ML023400575

A. Incorrect since the Tech Spec setpoint is < 150 F.

B. Correct answer since isolation capability is not maintained since all channels are

C. Incorrect since isolation capability is not maintained per Bases definition.

D. Incorrect since the instruments should be declared INOPERABLE immediately.

RO Tier:

SRO Tier:

T2G2

Keyword:

RWCU ISOLATION

Cog Level:

C/A 2.6/2.6

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

2

Friday, October 11, 2002 06:51:25 AM

QUESTIONS REPORT

for Revision2 HT2002

30. 204000K5.08 001

Unit I is at 100% RTP. The I & C Techs have just completed the quarterly functional

surveillance for the RWCU Area High Temperature isolation instruments. The foreman

..

-

ý,t-

.--

,

oIn+nninfc fnr all the area s. were set

is reviewing the paperwork andu nIotVI

Lild soLi.,

a,

,o, ....

a.

.....

to 1550F. He immediately notifies the Shift Supervisor.

Which ONE of the following describes the determination the Shift Supervisor should

make? (Provide TS Section 3.3.6.1 and Table 3.3.6.1-1)

A. This is not a problem because the setpoint per Tech Specs is < 160 0F.

B.' This is a problem and all of the instruments are INOPERABLE. The RWCU system

isolation capability must be restored within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

C. This is a problem and all of the instruments are INOPERABLE. Each channel must

be placed in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of INOPERABILITY.

D. This is not a problem if the I & C Techs can recalibrate the instruments within

tolerance provided the surveillance frequency hasn't expired.

References: Tech Spec section 3.3.6.1

Tech Spec Table 3.3.6.1-1

A. Incorrect since the Tech Spec setpoint is < 150 F.

B. Correct answer since isolation capability is not maintained since all channels are

C. Incorrect since isolation capability is not maintained per Bases definition.

D. Incorrect since the instruments should be declared

RO Tier:

SRO Tier:

Keyword:

RWCU ISOLATION

Cog Level:

Source:

N

Exam:

Test:

S

Misc:

Friday, September 20, 2002 09:23:23 AM

INOPERABLE immediately.

T2G2

C/A 2.6/2.6

HT02301

TCK

33

QUESTIONS REPORT

for HT2002

8. 204000K5.08 001

Unit I is at 100% RTP. The Instrument Maintenance Techs have just completed the

quarterly functional surveillance for the RWCU Area High Temperature isolation

instruments. The foreman is reviewing the paperwork and notes that isolation

setpoints for all the areas were set to 155 0F. He immediately notifies the Shift

Supervisor.

Which ONE of the following describes the determination theShift Supervisor should

make?

A. This is not a problem because the setpoint per Tech Specs is < 160 0F.

Bf This is a problem and all of the instruments are INOPERABLE. The RWCU system

isolation capability must be restored within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

C. This is a problem and all of the instruments are INOPERABLE. At least one

channel must be placed in the tripped condition witnin 1z hours.

D. This is not a problem if the Instrument Techs can recalibrate the instruments within

tolerance provided the surveillance frequency hasn't expired.

References: Tech Spec section 3.3.6.1

Tech Spec Table 3.3.6.1-1

A. incorrect since the Tech Spec setpoint is < 150 F.

B. Correct answer. Isolation capability is not maintained because no channels are

OPERABLE and none are in trip at this time.

C. Incorrect since Condition A allows 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to place a channel in trip.

D. Incorrect since the instruments should be declared

RO Tier:

SRO Tier:

Keyword:

RWCU ISOLATION

Cog Level:

Source:

N

Exam:

Test:

S

Misc:

INOPERABLE immediately.

T2G2

C/A 2.6/2.6

HT02301

TCK

Monday, June 24, 2002 08:16:46 AM

8

QUESTIONS REPORT

for HT2002

1. 204000K5.08 001

Unit 1 is at 100% RTP. The Instrument Maintenance Techs have just completed the

quarterly functional surveillance for the RWCU Area High Temperature isolation

instruments. The foreman is reviewing the paperwork and notes that isolation

setpoints for all the areas were set to 1551F. He notifies the Shift Supervisor

immediately.

Which ONE of the following describes the determination theShift Supervisor should

make?

A. This is not a problem because the setpoint per Tech Specs is < 1600F.

B.f This is a problem and all of the instruments are INOPERABLE. The RWCU system

isolation capability must be restored within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

C. This is a problem and all of the instruments are INOPERABLE. At least one

channel must be placed in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D. This is not a problem if the Instrument Techs can recalibrate the instruments within

tolerance provided the surveillance frequency hasn't expired.

References: Tech Spec section 3.3.6.1

Tech Spec Table 3.3.6.1-1

Tech Spec section 3.4.7, RHR Shutdown Cooling

A. Incorrect since the Tech Spec setpoint is < 150 F.

B. Correct answer. Isolation capability is not maintained because no channels are

OPERABLE and none are in trip at this time.

C. Incorrect since Condition A allows 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to place a channel in trip.

D. Incorrect since the instruments should be declared INOPERABLE immediately.

1

Friday, May 31, 2002 07:22:22 AM

Primary Containment Isolation Instrumentation

3.3.6.1

Table 3.3.6.1-1 (page 4 of 4)

Primary Containmnent Isolation Instrumrentation

APPLICABLE

CONDITIONS

MODES OR

REQUIRED

REFERENCED

OTHER

CHANNELS

FROM

SPECIFIED

PER TRIP

REQUIRED

SURVEILLANCE

ALLOWABLE

FUNCTION

CONDITIONS

SYSTEM

ACTION C.1

REQUIREMENTS

VALUE

4.

RCIC System Isolation

(continued)

g.

RCIC Suppression Pool

Area Differential

Temperature - High

h.

Emergency Area Cooler

Temperature - High

5.

RWCU System Isolation

a.

Area

Temperature - High

b.

Area Ventilation

Differential

Temperature - High

c.

SLC System Initiation

d.

Reactor Vessel Water

Level - Low Low,

Level 2

6.

RHR Shutdown Cooling

System Isolation

a.

Reactor Steam Dome

Pressure - High

b.

Reactor Vessel Water

Level - Low, Level 3

1,2,3

1,2

1,2,3

3,4,5

I

F

1 per

area

1 per

area

11(c)

SR

F

H

2

1

2 (d)

F

SR

F

SR

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

SR

3.3.6.1.6

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

Amendment No. 135

S 420F

S 169OF

5 1500 F

S67*F

NA

> -47 inches

Z 145 psig

Ž 0 inches

(c)

SLC System Initiation only inputs into one of the two trip systems.

(d) Only one trip system required in MODES 4 and 5 when RHR Shutdown Cooling System integrity maintained.

I

3.3-59

HATCH UNIT 2

Primary Containment Isolation Instrumentation

3.3.6.1

3.3

INSTRUMENTATION

3.3.6.1

Primary Containment Isolation Instrumentation

LCO

3.3.6.1

APPLICABILITY:

The primary containment isolation instrumentation

Function in Table 3.3.6.1-1 shall be OPERABLE.

According to Table 3.3.6.1-1.

for each

ACTIONS

-NOTE-

NOTE

Separate Condition entry is allowed for each channel.

CONDITION

REQUIRED ACTION

COMPLETION TIME

A.

One or more required

A.1

Place channel in

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for

channels inoperable,

trip.

Functions 2,a,

2.b, and 6.b

AND

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for

Functions other

than Functions

2.a, 2.b, and

6.b

B. ---------NOTE --------

8.1

Restore isolation

I hour

Not applicable for

capability.

Function 5.c.

One or more automatic

Functions with

isolation capability

not maintained.

(continued)

4

Amendment No.

135

if

1

3.3-52

HATCH UNIT 2

Primary Containment Isolation Instrumentation

B 3.3.6.1

BASES

ACTIONS

A.1 (continued)

the channel in trip (e.g., as in the case where placing the

inoperable channel in trip would result in an isolation),

Condition C must be entered and its Required Action taken.

B.1

Required Action B.1 is intended to ensure that appropriate

actions are taken if multiple, inoperable, untripped

channels within the same Function result in automatic

isolation capability being lost for the associated

penetration flow path(s).

The MSL Isolation Functions are

considered to be maintaining isolation capability when

sufficient channels are OPERABLE or in trip, such that both

trip systems will generate a trip signal from the given

Function on a valid signal.

The other isolation functions

are considered to be maintaining isolation capability when

sufficient channels are OPERABLE or in trip, such that one

trip system will generate a trip signal from the given

Function on a valid signal.

This ensures that one of the

two PCIVs in the associated penetration flow path can

receive an isolation signal from the given Function.

As

noted, this Condition is not applicable for Function 5.c

(SLC System Initiation), since the loss of the single

channel results in a loss of the Function (one-out-of-one

logic).

This loss was considered during the development of

Reference 5 and considered acceptable for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

allowed by Required Action A.1.

The Completion Time is intended to allow the operator time

to evaluate and repair any discovered inoperabilities.

The

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes

risk while allowing time for restoration or tripping of

channels.

C.I

Required Action C.1 directs entry into the appropriate

Condition referenced in Table 3.3.6.1-1.

The applicable

Condition specified in Table 3.3.6.1-1 is Function and

MODE or other specified condition dependent and may change

as the Required Action of a previous Condition is completed.

(continued)

REVISION I

B 3.3-169

HATCH UNIT 2

QUESTIONS REPORT

for HT2002

9. 205000G2.1.22 001

Unit 2 is shutting down for a maintenance outage due to the failure of the "B" Recirc

pump which is out-of-service electrically. At 0100 on 4/12/02 reactor pressure went

below 145 psig and reactor temperature went below 300 0F. The following conditions

exist at 0400 on 4/12/02:

Reactor pressure

130 psig

Reactor temperature

285°F

Mode Switch position

S/D

At 0415 on 4/12/02 the "A" Recirc pump tripped and cannot be restarted due to bus

overcurrent.

Which ONE of the following is required to be taken per Tech Specs?

(Provide copy of Tech Spec sections 3.3.6.1, 3.4.1, 3.4.7)

A. No action is required to be taken since Recirc Pumps are only required to be in

operation in Modes 1 and 2.

B. Initiate action to place Shutdown Cooling in operation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND monitor

reactor coolant temperature and pressure once per hour.

-/

C. Initiate action to restore Shutdown Cooling to OPERABLE status immediately AND

verify reactor coolant circulation by an alternate method within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from

discovery of no reactor coolant circulation AND be in Mode 4 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D' No action is required for up 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months at which time a Recirc Pump must be running

or Shutdown Cooling must be in operation.

n9

Monday, June 24, LUUt vo: i0,uv inVl

QUESTIONS REPORT

for HT2002

References: Tech Spec section 3.3.6.1, Primary Containment Isol Instrument

Tech Spec section 3.4.1, Recirc Loops Operating

A. Incorrect since Reactor Coolant circulation is required in Mode 3 by Shutdown

Cooling or Recirc Pumps.

B. Incorrect since 3.4.7 requires action to place Shutdown Cooling or a Recirc Pump in

operation IMMEDIATELY provided NOTE 1 of the LCO is not used or expired.

C. Incorrect since these are the actions to take if Shutdown Cooling is INOPERABLE.

At this time Shutdown Cooling is OPERABLE since reactor pressure is below the

Shutdown Cooling low pressure permissive.

D. Correct answer.

RO Tier:

SRO Tier:

T2G2

Keyword:

SHUTDOWN COOLING

Cog Level:

C/A 2.8/3.3

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

Monday, June 24, 2002 08:16:46 AM

10

QUESTIONS REPORT

for HT2002

1. 205000G2.1.22 001

Unit 2 is shutting down for a maintenance outage due to the failure of the "B" Recirc

pump which is out-of-service electrically. At 0100 on 4/12/02 reactor pressure went

below 145 psig and reactor temperature went below 300 F. The following conditions

exist at 0400 on 4/12/02:

Reactor pressure

130 psig

Reactor temperature

285 F

Mode Switch position

S/D

At 0415 on 4/12/02 the "A" Recirc pump tripped and cannot be restarted due to bus

overcurrent. What action is required to be taken per Tech Specs?

A. No action is required to be taken since Recirc Pumps are only required to be in

operation in Modes I and 2.

B. Initiate action to place Shutdown Cooling in operation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND monitor

reactor coolant temperature and pressure once per hour.

C. Initiate action to restore Shutdown Cooling to OPERABLE status immediately AND

verify reactor coolant circulation by an alternate method within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from

discovery of no reactor coolant circulation AND be in Mode 4 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Do No action is required for up 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months at which time a Recirc Pump must be running

or Shutdown Cooling must be in operation.

References: Tech Spec section 3.3.6.1, Primary Containment Isol Instrument

Tech Spec section 3.4.1, Recirc Loops Operating

Tech Spec section 3.4.7, RHR Shutdown Cooling

A. Incorrect since Reactor Coolant circulation is required in Mode 3 by Shutdown

Cooling or Recirc Pumps.

B. Incorrect since 3.4.7 requires action to place Shutdown Cooling or a Recirc Pump in

operation IMMEDIATELY provided NOTE 1 of the LCO is not used or expired.

C. Incorrect since these are the actions to take if Shutdown Cooling is INOPERABLE.

At this time Shutdown Cooling is OPERABLE since reactor pressure is below the

Shutdown Cooling low pressure permissive.

D. Correct answer.

Monday, May 06, 2002 07:17:22 AM

Primary Containment Isolation Instrumentation

3.3.6.1

Table 3.3.6.1-1 (page 4 of 4)

Primary Containment Isolation Instrumentation

APPLICABLE

CONDITIONS

MODES OR

REQUIRED

REFERENCED

OTHER

CHANNELS

FROM

SPECIFIED

PER TRIP

REQUIRED

SURVEILLANCE

ALLOWABLE

FUNCTION

CONDITIONS

SYSTEM

ACTION C.1

REQUIREMENTS

VALUE

4.

RCIC System Isolation

(continued)

g.

RCIC Suppression Pool

Area Differential

Temperature - High

h.

Emergency Area Cooker

Temperature - High

5.

RWCU System Isolation

a.

Area

Temperature - High

b.

Area Ventilation

Differential

Temperature - High

c.

SLC System Initiation

d.

Reactor Vessel Water

Level - Low Low,

Level 2

6.

RHR Shutdown Cooling

System Isolation

a.

Reactor Steam Dome

Pressure - High

b.

Reactor Vessel Water

Level - Low,

Level 3

F

1,2,3

1,2

1,2,3

3,4,5

1 per

area

1 per

area

1(c)

2

F

SR

F

H

F

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

SR 3.3.6.1.6

SR

F

SR

2(d)

SR

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

3.3.6.1.1

3,3.6.1.2

3.3.6.1.5

3.3.6.1.6

3.3.6.1.1

3.3.6.1.2

3.3.6.1.5

3.3.6.1.6

Amendment No. 135

S 42°F

S 169*F

5 150°F

S67°F

NA

?:

-47 inches

145 psig

ý 0 inches

(c) SEC System Initiation only inputs into one of the two trip systems.

(d) Only one trip system required in MODES 4 and 5 when RHR Shutdown Cooling System integrity maintained.

1

I

3.3-59

HATCH UNIT 2

Recirculation Loops Operating

3.4.1

3.4

REACTOR COOLANT SYSTEM (RCS)

3.4.1

Recirculation Loops Operating

LCO 3.4.1

Two recirculation loops with matched flows shall be in

operation,

One recirculation loop shall be in operation with the

following limits applied when the associated LCO is

applicable:

a.

LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION

(APLHGR)," single loop operation limits specified

COLR;

b.

LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"

loop operation limits specified in the COLR; and

RATE

in the

single

c.

LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation," Function 2.b (Average Power Range

Monitor Simulated Thermal Power--High), Allowable Value

of Table 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY:

MODES 1 and 2.

Amendment No.

154

HATCH UNIT 2

I

3.4-1

RHR Shutdown Cooling System--Hot Shutdown

3.4.7

3.4

REACTOR COOLANT SYSTEM (RCS)

3.4.7 Residual

LCO 3.4.7

APPLICABILITY:

Heat Removal

(RHR)

Shutdown Cooling System -

Hot Shutdown

Two RHR shutdown cooling subsystems shall be OPERABLE and,

with no recirculation pump in operation, at least one RHR

shutdown cooling subsystem shall be in operation.

--- -NOTES --------------------------

I.

Both RHR shutdown cooling subsystems and recirculation

pumps may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.

2.

One RHR shutdown cooling subsystem may be inoperable

for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of Surveillances.

MODE 3 with reactor steam dome pressure less than the RHR

low pressure permissive pressure.

ACTIONS

6

NOTES -----------------------------------

1.

LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each RHR shutdown cooling

subsystem.

CONDITION

REQUIRED ACTION

COMPLETION TIME

A. One or two RHR

shutdown cooling

subsystems inoperable.

A._

Initiate action to

restore RHR shutdown

cooling subsystem(s)

to OPERABLE status.

ANDý

Immediately

(continued)

E

Amendment No.

135

HATCH UNIT 2

3.4-16

RHR Shutdown Cooling System--Hot Shutdown

3.4.7

ACTIONS

CONDITION

A. (continued)

B. No RHR shutdown

cooling subsystem in

operation.

AND

No recirculation pump

in operation.

DFAnIIRFfl ACTION

I COMPLETION TIME

A.2

Verify an alternate

method of decay heat

removal is available

for each inoperable

RHR shutdown cooling

subsystem.

AND

A.3

B.1

Be in MODE 4.

Initiate action to

restore one RHR

shutdown cooling

subsystem or one

recirculation pump

operation.

AND

B.2

to

Verify reactor

coolant circulation

by an alternate

method.

AND

I hour

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

immediately

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from

discovery of no

reactor coolant

circulation

AND

Once per

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

thereafter

B.3

Monitor reactor

Once per hour

coolant temperature

and pressure.

,^TrU

IIMTT 9

3.4-17

Amendment No.

135

I%6*[V

A l*blJ

.Iv

! ....

nl/ iL~nU*

QUESTIONS REPORT

for Revision4HT2002

3. 206000K5.08 001

Unit I is operating at 100% RTP. The HPCI isolation valves are being stroked and

timed per the Inservice Testing program when MO I E41 -F1 11 HPCI Vacuum Breaker

Isolation Valve failed to close. The Shift Supervisor directed HPCI Vacuum Breaker

Isolation Valve MO 1E41 -F1l04 to be closed and deactivated.

Which ONE of the following describes the time limit for deactivating MO 1 E41 -F1 04 per

Tech Specs and the effect on the HPCI system after the action(s) is/are taken?

A. Actions must be taken within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . HPCI should be declared INOP and a 14 day

LCO entered per TS 3.5.1 .C.

B. Actions must be taken within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . HPCI should be declared INOP and a 14 day

LCO entered per TS 3.5.1 .C.

C. Actions must be taken within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . HPCI system should still be considered

OPERABLE because it can still perform its safety function.

D. Actions must be taken within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . HPCI system should still be considered

OPERABLE because it can still perform its safety function.

References: Tech Spec 3.6.1.3 for PCIVs

SI-LP-00501 Rev. 01, LT-00501 Fig. I

SI-LP-00501 Rev. 01, pg 8 of 46

A. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the line since

there is more than 1 PCIV in the penetration flow path and only 1 valve is

INOPERABLE. Also, HPCI can still perform its function and should still be considered

B. Incorrect since HPCI can still perform its function and should still be considered

C. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the line since

there is more than 1 PCIV in the penetration flow path and only 1 valve is

D. Correct answer.

RO Tier:

SRO Tier:

T2Gl

Keyword:

Cog Level:

C/A 3.0/3.2

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

3

Friday, October 11, 2002 06:51:25 AM

QUESTIONS REPORT

for Revision2 HT2002

31. 206000K5.08 001

Unit 1 is operating at 100% RTP. The HPCI isolation valves are being stroked and

timed per the Inservice Testing program when MO 1 E41-F1 11 HPCI Vacuum Breaker

Isolation Valve failed to close. The Shift Supervisor directed HPCI Vacuum Breaker

Isolation Valve MO 1E41-F104 to be closed and deactivated.

Which ONE of the following describes the time limit for deactivating MO 1 E41-F104 per

Tech Specs and the effect on the HPCI system after the action(s) is/are taken?

A. Actions must be taken within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . HPCI should be declared INOP and a 14 day

LCO entered per TS 3.5.1.C.

B. Actions must be taken within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . HPCI should be declared INOP and a 14 day

LCO entered per TS 3.5.1.C.

C. Actions must be taken within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . HPCI system should still be considered

OPERABLE because it can still perform its design function.

D. Actions must be taken within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . HPCI system should still be considered

OPERABLE because it can still perform its design function.

References: Tech Spec 3.6.1.3 for PCIVs

SI-LP-00501 Rev. 01, LT-00501 Fig. 1

SI-LP-00501 Rev. 01, pg 8 of 46

A. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the line since

there is more than 1 PCIV in the penetration flow path and only 1 valve is

INOPERABLE. Also, HPCI can still perform its function and should still be considered

B. Incorrect since HPCI can still perform its function and should still be considered

C. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the line since

there is more than I PCIV in the penetration flow path and only 1 valve is

D. Correct answer.

RO Tier:

SROTier:

T2G1

Keyword:

Cog Level:

C/A 3.0/3.2

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

r

nAn

o

A

oo

fO

A

34

Friday, September 0,

:

QUESTIONS REPORT

for HT2002

12. 206000K5.08 001

Unit I is operating at 100% RTP. The HPCI isolation valves are being stroked and

timed per the Inservice Testing program when MO F111 HPCI Vacuum Breaker

Isolation Valve failed to close. The Shift Supervisor directed HPCI Vacuum Breaker

Isolation Valve MO F104 to be closed and deactivated.

Which ONE of the following describes the time limit for deactivating MO F1 04 per Tech

Specs and the effect on the HPCI system after the action(s) is/are taken?

A. Actions must be taken within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . HPCI should be declared INOP and a 14 day

LCO entered per TS 3.5.1.C.

B. Actions must be taken within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . HPCI should be declared INOP and a 14 day

LCO entered per TS 3.5.1.C.

C. Actions must be taken within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . HPCI system should still be considered

OPERABLE because it can still perform its design function.

Df Actions must be taken within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . HPCI system should still be considered

OPERABLE because it can still perform its design function.

References: Tech Spec 3.6.1.3 for PCIVs

SI-LP-00501 Rev. 01, LT-00501 Fig. 1

SI-LP-00501 Rev. 01, pg 8 of 46

A. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the line since

there is more than 1 PCIV in the penetration flow path and only 1 valve is

INOPERABLE. Also, HPCI can still perform its function and should still be considered

B. Incorrect since HPCI can still perform its function and should still be considered

C. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the line since

there is more than 1 PCIV in the penetration flow path and only 1 valve is

D. Correct answer.

RO Tier:

SRO Tier:

T2G0

Keyword:

Cog Level: C/A 3.0/3.2

-

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

Monday, June 24, 2002 08:16:47 AM

13

QUESTIONS REPORT

for HT2002

3. 206000K5.08 001

Unit 1 is operating at 100% RTP. The HPCI isolation valves are being stroked and

timed per the Inservice Testing program when MO F1 11 HPCI Vacuum Breaker

Isolation Valve failed to close. The Shift Supervisor directed HPCI Vacuum Breaker

Isolation Valve MO F104 to be closed and deactivated.

Which ONE of the following describes the time limit for deactivating MO F1 04 per Tech

Specs and the effect on the HPCI system after the action(s) is/are taken?

A. HPCI Vacuum Breaker Isolation Valve MO F104 must be closed and deactivated

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . HPCI should be declared INOP and a 14 day LCO entered per TS

3.5.1 .C.

B. HPCI Vacuum Breaker Isolation Valve MO F104 must be closed and deactivated

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . HPCI should be declared INOP and a 14 day LCO entered per TS

3.5.1.C.

C. HPCI Vacuum Breaker Isolation Valve MO F104 must be closed and deactivated

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . HPCI system should still be considered OPERABLE because it can

still perform its design function.

DW HPCI Vacuum Breaker Isolation Valve MO F104 must be closed and deactivated

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . HPCI system should still be considered OPERABLE because it can

still perform its design function.

References: Tech Spec 3.6.1.3 for PCIVs

SI-LP-00501 Rev. 01, LT-00501 Fig. 1

SI-LP-00501 Rev. 01, pg 8 of 46

A. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the line since

there is more than 1 PCIV in the penetration flow path and only 1 valve is

INOPERABLE. Also, HPCI can still perform its function and should still be considered

B. Incorrect since HPCI can still perform its function and should still be considered

C. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the line since

there is more than I PCIV in the penetration flow path and only 1 valve is

D. Correct answer.

4

Friday, May 31, 2002 03:24:45 PM

QUESTIONS REPORT

for HT2002

1. 206000K5.08 001

Unit I is operating at 100% RTP. The HPCI isolation valves are being stroked and

timed per the Inservice Testing program when the HPCI Vacuum Breaker Isolation

Valve MO F1 11 failed to close. SELECT the answer that meets the requirements of

Tech Specs and indicates the effect on the HPCI system after the action(s) is/are

taken?

A. HPCI Vacuum Breaker Isolation Valve MO F104 must be closed and deactivated

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . HPCI should be declared INOP and a 14 day LCO entered per TS

3.5.1.C.

B. HPCI Vacuum Breaker Isolation Valve MO F104 must be closed and deactivated

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . HPCI should be declared INOP and a 14 day LCO entered per TS

3.5.1 C.

C. HPCI Vacuum Breaker Isolation Valve MO F104 must be closed and deactivated

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . HPCI system should still be considered OPERABLE because it can

still perform its design function.

D.* HPCI Vacuum Breaker Isolation Valve MO F104 must be closed and deactivated

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . HPCI system should still be considered OPERABLE because it can

still perform its design function.

References: Tech Spec 3.6.1.3 for PCIVs

SI-LP-00501 Rev. 01, LT-00501 Fig. 1

SI-LP-00501 Rev. 01, pg 8 of 46

A. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the line since

there is more than 1 PCIV in the penetration flow path and only 1 valve is

INOPERABLE. Also, HPCI can still perform its function and should still be considered

B. Incorrect since HPCI can still perform its function and should still be considered

C. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the line since

there is more than 1 PCIV in the penetration flow path and only 1 valve is

D. Correct answer.

Wednesday, April 10, 2002 10:59:13 AM

Note:

Minimum recommended speed for Turbine operation is 2000

rpm based on maintaining adequate oil pressure for governor

operation and bearing lubrication. Above this speed there is also

sufficient steam flow through the Turbine to prevent turbine

exhaust valve chatter.

8.

The Exhaust Line Drain Pot removes condensation from the HPCI Turbine

Exhaust line drain when the HPCI system is in standby. Level in the Drain

Pot is controlled automatically by drain valve F053. F053 is interlocked

closed IF BOTH F001 AND TSV ARE NOT FULLY CLOSED. EITHER

F001 OR the TURBINE STOP VALVE must be closed for F053 to open

(both units). The drain pot discharges to the Barometric Condenser.

9.

The HPCI System Rupture Disks (D003 and D004) located in the Torus

Area protect the HPCI Turbine casing from excessive exhaust pressure. The

two diaphragms are in series and are designed to rupture at 150 psig. The

space between them is vented to the Torus area through an orifice. High

pressure between the diaphragms will cause a HPCI System Isolation at 10

psig.

10.

HPCI Exhaust Line Vacuum Breakers F102 and F103 prevent drawing a

vacuum on the exhaust line by steam condensation following turbine

shutdown. This vacuum would result in siphoning of Suppression Pool

water into the HPCI Exhaust line and could cause exhaust line damage on a

subsequent start.

11.

Vacuum Breaker Isolation Valves F104 and Fll1 are normally open and

will isolate the vacuum breaker piping if conditions indicate a possible

HPCI system leak. F104 and Fl 11 are AC operated MOVs powered from

R24-S01 1 and S012 respectively.

These valves automatically close on a combined signal of High

Drywell Pressure (set at 1.85 psig) and Low HPCI Steam Line

Pressure (set at 128 psig).

B.

Gland Seal Condenser System

The Gland Seal Condenser System prevents steam outliakage from the turbine

shaft seals, turbine stop valve, turbine control valve, and turbine exhaust drain

from entering the HPCI room. This outleakage could cause potential safety (High

temperatures) or airborne radiological hazards. The system automatically starts

on an auto-initiation of HPCL and consists of:

3.6

CONTAINMENT SYSTEMS

3.6.1.3

Primary Containment Isolation Valves (PCIVs)

LCO

3.6.1.3

APPLICABILITY:

Each PCIV, except reactor building-to-suppression chamber

vacuum breakers, shall be OPERABLE.

MODES 1, 2, and 3,

When associated instrumentation is required to be OPERABLE

per LCO 3.3.6.1, "Primary Containment Isolation

Instrumentation."

ACTIONS

---------------- -

NO T ES ----

------ -----

-

1.

Penetration flow paths except for 18 inch purge valve penetration flow

paths may be unisolated intermittently under administrative controls.

2.

Separate Condition entry is allowed for each penetration flow path.

3.

Enter applicable Conditions and Required Actions for systems made

inoperable by PCIVs.

4.

Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary

Containment," when PCIV leakage results in exceeding overall containment

leakage rate acceptance criteria.

-.....-

___-----------------------------------------------

CONDITION

REQUIRED ACTION

COMPLETION TIME

A. - --------

NOTE ---------

A.1

Isolate the affected

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except

Only applicable to

penetration flow path

for main steam

penetration flow paths

by use of at least

line

with two PCIVs.

one closed and de

activated automatic

AND

valve, closed manual

One or more

valve, blind flange,

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for main

penetration flow paths

or check valve with

steam line

with one PCIV

flow through the

inoperable except due

valve secured.

to leakage not within

limit.

AND

(continued)

Amendment No.

135

HATCH UNIT 2

PCIVs

3.6.1.3

4

3.6-8

PCIVS

3.6.1.3

ACTIONS

CONDITION

A. (continued)

REQUIRED ACTION

A.2

- --------

NOTE------

Isolation devices in

high radiation areas

may be verified by

use of administrative

means.

Verify the affected

penetration flow path

is isolated.

3.6-9

COMPLETION TIME

Once per 31 days

for isolation

devices outside

primary

containment

AND

Prior to

entering MODE 2

or 3 from MODE 4

if primary

containment was

de-inerted while

in MODE 4, if

not performed

within the

isolation

devices inside

primary

containment

(continued)

Amendment No. 135

HATCH UNIT 2

PCIVs

3.6.1.3

6

1/2>_

C. - -------- NOTE------

Only applicable to

penetration flow paths

with only one PCIV.

One or more

penetration flow paths

with one PCIV

inoperable except due

to leakage not within

limits.

_____________________________________________________________i

C.1

AND

C.2

Isolate the affected

penetration flow path

by use of at least

one closed and de

activated automatic

valve, closed manual

valve, or blind

flange.

- --------

NOTE------

Valves and blind

flanges in high

radiation areas may

be verified by use of

administrative means.

Verify the affected

penetration flow path

is isolated.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except

for excess flow

check valve

(EFCV) line

AND

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for

EFCV line

a

Once per 31 days

(continued)

C

/

Amendment No.

135

HATCH UNIT 2

3.6-10

CONDITION

REQUIRED ACTION

COMPLETION TIME

B. ----------

NOTE ---------

B.1

Isolate the affected

I hour

Only applicable to

penetration flow path

penetration flow paths

by use of at least

with two PCIVs.

one closed and de

activated automatic

valve, closed manual

One or more

valve, or blind

.penetration flow paths

flange.

with two PCIVs

inoperable except due

to leakage not within

limit.

ArTIAtIC

f,-nn+in;iprl

Vacuum

Pump

(S022)

HPCI SYSTEM

(

(SIMPLIFIED DIAGRAM)

LT-00501 Fig 1

Page 33 of 4

QUESTIONS REPORT

for HT2002

14. 2090010G2.2.21 001

The "A" Core Spray system on Unit 1 was taken out-of-service to inspect the pump

internals due to high vibration. Foreign material was found inside the pump and no

additional repairs were necessary. The unit is in Day 5 of a 7 day LCO and the system

has been returned to service, filled and vented.

Which ONE of the following indicates the surveillances that are required to be

performed prior to declaring "A" Core Spray operable?

A. SR 3.5.1.1 piping filled from pump disch to injection valve, SR 3.5.1.7 flow rate test,

SR 3.5.1.10 subsystem actuates on initiation signal.

B. SR 3.5.1.1 piping filled from pump disch to injection valve, SR 3.5.1.2 valve position

verification, SR 3.5.1.7 flow rate test, SR 3.5.1.10 subsystem actuates on initiation

signal.

C. SR 3.5.1.2 valve position verification, SR 3.5.1.7 flow rate test, SR 3.5.1.10

subsystem actuates on initiation signal, SR 3.5.1.13 ECCS Response time.

D' SR 3.5.1.1 piping filled from pump disch to injection valve, SR 3.5.1.2 valve position

verification, SR 3.5.1.7 flow rate test.

Reference: Tech Spec Bases SR 3.0.1

A. Incorrect since SR 3.5.1.10 does not need to be performed since no work was done

on the Core Spray Logic and SR 3.5.1.2 does need to be performed since valves in the

system were out of the normal Operable lineup.

B. Incorrect since SR 3.5.1.10 does not need to be performed since no work was done

on the Core Spray Logic.

C. Incorrect since SR 3.5.1.10 and SR 3.5.1.13 do not need to be performed since no

work was done on the Core Spray Logic. Also, SR 3.5.1.1 does need to be performed

since system was drained.

D. Correct answer.

ROTier:

SROTier:

T2G1

Keyword:

TECH SPEC

Cog Level:

C/A 2.3/3.5

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

Monday, June 24, 2002 08:16:47 AM

16

QUESTIONS REPORT

for HT2002

1. 209001G2.2.21 001

The "A" Core Spray system on Unit I was taken out-of-service to inspect the pump

internals due to high vibration. Foreign material was found inside the pump and no

additional repairs were necessary. The unit is in Day 5 of a 7 day LCO and the system

has been returned to service, filled and vented. Which surveillances are required to be

performed prior to declaring A Core Spray operable?

A. SR 3.5.1.1 piping filled from pump disch to injection valve, SR 3.5.1.7 flow rate test,

SR 3.5.1.10 subsystem actuates on initiation signal.

B. SR 3.5.1.1 piping filled from pump disch to injection valve, SR 3.5.1.2 valve position

verification, SR 3.5.1.7 flow rate test, SR 3.5.1.10 subsystem actuates on initiation

signal.

C. SR 3.5.1.2 valve position verification, SR 3.5.1.7 flow rate test, SR 3.5.1.10

subsystem actuates on initiation signal, SR 3.5.1.13 ECCS Response time.

D' SR 3.5.1.1 piping filled from pump disch to injection valve, SR 3.5.1.2 valve position

verification, SR 3.5.1.7 flow rate test.

Reference: Tech Spec Bases SR 3.0.1

A. Incorrect since SR 3.5.1.10 does not need to be performed since no work was done

on the Core Spray Logic and SR 3.5.1.2 does need to be performed since valves in the

system were out of the normal Operable lineup.

B. Incorrect since SR 3.5.1.10 does not need to be performed since no work was done

on the Core Spray Logic.

C. Incorrect since SR 3.5.1.10 and SR 3.5.1.13 do not need to be performed since no

work was done on the Core Spray Logic. Also, SR 3.5.1.1 does need to be performed

since system was drained.

D. Correct answer.

Monday, May 06, 2002 07:42:08 AM

1

SR Applicability

B 3.0

B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY

BASES

SRs

SR 3.0.1 through SR 3.0.4 establish the general requirements

applicable to all Specifications and apply at all times, unless otherwise

stated.

SR 3.0.1

SR 3.0.1 establishes the requirement that SRs must be met during the

MODES or other specified conditions in the Applicability for which the

requirements of the LCO apply, unless otherwise specified in the

individual SRs. This Specification is to ensure that Surveillances are

performed to verify the OPERABILITY of systems and components,

and that variables are within specified limits. Failure to meet a

Surveillance within the specified Frequency, in accordance with

SR 3.0.2, constitutes a failure to meet an LCO.

Systems and components are assumed to be OPERABLE when the

associated SRs have been met. Nothing in this Specification,

however, is to be construed as implying that systems or components

are OPERABLE when:

a.

The systems or components are known to be inoperable,

although still meeting the SRs; or

b.

The requirements of the Surveillance(s) are known to be not

met between required Surveillance performances.

Surveillances do not have to be performed when the unit is in a

MODE or other specified condition for which the requirements of the

associated LCO are not applicable, unless otherwise specified. The

SRs associated with a Special Operations LCO are only applicable

when the Special Operations LCO is used as an allowable exception

to the requirements of a Specification.

Surveillances, including Surveillances invoked by Required Actions,

do not have to be performed on inoperable equipment because the

ACTIONS define the remedial measures that apply. Surveillances

have to be met and performed in accordance with SR 3.0.2, prior to

returning equipment to OPERABLE status.

Upon completion of maintenance, appropriate post maintenance

testing is required to declare equipment OPERABLE. This includes

ensuring applicable Surveillances are not failed and their most recent

performance is in accordance with SR 3.0.2. Post maintenance

(continued)

HATCH UNIT 1

B 3.0-9

REVISION 0

SR Applicability

B3.0

BASES

SR 3.0.1

testing may not be possible in the current MODE or other specified

(continued)

conditions in the Applicability due to the necessary unit parameters

not having been established. In these situations, the equipment may

be considered OPERABLE provided testing has been satisfactorily

completed to the extent possible and the equipment is not otherwise

believed to be incapable of performing its function. This will allow

operation to proceed to a MODE or other specified condition where

other necessary post maintenance tests can be completed.

Some examples of this process are:

a.

Control Rod Drive maintenance during refueling that requires

scram testing at > 800 psi. However, if other appropriate

testing is satisfactorily completed and the scram time testing of

SR 3.1.4.3 is satisfied, the control rod can be considered

OPERABLE. This allows startup to proceed to reach 800 psi

to perform other necessary testing.

b.

High pressure coolant injection (HPCI) maintenance during

shutdown that requires system functional tests at a specified

pressure. Provided other appropriate testing is satisfactorily

completed, startup can proceed with HPCI considered

OPERABLE. This allows operation to reach the specified

pressure to complete the necessary post maintenance testing.

SR 3.0.2

SR 3.0.2 establishes the requirements for meeting the specified

Frequency for Surveillances and any Required Action with a

Completion Time that requires the periodic performance of the

Required Action on a "once per..." interval.

SR 3.0.2 permits a 25% extension of the interval specified in the

Frequency. This extension facilitates Surveillance scheduling and

considers plant operating conditions that may not be suitable for

conducting the Surveillance (e.g., transient conditions or other

ongoing Surveillance or maintenance activities).

The 25% extension does not significantly degrade the reliability that

results from performing the Surveillance at its specified Frequency.

This is based on the recognition that the most probable result of any

particular Surveillance being performed is the verification of

conformance with the SRs. The exceptions to SR 3.0.2 are those

Surveillances for which the 25% extension of the interval specified in

the Frequency does not apply. These exceptions are stated in the

individual Specifications. The requirements of regulations take

(continued)

HATCH UNIT 1

B 3.0-10

REVISION 0

QUESTIONS REPORT

for HT2002

25. 215004K5.03 001

A startup is in progress on Unit I with all the IRM's on Range 4. The Control Board

Operator is in the process of withdrawing SRM's to keep the rod block cleared when it

is determined that SRM "A" will not retract. All attempts to free the SRM have failed

and Upper Management decides to continue with the startup and to leave the SRM

inserted.

Which ONE of the following states IF and WHEN the SRM should be declared

INOPERABLE?

A. Declare "A" SRM INOPERABLE immediately, since the SRM cannot be moved.

B/ Declare "A" SRM INOPERABLE when it is bypassed to continue with the startup.

C. You don't have to consider the SRM Inoperable since the SRM's are not required

with IRM's on range 3 or above.

D. Declare "A" SRM INOPERABLE when the "A" SRM reading deviates by >200 cps

from the other 3 SRM's.

References: Tech Spec 3.3.1.2, Source Range Monitor (SRM) Instrumentation

Tech Spec 3.3.1.2 Bases

Technical Requirements Manual Table 3.3.2-1

34SV-SUV-019-2S, Surveillance Checks Rev. 32.3 pg 21 of 59

(NOTE) If this question is unacceptable then HATCH99,1NK #96 may be

used in its place.

A. Incorrect since the SRM is currently performing its function and there isn't a

requirement that the detector move.

B. Correct answer. The SRM has to be bypassed prior to continuing with the startup

since there is a rod block inserted when the SRM reaches the high limit of 7 X 10e4.

C. Incorrect since the

on range 8 or above.

D. Incorrect since the

RO Tier:

Keyword:

SRM

Source:

N

Test:

S

Rod Block function of the SRM's are required until the IRM's are

deviation is figured as the Max

SRO Tier:

Cog Level:

Exam:

Misc:

divided by Min < 20.

T2G I

C/A 2.8/2.8

HT02301

TCK

27

Monday, June 24, 2002 08:16:48 AM

QUESTIONS REPORT

for HT2002

1. 215004K5.03 001

A startup is in progress on Unit 1 with all the IRM's on Range 4. The Control Board

Operator is in the process of withdrawing SRM's to keep the rod block cleared when it

is determined that SRM "A" will not retract. All attempts to free the SRM have failed

and Upper Management decides to continue with the startup and to leave the SRM

inserted. As Shift Supervisor, determine IF and WHEN the SRM should be declared

INOPERABLE?

A. Declare "A" SRM INOPERABLE immediately, since the SRM cannot be moved.

B! Declare "A" SRM INOPERABLE when it is bypassed to continue with the startup.

C. You don't have to consider the SRM Inoperable since the SRM's are not required

with IRM's on range 2 or above.

D. Declare "A" SRM INOPERABLE when the "A" SRM reading deviates by >200 cps

form the other 3 SRM's.

References: Tech Spec 3.3.1.2, Source Range Monitor (SRM) Instrumentation

Tech Spec 3.3.1.2 Bases

34SV-SUV-019-2S, Surveillance Checks Rev. 32.3 pg 21 of 59

7

(NOTE) If this question is unacceptable then HATCH99.BNK #96 may be

used in its place.

A. Incorrect since the SRM is currently performing its function and there isn't a

requirement that the detector move.

B. Correct answer. The SRM has to be bypassed prior to continuing with the startup

since there is a rod block inserted when the SRM reaches the high limit of 7 X 1 0e4.

C. Incorrect since the SRM's are required until the IRM's are on range 8 or above.

D. Incorrect since the deviation is figured as the Max divided by Min < 20.

Friday, May 03, 2002 07:48:40 AM

Control Rod Block Instrumentation

T 3.3.2

Table T3.3.2-1 (Page 1 of 2)

Control Rod Block Instrumentation

APPLICABLE

MODES OR

REQUIRED

OTHER

CHANNELS

SPECIFIED

PER

SURVEILLANCE

ALLOWABLE

FUNCTION

CONDITIONS

FUNCTION

REQUIREMENTS

VALUE

1.

SRM

3

a. Detector Not Full In

5 (e)

2(c)

b.

Upscale

2 (b)

3

5

c. Inoperative

d.

Downscale

2(c)

5

3

2 (b)

3

2 (b)

5

2.

IRM

a. Detector Not Full In

b. Upscale

c.

Inoperative

d. Downscale

2,5(e)

2,5

2(d)

6

TSR 3.3.2.1

TSR 3.3.2.3

TSR 3.3.2.1

TSR 3.3.2.3

TSR 3.3.2.1

TSR 3.3.2.3

TSR 3.3.2.1

TSR 3.3.2.3

TSR 3.3.2.1

TSR 3.3.2.3

TSR 3.3.2.1

TSR 3.3.2.3

NA

_O

105 cps

< 105 cps

NA

Ž 3 cps

> 3 cps

_<

N/A

108/125 of full

scale

NA

> 5/125 of full

scale

(continued)

(a) With IRMs on Range 2 or below.

(b) Only one SRM is required to be OPERABLE during spiral offload or reload when the fueled region

includes only that SRM detector.

(c) With IRMs on Range 7 or below.

(d)

With IRMs on Range 2 or above.

(e) This function is not required if the detector is verified to be in the fully inserted position and the drive

motor is deactivated.

Revision 24

T 3.3-5

HATCH UNIT 2 TRM

2(a)

SOUTHERN NUCLEAR

PAGE 21 OF 59

PLANT E. I. HATCH

DOCUMENT TITLE:

DOCUMENT NUMBER:

REVNER NO:

SURVEILLANCE CHECKS

34SV-SUV-O19-28

32.3

PANEL - INSTRUMENT / TECH SPEC.

2H11-P689 - 2011-K621A, W.R. Drywell Radiation

2H1 1-P690 - 2D1 1-K621B. W.R. Drywell Radiation

ConfimI max minus mi < 10 for Items in 7.5.1

(SR 3.3.3.1.1 for 3.3.3.1-1(5.)),

(SR 3.3.6.1.1 for 3.3.6.1-1(2.c.))

N

OPER COND

1,2,3

B

1 1,2,3

a

7.5.3

2D21-P600 - Area Rad Monitors

I Ba

6

I

a

2D21-P600 - 2D211-K601A, Area Rad. Monitor

- 2D21-K601M, Area RMd. Monitor

(TSR 3.3.7.1 for T3.3.7-1 14.))

Confirm Max Divided by

Min <- 5 for Items in 7.5.4.

(TSR 3.3.7.1 for T3.3.7-1

.241)

2H1 1-P606 - 2C51-K600A, SRM A CPS

- 2C51-K600B, SRM B CPS

- 2C51 -K600C, SIRM C CPS

- 2C51-K600D, SRM D CPS

(SR 3.3.1.2.4 for 3.3.1.2-1(1.))

Confirm max divided by min _< 20 for items in 7.5.6

(SR 3.3.1.2.1 AND 3.3.1.2.3 for 3.3.1.2-1 (1 .))

2H1 1-13606 - 2C51-K501 A thru H, IRM Channel

Check

(SR 3.3.1.1.1 for 3.3.1.1-1(1.a.))

2H11-P608 - 2C51-K615A thru D, APRM Channel

Check

(SR 3.3.1.1.1 for 3.3.1.1-1(2.a.)(2.b.)(2.c.)

(2 e Wh'2

f) SR 3.10.8.1)

B

B,

BB

B,

7.5.10

2H1 1-P608 - 2C51-K617A thru D, APRM 2 of 4 Voter

C,

,Logic

Module

cc

1,2,3,(*)

2(5-),3,4,

5

2,5($)

1,2,5(+)

Calculations verified

/

Ninht I Day

FREQ TS - OPER LIM

a

< 138 RIHR

On scale

On scale AND

< 20 mr/hr.

a

> 3 cps

AND detector

full-in

a

On scale and

a

difference

between

Highest Range

and Lowest

Range eS 3

SELF-TEST OK

a

NO ERROR

MESSAGES.

All APRMs read

within 3%

power of

each other

Lamp test,

a

All LED's lit

Initials

(*) During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, during OPDRVs.

($) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(+) During Shutdown Margin Testing.

(**) With IRM's on Range 2 OR below.

G16.30

MGR-0001 Rev 3

7.5.1

7.5.2

7.5.4

7.5.5

7.5.6

7.5.7

7.5.8

7.5.9

I

NIGHTFDAY

1,2,3,(*)

a

Time

Date

SRM Instrumentation

3.3.1.2

Table 3.3.1.2-1 (page 1 of 1)

Source Range Monitor Instrumentation

APPLICABLE

MODES OR OTHER

SPECIFIED

REQUIRED

SURVEILLANCE

FUNCTION

CONDITIONS

CHANNELS

REQUIREMENTS

1.

Source Range Monitor

2(a)

3,4

5

2

SR

2(b)(c)

3.3.1.2.1

3.3.1.2.4

3.3.1.2.6

3.3.1.2.7

3.3.1.2.3

3.3.1.2.4

3.3.1.2.6

3.3.1.2.7

3.3.1.2.1

3.3.1.2.2

3.3.1.2.4

3.3.1.2.5

3.3.1.2.7

(a)

With IRMs on Range 2 or below.

(b)

Only one SRM channel is required to be OPERABLE during spiral off load or reload when the fueled region includes

only that SRM detector.

(c)

Special movable detectors may be used in place of SRMs if connected to normal SRM circuits.

4

Amendment No. 195

HATCH UNIT 1

3.3-14

a

3

SRM Instrumentation

B 3.3.1.2

BASES

LCO

(continued)

APPLICABILITY

ACTIONS

indication can be generated. These special detectors provide more

flexibility in monitoring reactivity changes during fuel loading, since

they can be positioned anywhere within the core during refueling.

They must still meet the location requirements of SR 3.3.1.2.2 and all

other required SRs for SRMs.

For an SRM channel to be considered OPERABLE, it must be

providing neutron flux monitoring indication.

The SRMs are required to be OPERABLE in MODES 2, 3, 4, and 5

prior to the IRMs being on scale on Range 3 to provide for neutron

monitoring. In MODE 1, the APRMs provide adequate monitoring of

reactivity changes in the core; therefore, the SRMs are not required.

In MODE 2, with IRMs on Range 3 or above, the IRMs provide

adequate monitoring and the SRMs are not required.

A.1 and B.1

In MODE 2, with the IRMs on Range 2 or below, SRMs provide the

means of monitoring core reactivity and criticality. With any number of

the required SRMs inoperable, the ability to monitor neutron flux is

degraded. Therefore, a limited time is allowed to restore the

inoperable channels to OPERABLE status.

Provided at least one SRM remains OPERABLE, Required Action A.1

allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore the required SRMs to OPERABLE status.

This time is reasonable because there is adequate capability

remaining to monitor the core, there is limited risk of an event during

this time, and there is sufficient time to take corrective actions to

restore the required SRMs to OPERABLE status or to establish

alternate IRM monitoring capability. During this time, control rod

withdrawal and power increase is not precluded by this Required

Action. Having the ability to monitor the core with at least one SRM,

proceeding to IRM Range 3 or greater (with overlap required by

SR 3.3.1.1.6), and thereby exiting the Applicability of this LCO, is

acceptable for ensuring adequate core monitoring and allowing

continued operation.

With three required SRMs inoperable, Required Action B.1 allows no

positive changes in reactivity (control rod withdrawal must be

immediately suspended) due to inability to monitor the changes.

Required Action A.1 still applies and allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore

(continued)

HATCH UNIT 1

REVISION 0

B 3.3-35

SRM Instrumentation

B 3.3.1.2

BASES

ACTIONS

A.1 and B.1 (continued)

monitoring capability prior to requiring control rod insertion. This

allowance is based on the limited risk of an event during this time,

provided that no control rod withdrawals are allowed, and the desire to

concentrate efforts on repair, rather than to immediately shut down,

with no SRMs OPERABLE.

C.1

In MODE 2, if the required number of SRMs is not restored to

OPERABLE status within the allowed Completion Time, the reactor

shall be placed in MODE 3. With all control rods fully inserted, the

core is in its least reactive state with the most margin to criticality.

The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on

-operating experience, to reach MODE 3 from full power conditions in

an orderly manner and without challenging plant systems.

D.1 and D.2

With one or more required SRMs inoperable in MODE 3 or 4, the

neutron flux monitoring capability is degraded or nonexistent. The

requirement to fully insert all insertable control rods ensures that the

reactor will be at its minimum reactivity level while no neutron

monitoring capability is available. Placing the reactor mode switch in

the shutdown position prevents subsequent control rod withdrawal by

maintaining a control rod block. The allowed Completion Time of

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is sufficient to accomplish the Required Action, and takes into

account the low probability of an event requiring the SRM occurring

during this interval.

E.1 and E.2

With one or more required SRMs inoperable in MODE 5, the ability to

detect local reactivity changes in the core during refueling is

degraded. CORE ALTERATIONS must be immediately suspended

and action must be immediately initiated to fully insert all insertable

control rods in core cells containing one or more fuel assemblies.

Suspending CORE ALTERATIONS prevents the two most probable

causes of reactivity changes, fuel loading and control rod withdrawal,

from occurring. Inserting all insertable control rods ensures that the

reactor will be at its minimum reactivity given that fuel is present in the

(continued)

HATCH UNIT 1

B 3.3-36

REVISION 0

SRM Instrumentation

B 3.3.1.2

BASES

/

ACTIONS

E.1 and E.2 (continued)

core. Suspension of CORE ALTERATIONS shall not preclude

completion of the movement of a component to a safe, conservative

position.

Action (once required to be initiated) to insert control rods must

continue until all insertable rods in core cells containing one or more

fuel assemblies are inserted.

SURVEILLANCE

As Noted at the beginning of the SRs, the SRs for each SRM

REQUIREMENTS

Applicable MODE or other specified conditions are found in the SRs

column of Table 3.3.1.2-1.

The Surveillances are modified by a second Note to indicate that

when a channel is placed in an inoperable status solely for

performance of required Surveillances, entry into associated

Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

provided the other required channel (or channels when 3 channels are

required) is OPERABLE. Upon completion of the Surveillance, or

expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to

-'

OPERABLE status or the applicable Condition entered and Required

Actions taken. The Note is based upon a NRC Safety Evaluation

Report (Ref. 1) which concluded that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance

does not significantly reduce the probability of detecting power

changes, when necessary.

SR 3.3.1.2.1 and SR 3.3.1.2.3

Performance of the CHANNEL CHECK ensures that a gross failure of

instrumentation has not occurred. A CHANNEL CHECK is normally a

comparison of the parameter indicated on one channel to a similar

parameter on another channel. It is based on the assumption that

instrument channels monitoring the same parameter should read

approximately the same value. Significant deviations between the

instrument channels could be an indication of excessive instrument

drift in one of the channels or something even more serious. A

CHANNEL CHECK will detect gross channel failure; thus, it is key to

verifying the instrumentation continues to operate properly between

each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff based on a

combination of the channel instrument uncertainties, including

-j

(continued)

a ,_REVISION

15

HATCH UNIT 1

U 2.

"0

SRM Instrumentation

B 3.3.1.2

BASES

SURVEILLANCE

SR 3.3.1.2.1 and SR 3.3.1.2.3 (continued)

REQUIREMENTS

indication and readability. If a channel is outside the criteria, it may be

an indication that the instrument has drifted outside its limit.

The Frequency of once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for SR 3.3.1.2.1 is based on

operating experience that demonstrates channel failure is rare. While

in MODES 3 and 4, reactivity changes are not expected; therefore,

the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is relaxed to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for SR 3.3.1.2.3. The

CHANNEL CHECK supplements less formal, but more frequent,

checks of channels during normal operational use of the displays

associated with the channels required by the LCO.

SR 3.3.1.2.2

To provide adequate coverage of potential reactivity changes in the

core when the fueled region encompasses more than one SRM, one

SRM is required to be OPERABLE in the quadrant where CORE

ALTERATIONS are being performed, and the other OPERABLE SRM

must be in an adjacent quadrant containing fuel. Note 1 states that

the SR is required to be met only during CORE ALTERATIONS. It is

not required to be met at other times in MODE 5 since core reactivity

changes are not occurring. This Surveillance consists of a review of

plant logs to ensure that SRMs required to be OPERABLE for given

CORE ALTERATIONS are, in fact, OPERABLE. In the event that

only one SRM is required to be OPERABLE (when the fueled region

encompasses only one SRM), per Table 3.3.1.2-1, footnote (b), only

the a. portion of this SR is required. Note 2 clarifies that more than

one of the three requirements can be met by the same OPERABLE

SRM. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is based upon operating experience

and supplements operational controls over refueling activities that

include steps to ensure that the SRMs required by the LCO are in the

proper quadrant.

SR 3.3.1.2.4

This Surveillance consists of a verification of the SRM instrument

readout to ensure that the SRM reading is greater than a specified

minimum count rate, which ensures that the detectors are indicating

count rates indicative of neutron flux levels within the core. This

surveillance also requires the signal to noise ratio to be verified to be

2 2:1. A signal to noise ratio that meets this requirement ensures the

detectors are inserted to an acceptable operating level. Therefore, to

meet this portion of the surveillance, it is necessary only to verify the

(continued)

HATCH UNIT 1

B 3.3-38

REVISION 15

QUESTIONS REPORT

for HT2002

29. 218000G2.2.22 001

"On 3/2/02 at 0800 Unit 2 is in Mode 1 when RCIC is declared inoperable and day 1 of

a 14 day LCO is entered. On 3/7/02 at 1600 the Instrument Techs start a surveillance

on a Drywell Pressure instrument associated with ADS Trip system A by valving the

instrument out. At 2200 they report to the Shift Supervisor that the instrument cannot

be calibrated and that no other instruments are affected.

Per Tech Specs, which ONE of the following is the latest time the channel shall be

placed in the tripped condition?

(Provide Tech Spec section 3.3.5.1 and 3.5.3)

A.! 2200 on 3/11/02.

B. 1600 on 3/11/02.

C. 2200 on 3/15/02.

D. 1600 on 3/15/02.

Reference: Tech Spec section3.3.5.1 and 3.5.3.

A. Correct answer.

B. Incorrect answer. Can delay the actions for Condition F for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months due to note 2 for

surveillance requirements even though the instrument is inoperable.

C. Incorrect answer. Completion time in answer is 8 days from required action time.

This is wrong since RCIC is inoperable concurrent with this instrument.

D. Completion time in answer is 8 days from instrument being inoperable. This is wrong

since you can use the surveillance note of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and RCIC is also inoperable.

RO Tier:

SRO Tier:

T2G1

Keyword:

TECH SPEC

Cog Level:

C/A 3.4/4.1

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

31

Monday, June 24, 2002 08:16:49 AM

QUESTIONS REPORT

for HT2002

6. 218000G2.2.22 001

On 3/2/02 at 0800 Unit 2 is in Mode I when RCIC is declared inoperable and day 1 of

a 14 day LCO is entered. On 3/7/02 at 1600 the Instrument Techs start a surveillance

on a Drywell Pressure instrument associated with ADS Trip system A by valving the

instrument out. At 2200 they report to the Shift Supervisor that the instrument cannot

be calibrated and that no other instruments are affected. Per Tech Specs, what is the

latest time the channel shall be placed in the tripped condition?

A.r 2200 on 3/11/02.

B. 1600 on 3/11/0 2 .

C. 2200 on 3/15/02.

D. 1600 on 3/15/02.

Reference: Tech Spec section3.3.5.1 and 3.5.3.

A. Correct answer.

B. Incorrect answer. Can delay the actions for Condition F for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months due to note 2 for

surveillance requirements even though the instrument is inoperable.

C. Incorrect answer. Completion time in answer is 8 days from required action time.

This is wrong since RCIC is inoperable concurrent with this instrument.

D. Completion time in answer is 8 days from instrument being inoperable. This is wrong

since you can use the surveillance note of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and RCIC is also inoperable.

Thursday, April 04, 2002 11:36:26 AM

6

ECCS Instrumentation

3.3.5.1

3.3 INSTRUMENTATION

3.3.5.1

Emergency Core Cooling System (ECCS) Instrumentation

LCO 3.3.5.1

APPLICABILITY:

The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be

According to Table 3.3.5.1-1.

---.

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

Separate Condition entry is allowed for each channel.

CONDITION

REQUIRED ACTION

COMPLETION TIME

A.

One or more channels

B.

As required by Required

Action A.1 and referenced

in Table 3.3.5.1-1.

A.1

Enter the Condition

referenced in

Table 3.3.5.1-1 for the

channel.

B.1 -

NOTES ----------

1.

Only applicable in

MODES 1,2,

and 3.

2.

Only applicable for

Functions 1 .a, 1.b,

2.a, and 2.b.

- --------------------------------

Declare supported

feature(s) inoperable.

Immediately

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery

of loss of initiation

capability for

feature(s) in both

divkinn*

AND

(continued)

3.3-33

Amendment No. 135

HATCH UNIT 2

ACTIONS

ECCS Instrumentation

3.3.5.1

ACTIONS (continued)

CONDITION

F.

As required by Required

Action A.1 and referenced

in Table 3.3.5.1-1.

G.

As required by Required

Action A.1 and referenced

in Table 3.3.5.1-1.

H. Required Action and

associated Completion

-r;m- ,.*f rnndfltinn R fl. fl.

REQUIRED ACTION

F.1

Declare Automatic

Depressurization

System (ADS) valves

AND

F.2

G.1

Place channel in trip.

Declare ADS valves

AND

G.2

Restore channel to

OPERABLE status.

H.1

Declare associated

supported feature(s)

COMPLETION TIME

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery

of loss of ADS

initiation capability in

both trip systems

96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from

discovery of

inoperable channel

concurrent with HPCI

or reactor core

isolation cooling

(RCIC) inoperable

AND

8 days

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery

of loss of ADS

initiation capability in

both trip systems

96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from

discovery of

inoperable channel

concurrent with HPCI

or RCIC inoperable

AND

8 days

Immediately

E, F, or G not met.

,, ,i,,-r ,

3.3-36

Amendment No. 135

rim I Uil

U1,4l I

I

4

ECCS Instrumentation

3.3.5.1

SURVEILLANCE REQUIREMENTS

.----------..........---------------------.......

NOTES ----------------------------------------------------------

1.

Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.

2.

When a channel is placed in an inoperable status solely for performance of required

Surveillances, entry into associated Conditions and Required Actions may be delayed as

follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 3.c and 3.f; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for

Functions other than 3.c and 3.f provided the associated Function or the redundant

Function maintains initiation capability.

SURVEILLANCE

Perform CHANNEL CHECK.

Perform CHANNEL FUNCTIONAL TEST.

SR 3.3.5.1.3

Perform CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION.

Perform LOGIC SYSTEM FUNCTIONAL TEST.

FREQUENCY

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

92 days

18 months

3.3-37

Amendment No. 137

HATCH UNIT 2

I

ECCS Instrumentation

3.3.5.1

Table 3.3.5.1-1 (page 4 of 5)

Emergency Core Cooling System Instrumentation

APPLICABLE

CONDITIONS

MODES

REQUIRED

REFERENCED

OR OTHER

CHANNELS

FROM

SPECIFIED

PER

REQUIRED

SURVEILLANCE

ALLOWABLE

FUNCTION

CONDITIONS

FUNCTION

ACTION A.1

REQUIREMENTS

VALUE

4.

Automatic Depressurization

System (ADS) Trip System A

a.

Reactor Vessel Water

1,

2

F

k-113 inches

Level - Low Low Low,

2 (d), 3(d)

Level 1

b.

Drywell

1,

2

F

s 1.92 psig

Pressure - High

2(d), 3(d)

c.

Automatic

1,

G

5 114 seconds

Depressurization

2(d), 3(d)

System Initiation

Timer

d.

Reactor Vessel Water

1,

F

0 Oinches

Level - Low, Level 3

2(d), 3(d)

94

(Confirmatory)

e.

Core Spray Pump

1,

2

G

>Ž137 psig

Discharge Pressure -

2 (d), 3 (d)

and

High

-< 180 psig

f.

Low Pressure Coolant

1,

4

G

SR 3.3.5.1.1 i

112 psig

Injection Pump

2(d), 3(d)

and

Discharge Pressure -

< 180 psig

High

g.

Automatic

1,

2

G

r 12 minutes

Depressurization

2 (d), 3(d)

18 seconds

System Low Water

Level Actuation Timer

(continued)

(d)

With reactor steam dome pressure> 150 psig.

Amendment No. 135

HATCH UNIT 2

3.3-41

QUESTIONS REPORT

for HT2002

48. 264000K5.06 001

Unit 1 is operating at 75% RTP when the following actions occur:

Reactor Scram

D/G "A" and "B" start and attain proper speed and voltage

D/G "C" fails to start

Reactor Water Level

-15" increasing

Drywell Pressure

4.5 psig

Drywell Temperature

200OF

Startup Transformers 1 C and 1D are De-Energized

Which ONE of the following lists the major loads on the "1 B" D/G and the sequence

that they started?

A. Core Spray "B", LPCI "C", LPCI "D".

B. LPCI "C", LPCI "D", PSW "C".

C. LPCI "B", LPCI "C", LPCI "D".

DO LPCI "C", LPCI "D", PSW "D".

Reference: LT-LP-02801 Rev 3 pg 49 and 50 of 87.

Copy of Electrical Lineup

A. Incorrect since "B" Core Spray pump is powered from DIG "C" only.

B. Incorrect since "C" PSW pump only starts if the "A" PSW pump fails to start.

C. Incorrect since "B" LPCI pump starts from D/G "C" only.

D. Correct answer. PSW pump "D" starts since the "C" D/G has failed to start which

powers the "B" PSW pump.

RO Tier:

SROTier:

T2G1

Keyword:

D/G START SEQUENCE

Cog Level: C/A 3.4/3.5

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

52

Monday, June 24, zuuz u0:16:5

AlVI

QUESTIONS REPORT

for HT2002

7. 264000K5.06 001

In regards to the Diesel Generator Loading Sequence, which ONE of the following

describes why the engineered safety feature (ESF) loads are applied automatically in

approx. 10 second intervals?

A. prevent overloading the DG output breaker and causing it to trip on overcurrent

from starting motor-driven pumps.

B. minimize the initial voltage increase due to starting the motor-driven pumps.

C. prevent a differential lockout due to high starting current from the motor-driven

pumps.

Df minimize the initial voltage drop due to starting the motor-driven pumps.

Reference: FSAR Section 8.3.1.1.3 F Sequential Loading

A. Incorrect since output breaker can handle current from starting all D/G loads at once.

B. Incorrect since starting pumps causes a voltage decrease instead of an increase.

C. Incorrect since the differential lockout is caused by an internal generator fault on

different phases.

D. Correct answer.

(This may be too easy for an SRO Only question)

8

Friday, May 31, 2002 03:24:45 PM

QUESTIONS REPORT

for HT2002

1. 264000K5.06 001

In regards to the Diesel Generator Loading Sequence, the engineered safety feature

(ESF) loads are applied automatically in approx. 10 second intervals to:

A. prevent overloading the DG output breaker and causing it to trip on overcurrent.

B. minimize the initial voltage increase due to starting the motor-driven pumps.

C. prevent a differential lockout due to high starting current from the motor-driven

pumps.

D. minimize the initial voltage drop due to starting the motor-driven pumps.

Reference: FSAR Section 8.3.1.1.3 F Sequential Loading

A. Incorrect since output breaker can handle current from starting all D/G loads at once.

B. Incorrect since starting pumps causes a voltage decrease instead of an increase.

C. Incorrect since the differential lockout is caused by an internal generator fault on

different phases.

D. Correct answer.

(This may be too easy for an SRO Only question)

Friday, May 10, 2002 08:07:23 AM

Page 49 of 87

LT-LP-02801-03

DIESEL GENERATORS

Review Attachment 2 with

NOTE:

Refer to Attachment 2 for detailed

students

sequence.

3.

Diesel Generator Start Failure

a.

If the time delayed (T2A and T2B) relays time

out (7 seconds) before the Low Speed Relay

deenergizes them, the Start Failure Relay (SFR)

is energized (diesel generator has cranked for 7

seconds without firing and has failed to start).

b.

The energized Start Failure Relay energizes the

Shutdown Relay (SDR) which seals in and stops

the diesel generator.

c.

The diesel generator can be returned to a standby

condition after the shutdown relay is manually

reset using the pushbutton on P652. The diesel

generator may now be restarted after a 100

second time delay.

EO 14

5.

Diesel Generator Loading Sequence

Table 02

a.

If the diesel generator started due to a LOSP

signal, it will tie to its respective 4160 VAC

emergency bus and initiate LOSP/LOCA Timers

and Load Shed Relays which will:

1)

Prior to the Diesel tying to the bus, all

loads on the respective 4160 VAC

emergency bus are Loadshed.

NOTE:

Loadshed is a term used to

describe the tripping of power

supply breakers to non-essential

loads. This in conjunction with

time delayed starts prevents

overloading the diesels.

2)

Immediately start both Core Spray Pumps

and LPCI RHR Pump C if a LOCA signal

exists. (T=12 seconds)

Figs 09 & 14

EO 15a

MKV-2(1) Normal fan

MKV-1(2) Backup fans

3)

10 seconds later, LPCI RHR Pumps A, D

and B start if a LOCA exists (T=22

seconds)

4)

8 seconds later, PSW Pumps A and B start

(T=30 seconds)

5)

2 seconds later, PSW Pump C start if PSW

pump A failed to start. (T=32 seconds)

6)

2 seconds later, PSW Pump D start if PSW

pump B failed to start. (T=34 seconds)

b.

This loading sequence prevents overloading the

diesel generator due to motor starting current.

6.

Diesel Generator Shutdown Sequence

NOTE:

Refer to Attachment 2 for detailed

sequence.

B.

Diesel Generator Heating Ventilation System Operation

1 .

Generator Room Heating and Ventilation System

Each generator room contains three 50% heaters.

Each heater is rated at 12 kW. These heaters are

normally automatically controlled by separate wall

mounted thermostats to maintain the area temperature

above 40'F. The operating range of the heaters is 40

43°F.

There are three vents fans located in each generator

room. The MK V-2 fan is controlled by thermostat

X41-N012, and the MK V-I fans are controlled by

thermostat X41-NO11. At 95'F increasing the MK V

2 fan is auto started and at 97°F increasing the MK V

1 primary fan is auto started. The MK V-I fan is

tripped off at 94°F decreasing and the MK V-2 fan at

92°F decreasing.

a.

Interlocks

Ei,_'OR INTERLOCKS

TIM

S WILLTRIPA DIESEL GENERATOR:

SH

,TURE HIGH

9 LOW

I

sOnmNT

21 psig

230 degrees

205 degrees

9 psig

0.5 incses Water

1000 rpm

<250 rpm, 7 seconds after Diesel Start

LICABLE IN THE TEST MODE

ATOR'S TEST RELAYS RESULTS IN THE FOLLOWING:

ESEL GENERATOR EMERGENCY START

)F ASSOCIATED DIESEL GENERATOR OUTPUT BREAKER

iSSOCIATED DIESEL GENERATOR AND EITHER ITS NORMAL OR

IE OF START-UP TRANSFORMER SUPPLY BREAKER TO 4160V

"TO FAST TRANSFER WILL STILL OCCUR)

OF ASSOCIATED EMERGENCY 4160V BUS TO ITS ALTERNATE

ENERATOR TRIPS

L DEENERGIZE THE DIESEL GENERATOR TEST RELAYS.

0 SEC

12 SEC

22 SEC

30 SEC

32 SEC

34 SEC

DIG/"A"

START

TIE

CORE SPRAY A

LPCI A

ING"lB

START

TIE

LPCI C

LPCI D

PSW A

PSW C

IF PSWA FAILED TO START

PSW D

IF PSW B FAILED TO START

UI

4160 VAC "2F"

4160 VAC G

RHRSW PUMP 1B

RHRSW PUMP ID

PSW PUMP 1B

RHR PUMP lB

CORE SPRAY PUMP IB

DRYWELL CHILLER IB

iD

['ION

@,,

ri N.

600 VAC "I'D"

1R42-S008

"rs 7-Th

T

Tl

1

T

-

001

B C

2R 24-S0181

S

1C01-S001B

(oN

7'> ( K

7 K I

/

.2

_J

)*RBCCW

PUMP IB

UNIT 1 AND 2 REACTOR BLDG CRANES

-STATION SERVICE AIR COMP lB

1R24-SO1O

WATER FROM DIV 11

)

TRASH RAKE IBl

1P41-F313B PSW SIRAINER

[

2B BACKWASH

VENT FANIB

1

i R

TRAVEI.NG WAIER SCREEN I R

DIESEL

GENERATOR

IC

CIRC

R-E

F

1R24-S033

MEDICAL BUILDING FE"I0

I R24-S0

CHEM/HPARICA HVAC

VITAL AC NON-ESS RNI

I7AI-B040, 27AI-B040

RFPTA & BOILPUMP,

I NS4C005BIC006B

25-S045

I R24.

f-

MA

GENE]

UN

DIESEL

GENERATOR

1B

)

)UTBD MSIVs

NBD MSIVs

& P924

.Y

Xl

/

HNP-2-FSAR-8

The supply breakers between startup auxiliary transformers 2C and 2D

on the associated essential 4160-V bus are tripped.

When the last condition above is met, the possibility of one diesel generator

operating in parallel with any other diesel generator is precluded.

E.

Load Shedding

When the diesel generator breaker closes, the following load shedding has already

taken place:

The 4160-V loads and most nonessential 600-V loads are tripped, but the feeder

breakers to the 4160-600-V station service transformers supplying the essential

600-V load centers and their associated MCCs remain closed. This ensures

power continuity to vital 600-V auxiliaries such as the generator seal oil pumps and

instrumentation transformers even when a reactor trip does not accompany loss of

normal power.

F.

Sequential Loading

The diesel generator loading sequence is shown in table 8.3-3. Emergency loads

are shown in tables 8.3-4, 8.3-5, and 8.3-6.

Timing devices are provided to sequentially start the motors for each essential

load. The engineered safety feature (ESF) loads are applied automatically in

sequence at - 10-s intervals to minimize the initial voltage drop due to starting the

induction motor-driven pumps. This method of starting motors provides flexibility

in timing adjustment and independence of control. The tabulation of tables 8.3-3

through 8.3-6 assumes three diesel generators are available.

At time t-plus-30 s after a LOCA with all three essential buses available, four

residual heat removal (RHR) and two core spray (CS) pumps would be in

operation. Full flow injection or spray may still be prohibited by flow- or

pressure-sensing ESF interlocks. Failure of any one diesel or diesel battery and

its buses cannot prevent attainment of minimum safe shutdown requirement

regardless of which bus fails. The plant operator can manually drop off any

excess pumping capacity at any time t-plus-30 s but prior to proceeding into the

second phase of accident control. This occurs at approximately time t-plus-10 min

when reactor water level is stabilized and containment cooling begins.

The automatic starting and load sequencing times in the current design are more

restrictive than the timing assumptions made in the SAFER/GESTR-LOCA

analysis. The LOCA analysis supports a 34-s response time for CS and a 64-s

response time for LPCI.

At time t-plus-10 min, all diesel generator loading can be controlled by the plant

operator. The plant operator makes decisions as to which emergency loads may

8.3-8

REV 19 7/01

QUESTIONS REPORT

for HT2002

58. 295002AA2.02 001

Unit 2 is holding load at 25% RTP. The main turbine is on line when the steam seal

regulator fails closed.

Which ONE of the following describes the impact this will have on reactor power with

no operator action?

A. Reactor power will remain constant since seal steam is not required at this power

level.

BY Reactor power will decrease since condenser vacuum will decrease.

C. Reactor power will decrease since less steam is required from the reactor.

D. Reactor power will remain constant since the steam seal bypass valve automatically

opens to maintain seal steam pressure constant.

Reference: SI-LP-02501 Rev. SI-00 pg 6 of 13

A. Incorrect since condenser vacuum will decrease with Rx Power <28%. This will

cause reactor power to decrease.

B. Correct answer since seal steam is required up to 28% RTP or condenser vacuum

will decrease. If condenser vacuum decreases then reactor power will decrease.

C. Incorrect since this amount of steam being drawn off of reactor is negligible..

D. Incorrect since steam seal bypass valve does not automatically open.

RO Tier:

SRO Tier:

TIG2

Keyword:

MAIN CONDENSER

Cog Level:

MEM 3.2/3.3

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

62

Monday, June 24, 2002 08:16:53 AM

Page 6 of 13

SI-LP-02501-00

MAIN CONDENSER

D.

Steam Seal System

Fig 2

1.

The Steam Seal System contains one 100% capacity gland seal condenser,

two 100% capacity steam packing exhauster fans, associated piping and

valves.

2.

The Steam Seal System supplies sealing steam to the main turbine, RFPT

and shafts and various turbine valves, (main stop, control, combined

intercept, and bypass valves). Sealing steam can is supplied by the main

steam system.

3.

The excess steam from the steam seals is condensed in the gland seal

condenser and is returned as condensate to the main condenser. The air and

non-condensable gases are removed by the Steam Packing Exhauster and are

discharged via a 1.75 minute holdup volume to the main stack.

4.

At low power (< 30%) sealing steam prevents air from being drawn across

the seals and into the turbine which could result in a loss of condenser

vacuum or damage to the main turbine.

5.

At higher power (> 30%) pressure inside the turbine casing is greater than

sealing steam pressure therefore the internal steam tends to leak into the seal

cavity.

Fig 4

6.

When the steam pressure leaking out past the inner labyrinth exceeds the

seal steam supply pressure the steam seal regulator valve will close.

Pressure is controlled in the seal steam header by two pressure control

valves which dump to the main condenser.

IV. System Interfaces

A.

Power Supplies

Mechanical Vacuum pump is powered from 600V 2A (2R23-S001) [U1 600V IB

(1R23-S002)].

Steam Packing Exhauster fans are powered from 208V MCC2 A2 2R24-S039

(600V MCC IA R24-S005 for Ul).

QUESTIONS REPORT

for HT2002

1. 295002AA2.02 001

According to procedure 34AB-N61-002-1S, Main Condenser Vacuum Low, SELECT

the condition below which would require reducing reactor power to maintain condenser

vacuum > 25".

A. Inlet flow to holdup line is high.

B. Circulating Waterbox dP's are erratic.

C.r Circulating Water dT exceeds > 25 F.

D. Circulating Water Suction Bay has been < 116'.

Reference: Procedure 34AB-N61-002-1S, Rev. 4 pg 3 of 6

A. Incorrect since this requires the operator to check proper SJAE operation.

B. Incorrect since this would require the operator to vent the waterboxes and pumps.

C. Correct answer.

D. Incorrect since this would require the operator to vent the waterboxes and pumps.

Wednesday, April 24, 2002 10:57:46 AM

SOUTHERN NUCLEAR

PAGE

PLANT E. I. HATCH

3 OF 6

DOCUMENT TITLE:

DOCUMENT NUMBER: REVISIONNERSION

MAIN CONDENSER VACUUM LOW

34AB-N61-002-1S

NO:

0.4

4.0

SUBSEQUENT OPERATOR ACTIONS

NOTE

Decreasing Condenser Vacuum may result in saturated conditions within the Condenser and the

flashing of condensate. IF it becomes necessary OR desirable to break Vacuum, THEN refer to

34AB-C71-001-1 S, Scram Procedure, for placing the Condenser Low Vacuum Trip Bypass switches in

IBYPASS.

4.1

IF main condenser vacuum decreasing trend cannot be stopped prior to any automatic action

that could cause a scram OR complicate/prohibit continued operation, enter 34AB-C71-001-1S,

Scram Procedure, AND SCRAM the reactor.

4.2

IF RTP is <5%, attempt to restore main condenser vacuum with the mechanical vacuum pump,

per 34SO-N61-001-1S, Main Condenser Vacuum System and Closeout.

4.3

Confirm proper Circ Water System operation. IF a singlecirculating pump has tripped and

cannot be restarted enter 34AB-N71-001-1S, Loss of Circulating Water System AND

34SO-N71-001-1S, Circulating Water System, AND reduce reactor power per

34GO-OPS-005-1S, Power Changes to maintain a constant vacuum.

-J

4.4

CONFIRM CLOSED 1N22-F058A & B, Condenser Vacuum Breakers at IH11-P650.

4.5

CONFIRM/START all available cooling tower fans per 34SO-W24-001-IN, Cooling Towers.

4.6

CONFIRM normal hotwell level and adjust level as necessary per 34SO-N21-007-1S,

Condensate System.

4.7

IF Circulating Water dT exceeds > 25 OF, maintain condenser vacuum > 25" by reducing

reactor power per 34GO-OPS-005-1S, Power Changes.

,

4.8

IF Circulating Waterbox dP's are erratic ORcirc water suction bay has been <116', vent the

waterboxes and Circ water pumps per 34-SO-N71-001-1S, Circulating Water System. Notify

I&C to vent condenser waterbox dP instrument lines.

4.9

IF inlet flow to holdup line is high, confirm proper operation of the SJAE in accordance with

34SO-N61-001-1 S, Main Condenser Vacuum System and Closeout.

4.10

CONFIRM proper operation of the Steam Packing Exhauster per 34SO-N33-001-IS, Seal

Steam System.

4.11

Verify 1G31-F034 (Disch to Main Cndsr) AND 1G31-F035 (Drain to Waste Coil Tnk) on panel

1H1 1-P602 are not open at the same time, if so, restore to proper lineup per

34SO-G31-003-1 S, Reactor Water Cleanup System.

MGR-0001 Rev 3

QUESTIONS REPORT

for Revision2 HT2002

32. 295006G2.2.22 001

Unit 1 is operating at 100% RTP. The I & C Techs notify you at 0900 that "A" and "B"

APRM's will not generate a scram signal until the reactor is at 122% RTP.

Adjustments to the APRM's cannot be made for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Which ONE of the following describes the condition of the plant after the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has

expired? (Provide copy of Tech Spec 3.3.1.1 conditions and SR's)

A. The Unit is in Mode 2 as required by Required Action F.1.

B. The Unit is in Mode 3 as required by Required Action G.1.

C. The Unit is <28% RTP as required by Required Action E.I.

D. The Unit is at 100% RTP with one APRM bypassed and the other APRM channel in

trip.

Reference: Tech Spec section 3.3.1.1

Tech Spec table 3.3.1.1-1

A. Incorrect since this action would only be required if you cannot meet the Required

Action of Condition A. This can be met by bypassing one APRM and placing the other

APRM in trip.

B. Incorrect since Condition G would be entered if the APRM were INOPERABLE for

the Inop function. They are Inop for the Neutron Flux-High function.

C. Incorrect since you would only go below 28% RTP if the problem was turbine

related.

D. Correct answer. Bypass one APRM and then you can meet Condition A by placing

the other APRM in trip. This maintains the unit at the current power level as long as

you want.

RO Tier:

SROTier:

T1GI

Keyword:

SCRAM

Cog Level:

C/A 3.4/4.1

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

35

Friday, September 20, 2002 09:23:23 AM

QUESTIONS REPORT

for HT2002

64. 295006G2.2.22 001

Unit I is operating at 100% RTP. The Instrument Maintenace Techs notify you at 0900

that "A" and "B" APRM's will not generate a scram signal until the reactor is at 122%

RTP. Adjustments to the APRM's cannot be made for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Which ONE of the following describes the condition of the plant after the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has

expired?

A. The Unit is in Mode 2 as required by Required Action F.1.

B. The Unit is in Mode 3 as required by Required Action G.1.

C. The Unit is <28% RTP as required by Required Action E.1.

Df The Unit is at 100% RTP with one APRM bypassed and the other APRM channel in

trip.

Reference: Tech Spec section 3.3.1.1

Tech Spec table 3.3.1.1-1

A. Incorrect since this action would only be required if you cannot meet the Required

Action of Condition A. This can be met by bypassing one APRM and placing the other

APRM in trip.

B. Incorrect since Condition G would be entered if the APRM were INOPERABLE for

the Inop function. They are Inop for the Neutron Flux-High function.

C. Incorrect since you would only go below 28% RTP if the problem was turbine

related.

D. Correct answer. Bypass one APRM and then you can meet Condition A by placing

the other APRM in trip. This maintains the unit at the current power level as long as

you want.

RO Tier:

SROTier:

TIG1

Keyword:

SCRAM

Cog Level:

C/A 3.4/4.1

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

69

Monday, June 24, 2002 U0:I:54 M

QUESTIONS REPORT

for HT2002

1. 295006G2.2.22 001

Unit 1 is operating at 100% RTP. The Instrument Maintenace Techs notify you at 0900

that "A" and "B" APRM's will not generate a scram signal until the reactor is at 122%

RTP. Adjustments to the APRM's cannot be made for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Which ONE of the following describes the condition of the plant after the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has

expired?

A. The Unit is in Mode 2 as required by Required Action F.1.

B. The Unit is in Mode 3 as required by Required Action G.1.

C. The Unit is <28% RTP as required by Required Action E.1.

Do The Unit is at 100% RTP with one APRM bypassed and the other APRM channel in

trip.

Reference: Tech Spec section 3.3.1.1

Tech Spec table 3.3.1.1-1

A. Incorrect since this action would only be required if you cannot meet the Required

Action of Condition A. This can be met by bypassing one APRM and placing the other

APRM in trip.

B. Incorrect since Condition G would be entered if the APRM were INOPERABLE for

the Inop function. They are Inop for the Neutron Flux-High function.

C. Incorrect since you would only go below 28% RTP if the problem was turbine

related.

D. Correct answer. Bypass one APRM and then you can meet Condition A by placing

the other APRM in trip. This maintains the unit at the current power level as long as

you want.

Friday, May 31,2002 11:21:37 AM

RPS Instrumentation

3.3.1.1

3.3 INSTRUMENTATION

3.3.1.1

Reactor Protection System (RPS) Instrumentation

LCO 3.3.1.1

APPLICABILITY:

The FPS instrumentation for each Function in Table 3.3.1.1-1 shall be

According to Table 3.3.1.1 -1.

ACTIONS

S--------------------------------------------------- NOTE -------.-------..............................................

Separate Condition entry is allowed for each channel.

CONDITION

A.

One or more required

channels inoperable.

B. --------

---

NOTE ------------

Not applicable for

Functions 2.a, 2.b, 2.c,

2.d, and 2.f.

One or more Functions with

one or more reqluired

channels inoperable in both

trip systems.

REQUIRED ACTION

I COMPLETION TIME

A.1

A.2

B.1

B.2

Place channel in trip.

.

NOTE

Not 'applicable for

Functions 2.a, 2.b, 2.c,

2.d, and 2.f.

Place associated trip

system in trip.

Place channel in one

trip system in trip.

Place

trip.

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

6 hours

one trip system in

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

(continued)

Amendment No. 213

HATCH UNIT 1

3.3-1

RPS Instrumentation

3.3.1.1

ACTIONS (continued)

CONDITION

REQUIRED ACTION

COMPLETION TIME

C.

One or more Functions with

C.1

Restore RPS trip

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

RPS trip capability not

capability.

maintained.

D.

Required Action and

D.1

Enter the Condition

Immediately

associated Completion

referenced in

Time of Condition A, B,

Table 3.3.1.1-1 for the

or C not met.

channel.

E.

As required by Required

E.1

Reduce THERMAL

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

Action D.1 and referenced

POWER to < 28% RTP.

in Table 3.3.1.1-1.

F.

As required by Required

F.1

Be in MODE 2.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

Action D.1 and referenced

in Table 3.3.1.1-1.

G.

As required by Required

GA

Be in MODE 3.

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

Action D.1 and referenced

in Table 3.3.1.1-1.

H.

As required by Required

H.1

Initiate action to fully

Immediately

Action D.1 and referenced

insert all insertable

in Table 3.3.1.1-1.

control rods in core

cells containing one or

more fuel assemblies.

(continued)

6

4

Amendment No. 214

HATCH UNIT 1

3.3-2

RPS Instrumentation

3.3.1.1

.

ACTIONS (continued)

CONDITION

1.

As required by Required

Action D.1 and referenced

in Table 3.3.1.1-1.

J.

Required Action and

associated Completion

Time of Condition I not met.

REQUIRED ACTION

1.1

Initiate alternate method

to detect and suppress

thermal-hydraulic

instability oscillations.

AND

1.2

Restore required

channels to OPERABLE.

J.1

Be in MODE 2.

COMPLETION TIME

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

120 days

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

SURVEILLANCE REQUIREMENTS

- ---------------------------------------------------------

NOTES ----------------------------------------------------------

1.

Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2.

When a channel is placed in an inoperable status solely for performance of required

Surveillances, entry into associated Conditions and Required Actions may be delayed

for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE

FREQUENCY

SR 3.3.1.1.2

Perform CHANNEL CHECK.

NOTE ---------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

THERMAL POWER >:25% RTP.

-

Verify the absolute difference between the

average power range monitor (APRM) channels

and the calculated power is S 2% RTP while

operating at a 25% RTP.

7 days

(continued)

3.3-3

Amendment No. 213

<-

HATCH UNIT 1

If-- ggUl *)

RPS Instrumentation

3.3.1.1

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE

SIR 3.3.1.1.3

(Not used.)

SR 3.3.1.1.4

NOTE -------------------

Not required to be performed when entering

MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

entering MODE 2.

- --------------------------------------------------

Perform CHANNEL FUNCTIONAL TEST.

SIR 3.3.1.1.5

Perform CHANNEL FUNCTIONAL TEST.

SR 3.3.1.1.6

Verify the source range monitor (SRM) and

intermediate range monitor (IRM) channels

overlap.

SR 3.3.1.1.7

NOTE

-

Only required to be met during entry into MODE 2

from MODE 1.

- --------------------------------------------------

Verify the IRM and APRM channels overlap.

Calibrate the local power range monitors.

SR 3.3.1.1.9

Perform CHANNEL FUNCTIONAL TEST.

NOTE -------------

.....----------

For Function 2.a, not required to be performed

when entering MODE 2 from MODE 1 until

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

- -

eT---------------------------

Perform CHANNEL FUNCTIONAL TEST.

6

FREQUENCY

7 days

Prior to

withdrawing SRMs

from the fully

inserted position

a

7 days

1000 effective full

power hours

92 days

184 days

(continued)

Amendment No. 213

"->

HATCH UNIT 1

4

3.3-4

RPS Instrumentation

3.3.1.1

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE

SR 3.3.1.1.11

Verify Turbine Stop Valve - Closure and

Turbine Control Valve Fast Closure, Trip Oil

Pressure - Low Functions are not bypassed

when THERMAL POWER is; *28% RTP.

SR 3.3.1.1.12

Perform CHANNEL FUNCTIONAL TEST.

NOTES ---------------------------

1.

Neutron detectors are excluded.

2.

For Function 1, not required to be

performed when entering MODE 2 from

MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering

MODE 2.

-- --------------------------------------------------

Perform CHANNEL CALIBRATION.

SR 3.3.1.1.14

(Not used.)

Intermediate Range Monitor

Perform LOGIC SYSTEM FUNCTIONAL TEST.

SR 3.3.1.1.16 -

- ------------- NOTE ---------------------------

Neutron detectors are excluded.

Verify

Verify the RPS RESPONSE TIME is within limits.

FREQUENCY

184 days

18 months

18 months on a

STAGGERED

TEST BASIS

(continued)

Amendment No. 214

HATCH UNIT 1

I

3.3-5

RPS Instrumentation

3.3.1.1

MiIM~~I I ANhlF; REQUIIREMENTS (continued)

SURVEILLANCE

SR 3.3.1.1.17

t

Verify OPRM is not bypassed when APRM

Simulated Thermal Power is 2 25% and

recirculation drive flow is < 60% of rated

recirculation drive flow.

FREQUENCY

18 months

I

Amendment No. 213

-

HATCH UNIT 1

6

3.3-6

RPS Instrumentation

3.3.1.1

Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation

APPLICABLE

CONDITIONS

MODES OR

REQUIRED

REFERENCED

OTHER

CHANNELS

FROM

SPECIFIED

PER TRIP

REQUIRED

SURVEILLANCE

ALLOWABLE

FUNCTION

CONDITIONS

SYSTEM

ACTION D.1

REQUIREMENTS

VALUE

1.

a.

Neutron Flux- High

2

3

G

5120/125

divisions of full

scale

5(a)

3

H

%1201125

divisions of full

scale

b.

Inop

2

3

G

SR 3.3.1.1.4

NA

5 (a)

3

H

NA

2.

Average Power Range

Monitor

a.

Neutron Flux - High

2

3(c)

G

s 20% RTP

(Setdown)

SR 3.3.1.1.7

b. Simulated Thermal

1

3(c)

F

0.58W +

Power-High

SR 3.3.1.1.2

58% RTP

and

5115.5%

RTP(b)

c.

Neutron Flux -High

1

3()

F

!5 120% RTP

SR 3.3.1.1.2

d.

Inop

1,2

S(i)

G

Reactor Vessel Water Level

NA

(continued)

(a)

With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b)

0.58 W + 58% -0.58 AW RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating."

(c)

Each APRM channel provides inputs to both trip systems.

Amendment No. 214

HATCH UNIT 1

3.3-7

RPS Instrumentation

3.3.1.1

Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation

E

APPLICABLE

CONDITIONS

MODES OR

REQUIRED

REFERENCED

OTHER

CHANNELS

FROM

SPECIFIED

PER TRIP

REQUIRED

SURVEILLANCE

ALLOWABLE

FUNCTION

CONDITIONS

SYSTEM

ACTION D.1

REQUIREMENTS

VALUE

2.

Average Power Range Monitor

(continued)

e.

Two-out-of-Four Voter

f.

OPRM Upscale

3.

Reactor Vessel Steam Dome

Pressure - High

4.

Low, Level 3

5.

Main Steam Isolation Valve

Closure

6.

DryWenl Pressure - High

7.

Scram Discharge Volume

Water Level - High

a.

Resistance Temperature

Detector

b.

Float Switch

1,2

1

1,2

1

1,2

5 (a)

1,2

5(a)

G

2

3(c)

2

G

2

SR

F

8

2

G

H

G

H

SR

3.3.1.1.1

3.3.1.1.10

3.3A.1.15

3.3.1.1.13

3.3.1.1.1

33.1.1.18

3.3.1.1.10

3.3.1.1.13

3.3.1.1.17

3.3.1.1.1

3.3.1.1.9

3.3.1.1.13

3.3.1.1.15

3.3.1.1.1

3.3.1.1.9

3.3.1.1.13

3.3.1.1.15

3.3.1.1.9

3.3.1.1.13

3.3.1.1.15

3.3.1.1.1

3.3.1.1.9

3.3.1.1.13

3.3.1.1.15

3.3.1.1.9

3.3.1.1.13

3.3.1.1.15

3.3.1.1.9

3.3.1.1.13

3.3.1.1.15

NA

< 1085 psig

2: 0 inches

a

10% closed

1.92 psig

71 gallons

S 71 gallons

s 71 gallons

< 71 gallons

(continued)

(a)

With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(c)

Each APRM channel provides inputs to both trip systems.

4

Amendment No. 213

'w.

HATCH UNIT 1

3.3-8

RPS Instrumentation

3.3.1.1

Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation

APPLICABLE

CONDITIONS

MODES OR

REQUIRED

REFERENCED

OTHER

CHANNELS

FROM

SPECIFIED

PER TRIP

REQUIRED

SURVEILLANCE

ALLOWABLE

FUNCTION

CONDITIONS

SYSTEM

ACTION D.1

REQUIREMENTS

VALUE

8.

Turbine Stop Valve - Closure

a 28% RTP

4

E

SR 3.3.1.1.9

s 10% closed

SR 3.3.1.1.11

9.

Turbine Control Valve Fast

2:28% RTP

2

E

SR 3.3.1.1.9

Ž600 psig

Closure, Trip Oil Pressure-

SR 3.3.1.111

LOW

SR 3.3.1.1.16

10.

Reactor Mode Switch -

1,2

1

G

SR 3.3.1.1.12

NA

Shutdown Position

5(a)

1

H

SR 3.3.1.1.12

NA

11.

Manual Scram

1,2

1

G

NA

5 (a)

1

H

NA

a

(a)

With any control rod withdrawn from a core cell containing one or more fuel assemblies.

Amendment No. 214

<"->

HATCH UNIT 1

3.3-9

QUESTIONS REPORT

for HT2002

70. 295010AA2.06 001

A DBA LOCA has occurred on Unit 2 and the following conditions exist:

Drywell Pressure

51 psig and increasing at 2 psig/min

Reactor Water Level

-230 inches and increasing at 10"/min with RHR pumps

Bulk Drywell Temp

280°F

Torus Water Level

218" and increasing slowly

Which ONE of the following should be ordered by the Shift Supervisor?

A. Vent the Drywell IRRESPECTIVE of offsite radioligical release rates.

B!. Vent the Torus IRRESPECTIVE of offsite radioligical release rates.

C. Spray the Drywell after verifying within Drywell Spray Initiation Limit.

D. Enter the Severe Accident Guidelines (SAG's).

Reference: PC-1 Primary Containment Control

Drywell Spray Initiation Limit (Graph 8)

(Consider providing PC-1 and Graph 8)

A. Incorrect since Torus water level is below 300".

B. Correct answer.

C. Incorrect since spraying the Drywell is not allowed since RWL is below Top of Active

Fuel and RHR pumps are required to maintain adequate core cooling.

D. Incorrect since EOP's have direction to cover this situation.

Need to verify these answers with drawings.

RO Tier:

SROTier:

TIGI

Keyword:

DRYWELL PRESSURE

Cog Level: C/A 3.6/3.6

Source:

B

Exam:

HT02301

Test:

S

Misc:

TCK

Monday, June 24, 2002 08:16:55 AM

76

QUESTIONS REPORT

for HT2002

10. 295010AA2.06 001

A DBA LOCA has occurred on Unit 2 and the following conditions exist:

Drywell Pressure

51 psig and increasing at 2 psig/min

Reactor Water Level

-230 inches and increasing at 10"/min with RHR pumps

Bulk Drywell Temp

280 F

Torus Water Level

218" and increasing slowly

The Shift Supervisor SHOULD DIRECT:

A. Venting the Drywell IRRESPECTIVE of offsite radioligical release rate.

B.e Venting the Torus IRRESPECTIVE of offsite radioligical release rate.

C. Spray the Drywell after verifying within Drywell Spray Initiation Limit (Graph 8).

D. Entering the Severe Accident Guidelines (SAG's).

Reference: PC-1 Primary Containment Control

Drywell Spray Initiation Limit (Graph 8)

A. Incorrect since Torus water level is below 300".

B. Correct answer.

C. Incorrect since spraying the Drywell is not allowed since RWL is below Top of Active

Fuel and RHR pumps are required to maintain adequate core cooling.

D. Incorrect since EOP's have direction to cover this situation.

Need to verify these answers with drawings.

Thursday, April 04, 2002 11:36:27 AM

10

QUESTIONS REPORT

for HT2002

77. 295017G2.3.4 001

An event has occurred on Unit I that resulted in an individual getting injured. The

individual is disabled and is in a 100 R/Hr field. An individual is standing by to save the

disabled individuals life (He has NOT volunteered). The job will require being in the

radiation field for 13 minutes. After conferring with HP Supervision the

has determined that

is the maximum amount of dose allowed per

73EP-EIP-017-OS, Emergency Exposure Control, for this lifesaving attempt.

(CHOOSE the answer that correctly fills in the blanks.)

A. Shift Supervisor, 10 Rem

B. Emergency Director, 10 Rem

C. Shift Supervisor, 25 Rem

DO Emergency Director, 25 Rem

References: Procedure 73EP-EIP-017-OS, Emergency Exposure Control pg 4 & 6 of

13.

A. Incorrect since the Emergency Director has the responsibility to make these

decisions.

B. Incorrect since the maximum dose allowed without volunteering is 25 Rem.

C. Incorrect since the Emergency Director has the responsibility to

decisions.

D. Correct answer.

RO Tier:

Keyword:

EMERGENCY EXPOSURE

Source:

N

Test:

S

SRO Tier:

Cog Level:

Exam:

Misc:

make these

TIG1

C/A 2.5/3.1

HT02301

TCK

Monday, June 24, 2002 08:16:56 AM

83

QUESTIONS REPORT

for HT2002

1. 295017G2.3.4 001

An event has occurred on Unit 1 that resulted in an individual getting hurt. The

individual is disabled and is in a 100 R/Hr field. An individual is standing by to save the

disabled individuals life (He has NOT volunteered). The job will require being in the

radiation field for 13 minutes. After conferring with HP Supervision the

has determined that

is the maximum amount of dose allowed per

73EP-EIP-017-OS, Emergency Exposure Control, for this lifesaving attempt.

(CHOOSE the answer that correctly fills in the blanks.)

A. Shift Supervisor, 10 Rem

B. Emergency Director, 10 Rem

C. Shift Supervisor, 25 Rem

D., Emergency Director, 25 Rem

References: Procedure 73EP-EIP-017-OS, Emergency Exposure Control pg 4 & 6 of

13.

A. Incorrect since the

decisions.

Emergency Director has the responsibility to make these

B. Incorrect since the maximum dose allowed without volunteering is 25 Rem.

C. Incorrect since the

decisions.

D. Correct answer.

Monday, April 29, 2002 01:35:39 PM

Emergency Director has the responsibility to make these

1

GEORGIA POWER COMPANY

PAGE

4

OF 13

PLANT E.I.

HATCH

_

DOCUMENT TITLE:

DOCUMENT NUMBER:

REVISION

NO:

EMERGENCY EXPOSURE CONTROL

73EP-EIP-017-OS

2 ED 1

REFERENCE

7.0

PROCEDURE

Emergency response personnel may receive exposure under a variety of

circumstances in order to assure protection of others and of valuable property.

These exposures will be justified if the risks permitted to the workers are

acceptably low, AND the costs to others that are avoided by their actions

outweigh the risks to which workers are subjected.

7.1

SAVING OF HUMAN LIFE

Where the potential risk of radiation hazard following the nuclear incident

is

such that life would be in jeopardy, or that there would be severe effects

on the public health or loss of property detrimental to the public safety,

the following criteria for saving of human life shall apply:

7.1.1

In consultation with HP supervision, the Emergency Director will evaluate

the risks involved versus the benefits to be gained by considering the

following:

7.1.1.1

The reliability of the prediction of radiation injury.

Consideration

must be given to limits of error associated with specific instruments

AND techniques used to estimate the dose rate.

This is especially

crucial when the estimated dose approximates 100 REM or more.

7.1.1.2

Assessment of the capability of reducing inherent risks from the

hazard through the use of appropriate mechanisms such as protective

equipment, remote manipulation equipment or similar means.

7.1.1.3

The probable effects of acute exposure that may be incurred AND

numerical estimates of the delayed effects.

These effects are listed

in Attachment 3,

Emergency Worker Risks and Delayed Health Effects

Associated With Large Doses of Radiation.

7.1.1.4

The probability of success of the emergency action.

7.1.2

Make exposure authorizations in accordance with subsection 7.4 Emergency

Exposure Guidelines.

7.2

PROTECTION OF HEALTH AND PROPERTY

7.2.1

When the Emergency Director in consultation with HP supervision, deems-it

necessary to reduce a hazard OR potential hazard to acceptable levels to

prevent a substantial loss of property, an exposure of up to, but not to

exceed,

10 REM may be received by individuals participating in

the

operation.

MGR-0001 Rev.

1

GEORGIA POWER COMPANY

PLANT E.I.

HATCH

DOCUMENT TITLE:

EMERGENCY EXPOSURE CONTROL

DOCUMENT NUMBER:

73EP-EIP-017-OS

PAGE

6

OF 13

REVISION

NO:

REVISION

NO:

2 ED I

7.4

EMERGENCY EXPOSURE GUIDELINES

7.4.1

The Emergency Director will establish the exposure limits for the

emergency response personnel based on the following Emergency Response

Personnel Exposure Guides:

NOTE

These guidelines do not establish a rigid upper

limit of exposure.

The Emergency Director may use

his/her judgment in establishing the appropriate

limit.

NOTE

No thyroid limit is specified for lifesaving action

since the complete loss of the thyroid may be

considered an acceptable risk for saving a life;

however, thyroid exposure must be minimized through

the use of respiratory protection and/or KI

tablets.

C

EMERGENCY RESPONSE PERSONNEL EXPOSURE GUIDES

Dose Limit*

Activity

Condition

(REM)

5

all

n/a

10

protecting valuable

lower dose not practicable

property

25

life saving or protection

lower dose not practicable

of large populations

1

>25

life saving or protection

only on a voluntary basis to

of large populations

persons fully aware of the

risks involved

This limit is expressed as the sum of the effective dose equivalent (EDE) and the

committed effective dose equivalent (CEDE)

I

M

R/

MGR-000l Rev.

1

K->

-

6

I

QUESTIONS REPORT

for Revision4HT2002

10. 295025EA2.04 001

Unit 2 Scrammed from High Drywell pressure due to a small leak in the Recirc piping.

The following conditions currently exist:

Drywell pressure

3.5 psig (decreasing)

Drywell temperature

245OF (decreasing)

Torus level

185 inches (increasing)

Torus temperature

160OF (decreasing)

Reactor pressure

600 psig (decreasing)

Torus sprays and cooling

running

Based upon the above conditions the Shift Supervisor has determined that injection

into the RPV from sources external to primary containment must be terminated.

Which ONE of the following identifies the systems that are EXEMPTED from

termination?

A. Systems needed for adequate core cooling, boron injection and CRD.

B. Systems needed to shutdown the reactor, boron injection and CRD.

C. Systems needed for boron injection, CRD and RCIC.

D. Systems needed for adequate core cooling, fire fighting and boron injection.

References: PC-1 Primary Containment Control

Suppression Pool Level High

SRV Tail Pipe Level Limit (Graph 6)

A. Correct answer due to torus water level CANNOT be maintained below SRV Tail

Pipe Level Limit (graph 6).

B. Incorrect since systems needed to shutdown the reactor are not exempted.

C. Incorrect since these systems are excepted when preventing all injection during an

ATWS.

D. Incorrect since fire fighting systems are only excepted during SC Secondary

Containment Control.

RO Tier:

SRO Tier:

TIGI

Keyword:

TORUS LEVEL

Cog Level:

C/A 3.9/3.9

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

11

Friday, October 11, 2002 06:51:26 AM

QUESTIONS REPORT

for HT2002

85. 295025EA2.04 001

Unit 2 Scrammed from High Drywell pressure due to a small leak in the Recirc piping.

The following conditions currently exist:

Drywell pressure

3.5 psig (decreasing)

Drywell temperature

245OF (decreasing)

Torus level

185 inches (increasing)

Torus temperature

160°F (decreasing)

Reactor pressure

600 psig (decreasing)

Torus sprays and cooling

running

Based upon the above conditions the Shift Supervisor has determined that injection

into the RPV from sources external to primary containment must be terminated.

Which ONE of the following identifies the systems that are EXEMPTED from

termination?

A! Systems needed for adequate core cooling, boron injection and CRD.

B. Systems needed for boron injection and CRD.

C. Systems needed for boron injection, CRD and RCIC.

D. Systems needed for adequate core cooling, fire fighting and boron injection.

References: PC-I Primary Containment Control

Suppression Pool Level High

SRV Tail Pipe Level Limit (Graph 6)

A. Correct answer due to torus water level CANNOT be maintained below SRV Tail

Pipe Level Limit (graph 6).

B. Incorrect since these systems are excepted per CP-2 RPV Flooding.

C. Incorrect since these systems are excepted when preventing all injection during an

ATWS.

D. Incorrect since fire fighting systems are only excepted during SC Secondary

Containment Control.

RO Tier:

SRO Tier:

TIGI

Keyword:

TORUS LEVEL

Cog Level:

C/A 3.9/3.9

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

Monday, June 24, 2002 08:16:57 AM

92

QUESTIONS REPORT

for HT2002

1. 295025EA2.04 001

Unit 2 Scrammed from High Drywell pressure due to a small leak in the Recirc piping.

The following conditions currently exist:

Drywell pressure

3.5 psig (decreasing)

Drywell temperature

245 F (decreasing)

Torus level

185 inches (increasing)

Torus temperature

160 F (decreasing)

Reactor pressure

600 psig (decreasing)

Torus sprays and cooling

running

Based upon the above conditions the Shift Supervisor has determined that injection

into the RPV from sources external to primary containment must be terminated. Which

systems are EXCEPTED from termination?

A.r Systems needed for adequate core cooling, boron injection and CRD.

B. Systems needed for boron injection and CRD.

C. Systems needed for boron injection, CRD and RCIC.

D. Systems needed for adequate core cooling, fire fighting and boron injection.

References: PC-1 Primary Containment Control

Suppression Pool Level High

A. Correct answer due to torus water level CANNOT be maintained below SRV Tail

Pipe Level Limit (graph 6).

B. Incorrect since these systems are excepted per CP-2 RPV Flooding.

C. Incorrect since these systems are excepted when preventing all injection during an

ATWS.

D. Incorrect since fire fighting systems are only excepted during SC Secondary

Containment Control.

Tuesday, April 09, 2002 10:51:43 AM

RCIC or HPCI is operaling

defeat high torus water level

suction transfer logic per

31 EO-EOP-1 00-2S

PERFORM CONCURRENTLY

I Maintain torus water level below

SRV Tail Pipe Level Umit(Graph 6) per

3480-Ell-0 10-2S or 34GO-OPS-087-2S

SMWAIT UNTIL

Maintain torus water level

below 215 in. per 34SO-Ell-010-2S or

34GO-OPS-087-28

WAIT UNTIL

torus water level

CANNOT be maintained

below SRV Tail Pipe Level Limitf

(Graph 6)

RCAPERFORMon ACONCURRENTLY

SWAIT LINT L

/

~torus water level

S~ANDI

/

~reactor pressure/

o usANDNOT

be maintand

J

below SRV Tail Pipe Level Umit

T(Graph 6)

torus water level

CANNOT be maintained

C

below 215 in.

-

ýWAIT UNTIL

torus water level

QCANNOT

be maintained

below 300 in.

Determine primary containmentwater level

per 31EO-EOP-105-2S

IF primary contal

and torus prea

I

maintainedb

Pressure Umi

E LF drell sprat

BEFORE d

ýWAIT UNTIL.

torus water level

AND

reactor pressure

CANNOT be restored and maintained

below SRV Tail Pipe Level Uimit

(Graph 6)

-. , (EMERGENCY

DEPRESS IS REQUIRED

MflTF 9

Terminate drywell sprays

AND

Terminate injection into RPV from sources

external to primary containment EXCEPT

systems required for:

"o adequate core cooling

"C boron injection

o CRD

Terminate injection into RPV from sources

external to primary containment EXCEPT

systems required for.

o adequate core cooling

o boron injection

o CRD

SGRAPH 6

180

174

S

U)

C

168

162

156

UNIT -1

SRV TAIL PIPE LEVEL LIMIT

150

0

200

400

600

NOTE: M,

SPDS Emergency Displays in place of this Graph.

bv

800

RPV PRESSURE (psig)

1000

1200

Q

QUESTIONS REPORT

for Revision2 HT2002

33. 295030EA2.01 001

Unit 2 has developed a leak in the Torus and the Shift Supervisor has entered PC-1

Primary Containment Control. Mechanical Maintenance and Health Physics have been

dispatched to investigate and repair the leak. Mechanical Maintenance has reported

that the leak should be stopped in approximately 30 minutes. Torus level is currently at

120" and is dropping at a rate of 1" per minute.

Which ONE of the following should be directed by the Shift Supervisor?

A. wait until Torus level drops to 110" and order HPCI tripped.

B. wait until Torus level drops to 98" and order Emergency Depressurization per CP-1.

C' order HPCI tripped and depressurize reactor through the Main Turbine Bypass

Valves irrespective of cooldown rate.

D. order HPCI tripped and depressurize reactor through the Main Turbine Bypass

valves without exceeding a 100OF/hr cooldown rate.

References: LR-LP-20310 Rev. 05 pg. 20 - 23 of 96

RC RPV CONTROL (NON-ATWS)

PC-1 PRIMARY CONTAINMENT CONTROL

A. Incorrect since actions should be taken before they are hit if the trend indicates that

the limit will be met.

B. Incorrect since actions should be taken before they are hit if the trend indicates that

the limit will be met.

C. Correct answer.

D. Incorrect since it is acceptable to exceed the 100 F/hr cooldown rate when you

anticipate blowdown per overide in PC RPV Control (Non-ATWS).

RO Tier:

SROTier:

TIGI

Keyword:

TORUS LEVEL

Cog Level: C/A 4.1/4.2

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

...

36

Friday, September 20, ZuUL u0:23:23 AMl

QUESTIONS REPORT

for HT2002

89. 295030EA2.01 001

Unit 2 has developed a leak in the Torus and the Shift Supervisor has entered PC-1

Primary Containment Control. Mechanical Maintenance and Rad Protection have been

dispatched to investigate and repair the leak. Mechanical Maintenance has reported

that the leak should be stopped in approximately 30 minutes. Torus level is currently at

120" and is dropping at a rate of 1" per minute.

Which ONE of the following should be directed by the Shift Supervisor?.

A. wait until Torus level drops to 110" and order HPCI tripped.

B. wait until Torus level drops to 98" and order Emergency Depressurization per CP-1.

C' order HPCI tripped and depressurize reactor through the Main Turbine Bypass

Valves irrespective of cooldown rate.

D. order HPCI tripped and depressurize reactor through the Main Turbine Bypass

valves without exceeding a 1 00OF/hr cooldown rate.

References: LR-LP-20310 Rev. 05 pg. 20 - 23 of 96

RC RPV CONTROL (NON-ATWS)

PC-1 PRIMARY CONTAINMENT CONTROL

A. Incorrect since actions should be taken before they are hit if the trend indicates that

the limit will be met.

B. Incorrect since actions should be taken before they are hit if the trend indicates that

the limit will be met.

C. Correct answer.

D. Incorrect since it is acceptable to exceed the 100 F/hr cooldown rate when you

anticipate blowdown per overide in PC RPV Control (Non-ATWS).

ROTier:

SROTier:

TIGI

Keyword:

TORUS LEVEL

Cog Level:

C/A 4.1/4.2

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

96

Monday, June 24, 2002 08:16:57 AM

QUESTIONS REPORT

for HT2002

1. 295030EA2.01 001

Unit 2 has developed a leak in the Torus and the Shift Supervisor has entered PC-1

Primary Containment Control. Mechanical Maintenance and Rad Protection have been

dispatched to investigate and repair the leak. Mechanical Maintenance has reported

that the leak should be stopped in approximately 30 minutes. Torus level is currently at

120" and is dropping at a rate of 1" per minute. The Shift Supervisor should:

A. wait until Torus level drops to 110" and order HPCI tripped.

B. wait until Torus level drops to 98" and order Emergency Depressurization per CP-1.

C. order HPCI tripped and depressurize reactor through the Main Turbine Bypass

Valves irrespective of cooldown rate.

D. order HPCI tripped and depressurize reactor through the Main Turbine Bypass

valves without exceeding a 100 F/hr cooldown rate.

References: LR-LP-20310 Rev. 05 pg. 20 - 23 of 96

RC RPV CONTROL (NON-ATWS)

PC-1 PRIMARY CONTAINMENT CONTROL

A. Incorrect since actions should be taken before they are hit if the trend indicates that

the limit will be met.

B. Incorrect since actions should be taken before they are hit if the trend indicates that

the limit will be met.

C. Correct answer.

D. Incorrect since it is acceptable to exceed the 100 F/hr cooldown rate when you

anticipate blowdown per overide in PC RPV Control (Non-ATWS).

Tuesday, April 09, 2002 02:14:15 PM

Page 20 of 96

LR-LP-20310-05

PRIMARY CONTAINMENT CONTROL (PC-1 & 2)

Q: Why are these paths

concurrent and not in series?

A: As level decreases, HPCI

would need to be secured,

however, the plant would not

necessarily need to be

depressurized.

PERFORM CONCURRENTLY

I

Maintain torus water level

above 98 in. per 34SO-E21-001-2S

or 34GO-OPS-087-2S

I

Maintain torus water level

above 110 in. per 34SO-E21-001-2S

or 34G0-OPS-087-2S

The two paths for the low level condition direct control of torus water

level relative to the elevation of the downcomer openings and the

elevation of the top of the HPCI exhaust line, respectively.

Therefore, these two paths must be executed in parallel since the

actions required in one path may or may not have to be accomplished at

the same time as the actions in the other path.

HPCI PATH WILL BE DISCUSSED LATER

98" is the elevation of the

downcomer openings

PRIOR to tows level dropping

below 98", the RPV should be

depressurized.

The RPV is not permitted to remain at pressure if suppression of steam

discharged from the RPV into the drywell cannot be assured. When the

downcomer vent openings are not adequately submerged, any steam

discharged from the RPV into the drywell may not condense in the

torus before tows pressure reaches unacceptable levels.

Emergency RPV depressurization will be required at or before the point

at which this low water level condition occurs (98in.).

I

4

WAIT UNTIL

torus water level

CANNOT be maintained

above 98in.

4

Page 21 of 96

LR-LP-20310-05

PRIMARY CONTAINMENT CONTROL (PC-1 & 2)

At this point, if the torus level is

above 98", there is no need for

emerg depressurization.

NOTE: Results of the Bodega Bay Mark I containment tests

indicate 95% steam condensation may be achieved from a

vertical downcomer vent that discharges at a level six inches

above the torus surface.

As long as torus water level remains at or above the elevation of the

downcomer vent openings, the need to emergency depressurize the

RPV due to torus heatup is dictated by the Heat Capacity Temperature

Limit. Actions associated with the Heat Capacity Temperature will be

discussed in the SP/F path.

PERFORM CONCURRENTLY

RC[A] point A

Stress that if torus level cannot be

prevented from decreasing to 98

inches, enter the RC flowchart.

EO 9

ASK

What does entering

RC flowchart accomplish ?

EO 10

ASK How is the Emergency

Depressurization Flowchart CP-l

reached ?

The determination that torus level cannot be maintained above 98

inches should be made BEFORE reaching the actual limit based on

level trend or other plant conditions.

In other words, the operator should ANTICIPATE the need to

perform RC[A] and enter the correct chart.

Entering either RC or RCA at point "A" assures that, if possible, the

reactor is scrammed and shutdown by control rod insertion before RPV

depressurization is initiated.

Directing that the RPV Control EOP be entered, rather than explicitly

stating here "Initiate a reactor scram," coordinates actions currently

being executed if the RPV Control has already been entered. (Note that

the RPV Control requires initiating a reactor scram only if one has not

previously been initiated.)

In addition, entry into RC or RCA must be made

because it is through these flowcharts that the transfer

to the "Emergency RPV Depressurization"

contingency is effected.

j

I

EMERGENCY DEPRESS IS REQUIRED

EO 11

Discuss reason for Emerg RPV

Depressurization if unable to

maintain above 98inches.

Depressurizing the RPV ensures

steam can be condensed while

adequate water level exists.

<S

EO 12

Depressurizing the RPV when torus water level cannot be maintained

above 98 inches is done to prevent failure of the containment or

equipment necessary for the safe shutdown of the plant.

The RPV is not permitted to remain at pressure if suppression of steam

discharged from the RPV cannot be assured.

Therefore the vessel is depressurized while the tows can still support a

blow down and places the RPV in the lowest possible energy state.

This precludes the possibility of a primary containment failure and the

resultant uncontrolled radioactivity release,

This override flag DOES NOT direct the operator to leave the

SP/L flow path.

The override flag DOES direct the operator to exercise the

EMERGENCY DEPRESSURIZATION IS REQUIRED

overrides in all other flow paths that are currently being

performed, especially the override on the RC[A] pressure

path that directs the operator to the Emergency

Depressurization path on CP-1.

HPCI PATH WILL NOW BE DISCUSSED

1

PERFORM CONCURRENTLY

I

Maintain torus water level

above 110 in per 34SO-E21-001-2S or

34GO-OPS-087-2S

I

EO 13

If level decreases < 110 inches,

the torus air space will be directly

pressurized with HPCI running.

EO 14

HPCI is secured and prevented

from operating (PTL) even if

needed to maintain RWL

The torus level needs to be maintained above the discharge of the HPCI

steam turbine exhaust line to ensure adequate steam condensing. This

precludes possible primary containment failure due to over

pressurization caused by HPCI steam exhaust discharging directly into

the torus air space.

The determination that the torus level cannot be maintained above 110

inches CAN BE MADE BEFORE reaching the actual limit based on

trend or other plant conditions. As soon as this determination is made,

the operator proceeds to the next step and secures HPCI.

I

PTL.

Aux Oil Pump is placed in

ASK Why must HPCI be tripped

at 110" "?

ASK

Why is RCIC operation is

allowed ?

HPCI exhaust press trip approx

140 psig where as RCIC is about

40 psig.

Trip and prevent operation of HPCI

irrespective of adequate core cooling

Operation of the HPCI System with its exhaust discharge line (located

at 110 in.) not submerged will directly pressurize the torus air space.

HPCI operation is therefore secured, and prevented from restarting, to

preclude the occurrence of this condition.

The consequences of not doing so may extend to failure of the primary

containment from over pressurization, and thus HPCI must be secured

irrespective of adequate core cooling concerns.

No instruction regarding RCIC operation is included in this step (or in

an equivalent step) for two reasons:

1.

The exhaust flow rate of RCIC is approximately equal to

that of decay heat, and is thus consistent with the basis used

for determining the Primary Containment Pressure Limit.

2.

Elevated torus pressure will cause the RCIC turbine to trip

much sooner than the HPCI turbine.

WAIT UNTIL

torus water level

CANNOT be maintained

above 110 in.

I

4*

RC RPV CC

C

RWL below + 3 in.

A condition which requires reactors

and reactor power is above 5%

)

C

IF ALL control roc

or beyond posi

control rods ar

shutdown rod

I

[

I

S

WHILE PERFORMING THE FOLLOWING

IF Emergency Depress is anticipated,

biTEN rapidly depress with

EXCEPT for low RWLtI

main turbine bypass valves,

I

irrespective of the resulting

cooldown rate.

LF EMERGENCY DEPRESS IS,

THEN perform Emergency Depress

OR HAS BEEN, REQUIRED

I

IF RWL CANNOT be determined

THEN perform RPV Flooding

IF drywell pressure is above 1.85 psig

THEN prevent injection from CS and LPCI

pumps per 31EO-EOP-114-2S

C_

A

B

C

C $

I

1

0

I I

I

V

SP/

NOTE2

If fuel failure is suspected consult with

I

Plant Chemistry prior to discharging water 1

GO TO ANY entry condition

on PC-2

B

PERFORM CONCURRENTLY

I

Maintain torus water level

above 98 in. per 34$0-E21-001-2S or

34GO-OPS-087-2S

I

ýWAIT U`NTILý

trswtr

level

CANNOT be maintained

above 98 in.

PERFORM CONCURRENTLY

RC[A] point A

Maintain torus water level

above 110 in. per 34SO-E21-001-2S or

34GO-OPS-087-2S

,-ý WAIT UNTIL

" ~

~torus

water level

'

CANNOT be maintained

above 110 in.

Trip and prevent operation of HPCI

irrespective of adequate core cooling

EMERGENCY DEPRESS IS REQUIRED'

I

ir PRIMP

I%*Ss+/-Ig

7

"torus water level

NNOT

146b main* a

in.

BELOW 146

IsABOVE

15

toruswaterleve

Monitor and control torus water level

between 146 in. and 150 in. using:

"o CS for low water levels per

3480-E21-001-28

"o RHR for high water levels per

3480-Ell-010-2I

D

I

-

/

WAI T UN T I L

U

QUESTIONS REPORT

for Revision2 HT2002

34. 29503 1G2.4.4 001

Unit 1 is operating at 100% power when a leak in the Drywell develops. Reactor water

level is trending down and Drywell pressure, temperature and level are trending

upward. The SRO orders a reactor SCRAM with the following conditions occuring:

RWL initially reaches +2" and stabilizes at +15"

Drywell pressure reaches 1.83 psig and stabilizes

Drywell temperature currently at 147 0F and rising slowly

Torus level currently at 148" and rising slowly

Reactor pressure at 920 psig and steady

6 control rods stuck at position 02 and all others fully inserted

Which ONE of the following actions are required by the Shift Supervisor under these

conditions?

A. Enter RCA RPV Control (ATWS) and take actions to ensure reactor stays shutdown

under all conditions.

B. Enter PC-1 and PC-2 and take actions to prevent reaching entry conditions.

C" Enter RC RPV Control (Non-ATWS) and take actions to stabilize plant.

D. Entry into EOP's not required since an entry condition does not exist at this time.

S~34AB-C71-OO1-2S,

Scram Procedure is entered.

References: LR-20308 Entry Conditions

LR-20310 Entry Conditions

LR-20328 ATWS Conditions

A. Incorrect answer. Does not meet conditions for ATWS. All rods are at position 02 or

beyond.

B. Incorrect answer. Does not meet entry conditions yet. Not appropriate to take

actions to prevent meeting entry conditions.

C. Correct answer.

D. Incorrect answer. EOP entry is required since conditions were previously met. RWL

<+3".

-

-.

-

-.-

~.

-,~a

an

AftA37

Friday, September 2u, 2UU2 U0:2J3: 3wVI

RO Tier:

Keyword:

EOP RPV CONTROL

Source:

N

Test:

S

QUESTIONS REPORT

for Revision2 HT2002

SRO Tier:

Cog Level:

Exam:

Misc:

Friday, September 20, 2002 09:23:23 AM

T1GI

C/A 4.0/4.3

HT02301

TCK

38

QUESTIONS REPORT

for HT2002

91. 295031G2.4.4 001

Unit I is operating at 100% power when a leak in the Drywell develops. Reactor water

level is trending down and Drywell pressure, temperature and level are trending

upward. The SRO orders a reactor SCRAM with the following conditions occuring:

RWL innitially reaches +2" and stabilizes at +15"

Drywell pressure reaches 1.83 psig and stabilizes

Drywell temperature currently at 147 0F and rising slowly

Torus level currently at 148" and rising slowly

Reactor pressure at 920 psig and steady

6 control rods stuck at position 02 and all others fully inserted

Which ONE of the following actions are required by the Shift Supervisor under these

conditions?

A. Enter RCA RPV Control (ATWS) and take actions to ensure reactor stays shutdown

under all conditions.

B. Enter PC-1 and PC-2 and take actions to prevent reaching entry conditions.

C. Enter RC RPV Control (Non-ATWS) and take actions to stabilize plant.

D. Entry into EOP's not required since an entry condition does not exist at this time.

34AB-C71-001-2S, Scram Procedure is entered.

References: LR-20308 Entry Conditions

LR-20310 Entry Conditions

LR-20328 ATWS Conditions

A. Incorrect answer. Does not meet conditions for ATWS. All rods are at position 02 or

beyond.

B. Incorrect answer. Does not meet entry conditions yet. Not appropriate to take

actions to prevent meeting entry conditions.

C. Correct answer.

D. Incorrect answer. EOP entry is required since conditions were previously met. RWL

<+3".

98

Monday, June 24, 2002 08:16:57 AM

RO Tier:

Keyword:

EOP RPV CONTROL

Source:

N

Test:

S

Monday, June 24, 2002 08:16:57 AM

QUESTIONS REPORT

for HT2002

SRO Tier:

TIGI

Cog Level:

C/A 4.0/4.3

Exam:

HT02301

Misc:

TCK

99

QUESTIONS REPORT

for HT2002

14. 29503 1G2.4.4 001

Unit 1 is operating at 100% power when a leak in the Drywell develops. Reactor water

level is trending down and Drywell pressure, temperature and level are trending

upward. The SRO orders a reactor SCRAM with the following conditions occuring:

RWL innitially reaches +2" and stabilizes at +15"

Drywell pressure reaches 1.83# and stabilizes

Drywell temperature currently at 147 F and rising slowly

Torus level currently at 148" and rising slowly

Reactor pressure at 920# and steady

6 control rods stuck at position 02 and all others fully inserted

What actions are required by the Shift Supervisor under these conditions?

A. Enter RCA RPV Control (ATWS) and take actions to ensure reactor stays shutdown

under all conditions.

B. Enter PC-1 and PC-2 and take actions to prevent reaching entry conditions.

C. Enter RC RPV Control (Non-ATWS) and take actions to stabilize plant.

D. Entry into EOP's not required since an entry condition does not exist at this time.

References: LR-20308 Entry Conditions

LR-20310 Entry Conditions

LR-20328 ATWS Conditions

A. Incorrect answer. Does not meet conditions for ATWS. All rods are at position 02 or

beyond.

B. Incorrect answer. Does not meet entry conditions yet. Not appropriate to take

actions to prevent meeting CorrectnLamer.entry conditions. *

C. Correct answer.

D. Incorrect answer. EOP entry is required since conditions were previously met. RWL

<+3".

Thursday, April 04, 2002 11:36:28 AM

14

QUESTIONS REPORT

for Revision2 HT2002

35. 295033G2.3.10 001

Unit 1 is operating at 100% RTP. The Fuel Movement Team is moving fuel bundles in

the Unit 1 Fuel Pool to get ready for an upcoming outage. There is a malfunction

associated with the mast and a fuel bundle is dropped in the pool. This caused some

fuel pins to break and radiation levels are increasing on the Refuel Floor and around

the Fuel Pool pumps. The following radiation levels exist on Unit 1:

Refuel Floor

1500 mR/hr

Spent Fuel Pool Demin Equip

2000 mR/hr

Fuel Pool Demin Panel

75 mR/hr

Which ONE of the following actions should the Shift Supervisor order?

(Provide copy of Unit 1 SC-Secondary Containment Control)

A. Scram the reactor and evacuate the Reactor Building per 73EP-RAD-O01-OS,

Radiological Event.

B.' Commence a Normal Unit Shutdown and evacuate the associated High Rad areas.

C. Scram the reactor and commence a Reactor Blowdown per the EOP's.

D. Announce the High Rad condition over the public address system and evacuate the

affected areas. Reactor operation should not be affected.

References: 73EP-RAD-001-OS, Radiological Event Rev. 1.1 pg 4 of 7

SC - SECONDARY CONTAINMENT CONTROL

A. Incorrect since a reactor scram is not required by the EOP's.

B. Correct answer.

C. Incorrect, a reactor blowdown is not required since there is not a primary system

discharging into secondary containment.

D. Incorrect since the reactor should be shutdown per the EOP's.

RO Tier:

SROTier:

T1G2

Keyword:

EOP PC CONTROL

Cog Level: C/A 2.9/3.3

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

flfl

q,ýo

ll.O

A"A

39

Friday, S)eptembrn[ 20, 2004 09.:.*

,v

QUESTIONS REPORT

for HT2002

93. 295033G2.3.10 001

Unit 1 is operating at 100% RTP. The Fuel Handlers are moving fuel bundles in the

Unit 1 Fuel Pool to get ready for an upcoming outage. There is a malfunction

associated with the mast and a fuel bundle is dropped in the pool. This caused some

fuel pins to break and radiation levels are increasing on the Refuel Floor and around

the Fuel Pool pumps. The following radiation levels exist on Unit 1:

Refuel Floor

Spent Fuel Pool Demins

Fuel Pool Demin Panel

1500

2000

75

mR/hr (Max Safe Operating value = 1000

mR/hr (Max Normal Operating value = 50

Which ONE of the following actions should the Shift Supervisor order?

A. Scram the reactor and evacuate the Reactor Building per 73EP-RAD-O01-OS,

Radiological Event.

B! Commence a Normal Unit Shutdown and evacuate the associated High Rad areas.

C. Scram the reactor and commence a Reactor Blowdown per the EOP's.

D. Announce the High Rad condition over the public address system and evacuate the

affected areas. Reactor operation should not be affected.

73EP-RAD-001-0S, Radiological Event Rev. 1.1 pg 4 of 7

SC - SECONDARY CONTAINMENT CONTROL

A. Incorrect since a reactor scram is not required by the EOP's.

B. Correct answer.

C. Incorrect, a reactor blowdown is not required since there is not

discharging into secondary containment.

D. Incorrect since the reactor should be shutdown per

RO Tier:

SRO Tier:

Keyword:

EOP PC CONTROL

Cog Level:

Source:

N

Exam:

Test:

S

Misc:

a primary system

the EOP's.

TIG2

C/A 2.9/3.3

HT02301

TCK

101

Monday, June 24, 2002 08:16:58 AM

mR/hr))

mR/hr)

References:

QUESTIONS REPORT

for HT2002

1. 295033G2.3.10 001

Unit 1 is operating at 100% RTP. The Fuel Handlers are moving fuel bundles in the

Unit 1 Fuel Pool to get ready for an upcoming outage. There is a malfunction

associated with the mast and a fuel bundle is dropped in the pool. This causes some

fuel pins to break and radiation levels are going up on the Refuel Floor and around the

Fuel Pool pumps. The following radiation levels exist on Unit 1:

Refuel Floor

1500 mR/hr (Max Safe Operating value = 1000 mR/hr))

Spent Fuel Pool Demins

2000 mRPhr (Max Safe Operating value = 1000 mR/hr)

Fuel Pool Demin Panel

75 mRlhr (Max Normal Operating value = 50 mR/hr)

The Shift Supervisor should:

A. Order the Control Board Operator to Scram the reactor and evacuate the Reactor

Building per 73EP-RAD-001-OS, Radiological Event.

B!. Order the Control Board Operator to commence a Normal Unit Shutdown and

evacuate the associated High Rad areas.

C. Order the Control Board Operator to Scram the reactor and commence a Reactor

Blowdown per the EOP's.

D. Have the Control Board Operator announce the High Rad condition over the public

address system and evacuate the affected areas. Reactor operation should not be

affected.

References: 73EP-RAD-001-0S, Radiological Event Rev. 1.1 pg 4 of 7

SC - SECONDARY CONTAINMENT CONTROL

A. Incorrect since a reactor scram is not required by the EOP's.

B. Correct answer.

C. Incorrect, a reactor blowdown is not required since there is not a primary system

discharging into secondary containment.

D. Incorrect since the reactor should be shutdown per the EOP's.

Tuesday, April 30, 2002 01:41:50 PM

GEORGIA POWER COMPANY

PAGE

4 OF

7

PLANT E.I.

HATCH

I

DOCUMENT TITLE:

DOCUMENT NUMBER:

REVISION NO:

RADIOLOGICAL EVENT

73EP-RAD-001-OS

1 ED 1

4

I

REFERENCE

7.0

PROCEDURE

7.1

INVOLVED PERSONNEL

CAUTION

PERSONNEL MUST NOT BE SENT INTO AN AREA OF

UNKNOWN RADIATION CONDITIONS WITHOUT HP COVERAGE,

DOSIMETRY AND APPROPRIATE PROTECTIVE EQUIPMENT.

i

7.1.1

Control RoomPersonnel

Upon determining that a Radiological Event has occurred, the Shift

Supervisor will perform the following actions:

7.1.1.1

Direct the Control Room Operator to make the following announcement

over the public address system:

A RADIOLOGICAL EVENT IS OCCURRING.

ABOVE NORMAL RADIATION

(OR

AIRBORNE RADIOACTIVITY)

EXISTS IN THE

(location)

AREA.

EVACUATE AND STAY CLEAR OF THE

(location)

AREA(S).

7.1.1.2

Direct the Control Room Operator to repeat the announcement a second

time.

7.1.1.3

Contact the Health Physics Office to assist in investigating the

condition.

Inform HP of the indicated dose rate of the area, IF the

event was initiated due to an alarming ARM,

and any other pertinent

information, e.g., dropped fuel bundle, indication of leak, etc..

7.1.1.4

Attempt to confirm accuracy of alarmed ARMs and effluent monitors by

directing that the status of ARMs and effluent monitors near or

associated with incident area be checked for recent or sudden change.

7.1.1.5

Check habitability of Control Room by observing radiation monitors OR

possible automatic isolation of control room ventilation.

7.1.1.6

Ensure that START HIST (history) light on the SPDS keyboard is

ILLUMINATED;

if

not, simultaneously DEPRESS the CTRL and START HIST

keys.

Cancel or continue history as directed by the SOS.

7.1.1.7

Observe Control Room instrumentation and controls.

Implement

corrective action to eliminate cause of this abnormal condition, IF

possible, from the Control Room.

MGR-0001 Rev.

1

HVAC Isolation

"o

Unit 1 and Unit 2 Refuel Floor

HVAC isolation

"o Unit 1 and Unit 2 SBGT initiation

per 34AB-T22-O03-1 S

TH~FN restart Refuel Floor HVAC per

34SO-T41-006-IS

If necessary defeat high dqywell pressure

'/L isolation interlocks

EO-EOP-100-1S

THEN 'reseetd Reactor Building HVAC per

34SO-T41-005-1S

If necessary defeat high drywell pressure

ANDl

low RWL isolation Interlocks

per31EO-EOP-100-1S

3NCURRENTLY

SIC'

WAIT UNTIL

area radiation level

is above

Maximum Normal Operating Radiation Level

(Table 6)

INTIL

ollowing

ye

crating Water Level

) 5):

n sump water level

iter level

ump

Pumps to restore

level below Maximum

Vater Level (Table 5)

tg CANNOT be

alned below Maximum

Nater Level (Table 5):

in sump water level

Iar level

aems disc&,ning water

ua EXC

-tems

luate &,.Ang

)actor

nary containment

Isolate ALL systems discharging into area

EXCEPT systems required to:

"o assure adequate core cooling

"o shut down reactor

"o suppress fire

"o maintain primary containment

integrity

PERFORM CONCURRENTLY

I

primary system

Into secondary containment

(Tdi

arg bgleac o cooan

Shut down reactor per 34GO-OPS-013-IS

or34GO-OPS-f14-1S

I

I ANY area radiation level reaches

Maximum Safe Operating Radiation Level

(Table 6)

PERFORM CONCURRENTLY

RC(A) pantA

WAIT UNTIL

area radiation level

F0Is

above

Maximum Safe Operating Radiation Level

In more than one area

(TableS6)

EMERGENC

DEPRESS IS REQUIRIED

RR - RADIOACTIVITY RELEASE CONTROL

C

rr

Offslte radioactivity release rate

above 0.57 mR/hr

WHII F= PFRFORMIWG TIHI FOLLOWING

'I

IF

PR IM AR Y

tfN

TA INM FM TFV 1 fl

rlf

lM f

TR FN prit

f

lPq n

Hta

ntor

Thn

B

C

D

E

F

I

WHL

PE F RMN

THE* FOUL

ING.....

.

.....

area radiation level

is above

Maximum Safe OperatingRaitoLel

in more than one area

(Table 6)

I

QUESTIONS REPORT

for Revision2 HT2002

36. 295034EA2.02 001

Which ONE of the following conditions would most likely cause a Secondary

Containment Ventilation High Radiation isolation?

A. A leak has developed on the RHRSW side of the RHR Heat Exchanger and the

water level in the room is approximately 1/2" deep.

B. A Recirc Pump seal failure which causes Drywell Pressure to exceed 1.85 psig.

C. The packing is blown on a Startup Level Control Valve (SULCV) and the area is

blanketed in steam.

Df A leak has developed upstream of the HPCI Stop Valve but it is not large enough to

cause a high temperature isolation of HPCI.

References: SI-LP-01303 Rev. Sl-00, Figures 10.

A. Incorrect since the RHRSW side of the HX is not highly radioactive.

B. Incorrect since the drywell will contain the Recirc Pump seal leakage.

C. Incorrect since the feedwater reg valve is in the turbine building.

D. Correct answer since this steam leak is directly off of main steam and has the

potential of being highly radioactive.

RO Tier:

SROTier:

TIG2

Keyword:

SECONDARY CONTAIN

Cog Level:

C/A 3.7/4.2

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

Friday, September 20, 2002 09:23:23 AM

40

QUESTIONS REPORT

for HT2002

94. 295034EA2.02 001

Which ONE of the following conditions would most likely cause a Secondary

Containment Ventilation High Radiation isolation?

A. A leak has developed on the RHRSW side of the RHR Heat Exchanger and the

water level in the room is approximately 1/2" deep.

B. A Recirc Pump seal failure which causes Drywell Pressure to exceed 1.85 psig.

C. The packing is blown on a Feedwater Reg Valve and the area is blanketed in

steam.

D. A leak has developed upstream of the HPCI Stop Valve but it is not large enough to

cause a high temperature isolation of HPCI.

References: SI-LP-01303 Rev. SI-00, Figures 10.

A. Incorrect since the RHRSW side of the HX is not highly radioactive.

B. Incorrect since the drywell will contain the Recirc Pump seal leakage.

"-*~

C. Incorrect since the feedwater reg valve is in the turbine building.

D. Correct answer since this steam leak is directly off of main steam and has the

potential of being highly radioactive.

RO Tier:

SRO Tier:

T1G2

Keyword:

SECONDARY CONTAIN

Cog Level:

C/A 3.7/4.2

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

102

Monday, June z4, zuuz u0:16:58 AM

QUESTIONS REPORT

for HT2002

1. 295034EA2.02 001

SELECT the condition that would most likely cause a Secondary Containment

"Ventilation High Radiation isolation.

A. A leak has developed on the RHRSW side of the RHR Heat Exchanger and the

water level in the room is approximately 1/2" deep.

B. A firemain leak has developed in the CRD repair area on Unit 1 and is draining into

the floor drains.

C. The packing is blown on a Feedwater Reg Valve and the area is blanketed in

steam.

D.r A leak has developed upstream of the HPCI Stop Valve but it is not large enough to

cause a high temperature isolation of HPCI.

References: SI-LP-01303 Rev. SI-00, Figures 6, 10, 11 and 12.

A. Incorrect since the RHRSW side of the HX is not highly radioactive.

B. Incorrect since firemain is not highly radioactive and whatever is getting wet should

not go airborne until it is isolated and dries.

C. Incorrect since the feedwater reg valve is in the turbine building.

D. Correct answer since this steam leak is directly off of main steam and has the

potential of being highly radioactive.

Thursday, April 11, 2002 02:31:54 PM

C

-4

0

w

0 w

-L

0

c

Xh

x

ca

FL

to m

x4

-D

CD

C

AO

ah

QUESTIONS REPORT

for Revision5HT2002

1. 295035G2.1.7 001

A tornado was observed moving toward the plant 15 minutes ago. Meteorological

instruments have detected wind speeds in excess of 100 mph. The annunciator for

"RB INSIDE TO OUTSIDE AIR DIFF PRESS LOW" has just alarmed and the Inside

Rounds SO reports air rushing in and then out through a crack in the Reactor Bldg wall

on the 158' EL. The following conditions exist for Secondary Containment:

Reactor Power

Both Units at 100% RTP

Rx Bldg Dp

fluctuating between 0" and +.25" Hg

Rx Bldg Vent System

system isolated

Rx Bldg Vent Rad level

I mR/hr

Area water levels

normal

Which ONE of the following describes the appropriate actions the Shift Supervisor

should take?

(Provide copy of 73EP-EIP-001-OS)

A. Declare an ALERT. Initiate actions for Secondary Containment System being

Inoperable.

B! Declare a SITE AREA EMERGENCY. Initiate actions for Secondary Containment

System being Inoperable.

C. Declare an ALERT. No actions required for Secondary Containment System.

D. Declare a SITE AREA EMERGENCY. No actions required for Secondary

Containment System.

References: 73EP-EIP-001-0S Rev. 14.2 pg 16 & 22 of 47

SC - Secondary Containment Control

A. Incorrect since should declare a Site Area Emergency due to tornado damage.

B. Correct answer.

C. Incorrect since should declare a Site Area Emergency due to tornado damage.

D. Incorrect since actions are required by Tech Specs since the Containment is

Inoperable.

RO Tier:

SRO Tier:

T1G2

Keyword:

SECONDARY CONTAIN

Cog Level:

C/A 3.7/4.4

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

I

Monday, October 28, 2002 08:02:11 AM

QUESTIONS REPORT

for Revision2 HT2002

37. 295035G2.1.7 001

A tornado was observed moving toward the plant 15 minutes ago. Meteorological

instruments have detected wind speeds in excess of 100 mph. The annunciator for

"RB INSIDE TO OUTSIDE AIR DIFF PRESS LOW" has just alarmed and the Inside

Rounds SO reports air rushing in and then out through a crack in the Reactor Bldg wall

on the 158' EL. The following conditions exist for Secondary Containment:

Reactor Power

Rx Bldg Dp

Rx Bldg Vent System

Rx Bldg Vent Rad level

Area water levels

Both Units at 100% RTP

fluctuating between 0" and +.25" Hg

system isolated

I mR/hr

normal

Which ONE of the following describes the appropriate actions the Shift Supervisor

should take?

(Provide copy of 73EP-EIP-001-OS)

A. Declare an ALERT due to loss of containment. Restart Reactor Bldg Vent System,

commence a shutdown on both units and be in Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B! Declare a SITE EMERGENCY due to damage caused by the tornado. Restart

Reactor Bldg Vent System, restore containment to OPERABLE status within 4

nlnrn hnth ,nite

in M*,,n4

q within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SVerify or Start the SBGT S stem reduce reactor power on both units t

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , declare an UNUSUAL EVENT f power reduction not

D. Declare an ALERT due to damage caused by the tornado. Restart Re

Vent System and SBGT System, maintain reactor power at Shift Supe

discretion.

References: 73EP-EIP-001-0S Rev. 14.2 pg 16 & 22 of 47

SC - Secondary Containment Control

A. Incorrect since declare a Site Emergency due to tornado damage. 4

B. Correct answer.

o < 15% RTP

completed on

actor Bldg

rvisors

AV

C. Incorrect since there is no direction to start SBGT and no direction to lower power to

<15%. The 15% is the power level where primary containment is not required.

D. Incorrect since no direction to start SBGT and reactor power should be lower due to

Friday, September 20, 2002 09:23:24 AM

41

RO Tier:

Keyword:

SECONDARY CONTA1

Source:

N

Test:

S

QUESTIONS REPORT

for Revision2 HT2002

SRO Tier:

N

Cog Level:

Exam:

Misc:

Friday, September 20, 2002 09:23:24 AM

TIG2

C/A 3.7/4.4

HT02301

TCK

42

QUESTIONS REPORT

for HT2002

95. 295035G2.1.7 001

A tornado was observed moving toward the plant 15 minutes ago. Meteorological

instruments have detected wind speeds in excess of 100 mph. The annunciator for

"RB INSIDE TO OUTSIDE AIR DIFF PRESS LOW" has just alarmed and the outside

operator reports that it looks like part of the Reactor Bldg siding is loose. The following

conditions exist for Secondary Containment:

Reactor Power

Both Units at 100% RTP

Rx Bldg Dp

fluctuating between 0" and +.25" Hg

Rx Bldg Vent System

system isolated

Rx Bldg Vent Rad level

1 mR/hr

Area water levels

normal

Which ONE of the following describes the appropriate actions the Shift Supervisor

should take?

A. Restart Reactor Bldg Vent System, commence a shutdown on both units and be in

Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , declare ALERT due to loss of containment.

B.! Restart Reactor Bldg Vent System, restore containment to OPERABLE status

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or place both units in Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , declare a SITE

EMERGENCY due to damage caused by the tornado.

C. Verify or Start the SBGT System, reduce reactor power on both units to < 15% RTP

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , declare an UNUSUAL EVENT if power reduction not completed on

time.

D. Restart Reactor Bldg Vent System and SBGT System, maintain reactor power at

Shift Supervisors discretion, declare ALERT due to damage caused by tornado.

References: 73EP-EIP-001-0S Rev. 14.2 pg 16 & 22 of 47

SC - Secondary Containment Control

A. Incorrect since declare a Site Emergency due to tornado damage.

B. Correct answer.

C. Incorrect since there is no direction to start SBGT and no direction to lower power to

<15%. The 15% is the power level where primary containment is not required.

D. Incorrect since no direction to start SBGT and reactor power should be lower due to

loss of containment and Site Emergency should be declared.

103

Monday, June 24, 2UU2 ub:16:b8 AM

RO Tier:

Keyword:

Source:

Test:

SECONDARY CONTAIN

N

S

UESTIONS REPORT

for HT2002

SRO Tier:

Cog Level:

Exam:

Misc:

Monday, June 24, 2002 08:16:58 AM

TIG2

C/A 3.7/4.4

HT02301

TCK

104

QUESTIONS REPORT

for HT2002

1. 295035G2.1.7 001

A tornado was observed moving toward the plant 15 minutes ago. The annunciator for

"RB INSIDE TO OUTSIDE AIR DIFF PRESS LOW" has just alarmed and the outside

operator reports that it looks like part of the Reactor Bldg siding is loose. The following

conditions exist for Secondary Containment:

Reactor Power

Both Units at 100% RTP

Rx Bldg Dp

fluctuating between -.1" and +.25" Hg

Rx Bldg Vent System

system isolated

Rx Bldg Vent Rad level

1 mR/hr

Area water levels

normal

The Shift Supervisor actions should be:

A. Restart Reactor Bldg Vent System, commence a shutdown on both units and be in

Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , declare ALERT due to loss of containment.

B! Restart Reactor Bldg Vent System, restore containment to OPERABLE status

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or place both units in Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , declare a SITE

EMERGENCY due to damage caused by the tornado.

C. Verify or Start the SBGT System, reduce reactor power on both units to < 15% RTP

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , declare an UNUSUAL EVENT if power reduction not completed on

time.

D. Restart Reactor Bldg Vent System and SBGT System, maintain reactor power at

Shift Supervisors discretion, declare ALERT due to damage caused by tornado.

References: 73EP-EIP-001-0S Rev. 14.2 pg 16 & 22 of 47

SC - Secondary Containment Control

A. Incorrect since declare a Site Emergency due to tornado damage.

B. Correct answer.

C. Incorrect since there is no direction to start SBGT and no direction to lower power to

<15%. The 15% is the power level where primary containment is not required.

D. Incorrect since no direction to start SBGT and reactor power should be lower due to

loss of containment and Site Emergency should be declared.

Wednesday, May 08, 2002 04:26:53 PM

1

SOUTHERN NUCLEAR

PLANT E.I. HATCH

DOCUMENT TITLE:

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS

PAGE 22 OF 47

DOCUMENT NUMBER:

73EP-EIP-001-OS

REVISIONNERSIONT

NO:

14.2

6s

10.0 - NATURAL PHENOMENON, (continued)

I

Emergency conditions exist WHEN:

HIGH WINDS EXIST:

HIGH WINDS are indicated by:

Any tornado observed onsite

Any hurricane force winds projected onsite with windspeed > 75 mph

SAny tornado observed striking the operating facility (areas within the protected

area and the 230 Kv and 500 Kv switchyards)

Any hurricane observed. onsite with sustained windspeeds at design level

L> 94.5 mph)

SOS/ED judgment

The observation of damage from an onsite tornado with windspeed in excess of

meteorological instruments range (>100 mph)

Sustained windspeeds in excess of meteorological

mph)

instruments range (>100

AND

Either unit NOT in Cold Shutdown

END - HIGH WINDS

->

[NATURAL PHENOMENON - CONTINUED TO NEXT PAGE]->

MGR-0001 Rev. 3

CAUTION

The value of any emergency actions, which may require movement of

plant personnel, must be judged against the danger to personnel or

nuclear safety.

S

A

E

i--

G

E

N

N A

U L

E E

R

T

4

I

SOUTHERN NUCLEAR

PLANT E.I. HATCHI

DOCUMENT TITLE:

DOCUMENT NUMBER:

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS

73EP-EIP-001-OS

PAGE 16 OF 47

REVISIONNERSION

NO:

14.2

Is_

7.0 - LOSS OF CONTAINMENT

Emergency conditions exist WHEN:

NOTE

NUE is to be declared upon commencing Load Reduction.

A LOSS OF PRIMARY OR SECONDARY CONTAINMENT INTEGRITY OCCURS as indicated by the

inability to meet any one of the requirements WITHIN the time limit established by the

applicable unit's TS.

See Section 11.0, Hazards to Plant Operation, for determination of Alert Classification.

See Section 11.0, Hazards to Plant Operation for determination of Site Area Emergency

Classification.

-See Section 22.0, Multiple Symptoms and Other Conditions, for determination of General

Emergency Classification.

7-

LOSS OF

END

CONTAINMENT

E

MGR-0001 Rev. 3

SOUTHERN

NUC

PLANT

. .

C

I

STPLANT E

SC - SECONDARY CONTAINMENT CONTROL

to:Nucxhr

C

Areaorfloor dmiaiu

tv a rn s levell eove

Tube Maximums Nomna, Dapmersau Water Level]

ken or NAC

ieaest redleton Wlee Mabo.e

lTulfbte 6 Maximum NMusl Operftan Fraatio

ee

D0feantiat iressre

wateorebo.

in o' . watera

T

SPRIMARYCONTAINMENTFLOODING

THE

erdnhellOpsand

IS OR HAS BEEN REQUIRED

enter the Severe Atcideut Guidlines

Gs

PERFORM CONCURRENTLY

I

eus*iluble area coolers

ad areas,

sacondary conalinment radnation

niion does NOT ealst

W

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

...

arate

tititowig:

Rental Fluor HYAC

per 34S0-T41 -1W-S

Reactor

rteding

HVAD

per 54SO-T41"0t-IS

WAIT UN"TIL

'blent E- dsltrenliat tsmeatelre

It abtove

oINorma Oprefin Tmpretroiar

(Table 4)

7LL a,

c9n10 into ares

T.s,

I to:

ire a1

cooling

stain ptimay cornamemantre

Idly

I

.RFORM CON C UINTLy

atur

SCIL

WAIT UNTIL

ONE of the flloiine

Is shone

Maximum No"rn Opeationg Water tenv

(Table 6):

O A oor dranlnamp etatertenl

Operelt alilable aump pumps to retore

and maintaln mter tevne below MEourrm

Normal Oueatno Water Level (Tubla 5)

I

IF ONE orf t

foloomig CANNOT be

rent oed and maintained below Maximum

Natinal Operating Water Level (Table 5):

o My floor drain somp maret levl

o ANy areawater leel

THENISOte ALL

Uystems dbPmcharging

muter

Into strmp or area EXCEPT syStears

reluired to:

O0 catureaderurate core colln,

O shut down reacor

O .upprfss fire

O maintain pdmrry containment

Intagrity

PERPORMCONCURRENTLy

WAIT UNTIL

ia flacurein

reacor co~olan

rtutusecondary containment

I-'

BFORE

ANY area esther level reaches

,.Mo.ma Sate Domitno, Water Level

(Tablet5)

PERFORM CONCURRENTLY

ROIA poNtA

WAIT UNTIL

area wother level

Masinahm SaeOero~at'ingWter Level

in morellaren ometae

(Teble 0)

(EMERGENCY DEPRESS IS REQUIRED)

WAIT UNTIL

eree water level

is ubove

Maximum Sate Operutna Water Level

in mare then ore area

(TableS)

Shot down reactor per3400OPS0fl45

or 3400OFS.014.1S

lENT

"URES

aF Nomuaw as Sare

SIC'

" WAIT Tul

ra

rem radat_

leveo

tint

No thl~

.a NO .

Isttm

sh

vel

Matniuam Nortnut Operetin Radiautio

(Table S)

toailst ALLsyttnmt ditchurat

into

EXCEPT sYstemsrlresmoa to:

0 assuraxeeeuatoe cmcaniog

o supress fire

o

enslnaIn pri*naXcnainment

PERFORM O*NCURRENTLY

WAIT UNTIL

pitmaty syttern

isdiec~arguiimug

racor oon

into econdorycontainmentm

(Table 7)

BEFORE

kYnareardan

NOW

vel. reaches

Maximumn Sore Operebu Rodlaton Level

(Table 6)

PERFORM CONCURRENTLY

RC(A) pointA

WAIT UNTIL

area

raiamnlee

PMasinvm Sama

Opeatming Radiation Level

in mare Soon ove arear

(EMERGENCY.F DERES ISREOUIRED

RR -RADIOACTIVITY RELEAS

SOtter. redsOact~vty rleat

aret

shove 0 S7 ruPlhr

i t

LE PRIMARY CONTAINMENT FLOODING IM

saxt the tOPic

IS OR HAS ShEN REOLIIRPED

I

Ser

Acl*,d,

C Ez

i[

WHMpTHFORM'GTEGOLLOWINO

L

T TurbieeBulkifitnHVACioshuldOWn

HE r.a Trarurm

roiurel Per 34

isolate ALL Primary systoem discharogng

Nomanc coolant into ares outsie pnmsry

aMd secondary ronanarrnts (Tuse 7)

EXCEPT systems raruired to:

o asreadesutne cor coling

o shut dow reantor

o nalnottin tmarycntttelmrnent

ruintnfd

WtHILE PERFORMHING

THE FOLLOWING

4,5

IE MOYUniitI orUn 2 secondary

co'.

a'rent ':

conS

2t,

tVAC eahsaus radationtlevetexcsadsee

0 Unt 1 and Unit 2 Readclorul~ding

isolation serpoint (Table 14)

I

WAC Iscoalo

"0

Unit I ntd Unit2 RatedaFloor

HVAC isecation

"I o Unit 1 n

Unit 2 SBGT initiation

I per 34A,0.Tfl-0O)-1S

fRefetl Floor WAD isolates

'T

n restar Reruel Floor (aAD per

SI

04S0-T41-OC-1S

8 Unit1 I orUnit 2 secondary conainment

i

If ,eceasetay deot h*l dt

dell

pressure

radiation conditiondoes NOT exist

R

low

F

soruonUintadncts

I

per 31EO.EOP.I00-1S

if Reactor Buei* HVAC Isolates

If

restart Reodor sBilding HVAC per

I

34S0-T41-005-1S

a Unit 1 or U=R 2

rec

arsconlainnment

If nocestay defeuthta

llt

d"

pressrue

mrdabon

Boaditon

does NOT avast

I

.

i

RWL holatimon intedos

I

per 31ECEOP-100-iS

r M--- an

Shut unnon a

or 34GO-O

J

T1

QUESTIONS REPORT

for HT2002

107. G2.1.22 001

Unit 2 has been shutdown for a refueling outage. After 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months the following

conditions exist:

Reactor Mode Switch:

Refuel

Reactor Temperature:

1650F and steady

Reactor Pressure:

0 psig

All reactor vessel head closure bolts are fully tensioned.

All rods are IN.

Which ONE of the following is correct Mode of Operation for Unit 2?

A! Mode 2

B. Mode 3

C. Mode 4

D. Mode 5

References: Tech Spec section 1.1, Table 1.1-1

Modified from question #84 on 1995 SRO exam

A. Correct answer.

B,C,D Incorrect (See table 1.1-1) Unless a Special Operations Tech Spec is invoked

then the reactor changes modes when moving the mode switch to refuel.

RO Tier:

SRO Tier:

T3

Keyword:

MODE

Cog Level: C/A 2.8/3.3

Source:

M

Exam:

HT02301

Test:

S

Misc:

TCK

116

Monday, June 24, 2002 08:16:59 AM

Definitions

1.1

6

Table 1.1-1 (page 1 of 1)

MODES

AVERAGE REACTOR

COOLANT

REACTOR MODE

TEMPERATURE

MODE

TITLE

SWITCH POSITION

(OF)

1

Power Operation

Run

NA

Dfa

nr .trrti n/Hnt

NA

StartupL

Hot Shutdown(a)

Cold Shutdown(a)

Refueling(b)

Standby

Shutdown

Shutdown or Refuel

> 212

S 212

NA

(a)

<-'

(b)

E

All reactor vessel head closure bolts fully tensioned.

One or more reactor vessel head closure bolts less than fully tensioned.

4

Amendment No. 195

HATCH UNIT 1

3

4

5

2

1.1-6

84.

Preparations are presently being made to startup the Unit One reactor. The following conditions

exist:

Reactor Mode Switch:

Shutdown

Reactor Pressure:

125 psig

All reactor vessel head closure bolts are fully tensioned

All rods are IN.

The reactor is in:

a.

Mode 2

b.

Mode 3

c.

Mode 4

d.

Mode 5

ANS:

b

a,c,d incorrect, see table 1.1-1 (pg 1.1-8) in Unit 1 Tech Specs.

NEW

KA#

Generic 2.1.22

OBJ# 400.067.a.05

REF LR-LP-30005

COGNITIVE LVL 3

5.

During a valve lineup, an operator needs to check a valve open. It is noted that the valve has a

locking device on it. To check the valve position the operator should:

a.

unlock the valve, turn it in the closed direction no more than 1/4 turn, place it full

open, and replace the locking device

b.

unlock the valve, turn it in the open direction, verify that the hand wheel moves less than

1/4 turn, and replace the locking device

c.

leave the locking device installed, try to move the hand wheel to ensure locking device

integrity, and verify stem position

d.

leave the locking device installed, verify stem position, and verify administratively that

the valve has not been repositioned.

ANS:

a

b incorrect, check it in the closed direction

c,d incorrect, the locking device needs to be remove to check actual valve position.

NEW

IKA#

Generic 2.1.29

OBJ# 300.022.a.06 i

REF LT-LP-30004

[COGNITIVE LVL 1

1[

49

QUESTIONS REPORT

for HT2002

110. G2.1.4 001

Fuel movement is on progress on Unit 2 with the following plant conditions:

Mode Switch

Coolant

Reactor

Position

Temperature

Power

Unit I

Run

5450F

80%

Unit 2

Refuel

128 0F

0%

Which ONE of the following is the minimum on-site shift staffing required by the Unit 2

Technical Specifications?

(Provide Tech Spec section 5.2.2)

A. SRO

I + I for Fuel Handling

2

2

1

B!3 SRO

1 + 1 for Fuel Handling

2

3

1

C. SRO

2 + 1 for Fuel Handling

2

3

0

D. SRO

2 + 1 for Fuel Handling

3

3

Tech Specs Section 5.2.2.c

0

119

Monday, June 24, LuuL u0:1t 7uu ,0lv

QUESTIONS REPORT

for HT2002

References: Tech Spec Section 5.2.2

-

99 exam Question #6

LT-ST-30003-05, p.7 & 8

10CFR50.54(m)(2)(i)

Modified answer A as follows: PEO from 3 to 2.

A. Incorrect since 3 PEO's are required at all times.

B. Correct answer.

C. Incorrect since Unit 2 does not need an SRO since it is in Mode 5.

D. Incorrect since only 2 RO's are required (one for each unit that has fuel).

RO Tier:

Keyword:

STAFFING

Source:

B

Test:

S

Monday, June 24, 2002 08:17:00 AM

SRO Tier:

T3

Cog Level: MEM 2.3/3.4

Exam:

HT02301

Misc:

TCK

120

QUESTIONS REPORT

for HT2002

18. G2.1.4 001

Fuel movement is on progress on Unit 2 with the following plant conditions:

Mode Switch

Position

Run

Refuel

Coolant

Temperature

545 F

128 F

Which one of the following is the minimum on-site shift staffing required by the Unit 2

Technical Specifications?

A. SRO

B.r SRO

C. SRO

D. SRO

1 + 1 for Fuel Handling

2

1

1 + 1 for Fuel Handling

2

3

1

2 + 1 for Fuel Handling

2

3

1

2 + 1 for Fuel Handling

3

1

References: Tech Spec Section 5.2.2

A. Incorrect since 3 PEO's are required at all times.

B. Correct answer.

C. Incorrect since Unit 2 does not need an SRO since it is in Mode 5.

D. Incorrect since only 2 RO's are required (one for each unit that has fuel).

18

Thursday, April 04, 2002 11:36:29 AM

Unit 1

Unit 2

Reactor

Power

80%

0%

QUESTIONS REPORT

for HT2002

SRO Only

99 exam Question #6

LT-ST-30003-05, p. 8

10CFR50.54(m)(2)(i)

Modified answer A as follows:

Thursday, April 04, 2002 11:36:29 AM

STA from 0 to 1, PEO from 3 to 2.

19

Organization

5.2

5.0

ADMINISTRATIVE CONTROLS

5.2 Organization

5.2.1

Onsite and Offsite Organizations

Onsite and offsite organizations shall be established for unit operation and

corporate management, respectively. The onsite and offsite organizations shall

include the positions for activities affecting safety of the nuclear power plant.

a.

Lines of authority, responsibility, and communication shall be defined and

established throughout highest management levels, intermediate levels,

and all operating organization positions. These relationships shall be

documented and updated, as appropriate, in organization charts,

functional descriptions of departmental responsibilities and relationships,

and job descriptions for key personnel positions, or in equivalent forms of

documentation. These requirements, including plant specific titles of

those personnel fulfilling the responsibilities of the positions delineated in

these Technical Specifications, shall be documented in the Plant Hatch

Unit 2 FSAR;

b.

An assistant plant manager shall be responsible for overall safe operation

of the plant and shall have control over those onsite activities necessary

for safe operation and maintenance of the plant;

c.

The corporate executive responsible for Plant Hatch shall take any

measures needed to ensure acceptable performance of the staff in

operating, maintaining, and providing technical support to the plant to

ensure nuclear safety; and

d.

The individuals who train the operating staff, carry out health physics, or

perform quality assurance functions may report to the appropriate onsite

manager; however, these individuals shall have sufficient organizational

freedom to ensure their independence from operating pressures.

5.2.2

Unit Staff

The unit staff organization shall include the following:

a.

A total of three plant equipment operators (PEOs) for the two units is

required in all conditions. At least one of the required PEOs shall be

assigned to each reactor containing fuel.

(continued)

Amendment No. 135

HATCH UNIT 2

5.0-2

Organization

5.2

5.2 Organization

5.2.2

Unit Staff (continued)

b.

At least one licensed Reactor Operator (RO) shall be present in the

control room for each unit that contains fuel in the reactor. In addition,

while the unit is in MODE 1, 2, or 3, at least one licensed Senior Reactor

Operator (SRO) shall be present in the control room.

c.

The minimum shift crew composition shall be in accordance with

10 CFR 50.54(m)(2)(i). Shift crew composition may be less than the

minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a for a period of

time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence

of on duty shift crew members provided immediate action is taken to

restore the shift crew composition to within the minimum requirements.

d.

An individual qualified to implement radiation protection procedures shall

be on site when fuel is in the reactor. The position may be vacant for not

more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided

immediate action is taken to fill the required position.

e.

Administrative procedures shall be developed and implemented to limit

the working hours of unit staff who perform safety related functions (e.g.,

licensed and non-licensed operations personnel, health physics

technicians, key maintenance personnel, etc.).

Adequate shift coverage shall be maintained without routine heavy use of

overtime. The objective shall be to have operating personnel work a

nominal 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months week while the unit is operating. However, in the event

that unforeseen problems require substantial amounts of overtime to be

used, or during extended periods of shutdown for refueling, major

maintenance, or major plant modification, on a temporary basis the

following guidelines shall be followed:

1.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months

straight, excluding shift turnover time;

2.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months

in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months

period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding

shift turnover time;

3.

A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work

periods, including shift turnover time;

4.

Except during extended shutdown periods, the use of overtime

should be considered on an individual basis and not for the entire

staff on a shift.

(continued)

Amendment No. 135

HATCH UNIT 2

5.0-3

Page 7 of 28

LT-LP-30003-05

C,

I.

Tech Spec Administrative Controls

Cover Tech Spec section 5.1,

Responsibility

A.

Responsibility

1.

The Plant Manager provides direct executive oversight

over all aspects of Plant Hatch.

2.

The assistant plant manager is responsible for overall

unit operation.

3.

The Plant Manager, or his designee, is responsible for

the Radiological Environmental Monitoring Program.

4.

The Superintendent of Shift (SOS) is responsible for

the control room command function. During his

absence an active Senior Reactor Operator (SRO), OR

Licensed Reactor Operator (RO) if both units are in

Mode 4 or 5, shall be designated to assume the control

room command function.

L02

EO la,b

B.

Unit Staff

fEN 91-024 (CO 9100131)

I.

A total of three Plant Equipment Operators (PEOs) for

the two units is required at all times. At least one of

the required PEOs shall be assigned to the reactor

containing fuel.

2.

At least one Licensed RO shall be in the Main Control

Room (MCR) for each reactor containing fuel. Also at

least one SRO shall be present in the MCR while the

unit is in MODE 1, 2, or 3.

3.

Minimum shift crew composition

The MCR shall be manned as a minimum per

1OCFR50.54(m)(2)(i). The chart below outlines the

requirements. Shift crew composition may be less

than the minimum requirements for short periods, not

to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , to accommodate unexpected

absences provided immediate action is taken to

restore shift crew composition.

TECH SPECS / ADMINISTRATIVE CONTROLS

Page 7 of 28

Minimum # Required

Units in Cold'Shutdown

,

2

0

4.

Overtime

Question:

If employee came to work on night

shift at 7:00 p.m. Friday and worked

night shifts for 4 days, then came in

at 7:00 a.m. Wednesday and worked

day shift for 3 days; would employee

violate overtime limits?

a.

Section 5.2.2.e limits the working hours of Unit

staff who perform safety-related functions (e.g.,

SROs, ROs, Plant Equipment Operators, HPs,

and key maintenance personnel, etc.).

b.

In the event that unforeseen problems require

substantial amounts of overtime to be used or

during periods of shutdown for refueling, major

maintenance, or major plant modifications, the

following guidelines shall be followed on a

temporary basis:

1)

An individual should not be permitted to

work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight,

excluding shift turnover time.

2)

An individual should not be permitted to

work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48

hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any

7 day period, all excluding shift turnover

time.

4

Both

Position

QUESTIONS REPORT

for Revision2 HT2002

38. G2.2.27 001

Unit 1 is in Mode 5 with a core shuffle in progress. The bridge operator has just

inserted a fuel bundle into the core when he notices that the adjacent fuel bundle is

mis-oriented.

Which ONE of the following actions are required to be performed by the fuel handling

crew with regards to the fuel shuffle?

A/ The bridge operator stops fuel movements and informs SRO of condition. The SRO

contacts Reactor Engineering to prepare a Fuel Movement Sheet change. The

crew reviews the approved change sheet, corrects the orientation error and

continues with fuel movements with SRO approval.

B. The SRO allows the crew to continue with fuel movements after correcting the

orientation error and notifies Reactor Engineering for documentation on the Core

Loading Verification sheet.

C. The SRO stops fuel movements and the crew determines the proper orientation of

the adjacent fuel bundle. The SRO approves the actions to re-orient the fuel bundle

and the bridge operator notifies the control room when move is complete.

D. The SRO allows fuel movements to continue and contacts Reactor Engineering to

prepare a Fuel Movement Sheet change. At the next appropriate opportunity the

crew will correct the orientation error per the Movement Sheet as long as it is done

on their shift.

References: 42FH-ERP-014-OS Rev. 15.2, pg 10 of 28

34FH-OPS-001-OS Rev. 21.1, pg 6 and 7 of 42

A. Correct answer.

B. Incorrect since ALL fuel movements must be stopped when an error is found.

C. Incorrect since the SRO cannot approve a movement without proper authorization.

D. Incorrect since the Fuel Movement Change sheet must be approved prior to any

further fuel movements.

RO Tier:

SRO Tier:

T3

Keyword:

FUEL MOVEMENTS

Cog Level:

C/A 2.6/3.5

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

o OnlflO'23:2 AM

43

Fida~i:y, Septt*ll~ em

U

e

-r

,¢

I

't'~,"

v

QUESTIONS REPORT

for HT2002

113. G2.2.27 001

Unit 1 is in Mode 5 with a core shuffle in progress. The bridge operator has just

inserted a fuel bundle into the core when he notices that the adjacent fuel bundle is

mis-oriented.

Which ONE of the following actions are required to be performed by the fuel handling

crew with regards to the fuel shuffle?

Af The bridge operator stops fuel movements and informs SRO of condition. The SRO

contacts Reactor Engineering to prepare a Fuel Movement Sheet change. The

crew reviews the approved change sheet and continues with fuel movements with

SRO approval.

B. The SRO allows the crew to continue with fuel movements and notifies Reactor

Engineering for documentation on the Core Loading Verification sheet.

C. The SRO stops fuel movements and the crew determines the proper orientation of

the adjacent fuel bundle. The SRO approves the actions to re-orient the fuel bundle

and the bridge operator notifies the control room when move is complete.

D. The SRO allows fuel movements to continue and contacts Reactor Engineering to

prepare a Fuel Movement Sheet change. At the next appropriate opportunity the

crew will complete the new movement as long as it is done on their shift.

References: 42FH-ERP-014-OS Rev. 15.2, pg 10 of 28

34FH-OPS-001-OS Rev. 21.1, pg 6 and 7 of 42

A. Correct answer.

B. Incorrect since ALL fuel movements must be stopped when an error is found.

C. Incorrect since the SRO cannot approve a movement without proper authorization.

D. Incorrect since the Fuel Movement Change sheet must be approved prior to any

further fuel movements.

RO Tier:

SRO Tier:

T3

Keyword:

FUEL MOVEMENTS

Cog Level: C/A 2.6/3.5

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

0

....

,,O

.o

.7.nn

AhA

123

IVIUIIay, "JUIn

I-' U {

O II.v,*v

QUESTIONS REPORT

for HT2002

1. G2.2.27 001

Unit I is in Mode 5 with a core shuffle in progress. The bridge operator has just

inserted a fuel bundle into the core when he notices that the adjacent fuel bundle is

mis-oriented. SELECT the actions required to be performed by the fuel handling crew

with regards to the fuel shuffle?

A. The bridge operator stops fuel movements and informs SRO of condition. SRO

allows bridge operator to correct orientation after verifying that the orientation was

not correct.

B. The SRO allows the crew to continue with fuel movements and notifies Reactor

Engineering for documentation on the Core Loading Verification sheet.

C. The SRO stops fuel movements and contacts Reactor Engineering to prepare a

Fuel Movement Sheet change. The crew reviews the approved change sheet and

continues with fuel movements with SRO approval.

D. The SRO allows fuel movements to continue and contacts Reactor Engineering to

prepare a Fuel Movement Sheet change. At the next appropriate opportunity the

crew will complete the new movement as long as it is done on their shift.

References: 42FH-ERP-014-0S Rev. 15.2, pg 10 of 28

34FH-OPS-001-0S Rev. 21.1, pg 6 and 7 of 42

A. Incorrect since all moves must be controlled by a Fuel Movement Sheet.

B. Incorrect since ALL fuel movements must be stopped when an error is found.

C. Correct answer.

D. Incorrect since the Fuel Movement Change sheet must be approved prior to any

further fuel movements.

Friday, May 03, 2002 02:07:34 PM

PAGE

10 OF 28

PL

T

E

I-

HATCH..-.---

DOCUMENT TITLE:

FUEL MOVEMENT

7.1.4.1.9

7.1.4.1.10

7.1.4.1.11

7. 1.4 .1.12

DOCUMENT NUMBER:

42FH-ERP-014-OS

R1V5SEDN

15 ED 2

Present the change request, original marked-up move sheet(s),

any additional sheets to another Reactor Engineer/designated

alternate, OR a Shift Supervisor/SRO for review.

Present the reviewed change request, original marked-up move

sheet(s),

AND any additional sheets to the Reactor Engineering

Supervisor, Manager of Engineering Support,

OR the SOS for

approval.

Upon approval, the Reactor Engineering Supervisor,

Manager of Engineering Support,

OR the SOS will sign the change

request and any additional sheets and initial

the changes on the

original move sheet(s).

Make a copy of the completed Change Request AND forward to the

Reactor Engineering Supervisor for review and possible Computer

Database Update.

Attach the completed change request to the original move sheet.

7.1.4.2

IF major changes to the move sheet are needed AND these changes are

such that they cannot be adequately handled by the above rules for

changes,

THEN all fuel movements must stop UNTIL a new move sheet can

be prepared incorporating the new moves.

The Reactor

Engineer/designated alternate and the SRO on the refueling

floor/Shift Supervisor decide IF a new move sheet is required.

IF a

new move sheet is required, make a copy of the approved move sheet

and forward to the Reactor Engineering Supervisor.

4

MGR-0001 Rev.

2

SOUTHERN NUCLEAR

6

NOTE

In the special case of fuel movements that only

involve the Spent Fuel Pool or a Spent Fuel Pool

location, a Reactor Engineer may serve as the

approval authority for the Reactor Engineering

Supervisor.

The Reactor Engineering Supervisor or

superior must approve changes to move sheets being

used to load an MPC-68.

<-V

C

NO:

SOUTHERN NUCLEAR

PAGE

PLANT E. I. HATCH

6 OF 42

DOCUMENT TITLE:

DOCUMENT NUMBER: REVISIONNERSIK

FUEL MOVEMENT OPERATION

34FH-OPS-001-OS

NO:

21.1

5.2.7

Irradiated fuel must NOT be ungrappled in any Fuel Preparation Machine (FPM) UNLESS

the FPM is in the full down position.

5.2.8

Fuel must NOT be ungrappled in the core, fuel storage canister OR in the fuel storage

racks UNLESS proper depth and seating are verified.

5.2.9

IF, during fuel movement, it is found that conditions have changed such that any of the

requirements of this procedure are no longer satisfied, any member of the refueling

bridge team has the authority to halt fuel movement. Prior to halting fuel movement, the

bundle will be loaded in a safe Fuel Pool location or returned to its proper In-Core

location, if possible, UNTIL all requirements are again satisfied.

5.2.10

The Fuel Grapple must be in the full up position prior to moving the bridge OR trolley,

except WHEN:

"* making small adjustments of the bridge OR trolley position to allow alignment for

latching OR discharging a fuel bundle, blade guide,or weight.

"

transporting a blade guide/fuel bundle from one core location to another core location

provided the "BUNDLE CLEAR OF CORE" light is illuminated. WHILE moving the

bridge/trolley, the travel must be slow enough such that the mast does not contact the

trolley.

"* performing refueling interlock checks.

"* core load OR fuel pool verification as long as the bridge is not moved at a high rate of

speed and no load other than the camera and bracket are attached.

"

after discharging a load in the fuel pool/core, THEN raise the grapple several feet OR

as high as necessary to clear any obstruction. WHILE moving the bridge/trolley to a

new location for the purpose of grappling onto another load, the travel must be slow

enough such that the mast does not contact the trolley.

5.2.11

The trolley (operator's cab) must be aligned to allow passage through the transfer canal

prior to moving the bridge forward OR backward WHEN transferring fuel OR blade guides

from the Spent Fuel Pool to the Reactor Cavity, OR vice versa, OR WHEN transferring

fuel OR blade guides from one spent Fuel Pool to another.

MGR-0001 Rev 3

SOUTHERN NUCLEAR _-

PAGE

PLANT E. I. HATCH

7 OF 42

DOCUMENT TITLE:

DOCUMENT NUMBER: REVISIONNERSIOIT

FUEL MOVEMENT OPERATION

34FH-OPS-001-OS

NO:

21.1

5.2.12

The Refueling Bridge Operator will NOT engage in any other activities WHILE moving the

Refueling Bridge OR manipulating any Refueling Bridge controls that will divert full

attention from being devoted to the operation of the Refueling Bridge. Any movement of

the Refueling Bridge OR Hoists must be terminated IF full attention cannot be devoted to

the Refueling activity in progress.

5.2.12.1

The Refueling Bridge Operator will be fully cognizant of all procedural requirements

for fuel movement and must immediately inform the SRO of any problems, whether

with equipment or procedures, which could prevent compliance with these procedural

requirements.

5.2.12.2

The Refueling Bridge Operator must remain aware of the critical nature of his task,

utilize STAR techniques to preclude any errors, and maintain a questioning attitude.

5.2.12.3

The Refueling Bridge Operator is expected to ask the SRO to verify or clarify any

movement which appears unusual.

5.2.13

A Senior Reactor Operator (SRO) shall be on the Refueling Bridge during all core

alterations, and other fuel movements as directed by the Operations Manager. His

responsibilities will be as follows:

5.2.13.1

The SRO will not allow any fuel movement unless the Refuel crew is able to devote

100% attention to the task. The crew, as a minimum, will consist of an SRO, a bridge

operator, and a second verifier. The SRO must ensure that each member is fully

qualified and capable of performing their task, and that each is aware of their

responsibilities.

5.2.13.2

The SRO must ensure compliance with all applicable procedures at all times, and

must ensure that procedural & equipment problems are properly documented.

Procedure problems, or conditions which preclude compliance, must be corrected

before proceeding.

5.2.13.3

The SRO will not allow any crew member to conduct a turnover or be relieved when

the bridge or grapple is moving, or when the grapple is loaded.

5.2.13.4

The SRO must ensure that the control room is aware of conditions on the refueling

floor. Constant communications will be maintained with a licensed individual in the

control room when core alterations are in progress.

5.2.13.5

The SRO will ensure that the crew has all the material necessary (i.e., procedures,

core maps, movement sheets, etc.) to conduct fuel movement promptly and correctly.

One required item will be a core map with the cell removal sequence for upcoming

cells already marked and verified (not required for fuel shuffle). Laptop computer

displays of the core may be used instead of paper copies of core maps.

5.2.13.6

The SRO will ensure that the work pace is comfortable to all members, and that no

one feels any urgency in completing the task.

MGR-0001 Rev 3

-- V

QUESTIONS REPORT

for HT2002

115. G2.2.32 001

Unit 2 is in a refueling outage with a fuel shuffle in progress. A fuel bundle is being

transfered from the core to the fuel pool when the Control Board Operator reports that

reactor cavity water level is decreasing.

Per 34AB-G41-002-2S, Decreasing Rx Well/Fuel Pool Water Level, which ONE of the

following actions should the refueling SRO direct the bridge operator to perform?

A. Return the fuel bundle to the closest in-core location as possible.

B. Stop all movement and evacuate the refueling floor immediately.

C. Continue with movement to the fuel pool and lower it as deep into the pool as

possible.

Df Return the fuel bundle to its proper in-core location.

References: 34AB-G41-002-2S Rev. 2 pg 2 of 5.

1999 Hatch Exam Question 21

Modified answers slightly to make different answer correct.

A. Incorrect since the direction should be to move the fuel bundle to its proper in-core

location.

B. Incorrect since direction should be to lower the bundle and you only have to

evacuate the refueling floor if there are radiation alarms.

C. Incorrect since direction should be to place the fuel bundle in any fuel pool rack.

D. Correct answer.

RO Tier:

SRO Tier:

T3

Keyword:

FUEL POOL

Cog Level:

MEM 2.3/3.3

Source:

B

Exam:

HT02301

Test:

S

Misc:

TCK

125

Monday, June 24, 2002 08:1 t:uu AM

QUESTIONS REPORT

for HT2002

1. G2.2.32 001

Unit 2 is in a refueling outage with a fuel shuffle in progress. A fuel bundle is being

transfered from the core to the fuel pool when the Control Board Operator reports that

reactor cavity water level is decreasing. Per 34AB-G41-002-2S, Decreasing Rx

Well/Fuel Pool Water Level, which one of the following actions should the refueling

SRO direct the bridge operator to perform?

A. Return the fuel bundle to the closest in-core location as possible.

B. Stop all movement and evacuate the refueling floor immediately.

C. Continue with movement to the fuel pool and lower it as deep into the pool as

possible.

D. Return the fuel bundle to its proper in-core location.

References: 34AB-G41-002-2S Rev. 2 pg 2 of 5.

1999 Hatch Exam Question 21

Ro

Aro'o:rli

1

_

I2a

tiferari -.

a :..

A. Incorrect since the direction should be to move the fuel bundle to its properin-core ..

S...

l-

ocation. -

.- ncorrect-since-dtirectioflhOU-oýowerh-bInllC

ad you oly ....

t.....

evacuate the refueling floor if there are radiation alarms.

C. Incorrect since direction should be to place the fuel bundle in any fuel pool rack.

D. Correct answer.

Friday, May 03, 2002 06:58:19 AM

SOUTHERN NUCLEAR

DI AKlT M

I WATW.

S......PAGE 2OF 5

DOCUMENT TITLE:

DOCUMENT NUMBER:

REVISION NO:

DECREASING RX WELL/FUEL POOL WATER LEVEL

34AB-G41-002-2S

j_-_----2

4.0

SUBSEQUENT OPERATOR ACTIONS

N>

4.1

Dispatch personnel to investigate the alarm.

4.2

IF a refueling floor radiation monitor alarms due to decreasing reactor well OR fuel pool water

level:

4.2.1

Evacuate the refueling floor immediately.

4.2.2

Have Health Physics assess the Radiological conditions on the Refueling Floor AND

establish a manned access control point to the Refueling Floor. IF fuel OR highly

irradiated components are actuallyý-ncovered - exposure rates may b&

evere

additional entries may NOT- be possible.

......

, ....

-- --

4.3

F

Wfuelmvment-is-in-progressr-pIth4U

Jt-)bundI

the following:

4.3.1

Return fuel bundle to its proper incore location, OR

-.

i-ti~f~flfl~

Tift

.4 MTN7MF1'W thdIMP151, Y_7

4.3.3

4.4

4.4.1

4.4.2

4.5

Lower fuel bundle as deep into the vessel as possible.

IF movement of other highly radioactive materials (irradiated control rods, fuel channels,

LPRM's, etc.) is in progress, PLACE the item in a safe condition by performing one of the

following:

Lower item as deep into vessel as possible

Lower item as deep into fuel pool, cask storage area, OR transfer canal, as possible

Contact Health Physics to provide continuous coverage WHILE personnel are on the refueling

floor.

E

MGR-0001 Rev 3

CAUTION

CONTROL ROD BLADES AND FUEL IN THE FUEL PREP MACHINE MAY BECOME UNCOVERED

ALONG WITH IRRADIATED MATERIAtýSýrýTH

IRFTHE PRO-tESSflFB

.

....

EXTREME RADIATION LEVELS COULD RESULT IF THESE ITEMS ARE EXPOSED ABOVE

WATER.

'riday, October 01, 1999 @ 06:18 PM

HATCH99.BNK

Page: 21

21. Unit 2 is in a refueling outage with a full core off load in progress. A fuel bundle is

being transfered from the core to the fuel pool when the control room operator reports

that reactor cavity water level is decreasing. Per 34AB-G41-002-2S, "Decreasing Rx

Well/Fuel Pool Water Level," which one of the following actions should the refueling

SRO direct the bridge operator to perform?

a. Return the fuel bundle to any in-core location that is available.

,1b Move the fuel bundle to any fue storage rack in the fuelpool.

c. Move the fuel bundle to the fuel pool and lower it as deep into the pool as possible.

d. Do not move the fuel bundle any further and lower it as deep as possible where it is.

Bank question (modified slightly)

34AB-G41-002-2S

LT-LP-04502-03, p. 36

KEY WORDS:

System

K/A No.

KA/Value

Difficulty

SampiePian

VendorL-

LiceAsee

Last used

GENERICS

.2.32

(2.3/33)

_1

IER3CAT2

1BWR-4

HATCH

BANK

QUESTIONS REPORT

for HT2002

116. G2.2.6 001

While reviewing a procedure that is required to be completed before the end of the

current shift, the SS notices a step that requires the use of a gauge which is broken.

Another gauge is available in the system and the SS has confirmed it will operationally

function as a substitute.

At a minimum, which ONE of the following actions must be done to perform the

procedure? The SS should:

A. Make a pen and ink change to the procedure.

B. Make a SRO change to the procedure.

C. Make a pen and ink change to the procedure with SOS concurrence.

D. Make a permanent change to the procedure obtaining manager approval prior to

use.

Reference: LT-LP-30004-04, Pg. 15-17

99 exam question #19

EO 300.002.a.02

A. Incorrect since this process is used for editorial changes. This change is not

editorial.

B. Correct answer.

C. Incorrect since this process is used for editorial changes. This change is not

editorial.

D. Incorrect since this is the normal process that is used if the procedure is not needed

now. Since this procedure change is needed prior to the end of the shift then an SRO

change is appropriate.

RO Tier:

Keyword:

Source:

Test:

PROCEDURE CHANGE

B

S

SRO Tier:

Cog Level:

Exam:

Misc:

T3

MEM 2.3/3.3

HT02301

TCK

126

Monday, June 24, 2002 08:17:01 AM

QUESTIONS REPORT

for HT2002

22. G2.2.6 001

While reviewing a procedure that is required to be completed before the end of the

current shift, the SS notices a step requireing the use of a gauge which is broken.

Another gauge is available in the system and the SS has confirmed it will operationally

function as a substitute. At a minimum, which one of the following actions must be

done to perform the procedure? The SS should:

A. Make a pen and ink change to the procedure.

B.r Make a SRO change to the procedure.

C. Make a pen and ink change to the procedure with SOS concurrence.

D. Make a permanent change to the procedure obtaining manager approval prior to

use.

Reference: LT-LP-30004-04, Pg. 15-17

99 exam question #19

A. Incorrect since this process is used for editorial changes. This change is not

editorial.

B. Correct answer.

C. Incorrect since this process is used for editorial changes. This change is not

editorial.

D. Incorrect since this is the normal process that is used if the procedure is not needed

now. Since this procedure change is needed prior to the end of the shift then an SRO

change is appropriate.

Thursday, April 04, 2002 11:36:30 AM

23

5.

Controlling equipment status, with special emphasis

on Technical Specifications Limiting Conditions for

Operation and for worker protection.

6.

Controlling the position/condition of all plant

components and systems except as allowed by other

approved procedures.

7.

Ensuring operations personnel are trained and

qualified.

8.

Ensuring that operations activities are governed by

effective administration controls.

9.

Being responsible for the operation of the Radwaste

facility.

E04

B.

The order to startup or to shutdown the reactor for planned

maintenance or refueling is issued by the General Manager

Nuclear Plant or his designated alternate.

E Q

C.

In an emergency or when it is judged that continued

operation would jeopardize plant or personnel safety, any

member of the plant staff holding an operator license has the

authority to order the reactor shutdown or to shut it down

himself.

Review 10AC-MGR-003-0S

IV.

1OAC-MGR-003-OS, "Preparation and Control of

Procedures"

Procedure is now a flowchart.

A.

This procedure shows how to process:

Review Attachment 3 of LP

30004.

Review lOAC-MGR-003-OS,

1.

Changes to existing procedures,

2.

New procedures,

3.

Special purpose procedures and,

4.

Vendor procedures controlled under the Plant Hatch

Quality Assurance Program

(10AC-MGR-003, Attachment 1)

EO 6

Question: May a SRO

change be made to a

procedure that addresses core

alteration?

Ans: No (IOAC-MGR-003,

Attachment 2)

B.

Editorial Changes must fit one of the examples listed in

Attachment 1. The Editorial Change process is restricted to

those examples and no other changes can be made using this

process.

C.

SRO Changes may be made provided the change is not

Editorial and the intent of the original procedure is not

altered. Intent of a procedure is what the procedure does

and how it does it.

1.

General Examples of Changes of Intent:

a.

Change in the method of performing a step or

the sequence of steps in such a way that it would

affect the results.

b.

Achieving the same result with different steps or

a different sequence of steps which have not

been previously evaluated.

2.

Specific Examples of Changes of Intent:

4

a.

Change in sequence of performing Core

Alterations.

b.

Changes to Limits/Setpoints/Acceptance Criteria

NOT previously evaluated (by 10 CFR 50.59

Evaluation).

c.

Changes that reduce control or design features

for ALARA.

d.

Changes to initial conditions.

e.

Deleting or reducing verifications or

requirements.

f.

Deleting or relocating Hold Points.

g.

Changes to authority or responsibility for review

or approval.

h.

Changes to SSC alignment not previously

evaluated.

3.

SRO Changes to procedures must be approved by two

members of management:

EO 7

a.

Any Supervisor familiar with the work and

knowledgeable of the procedure change process.

b.

Any licensed SRO whose license is active or

inactive. This does not have to be an individual

actually on shift. The role of the SRO here is to

ensure that there is no adverse impact on plant

operation.

4.

SRO Changes must be reviewed by the PRB or

Qualified Reviewer (QR) and approved by the

applicable manager within 14 days of becoming

effective.

V.

10AC-MGR-004-OS, "Deficiency Control System"

A.

Correction of Deficiencies

The normal methods for affecting change to improve

reliability or to correct defects that reduce reliability are the

various administrative controls that have been established to

identify and either correct or improve these conditions. The

administrative processes for correcting defects or causing

improvements are identified as follows:

1.

A Maintenance Work Order (MWO) is used to control

the correction of defects in plant equipment.

2.

A Request for Engineering Review (RER) is used to

initiate and control modifications to plant equipment.

3.

A Procedure Processing Form, or an Instruction

Request/ Development Form is used to control the

development of or revision to Plant Hatch Procedures

and Instructions.

Page 3 of 64

LT-LP-30004-04

ADMINISTRATIVE PROCEDURES

300.041.C

MAINTAIN key control per the guidelines outlined in 80AC-SEC-002-OS, "Key and

Annunciator Door Control."

400.059.A

Given a plant transient and/or accident has occurred, collect and analyze the required data to,

DETERMINE the root cause per 10AC-MGR-012-OS "Plant Event Analysis and Resolution

Program."

ENABLING OBJECTIVES

1.

Given procedure OOAC-REG-001-OS, IDENTIFY the events that require a report to be made to the

NRC within one hour. (SRO ONLY) (300.004.b.01)

2.

Given procedure OOAC-REG-00l-OS, IDENTIFY the reporting requirements to the NRC on the

actuation and injection of an Emergency Core Cooling System. (SRO ONLY) (300.004.b.03)

3.

Given procedure OOAC-REG-001 -OS, IDENTIFY those events that require a report to be made to the

NRC within four hours. (SRO ONLY) (300.004.b.02)

4.

LIST the personnel having the authority to order a startup or shutdown of the Reactor for planned

maintenance or refueling per 10AC-MGR-001-0S. (300.032.a.02)

5.

STATE who has the authority to shutdown the Reactor in an emergency condition per 1OAC-MGR

001-OS. (300.032.a.03)

6.

Given the applicable procedure, IDENTIFY the conditions in which a SRO change to a procedure is

allowed to be made. (SRO ONLY) (300.002.a.02)

7.

Given the applicable administrative procedure, IDENTIFY the approval requirements for an SRO

change to a procedure. (SRO ONLY) (300.002.a.03)

8.

Given a scenario, STATE whether the conditions given require the initiation of a Deficiency Card.

(300.023.a.01)

9.

Given a list of statements, SELECT the statement which best describes the purpose of analyzing

unusual plant events per 10AC-MGR-012-0S, "Plant Event Analysis and Resolution Program."

(400.059.a.01)

10.

Given a list of evolutions, SELECT the evolution that satisfies the procedure requirements for

independent verification. (300.016.a.02)

(2

11.

Given plant\\system status and a copy of 30AC-OPS-001-0S, DETERMINE if an equipment

clearance is being processed correctly. (300.016.a.01)

QUESTIONS REPORT

for Revision2 HT2002

39. G2.3.3 001

Unit 1 is at 75% RTP. At 1400 on 8/12/02, after performing scram time testing, the

Control Board Operator notes that the Offgas Flow has increased from the steady state

level as follows:

Offgas Inlet Flow to Stack prior to scram time testing

100 scfm

Offgas Inlet Flow to Stack after scram time testing

175 scfm

Which ONE of the following actions is/are required by Tech Specs for this condition?

(Provide copy of TS Section 3.7.6 along with SR's)

A. Notify Chemistry to sample the offgas system by 1900 to verify gross gamma

activity is < 240 mCi/second. If greater than 240 mCi/second then isolate SJAE

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OR BE in Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B! Notify Chemistry to sample the offgas system by 1800 to verify gross gamma

activity is < 240 mCi/second. If greater than 240 mCi/second then enter 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

LCO to restore within limits.

C. Notify Chemistry to sample the offgas system by 0200 on 8/13/02 to verify gross

gamma activity is < 240 mCi/second. If greater than 240 mCi/second then enter 72

hour LCO to restore within limits.

ID. Notify Chemistry to sample the offgas system by 0400 on 8/13/02 to verify gross

gamma activity is < 240 mCi/second. If greater than 240 mCi/second then enter 24

hour LCO to restore within limits.

References: Tech Spec section 3.7.6 (SR 3.7.6.1)

LT-LP-03101 Rev. 3 pg 29 of 44

A. Incorrect since the sample time does not allow for 25% grace period and if exceed

the LCO limit then have 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore.

B. Correct answer.

C. Incorrect since the sample time is based on 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months instead of the required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

D. Incorrect since the sample time is based on 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months plus 25% instead of the

required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Also, if exceed the LCO limit then have 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore.

RO Tier:

SRO Tier:

T3

Keyword:

OFF-GAS

Cog Level:

C/A 1.8/2.9

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

-

k

A

ýý

ý023:2

AM44

Friday,

e~ptllU

I

m

U

e

-r U-

I

/O*-

~¥

QUESTIONS REPORT

for HT2002

119. G2.3.3 001

Unit 1 is at 75% RTP. At 1400 on 8/12/02, after performing scram time testing, the

Control Board Operator notes that the Offgas Flow has increased from the steady state

level as follows:

Offgas Inlet Flow to Stack prior to scram time testing

100 scfm

Offgas Inlet Flow to Stack after scram time testing

175 scfm

Which ONE of the following actions is/are required by Tech Specs for this condition?

A. Notify Chemistry to sample the offgas system by 1800 at the latest to verify gross

gamma activity is < 240 mCi/second. If greater than 240 mCi/second then enter 24

hour LCO to restore within limits.

B! Notify Chemistry to sample the offgas system by 1900 at the latest to verify gross

gamma activity is < 240 mCi/second. If greater than 240 mCi/second then enter 72

hour LCO to restore within limits.

C. Notify Chemistry to sample the offgas system by 0200 on 8/13/02 at the latest to

verify gross gamma activity is < 240 mCi/second. If greater than 240 mCi/second

then enter 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months LCO to restore within limits.

D . Notify Chemistry to sample the offgas system by 0400 on 8/13/02 at the latest to

verify gross gamma activity is < 240 mCi/second. If greater than 240 mCi/second

then enter 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months LCO to restore within limits.

References: Tech Spec section 3.7.6 (SR 3.7.6.1)

LT-LP-03101 Rev. 3 pg 29 of 44

A. Incorrect since the sample time does not allow for 25% grace period and if exceed

the LCO limit then have 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore.

B. Correct answer.

C. Incorrect since the sample time is based on 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months instead of the required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

D. Incorrect since the sample time is based on 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months plus 25% instead of the

required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Also, if exceed the LCO limit then have 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore.

RO Tier:

SRO Tier:

T3

Keyword:

OFF-GAS

Cog Level: C/A 1.8/2.9

Source:

N

Exam:

HT02301

>

Test:

S

Misc:

TCK

,

n.

A

Ann

O

A.O*7t

AnA

129

IVloually, Julle

Z,r'pfU

,r..It.JI/*

QUESTIONS REPORT

for HT2002

1. G2.3.3 001

/

Unit 1 is at 75% RTP. At 1400 on 8/12/02, after performing scram time testing, the

Control Board Operator notes that the Offgas Flow has increased from the steady state

level as follows:

Offgas Inlet Flow to Stack prior to scram time testing

100 scfm

Offgas Inlet Flow to Stack after scram time testing

175 scfm

SELECT the action(s) that is/are required by Tech Specs for this condition.

A. Notify Chemistry to sample the offgas system by 1800 at the latest to verify gross

gamma activity is < 240 mCi/second. If greater than 240 mCi/second then enter 24

hour LCO to restore within limits.

B/ Notify Chemistry to sample the offgas system by 1900 at the latest to verify gross

gamma activity is < 240 mCi/second. If greater than 240 mCi/second then enter 72

hour LCO to restore within limits.

C. Notify Chemistry to sample the offgas system by 0200 on 8/13/02 at the latest to

verify gross gamma activity is < 240 mCi/second. If greater than 240 mCi/second

then enter 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months LCO to restore within limits.

D. Notify Chemistry to sample the offgas system by 0400 on 8/13/02 at the latest to

verify gross gamma activity is < 240 mCi/second. If greater than 240 mCi/second

then enter 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months LCO to restore within limits.

References: Tech Spec section 3.7.6 (SR 3.7.6.1)

LT-LP-03101 Rev. 3 pg 29 of 44

A. Incorrect since the sample time does not allow for 25% grace period and if exceed

the LCO limit then have 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore.

B. Correct answer.

C. Incorrect since the sample time is based on 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months instead of the required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

D. Incorrect since the sample time is based on 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months plus 25% instead of the

required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Also, if exceed the LCO limit then have 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore.

Friday, May 03, 2002 03:16:34 PM

Main Condenser Offgas

3.7.6

SURVEILLANCE REQUIREMENTS

SURVEILLANCE

NOTE -----------------------------

Not required to be performed until 31 days after

any main steam line not isolated and SJAE in

operation.

Verify the gross gamma activity rate of the noble

gases is s 240 mCi/second.

FREQUENCY

31 days

AND

Once within

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after a

t 50% increase in

the nominal steady

state fission gas

release after

factoring out

increases due to

changes in

THERMAL

POWER level

Amendment No. 195

HATCH UNIT 1

SR 3.7.6.1

3.7-17

Main Condenser Offgas

3.7.6

3.7 PLANT SYSTEMS

3.7.6

Main Condenser Offgas

LCO 3.7.6

APPLICABILITY:

The gross gamma activity rate of the noble gases measured at the main

condenser evacuation system pretreatment monitor station shall be

< 240 mCi/second.

MODE 1,

MODES 2 and 3 with any main steam line not isolated and steam jet air

ejector (SJAE) in operation.

ACTIONS

CONDITION

REQUIRED ACTION

COMPLETION TIME

A.

Gross gamma activity rate

A.1

Restore gross gamma

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

of the noble gases not

activity rate of the noble

within limit,

gases to within limit.

B.

Required Action and

B.1

Isolate all main steam

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

associated Completion

lines.

Time not met.

B.2

Isolate SJAE.

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

B.3.1

Be in MODE 3.

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

AND

B.3.2

Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

C

Amendment No. 195

C

3.7-16

HATCH UNIT 1

Page 29 of 44

LT-LP-03101-03

OFF GAS SYSTEM

,'

I

LT-03101 Table 1

N62-P600 METER AND RECORDER INDICATIONS

METERS

Glycol Pump Disch

O/G To Preheater

Recomb Inlet A

Recomb Inlet B

BOO 1A Temp (Preheater Inlet)

BOO1 B Temp (Preheater Inlet)

Prefilter Diff Press

Adsorber Train Inlet/Outlet Press

After Fltr Diff Press

0-50 psi

0-15 psi

0-600°F

O-4000F

0-4000 F

0-12"H20

0-3.5 psid

0-12" H20

RECORDERS

Inlet Temp (Reheater)

Outlet Dewpoint

Storage Tk Temp (Glycol)

Inlet Flow To Stack RED

BLK

- High

- Low

Adsorber Vault Temp

H2 Analyzers

RED - A

GRN - B(BLK on Ul)

Recombiner Temperatures (Multipoint)

Adsorber Vessel Temperatures (Multipoint)

O-lOO0F

0-1OO°F

0-100°F

3-256 SCFM

3-24 SCFM

O-100°F

0-5%

O-1000°F

O-150°F

I 1N .

I

R605

R600

R601A

R601B

R607A

R607B

R611

R612

R616

4

R608

R609

R606

R604

R615

R603

R602

R613

4

QUESTIONS REPORT

for HT2002

120. G2.3.4 001

The Emergency Director decides that is is necessary to send someone into the

Reactor Building (with Health Physics) to isolate a leak before the Core Spray and

RHR pumps are flooded. (No releases are underway and RPV level is being

maintained at 60 inches with the Condensate System)

Which ONE of the following is the maximum allowable dose limit that the Emergency

Director may authorize?

A. 5 REM

Bf 10 REM

C. 25 REM

D. > 25 REM

References: 73EP-EIP-017-OS Rev 2.1 pg 6 of 13.

SRO exam 95-01 question # 94.

B. Correct answer.

K'-

A,C and

RO Tier:

Keyword:

Source:

Test:

D. Incorrect. See reference above on page

SRO Tier:

DOSE RATE LIMITS

Cog Level:

B

Exam:

S

Mise:

6.

T3

MEM 2.5/3.1

HT02301

TCK

Monday, June 24, 2002 08:17:01 AM

130

QUESTIONS REPORT

for HT2002

2. 02.3.4 001

The Emergency Director decides that is is necessary to send someone into the

Reactor Building (with Health Physics) to isolate a leak before Core Spray and RHR

pumps are flooded. (No releases are underway and RPV level is being maintained at

60 inches with the Condensate System)

SELECT the

authorize:

maximum allowable dose limit that the Emergency Director may

A. 5 REM

Bf 10 REM

C. 25 REM

D. > 25 REM

References: 73EP-EIP-017-0S Rev 2.1 pg 6 of 13.

SRO exam 95-01 question # 94.

B. Correct answer.

A,C and D. Incorrect. See reference above on page 6.

Thursday, May 02, 2002 04:51:44 PM

2

GEORGIA POWER COMPANY

PLANT E.I.

HATCH

DOCUMENT TITLE:

EMERGENCY EXPOSURE CONTROL

DOCUMENT NUMBER:

73EP-EIP-017-0S

PAGE

6

OF 13

REVISIUN

P1(9:

2 ED 1

7.4

EMERGENCY EXPOSURE GUIDELINES

7.4.1

The Emergency Director will establish the exposure limits for the

emergency response personnel based on the following Emergency Response

Personnel Exposure Guides:

NOTE

These guidelines do not establish a rigid upper

limit of exposure.

The Emergency Director may use

his/her judgment in establishing the appropriate

limit.

NOTE

No thyroid limit is

specified for lifesaving action

since the complete loss of the thyroid may be

considered an acceptable risk

for saving a life;

however, thyroid exposure must be minimized through

the use of respiratory protection and/or KI

tablets.

E

EMERGENCY RESPONSE PERSONNEL EXPOSURE GUIDES

This limit is expressed as the sum of the effective dose equivalent

(EDE)

committed effective dose equivalent (CEDE)

and the

4

MGR-0001 Rev.

1

6

Dose Limit*

Activity

Condition

(REM)

5

all

n/a

protecting valuable

10

25

life saving or protection

lower dose not practicable

of large populations

>25

life saving or protection

only on a voluntary basis to

of large populations

persons fully aware of the

risks involved

i

.

...

i

lower dose not practicable

94.

KA#

Generic 2.3.4

OBJ# LT-30008.002

REF LT-LP-30008

COGNITIVE LVL 1

95.

An Alert Emergency has been declared and the OSC has been manned. A fire in the Service

Building breakroom kitchen requires that the OSC be evacuated due to excessive smoke. When

the evacuation is ordered, the OSC workers should go to the:

a.

East Wing of the Simulator Building

b.

Classroom 172 in the Simulator Building

C.

Simulator Building Cafeteria

d.

Technical Support Center conference room.

ANS:

c

a incorrect, normal for EOF

b incorrect, OSC supervision goes here

d incorrect, TSC is not an alternate

NEW

KA# Generic 2.4.42

OBJ# 200.052.h.01

REF EP-LP-30200

COGNITIVE LVL 1

54

The Emergency Director decides that it is necessary to send someone into the Reactor Building

(with Health Physics) to isolate a leak before Core Spray and RHR pumps are flooded. (No

releases are underway and RPV level is being maintained at 60 inches with the Condensate

System)

SELECT the maximum allowable dose limit that the Emergency Director may authorize:

a.

5 REM

b.

10 REM

C.

25 REM

d.

> 25 REM

ANS:

b

a,c,d incorrect, see 73EP-EIP-017-OS pg 6 of 13

NEW

QUESTIONS REPORT

for Revision2 HT2002

40. G2.4.48 001

Unit 1 is in an ATWS condition with Reactor Power oscillating between 15 and 45%

RTP. The following indications exist at this time:

SBLC Pump Select Switch in Start Sys A position.

SBLC Squib Vlv Ready lights are LIT.

Rx Water Cleanup VIv, 2G31-F004,Rx Wtr Cleanup Suction Vlv, is CLOSED.

SBLC Discharge Pressure is greater than reactor pressure.

Which ONE of the following describes the appropriate actions the Shift Supervisor

should order?

A. Inhibit ADS and bypass RWCU filter/demineralizers per 34SO-G31-003-1S.

B. Continue to monitor SBLC and secure when the Cold Shutdown Boron Weight has

been added.

C. Inhibit ADS, continue to monitor SBLC, and exit RC/Q when the reactor is

subcritical.

D' Initiate SBLC per 34SO-C41-003-1S using the manual-local initiation method.

References: RCA RPV CONTROL (ATWS) Rev. 6

LR-20328 Rev. 6 pg 44-45 of 58.

A. Incorrect since 2G31-F004 is closed.

B. Incorrect since boron is not injecting.

C. Incorrect since boron is not being injected and RC/A exit requires subcritical with no

boron injection.

D. Correct answer.

RO Tier:

SRO Tier:

T3

Keyword:

EOP PC CONTROL

Cog Level: C/A 3.5/3.8

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

Friday, September 20, 2002 09:23:24 AM

45

QUESTIONS REPORT

for HT2002

125. G2.4.48 001

Unit I is in an ATWS condition with Reactor Power oscillating between 15 and 35%

RTP. The following indications exist at this time:

SBLC Pump Select Switch in Start Sys A position.

SBLC Squib VIv Ready lights are LIT.

SBLC LOSS OF CONTINUITY TO SQUIB VALVE is ALARMED.

Rx Water Cleanup VIv, 2G31-F004,Rx Wtr Cleanup Suction VIv, is CLOSED.

SBLC Discharge Pressure is greater than reactor pressure.

Which ONE of the following describes the appropriate actions the Shift Supervisor

should order?

A. Inhibit ADS and bypass RWCU filter/demineralizers per 34SO-G31-003-1S.

B. Continue to monitor SBLC and secure when the Cold Shutdown Boron Weight has

been added.

C. Inhibit ADS and inject boron using HPCI, RCIC or CRD to shutdown the reactor.

Df Reset ARI and continue to insert control rods per 31 EO-EOP-103-1S to shutdown

the reactor.

References: RCA RPV CONTROL (ATWS) Rev. 6

LR-20328 Rev. 6 pg 44-45 of 58.

A. Incorrect since reactor power oscillations are less than 25% and these actions are

not yet directed by the ATWS procedure. Also, you don't isolate the RWCU demins

unless there is a failure to isolate the system.

B. Incorrect since the squib valves did not fire and boron is not going into the reactor.

C. Incorrect since reactor power oscillations are less than 25% and these actions are

not yet directed by the ATWS procedure.

D. Correct answer.

RO Tier:

SRO Tier:

T3

Keyword:

EOP PC CONTROL

Cog Level:

C/A 3.5/3.8

Source:

N

Exam:

HT02301

Test:

S

Misc:

TCK

.

...

...........

.

135

Monday, June 24, 2uu0 08: I1

A

Mvi

QUESTIONS REPORT

for HT2002

1. G2.4.48 001

Unit 1 is in an ATWS condition with Reactor Power oscillating between 15 and 35%

RTP. The following indications exist at this time:

SBLC Pump Select Switch in Start Sys A position.

SBLC Squib VIv Ready lights are LIT.

SBLC LOSS OF CONTINUITY TO SQUIB VALVE is ALARMED.

Rx Water Cleanup VIv, 2G31-F004,Rx Wtr Cleanup Suction VIv, is CLOSED.

SBLC Discharge Pressure is greater than reactor pressure.

Which ONE of the following describes the appropriate actions the Shift Supervisor

should order?

A. Inhibit ADS and bypass RWCU filter/demineralizers per 34SO-G31-003-1 S.

B. Continue to monitor SBLC and secure when the Cold Shutdown Boron Weight has

been added.

C. Inhibit ADS and inject boron using HPCI, RCIC or CRD to shutdown the reactor.

D. Reset ARI and continue to insert control rods per 31 EO-EOP-103-1S to shutdown

the reactor.

References: RCA RPV CONTROL (ATWS) Rev. 6

LR-20328 Rev. 6 pg 44-45 of 58.

A. Incorrect since reactor power oscillations are less than 25% and these actions are

not yet directed by the ATWS procedure. Also, you don't isolate the RWCU demins

unless there is a failure to isolate the system.

B. Incorrect since the squib valves did not fire and boron is not going into the reactor.

C. Incorrect since reactor power oscillations are less than 25% and these actions are

not yet directed by the ATWS procedure.

D. Correct answer.

Friday, May 31, 2002 03:35:56 PM

I

Page 45 of 58

LR-LP-20328-06

RPV CONTROL - ATWS (RCA)

Define Large

Threshold.

Oscillation

if > 25% (above LOT) Boron

injection is required.

Boron injected ONLY if

oscillation persist.

P if cannot maintain below HCTL,

depressurization is required.

5. Higher clad temperatures increase the surface heat flux,

generating steam and a pressure increase in the channel.

6. Moderator is discharged from both the top and bottom of the

bundle with void generation rapidly decreasing power.

7. Inlet flow is restored by the lower plenum pressure/flow

boundary conditions, and the process begins again.

To provide reasonable assurance that any rapidly growing oscillations

are mitigated in a timely manner, boron is injected when neutron flux

oscillations in excess of the Large Oscillation Threshold (LOT)

commence and continue. The LOT is a peak-to-peak neutron flux

oscillation amplitude equal to or less than 25% yet sufficiently large to

be distinguishable from the flux perturbations expected of a stable

thermal-hydraulic system. Flux oscillations at or below the LOT during

a failure-to-scam event are not expected to threaten fuel clad integrity.

Initiation of boron iniection is required for oscillations in excess of the

LOT only if they "commence and continue." This wording clarifies

that boron need not be injected in response to a single flux pulse which

subsequently subsides.

For conditions susceptible to oscillations, the oscillation growth is

directly related to core inlet subcooling. Since the length of time

required to raise in-core boron concentration is longer than the time

required to reduce core inlet subcooling, boron injection alone may not

prevent large irregular neutron flux oscillations from occurring.

However, the magnitude of the oscillations is reduced as the

concentration of boron in the core increases.

If Torus temperature and RPV pressure cannot be maintained below the

Heat Capacity Temperature Limit, rapid depressurization of the RPV

will be required.

BEFORE

Torus water temperature reaches

BUT curve limit

(graph 5)

I-

Efforts to insert rods and inject

boron occur simultaneously.

The initiation and growth of these

oscillations is principally

dependent upon the subcooling at

the core inlet.

Concurrent execution of boron initiation and control rod insertion is

needed to optimize efforts to achieve reactor shutdown.

The symptomatic approach to emergency response precludes

assignment of priorities to these actions since the time at which boron

must be injected into the RPV is dependent on the magnitude of the

failure to scram event.

Instabilities are manifested by oscillations in reactor power which, if

the reactor cannot be shutdown, may increase in magnitude. If the

oscillations remain small or moderately sized, they tend to repeat on

approximately a two second period. Under certain circumstances,

however, the oscillations may continue to grow and become

sufficiently large and irregular to cause localized fuel damage. The

initiation and growth of these oscillations is principally dependent upon

the subcooling at the core inlet, the greater the subcooling, the more

likely that oscillations will commence and increase in magnitude.

Although unlikely, it is possible for such oscillations to develop before

corrective actions can be taken. The process by which large irregular

neutron flux oscillations can develop within a fuel bundle assembly

occurs as follows:

1. Subcooled water enters the fuel bundle.

2. The resulting positive reactivity addition causes a rapid increase

in bundle power.

3. The increased energy deposition in the fuel increases the fuel

and clad temperature.

4. Doppler (fuel temperature) feedback terminates the power

increase.

WAIT UNTIL

Reactor power oscillations

exceed 25% peak to peak

IG THE FOLLOWING

THEN tominate boron injection

perform RPV Control (Non-ATWS)

GO TOR R

point A

OW

N*

LY

0@

ctions:

,er34AB-C71-001IS

V DIESEL GENERATO

-CP

TE_

ad =-

'.-g

/

Goto 34AB-C71-001-1S

I-

r

I

WHILE PERFORMING THE FOLLOWING

.LE reactor is shutdown (subcrifical with

THEN perform scram procedure

IRMs below range 6)

NO boron has been Injected into RPV

I

Confirm reactor mode

switch In SHUTDOWN

I

Confirm ARI initiation

'I

AN>

Confirm reciro flow runback

to minimum

IF reactor power is above 5%

OdR

CANNOT be determined

THEN trip reciro pumps

InhibitADS

Evaluate override on CP-3 Chart

at coordinate C-2

WHILE PERFORMING THE FOLLOWING

IF SBLCtankleveldropsto8%

THEN tip the SBLC pumps

JE RWCU is NOT isolated

TIEN bypass RWCU filterdomlnerolizers

per 34S0-G31-003-1S

CodShutdown Boo

eight

(Table 3) has beeWniJectdI

IF

boron CANNOT be injected with SBLC

THEN Inject boron using one ormore of the

following per 31 EO-EOP-109-1 S:

"o CRD

"O