV Water Level (refer to the discussion ofStep C3-3

in the C3t Steam Cooling, Bases), Minimum Steam Cooling RPV Water Level is slightly higher

than Minimum Zero-Injection RPV Water Level. This is attributed to two key factors:

.

1.

Injection ofsubcooled water requires that part ofthe energy that would be used to

generate steam for cooling the uncovered portion ofthe core must now be expended in

heating subcooled liquid to saturation temperature (Minimum Zero-Injection RPV Water

Level is calculated assuming no injection into the RPV).

2.

More steam is required to maintain clad temperature below 1500 OF as compared to the

1800 OF limit assumed for Minimum Zero-Injection RPV Water Level calculation.

The injection sources listed for use in controlling RPV water level comprise all ofthose that inject

outside ofthe core shroud. These are used, preferentially, because the flowpath outside the core

shroud mixes the relatively cold injected water with warmer water in the lower plenum prior to

reaching the core. No priority between use ofeach listed system is intended, therefore the

operator should use the most appropriate means available under current plant conditions.

EOI Appendices 5A t 5B t and 6A provide guidance to operate CondensatelFeedwater, CRD t and

only Condensate respectively. These systems are preferred sources ofinjection since they are of

high quality water and are used for RPV water level control during normal plant power

operations. Feedwater and CRD both provide high pressure injection from either a steam or

motor-driven supply, and Condensate by itselfprovides for lowpressure injection.

EOI Appendices 5C and 5D provide guidance to operate RCIC and HPCI respectively. The

operator is instructed to operate RCIC and HPCI with suction from the CST ifavailable, to

ensure that the highest quality water is used for injection into the RPV. The CST is the preferred

suction source not only because ofhigher water quality, but also because the CST is not subjected

to the temperature increase that the suppression pool is. For these reasons, defeating HPCI high

suppression pool suction transfer logic in EOI Appendix 5D, allows the operator to maintain the

CST as the suction source. EOI Appendix 5C provides direction to defeat the RCIC low RPV

pressure isolation interlock, that allows operation ofthe RCIC turbine at low pressure. Even if

RPV pressure is below the isolation setpoint, but above turbine stall pressure, RCIC can still

provide some injection into the RPV.

EOI Appendices 6B and 6C provide guidance to operate LPCI Systems I and II respectively. The

operator is instructed to only operate RHR in LPCI mode when suppression pool level is above

<A.62>. Engineering calculations have determined that operation ofRHR pumps below a

suppression pool level of<A.62> may induce vortex formation at the system suction strainer.

REVISION 0

PAGE 45 OF 110

SECTION O-Y-K

EOI PROGRAM MANUAL

SECTION O-V-K .

DISCUSSION: STEP C5-30

(

CS, LEVEUPOWER CONTROL BASES

~ ~-------~I

This signal step informs the operator that actions to control RPV pressure control must

immediately change because ofpresent plant conditions.

When emergency RPV depressurization is required, the operator shall transfer RPV pressure

control actions from the RCIP Section ofEO1-1, RPV Control, to C2, Emergency RPV

Depressurization.

This step has been reached in this procedure because previous attempts to maintain adequate core

cooling have been unsuccessful, or plant conditions are such that emergency RPV

depressurization is required, as indicated by a signal step in another EOI being concurrently

executed.

Ifadequate core cooling cannot be assured, then plant conditions may be such that RPV water

level is at or below TAF, and RPV pressure is high enough to prevent injection from low-head

pumps. Therefore, emergency RPV depressurization is required for the purpose ofmaximizing

injection flow from high-head pumps and to permit injection from low-head pumps.

Depressurizing the RPV is preferred over restoring RPV water level through the use ofsystems

that inject inside the shroud because:

1.

A large reactor power excursion may result from the in-shroud injection ofrelatively cold

water.

2.

Rapid depressurization, by itself, will shut down the reactor due to a substantial increase in

voids.

3.

Following the depressurization, reactor power will stabilize at a lower level.

REVISION 0

PAGE 81 OF 110

EOI PROGRAM MANUAL

SECTiON O-V-K .

I

1

Caution #5 applies throughout performance ofStep C5-38. Caution #5 is identified at this step to

highlight the potential for large power excursions and subsequent core damage ifcold, unborated

water is rapidly injected using injection sources within Step C5-38.

,

This action step directs the operator to use injection sources listed to restore and maintain RPV

water level above <A.71>. System specific EOr Appendices provide step-by-step guidance for

lining up and injecting into the RPV. Injection pressures <<A.1>> have also been provided as

additional information to the operator.

Engineering calculations have determined that when RPV water level is at or above <A.71>,

adequate core cooling is still assured. The value of<A.71> RPV water level is Minimum Steam

Cooling RPV Water Level. Refer to discussion ofStep C5-16 for more information on Minimum

Steam Cooling RPV Water Level.

This step has been reached only when RPV water level cannot be restored and maintained above

<A.71> using preferred systems. Therefore, use ofadditional systems is required that either inject

inside the core shroud, are difficult to lineup, or take suction on sources ofcomparatively lower

water quality. No priority between use ofeach listed system is intended, therefore, the operator

should use the most appropriate means available under current plant conditions.

Eor Appendices 6D and 6E provide guidance to operate CS Systems I and II. CS provides

relatively high quality water from the suppression pool and can provide injection into the RPV

quicker than other sources listed in this step. However, reactor power excursions are more

probable since CS injects directly into the core shroud at high flowrates. Therefore, extreme

caution should be used for CS injection at this step.

Unlike directions given for use ofmotor driven pumps in EOr-I, RPV Control, CS System

operation is not restricted by pump NPSH and Vortex limits (suppression pool level). Even

though risk ofequipment damage exists ifNPSH and Vortex limits are exceeded, immediate and

catastrophic pump failure is not expected should operation beyond these limits be required. Since

prolonged operation under these conditions is most likely required before degraded system and

pump performance may result, the undesirable consequences ofuncovering the reactor core

outweigh risk ofequipment damage.

Eor Appendices 7C, 7E, and 7F provide guidance to inject RHR into the RPV from crossties to

other units or through RHR Drain Pumps A and B. EOr Appendix 7G provides guidance to inject

into the RPV with PSC Head Tank Pumps. All ofthese injection sources provide suppression

pool water at low pressure, but are relatively complicated to line up.

REVISION 0

PAGE101 OF 110

SECTION O-V-K

5. RQ 211000AK2.01 00l/C/A/T2Gl/3/06/63N.B.5/211000AK2.0l/2.9/3.IIRO/SRO/l0/27/2007

Given the follow ing plant conditions:

Unit 1 is operating at 75% power.

A fire is discovered inside 480V Shutdown Board 1B causing a loss of the 480v Shutdown

Board 1B.

Fire Protection reports that the fire cannot be extinguished.

The US directs a manual scram .

Not all control rods insert, and the following conditions are noted:

- Reactor Power

- Suppression Pool Temperature

15%

1080F and rising

(

The "A" 4KV Shutdown Board deenerg ized when 1A RHR pump was started for pool cooling.

Which ONE of the following describes the action and method of injecting boron into the reactor?

A.

Transfer 1B 480v Shutdown board and inject SLC using 1B SLC pump.

B"" Transfer 1A 480v Shutdown board and inject SLC using 1A SLC pump.

c. Transfer 1B 480v RMOV board and inject SLC using 1B SLC pump.

D. Transfer 1A 480v RMOV board and inject SLC using 1A SLC pump.

KIA Statement:

211000 SLC

K2.01 - Knowledge of electrical power supplies to the following : SBLC pumps

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

the power supplies to the SLC pumps .

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 solve a problem. This requires mentally using this

knowledge and its meaning to resolve the problem.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Justification:

Answer A is not correct. Due to the loss if "A" 4KV Shutdown Board, the 1A 480V Shutdown Board has

lost power. This is plausible if the candidate is not aware that the 480V Shutdown Board does not

automatically transfer to Alternate the same as the 4KV Shutdown Board.

Answer B is not correct because the B 480V Shutdown Board is unavailable due to a fire. This answer is

plausible if the operator does not know the correct power supply to SLC pumps.

Answer C is the correct answer. Manually transferring 1A 480V Shutdown Board to Alternate will restore

power to the board and allow starting 1A SLC pump.

Answer D is not correct because the power supply to 1A SLC is not 1A 480V RMOV Board. It is plausible

because 1A 480V RMOV Board is safety related and powered from the same DG as 1A 480V Shutdown

Board.

-O:l>::CO

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4KV SID Bd.

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4KV SID Bd.

C

,

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) NC

LLJ

I

L

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LLJ

rTl

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NO!)I ~

480V DSL Aux Bd.

IL--.__....

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NO

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NC

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NC

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480V

RMOVBd.

2C

4KV SID Bd.

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.--....---..---.---.----') NC

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4KV SID Bd.

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480V DSL Aux Bd.

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1r.

OPL171.039

Revision 16

Page 15 of 48

(

INSTRUCTOR NOTES

4.

SLC Pumps

a)

Two 100% capacity, triplex, positive displacement

Obj. V.B.5.c

piston pumps are installed in parallel.

Obj. V.C.4.d

b)

'A' pump is powered from 480V Shutdown Board A.

Obj. V.D.4

Obj. V.E.4

c)

'B' pump is powered from 480V Shutdown Board B.

Obj. V.B.5.c

d)

Electrically interlocked so that only one pump will run at

Obj. V.C.4.d

a time. This prevents system overpressurization.

Obj. V.B.3.f

e)

The pumps are manually started from the main control

Obj. V.C.2.f

room using the key-lock switch on panel 9-5, or locally,

using the Test Permissive Transfer Switch at Panel 25-

19.

f)

A control room start signal will fire the explosive valves.

A local start will not fire the explosive valves.

g)

Either pump is capable of supplying a system flow of

Obj. V.D.3.d

approximately 50 gpm at a system pressure of 1275

Obj. V.E.3.d

psig.

h)

Each pump discharge has a relief valve, set at 1425 +/-

Obj. V.B.3.f

75 psig, to protect the pump and the system from

Obj. V.C.2.f

overpressurization.

Obj. V.D.3.e

Obj. V.E.3.e

i)

Each pump contains internal suction and discharge

check valves, which open at approximately 5 psid,

allowing only forward flow through an idle pump. (INPO

O&MR 341).

j)

Pump motors are protected by an undervoltage trip.

5.

Accumulators

a)

An accumulator is installed between each pump and its

discharge check valve.

b)

Dampens the pressure pulsations that are inherent with

Obj. V.D.3.d

piston-type, positive-displacement pumps.

Obj. V.E.3.d

c)

A steel vessel accumulator, containing a synthetic

bladder, with one side charged to -450 psig nitrogen

gas and SLC solution on the other side.

(~

(

6. RO 212000K6.03 OOl/MEM/T2Gl/RPS//212000K6.03//RO/SRO/

Given the following plant conditions:

Reactor water level instrument L1S-3-203A has failed downscale.

Which ONE of the following describes the Analog Trip System Response?

The trip relay will be

and the contact in the RPS logic will be

_

A.

energized, closed

B.

energized, open

C.

deenergized, closed

D..... deenergized, open

KIA Statement:

212000 RPS

K6.03 - Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR

PROTECTION SYSTEM : Nuclear boiler instrumentation

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

plant conditions involving level instrumentation to determine the response of RPS logic components.

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. Whether RPS relays are normally emergeized or de-energized.

2. Whether RPS contacts fed from relays are normally open or closed.

3. Recall which scram signal, if any, is fed from L1S-3-203A.

A is incorrect. This is plausible if the novice operator fails to recognize a valid trip has been generated

by L1S-3-203A.

B is incorrect. This is plausible if the novice operator confuses PCIS logic relay response with RPS logic

relay response.

C is incorrect. This is plausible if the novice operator confuses PCIS logic relay response with RPS logic

relay response and fails to recognize a valid trip has been generated by L1S-3-203A..

D is correct.

OPL171.028

Revision 17

Page 13 of 50

INSTRUCTOR NOTES

(2)

The third is used to produce manual SCRAM

trip signals (trip channel A3).

(3)

The channels for trip system Bare

designated B1, B2 and B3.

c.

Both of the automatic channels in each trip system

monitor critical reactor parameters.

(1)

At least four channels for each monitored

parameter are required for the trip system

logic.

(2)

If either of the two channels sense a

parameter which exceeds a setpoint, then

this would place the associated trip system (A

or B) into a tripped condition.

(3)

To produce a SCRAM, both trip systems

must be tripped. This is called a "one-out-or-

two-taken twice" arrangement.

Obj. V.B.5.c

Obj. V.D.4

TP-3

Drawing

2-730E915RF-11

2-730E915RF-12

d.

Each trip system logic may also be manually

TP-4

tripped .

(1)

Each Trip system contains manual SCRAM

switches on Panel 9-5 which cause a trip in

the respective trip system when actuated .

(2)

The Reactor mode switch has contacts in

both the A3 and B3 channels. Placing the

reactor mode switch in SHUTDOWN will

result in a trip of both trip systems.

(3)

A trip in both channels A3 and B3 initiates a

reactor SCRAM.

0) -: All se~~9r:ctrl(rt'ripG9ntciCtsessential to

safetyareclo$e9,

(?)~,{1Eh~*~~:~1~~'lqgig~ ;'\\C:lIJ~f~~~t~C:l,t¢r~ .~(~

~n~rgl?~dx

Drawing

2-730E915RF-11

2-730E915RF-12

SER 3-05 Operator

fundamentals

(

(3)':'. Wh~n~a 'S.<<,~<~ signal.is~n3ce ived , the logic

relays :,geen~rgize toc ause a SCRAM.

(4)

Loss of power to one RPS bus will result in a

half-SCRAM. Loss of power to both RPS

buses will result in a full SCRAM.

4.

Reactor SCRAM Signals and Arrangement

Refer to 01-99 for the setpoints for each SCRAM.

a.

Channel test switch

OPL171.028

Revision 17

Page 14 of 50

INSTRUCTOR NOTES

Drawing

2-730E915-13

Obj. V.B.3

Obj. V.C.3

Obj. V.D.8

Obj. V.B.6

Obj. V.C.4

SER 3-05 Operator

fundamentals

(1)

Allows for testing each channel's trip function.

Drawing

2-730E915RF-11

2-730E915RF-12

(2)

Four, one per channel located on Panel 9-15

REACTIVITY

and 9-17 in Aux. Inst. Room.

MANAGMENT

Discuss when switches

(3)

Key-locked, two positions - NORMAL and

can be used.

TRIP

(4)

TRIP de-energizes that channel's relays

producing a half-SCRAM.

b.

Turbine Stop Valves, 10 percent closure

anticipates the pressure and neutron flux rise

caused by the rapid closure of the Turbine Stop

Valves.

(1)

Each of the four Turbine Stop Valves is

equipped with two limit switches. One limit

switch is assigned to RPS "A" and one to

RPS"B".

(2)

These switches will provide a valve-closed

signal to the RPS trip logic.

(3)

The position switch contacts are arranged so

that any two Stop Valves can be closed

causing no more than a half-SCRAM.

(4)

Closure << 90% full open) of any combination

of three Stop Valves will cause a full SCRAM

in all cases.

TP-5

Drawing

2-730E915-9, 10

Obj. V.D.5

Drawing

2-730E915RF-11

2-730E915RF-12

(

C.

Typical PCIS Isolation Logic

OPL171.017

Revision 13

Page 12 of 56

INSTRUCTOR NOTES

TP-1

1.

A typical logic arrangement for the PCIS valves

(except MSIVs) is shown in TP-1. This figure shows

that two separate trip channels (A and B) are each

provided with two sensor relay contacts (AIC and

BID).

a.

This arrangement creates trip subchannels

A1/A2 and B1/B2.

PCIS de-energizes

to isolate (except

HPCIIRCIC)

Obj. V.B.1

Obj. V.C.1

b.

A trip of either sensor relay within a trip

channel will cause opening of the associated

contact and de-energization of the

associated relay. This condition will create a

"half isolation" signal within both logic

channels but NO VALVE MOVEMENT.

HPCI/RCIC are

energize to actuate

Obj. V.B.3

Obj. V.C.3

c.

Should a trip of either sensor relay in the

other trip channel occur, conditions will exist to

de-energize the valve actuation relays in each

logic channel, causing both isolation valves

to close.

PCIS logic is arranged as follows:

A10RA2

AND

= Inboard AND Outboard valve closure

B1 OR B2

Note: Most PCIS logic is assembled as above.

The MSL drains however are an exception .

The MSL drain logic is as follows:

A1 AND B1 = liB valve closure

A2 AND B2 = O/B valve closure

(

7. RO 215003A4.04 OOI/MEM/SYS/IRM/B6/215003A4.04//RO/SRO/

Given the following plant conditions:

Unit 1 reactor startup preparations are in progress with no rods withdrawn.

Instrument Mechanics are performing the IRM functional surveillance.

No IRMs are currently bypassed.

The Instrument Mechanic Technician has depressed (and held) the "INOP INHIBIT"

pushbutton for "H" ChannellRM.

Which ONE of the following describes the IRM trips that are bypassed as a result of this action, if any?

A.

The IRM "Loft of +/-24 VDC" inop trip is bypassed .

B.

IRM "High Voltage Low" trip is bypassed.

C." The IRM mode switch "out of operate" inop trip is bypassed.

D.

The IRM "Module unplugged" inop trip is bypassed .

KIA Statement:

2150031RM

A4.04 - Ability to manually operate and/or monitor in the control room: IRM back panel switches, meters,

and indicating lights

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

component manipulations to correctly determine the response of the IRM system.

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. The function wich is bypassed by the INOP INHIBIT pushbutton.

NOTE: Each of the possible answers below will typically initiate an INOP trip of it's associated IRM

channel, therefore each distractor is plausible.

A is incorrect. This INOP trip will still function.

B is incorrect. This INOP trip will still function.

C is correct.

D is incorrect. This INOP trip will still function.

(

E.

OPL171.020

Revision 10

Page 20 of 42

INSTRUCTOR NOTES

TP-10

-HV low <<90v)

RUN Mode

-Module unplugged

-Function switch not in OPERATE

-Loss of j:24VDC

1.

ad b1bcks

Block

Setpoint

Downscale

~ 7.5

~90

When Bypassed

Range 1 or RUN

RUN Mode

Obj.V.D.7, V.B.5

Obj. V.C.3.,

Obj. V.B.6.

Obj. V.CA.

Obj. V.B.5

Detector Wrong

Position

Scrams

High-High

Detector

Not Full IN

Setpoint

~ 11604

RUN Mode

When Bypassed

In RUN Mode

Obj'v.B.13

TP-11

Obj. V.B.7.

Obj. V.C.5. Obj.V.D.8

(

INOP

-HV low <<90v)

In RUN Mode

-Module unplugged

-Function switch not in OPERATE

-Loss of j:24VDC

F.

Controls Provided

1.

Panel 9-5

a.

Recorder switches switch between IRM channels

and APRM/RBM channels

b.

Range switches allow operator to select

appropriate IRM range to maintain indications

between 25 to 75 on 0-125 scale. 0-40 scale is

no longer utilized.

(

c.

(2) 'Standby' - same as operate, except gives

Inop trip to yield maximum design

protection before channel is

removed from service.

(3) 'Zero l' - Removes signal from output

amplifier so that output amplifier,

local meter and recorder can be

zeroed.

(4) 'Zero 2' - Removes voltage from range

switch. This deselects all ranges.

This, in turn, causes no input to be

sent to attenuator and allows

setting the zero adjust on output

amplifier.

(5) '125 '- Input is removed from attenuator same

as Zero 2 position. A calibration signal

is substituted which will yield 125 on

the 125 scale. Used to set gain of

output amplifier.

(6) '40' - Produces a 40 reading on the 125

scale.

INOP/INHIBIT Pushbuttons

(1) Pushed to bypass the INOP trip that results

from taking mode switch S-1 out of

"operate."

(2) Used to allow testing of other scram or rod

block signals from the IRM drawer into

RPS/RMCS without them being masked by

the INOP trip.

OPL171.020

Revision 10

Page 22 of 42

INSTRUCTOR NOTES

Obj. V.B.6

Obj. V.C.4

DCN W18726A

replaced the

INOPIINHIBIT

Pushbutton with a

toggle switch for the

U-3 IRM drawers.

(UNIT DIFFERENCE)

(

8. RO 215004A3 .03 OOl/C/A/T2Gl/SRM/B8/215004A3.03//RO/SRO/

Given the following plant conditions:

A reactor startup is in progress following refueling, with all RPS shorting links removed .

The reactor is approaching criticality.

A loss of the High Voltage power supply to the B SRM detector results in the INOP trip and

Panel 9-5 alarm on SRM HIGH/INOP.

Which ONE of the following describes the plant response?

A.

Alarms only.

B." A Rod Out Block only.

C.

A Rod Out Block and 1/2 Scram.

D.

A Rod Out Block and Full Reactor Scram.

KIA Statement:

215004 Source Range Monitor

A3.03 - Ability to monitor automatic operations of the SOURCE RANGE MONITOR (SRM) SYSTEM

including: RPS status

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

plant conditions to correctly determine the response of the SRM with the shorting links removed.

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. The IRM response to the high voltage power supply failure under typical conditions.

2. The IRM response to the same conditions with the shorting links removed.

NOTE: This question initially appears to have Low Discriminatory Value but received a 100% failure rate

during validation. Every Licensed Operator chose Answer 0, believing a scram signal was generated.

However, I feel this question is appropriate for the KIA and SHOULD remain in the exam.

A is incorrect. A Control Rod Block is generated.

B is correct.

C is incorrect. This is plausible if the novice operator determines a scram signal is generated with the

typical "t-out of-2 taken twice" logic. However, only SRM Hi-Hi generates an input to RPS.

D is incorrect. This is plausible if the novice operator determines a scram signal is generated . However,

only SRM Hi-Hi generates an input to RPS.

(

Source Range Monitors

1-01-92

Unit 1

Rev. 0006

Page 6 of 14

3.0

PRECAUTIONS AND LIMITATIONS

A.

To prevent a rod withdrawal block when withdrawing SRMs, SRM count rate is

required to be above Retract Permit (145 counts per second) or all unbypassed

IRM channels are set to Range 3 or above and indicating above their

downscale trip point (7.5 on 125 scale).

B.

Only one SRM channel can be bypassed at a time.

C.

In order to prevent an inadvertent rod withdrawal block or Reactor scram (with

shorting links removed) while operating the SRM BYPASS selector switch,

1-HS-92-7A1S3,

Verify the previously bypassed channel returns to normal status by

observing the applicable HIGH HIGH and HIGH or INOP status lights are

extinguished prior to selecting any other channel to be bypassed.

After bypassing a channel, the applicable BYPASSED status light should

be illuminated prior to testing, operating, or working on that channel.

D.

To prevent SRM detector drive damage, the CRD service platform should be

locked in the stored position with key removed to allow free movement of

SRMs.

E.

In order to minimize their exposure, SRM detectors should be fully withdrawn

from the core when IRMs are on range 3 or above and indicating above their

downscale trip point.

F.

Illustration 1 lists trip signals and associated actions for the Source Range

Monitoring System.

G.

The Reactor Protection System in conjunction with the Neutron Monitoring

System (SRM and IRM) has non-coincident trip logic if all eight shorting links

are removed. If only the yellow, green, and red shorting links (six total) are

removed, the SRM High-High trips will be placed in a one-out-of-two taken

twice logic.

H.

The time required to drive a detector from full-out to full-in is approximately

3 minutes.

..

1.

Ii

I

1:11 f

swi clies, located on Panel 1-9-12 SRM drawers, bypass the

SRM switch position out-of-oRerate trip. I hese switches

re to be used only '

timing testing of. SRJ\\i1 cnannels.

J.

[NRC/C) Upon return to service of 24 VDC Neutron Monitoring Battery A or B,

Instrument Maintenance is required to perform functional tests on SRMs and

IRMs that are powered from the affected battery board.

[NRC IE Inspector Followup Item

86-40-03]

1

"'tJ~::oo

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CD

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CD

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CD

LOCAL

PERIOD

METER

PERIOD

51

.0

10

105

REF

LOCAL

LCR

METER

OP

PERIOD

AMP

SCRAMWITH SHORTING LINKS REMOVED

PERIOD

TRIP

51

~

o

~

LCRTR'P

- I~

RAMP

FIXEO -0 1 0-

VARIABLE

OP

10

51

REF

OP

r.------.,

g

~

FLEXIBLE

DRIVE

5HAFT

SHIELD

WALL

+20VDC

15VOC

REACTOR CONTAINMENT

+/- 24vdc

eno

C

0

om

~z

(j)

m

s:oz

=io

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Zo

o

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Z

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(

9. RO 215005A2.03 00 lIe/AlTIG l/PRNM/APRMlB7/215005A2.03/IRO/SRO/

Which ONE of the following describes the expected response due to a "FAULT" in an APRM channel and

the required action(s), if any, to address this condition?

A.

An APRM channel Non-critical Fault will result in an INOP trip input to all four

2/4 logic modules (voters) . Bypass the APRM per 01-92C and continue operation.

B.

An APRM channel Non-critical Fault will result in an INOP trip input to only the respective 2/4 logic

module (voter). Bypassing the APRM is not required to continue operation.

C...; An APRM channel Critical Fault will result in an INOP trip input to all four 2/4 logic module (voters).

Bypass the APRM per 01-92B and continue operation.

D.

An APRM channel Critical Fault will result in an INOP trip input to only the respective 2/4 logic

module (voter) . Bypassing the APRM is not required to continue operation.

KiA Statement:

215005 APRM / LPRM

A2.03 - Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE

MONITOR/LOCAL POWER RANGE MONITOR SYSTEM; and (b) based on those predictions, use

procedures to correct, control, or mitigate the consequences of those abnormal conditions Inoperative trip

(all causes).

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

plant conditions to correctly determine equipment response and the corrective actions due to an APRM

INOP trip.

References: 01-92B Precautions and Limitations 3.0.1

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. The difference between a "Critical Fault" and "Non-critical Fault" with respect to PRNM response .

2. The operation of the 2/4 Logic Module Voter operation for each type of fault.

3. Based on the above, determine the appropriate course of action regarding the APRM cahnnel is

question.

A is incorrect. A "Non-critical Fault" will not result in an INOP trip input. This is plausible because a

"Non-critical Fault" generates a Trouble Alarm similar to a "Critical Fault". In addition, placing the APRM in

BYPASS would be correct IF an INOP trip was geneated.

B is incorrect. A "Non-critical Fault" will not result in an INOP trip input. This is plausible because a

"Non-critical Fault" generates a Trouble Alarm similar to a "Critical Fault". In addition, not placing the

APRM in BYPASS because of a "Non-critical Fault" is appropriate.

C is correct.

D is incorrect. This is plausible because a "Critical Fault" generates an INOP trip input to the 2/4 Logic

Module Voters, but to all four modules, not just two. In addition, not placing the APRM in BYPASS

because of a "Critical Fault" is inappropriate since any additional equipment failure could result in an

unnecessary scram.

(

b.

Each LPRM instrument provides a brief

description of the self-test faults which are

divided into two categories, "Critical" and

"Non-Critical" faults.

(1)

Critical faults are those that affect

the instrument's capability to

ertorrn

its intended function and will cause

an instrument I a

trip- and a

T.rol:Jble Alarm indication.

(2)

Non-critical faults do not prevent

the instrument from performing its

intended function and will cause a

Trouble Alarm indication only.

OPL171.148

Revision 8

Page 31 of 150

Obj, V,D,4

V.B.? V.C.2

The Trouble

Alarm is indicated

in the Status

Header for each

instrument.

(

c.

The LPRM instrument transmits its self-test

status to its associated APRM and RBM

instruments.

(

8FN

Average Power Range Monitoring

1-01-928

Unit 1

Rev.OOOa

Page 7 of 27

3.0

PRECAUTIONS AND LIMITATIONS (continued)

I.

Each of the four APRM/OPRM channels input to the four Voters, such that

when a signal is generated from an APRM/OPRM channel, all four Voters see

and reflect that signal. Each Voter is directly associated with one RPS sub-

channel.

When operating in a 2 out of 4 voting configuration, the first un-bypassed input

will be seen as a single input with no trip outputs. When the second un-

bypassed signal of the same type [The SAME TYPE inferring that one type is

an APRM function and a different type is an OPRM function] is received it will

also be seen by all four Voters resulting in a trip output from all four Voters

consequently producing a full reactor scram.

J.

Bypassing an APRM does not preclude testing a Voter, such that with an APRM

in bypass, the Voters can still be tested and produce half scrams. Voters are

not bypassed with the APRM joystick.

K.

The Recirc Flow Indication and the Voters are never bypassed unless they are

removed for testing. There is no bypass capability for the Recirc flow signal

input or Voters .

L.

A reactor scram will be produced when at least two of the SAME TYPE of trip

inputs are received by the Voters:

Either: APRM HIGH/INOP {i.e., APRM High Flux/STP Flow Biased

Scram/INOP}

OR: Any OPRM ABA, PBA, or GRBA algorithm trip conditions met.

The SAME TYPE inferring that one type is an APRM function and a different

type is an OPRM function .

M.

The new APRM modules contain an automatic power oscillation detection and

suppression function (Oscillation Power Range Monitoring) which detects and

protects against thermal hydraulic instabilities. OPRM monitors local cell area

for thermal hydraulic core instabilities. There are 4 channels each containing 33

cells. Each cell contains up to 4 LPRM inputs per OPRM channel for power

monitoring.

Oscillations are detected using anyone of three algorithms; Period Based

Algorithm, Growth Rate Based Algorithm, and, Amplitude Based Algorithm .

When power oscillations are detected a trip signal inputs to the Voters which

will in turn, send a trip output to the RPS sub-channels and will produce a trip

signal. Two of these types of signals will produce a full reactor scram.

(

8FN

Average Power Range Monitoring

1-01-928

Unit 1

Rev.OOOa

Page 19 of 27

Illustration 1

(Page 1 of 6)

APRM/OPRM Trip Outputs and PRNMS Overview

APRM Trip Outputs

TRIP SIGNAL

SETPOINT

ACTION

APRM Downscale

5%

1.

Rod Block if REACTOR MODE

SWITCH in RUN.

APRM Inop

1.

APRM Chassis Mode not in

1.

One Channel detected, no alarm or

OPERATE (keylock to INOP).

RPS output signal.

2.

Loss of Input Power to APRM.

2.

Two Channels detected, RPS output

signal to all four Voters (Full Reactor

Scram).

3.

Self Test detected Critical Fault in

the APRM instrument.

4.

Firmware Watchdog timer has

timed out.

APRM Inop

1.

< 20 LPRMs in OPERATE, or

1.

<20 LPRMs total or <3 per level results

Condition

< 3 LPRMs per level.

in a Rod Block and a trouble alarm on

the display panel. This does not yield

an automatic APRM trip, but does,

however, make the associated APRM

INOP.

APRM High

1.

DLO

1.

Rod Block if REACTOR

s (0.66W + 59%)

MODE SWITCH in RUN.

SLO

s (0.66 (W-D.W) + 59%)

[W = Total Recirc drive flow in %

rated].

2.

Neutron Flux Clamp Rod Block

2.

Rod Block if REACTOR

s 113%

MODE SWITCH in RUN.

3.

s 10% APRM Flux.

3.

Rod Block in all REACTOR MODE

SWITCH positions except RUN.

APRM High High

1.

Scram

a.

DLO

s (0.66W + 65%)

SLO

s (0.66(W-D.W) + 65%)

[W = Total Recirc drive in %

rated].

b.

s 119% APRM FLUX.

2.

s 14% APRM FLUX.

2.

Scram in all REACTOR

MODE SWITCH positions except RUN.

Recirc Flow

1.

~ 5% mismatch between APRM

1.

Flow compare inverse video alarm.

Compare

Channels .

Recirc Flow

2.

107% Flow Monitor upscale.

2.

Rod Block.

Upscale

I

(

10. RO 217000K2.03 00 lIC/A/TIG1IRCIC/7/217000K2.03//RO/SRO/

Given the following Unit 2 conditions:

The Control Room has been evacuated.

RCIC is controlling Reactor level.

A loss of Div I ECCS inverter occurs.

Assuming no further operator action...

Which ONE of the following describes the RCIC turbine speed control response?

A.

Lowers to minimum in manual ONLY.

B.

Raises to maximum in manual ONLY.

C." Lowers to minimum in either manual or auto mode.

D.

Raises to maximum in either manual or auto mode.

KIA Statement:

217000 RCIC

K2.03 - Knowledge of electrical power supplies to the following : RCIC flow controller

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

plant conditions to correctly determine a loss of the logic power supply to the controller has occurred and

the RCIC system response to that loss.

References: 2-01-71 Precautions and Limitations 3.0.W

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. The current power supply to the RCIC controller while being operated from Panel 25-32.

2. The Power Transfer Switch (XS-256-1) is NOT part of the RCIC initiation procedure in

2-AOI-100-2.

3. The failure mode of the Yokogawa Flow Controller used for RCIC while operating from

Panel 25-32.

NOTE: Due to the wide spread use and various failure modes of Yokogawa Flow Controllers at BFN,

each of the four answers become plausible for a novice operator. These controllers can be set to fail

"as-is", "fail high" or "fail low" depending on the system and application.

A is incorrect. The RCIC controller at Panel 25-32 fails low in AUTO or MANUAL.

B is incorrect. The RCIC controller at Panel 25-32 fails low.

C is correct.

D is incorrect. The RCIC controller at Panel 25-32 fails low.

Reactor Core Isolation Cooling

2-01-71

Unit2

Rev. DOSS

Page 11 of 70

3.0

PRECAUTIONS AND LIMITATIONS (continued)

Q.

Suppression pool water temperature should not exceed 95°F without

suppression pool cooling in service to restore temperature to less than or equal

to 95°F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

R.

RCIC Testing is NOT permitted with suppression pool water temperature above

105°F.

S.

After RCIC steam lines have been hydrostatically tested, leak tested, or

exposed to other conditions which could fill the 2-FCV-71-2 valve bonnet with

water, 2-FCV-71-2 should be cycled to prevent overpressurization.

T.

[II/F] Prior to initiating any event which adds , or has the potential to add, heat

energy to the suppression chamber, the Unit Supervisor will evaluate the

necessity of placing suppression pool cooling in service. This is due to the

potential of developing thermal stagnation during sustained heat additions.

[11-8-91-129]

U.

Calculations have shown that 16 min. of RCIC operation without RHR operating

in the Suppression Pool Cooling Mode will result in a one deg F rise in bulk

suppression pool temperature.

V.

[NER/C] Extended RCIC System operation may raise suppression chamber O2

concentration above TRM 3.6.2 limits because of air-inleakage from RCIC

Turbine Gland Seal System.

[GE SIL 548]

W.

Whenever the 1E ECCS ATU Inverter (Division I) becomes INOP, RCIC Is

considered INOP.

DCN W17726B changed power supply for RCIC flow

controller from 1E Unit Preferred MMG set busses to the Unit 2 1E ECCS ATU

Inverter (Division I).

X.

The RCIC STEAM LINE OUTBD ISOLATION VLV hand switch, 2-HS-71-3A,

must be held in the OPEN position until2-FCV-71-3 is fully open because the

open seal-in circuit has been removed per ECN P0161 .

Y.

[INPO/C] A buildup of corrosion products in the RCIC TURBINE CONTROL

VALVE stem packing could result in speed oscillations, failure to control at the

desired speed, and mechanical overspeed of the RCIC Turbine. During

operation, RCIC Turbine parameters such as time to reach operating speed,

speed stability, and governor response should be monitored to identify possible

corrosion product buildup in the RCIC TURBINE CONTROL VALVE.

[INPOSER

95004]

Z.

(II/C) During routine plant evolutions, notify RADCON prior to making changes in

the RCIC System which could cause a rise in area radiation levels . Confirm

RADCON has implemented appropriate radiological controls/barriers for the

expected RCIC System alignment prior to performing the alignment.

(BFPER961778)

RCIC TURBINE

SI-71-42A

SPEED

RCIC TURBINE

FI-71-1A

STEAM FLOW

FI-71-1 B

2.

Flow Controller (FIC-71-36A & B)

RCIC PUMP

SUCTION

PRESSURE

RCIC TURBINE

STEAM LINE

PRESSURE

RCIC TURBINE

EXHAUST

PRESSURE

PI-71-20A

PI-71-4A

PI-71-12A

OPL171.040

Revision 22

Page 21 of 74

INSTRUCTOR

NOTES

0-50 psig

0-1500 psig

0-50 psig

BFPER971133

Indicator could read

from 0-200 rpm in

standby

0-6000 rpm

readiness due to

non-linearity in low

0-80 Ibm x1000

RPM range

a. One located on Panel 9-3 and one on Panel 25-32

(Remote Shutdown Panel) .

b. Power Supply to the Pnl 9-3 controller (FIC-71-

36A) is the Div I ECCS Inverter

c. Power Supply to the Pnl 25-32 Controller (FIC-71-

36B) is also the Div I ECCS Inverter.

d. AT Pnl 25-32, there is a power transfer switch

(XS-256-1) which, if placed in the Alternate

position, will transfer both (36A & B) Flow

Controller power supplies from the Div I ECCS

Inverter to the Unit Preferred 120VAC Power

Supply.

3.

Yokogawa Flow Controller

a. AUTO - output signal is changed by changing the

setpoint. Full Scale travel of setpoint is 40

seconds. Momentary depressing of either the

raise or lower keys will cause ~0.7 gpm change

(~1 %).

Obj. V.B.7.

TP-13

(

Control Room Abandonment

2-AOI-100-2

Unit 2

Rev. 0051

Page 11 of 95

4.2

Unit 2 Subsequent Actions (continued)

NOTES

1)

Attachment 1 provides normal backup control stations and available communications.

2)

Attachment 10 provides PAX extensions and locations.

[7]

ESTABLISH communication with the following personnel and

DIRECT attachments be completed as follows:

U-2 Unit Operator complete Attachment 2, Part A.

U-2 Rx Bldg AUO complete Attachment 3, Part A.

U-2 Turb Bldg AUO complete Attachment 4, Part A.

o

CAUTION

RCIC TURBINE STEAM SUPPLY VALVE, 2-FCV-71-8, transfer switch has been placed in

EMERGENCY and will NOT trip on Reactor Water Level High (+51 inches). Failure to

maintain level below this value may result in equipment damage.

RCIC will still trip on low suction pressure, high turbine exhaust pressure, mechanical

overspeed, and trip push button on pnl 25-32.

[8]

Upon completion of attachments, RE-ESTABLISH

communication using the best available means and continue

procedure.

_A

A.r

o-c-e<<,

I

_~

.J -

~

0

7<r I v D

f'-r=i U'-V lie:. ff,~, CI'\\J I

7-0

7MtVSF61t.

[9]

INITIATE RCIC as follows:

Rc.te..

LoNTjZOLL~

'PDw eY2-

"> CJff>L'j.

[9.1]

At Panel 2-25-32, CHECK OPEN 2-FCV-71-9 (Red Light

above switch) RCIC TURB TRIPITHROT VALVE

RESET,2-HS-71-9D.

0

[9.2]

[9.3]

At 250V DC RMOV Bd 2B, compt. 50, PLACE RCIC

PUMP MIN FLOW VALVE EMER HAND SWITCH ,

2-HS-071-0034C, in OPEN. (Unit 2 Turbine Building

AUO)

At 250V DC RMOV Bd 2C, compt. 4B, PLACE RCIC

TURB STM SUPPLY VALVE EMER HAND SWITCH,

2-HS-071-0008C, in OPEN. (Unit 2 Reactor Building

AUO)

o

Control Room Abandonment

2-AOI-100-2

Unit 2

Rev. 0051

Page 12 of 95

4.2

Unit 2 Subsequent Actions (continued)

NOTE

RCIC Turbine should start and flow should stabilize at 600 gpm.

[904]

[9.5]

[9.6]

[9.7]

At Panel 2-25-32, CHECK turbine speed 2100 rpm or

above using RCIC TURBINE SPEED, 2-SI-71-42B.

At 250V DC RMOV Bd 2B, compt. 50, PLACE RCIC

PUMP MIN FLOW VALVE EMER HAND SWITCH,

2-HS-071-0034C, in CLOSE. (Unit 2 Turbine Building

AUO)

At Panel 2-25-32, ADJUST flowrate as necessary using

RCIC SYSTEM FLOW/CONTROL, 2-FIC-71-36B.

At Panel 2-25-32, MAINTAIN Reactor Water Level

between +2 and +50 inches using RX WATER LEVEL A

& B, 2-L1-3-46A & B.

o

NOTE

The following step prevents HPCI operation and automatic opening of HPCI MAIN PUMP

MINIMUM FLOW VALVE, 2-FCV-73-30.

[10]

At 250V Reactor MOV Bd 2A, PERFORM the following :

[10.1]

Compt. 3D, VERIFY CLOSED HPCI STEAM SUPPLY

VALVE TO TURB FCV-73-16 (MO 23-14).

0

[10.2]

Compt. 3D, PLACE HPCI TURBINE STEAM SUP VLV

TRANS, 2-XS-73-16, in EMERG.

0

[10.3]

IF desired to verify HPCI MIN FLOW BYPASS TO

SUPPRESSION CHAMBER VALVE, 2-FCV-73-30,

closed prior to opening breaker, THEN

DIRECT operator to verify locally.

0

[1004]

Compt. 80, PLACE HPCI MAIN PUMP MIN FLOWVLV

(

FCV-73-30, breaker in OFF.

0

~

(

11. RO 218000K1.05 00l/MEM/T2Gl/lOO-2/5/218000K1.05//RO/SRO/

Given the following plant conditions:

The Unit 1/2 control room has been abandoned.

All MSRV transfer switches at panel 25-32 have been placed in EMERGENCY.

All MSRV control switches at panel 25-32 have been checked in CLOSE.

Which ONE of the following describes the operation of the MSRVs?

A.

The associated ADS valves will open upon receipt of an ADS initiation signal.

B.'" The associated ADS valves will open if their respective pressure relief setpoints are exceeded.

C.

The associated ADS valves will open if their respective control switches on Panel 9-3 are placed in

OPEN.

D.

Any associated ADS valve will open ONLY when its control switch is placed n OPEN.

KIA Statement:

218000 ADS

K1.05 - Knowledge of the physical connections and/or cause- effect relationships between AUTOMATIC

DEPRESSURIZATION SYSTEM and the following: Remote shutdown system: Plant-Specific

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

plant conditions to correctly determine their effect on MSRV operation during a Remote Shutdown

condition.

References: OPL171.208 Rev. 5 page 8 and 2-AOI-100-2 page 8

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 determ ine the following:

1. Transferring the MSRV control to Panel 25-32 disables the ADS function.

2. Transferring the MSRV control to Panel 25-32 disables the Panel 9-3 control switch.

3. Transferring the MSRV control to Panel 25-32 does NOT disable the Pressure Relief function .

A is incorrect. Transferring the MSRV control to Panel 25-32 disables the ADS function .

B is correct.

C is incorrect. Transferring the MSRV control to Panel 25-32 disables the Panel 9-3 control switch.

D is incorrect. Transferring the MSRV control to Panel 25-32 does NOT disable the Pressure Relief

function.

(

Control Room Abandonment

2-AOI-100-2

Unit 2

Rev. 0051

Page 8 of 95

4.2

Unit 2 Subsequent Actions

[1]

IF ALL control rods were NOT fully inserted AND RPS failed to

deenergize, THEN (Otherwise N/A)

DIRECT an operator to Unit 2 Auxiliary Instrument Room to

perform Attachment 11.

o

NOTES

1)

The following transfers Reactor Pressure Control to Panel 2-25-32 to allow for

pressure control while completing the Panel Checklist.

2)

Attachment 9, Alarm Response Procedure Panel 2-25-32, provides for any alarms

associated with this instruction.

CAUTION

Failure to place control switch in desired position prior to transferring to emergency position

may result in inadvertent actuation of the component.

[NER/C) Operation from Panel 2-25-32 bypasses logic and interlocks normally associated with

the components.

[GE SIL 326,51)

[2]

At Panel 2-25-32, PLACE the following MSRV control switches

in CLOSE/AUTO:

Switch No.

Description

2-HS-1-22C

MAIN STM LINE B RELIEF VALVE

0

2-HS-1-5C

MAIN STM LINE A RELIEF VALVE

0

2-HS-1-30C

MAIN STM LINE C RELIEF VALVE

0

2-HS-1-34C

MAIN STM LINE C RELIEF VALVE

0

(

9.

Trip reactor feed pumps as necessary to prevent tripping

on high water level.

OPL171.208

Revision 5

Page 8 of 30

INSTRUCTOR NOTES

Obj. V.B.8

Obj. V.C.5

F.

10.

Start the diesel generators. (9-8 Switch starts respective

units DIG only)

11.

Verify each EECW header has one pump in service.

12.

Announce to all plant personnel that the Control Room is

being evacuated and all operators are to report to their

assigned backup control stations .

13.

Obtain hand held radios from the control room.

14.

Proceed to the Backup Control Panel (25-32)

SUbsequent Actions

1.

If rods failed to fully insert and RPS did not deenergize,

an operator is directed to pull RPS fuses . However, this

is beyond the actual design bases.

See AOI-100-2 for

details for actions

HU Tools: Procedure

Use

Obj V.C.2

See AOI-100-2

Attachment 11

2.

3.

Transfer reactor pressure control to Panel 25-32 to allow

for pressure control while the rest of the panel checklist

is being completed.

Before any transfer switch is placed in EMERGENCY, its

associated control switch must be verified to be in the proper

position . Placing a transfer switch in the EMERGENCY

position enables the local control switch, and the device will

assume the condition called for by the local control switch .

For example, if a transfer switch for an ADS valve is placed

in EMERGENCY with the local control switch in OPEN, the

ADS valve will open.

Note: System Status

prior to abandonment

maintained by GOI-300-

1 checklists.

Obj. V.B.2

Obj. V.B.3.

(

a.

Place the transfer switches for the ADS valves, and

the disconnect switches for the non-ADS valves in

EMERGENCY after making sure the control

switches are in the AUTO position. This action

disables the Control Room hand switches and the

ADS function and is performed to prevent spurious

blowdown of the primary system. The other 3 SRVs

are disabled by opening their breakers on 250VDC

RMOV board 2B(3B) .

Four ADS valves can be controlled from Panel

25-32. Six SRVs (Non-ADS) have only

disconnect switches at Panel 25-32 .

TP-1

Obj. V.B.7

Obj. V.B.8

Obj. V.B.7

I

(

12. RO 218000G2.1.24 00l/MEM/T2G1///218000G2.1.24//BOTH/12/17/2007 RMS

Given the following plant conditions:

Unit-2 is at rated power.

A loss of 2B 250 Volt RMOV Board has occurred.

Which ONE of the following describes the affect on the Unit 2 ADS valves and ADS logic? (Do not

consider the mechanical relief function)

A':! Both Div I & II ADS logic inoperable

No ADS valves will operate automatically

4 ADS valves can still be operated manually.

B. Div I ADS logic inoperable; Div II ADS logic operable

All ADS valves will still actuate automatically.

All ADS valves can still be operated manually.

C.

Div I ADS logic operable, Div II ADS logic inoperable

ADS logic is only capable of opening 3 ADS valves automatically

4 ADS valves can still be operated manually .

D. Both Div I & II ADS logic is still operable

All ADS valves will operate automatically

All ADS valves can be manually operated.

KIA Statement:

218000 ADS

2.1.24 - Conduct of Operations Ability to obtain and interpret station electrical and mechanical drawings

KIA Justification: This question satisfies the KiA statement by requiring the candidate to recall and

interpret the electrical logic drawing of the ADS system to determine the effect of a loss of power to that

logic.

References:

OPL171.043

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. 2B 250V RMOV Board supplies Div 1, 2A 250V RMOV Board supplies Div II. This is the opposite of

conventional logic.

2. 2A 250V RMOV Board supplies only relay power and only in Div II.

3. 2B 250V RMOV Board supplies BOTH Div I and Div II logic.

4. Four ADS valves have an alternate power supply that is NOT 2B 250V RMOV Board.

5. Two ADS valves have no alternate power supply and ONLY powered from 2B 250V RMOV Board.

NOTE: Answers A & B are plausible since conventional logic on other ECCS systems is

"cross-connected" to ensure initiation capability is maintained by either division being energized. Answer

D is plausible if the novice operator uses conventional Division assignments. (A to Div 1 and B to Div II)

A is correct.

B is incorrect. Div II ADS logic is inoperable. No ADS logic is available and only four ADS valves have

power.

C is incorrect. Div I ADS logic is inoperable . No ADS logic is available.

D is incorrect. Div I and Div II logic are inoperable. No ADS logic is available and only four ADS valves

have power.

(

2)

OPL171.043

Revision 12

Page 10 of 30

INSTRUCTOR NOTES

(i)

When pressure has lowered to the

valve reseat pressure (50 psig below

setpoint), the pressure-sensing

stabilizer disc will be unseated by the

pilot disc via the setpoint adjust spring.

This, in turn, causes main piston

chamber repressurization, which

results in closing of the main stage.

Pilot actuation

(a)

DC solenoid admits air pressure to

remote air actuator.

(b)

This unseats the pilot valve disc which

depressurizes the upper main piston

chamber.

(c)

This creates a ~P across the main

valve piston which causes it to move,

against spring tension, opening the

valve.

(d)

The solenoid is actuated by:

i.

Manual demand

(hand switch)

ii.

Automatic blowdown demand

(ADS) for 6 valves which are

controlled by ADS.

iii.

RPV high pressure

(e)

The operating air is supplied from the

drywell control air system.

(f)

The SRV solenoids are powered from

250 VDC RMOV Boards or Battery

Boards. Some SRV power supplies

have relays in the bottom of panel 25-

32 that allow them to swap to an

alternate supply when the normal

supply is lost.

(

(g)

2.

Vacuum breaker

OPL171.043

Revision 12

Page 11 of 30

INSTRUCTOR NOTES

(i)

On Unit 1, SRV's 1-5,1-22,1-

3D, and 1-34 have auto transfer

capabilities (for power supplies)

(ii)

On Unit 2, SRV's 1-5, 1-22, 1-

3D, and 1-34 have auto transfer

capabilities (for power

supplies).

(iii)

On Unit 3, SRV's 1-5, 1-22, 1-

34, and 1-41 have auto transfer

capabilities (for power

supplies).

Loss of air or power to an SRV would

inhibit the relief function but not the

safety function. Per TS 3.4.3 MSRV

operability is based on the safety

function (spring action) and not the

'relief function

a.

Two check valves are provided in each SRV discharge

line to prevent drawing water up into the line due to

steam condensation following termination of valve

operation

b.

Without the vacuum breakers, water in the discharge

lines above suppression pool water level could cause

excessive hydraulic stresses to the T-quenchers and

other torus structural components

3.

Accumulator and check valve arrangement

a.

Only ADS valves are provided with the accumulator

arrangement

TP-1

b.

c.

d.

Accumulators are provided to assure that the ADS

valves can be held open for 30 minutes following a

failure of the air supply to the accumulators

Accumulators are sized to contain sufficient air for that

minimum of five valve operations following a loss of

Drywell Control Air

2/3-EOI Appendix 8G crossties CAD to DWCA

Obj. V.B.2

Obj. V.C.1

Obj. V.D.1

Obj. V.E.2

A CAD supplies 3

valves

B CAD supplies 3

valves

PROCEDURE USE

(

OPL171.043

Revision 12

Page 15 of 31

INSTRUCTOR NOTES

4)

EOls will direct the operator when this action is

FLAGGING

appropriate. Both keylocks must be placed in

inhibit to prevent ADS blowdown

5)

ADS Logic can be inhibited by removing fuses

in Panel 9-30 in Auxiliary Instrument Room.

6)

The fuses for "A" logic are on terminal block "BB

104 & 105" (FU2-1-2EK3)

7)

The "B" logic fuses are on terminal "AA94 & 95"

(FU2-1-2EK13)

8)

The time delay setting is chosen to be long

enough so that HPCI has time to start and yet

not so long that Core Spray and LPCI are

unable to adequately cool the fuel if HPCI

should fail to start

3-WAY

COMMUNICATIONS

6.

e.

The 100 psig and 185 psig ECCS interlocks are

provided to ensure that there is a vessel level

inventory medium available prior to initiating blowdown

of steam from the vessel

ADS Trip Systems

a.

Redundant trip systems from the same power supply

b.

AlC interlock ensures ADS functions when needed

c.

There are two channels in each trip system

1)

A and C in

System I

2)

Band Din

System II

d.

Both channels of a trip system are required to function

to initiate ADS from a given trip system

Obj. V.BA

Obj. V.C.3

Obj. V.CA

Obj. V.D.3

Obj. V.EA

Obj. V.BA

Obj. V.C.5

Obj. V.D.5

Obj. V.E.4

(

e.

This two-channel interlock is called the A-C interlock

and is provided to ensure that all signals to initiate

ADS response are confirmed, thus preventing an ADS

response from an erroneous or failed signal

Obj. VB.5

Obj. V.C.5

Obj. V.D.5

Obj. V.E.5

A Loss of 250V RMOV Bd B would prevent actuation

(

f.

g.

h.

The power supply for the LOGIC and the solenoid

valves is 250VDC

250V RMOV Bd B supplies LOGIC Power for both

system I & II

OPL 171.043

Revision 12

Page 16 of 30

INSTRUCTOR NOTES

All ADS valves

with alternate

power supplies

can be manually

operated from

backup control

panel (25-32)

i.

250V RMOV Bd A supplies Power for relays in system

II of ADS Logic

j.

A Loss of 250V RMOV Bd A would prevent system II

actuation

I.

PCVs 1-19, 1-31 are powered from 250V RMOV Board

2B. There is no alternate power to these valves

k.

PCV 1-22 is powered from 250V RMOV Board 2A with

alternate supply from 250V RMOV Board 2B

See section F. Unit

Differences for U-3

Power Supplies

c...

m.

PCVs 1-5 and 1-34 are normally powered from 250V

RMOV Board 2C with alternate power supply from

Battery Board 1 panel

n.

PCV 1-30 is normally powered from 250V RMOV

Board 2A with a first alternate to 250V RMOV Board

2C and a second alternate to Battery Board 1 panel 7

o.

Valves powered from 250V RMOV Bd 2C required

alternate sources due to RMOV Board 2C not being

environmentally qualified for a line break in secondary

containment

p.

The transfer occurs automatically when undervoltage

DCN 51106

relays (mounted on panel 2-25-32) sense a loss of

power to 250V RMOV Bd B

B.

Instrumentation

1.

SRV discharge piping temperatures are measured by a

multipoint recorder in the Control Room located on Panel 9-

47 (range 0-600°F)

13. RO 223002A2.06 00 lICIA/T2G IIPCIS11223002A3.0 lI/RO/SROI

Given the following plant conditions:

(

During performance of 2-SR-3.3.1.1.13(4A), Reactor Protection and Primary Containment

Isolation Systems Low Reactor Water Level Instrument Channel A1 Calibration, 2-L1S-3-203A

fails to actuate.

(

c

It is determined that the failure is due to an inoperable switch and a replacement is not

available for 4 days.

The Shift Manager has determined that the proper action is to trip the inoperable channel,

only.

Which ONE of the following describes how this is accomplished and the effect on Unit status?

A. >I

Remove fuse 2-FU1-3-203AA associated with 2-L1S-3-203A, a half scram will result and no PCIVs

will realign.

B.

Remove fuse 2-FU1-3-203AA associated with 2-L1S-3-203A, a half scram will result and PCIS

Groups 2, 3 and 6 inboard isolation valves will close.

C.

Place a trip into the ATU associated with 2-L1S-3-203A, no half scram will result and no PCIVs will

realign.

D.

Place a trip into the ATU associated with 2-L1S-3-203A, a half scram will result and PCIS Groups 2,

3 and 6 outboard isolation valves will close.

KIA Statement:

223002 PCIS/Nuclear Steam Supply Shutoff

A2.06 - Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION

SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF ; and (b) based on those predictions, use procedures

to correct, control, or mitigate the consequences of those abn cond or ops. Containment instrumentation

failures

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

plant conditions to determ ine the effect of an instrumentation failure and the corrective actions required as

a result of that failure.

References: 2-01-99 Illustration 3 (page 6 of 11)

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. Recognize the appropriate action is to ensure trip input by de-energizing the level switches.

2. Recognize the effect on RPS logic based on the trip input.

3. Recognize the effect on PCIS logic based on the trip input.

A is correct.

B is incorrect. PCIS logic cuases a "114-isolation"signal BUT no PCIV devices actuate. This is plausible

because the action to ensure the trip input is correct. In addition, the "1/4-isolation" applies to the PCIS

groups identified in the distractor.

C is incorrect. The method of inputting the trip is incorrect. Tripping an ATU cannot be ensured via a

clearance. This is plausible because the RPS and PCIS response is correct.

D is incorrect. The method of inputting the trip is incorrect. Tripping an ATU cannot be ensured via a

clearance. This is plausible because the "1/4-isolation" applies to the PCIS groups identified in the

distractor. This distractor is also similar to answer "A" except it is applied to outboard PCIVs.

(

r>

Reactor Protection System

2-01-99

Unit 2

Rev. 0073

Paae 72 of 77

Illustration 3

(Page 6 of 11)

Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1)

DEVICE

FUSE

RELAY

PANEL

PRINT

ALARMS

REMARKS

2-L1S-3-203A *

2-FU1-3-203AA

2-RLY-099-05AK06A

9-15

2-730E915-9

2-XA-55-4A-2

ALARMS AND 1/2 SCRAM IN CHANNEL A

RXWATER

(5AF6A)

2-RLY-099-5A-K25A

2-730E927-7

NO PCIS DEVICES ACTUATE.

LEVEL LOW

2-RLY-064-16AK5A

2-45E671-26

RX VESSEL WfR LEVEL LOW

(Level 3)

2-RLY-064-16AK6A

A1 CHANNEL

2-XA-55-5B-1

REACTOR CHANNEL A AUTO

1 channel actuated for secondary

Function : 4

containment and CREV initiation

2-L1S-3-203B

2-FU 1-3-203BA

2-RLY-099-05AK06B

9-17

2-730E915-10

2-XA-55-4A-2

ALARMS AND 1/2 SCRAM IN CHANNEL B

RXWATER

(5AF6B)

2-RLY-099-5A-K25B

2-730E927-8

NO PCIS DEVICES ACTUATE.

LEVEL LOW

2-RLY-064-16AK5B

2-45E671-38

RX VESSEL WfR LEVEL LOW

(Level 3)

2-RLY-064-16AK6B

B1 CHANNEL

2-XA-55-5B2

REACTOR CHANNEL B AUTO

1 channel actuated for secondary

Function: 4

containment and CREV initiation

2-L1S-3-203C

2-FU 1-3-203CA

2-RLY-099-05AK06C

9-15

2-730E915-9

2-XA-55-4A-2

ALARMS AND 1/2 SCRAM IN CHANNEL A

RXWATER

(5AF6C)

2-RLY-099-5A-K25C

2-730E927-7

NO PCIS DEVICES ACTUATE.

LEVEL LOW

2-RLY-064-16AK5C

2-45E671-32

RX VESSEL WfR LEVEL LOW

(Level 3)

2-RLY-064-16AK6C

A2 CHANNEL

2-XA-55-5B-1

REACTOR CHANNEL A AUTO

1 channel actuated for secondary

Function: 4

containment and CREV initiation

2-L1S-3-203D

2-FU 1-3-203DA

2-RLY-099-05AK06D

9-17

2-730E915-10

2-XA-55-4A-2

ALARMS AND 1/2 SCRAM IN CHANNEL B

RXWATER

(5AF6D)

2-RLY-099-5A-K25D

2-730E927-8

NO PCIS DEVICES ACTUATE.

LEVEL LOW

2-RLY-064-16AK5D

2-45E671-44

RX VESSEL WfR LEVEL LOW

(Level 3)

2-RLY-064-16AK6D

2-XA-55-5B2

B2 CHANNEL

REACTOR CHANNEL B AUTO

1 channel actuated for secondary

Function: 4

containment and CREV initiation

NOTE:

Device Function corresponds to the TS Table 3.3.1.1 Functions.

~

I

Primary Containment System

2-01-64

Unit2

Rev. 0106

Page 102 of 194

Illustration 2

(Page 1 of 10)

Actions to Place PCIS in Tripped Condition

NOTE

Water level designators (1-8)are listed for relationship to the applicable device only.

(T.S. Tables 3.3.6.1-1,3.3.6.2-1, & 3.3.7.1-1)

~

DEVICE

FUSE

RELAY

PANEL

PRINT

ALARM

REMARKS

2-L1S-3-203A

2-FU1-3-203AA

5AK6A

9-15

2-730E9 15-9

2-XA-55-4A-2

ALARMS AND 1/2 SCRAM IN

RXWATER

(5A-F6A)

5AK25A

2-730E927-7

RX VESSEL WTR LEVEL LOW HALF SCRAM

CHANNELA. CAUSES 1M

LEVEL LOW

16AK5A

2-45E671-26

2-XA-55-5B-1

ISOLATION IN PCIS GROUPS 2,3,

(Level 3)

16A6A

REACTOR CHANNEL A AUTO SCRAM

6 AND 8. NO PCIS DEVICES

ACTUATE.

2-L1S-3-203B

2-FU1-3-203BA

5AK6B

9-17

2-730E915-10

2-XA-55-4A-2

ALARMS AND 1/2 SCRAM IN

RXWATER

(5A-F6B)

5AK25B

2-730E927-8

RX VESSEL WTR LEVEL LOW HALF SCRAM

CHANNEL B. CAUSES 1/4

LEVEL LOW

16AK5B

2-45E671-38

2-XA-55-5B-2

ISOLATION IN PCIS GROUPS 2,3,

(Level 3)

16AK6B

REACTOR CHANNEL B AUTO SCRAM

6 AND 8. NO PCIS DEVICES

ACTUATE.

2-L1S-3-203C

2-FU 1-3-203CA

5AK6C

9-15

2-730E915-9

2-XA-55-4A-2

ALARMS AND 1/2 SCRAM IN

RXWATER

(5A F6C)

5AK25C

2-730E927-7

RX VESSEL WTR LEVEL LOW HALF SCRAM

CHANNELA. CAUSES 1M

LEVEL LOW

16AK5C

2-45E671-32

2-XA-55-5B-1

ISOLATION IN PCIS GROUPS 2,3,

(Level 3)

16AK6C

REACTOR CHANNEL A AUTO SCRAM

6 AND 8. NO PCIS DEVICES

ACTUATE.

2-L1S-3-203D

2-FU1-3-203DA

5AK6D

9-17

2-730E915-10

2-XA-55-4A-2

ALARMS AND 1/2 SCRAM IN

RXWATER

(5A-F6D)

5AK25D

2-730E927-8

RX VESSEL WTR LEVEL LOW HALF SCRAM

CHANNELB. CAUSES 1M

LEVEL LOW

16AK5D

2-45E671-44

2-XA-55-5B-2

ISOLATION IN PCIS GROUPS 2,3,

(Level 3)

16AK6D

REACTOR CHANNEL B AUTO SCRAM

6 AND 8. NO PCIS DEVICES

ACTUATE.

Table 3.3.6.1-1: Function 2a and 5h

Table 3.3.6.2-1: Function 1

Table 3.3.7.1-1: Function 1

(

14. RO 223002A3.01 OOl/C/A/T2Gl/ADS/B5/223002A2.06//RO/SRO/

Given the following plant conditions:

Unit 3 is in Mode 1 with 4 Bypass valves open.

3-SI-3.4.3.2 "Main Steam Relief Valve Manual Cycle Test" is in progress.

The unit operator performing the test notices that the 3-FCV-1-5, which was just cycled 1

minute eariler, has lost it's indication lights.

The outside US is dispatched and reports that the troubleshooting indicates that a ground in

the normal feeder breaker from 250V RMOV Bd 3C to the 3-FCV-1-5 SRV is causing the

breaker to trip, all other circuits associated with the SRV are functional and normal.

Regarding the 3-FCV-1-5, which ONE of the following statements describes the result of a loss of it's

normal power source?

A.

3-FCV-1-5 cannot be controlled from 25-32 and will automatically transfer to an alternate power

source but will NOT retain it's operability for SRV safety relief mode (non ADS).

B."; 3-FCV-1-5 can be controlled from panel 25-32 and auto transfers to an alternate power source and

WILL retain it's operability for SRV safety relief mode (non ADS) .

C.

3-FCV-1-5 can be controlled from panel 25-32 and can be manually transferred to an alternate

power source but will NOT retain it's operability for SRV safety relief mode (non ADS).

D.

3-FCV-1-5 cannot be controlled from 25-32 but it can be manually transfered to an alternate power

source and WILL retain it's operabity for SRV safety relief mode (non ADS).

KIA Statement:

223002 PCIS/Nuclear Steam Supply Shutoff

A3.01 - Ability to monitor automatic operations of the PRIMARY CONTAINMENT ISOLATION

SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including : System indicating lights and alarms.

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

plant conditions to correctly determine the effect a loss of indication has on MSRV operability.

References:

3-AOI-100-2, OPL 171.043

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 MSRV 1-5 can be controlled from Panel 25-32 on Unit-3.

2. Whether MSRV 1-5 has an alternate power supply available.

3. Whether MSRV 1-5 will AUTO transfer to Alternate or must be manually transferred.

4. Recognize that electrical power is sufficient for Safety Relief Mode OPERABILITY.

A is incorrect. 3-FCV-1-5 can be controlled from 25-32. In addition, the valve will retain it's operability for

Safety Relief Mode. This is plausible because the valve automatically transfers to Alternate.

B is correct.

C is incorrect. 3-FCV-1-5 automatically transfers to alternate upon a loss of normal power. In addition,

the valve will retain it's operability for Safety Relief Mode. This is plausible because the valve CAN be

controlled from 25-32.

D is incorrect. 3-FCV-1-5 can be controlled from 25-32. In addition, the valve automatically transfers to

alternate upon a loss of normal power. This is plausible because the valve will retain it's operability for

Safety Relief Mode.

(

OPL171.043

Revision 12

Page 16 of 30

INSTRUCTOR NOTES

f.

The power supply for the LOGIC and the solenoid

valves is 250VDC

g.

250V RMOV Bd B supplies LOGIC Power for both

system I & II

h.

A Loss of 250V RMOV Bd B would prevent actuation

All ADS valves

with alternate

power supplies

can be manually

operated from

backup control

panel (25-32)

i.

250V RMOV Bd A supplies Power for relays in system

II of ADS Logic

j.

A Loss of 250V RMOV Bd A would prevent system II

actuation

I.

PCVs 1-19, 1-31 are powered from 250V RMOV Board

28. There is no alternate power to these valves

k.

PCV 1-22 is powered from 250V RMOV Board 2A with

alternate supply from 250V RMOV Board 2B

See section F. Unit

Differences for U-3

Power Supplies

(

m.

PCVs 1-5 and 1-34 are normally powered from 250V

RMOV Board 2C with alternate power supply from

Battery Board 1 panel

n.

PCV 1-30 is normally powered from 250V RMOV

Board 2A with a first alternate to 250V RMOV Board

2C and a second alternate to Battery Board 1 panel 7

o.

Valves powered from 250V RMOV Bd 2C required

alternate sources due to RMOV Board 2C not being

environmentally qualified for a line break in secondary

containment

p.

The transfer occurs automatically when undervoltage

DCN 51106

relays (mounted on panel 2-25-32) sense a loss of

power to 250V RMOV Bd 2

B.

Instrumentation

1.

SRV discharge piping temperatures are measured by a

multipoint recorder in the Control Room located on Panel 9-

47 (range 0-600°F)

(

Control Room Abandonment

3-AOI-100-2

Unit 3

Rev. 0017

Page 7 of90

Date

_

4.2

Unit 3 Subsequent Actions

[1]

IF ALL control rods were NOT fully inserted AND RPS failed to

deenergize, THEN (Otherwise N/A)

DIRECT an operator to Unit 3 Auxiliary Instrument Room to

perform Attachment 9.

CAUTIONS

o

1)

Failure to place control switch in desired position prior to transferring to emergency

position may result in inadvertent actuation of the component.

2)

(NERlC] Operation from Panel 3-25-32 bypasses logic and interlocks normally associated

with the components.

[GE SIL 326,51)

NOTES

1)

The following transfers Reactor Pressure Control to Panel 3-25-32 to allow for

pressure control while completing the Panel Checklist.

2)

Attachment 7, Alarm Response Procedure Panel 3-25-32, provides for any alarms

associated with this instruction.

[2]

PLACE the following MSRV control switches in CLOSE/AUTO

at Panel 3-25-32 :

0

Switch No.

Description

(-.J)

3-HS-1-22C

MAIN STM LINE B RELIEF VALVE

0

3-HS-1-5C

MAIN STM LINE A RELIEF VALVE

0

3-HS-1-41C

MAIN STM LINE D RELIEF VALVE

0

3-HS-1-34C

MAIN STM LINE C RELIEF VALVE

0

(

Control Room Abandonment

3-AOI-100-2

Unit3

Rev. 0017

Page 8 of 90

Date

4.2

Unit 3 Subsequent Actions (continued)

[3]

PLACE the following MSRV disconnect switches in DISCT at

Panel 3-25-32:

0

Switch No.

Description

(-.J)

3-XS-1-4

MAIN STM LINE A RELIEF VALVE DISCT

0

3-XS-1-42

MAIN STM LINE D RELIEF VALVE DISCT

0

3-XS-1-23

MAIN STM LINE B RELIEF VALVE DISCT

0

3-XS-1-30

MAIN STM LINE C RELIEF VALVE DISCT

0

3-XS-1-180

MAIN STM LINE D RELIEF VALVE DISCT

0

[4]

PLACE the following MSRV transfer switches in EMERG at

Panel 3-25-32:

0

Switch No.

Description

(-.J)

3-XS-1-22

MAIN STM LINE B RELIEF VALVE XFR

0

3-XS-1-5

MAIN STM LINE A RELIEF VALVE XFR

0

3-XS-1-41

MAIN STM LINE D RELIEF VALVE XFR

0

3-XS-1-34

MAIN STM LINE C RELIEF VALVE XFR

0

NOTE

Use of the following sequence when opening MSRVs should distribute heat evenly in the

Suppression Pool.

[5]

MAINTAIN Reactor Pressure between 800 and 1000 psig

using the following sequence at Panel 3-25-32:

0

A.

3-HS-1-22C, MAIN STM LINE B RELIEF VALVE

0

B.

3-HS-1-5C, MAIN STM LINE A RELIEF VALVE

0

(

C.

3-HS-1-41C, MAIN STM LINE D RELIEF VALVE

0

D.

3-HS-1-34C, MAIN STM LINE C RELIEF VALVE

0

(

15. RO 239002A3.03 OOl/C/A/T2GlIMAIN STEAM/C/A/239002A3.03/IRO/SRO/

Given the following plant conditions:

The reactor is operating at 100% power and 1000 psig.

A turbine control valve malfunction resulted in reactor safety relief valve (SRV) 1-4 lifting and

failing to reseal.

Which ONE of the following describes the expected SRV tailpipe temperature?

REFERENCE PROVIDED

KIA Statement:

239002 SRVs

A3.03 - Ability to monitor automatic operations of the RELIEF/SAFETY VALVES including: Tail pipe

temperatures

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

plant conditions to determine the expected tailpipe temperature of an open MSRV using steam tables.

References:

Steam Tables

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: Steam Tables

Plausibility Analysis:

In order to answer this question correctly the candidate must:

1. Use the Steam Table Mollier Diagram to determine the correct process and temperature for an

open MSRV.

NOTE: This question is typical for a GFES examination, however the KIA provides little lattitude for a

question with discriminatory value based on reading a multi-point recorder. In addition, with it's direct

connection to an issue identified following the accident at TMI, the importance of understanding this

process becomes self-evident.

A is incorrect. This temperature is indicative of saturation temperature for steam at tailpipe pressure

(atmospheric).

B is correct. This is a throttling process and is therefore isoenthalpic.

C is incorrect. 340°F would be incorrectly determined if the candidate considered the process to be

isoenthalpic to the saturation line, then followed the constant superheat line to atmospheric pressure.

D is incorrect. This temperature is indicative of saturation temperature for reactor pressure.

E

MINATION

REFERENCE

PROVIDED TO

CANDIDATE

C

Combustion Engineering Steam Tables

..

(

16. RO 239002A4.08 OOl/C/A/TIGl/32A-l/4/239002A4 .08/IRO/SRO/

Given the following plant conditions:

Unit 2 is operating at 100% power.

A complete loss of Drywell Control Air occurs (both headers).

NEITHER crosstie with CAD nor plant Control Air can restore system pressure.

Which ONE of the following statements describes the effect on pneumatically operated valves inside the

Primary Containment in accordance with 2-AOI-32A-1 , Loss of Drywell Control Air?

A.

All inboard MSIVs can still be cycled once.

B.

All MSRV's can still be cycled five times.

C.

All inboard MSIVs can still be cycled with the test switch.

D." ADS MSRVs can still be cycled five times.

KJA Statement:

239002 SRVs

A4.0B - Ability to manually operate and/or monitor in the control room: Plant air system pressure :

Plant-Specific

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

plant conditions to determine the operability of MSRVs following a loss of pneumatic supply .

References: 2-AOI-32A-1

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. Recognize that ADS MSRV accumulators allow for five cycle operations.

2. Recognize that Inboard MSIVs are capable of being CLOSED one time, but not CYCLED one time.

3. Recognize that using the TEST switch to close an MSIV has no impact on the AMOUNT of

pneumatic pressure required.

A is incorrect. MSIVs can be CLOSED once, but not CYCLED. This is plausible because the

accumulator does not fully discharge with one closure, but there is insufficient pressure remaining to

overcome the spring pressure to open the MSIV.

B is incorrect. Only ADS MSRVs have accumulators sufficient to cycle five times. The remaining

MSRVs will not function without a pneumatic supply.

C is incorrect. MSIVs can be CLOSED once, but not CYCLED. Using the TEST switch to close an MSIV

has no impact on the AMOUNT of pneumatic pressure required.

D is correct.

(

Loss of Drywell Control Air

2-AOI-32A-1

Unit 2

Rev. 0021

Page 5 of 9

4.0

OPERATOR ACTIONS

4.1

Immediate Actions

None

4.2

SUbsequent Actions

[1]

IF ANY EOI entry condition is met, THEN

ENTER the appropriate EOI(s).

NOTES

o

1)

The MSIV air accumulators are designed to provide for one closing actuation following

loss of air supply. Once closed the valve is held closed by the springs.

2)

The ADS MSRV air accumulators are provided to assure that the valves can be held

open following failure of the air supply to the accumulators, and they are sized to

contain sufficient air for a minimum of five valve operations. 0 peratlons of the AGS

MSR\\i' should be limited to 5 times.

3)

Nitrogen Tanks supply pressurized nitrogen to the Drywell Control Air System via the

DWCA SUPPLY REGULATORS 2-PREG-32-49A and 2-PREG-32-49A (lead regulator

will be set at 100 psig and backup regulator set at 5-8 psig lower)

4)

DWCA NITROGEN REG STATION BYPASS VLV, 2-BYV-032-0141 can be used to

maintain approximately 98 psig in DWCA Receiver Tanks A & B when required by

plant conditions

.

[2]

CHECK Drywell Control Air System operating properly.

(

REFER TO 2-01-32A.

[3]

IF Operation with DWCA Nitrogen Regulattion Bypass Valve

OpenlThrottled is required, THEN

REFER TO 2-01-32A.

o

(

17. RO 259002A4.03 00lIe/A/TIG lIOI-3 //259002A4.03/IRO/SRO/lll28/07 RMS

In accordance with 1-GOI-100-1A, Unit Startup, the RFPT is not placed in AUTOMATIC control until

____ to prevent.

_

A.

power is above 15%, RPV level oscillations due to low steam flow vs. feed flow error signals.

B. the Mode switch is in RUN, an uncontrolled reactivity insertion.

C~ power is above 15%, an uncontrolled reactivity insertion.

D.

the Mode switch is in RUN, RPV level oscillations due to low steam flow vs. feed flow error signals.

KIA Statement:

259002 Reactor Water Level Control

A4.03 - Ability to manually operate and/or monitor in the control room: All individual component controllers

when transferring from manual to automatic modes

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

plant conditions to determine the condition and basis for transferring reactor water level control to

automatic operation.

References:

1-GOI-100-1A

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. Recognize the power level or plant condition when it is appropriate to place RFPs in AUTOMATIC.

2. The basis for establishing the required condition is a reactivity control issue.

A is incorrect. This is plausible because the procedural limit is correct. In addition, maintaining steady

RPV level at low steam flows has been an issue prior to more advanced electronic control systems

becoming available.

B is incorrect. The procedural limit is incorrect. This is plausible because the basis is correct. In

addition, previous revisions to 1-GOI-100-1A had the RFPs placed in AUTOMATIC after placing the Mode

Switch in RUN.

C is correct.

D is incorrect. The procedural limit is incorrect. This is plausible because the basis is correct. In

addition, previous revisions to 1-GOI-100-1A had the RFPs placed in AUTOMATIC after placing the Mode

Switch in RUN. In addition, maintaining steady RPV level at low steam flows has been an issue prior to

more advanced electronic control systems becoming available.

(

Unit Startup

1-GOI-100-1A

Unit 1

Rev. 0011

Page 115 of 173

5.0

INSTRUCTION STEPS (continued)

MODE/CONDITION CHANGE

NOTES

1)

Drywell to Torus differential pressure must be established within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after

reaching 15% RTP per Tech Specs Section 3.6.2.6. (1-01-64) .

2)

Primary Containment must be inerted within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of reaching 15% RTP per Tech Specs Section 3.6.3.2. (1-01-76).

[81]

WHEN Reactor is at 15% RTP, THEN

RECORD the time 15% RTP was obtained in the NOMS Narrative Log.

Initials

Date

Time

ENTER 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months LCO for Drywell to Suppression Pool Differential

Pressure. REFER TO Tech Specs LCO 3.6.2.6. (N/A if Drywell to

Suppression Pool Differential Pressure already established)

Initials

Date

Time

ENTER 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months LCO for Primary Containment Oxygen Concentration.

REFER TO Tech Specs LCO 3.6.3.2. (N/A if Primary Containment is

already inerted)

Initials

Date

Time

RECORD Time LCO entered. (N/A if no LCO entry is required.)

Date -----

Initials

Time

Date

Time

(

Unit Startup

1-GOI-100-1A

Unit 1

Rev. 0011

Page 116 of 173

5.0

INSTRUCTION STEPS (continued)

CAUTIONS

1)

Failure to monitor SJAE/OG CNDR CNDS FLOW, 1-FI-2-42, on Panel 1-9-6 for proper

flow may result in SJAE isolation.

2)

Changes in condensate system flow may require adjustment to SPE CNDS BYPASS,

1-FCV-002-0190, either in the Control Room or locally. Personnel adjusting this valve

locally shall be in direct communication with the Control Room.

[82]

WHEN stable operation can be maintained, THEN

PLACE operating RFPT in automatic operation. REFER TO 1-01-3.

Initials

Date

Time

[83]

PERFORM the following for IRMs:

[83.1]

WITHDRAW all operable IRMs.

Time

Date

Initials

(R)

_

[83.2]

PLACE all range switches to a position such that associated alarms are

reset.

Time

Date

Initials

(R)

--

[83.3]

VERIFY alllRM upscale or downscale alarms are reset.

Time

Date

Initials

(R)

_

[83.4]

VERIFY IRM recorder High Alarm setpoint programmed OFF.

Initials

Date

1M

Time

(

18. RO 261000K3.06 OOl/C/A/TIGl/CONT/PRIN.B.8/261000K3.06/IRO/SRO/

Unit-2 has experienced a LOCA with the following plant conditions:

Drywell pressure is 50 psig and rising.

Drywell O2 concentration is 16%.

Drywell H2 concentration is 5%.

The Drywell is being vented through SGT "A" train.

SGT "B" and "C" are unavailable and INOP .

Which ONE of the following can be used to exhaust primary containment atmosphere if SGT "A" were to

become INOPERABLE?

A."

Vent the Suppression Chamber via the HARDENED SUPPR CHBR VENT in accordance with

2-EOI Appendix 13, Emergency Venting Primary Containment.

B.

Vent the Drywell in accordance with 2-EOI Appendix 13, Emergency Venting Primary Containment,

allowing the primary containment vent ducts to fail.

c.

Vent the Suppression Chamber in accordance with 2-EOI Appendix 13, Emergency Venting

Primary Containment, allowing the primary containment vent ducts to fail.

D.

Vent the Suppression Chamber in accordance with 2-AOI-64-1, Drywell Pressure and/or

Temperature High, or Excessive Leakage into Drywell.

KJA Statement:

261000 SGTS

K3.06 - Knowledge of the effect that a loss or malfunction of the STANDBY GAS TREATMENT SYSTEM

will have on following: Primary containment oxygen content: Mark-I&II

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

knowledge of the relationship between SGT and the inerting process.

References:

2-01-76, Containment Inerting System, 2-EOI Appendix 13, Emergency Venting Primary

Containment

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. Drywell pressure is approaching the 55 psig pressure limit.

2. Hydrogen an oxygen have reached explosive concentrations.

3. Recognize which available vent path does NOT require SGT to be OPERABLE.

4. Recognize that venting the Suppression Chamber is preferred over the Drywell to facilitate scrubbing

of radioactive fission products.

NOTE: The following distractors are plausible because the vent lineups are physically possible.

A is correct.

B is incorrect. Venting the Suppression Chamber is preferred over the Drywell to facilitate scrubbing

of radioactive fission products. In addition, SGT is required for that vent path.

C is incorrect. Destructive venting of the Suppression Chamber is physically possible, but not

procedurally authorized.

D is incorrect. Venting per 2-AOI-64-1 requires SGT Operability.

(

2-EOI APPENDIX-13

Re v .

5

Page 1 of 6

...

2-EOI APPENDIX-13

EMERGENCY VENTING PRIMARY CONTAINMENT

LOCATION:

Unit 2 Control Room

ATTACHMENTS:

1.Tools and Equipment

2.Vent System Overview

3.Hardened Vent Flow Path

(

1.

NOTIFY SHIFT MNGR./SED of the following:

Emergency Venting of Primary Containment is in

progress.

Off-Gas Release Rate Limits will be exceeded.

2.

VENT the Suppression Chamber as follows

(Panel 9-3):

a.

IF

EITHER of the following exists:

Suppression Pool water level CANNOT be determined

to be below 20 ft,

OR

Suppression Chamber CANNOT be vented,

THEN ..... CONTINUE in this procedure at Step 3.

b.

PLACE keylock switch 2-HS-64-222B,

HARDENED SUPPR

CHBR VENT OUTBD PERMISSIVE, in PERM.

c.

CHECK blue indicating light above 2-HS-64-222B,

HARDENED SUPPR CHBR VENT OUTBD PERMISSIVE,

illuminated.

d.

OPEN 2-FCV-64-222,

HARDENED SUPPR CHBR VENT OUTBD

ISOL VLV.

e.

PLACE keylock switch 2-HS-64-221B,

HARDENED SUPPR

CHBR VENT

INBD PERMISSIVE,

in PERM.

f.

CHECK blue indicating light above 2-HS-64-221B,

HARDENED SUPPR CHBR VENT INBD PERMISSIVE,

illuminated.

g.

OPEN 2-FCV-64-221,

HARDENED SUPPR CHBR VENT

INBD

ISOL VLV.

(

2.

2-EOI APPENDIX-13

Rev.

5

Paqe 2 of 6

(continued from previous page)

h.

CHECK Drywell and Suppression Chamber Pressure

lowering.

i.

MAINTAIN Primary Containment Pressure below 55 psig

using 2-FCV-64-222,

HARDENED SUPR CHBR VENT OUTBD

ISOL VLV, as directed by SRO.

3.

IF

Suppression Chamber vent path is

NOT available,

THEN

VENT the Drywell as follows:

a.

NOTIFY SHIFT MNGR./SED that Secondary Containment

integrity failure is possible.

b.

NOTIFY RADCON that Reactor Building is being

evacuated due to imminent failure of Primary

Containment vent ducts.

c.

EVACUATE ALL Reactor Buildings using P.A.

System.

d.

START ALL available SGTS trains.

e.

VERIFY CLOSED 2-FCV-64-36,

DW/SUPPR CHBR VENT TO SGT

(Panel 9-3).

f.

VERIFY OPEN the following dampers

(Panel 9-25):

2-FCO-64-40,

REACTOR ZONE EXH TO SGTS

2-FCO-64-41,

REACTOR ZONE EXH TO SGTS.

g.

VERIFY CLOSED 2-FCV-64-29,

DRYWELL VENT

INBD ISOL

VALVE

(Pane l

9-3 or Panel 9-54).

h.

DISPATCH personnel to Unit 2 Auxiliary Instrument

Room to perform the following:

1)

REFER TO Attachment 1 and OBTAIN one 12-in.

Banana Jack Jumper from EOI Equipment Storage

Box.

2)

LOCATE terminal strip

DO in Panel 9-43,

Front.

3)

JUMPER 00-76 to 00-77

(Panel 9-43).

4)

NOTIFY Unit Operator that jumper for 2-FCV-64-

30,

DRYWELL VENT OUTBD ISOLATION VLV, is in

place.

i.

VERIFY OPEN 2-FCV-64-30,

DRYWELL VENT OUTBD SOLATION

VLV

(Panel 9-3).

""

~

TO REACTOR BLDG

EXHAUST FANS

VENT SYSTEM OVERVIEW

64-29

18"

64-30

co

C'"l

...rco.

I

II

J::'O-O:;ON

I-3OJ(J)

I

1-3 \\Q

<1

t':l

J::' (J).

0

()

H

r:U1U1

3:

J::'

t':l 0

0-0

ZH1

0-0

1-3

t':l

(j)

Z

N

t:l

H

><:

I

f--'

W

DRYWELL

- -1

---

I

I ~

I

Ul

t.,

~

,

w

,

en

I,,

I,,,

_ _ I

--- - --

f---~~

o

t-

o

w

U1

I

_________ _ J

~

~w

~

W

.".

! I--§9

~

2"

,..---------

I

64-32

18"

!~--§9t

co+

0

o

2"

a

64-33

DRYWELL PURGE

FANS AND FILTERS

I

REACTOR BLDG

VENTILATION

EXHAUST

(

2-EOI APPENDI X-13

Rev.

5

Page

6 of 6

ATTACHMENT 3

Sl

~

9

~

s

~

..I

LL

~

I-

Z

~

W>

CW

~

Z

W

~

C

0::

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~

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(

LAST PAGE

(

Drywell Pressure and/or Temperature

2-AOI-64-1

Unit2

High, or Excessive Leakage into

Rev. 0023

Drywell

Page 7 of 12

4.0

OPERATOR ACTIONS

NOTE

This procedure covers possible multiple symptoms of a problem within primary

containment. Any or all of the symptoms may exist. The SRO will direct actions based on

symptoms and experience.

4.1

Immediate Actions

None

4.2

Subsequent Actions

[1]

IF any EOI entry condition is met, THEN

ENTER appropriate EOI(s). (Otherwise N/A)

[2]

IF Drywell Pressure is High, THEN

PERFORM the following: (Otherwise N/A)

o

[2.1 ]

[2.2]

CHECK Drywell pressure using multiple indications.

ALIGN and START additional Drywell coolers and fans

as necessary. REFER TO 2-01-64.

o

CAUTION

Stack release rates exceeding 1.4 X 107 uci/sec, or a SI-4.8.B.1.a.1 release fraction above

one will result in ODCM release limits being exceeded.

[2.3]

VENT Drywell as follows:

(

[2.3.1]

[2.3.2]

[2.3.3]

CLOSE SUPPR CHBR INBD ISOLATION VLV

2-FCV-64-34 (Panel 2-9-3).

VERIFY OPEN, DRYWELL INBD ISOLATION VLV,

2-FCV-64-31 (Panel 2-9-3).

VERIFY 2-FIC-84-20 is in AUTO and SET at

100 scfm (Panel 2-9-55).

o