ML071570565

4/2007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday , March 12, 2007 2:35:28 PM

22

(

OPL271E-O

Revision°

Page 3 of 16

I.

PROGRAM:

OPERATOR TRAINING - LICENSED

II.

COURSE:

LICENSE TRAINING

III.

LESSON TITLE:

E-O, "Reactor Trip or Safety Injection"

IV.

LENGTH OF LESSON/COURSE:

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

V.

TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of HLC Procedures training, the participant shall be able to explain,

using classroom evaluations and/or simulator scenarios, the requirements of E-O,

"Reactor Trip or Safety Injection".

B. Enabling Objectives

Obiectives

O.

Demonstrate an understanding of NUREG 1122 knowledge's and abilities

associated with Reactor Trip or Safety Injection that are rated <: 2.5 during Initial

LicensEfTraining and <: 3.0 during License Operator Requalification Training for

the appropriate position as identified in Appendix A

I.

State the purpose/goal of this E-O.

2.

Describe the E-O entry conditions .

3.

Summarize the mitigating strategy for the failure that initiated entry into E-O.

4.

Describe the bases for all limits, notes, cautions, and steps of E-O.

5.

Describe the conditions and reason for transitions within this procedure and

transitions to other procedures.

6.

Given a set of initial plant conditions use E-O to correctlv:

a.

Recoanize entrv conditions.

b.

Identifv reauired actions.

c.

Respond to Continaencies.

d.

Observe and Interpret Cautions and Notes.

7.

Apply GFE and system response concepts to the abnormal condition - prior to,

durina and after the abnormal condition .

OPl271E-O

Revision 0

(

Page 8 of 16

X.

LESSON BODY:

INSTRUCTOR NOTES

1.

VERIFY reactor TRIPPED:

Provides indication to be

a.

Refer to E-O for Substeps

used to determine if a valid

b.

Refer to E-O for RNa

RX trip signal exist and

directs tripping RX if NOT

tripped. If RX trip

indications exist and RX is

NOT tripped then GO TO

FR-S.1 for ATWS

2.

VERiFY turbine TRIPPED:

Turbin e trip is required to

a.

Refer to E-O for Substeps

prevent uncontrolled cool

b.

Refer to E-O for RNa

down on RX trip.

3.

VERIFY shutdown boards ENERGIZED:

Objective 5

a.

Refer to E-O for RNa

RX trip/SI response will

require safeguard

equipment operation. If

power cannot be

immediately restored, then

GO TO ECA-O.O for Loss of

(

All AC Power since power

is NOT availabie to

safeguard equipment

4.

DETERMINE if SI actuated:

Objective 5

a.

Refer to E-O for Substeps

if SI is required and not

b.

Refer to E-O for RNa

actuated, then manuaiiy

actuate. If SI is not

required, then GO TO

ES-O.1

5.

PERFORM E-OA, Equipment Verifications, WHILE

Allow reader-doer

continuing in this procedure.

verification of auto ESF

actuations.

6.

DETERMINE if secondary heat sink available:

Objective 5

a.

Refer to E-Ofor Substeps

Verifies AFW pumps

b.

Refer to E-O for RNO

started and operated as

designed and S/G level

assures a heat sink is

available OR GO TO

FR-H.1 to restore heat sink.

7.

CHECK if main steamlines should be isolated:

Determines if MSIVs should

(

a.

Refer to E-O for Substeps

have auto isolated. If NOT

b.

Refer to E-O for RNa

and condition are met, then

manually isolate them.

Prevent RCS cooldown.

REACTOR TRIP OR SAFETY INJECTION

(

',.

saN

ISTEP II ACTION/EXPECTED RESPONSE

E-O

Rev. 28

IIRESPONSE NOT OBTAINED

NOTE 1

NOTE 2

Steps 1 through 4 are immediate action steps.

This procedure has a foldout page.

(

l

1.

2.

3.

VERIFY reactor TRIPPED:

Reactor trip breakers OPEN

Reactor trip bypass breakers

DISCONNECTED or OPEN

Neutron flux DROPPING

Rod bQ.ltom lights LIT

- e,

Rod position indicators

less than or equal to 12 steps.

VERIFY turbine TRIPPED:

Turbine stop valves CLOSED.

VERIFY at least one train of shutdow n

boards ENERGI ZED.

TRIP reactor.

IF reactor CANNOT be tripped,

THEN

PERFORM the following:

a, MONITOR status trees.

b.

GO TO FR-S.1, Nuclear Power

Generation/ATWS.

---a--

TRIP turbine.

IF turbine CANNOT be tripped,

THEN

CLOSE MSIVs and MSIV bypass valves.

ATTEMPT to restore power to

at least one train of shutdown boards.

IF power CANNOT be immediately

restored to at least one train of

shutdown boards,

THEN

GO TO ECA-O.O, Loss of All AC Power.

---a--

Page 4 of 21

r:

-,

SON

REACTOR TRIP RESPONSE

IES-O.1

~

R_e_v_._3_0

_

ISTEPIIAcnONIEXPECTED RESPONSE

II RESPONSE NOT OBTAINED

5.

CHECK if emergency boration is required:

(

a.

VERIFY all control rods fully inserted:

Rod bottom lights LIT

Rod position indicators

less than or equal to 12 steps.

b.

MONITOR RCS temperature:

T-avg greater than 540°F

if any RCP running

T-cold greater than 5400F

if all RCPs stopped.

6.

ANNOUNCE reactor trip

USING PA system.

a.

IF any of the following conditions

exists:

two or more RPls indicate

greater than 12 steps

two or more control rod positions

CANNOT be determined,

Tl-IEN

EMERGENCY BORATE

USING EA-68-4, Emergency

Boration.

b.

EMERGENCY BORATE

as necessary to maintain shutdown

margin USING EA-68-4, Emergency

Boration.

(

I

Page 6 of 15

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

16. 008 G2.4.4 001

(

Given the following plant conditions:

- A LOCA has occurred

- The crew is performing E-1 . Loss of Reactor or Secondary Coolant

- The following parameters exist:

- All SG pressures - 730 psig and slowly trending down

- All SG levels - being controlled at 40% NR

- PRZR level - off-scale high

- RVLlS Lower Range indicates 50%

- Containment Pressure - 3 psig

- RWST level - 74% and decreasing slowly

- RCS pressure - 875 psig and decreasing slowly

- Highest CET - 500°F

Based on these indications, which ONE (1) of the following procedures will the crew

enter next?

A. ES-1.1 , SI Termination

B:' ES-1.2, Post LOCA Cooldown and Depressurization

(

C. ES-1.3, Transfer to Cold Leg Recirculation

D. E-2, Faulted Steam Generator Isolation

B-Correct. RCS Pressure not stable, and low RCS inventory (low reactor vessel level

and high PRZR level indicates a large head bubble).

A-Incorrect. (see B)

C-Incorrect. RCS pressure and RWST level are high. Entry to ES-1.3 on low RWST

level.

D-Incorrect. SG pressures are trending down because RCS temperature is trending

down. (RCS pressure lower than SG pressure)

Monday, March 12, 2007 2:35;28 PM

29

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Emergency Procedures I Plan Ability to recognize abnormal indications for system operating parameters which are entry-level

conditionsfor emergency and abnormal operating procedures.

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

77

SRO 4.3

E-1

Proposed references to be provided to applicants during examination :

None

Learning Objective:

Question Source:

Question History:

Question Cognitive Level:

10 CFR Part 55 Content:

Comments:

OPL271E-1 Objective 5

Bank

North Anna 2006

Higher

43.5

(

(,

Source:

Cognitive Level:

Job Position:

Date:

BANK

HIGHER

4/2007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

NORTH Al'lNA 2006 NRC

SEQUOYAH

NO

Monday, March 12, 2007 2:35:28 PM

30

(

(,

OPL271 E-1

Revision 1

Page 3 of 19

I.

PROGRAM:

OPERATOR TRAINING* LICENSED

II.

COURSE:

LICENSE TRAINING

III.

LESSON TITLE:

E-1 , "Loss of Reactor or Secondary Coolant"

IV.

LENGTH OF LESSON/COURSE:

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

V.

TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or

explain, using classroom evaluations and/or simulator scenarios, the requirements of

E-1, "Loss of Reactor or Secondary Coolant.

B. Enabling Objectives

O.

Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities

-_associated with E-1 , "Loss of Reactor or Secondary Coolant that are rated ;:,

2.5 during Initial License Training and z 3.0 during License Operator

Requalification Training for the appropriate license position as identified in

Appendix A.

1.

Explain the purpose/goal of E-1.

2.

Discuss the E-1 entry conditions.

3.

Summarize the mitigating strategy for the failure that initiated entry into E-1.

4.

Describe the bases for all limits, notes, cautions, and steps of E-1.

5.

Describe the conditions and reason for transitions within this procedure and

transitions to other procedures.

6.

Given a set of initial plant conditions use E-1 to correctly:

a.

Identify required actions

b.

Respond to Contingencies

c.

Observe and Interpret Cautions and Notes

7.

Apply GFE and system response concepts to the performance of E-1

conditions.

(

X.

LESSON BODY:

14.

INITIATE evaluation of plant status:

a.

Refer to EOP for Substeps

b.

Refer to EOP for RNO

NOTE: Power is removed to the ice condenser

AHUs in prepara tion for the following step which

energizes hydrogen mitigation equipment.

15.

MONITOR if hydrogen igniters and recombiners

should be turned on:

a.

Refer to EOP for Substeps

b.

Refer to EOP for RNO

16.

DETERMINE if RCS cooldown and

depressurization is required:

a.

Refer to EOP for Substeps

b.

Refer to EOP for RNO

NOTE: 300 psig is based on RHR shutoff head

including post accident instrument errors

17.

DETERMINE if transfer to cold leg recirculation is

required:

a.

Refer to EOP for Substeps

b.

Refer to EOP for RNO

18.

MONITOR if CLAs should be isolated:

a.

Refer to EOP for Substeps

b.

Refer to EOP for RNO

OPL271 E-1

Revision 1

Page 11 of 19

INSTRUCTOR NOTES

If recirculation capability is

lost. transition to ECA for

required actions.

If break in Aux. Bldg. is

indicated, transitio n to ECA

for required actions.

If break is in RX. Bldg, then

ensure equipment is

available for long term

recovery.

Ensure hydrogen analyzers

in service, and place the

igniters in service if

explosive concentration is

not present. If needed,

hydrogen recombiners are

place in service to lower

hydrogen concentration.

For a small to intermediate

break, transition to ES-1.2

for response.

If RWST is <27% swap to

cold leg injection mode. If

level is >27% GO TO step

14 and repeat steps until

<27%.

For an intermediate or large

break LOCA isolate CLA to

prevent injection of nitrogen

into the RCS and allow

further RCS

depressurization

(

LOSS OF REACTOR OR SECONDARY COOLANT

E-1

Rev. 22

(

1STEP II ACTION/EXPECTED RESPONSE

15.

c.

WHEN ice condenser AHU breakers

have been opened,

THEN

ENERGIZE hydrogen igniters: [M-10]:

HS-268-73 ON

HS-268-74 ON.

d. CHECK containmen t hydrogen

concentration less than 0.5%. [M-10]

16.

DETERMINE if RCS cooldown and

depressurization is required:

a.

CHECK RCS pressure

greater than 300 psig.

b. GO TO ES-1.2, Post LOCA

Cooldown and Depressurization.

---.--

IIRESPONSE NOT OBTA1NED

d. PLACE hydrogen recombiners in

service USING EA-268-1, Placing

Hydrogen Recombiners in Service.

IF hydrogen recombiners

NOT available,

THEN

CONSULT TSC.

a. IF RHR injection flow

greater than 1000 gpm,

THEN

GO TO Step 17 .

Page 17 of 25

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

18. Oil EA2.10 001

(

Given the following plant conditions:

-

A Reactor Trip and Safety Injection have occurred .

-

Containment pressure is 5.6 psig and stable.

RCPs have been stopped.

RVLlS Lower Range is indicating 35%.

-

Core Exit Thermocouples are indicating 710°F.

PZR level is off scale low.

PZR pressure is 400 psig.

RCS Wide Range Hot Leg Temperatures are indicating 680°F.

Which ONE (1) of the following conditions currently exists?

A. A PZR steam space break has occurred and a transition to FR-C.1 , Response to

Inadequate Core Cooling, is required.

B. A PZR steam space break has occurred and a transition to FR-C.2, Response to

Degraded Core Cooling, is required.

C\\" An RCS hot or cold leg break has occurred and a transition to FR-C.1, Response to

Inadequate Core Cooling, is required.

D. An RCS hot or cold leg break has occurred and a transition to FR-C.2, Response to

Degraded Core Cooling, is required.

A. Incorrect. Incorrect event, correct procedure entry

B. Incorrect. Incorrect event, incorrect procedure entry

C. Correct. Requires applicant to distinguish between Orange and Red Path on Core

Cooling CSF. Also requires determination ofbreak location by determining that PZR

level is not off-scale high

D. Incorrect. Correct event, incorrect procedure entry

Monday, March 12, 20072:35:28 PM

32

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Abilityto determine or interpret the following as they apply to a Large Break LOCA: Verification of adequate core cooling

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

78

SR0 4.7

CSFSTs Core Cooling F.02

Proposed references to be provided to appiicants during examination:

None

Learning Objective:

OPL271FR-C.1 Objective 2.b

Question Source:

Modified

Question History:

WTSI Various (Harris)

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments:

(

Source:

MODIFIED

Source If Bank:

Cognitive Level:

HIGHER

Difficulty:

Job Position:

Plant:

SEQUOYAH

Date:

4/2007

Last 2 NRC?:

NO

(

Monday. March 12, 2007 2:35:28 PM

33

(

OPL271FR-C.1

Revision 1

Page 3 of 16

I.

PROGRAM:

OPERATOR TRAINING - LICENSED

II.

COURSE:

LICENSE TRAINING

III.

LESSON TITLE:

FR-C.1 , INADEQUATE CORE COOLING

IV.

LENGTH OF LESSON/COURSE:

1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

V.

TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or

explain, using classroom evaluations and/or simulator scenarios, the requ irements of

FR-C. l , INAD EQUATE CORE COOLING.

B. Enabling Objectives

Objectives

O.

Demonstrate an understanding of NUREG 1122 knowledge's and abilities

associated with Inadequate Core Cooling that are rated z 2.5 during Initial License

Training and 2: 3.0 during License Operator Requalification Training for the

appropriate position as identified in Appendix A

I.

State the purpose/goal of this FR-C.1.

2.

Describe the FR-C.1entrv conditions.

a.

Describe Ihe planI parameters and setpoints associated with FR-C.1

entrv conditions.

b.

Demonstrate an understanding of the use of F-G, Status Trees to

indicate when FR-C.1 must be implemented.

3.

Summarize the mitigating strategy for the failure that initiated entry into FR-C.1.

4.

Describe the bases for all limits, notes, cautions, and steps of FR-C.1.

5.

Describe the conditions and reason for transitions within this procedure and

transitions to other procedures.

6.

Given a set of initial plant conditions use FR-C.1 to correctlv:

a.

Recoanize entrv conditions.

b.

Identifv required actions.

c.

Resoond to Continoencies.

d.

Observe and Interpret Cautions and Notes.

7.

Apply GFE and system response concepts to the abnormal condition - prior to,

durinn and after the abnormal condition.

(

X.

LESSON BODY:

A.

Purpose

1.

This instruction provides guidance for operators to

mitigate the effects of inadequate core cooling.

2.

The diagnosis of the condition is assumed to be

complete prior to this step. Section 2.0 SYMPTOMS

AND ENTRY CONDITIONS contains a listing of

symptoms and entry conditions.

3.

Mitigating Strategy

a.

Ensure a cool water supply to RHR pumps

suction

b.

Establish ECCS flow including opening the CLA

isolation valve if closed

c.

Control containment hydrogen concentration

d.

lsolgte RCS inventory loss paths

e.

Depressurize the S/Gs to reduce RCS pressure

to <100 psig to allow increased ECCS flow

f.

Start one RCP at a time to get CET <12000f

g. Open RCS inventory loss paths to lower RCS

pressure to establish feed and bleed of RCS.

h.

Depressurize S/G to atmospheric pressure.

i.

If none of the above restores core cooling. then

transition to a Severe Accident Guideline.

j.

Verify core cooling and transition to E-1 for

restoring the unit to post accident conditions.

B.

Operator Act ions

1.

MONITOR RWST level greater than 27%.

a.

Refer to FR-C.1 for RNO

CAUTIO N:

Refer to CAUTION in AOP

OPL271FR-C.1

Revision 1

Page 7 of 16

INSTRUCTOR NOTES

Objective 1

Objective 2

Refer to Section 2.0 of this

procedure and discuss

entry conditions.

Only entry condition is from

FR-O. Status Trees

Objective 3

Objective 4

Objective 6

Objective 5

Ensures a suction flow path

to the RHR pumps for core

cooling

Running RHR pumps on

recirculation wlo CCS on

HX may cause pump failure

due to overheating or

cavitation

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

32. 026 G2.2.25 001

(

Given the following plant conditions:

A loss of Component Cooling Water has occurred on Unit 1.

The crew is performing actions of AOP M.03, Loss of Component Cooling

Water.

1A Surge Tank level indicates 0%.

1B Surge Tank level indicates 57% and stable.

Which ONE (1) of the following describes the effect of this condition , and the action

required?

A'!' Component Cooling Water System operability meets safety analysis assumptions.

Trip the reactor, Trip RCPs, and enter E-O, Reactor Trip or Safety Injection.

B. Component Cooling Water System operability meets safety analysis assumptions.

Monitor RCP temperatures and continue attempts to identify and isolate the leak in

accordance with M.03. Mode 1 operations may continue until technical

specifications requires a unit shutdown.

C. Component Cooling Water operability no longer meets safety analysis assumptions.

Trip the reactor, Trip RCPs, and enter E-O, Reactor Trip or Safety Injection.

(

D. Component Cooling Water operability no longer meets safety analysis assum ptions.

Monitor RCP temperatures and continue attempts to identify and isolate the leak in

accordance with M.03. Mode 1 operations may continue until technical

specifications requires a unit shutdown.

A is correct. Single failure loss of 1 train, still have B train available. Trip Rx due to

loss of surge tank level

B is incorrect because reactor trip is required

C and 0 are incorrect because safety function is still met for these conditions. C

contains the correct actions, 0 contains plausible actions for loss of train with TS

implications

Monday, March 12, 2007 2:35:30 PM

60

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Equipment Control Knowledgeof bases in technical specifications for limiting conditions foroperations and safetylimits.

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

79

SRO 3.7

TS 3.7.3, M.03

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPL271AOP-M.03 Objective 5,6,9

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments:

(

Source:

Cognitive Level:

Job Position:

Date:

NEW

HIGHER

SURVEILLANCE REQUIREMENTS

412007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday, March 12, 2007 2:35:30 PM

61

(

PLANT SYSTEMS

3/4.7.3 COMPONENT COOLING WATER SYSTEM

LIMITING CONDITION FOR OPERATION

3.7.3 At least two independent component cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72

hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4.7.3 At least two component cooling water loops shall be demonstrated OPERABLE :

a.

At least once per 31 days on a STAGGERED TEST BASIS by verifying that each valve (manual, power

operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured

in position, is in its correct position.

(

b.

At least once per 18 months, during shutdown, by verifying that each component cooling system pump

starts automatically on a Safety Injection test signal.

March 25, 1982

(

PLANT SYSTEMS

BASES

3/4.7.3 COMPONENT COOLING WATER SYSTEM

The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available

for continued operation of safety related equipment during normal and accident conditions. The redundant cooling

capacity of this system . assuming a single failure, is consistent with the assumptions used in the accident analyses .

3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM

The OPERABILITY of the essential raw cooling water system ensures that sufficient cooling capacity is

available for continued operation of safety related equipment during normal and accident conditions. The redundant

cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident

conditions Within acceptable limits.

(

LOSS OF COMPONENT COOLING WATER

AOP*M.03

Rev. 11

(

I STEP I

ACTION/EXPECTED RESPONSE

2.3

7.

ENSURE surge tank auto makeup starts

at 64% level.

8.

MONITOR Tra in A surge tank level

greater tha!:i 0%.

RESPONSE NOT OBTAINED

DISPATCH operator to PERFORM the

following:

Manually make up from demin water,

ALIGN ERCW supply USING

Appendix E, Aligning ERCW

Emergency Makeup. [C.1]

IF surge tank level drops to 0%,

THEN

PERFORM the following for the affected unit:

a.

IF affected unit in Mode 1 or 2,

THEN

PERFORM the following:

1)

TRIP reactor.

2)

STOP RCPs.

3)

GO TO E-O, Reactor Trip or

Safety Injection, WHILE continuing

in this procedure. [C.2]

---.--

b.

ENSURE RCPs are TRIPPED.

c.

STOP and LOCK OUT affected Unit's

Thermal Barrier Booster Pump s.

(

(Step continued on next page.)

Page 16 of 64

("

(

OPL271AOP-M.03

Revision 0

Page 3 of 32

I. PROGRAM:

OPERATOR TRAINING - LICENSED

II. COURSE:

LICENSE TRAINING

III. LESSON TITLE:

AOP-M.03 "LOSS OF COMPONENT COOLING WATER"

IV. LENGTH OF LESSON/COURSE:

- 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

V. TRAINING OBJ ECTIVES:

A.

Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or

explain, using classroom evaluations and/or simulator scenarios, the requirements of

AOP-M .03, LOSS OF COMPONENT COOLING WATER.

B. Enabling Objectives

Obiectives

O.

Demonstrate an understanding of NUREG 1122 knowledge's and abilities

associated with Loss of Component Cooling Water that are rated z 2.5 during Initial

License Training and z 3,0 during License Operator Requalification Training for the

appropriate position as identified in Appendix A

1.

State the purpose/goal of AOP-M,03.

2.

Describe the AOP-M.03 entry conditions.

a.

Describe the setpoints, interlocks, and automatic actions associated with

AOP-M.03 entry conditions.

b.

Describe the ARP requirements associated with AOP-M.03 entry conditions.

c.

Interpret, prioritize, and verify associated alarms are consistent with AOP-

M.03 entry conditions.

d.

Describe the Administrative and Tech Spec conditions resulting from a Loss

of Component Cooling Water.

3.

Describe the initial operator response to stabilize the plant upon entry into AOP-

M.03.

4.

Upon entry into AOP-M.03, diagnose the applicable condition and transition to the

appropriate procedural section for response.

5.

SUmmarize themitigating'strategy for the"condition that initiated entry into AOP-

M:03.

6.

<f Describe the bases for,alllimits,"notes,. cautions; and steps of AOP-M.03.: n

..........'. - .~. ,

. , ..'.'

... ,.. -...-' ,'

(

OPL271AOP-M.03

Revision 0

Page 4 of 32

7.

Describe the conditions and reason for transitions within this procedure and

transitions to other procedures.

8.

Given a set of initial plant conditions use AOP-M.03 to correctly:

a.

Recognize entry conditions

b.

Identify required actions

c.

Respond to Contingencies

d.

Observe and Interpret Cautions and Notes

9.

p~~.crjbeJ!l~Je cMpec ~ !l1.."f!3.~ .£<;:t!C?IJ~..~pplipa bJe 9u rir)g Jhe. pe.!i~~~~~e of

A.Q~-M .03 .

10.

Apply GFE and system response concepts to the abnormal condition

- prior to. during and after the abnormal condition.

(

c

X.

LESSON BODY:

1.

Monitor Reactor Coolant Pumps Motor Thrust

Bearing TEMP HIGH ANNUNCIATOR Dark

(M-5B, E-3).

RNO directs trip of the reactor & RCPs, then Go

to E-O if in Mode 1 or 2 while continuing in AOP.

If in any other mode then Trip RCPs, and if in

Mode 4, 5,or 6 directs stabilizing RCS temp

using RHR cooling.

2.

DISPATCH operators with radios to Aux Bldg to

locate failure and perform valve manipulations.

3.

DISPATCH an operator with radio to perform

Appendix A, Operation of Appendix R Valves

required by section 2.3.

Appendix A identifies valve, breaker location,

and transfer switch. Also gives step tex1 to

place xfer sw. to auto, close breaker, operate

valve when directed and remove power when

directed.

2 valves per unit -ESF Header A isolation &

Misc Equip& Bldg Supply Isolation

4.

ENSURE all affected Units CCS Pumps Running.

5.

CHECK ERCW flows normal for present

conditions:

RNOdirects Go To AOP-M.01 Loss of ERCW

NOTE I:rlirtliifeventof a'A';.Train linebreak.

the surqetankbaffle prevents'trie B

Train from draining to less than 57.%

indicated levelI

6.

MONITOR Train A CCS Surge Tank level

between 65% and 85%.

1(2)-L1-70-63A, UNIT 1(2) A CCS Surge

Tank level.

OPL271AOP-M.03

Revision 0

Page 16 of 32

INSTRUCTOR NOTES

This appendix places power

on the appendix R valves

for use in subsequent

steps.

Discuss Appendix A

Including the requirement

that if power is on for a

listed valve - an operator

contact with the MCR is

to remain at the breaker

so that if a fire develops

the valve can be

positioned as required &

the power removed

(

X.

LESSON BODY:

7.

ENSURE Surge Tank AUTO Makeup starts at

64% level.

RNO provides alternate makeup paths

(manually from Demin water or ERCW via

Appendix E)

8.

MONITORA ;Train .CCS.Surge Tanklevel.qreater

<than 0%....

OPL271AOP-M.03

Revis ion 0

Page 17 of 32

INSTRUCTOR NOTES

App E discussed in

Introd uction system review

RNO directs trip of the reactor & RCPs, then Go

to E-O if in Mode 1 or 2 while continuing in AOP.

If in any other mode then Trip RCPs, and if in

Mode 4, 5,or 6 directs stabilizing RCS temp

using RHR cooling

RNO also directs the Stop & Lock out of

affected Units TBBP and Train A Pumps

(A-A CCS , B-B CCS [if aligned to A train] and

A-A CS pumps)

RNQ also Ensures Ltdn & Excess Ltdn isoiated,

RCP seal return & charging valves closed, CCP

suction to RWST, and if in Modes 4-6, Train B

(

RHR in service. Then directs to applicable step

NOTE: A high flow indication on FI-70-42 may

indicate a line break on the Rx Bldg

Supply Return Header. A line break in

any other areas of the CCS may rob

flow to the Rx Bldg Supply Header

resulting in inadequate flow to the RCP

oil coolers.

(

Note preceding step to

monitor Rx bldg header flow

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

35. 02702.4.49 001

("

Given the following plant conditions:

-

Unit 1 is operating at 100% RTP.

-

Pressure Control Channel Selector Switch, 1-XS-68-340D is selected to

PT-68-340 & 334 position.

-

PT-68-340 fails LOW.

-

All equipment functions as designed.

Which ONE (1) of the following statements describes (1) the first Technical

Specification parameter to be challenged, and (2) the procedural action required?

A. PZR Low Pressure Reactor Trip; Direct a manual reactor trip and enter E-O,

Reactor Trip or Safety Injection.

B:" PZR High Pressure Reactor Trip ; Direct manual control of PZR pressure and enter

AOP 1.04, Pressurizer Instrument Malfunctions.

C. PZR High Pressure Reactor Trip; Direct action to manually close PORV 68-334 and

68-340 and enter AOP 1.04, Pressurizer Instrument Malfunctions.

(

D. PZR Low Pressure Reactor Trip; Direct action to restore RCS pressure to within

Technical Specification limits using applicable annunciator response procedures.

A. Incorrect. Wrong TS challenge. If channel fails low, actual pressure will rise.

Wrong procedure

B. Correct.

C. Incorrect. Only 1 PORV wiff be affected, but procedure entry is correct

D. Incorrect. Wrong TS challenge. If channel fails low, actual pressure will rise. Would

be partially correct action if this failure occurred

Monday, March 12. 2007 2:35:30 PM

66

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Em ergency Procedures ! Plan Ability to perform without reference to procedures those actions that require immed iate ope ration of

system components and controls.

c

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

80

SR04.0

AOP-1.04

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

Question Cognitive Level:

10 CFR Part 55 Content:

Comments:

OPL271AOP-1.04 Objective 3

New

Higher

43.5

(

Maybe this is better and still passes the SRO test?

Source:

NEW

Source If Bank:

Cognitive Level:

HIGHER

Difficulty:

Job Position:

Plant:

Date:

4/2007

Last 2 NRC?:

Monday, March 12,20072:35:30 PM

SEQUOYAH

NO

67

(

PRESSURIZER INSTRUMENT MALFUNCTION

AOP-1.04

Rev. 8

I STEP I

ACTION/EXPECTED RESPONSE

RESPONSE NOT OBTAINED

2.1

Pressurizer Pressure Instrument OR Controller Malfunction

NOTE 1:

NOTE 2:

NOTE 3:

Appendixes H is a layout of PZR pressure control provided for operator

reference .

A failure of channel III (P-68-323) will affect the automatic actuation of

PCV 68-334, PZR PORV, in the normal pressure control circuit. LTOPS

operation of this PORV is unaffected by this failure.

A failure of channel IV (P-68-322) will affect the automatic actuation of

PCV 68-340A, PZR PORV, in the normal pressure control circuit. LTOPS

operation of this PORV is unaffected by this failure.

(

1.

MONITOR pressurizer pressure stable or

trending to desired pressure.

RESTORE pressurizer pressure USING

manual control of the following :

PIC-68-340A

PZR Spray controllers

PIC-68-340D (Loop 1)

AND/OR

PIC-68-340B (Loop 2)

Pressurizer Heaters

Page 4 of 60

(

I.

PROGRAM:

OPERATOR TRAINING - LICENSED

OPL271AOP-1.04

Revision 1

Page 3 of 19

II.

COURSE:

LICENSE TRAIN ING

III.

LESSON TITLE:

AOP-1.04, PRESSURIZER INSTRUMENT MALFUNCTIONS

IV.

LENGTH OF LESSON/COURSE :

V.

TRAINING OBJECTIVES:

A. Terminal Objective:

1.0 hour0 days 0 hours 0 weeks 0 months (s)

(

Upon completion of License Training. the participant shall be able to demonstrate or

explain. using classroom evaluations and/or simulator scenarios, the requirements of

AOP-1.04, PRESSURIZER INSTRUMENT MALFUNCTIONS.

B. Enabling Objectives:

Objectives

O.

Demonstrate an understand ing of NUREG 1122 knowledge 's and abilities

associated with Pressurizer Instrument Malfunctions that are rated z 2.5 during

Initial License Training and z 3.0 during License Operator Requalification Training

for the appropriate position as identified in Appendix A.

1.

State the purpose/goal of this AOP-1.04.

2.

Describe the AOP-1.04 entry conditions.

a.

Describe the setpoints , interlocks, and automatic actions associated with

AOP-1.04 entry conditions.

b.

Describe the ARP requirements associated with AOP- 1.04 entry conditions.

c.

Interpret. prioritize, and verify associated alarms are consistent with AOP-

1.04 entry conditions.

d.

Describe the plant parameters that may indicate a Pressurizer Instrument

Malfunction.

3.

Describe the initial operator response to stabilize the plant upon entry into

AOP-1.04.

4.

Upon entry into AOP-1.04, diagnose the applicable condition and transition to the

appropriate procedural section for response.

5.

Summarize the mitigating strategy for the failure that initiated entry into AOP-1.04.

6.

Describe the bases for all limits, notes. cautions, and steps of AOP-1.04.

7.

Describe the conditions and reason for transitions within this procedure and

transitions to other procedures.

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

52. 058 AA2.02 001

C'

Given the following plant conditions:

- Unit 1 is in Mode 3.

- The following alarms are received in the control room:

- AR-M-01-C, B4, 125V DC VITAL CHGR II FAILURE OR VITAL BAT II

DISCHARGE.

- AR-M-01-C, B5, 125V DC VITAL BAT BD II ABNORMAL.

- Battery Board II Voltage indicates 119 VDC and lowering slowly.

- Battery Charger II DC Output Breaker is tripped open.

Which ONE (1) of the following describes the operability of the Battery Board , and the

action required?

A'I Declare Battery Board II INOPERABLE because voltage is less than 125 VDC.

Consider aligning the Spare Charger in accordance with 0-SO-250-1, 125 Volt dc

Vital Battery Boards.

B. Declare Battery Board II INOPERABLE because the battery is not connected to a

charger. Reduce Battery loading as necessary in accordance with 0-SO-250-1, 125

Volt dc Vitai Battery Boards. Do not align the spare charger until the cause of the

C

breaker trip has been identified

C. The battery remains OPERABLE as long as it remains connected to the Battery

board. Consider aligning the Spare Charger in accordance with 0-SO-250-1, 125

Volt dc Vital Battery Boards.

D. The battery remains OPERABLE as long as it remains connected to the Battery

board . Reduce Battery loading as necessary in accordance with 0-SO-250-1, 125

Volt dc Vital Battery Boards. Do not align the spare charger until the cause of the

breaker trip has been identified.

A is correct. Less than 125 VDC, Annunciator 84 and TS requires LCO entry

8 is incorrect because it is not inop due to charger disconnect.

C and 0 are incorrect because the battery is not operable. C contains correct action,

and 0 remains credible because the plant status and action taken do not eliminate

operability of the battery by themselves. If the battery is connected to the board, it may

be functioning, but it may not be operable (Low Volts)

(

Monday, March 12, 2007 2:35:31 PM

99

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Ability to determine and interpret the following as theyapplyto the Lossof DC Power: 125V de busvoltage. low/critical low,alarm

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

81

SRO 3.6

AR-M-01C

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

Question Cognitive Level:

10 CFR Part 55 Content:

Comments:

OPL271AOP-P.02, B.8.b & 9

New

Higher

43.5,43.2

(

Source:

Cognitive Level:

Job Position:

Date:

NEW

J1IGHER

4/2007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday, March 12,2007 2:35:31 PM

100

(

OPL271AOP-P.02

Revision 0

Page 3 of 17

I.

PROGRAM:

OPERATOR TRAINING - LICENSED

II.

COURSE :

LICENSE TRAINING

III.

LESSON TITLE:

AOP-P.02. LOSS OF 125V DC VITAL BATTERY BOARD

IV.

LENGTH OF LESSON/COURSE:

2.0 hour0 days 0 hours 0 weeks 0 months (s)

V.

TRAINING OBJECTIVES:

A.

Terminal Objective:

Upon completion of License Training, the participant shall be able to

demonstrate orexplain, using classroom evaluations and/or simulatorscenarios,

the requirements of AOP-P.02, LOSS OF 125V DC VITAL BATTERY BOARD.

B.

Enabiing Objectives:

Obiectives

O.

Demonstrate an understanding of NUREG 1122 knowledge's and abilities associated

with Loss of 125V DC Vital Battery Board that are rated ~ 2.5 during Initial License

Training and ~ 3.0 during License Operator Requalification Training for the

appropriate position as identified in Appendix A.

1.

State the purpose/goal of this AOP-P.02.

-

2.

Describe the AOP-P.02 entry conditions.

a.

Describe the setpoints, interlocks, and automatic actions associated with

AOP-P.02 entry conditions.

b.

Describe the ARP reouirements associated with AOP*P.02 entry conditions.

c.

Interpret. prioritize, and verify associated alarms are consistent with AOP-P.02

entry conditions.

d.

Describe the piant parameters that may indicate a Loss of 125V DC Vital Battery

Board.

3.

Describe the initial operator response to stabilize the plant upon entry into AOP-P.02.

4.

Upon entry into AOP-P.02, diagnose the applicable condition and transition to the

appropriate procedural section for response.

(

OPL27 1AOP-P,02

Revision 0

Page 4 of 17

5,

Summarize the mitigating strategy for the faiiure that initiated entry into

AOP-P ,02,

6.

Describe the bases for all limits, notes, cautions, and steps of AOP-P.02.

7.

Describe the conditions and reason for transitions within this procedure and

transitions to other procedures.

8,

Given a set of initial otant conditions use AOP-P.02 to correctiv:

a,

Recoanize entrv conditions,

b.

Identifv reauired actions.

c.

Respond to Continaencies.

d.

Observe and Interpret Cautions and Notes.

g,

Describe the Tech Spec and TRM actions applicable during the performance

of AOP-P .02,

10.

Apply GFE and system response concepts to the abnormal condition - prior

to, durina and after the abnormal condition,

(

LOSS OF 125V DC VITAL BATTERY BOARD

AOP-P.02

Rev. 10

I STEP I

ACTION/EXPECTED RESPO NSE

2.2

Loss of 125V DC Vital Battery Board II (cont'd)

RESPONSE NOT OBTAINED

(

NOTE

Restoring power from a charger is preferred after a fault on the battery board.

9.

RESTORE 125V DC Vital Ballery

Board II from one of the following

USING 0-SO-250-1, 125 Volt DC Vital

Power System: [C .1]

125V DC Ballery II

125V DC Vital Ballery Charger II

125V DC Vital Ballery Charger 1-S

Spare Vital Ballery II with Ballery V

(

10.

CHECK 125V DC Vital Ballery Board II

voltage between 125V and 140V.

11.

GO TO Step 20.

CONTINUE with Step 12.

WHEN voltage returned to normal,

THEN

GO TO Step 20.

Page 17 of 97

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

1. 068 AAl.02 00 1

(

Given the following plant conditions:

-

A fire is in progress in the Unit 1 Control Building Cable Spreading Room.

-

The Fire Brigade is on the scene and has requested backup assistance.

Which ONE (1) of the following describes the procedure selection required for this

event, and subsequent actions to cool down the unit to Mode 5?

A'! Enter AOP-N .01, Plant Fires. Transition to AOP-C.04 , Control Room Inaccessibility

when directed by AOP-N.01. Trip the reactor; Initiate and monitor boration to cold

shutdown conditions from the Aux Control Room OR from Aux Bldg 690 ft.

Penetration Room.

B. Enter AOP-N.08 , Appendix R Fire Safe Shutdown. Trip the reactor; Initiate and

monitor boration to cold shutdown conditions from the Aux Control Room OR from

Aux Bldg 690 ft. Penetration Room.

C. Enter AOP-N .01, Plant Fires. Trip the reactor; Enter E-O, Reactor Trip or Safety

Injection. Monitor for entry to AOP-C.04 , Control Room Inaccessibility. Initiate and

monitor boration in accordance with 1-GO-7, Plant Cooldown from Hot Standby to

Cold Shutdown.

(

D. Enter AOP-N.08, Appendix R Fire Safe Shutdown. Trip the reactor; Enter E-O,

Reactor Trip or Safety Injection. Monitor for entry to AOP-C.04, Control Room

Inaccessibility. Initiate and monitor boration in accordance with 1-GO-7, Plant

Cooldown from Hot Standby to Cold Shutdown .

A. Correct.

B. Incorrect. Initial procedure entry is incorrect. AOP C.04 will be entered

C. Incorrect. AOP C.04 will direct the trip. Boration will be done in accordance with teh

AOP

D. Incorrect. Initial procedure entry is incorrect. AOP C.04 will direct the trip

Monday, March 12, 2007 3:40:07 PM

1

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Abilityto determine and interpretthe following as theyapply to the Control Room Evacuation: Local boric acid flow

('

Question No.

Tier 1 Group 2

Importance Rating:

Technical Reference:

82

SRO 4.2

AOPs N.01, N.08, C.04

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPL271AOP N.01 B.6

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

(

Comments:

Source:

Cognitive Level:

Job Position:

Date:

NEW

HIGHER

4/2007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday, March 12.20073:40:07 PM

2

(

--.'

OPL271AOP-N.01

Revision 1

Page 3 of 15

I.

PROGRAM:

OPERATOR TRAINING - LICENSED

II.

COURSE:

LICENSE TRAINING

III.

LESSON TITLE:

AOP-N.01 , PLANT FIRES

IV.

LENGTH OF LESSON/COURSE:

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

V.

TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participa nt shall be able to

demonstrate or explain, using classroom evaluations and/or simulator scenarios,

the requirements of AOP-N .01, PLANT FIRES.

B.

Enabling Objectives

Obiectives

O.

Demonstrate an understanding of NUREG 1122 knowledge's and abilities associated

with Plant Fires that are rated ~ 2.5 during Initial License Training and ~ 3.0 during

License Operator Requalification Tra ining for the approp riate position as identified in

Appendix A.

1.

State the purpose/goal of this AOP-N.01.

2.

Describe the AOP-N .01 entry cond itions.

a.

Describe the setpoints, interlocks, and automatic actions associated with

AOP-N.01entrv conditions.

b.

Describe the ARP reauirements associated with AOP-N .01 entry conditions.

c.

Interpret, prioritize, and verify associated alarms are consistent with AOP-N.01

entry conditions.

d.

Describe the plant parameters that mav indicate a Plant Fire.

3.

Describe the initial operator response to stabilize the plant upon entry into AOP-N.01.

4.

Summarize the mitigating strategy for the failure that initiated entry into AOP-N .01.

5.

Describe the bases for all limits, notes, cautions, and steps of AOP-N.01.

6.

Describe the conditions and reason for transitions within this procedure and transitions

to other procedures.

---_. ._-- *. -

- -

(

OPL271AOP-N.01

Revision 1

Page 4 of 15

(

7.

Given a set of initial olant conditions use AOP-N.01 to correctlv:

a.

Recoqnize entry conditions.

b.

Identify reauired actions.

c.

Respond to Continqencies.

d.

Observe and Interpret Cautions and Notes.

8.

Describe the Tech Spec and TRM actions applicable during the performance of

AOP-N.01 .

9.

Apply GFE and system response concepts to the abnormal condition - prior to, during

and after the abnormal condition.

OBJECTIVES TO BE COVERED IN THESE SEQUOYAH OPERATOR TRAINING PROGRAMS

I

OBJECTIVE J NONLICENSED

LICENSE TRAINING

NO.

RO

REQUAL/SPECIAL

OPERATORS

O.

X

1.

X

2.

X

3.

X

4.

X

5.

X

6.

X

7.

X

8.

X

9.

X

10.

X

NOTE: The following approval is required for License Requalification and special training only:

Selected objectives to be covered in:

PowerPoint presentation to be used:

Sequoyah Operato r Training Manager

I

Date

Seq uoyah Operations Manager

I

Date

(

PLANT FIRES

IAOP....01

Rev. 22

I STEP I

ACTION/EXPECTED RESPONSE

2.0

OPERATOR ACTIONS (Coot'd)

RESPONSE NOT OBTAINED

(

21.

c.

CHECK if fire threatens ability to

safely maintain hot standby or to

achieve cold shutdown:

Multiple failures/spurious

operations of plant equipment.

Multiple trains/channels of

safety-related equipment

threatened by fire.

d.

IF fire is in Control Building,

THEN -

GO TO AOP-C.04 , Shutdown from

Auxiliary Control Room.

--~.--

e.

IF fire is NOT in Control Building,

THEN

GO TO AOP-N.08, Appendix R Fire

Safe Shutdown .

--~.--

c.

GO TO Step 22.

Page 17 of45

PLANT FIRES

(

I

AOP-N.01 I

Rev. 22

Page 1 of 1

APPENDIX I

ACTIONS FOR FIRE IN CONTROL BUILDING

NOTE

Various fire dampers may not close fully due to high air flow.

The following step should be evaluated based upon the

severity and duration of the fire.

(

1.

IF fire is in one of the following locations

250V Battery Room 1

250V Battery Room 2

Unit 1 Aux Instrument Room

THEN

EVALUATE need to stop Electric Board Room AHUs

to ensure all fire dampers will fully close.

CAUTION

NOTE

Operation of shunt trip breakers will result in all thermal barrier

containment isolation valves inoperable on both units.

The following step opens shunt trip breakers associated with

various motor-operated valves in ERCW system and CCS. The

following step should be evaluated based upon the severity and

duration of the fire.

(

'..

2.

IF fire is in Cable Spreading Rm,

THEN

EVALUATE the need to de-energize motor-operated valves by performing the following:

a,

PLACE the following handswitches in TRIP [1-M-15]:

O-HS-13-204

O-HS-13-205

b. VERIFY power removed from listed valves

USING Appendix C, Cable Spreading Room Fire.

END OF TEXT

Page 43 0145

(

SHUTDOWN FROM AUXILIARY CONTROL ROOM

AOP-C.04

Rev. 13

I STEP I

ACTION/EXPECTED RESPONSE

2.1

Control Room Abandonment

RESPONSE NOT OBTAINED

(

NOTE

EOPs are NOT applicable when evacuating MCR.

1.

ENSURE reactor TRIPPED. [M-4]

2.

ENSURE MSIVs and MSIV bypass valve

handswitches in CLOSE. [M-4]

3.

DISPATCH CRO with radio and Appendix Z

to perform the following:

a.

GO TO AOP-C.04 Cabinet.

[6.9KV Shutdown Board Rm A]

b.

ENSURE personnel dispatched

to perform applicable checklists and

appendices USING Appendix Z,

Task Assignment Sheet.

4.

ENSURE one CCP placed in

PULL TO LOCK.

Page 4 of 183

(

SHUTDOWN FROM AUXILIARY CONTROL ROOM

AOP-C.04

Rev. 13

I STEP I

ACTION/EXPECTED RESPONSE

2.1

Control Room Abandonment

5.

WHEN MCR must be immediately evacuated

due to life-threatening conditions,

THEN

PERFORM the following :

a.

EVACUATE MeR on affected unit(s).

b.

NOTIFY AUOs of MCR evacuation

using radio or PA system.

c.

GO TO Step 11.

6.

ANNOUNCE "Unit __ Reactor trip,

abandoning the Main Control Room"

USING PA System or radio.

7.

PLACE RCP handswitches

in STOP/PULL TO LOCK. [M-5)

Page 5 of 183

RESPONSE NOT OBTAINED

(

saN

SHUTDOWN FROM AUXILIARY CONTROL ROOM

AOP-C.04

Rev. 13

[STEP]

ACTION/EXPECTED RESPONSE

2.1

Control Room Abandonment

8.

PLACE TO AFW LCV handswitches

in CLOSE/PULL-TO-LOCK. [M-3)

RESPONSE NOT OBTAINED

("

NOTE

The following step trips shunt trip breakers for thermal barrier isolation valves

on both units and various ERCW and CCS valves.

9.

ENSURE the following handswitches

placed in TRIP: [1-M-15]

0-HS-13-204

0-HS-13-205

10.

EVACUATE Main Control Room

on affected unit(s).

Page 6 of 183

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

1. E07 EA2.1 001

(

Given the following plant conditions:

-

A Steam Generator Tube Rupture has occurred on Unit 1.

Due to equipment failures, the crew is performing actions contained in ECA-3.2,

SGTR and LOCA - Saturated Recovery.

-

The STA informs you that all CSF Status Trees are GREEN with the exception

of the following:

Core Cooling - YELLOW due to RVLlS level

Inventory - YELLOW due to RVLlS level

Which ONE (1) of the following describes the correct implementation of procedures for

this event?

A. Remain in ECA-3.2 . Do NOT address either procedure. Implementation of Yellow

Path procedures is not allowed in the ECA procedures.

B. Remain in ECA-3.2 while addressing BOTH Yellow Path procedures. You may

choose to implement the actions of either procedure as desired.

C. Transition from ECA-3.2 and address BOTH Yellow Path procedures. The actions

of the Yellow condition on the Core Cooling Safety Function take precedence over

C

the actions of the Inventory Safety Function.

D~ Remain in ECA-3.2 while addressing BOTH Yellow Path procedures. The actions

for the Yellow condition on the Core Cooling Safety Function will not be performed

due to conflict with ECA-3.2 actions. You may choose to implement the actions for

the Yellow Path on Inventory as desired .

A. Incorrect. The crew may treat the actions is this way, but ECAs do not preclude the

use of yel/oow path procedures

B. Incorrect. Normally this would be true, but a caution in ECA-3.2 prohibits

performance of FR-C.3 due to conflict.

C. Incorrect. Transition is not required, and although FR-C.3 is a higher priority, it

would not be performed due to the conflict with ECA-3.2

D. Correct.

Monday, March 12,2007 3:41:07 PM

1

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Abilityto determine and interpretthe following 85 theyapply to the (Saturated Core Cooling) Facility conditions and selection of

appropriate procedures duringabnormal and emergency operations.

(

Question No.

Tier 1 Group 2

Importance Rating:

Technical Reference:

83

SR04.0

EOP User's Guide, FR-C.3 step 1

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

Question Cognitive Level:

10 CFR Part 55 Content:

Comments :

OPL271EPM-4, B.11.c

New

Higher

43.5

(

c.

Source:

Cognitive Level:

Job Position:

Date:

NEW

HIGHER

4/2007

Source If Bank:

Diffi culty:

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday, March 12, 2007 3:41:07 PM

2

(

OPL27 1EPM-4

Revision 0

Page 4 of 26

8.

Given plant operating conditions, determine if AOP entry conditions have been met and state

the resultant appropriate actions for those conditions.

9.

Identify general operating crew responsibilities during emergency operations including

appropriate implementation of prudent operator actions.

10. Identify general operating crew responsibilities during emergency operations including

requirements for actions outside Technicai Specifications/plant licensed conditions

(10CFR50.54x application).

1t.

Given a set of conditions. analyze the EOP/FRP implementation:

a. identify the basis for the implementation;

b. determine the correct implementation hierarchy;

c.

determine if Critical Safety Function Status Trees (CFSTs) implementation is required;

d.

identify the status tree colors by priority and summarize each tree's purpose;

e.

identify conditions which will allow a FRP to be exited once it is entered (a RED or

ORANGE condition);

f.

state the monitoring frequency of CFSTs and when this can be relaxed;

g. determine correct coordination with other support procedures

h.

identify conditions permissible to terminate CFSTs monitoring.

12. Given an operational situation, analyze a crew brief and determine if it meets Management

expectations.

SATURATED COR E COOLING

(

saN

I

FR-C.3

Rev. 4

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

(

l

ISTEP I IACTION/EXPECTED RESPONSE

1.

DETERMINE procedure applicability:

a. CHECK ECA-3.2, SGTR and LOCA

- Saturated Recovery, in effect.

b. RETURN TO procedure and step

in effect.---.--

2.

MONITOR RWST level greater than 27%.

3.

DETERMINE RHR system status:

a. CHECK RHR System NOT aligned

for normal shutdown cooling mode.

II RESPONSE NOT OBTAINED

a. GO TO Step 2.

IF RHR pumps aligned to RWST,

THEN

GO TO ES-1.3. Transfer to RHR

Containment Sump.

--~.--

a. IF RHR aligned for shutdown cooling,

THEN

GO TO AOP-R.03, RHR System

Malfunction.

--~.--

Page 3 of 6

SO N EOt

PROGRAM

MANUAL

USER'S GUIDE

EPM-4

Rev. 18

Page 46 of 92

(

3.10 .5

Status Tree Rules of Usage

3.

Status trees are designed to monitor for the most severe challenges first

to shorten response time in addressing those conditions. Therefore,

typically, RED paths will be at the top of a status tree, following by

ORANGE, YELLOW, and GREEN as the status tree branches downward.

4.

If any RED challenge is detected, the person monitoring status trees

informs the procedure reader immediately before continuing with

monitoring any subsequent status trees. Since they are monitored in

order of importance, the first RED challenge encountered will be the

highest priority RED and thorefore, the highest priority challenge.

5.

If any ORANGE challenge is encountered, the person monitoring status

trees continues monitoring until all six status trees have been evaluated.

This is necessary because a subsequent RED challenge has priority over

any ORANG E challenge. If any RED is encountered, then Rule

3.10.5.0.4 applies. Otherwise, once it is determined that no RED

. chauenqes exist, then the person monitoring status trees informs the

- procedure reader of the highest priority ORANGE challenge.

6.

RED or ORANGE challenges must be addressed immediately by

implementing appropriate FRPs in order of priority and per the rules of

usage . When the person monitoring status trees informs the procedu re

reader that a RED or ORANGE challenge exists, the procedure reader

imme diately suspends the ORP (or lower priority FRP) in progress and

implements the appropriate FRP, as indicated at the terminus point of the

CSF under challenge.

7.

YELLOW challenges may be addressed by implementing appropriate

FRPs if desired, but do not require immediate operator action.

Addressing YELLOW challenges is optional since these are usually

temporary, off-normal conditions that will be restored to normal status by

actions already in progress. In other cases, the YELLOW path might

provide an early indication of a developing RED or ORANGE condition.

Following FRP implementation, a YELLOW might indicate a residual off-

normal condition. When the person monitoring status trees informs the

procedure reader that a YELLOW challenge exists, the procedure reader

should evaluate if the YELLOW challenge FRP should be implemented.

This decision will be based on the following:

Whether the procedures in effect will address the challenge as a

matter of course.

Whether the procedures in effect are more important at that time

based upon available time and current plant conditions.

Whether the challenge is of a nature that it will likely develop into an

ORANGE or RED condition if action is not taken early.

....

SO N EOI

PROGRAM

MANUAL

USER'S GUIDE

EPM-4

Rev. 18

Page 47 of 92

3.10.5

Sta tus Tree Rules of Usage (Continued)

E.

CSF challenges are addressed as follows :

1.

2.

3.

4.

5.

6.

When addressing RED or ORANGE challenges, all ORPs and lower

priority FRPs are suspended . Status tree monitoring continues in case

higher priority challenges occur.

The FRP associated with the highest priority challenge is entered. If an

FRP is being performed and a higher priority RED or ORANGE path

comes in, the current FRP should be suspended and a transition made to

the higher priority FRP unless stated otherwise in the procedure. After

the new FRP has been completed and guidance is provided to "RETURN

TO procedure and step in effect", the operator should go back to the

previous FRP which had been implemented and was, therefore, the

procedure in effect. (DW-99-061)

If an FRP is in progress due to an ORANG E path condition and the same

path turns to RED, then the following guidance is applicable:

a. If the FRP in progress addresses both the RED and ORANGE

condition, operators should continue in the guideiine in progress,

since the actions are the same. 11 the appiicable FRP was exited

prior to the condition degrading from ORANGE to RED, then

operators should re-enter the FRP at step 1 since plant conditions

have changed creating a higher priority challenge. (OW-97-001)

b. If the FRP is different for the RED path conditio n, operators should

transition to the applicable RED path FRP.

The initiation of FRPs is dependent upon current plant parameters. If a

RED or ORANG E priority condition comes in and clears, the FRG does

not need to be performed, If conditions degrade, the safety function will

become a continuous RED or ORANGE condition, at which time the

appropriate FRP should be implemented.

It is expected that FRP actions will clear the RED or ORANGE challenge

before all the FRP actions are complete. The FRP should be performed

to completion (until a defined exit point is reached) even if the RED or

ORANGE challenge is cleared prior to completion of the FRP.

YELLOW path FRPs are considered lower in priority than ORPs

(including applicable foldout page items). While performing YELLOW

path actions, the ORP that the operator was in when he transitioned to

the YELLOW path FRP is considered the controlling procedure.

Continuous actions or foldout page items of the ORP in effect are still

applicable and should be monitored by the operator.

(

c

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

79. E09 02.2.22 001

Given the following plant conditions:

A reactor trip has occurred on Unit 1.

Off-Site power has been lost.

Plant Cooldown to Mode 5 is anticipated.

The crew is preparing to exit E-O, Reactor Trip or Safety Injection.

Which ONE (1) of the following describes how the cooldown will be performed, and

how Technical Specification Shutdown Margin requirements will be maintained during

the cooldown?

RCS cooldown will be performed in accordance with...

A. 0-GO-7, Unit Shutdown from Hot Standby to Cold Shutdown . Perform Emergency

Boration in accordance with EA-68-4, Emergency Boration, to Cold Shutdown

conditions.

B. 0-GO-7, Unit Shutdown from Hot Standby to Cold Shutdown. Perform normal

boration in accordance with SO-62-7, Boron Concentration Control.

C~ ES-0.2, Natural Circulation Cooldown. Perform Emergency Boration in accordance

with EA-68-4, Emergency Boration, to Cold Shutdown conditions.

D. ES-0.2 , Natural Circulation Cooldown . Perform normal boration in accordance with

SO-62-7, Boron Concentration Control.

A. Incorrect. Correct boration procedure but incorrect shutdown procedure

B. Incorrect.

The shutdown and cooldown is performed lAW ES-0.2, and an

emergency boration is performed.

C. Correct. Because the EOPs are being used, Natural Circ Cooldown will be

performed. SOM is maintained and verified at 100 degree increments during the

cooldown, and the Emergency Boration procedure is used

O. Incorrect.

Correct EOP, but emergency boration is performed

KA is matched because the actions required during a natural circulation cooldown are

for the purpose of maintaining minimum TS SOM requirements.

This is the only TS

requirement directly related to performance of this procedure

Monday, March 12, 2007 2:35:35 PM

152

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Equipment ControlKnowtedge of limiting conditions for operations and safety limits.

(

Question No.

Tier 1 Group 2

Importance Rating:

Technical Reference: *

84

SRO 4.1

E-O, ES-O.2

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPL271ES-O.2, B.6.a

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments:

(

Source:

Cognitive Level:

Job Position:

Date:

NEW

HIGHER

4/2007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday, March 12, 20072:35:35 PM

153

c

(

OPL271 ES-0.2

Revision 0

Page 3 of 17

I.

PROGRAM:

OPERATOR TRAINING - LICENSED

II.

COURSE:

LICENSE TRAINING

III.

LESSON TITLE:

ES-0.2, "Natural Circulation Cooldown"

IV.

LENGTH OF LESSON/COURSE:

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

V.

TRAINING OBJECTIVES:

A. Terminal Objecti ve:

Upon completion of HLC Procedures training , the participant shall be able to explain,

using classroom evaluations and/or simulator scenarios, the requirements of EOP

ES-O.2, "Natural Circulation Cooldown".

B. Enabling Objectives

O.

Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities

.associated with Natural Circulation Cooldown that are rated ~ 2.5 during

Initial License Training and ~ 3.0 during License Operator Requalification

Training for the appropriate license position as identified in Appendix A.

1.

Explain the purpose/goal of ES-0.2.

2.

Discuss the ES-0.2 entry conditions.

a.

Describe the setpoints, interlocks, and automatic actions associated with

ES-0.2 entry conditions.

b.

Describe the requirements associated with ES-0.2 entry conditions.

3.

Summarize the mitigating strategy for the failure that initiated entry into ES-O.2.

4.

Describe the bases for all limits, notes, cautions, and steps of ES-0.2.

5.

Describe the conditions and reason for transitions within this procedure and

transitions to other procedures.

6.

Given a set of initial plant conditions use ES-O.2 to correctly:

a.

Identify required actions

b.

Respond to Contingencies

c.

Observe and Interpret Cautions and Notes

7.

Apply GFE and system response concepts to the performance of ES-0.2

conditions.

(

E-O

SON

REACTOR TRIP OR SAFETY INJECTION

Rev. 28

ISTEP IIACTION/EXPECTED RESPONSE

IIRESPONSE NOT OBTAINED

(

4.

DETERMINE if 51 actuated:

ECC5 pumps RUNNING.

Any 5 1alarm LIT [M-4D].

DETERMINE if 5 1require d:

a.

IF any of the following conditions

exists:

5/G pressure less than 600 psig,

RC5 pressure

less than 1870 psig,

Containment pressure

greater than 1.5 psig,

THEN

ACTUATE 51.

b.

IF 51is NOT required,

THEN

PERFORM the following:

1)

MONITOR status trees.

2) GO TO E5-0.1, Reactor Trip

Response .

--~.--

Page 5 of 21

(

saN

REACTOR TRIP RESPONSE

IES-O.1

Rev. 30

l-..-

...L:....-

---'

...l

ISTEPII ACTIONJEXPECTED RESPONSE

II RESPONSE NOT OBTAINED

17.

DETERMINE if natural circulation

cooldown is required:

(

a. CHECK at least one RCP RUNNING.

b.

CHECK at least one AFW pump

AVAILABLE.

c.

SELECT appropriate procedure:

o-GC>-8, Power Reduction from

30% Reactor Power to Hot

Standby (if maintaining hot

standby)

o-G0-7, Unit Shutdown from Hot

Standby to Cold Shutdown

other appropriate procedure

as determined by Shift Manager

or TSC (if manned).

d.

GO TO appropriate plant procedure.

---.--

END

a. IF plant cooldown required

with NO RCP available,

THEN

GO TO E5-0.2, Natural Circulation

Cooldown.---.--

DO NOT CONTINUE this procedure

UNTIL at least one RCP restarted.

b. DO NOT CONTINUE this procedure

UNTIL at least one AFW pump

AVAILABLE.

Page 15 of 15

(

\\ ,

NATURAL CIRCULATION COOLDOWN

ES*O.2

Rev. 15

ISTEP IIACTION/EXPECTED RESPONSE

IIRESPONSE NOT OBTAINED

NOTE

This procedure has a foldout page.

1.

MONITOR SI NOT actuated.

Sl ACTUATED permissive DARK

[M-4A, D4]

IF SI actuated,

THEN

GO TO E*O, Reactor Trip or Safety

Injection. ---

(

CAUTION

NOTE

Loss of all RCP seal cooling may cause RCP seal damage and will

require a TSC status evaluation prior to restarting affected RCPs.

Starting RCP #2 is preferred to provide normal pressurizer spray

capability.

2.

MONITOR if an RCP can be started:

a.

IF all RCP seal cooling has previously

been lost,

THEN

NOTIFY TSC to initiate RCP restart

status evaluation.

b.

ESTABLISH conditions for starting

!:'.

GO TO Step 3.

an RCP USING EA-68-2, Establishing

m,...

RCP Start Conditions.

~

(

c.

START one RCP.

d.

GO TO appropriate plant procedure.

-_.....--

c.

GO TO Slep 3.

Page 30f27

(

SON

NATURAL CIRCULATION COOLDOWN

ES-O.2

Rev. is

ISTEP IIACTION/EXPECTED RESPONSE

3.

INITIATE emergency boration

for cooldown to cold shutdown

USING EA*68-4. Emergency Boration.

II RESPONSE NOT OBTAINED

4.

VERIFY RCS boron concentration

required for cooldown to 450' F:

a.

NOTIFY STA or US to determine

RCS boron concentration required

for ~hutdown margin at 450'F

USING O-SI-NUC-OOo-038.0.

ppm

b.

NOTIFY Chem Lab to periodically

sample RCS boron concentration

at following sample points:

RCS hot legs 1 and 3

CVCS letdown line

....

Pressurizer liquid space.

c.

PERFORM EA*68*5, Determining

Total RCS Boron Concentration

on Natural Circulation.

d.

CHECK Total ReS boron

concentration from EA*68*5

greater than boron concentration

determined in Substep 4.a.

d.

CONTINUE Boralion.

GO TO Substep 4.b.

Page 4 of27

c

(

c..

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

83. E15 G2.1.32 00 1

Given the following plant conditions:

A LOCA has occurred and the following containment conditions exist upon transition

from E-O:

Pressure is 2.26 psig and rising.

-

Sump level is 70% and rising.

Upper and Lower Containment Radiation level is 102 Rem per hour.

Which ONE (1) of the following describes the impact on the unit and which ONE (1) of

the following procedures will be used to mitigate the event?

A. Containment Pressure is rising to a level that post-accident integrity could be

threatened. Transition to FR-Z.1, Response to High Containment Pressure.

B~ Containment Sump is rising to a level that the operation of critical components may

be affected. Transition to FR-Z.2, Response to Containment Flooding.

C. Containment Radiation level requires transition to FR-Z.3, Response to High

ContainmentRadiation, to ensure proper ventilation alignment to minimize

potential for off-site release.

D. All Containment Critical Safety Function conditions are GREEN. Go to E-1, Loss of

Reactor or Secondary Coolant.

A. Incorrect. Preessure is below the criteria for an orange path that would require entry

to FR-Z.1

B. Correct.

C. Incorrect. A yellow condition for radiation level does exist, but the orange condition

on cntmt flood level requires entry to a higher priority procedure

D. Incorrect. There is an orange and yesllow condition on the Containment CSF. E-1

will be performed after addressing the CSF

Monday, March 12,20072:35:36 PM

160

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Conduct of Operations: Ability to explain and apply all system limitsand precautions.

c

Question No.

Tier 1 Group 2

Importance Rating:

Technical Reference:

85

SRO 3.8

FR-Z.2 background, FR-O

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPL271FR-Z.2,8.5

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments:

(

Source:

Cognitive Level:

Job Position:

Date:

NEW

HIGHER

412007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday, March 12,20072:35:36 PM

161

(

OPL271FR-Z.2

Revision 1

Page 3 of 14

I.

PROGRAM:

OPERATOR TRAINING - LICENSED

II.

COURSE:

LICENSE TRAINING

III.

LESSON TITLE:

FR-Z.2, CONTAINMENT FLOODING

IV.

LENGTH OF LESSON/COURSE:

.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

V.

TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or

explain. usingclassroom evaluationsandlor simulator scenarios, the requirements of

FR-Z.2, Containment Flooding.

B. Enabling Objectives

O.

Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities

associated with FR-Z.2, Containment Flooding, that are rated " 2.5 during

Initial License Training and " 3.0 during License Operator Requalification

Training for the appropriate license position as identified in Appendix A.

1.

Explain the purpose/goal of FR-Z.2.

2.

Discuss the FR-Z.2 entry conditions.

a.

Describe the setpoints, interlocks, and automatic actions associated with

FR-Z.2 entry conditions.

b.

Describe the requirements associated with FR-Z.2 entry conditions.

3.

Summarize the mitigating strategy for the failure that initiated entry into FR-Z.2.

4.

Describe the bases for ali limits, notes, cautions. and steps of FR-Z.2.

5.

Describe the conditions and reason for transitions within this procedure and

transitions to other procedures.

6.

Given a set of initial plant conditions use FR-Z.2 to correctly:

a.

Identify required actions

b.

Respond to Contingencies

c.

Observe and Interpret Cautions and Notes

7.

Apply GFE and system response concepts to the performance of FR-Z.2

conditions.

c

(

QUESTIONS REPORT

for SEQUOYAH 2007 * NRC EXAM REV DRAFT

13. 007 G2.1.12 001

Given the following plant conditions:

Unit 1 is operating at 100% power.

-

The following alarm is received:

PRT HIGH PRESS

PZR PORV 69-340 acoustic monitor indicates discharge.

The RO places the PORV Control Switch in CLOSE.

PZR Pressure continues to drop with both Red and Green light indication

for the PORV extinguished.

-

The RO closes the associated block valve and PZR Pressure stops dropping

and is now at 2110 psig and increasing slowly.

Which ONE (1) of the following describes the MINIMUM required actions in accordance

with Technical Specifications?

A. Close and maintain power available to associated block valve within one hour;

restore RCS pressure to a minimum of 2220 psia within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

-

B. Close and maintain power available to associated block valve within one hour;

restore RCS pressure to a minimum of 2185 psia within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

C. Close and remove power from the associated block valve within one hour; restore

RCS pressure to a minimum of 2185 psia within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

D~ Close and remove power from associated block valve within one hour; restore RCS

pressure to a minimum of 2220 psia within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

o is correct.

A is incorrect.

Power would not be maintained because unable to cycle the leaking

PORV. Credible because conditions could exist that would allow power to be

maintained, and applicant must interpret indication to make that decision. Time for

restoring pressure is correct

B is incorrect.

Power would not be maintained because unable to cycle the leaking

PORV. Credible because conditions could exist that would allow power to be

maintained, and applicant must interpret indication to make that decision

C is incorrect. Credible because action to remove power is correct, and time for

restoring pressure is incorrect

Monday, March 12, 20072:35:28 PM

23

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Conduct of Operations: Abilityto applytechnical specifications for a system.

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

86

SR04.0

TS 3.2.5, TS 3.4.3.2

Proposed references to be provided to applicants during examination:

None

Leaming Objective:

OPT200PZRPCS Objective 6

Question Source:

Bank

Question History:

WTSI Bank

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.2

Comments:

Source:

BANK

Source If Bank:

WTSI

Cognitive Level:

HIGHER

Difficulty:

(

Job Position:

Plant:

SEQUOYAH

Dale:

4/2007

Last ZNRC?:

NO

(

Monday, March 12,20072:35:28 PM

24

(

OPT200.PZRPCS

Rev. 2

(

Page 4 of 109

V.

T RAINING OBJECTIVES (Cont'd):

B. Enabling Objectives (Cont'd):

5.

Describe the normal, abnormal, and emergency operation of the Pressurizer

Pressure Control System & Pressurizer Relief Tank as it relates to the following:

a. Precautions and limitations

b. Major steps performed while placing the Pressurizer Pressure Control System

& Pressurizer Relief Tank in service

e. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect Pressurizer Pressure Control System

& Pressurizer ReliefTank operation

f. How an instrument failure will affect system operation

6.

Describe the administrative controls and limits for the Pressurizer Pressure

Control System & Pressurizer ReliefTank as explained in this lesson:

a. State Tech Specs/TlcM LCOs that govern the Pressurizer Pressure Control

System & Pressurizer Relief Tank.

b: State the :51 hour action limit TS LCOs

c. Given the conditions/status of the Pressurizer Pressure Control System &

Pressurizer Relief Tank components and the appropriate sections of the Tech

Spec, dctenn ine if operability requirements are met and what actions are

required

7.

Discuss related Industry Events:

a.

OE21152:

Pressurizer Pressure Control System Design Deficiency with

Westinghouse 7100control system (Turkey Point)

b.

OE22498:

Boron Concentrationchanges when Backup Pressurizer Heaters were

Placed in Service(Salem)

c.

Inadvertant Actuation of Pressurizer Spray and Power Operated Relief Valves

During Controller Transfer (South Texas Project - U-2)

VI.

TRAINING AIDS:

A.

Classroo m Computer and Local Area Network (LAN) Access

B.

Computer projector

(

OPT200.PZRPCS

Rev. 2

Page 93 of 109

Tech Spec Exercise

~ U-1 reactor power is 100% RTP. All

equipment is operable when the operating

crew experiences a failed Pzr pressure

channel PT-68-323.

~ PT-68-323 failed high. The crew completes

the appropriate response using AOP-1.04. All

plant parameters are stable.

-- QUESTION:

What Tech Spec and/or TRM

LCOs should be entered as a result of the

failed Pzr pressure instrument, 1-PT-68-323?

E06

All ow time for stude nts to reference Tech Specs and the TRJI,I. Ask students to present their answer aloud

in class for discu ssion. The instructor sho uld make corrections and additions as necessary.

U- I reactor power is 100% RTP . All equipment is when the operating cre w experiencing a failed

pressurizer pressure channel PT-68-323.

PT -68-323 failed high . The crew completes the appropriate response using AOP-l.04. All plant

parameters arc stable.

QUESTION:

What Tech Spec and/or TRJI,I LCOs should be entered as a result of the failed Pzr pressure

instrument PT-68-323?

A:'iSWER: PCV-68-334 interlock is inoperable.

3.3.1.1

Reactor Trip System Instrumentation. As a minimum, the reactor trip system instrumentation channels

and interlocks ofT able 3.3-1 shall be OPERABLE.

Item 7 - Overtemperature 6 T Four Loop Ope ration. Moues 1 & 2 - Action 6

Item 9 - Pressurizer Pressure - Low. Modes 1 & 2 - Action 6

Item 10 - Pressurizer Pressure - High. . Modes 1 & 2 - Action 6

3.3.2.1

The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown

in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the

Trip Setpoint column of Table 3.3-4. Action b.

Item I - SAFETY INJECTIO:-l, TRUBINE TRIP AND FEEDWATER ISOLATIOIN

d. Pressurizer Pressure - Low. Modes I, 2, & 3# - Action 17

Item 8. - Engineered Safety Feature Actuation System Interlocks

a. Pressurizer Pressure P-lllNot P-I l.

Modes 1,2, & 3 - Action 22a

3.4.3.2

Relief Valves - Operating - Two power operated relief valves (PORVs) and their associated block valves

shall be OPERABLE.

Action a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months close PORV block valve, FCV-68-333.

(

l

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

26. 022 A2 .03 00 1

Given the following plant conditions:

Unit 1 is at 100% power.

-

The following alarm is received:

1-AR-M6E-E5, MOTOR TRIP OUT PNL 1-M-9

The RO determines that Lower Compartment Cooling Fan 1B-B control switch

has a white light.

Fans 1C-A and 1A-A are running.

Which ONE (1) of the following describes the primary concern related to this failure,

and the procedural actions required?

A. Lower Ice Condenser doors may open due to high DP; Stop one Upper

Compartment Cooling fan to equalize pressure in accordance with the alarm

response procedures.

B. Lower Ice Condenser doors may open due to high DP; Start any Lower

Cornpartrnent Cooling fan in standby using the alarm response and 0-SO-30-5,

Lower Compartment Cooling.

C. PZR Enclosure may heat up and cause PZR Safety Valves to leak; Isolate the

tripped cooler TCV and bypass the in-service Lower Compartment Cooler TCVs to

increase cooling to the PZR Enclosure in accordance with the alarm response

procedures.

D~ PZR Enclosure may heat up and cause PZR Safety Valves to leak; Start any Lower

Compartment Cooling fan in standby using the alarm response and 0-SO-30-5,

Lower Compartment Cooling.

A. Incorrect. Containment DP is a concern for other ventilation systems such as purge,

or an event such as a LOCA.

B. Incorrect. Containment DP is a concern for other ventilation systems such as purge,

or an event such as a LOCA. Action to start a fan is correct

C. Incorrect. rcv for the tripped fan is allowed to go open. Not isolated

D. Correct.

Three fans should be running. With one in standby, itmay be started.

Monday, March 12, 20072:35:29 PM

48

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Abilityto (a) predict the impacts of the following malfunctionsor operations on the CCS: and (b) based on those predictions. use

procedures to correct, control. or mitigate the consequences of those malfunctions oroperations: Fan motor thermal

overload/hlgh-speed operation.

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

87

8RO 3.0

AR-M6E. E5; 0-80-30-5

Proposed references to be provided to applicants during examination:

None

Leaming Objective:

Question Source:

Question History:

Question Cognitive Level:

10 CFR Part 55 Content:

OPT200.CONTCOOLING Objective 7

New

Higher

43.5

(

Comments:

Source:

Cognitive Level:

Job Position:

Date:

NEW

HIGH ER

4/2007

Source [fBank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday, March 12, 20072:35:29 PM

49

(

OPT200.CONTCOOLlNG

Revision 0

Page 4 of 14

6

For the containment cooling system, describe the differences between unit's design,

control board layouts, and instrumentation. (KIA 2.2.3,2.2.4)

7

Explain and apply the containment cooling systems' precautions and limitations.

(KIA 2.1.32)

8

Explain and apply the Alarm Response Procedures associated with containment cooling

systems. (KIA 2.4.46, 2.4.47, 2.4.48, 2.4.50)

c

X.

LESSON BODY:

17. Given a plant situation, demonstrate the ability to monitor the

automatic operation of the Containment Cooling system.

(KlA022A3)

a.

Initiation of safeguards mode operation

18. Given a plant situation for the Contairunent Cooling system,

demonstrate the ability to monitor and, as appropriate, perform

manual operation of the system in the control room. (KIA 022A4)

a. Cntmt Cooling fans

b.

Dampers in the Cntmt Cooling system

c.

Valves in the Cntmt Cooling system

d. Containment readings of temperature, pressure, and humidity

system

B.

Discuss, as applicable, the following procedures.

1.

O-SO-30-4

ERCW must be established to the Unit I lA-A and IB-B lower

compartment coolers before starting the IA-A and IB-B

upper compartment coolers since the upper and lower

coolers share the same ERCW headers

Normal aligrunent has 3 coolers operating and one either in A-

Auto or A-P Auto

2.

O-SO-30-S

a.

Major Precautions and Limitations

Before starting/stopping LCCUs, the thermal effect on the

pressurizer safety valves must be evaluated. PZR

enclosure heatup rates> SOF should be avoided. Stop

the enclosure heatup and cooldown the enclosure

If a TCV fails open, place its TIC to manual and adjust to

other TCVs

b.

Normal Operation

Normal aligrunent has 3 coolers operating and one in A-

Auto

3.

O-SO-30-6

a.

Major Precautions and Limitations

Starting/Stopping a CRDM coolers/dampers may impact

the operability of rod position indication.

Changes in dip across the containment divider where the

lower compartment is higher than the upper

compartment could cause the ice condenser doors to

open.

CRDM units should be in operation before exceeding

3S0oF

OPT200.CONTCOOLING

Revision 0

Page 12 of 14

INSTRUCTOR NOTES

Objective 17

Drawings, Student Research

Objective 18

Drawings, Student Research

Discuss when presenting

system

Objective 7

Unit 1 Specific

Objective 7

c

Source

Setpoint

33

(E-5)

SER 633

N/A

Control switch in "normal-alter-close" and

the breaker is open on any of the 480V

motor power supply.

MOTOR

TRIPOUT

PANEL 1-M-9

(

c

Probable

Causes

Corrective

Actions

References

1.

Motor overload or fault.

2.

Process control trip.

[1] DETERMINE affected equipment by checking for white indicating

light on 1-M-9.

[2] DISPATCH operator to affected equipment to determine cause of

motor tripout.

[3] PLACE additional equipment in service in accordance with

applicable system instructions.

45B655-06E-0, 45N657-20

1-AR-M6-E

Page 36 of 40

1

Rev. 18

(

0-SO-30-5

LOWER COMPARTMENT

Rev: 30

0

COOLING UNITS

Page 5 of 46

3.0

PRECAUTIONS AND LIMITATIONS

A. Failureto observe all posted radiation control requirements may result in

unnecessary radiation absorbed dose.

B. The cooling unit inlets shall be free of debris before placing the system in

service.

C. Before starting or stopping the Lower Compartment Coolers. the thermal effect

on the pressurizer safety valves must be evaluated . Rapid heatup rates have

a definite affect on pressurizer safety valve leakage .

D. The pressurizer enclosure temperature must be monitored prior to and after

starting or stopping the Lower Compartment Coolers (Comp uter Point

T1001A). The US shall be notified of any drastic change in the pressurizer

enclosure temperature.

E, Unif1 and Unit 2 shall have at least 3 Lower Compartment Coolers operating

in modes 1-4 in order to provide adequate air ftow to the PZR enclosure, Prior

to Mode 4 entry, 4 LCCUs shall be running with two of the TICs in manual and

two in automatic for a smooth startup. (Automatic opera tion of TIC's not

required if performing section 8.3)

F. Each LCCU temperature control valve (TCV) will go open when its associated

fan is stopped,

G. Section 5,2, LCCU Start-up Following Mode 5 Outage, should be performed to

prevent PZR safety valve leakage due to PZR enclosure temperature

changes. This section should be completed prior to RCS pressure exceed ing

1500 psig ,

H. Controlling the rate of PZR enclosure heatup is more important in preventing

safety valves from leaking than the actual temperature limit (110°F),

I. When> 2000 psig RCS pressure, it is imperative to minimize PZR enclosure

heatup. A PZR enclosure heatup rate as low as 5°F/hour should be avoided; if

it occurs, it must be evaluated for cause and effect. A 1O°F/hour heatup

requires immediate corrective action to prevent safeties from leaking. Merely

slowing down the heatup will not prevent the safeties from leaking, rather, the

heatup must be stopped and the PZR enclosure cooled down rapidly.

(

SON

0-SO-30-5

LOWER COMPARTMENT

Rev: 30

0

COOLING UNITS

Page 6 of 46

3.0

PRECAUTIONS AND LIMITATIONS (Continued)

J. When changing RCS pressure (during startup) or any time changes are made

to the ventilation system (either to air flow or ERCW flow) average lower

containment air temperature (computer point U0983) and PZR enclosure

temperature (computer point T1001A) should be trended and monitored.

K. If a fan is lost, try to return it to service as soon as possible. If no other fan is

available for service, and the in-service fans' TCVs are not full open, then their

TIC setpoints should be lowered to get more cooling.

L. The LCCU TICs control within a band +/- 5°F of setpoint. Being proportional

only controllers , their valves will be full open when lower containment

temperature is 5°F above setpoint and full closed when the temperature is 5°F

bel ~w setpoint.

M. If a TCV fails open, its TIC should be placed in manual and the valve should

be closed to a position slightly more open than the other in-service fans' TCVs.

N. For small changes in PZR enclosure temperature, it is preferable to adjust the

setpoint of the TCV for the cooler supplying most airflow to the PZR enclosure,

rather than starting or stopping fans. Due to the physical layout of the fans

and their duct work, LCCs A-A or B-B supply most of the air to the PZR

enclosure. Appendix A provides a simplified schematic plan of the coolers'

layout.

O. Unit 1 Only: The 1A-A LCCU is a more effective cooler and should be one of

the coolers placed in service to ensure pressurizer enclosure temperature is

maintained. LCCU's 1B-B, 1C-A and 1D-B when operated together may not

be capable of providing sufficient cooling to the pressurizer enclosu re.

(SQ980504PER)

(

0-SO-3 0-5

LOWER COMPARTMENT

Rev: 30

0

COOLING UNITS

Page 13 of 46

Unit.

_

5.2

LCCU Start-up Following Mode 5 Outage (Continued)

Date

_

(

NOTE

The setpoint and/or position listed below may be varied based on

recommendations from Engineering.

[5]

ENSURE controllers for LCCU TCVs are set as follows:

LCCU

CONTROLLER

POSITION

SETPOINT

INITIALS

A-A

TIC-67-84

Manual

0% (Full Open) or

1st

CV

Eng recomm.

B-B

TIC-67-100

Manual

0% (Full Open) or

1st

CV

Eng recomm.

C-A

-

TIC-67-92

AUTO

110' F or

1st

CV

Engrecomm.

D-S

TIC-67-108

AUTO

110'F or

1st

CV

Engrecomm.

CAUTION

NOTE

The pressurizer enclosure temperature (computer point T1001A)

must be monitored and the US notified if temperature increases.

The temperature should be trended on a recorder or an ICS

graphic display so that trends may be readily seen.

Pressurizer Enclosure temperature fluctuations should be minimized

to prevent Pressurizer Safety Valve leakage.

[6]

MONITOR Pressurizer enclosure temperature .

AND

NOTIFY US of temperature changes.

[7]

IF PZR cubicle temperature reaches >110' F, THEN

NOTIFY the US immediately that RCS pressurization

should be stopped.

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

29.025 A2.01 001

Given the following plant conditions:

Unit 1 is at 100% power.

The following alarm is received:

1-AR-M6E, C7, FLOOR COOLANT DP LO

Which ONE (1) of the following describes the potential effect of this failure , and the

action required to mitigate the effect?

M 1-AR-M6E, 86, FLOOR COOLANT TEMP HI, will alarm; Ensure at least 1 glycol

floor circulation pump is running in accordance with the alarm response procedure.

8. 1-AR-M6E, 86, FLOOR COOLANT TEMP HI, will alarm; 8wap Glycol Circ Pumps

and chiller packages if required in accordance with 1-80-61 -1, Ice Condenser.

C. 1-AR-M6E, 87, FLOOR COOLANT TEMP LO, will alarm; Ensure at least 1 glycol

floor circulation pump is running in accordance with the alarm response procedure.

D. 1-AR-M6E, ~7, FLOOR COOLANT TEMP LO, will alarm; Swap Glycol Circ Pumps

and chiller packages if required in accordance with 1-80-61-1 , Ice Condenser.

A. Correct. If DP is low, then the floor cooling pump must have tripped, and

temperature will rise

B. Incorrect. Would swap Glycol Pumps and chillers if the Glycol pump tripped.

C. Incorrect. Wrong direction on temperature. Action is correct.

D. Incorrect. Wrong direction on temperature Action is incorrect. Would be taken if a

Glycol Circ Pump tripped

Monday, March 12, 20072:35:29 PM

54

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Ability to (a) predict the impacts of the following malfunctions or operations on the ice condenser system; correct. control, or

mitigate the consequences of those malfunctions or operations: Trip of glycolcirculation pumps.

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:'

88

SRO 2.7

AR-M6E, C7, B6

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

Question Cognitive Level:

10 CFR Part 55 Content:

Comments:

OPT200.ICE Objective 5.d

New

Higher

43.5

(

Source:

Cognitive Level:

Job Position:

Date:

NEW

HIGHER

4/2007

Source IfBank:

Difficulty:

Plant:

Last 2 NRC"!:

SEQUOYAH

NO

Monday, March 12, 20072:35:29 PM

55

(

""

(

OPT200.lCE

Rev. 2

Page 4 of 56

V.

TRAINING OBJECTIVES (Cont'd):

B. Learning Objectives (Conl'd):

5.

Describe the operation of the Ice Condenser system:

a.

Precautions and limitations

b.

Major steps performed while placing the Ice Condenser system in service

c.

Alarms and alarm response

d.

How a component failure will affect system operation

e.

How a support system failure will affect Ice Condenser system operation

f.

How a instrument failure will affect system operation

6.

Describe the administrative controls and limits for the Ice Condenser system:

a.

State Tech Specs/TRM LCOs that govern the Ice Condenser

b.

State the :sI hour action limit TS LCOs

c.

Given the conditions/status of the Ice Condenser system components and the

appropriate sections of the Tech Spec, determine if operability requirements

are met and what actions are required

7. Discuss related Industry Events

SQN, Licensee Event Report, 327-92007, Numerous ice condenser lower

doors found inoperable. 27 of 48 lower ice condenser doors were found

inoperable on U-2.

SQN, Licensee Event Report, 327-92023, 12/31/1992, Low ice condenser

weights result in operation outside ofdesign basis.

VI.

TRAINING AIDS:

A.

Classroom Computer and Local Area Network (LAN) Access

B.

Computer projector

C.

Simulator (if available)

(

Source

Setpoint

21

(C-7)

SER 621

1-PDIS*61*128

Probable

Causes

26 psid decreasing

1.

Glycol floor circulation pump trip .

2.

Pipe rupture.

3.

Valve misalignment.

PDlS-61-128

FLOOR COOLANT

6PLO

(

Corrective

Actions

References

[1]

IF containment isolation phase A is not actuated, THEN

ENSURE 11-FCV-61-96], 11-FCV-61-971,

11-FCV-61-110] and [1-FCV-61-1221 are OPEN

on paneI1-M-10 .

[2]

DISPATCH operator to Reverse Osmosis room on elevation 734 in

Au xiliary Building to perform the following:

[a] ENSURE at least one glycol floor circulation pump is running.

[b] ENSURE proper valve alignment in accordance with

0-80-61-1 , fee Condenser System.

[c] IF [1-TCV-61-71] has failed, THEN

OPERATE [1-61-751] as necessary to regulate proper t,p

(ref. 0-80-61-1 section 8.9)..

[3] INITIATE Work Order as necessary.

45B655-06E-0

1-AR-M6-E

Page 24 of 40

1

Rev. 18

(

Source

Setpoint

13

(8-6)

SER 613

1-TIS*61-99N B

20' F increasing

TS*61-99A/8

FLOOR COOLANT

TEMP HI

(,

(

Probable

Causes

Corrective

Actions

NOTE

References

1.

Low glycol flow resulting from pump failure or failure of

temperature regulator.

2.

Faulty floor coils.

3.

Glycol chiller failure.

[1] VERIFY Glycol Circ Pumps NOT tripped (1-M-23B).

[a] IF glycol circ pump tripped, THEN

RESTART pump or swap to alternate pump per 0-SO-61-1.

[2]

CHECK ice condenser temperature recorder for increasing

temperatures (1-M-9).

[a] IF glycol temperature is increasing, THEN

DISPATCH AUO to ensure proper chiller operation per

0-SO-61-1.

[3]

DISPATCH operator to the old Reverse Osmosis room on

elevation 734 in the Auxiliary Building to perform the following:

[a] CHECK glycol floor coolant temperature by observing

[1-TI-61-90] in RO Room on el 734.

[b] ENSURE at least one glycol floor circulation pump running.

1-TCV-61-71 setpoint is 13°F.

[c] IF [1-TCV-61*71] not operating properly, THEN

OPERATE manual bypass valve [1-61-751] as necessary

to bring temperature within limits (ref. 0-S0-61-1 section 8.9).

[4]

INITIATE Work Order as necessary.

45B655-06E-0, 47W600-87, 47W610-61-3

1-AR-M6-E

Page 15 of 40

1

Rev. 18

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

45.039 G2.1.14 001

(

Given the following plant conditions:

" ..

Main Steam Line warmup is in progress in accordance with 1-S0-1-1, Main

Steam System, section 8.4.

-

Verification of Steam Trap Level Panel status lights indicates all of the status

lights lit.

Which ONE (1) of the following describes the status of the steam line warmup, and the

appropriate response?

A'! Indication of water exists in the Main Steam Lines. Notify lSI to determine the water

level in the steam lines.

B. Indication of water exists in the Main Steam Lines. Stop the warmup and re-close

any open MSIV Bypass valves. Notify Chemistry to sample the Main Steam Lines.

C. Indication exists that Main Steam Lines are clear of water. Notify lSI and System

Engineering that the warmup will continue.

D. Indication exists that Main Steam Lines are clear of water. Notify Chemistry to

sample Main-Steam Lines prior to continuing the warmup.

(

A. Correct. Switches lit means water in steam lines

B. Incorrect. Would not close MSIV Bypass valves if water existed. Would try to clear

the lines

C. Incorrect.

Indication is water in lines, not free of water

D. Incorrect. Lights on indicates water in the lines

If indication of water exists per the level switches, then notify lSI

Monday, March 12, 2007 2:35:31 PM

85

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Conduct of Operations: Knowledge of system status criteria which require the notification of plant personnel.

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

89

SRO 3.3

1-S0-1-1 , section 8.4 step 9

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

Question Cognitive Level:

10 CFR Part 55 Content:

Comments:

OPT200.MS, B.4.h

New

Higher

43.5

(

Source:

Cognitive Level:

Job Position:

Date:

NEW

HIGHER

4/2007

Source IfBank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday, March 12. 20072:35:31 PM

86

(

V.

TRAINING OBJECTIVES (continued)

4.

Describe the following features for each major component in the Main Steam

System as described in this lesson.

a. Location

b. Power supply (include control power as applicable)

c. Support equipment and systems

d. Normal operating parameters

e. Component operation

f. Controls

g. Interlocks (including setpoints)

h. Instrumentation and Indications

I.

Protective features (including setpoints)

J.

Failure modes

k. Unit differences

1.

Types of accidents for which the Main Steam components are designed

m. Location of controls and indications associated with the Main Steam in

the control room and auxiliary control room.

OPT200.MS

Rev. 3

Page 4 of 54

(

MAIN STEAM SYSTEM

1-50-1-1

Rev: 23

1

Page 51 of 82

Date.

_

8.4

Placing Main Steam Header In Service ~ 400'F. (Continued)

CAUTION

Main steam piping heat-up rate is limited to less than

200'F per hour.

[7]

MONITOR main steam temperature indications on the following

computer log points while continuing with this instruction:

COMPUTER LOG POINT

MAIN STEAM LINE

./

1T2300A

1

0

1T2301A

2

0

1T2302A

3

0

-

1T2303A

4

0

[8]

PERFORM Section 8.2 of this procedure to place by-pass

orifices in service AND

RETURN to step [9].

[9]

RECORD status of lights on Steam Trap Level Panel

(el. 685, T3-K near Air Dryers):

Level Switch

ON --- OFF

INITIALS

1-XI-206

0

1-XI-207

0

1-XI-208

0

1-XI-209

0

(

c.

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

70. 078 G2.4.6 001

Given the following plant conditions:

-

Unit 1 is in MODE 3 following a shutdown.

-

A loss of Auxiliary Air occurred.

-

Auxil iary Feedwater was aligned as directed by AOP-M.02 , "Loss Of

Control Air" to support current plant operations.

-

EA-3-4, "Local Alignment Of TD AFW LCV Backup Air Supply" has been

implemented

Which ONE (1) of the following describes the actions necessary to RESTORE Auxiliary

Feedwater to normal following the return of the plant air systems to normal?

MUsing EA-3-4, "LOCAL ALIGNMENT OF TD AFW LCV BACKUP AIR SUPPLY",

verify TO AFW LCVs remain closed while isolating back-up air from the back up air

supply bottles . Bottle pressure is required to be maintained at a minimum of 800

psig.

B. Using EA-3-4, "LOCAL ALIGNMENT OF TO AFW LCV BACKUP AIR SUPPLY""

verify TD AFW LCVs remain open while isolating back-up air supply bottles. Bottle

pressure is required to be maintained at a minimum of 800 psig.

C. Using M.02, Loss of Control Air, Section 2.1 for Loss of Auxiliary Air, verify TD AFW

LCVs are closed; maintain back-up air aligned from the back-up air supply until at

least one bottle is less than 800 psig.

D. Using M.02, Loss of Control Air, Section 2.1 for Loss of Auxiliary Air, verify TD AFW

LCVs are open; maintain back-up air aligned from the back-up air supply until at

least one bottle is less than 800 psig.

A is correct; EA-3-4, "LOCAL ALIGNMENT OF TO AFW LCV BACKUP AIR SUPPL Y"

Section 4.3

B is incorrect; TO LCVs are normally closed; backup air is isolated per EA-3.4 Section

4.3, Placing TO AFW LCV Backup Air Supply in Standby.

C is incorrect; Wrong procedure usage. Also incorrect application of the requirement

for bottle pressure

o is incorrect; Wrong procedure usage. Incorrect application ofrequirement for bottle

pressure, and valves would be closed

Monday, March 12. 2007 2:35:34 PM

134

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Emergency Procedures I Plan Knowledgesymptom based EOP mitigationstrategies.

c

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

90

SRO 4.0

M.02, EA 3-4

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPL271 C424, Obj. 804

Question Source:

Bank

Question History:

Sequoyah AOP-M.02B2

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41.10

(

Comments:

Source:

Cognitive Level:

Job Position:

Date:

BANK

LOWER

412007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH BANK

SEQUOYAH

NO

Monday, March 12, 2007 2:35:34 PM

135

c

(

OPL271AOP-M.02

Revision 0

Page 3 of 25

I.

PROGRAM:

OPERATOR TRAINING - LICENSED

II.

COURSE:

LICENSE TRAINING

III.

LESSON TITLE: AOP-M.02 LOSS OF CONTROL AIR

IV.

LENGTH OF LESSON/COURSE:

1.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (s)

V.

TRAINING OBJECTIVES:

A. Term inal Objective:

Upon completion of License Training. the participant shall be able to demonstrate or

explain. using classroom evaluations and/or simulator scenarios. the requirements of

AOP-M .02, LOSS OF CONTROL AIR.

B. Enabling Objectives

Obiectives

O.

Demonstrate an understanding of NUREG 1122 knowledge's and abilities

associated with Loss of Control Air events that are rated ~ 2.5 during Initial

License Training and ~ 3,0 during License Operator Requalification Training for

the appropriate position as identified in Appendix A

1.

State the purpose/goal of AOP-M.02.

2.

Describe the AOP-M.02 entry conditions.

a. Describe the setpoints, interlocks. and automatic actions associated with

AOP-M.02 entry conditions.

b. Describe the ARP requirements associated with AOP-M.02 entry

conditions.

c. Interpret. prioritize. and verify associated alarms are consistent with AOP-

M.02 entry conditions.

d. Describe the Administrative and Tech Spec conditions resulting from a

Loss of Control Air.

3.

Describe the initial operator response to stabilize the plant upon entry into AOP-

M.02.

4.

Upon entry into AOP-M.02. diagnose the applicable condition and transition to the

appropriate procedural section for response.

5.

Summarize the mitigating strategy for the condition that Initiated entry into AOP-

M.02.

6.

Describe the bases for all limits. notes, cautions, and steps of AOP-M.02.

7.

Describe the conditions and reason for transitions within this procedure and

(

OPL271AOP-M.02

Revision 0

Page 4 of 25

Obiectives

transitions to other procedures .

8.

Given a set of initial plant conditions use AOP-M .02 to correctly:

a.

Recognize entry conditions

b.

Identify required actions

c.

Respond to Contingencies

d.

Observe and Interpret Cautions and Notes

9.

Describe the Tech Spec and TRM actions applicable during the performance of

AOP-M.02.

10.

Apply GFE and system response concepts to the abnormal condition

- prior to, during and after the abnormal condition .

LOSS OF CONTROL AIR

(

SON

I

AOP-M.02

Rev. 12

I STEP I

ACTION/EXPECTED RESPONSE

2.1

Loss of Auxiliary Ai r (cont'd)

7.

MONITOR S /G levels STABLE and

CONTROLLED.

RESPONSE NOT OBTAINED

DISPATCH an operator to establish control

of any AFW LCV affected by loss of air

using the following:

EA-3-4, Local Alignment ofTDAFW LCV

Backup Air Supply

EA-3-10, MD AFW LCV Failures

NOTE:

S/~PORVs 1 and 4 can be operated manually using reach rods in 480 Volt SD

Board Room.

(

8.

MONITOR RCS temperature STABLE and

CONTROLLED.

9.

MONITOR RCS pressure STABLE and

CONTROLLED.

IF RCS temperature is rising,

THEN

DISPATCH an operator to control any S /G

PORVs affected by loss of air USING EA-1-2,

Local Control of SIG PORVs.

IF pressurizer sprays are inoperable,

THEN

a.

ENSURE letdown IN SERVICE .

b.

INITIATE auxiliary spray for pressurizer

pressure control USING EA-62-4,

Establishing Auxiliary Spray.

Page 7 of 59

(

SON

EA*3-4

LOCAL ALIGNMENT OF TO AFW LCV

Rev. 4

0

BACKUP AIR SUPPLY

Page 5 of6

4.3

Placin g TD AFW LeV Backup Air Supply in Standby

1.

SELECT the unit for local alignment of TOAFW LCVs.

Unit 1__

Unit2 __

2.

OBTAIN hand held lighting and radio.

0

3.

IF performing this procedure during loss of all AC power (ECA-D.D), THEN

OBTAIN the following keys:

[glass-faced box in Shift Manager's Office)

(

V ital Area key

Pr~ te ct e d Area key.

4.

WHEN directed by UO,

THEN

CLOSE the following backup air supply isolation valves:

[Aux Bldg, elev 714, Auxiliary Building General Supply Fan Room)

CLOSED

VALVE

DESCRIPTION

.,J

ISV-32-1950E

Isolation valve for LCV-3-172

0

ISV*32-1969E

Isolation valve for LCV-3*173

0

ISV-32-1866E

Isolation valve for LCV-3-175

0

ISV-32-1974E

Isolation valve fo r LCV-3*1 74

0

5.

CHECK the pressure in the 4 backup air supply bottles.

IF any high pressure air bottle supply pressure less than 800 psig,

THEN

NOTIFY UO.

6.

GO TO Section 4.1, step in effect.

--_....

END OF TEXT

o

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

1. 001 A2 .12 00I

Given the following plant conditions:

A reactor startup is in progress using control rods.

During the startup, it is determined that the RCS boron concentration used in

the ECP calculation was 1100 ppm.

Actua l RCS boron concentration is 1000 ppm.

Which ONE (1) of the following describe the correct action?

The actua l critical rod position will be.......

A. LOWER than estimated. Trip the reactor, enter E-O, and initiate emergency

boration until required boron concentration is reached.

B. HIGHER than estimated. Trip the reactor, enter E-O, and initiate emergency

boration until required boron concentration is reached.

C:-' LOWER than estimated. Insert Control Banks and recalculate the ECP prior to

re-initiating the startup if it becomes apparent that criticality will occur >1000 pcm

below the Estimated Critical Rod Position.

D. HIGHER than estimated. Insert Control Banks and recalculate the ECP prior to

re-initiating the startup if it becomes apparent that criticality will occur >1000 pcm

above the Estimated Critical Rod Position.

A. Incorrect. Correct effect; wrong action

B. Incorrect. Incorrect effect

C. Correct. With actual boron less than expected, it will take less reactivity added to

reach criticality.

D. Incorrect. Incorrect effect. Consistent with required action

Monday, March 12, 2007 2:35:27 PM

1

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Abilityto (a) predict the impacts of the following malfunction or operations on the CRDS- and (b) based on those predictions, use

procedures to correct. control, or mitigate the consequences of those malfunctions or operations: Erroneous ECP calculation.

(

Question No.

Tier 2 Group 2

Importance Rating:

Technical Reference:

91

SRO 4.2

O-SI-NUC-OOO-001 .0 R5

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source :

Question History:

Question Cognitive Level:

10 CFR Part 55 Content:

Comments:

OPT200RDCNT, 8.5.a, OPL271GO-2, 8.4

New

Higher

43.5

(

Source:

Cognitive Level:

Job Position:

Date :

NEW

HIGHER

4/2007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC? :

SEQUOYAH

NO

Monday, March 12, 2007 2:35:27 PM

2

(

OPL271GO-2

Revision 0

Page 3 of 28

I. PROGRAM:

OPERATOR TRAINING - LICENSED

II. COURSE:

LICENSE TRAINING

III. LESSON TITLE:

0-GO-2, "PLANT STARTUP FROM HOT STANDBY TO REACTOR

CRITICAL"

IV. LENGTH OF LESSON/COURSE:

V. TRAINING OBJECTIVES:

A.

Terminal Objective:

-2 hour(s)

(

Upon completion of this lesson and others presented, the student shall demonstrate

an understanding of General Operating Instruction 0-GO-2, "Plant Startup From Hot

Standby to Reactor Critical" by successfully completing a written examination with a

score ~ 80 percent or greater.

B.

Enabling Objectives:

1.

State the reason for each prerequisite and precaution as discussed in this

lesson or provided in 0-GO-2.

2.

State the reasons inverse count rate ratio (1 /m) plot is used when taking the

reactor critical. (C.1)

3.

Briefly describe the method for performing ICRR when taking the reactor

critical (C.1).

4.

Discuss the rod position limits per 0-GO-2 for achieving criticality.

5.

State the reasons power is leveled-off at 10-3% power to take critical data.

-~---

(

V.

OPT200.RDCNT

Rev. 2

Page 4 of71

TRAINING OBJECTIVES (Cont'd):

B. Enabling Objectives (Conl'd):

5. Describe the operation of the Rod Control system as it relates to the following:

a. Precautions and limitations

b. Major steps performed while placing the Rod Control system in service

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect Rod Control system operation

f. How a instrument failure will affect system operation

6. Describe the administrative controls and limits for the Rod Control system as

explained in this lesson:

a. State Tech Specs/TRM LCOs that govern the Rod Control system

b. State the S I hour action limit TS LCOs

c. Given the conditions/status of the Rod Control system components and the

appropriate sections of the Tech Spec, determine if operability requirements

are met and what actions are required

7. Discuss related Industry Events:

a. PER 72943, Sequoyah Unit I Rod Drop

b. NSD-TB-92-05-RO, Thermal Lockup of Control Rods during Cooldown

c. NRC Bltn 96-0 I, Rods Fail to Fully Insert Following Scrams

VI.

TRAINING AIDS :

A.

Classroom Computer and Local Area Network (LAN) Access

B.

Computer projector

C.

Simulator (if available)

(

ESTIMATED CRITICAL CONDITIONS

0-81-NUC-000*001 .0

0

Rev 5

Page 36 of 42

APPENDIXC

Page 9 of 10

MONITORING THE APPROACH TO CRITICALITY

DATA SHEET C-2

UNIT CONDITIONS AT CRITICALITY

______

ppm

OF

[1]

[2]

[3]

[4]

[5]

DatelTime at criticality.

Core Average Temperature.

Control Bank Position:

Bank 0 at

steps

Bank C at

steps

Critical boron concentration.

ACCEPTANCE CRITERIA

DYes DNa

B.

Actual critical rod position is further inserted than the

negative MTC withdrawal limits ofTl-28.

DYes

DNa

DNA

C.

Actual critical rod position is within the 1000 pcm

limits.

~

Actual critical rod position is further withdrawn than

the ZPIL of TS 3.1 .3.6.

(

DYes DNa

D Yes ONt

I

DYes DNa

E.

Actual critical rod position is within the 750 pcm

limits.

D.

DatelTime of criticality are within the applicable 4

hour time span.

NOTE 9

The following acceptan ce criteria is an

administrative limit:

F.

Actual critical rod position is within the 500 pcm

limits.

[6]

ACCEPTANCE CRITERIA VERIFICATION

A.

If acceptance criteria SA was not satisfied, notify the

8M that the action requirement of LCO 3.1.3.6 must

be satisfied.

DYes

DNa

DNA

(

"'.

UNIT STARTUP FROM HOT STANDBY

0-GO-2

UnitO

TO REACTOR CRITICAL

Rev. 0026

Page 39 of 85

STARTUP No.

_

Unit

_

Date

_

5.2

Reactor Startup after a refueling outage (continued)

[58]

IF ALL of the following conditions are met:

rod motion was stopped prior to reaching

doubling range or criticality

Bank D rods are below fully withdrawn position

Unit Supervisor concurrence is obtained to resume,

THEN

RESUME rod withdrawal USING 0-SO-85-1

to seventh doubling range or criticality (whichever comes first).

0

(

c.

[59]

IF any of the following conditions exist:

-

-

critical conditions cannot be achieved

within the +/-1000 pcm allowable limits

critical conditions cannot be achieved above

rod insertion limit

reactor startup must be aborted for other reasons,

THEN

PERFORM the following to abort reactor startup: [C.11]

[59.1]

STOP rod withdrawal.

[59.2]

INITIATE insertion of control banks.

[59.3]

PERFORM Appendix D, Actions if Reactor Startup

Must Be Aborted.

[59.4]

DO NOT CONTINUE this section.

o

(

UNIT STARTUP FROM HOT STANDBY

0-GO-2

UnitO

TO REACTOR CRITICAL

Rev. 0026

Paqe 80 of 85

Appendix D

(Page 1 of 1)

ACTIONS IF REACTOR STARTUP MUST BE ABORTED

STARTUP No.

_

1.0

Operato r Actions

Unit

_

CAUTION

Date

_

If reactor trip is required, E-O should be performed instead of t his appendix.

[1]

ENSURE all control bank rods FULLY INSERTED

in accordance with 0-50-85-1.

[2]

LOG Mode 3 entry in narrative log.

[3]

VERIFY adequate shutdown margin in accordance with

0-§I-NU C-000-038.0.

(

Initials

[4]

DETERMINE and CORRECT cause of the discrepancy.

[5]

WHEN reactor startup is to resume,

THEN

PERFORM the following:

Time

Date

o

(

[5.1]

[5.2]

[5.3]

[504]

RECALCULATE estimated critical conditions in

accordance with 0-RT-NUC-000-003.0 (Startup

after refueling) or 0-SI-NUC-000-001.0 (Startup

after non-refueling outage).

DILUTE/BORATE in accordance with 0-80-62-7

to the estimated critical boron concentration. [C.12]

EQUALIZE boron concentration (within 50 ppm)

between reactor coolant loops and pressurizer

by operating pzr heaters and spray. [C.12]

RE-INITIATE 0-GO-2.

End of Document

(

UNIT STARTUP FROM HOT STANDBY

0-GO-2

Unit 0

TO REACTOR CRITICAL

Rev. 0026

Pace 63 of 85

5.3

STARTUP No.

Unit

_

Reactor Startup after a non-refueling outage (continued)

[58]

IF ALL of the following conditions are met:

rod motion was stopped prior to reaching

doubling range or criticality

Bank D rods are below fully withdrawn position

Date

_

(

[59]

Unit Supervisor concurrence is obtained to resume,

THEN

RESUME rod withdrawal USING 0-50-85-1

to seventh doubling range or criticality (whichever comes first).

0

IF any of the following conditions exist:

-

critical conditions cannot be achieved

within the +/-750 pcm termination band

critical conditions cannot be achieved

within the +/-1000 pcm allowable limits

critical conditions cannot be achieved above

rod insertion limit

reactor startup must be aborted for other reasons,

THEN

PERFORM the following to abort reactor startup: [C.11]

[59.1]

[59.2]

[59.3]

STOP rod withdrawal.

INITIATE insertion of control banks.

PERFORM Appendix D, Actions if Reactor Startup

Must Be Aborted.

o

[59.4]

DO NOT CONTINUE this section.

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

46. 041 A2.03 001

(

..

Given the following plant conditions:

-,

Unit 1 is at 12% power during a plant startup.

-

A loss of Essential and Non-Essential Control Air has occurred.

The crew is attempting to restore Control Air in accordance with AOP-M.02,

Loss of Control Air.

RCS temperature is 559°F and rising.

PZR level is 92% and rising.

Which ONE (1) of the following describes the effect on the unit the action required to

control RCS temperature?

Trip the reactor based on...

Ay loss of PZR level control.

Control RCS temperature by manual control of all 4 SG

PORVs in accordance with M.02, concurrently with E-O.

B.

loss of PZR level control. Control RCS temperature by manual control of SG #1

and #4 PORVs ONLY in accordance with M.02, concurrently with E-O.

(

c

C. loss of RCS femperature control. Control RCS temperature by manual control of all

4 SG PORVs in accordance with E-O.

D. loss of RCS temperature control. Control RCS temperature by manual control of

SG 1 and 4 PORVs in accordance with E-O.

A. Correct.

Trip based on loss of PZR level control, with trip occurring automatically at

92%.

SG PORV 1 and 4 have reached rods and may be controlled manually, but

procedure calls for all 4.

B. Incorrect. PORVs 1 and 4 have reach rods, but all 4 PORVs are used

C. Incorrect.

Loss of RCS temperature control is not the reason for the trip.

Subsequent temperature control is correct

D. Incorrect. PORVs 1 and 4 have reach rods, but all 4 PORVs are used. RCS

temperature control is not the reason for the trip, although at 12% power with steam

dumps open, it does cause problems

92% PZR level is trip criteria in M.02

Monday, March 12, 2007 2:35:31 PM

87

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Ability to (a) predict the impactsof the following malfunctions or operations on the 5 0S; and (b) based on those predictions or

mitigate the consequences of those malfunctions or operations: Loss of lAS.

(

Question No.

Tier 2 Group 2

Importance Rating:

Technical Reference:

92

SRO 3.1

M.02

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPL271AOP-M.02, B.8.b

New

Source If Bank:

Difficulty:

Plant:

Last2 NRC?:

(

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41.7,41.10

Comments:

Correct answer should be A. Would they perform B????

reach rods, but would they use a1l4?

Source:

NEW

Cognitive Level:

HIGHER

Job Position:

Date:

4/2007

Monday, March 12,20072:35:31 PM

I know that 1 and 4 SGs have

SEQUOYAH

NO

88

(

OPL271AOP-M.02

Revision 0

Page 3 of 25

I.

PROGRAM:

OPERATOR TRAINING - liCENSED

II.

COURSE:

liCENSE TRAINING

III.

LESSON TITLE: AOp*M.02 LOSS OF CONTROL AIR

IV.

LENGTH OF LESSON/COURSE:

1.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (s)

V.

TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or

explain, using classroom evaluations and/or simulator scenarios, the requirements of

AOP-M.02, LOSS OF CONTROL AIR.

B. Enabling Objectives

Objectives

O.

Demonstrate an understanding of NUREG 1122 knowledge's and abilities

associated with Loss of Control Air events that are rated a 2.5 during Initial

License Training and z 3.0 during License Operator Requalification Training for

the appropriate position as identified in Appendix A

1.

State the purpose/goal of AOP-M .02.

2.

Describe the AOP-M .02 entry conditions.

a. Describe the setpoints, interlocks, and automatic actions associated with

AOP-M.02 entry conditions.

b. Describe the ARP requirements associated with AOP-M.02 entry

conditions.

c. Interpret, prioritize, and verify associated alarms are consistent with AOP-

M.02 entry conditions.

d. Describe the Admin istrative and Tech Spec conditions resulting from a

Loss of Control Air.

3.

Describe the initial operator response to stabilize the plant upon entry into AOP-

M.02.

4.

Upon entry into AOP-M.02, diagnose the applicable condition and trans ition to the

appropriate procedural section for response.

5.

Summarize the mitigating strategy for the condition that initiated entry into AOP-

M.02.

6.

Describe the bases for all limits, notes, cautions, and steps of AOP-M.02.

7.

Describe the conditions and reason for transitions within this procedure and

(

OPL271AOP-M.02

Revision 0

Page 4 of 25

Obiectives

transitions to other procedures.

8.

Given a set of initial plant conditions use AOP-M.02 to correctly:

a.

Recognize entry conditions

b.

Identify required actions

c.

Respond to Contingencies

d.

Observe and Interpret Cautions and Notes

9.

Describe the Tech Spec and TRM actions applicable during the performance of

AOP-M.02.

10.

Apply GFE and system response concepts to the abnormal condition

- prior to, during and after the abnormal condition.

LOSS OF CONTROL AIR

(

I

AOP-M.02

Rev. 12

I STEP I

ACTION/EXPECTED RESPONSE

RESPONSE NOT OBTAINED

2.2

Loss of Nonessential Control Air in MODE 1, 2, or 3 (cont'd)

NOTES:

ReS cooldown will reduce pressurizer level by shrinking RCS.

A CCP is maintained running when possible for RCP seal injection flow.

(

18.

MONITOR pressurizer level

less than or equal to 70%.

IF pressurizer level rises to greater than 70%,

THEN

EVALUATE initiation of plant shutdown

USING the following procedures as applicable:

0-GO-5, Normal Power Operations

0-GO-6, Power Reduction from 30%

Reactor power to Hot Standby

0-GO-7, Unit Shutdown From Hot

Standby to Cold Shutdown

IF pressurizer level approaches 92%,

THEN

TRIP the reactor, and

GO TO E-O, Reactor Trip or Safety

Injection while continuing with this

Instruction.---.--

Page 18 of 59

c

-- --- -

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

47.045 G2.4.30 001

Given the following plant conditions:

A power increase was in progress on Unit 1.

Power was at 56% when a Main Generator Differential overcurrent relay

actuation occurred .

Which ONE (1) of the following describes the reportability requirements for this event?

Reference Provided

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

B. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months reports

C. 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report only

Dr 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months reports

A. Incorrect. Generator trip does not require a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months report

B. Incorrect. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months not required, but 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report is.

C. Incorrect. 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report also required

D. Correct. SPP-3.5 requires a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report for generator trip that

cause a reactor trip

Monday. March 12,20072:35:31 PM

89

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Emergency Procedures / Plan Knowledgeof whichevents related to system operations/status should be reported tooutside

agencies.

(

Question No.

Tier 2 Group 2

Importance Rating:

Technical Reference:

93

SRO 3.6

SPP-3.5

Proposed references to be provided to applicants during examination:

SPP-3.5, Appendix A

Learning Objective:

Question Source:

Question History:

OPL271C168 Objective 2

New '

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments:

(

Source:

Cognitive Level:

Job Position:

Date:

NEW

HIGHER

412007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday, March 12, 2007 2:35:31 PM

90

(

TVAN Standard

Regulatory Reporting Requirements

SPP-3.5

Programs and

Rev. 0017

Processes

Pace 16 of 63

Appendix A

(Page 1 of 11)

Reporting of Events or Conditions Affecting

licensed Nucle ar Power Plants

1.0

PURPOSE

This Appendix identifies reporting requirements; and instructions for determining

reportability. preparation, and transmittal of LERs; and notification to NRC for events

occurring at TVA's licensed nuclear plants.

2.0

SCOPE

TVA is required by §50.72 and §50.73 to promptly report various types of conditions or

events and provide written follow-up reports, as appropriate. This appendix provides

reporting guidance applicable to licensed power reactors.

NOTES

1)

Append ix B provides additional reporting criteria found in §Part 20, 30, 40, and 70 that

mqy be applicable to events involving byproduct, source or special nuclear material

possessed by the licensed nuclear plant. Site licensing and Site RadCon are

responsible for making the reportability determinations for §Part 20, 30, 40, or 70

events associated with their site. Corporate Licensing and Corporate RadChem are

responsible for making the reportability determinations for §Part 20, 30, 40, or 70

events associated with all other TVA licensed activities. Licensing is responsible for

developing (with input from affected organiza tions) and submitting the immediate

notification and written reports to NRC in accordance with §Part 20, 30,40, or 70

requirements . Reporting requirements for person nel exposure required by §Part 20 .

are contained in RCDP-4, "Personnel lnprocessing and Dosimetry Administrative

Processes."

2)

Appendix C contains the criteria for reporting if events or conditions affecting ISFSI.

TVA, as the general licensee of the ISFSI, is required by §72.216 to make initial and

written reports in accordance with §72.74 and §72.75. Operations is responsible for

making the reportability determinations for §72.74 and §72.75 reports. Operations is

responsible for making the immediate notification to NRC in accordance with §72.74.

Operations is responsible for making the immediate, 4-hour, and 24-hour notifications

to NRC in accordance with §72.75. Licensing is responsible for developing (with input

from affected organizations) and submitting the written reports required by §72.75.

3)

Reporting requirements for events or conditions affecting the physical protection of

the licensed nuclear plant specified in §73.71 are contained in SPP-1.3 "Plant Access

and Security." Responsibilities for reportability determinations and immediate

notification requirements are assigned to Site Nuclear Security and Corporate Nuclear

Security. Licensing is responsible for developing (with input from affected

organizations) and submitting the written reports required by §73.71.

(

TVAN Standard

Regulatory Reporting Requirements

SPP*3.5

Programs and

Rev. 0017

Processes

Paoe 17 of 63

Appendix A

(Page 2 of 11)

3.0

REQUIREMENTS

NOTES

1)

Internal managemen t notification requirements for plant events are found in

Appe ndix D. Operat ions and the Plant Manager (or Duty Plant Manager) are

responsible for making these internal management notifications.

2)

NRC NUREG-1022, Supplements and subsequent revisions should be used as

guidance for determining reportability of plant events pursuant to §50.72 and §50.73.

3.1

Immediate Notification - NRC

TVA is required by §50.72 to notify NRC immediately if certain types of events occur. This

appendix contains the types of events and the allotted time in which NRC must be notified.

(Refer to Form SPP-3.5-1). Operations is responsible for making the reportability

determinations for §50.72 and §50.73 reports. Operations is responsible for making the

irnrnedlate.notificatlon to NRC in accordance with §50.72.

Notification is via the Emergency Notification System. If the Emergency Notification System

is not operative, use a telephone, telegraph , mailgram, or facsimile.

NOTE

The NRC Event Notification Worksheet may be used in preparing for notifying the

NRC.

A.

The Immediate Notification Criteria of §50.72 is divided into 1-hour, 4-hour, and 8- hour

phone calls. Notify the NRC Operations Center within the applicable time limit for any

item which is identified in the Immediate Notification Criteria.

B.

The following criteria require 1-hour notification:

1.

(Technical Specifications) - Safety Limits as defined by the Technical

Specifications which have been violated.

2.

§50.72 (a)(l )(i) - The declaration of any of the Emergency classes specified in the

licensee's approved Emergency Plan.

NOTE

If it is discovered that a condition existed which met the Emergency Plan

criteria but no emergency was declared and the basis for the emergency class

no longer exists at the time of discovery, an ENS notification (and notification of

the Operations Duty Specialist), within one hour of discovery of the undeclared

(or misclassified) event, shall be made. However, actual declaration of the

emergency class is not necessary in these circumstances.

(

TVAN Standard

Regulatory Reporting Requirements

SPP-3.5

Programs and

Rev. 0017

Processes

Pace 18 of 63

Appendi x A

(Page 3 of 11)

3.1

Immediate Notification* NRC (continued)

3.

§50.72(b).(1>> - Any deviation from the plant's Technical Specifications authorized

pursuant to §50.54(x).

C.

The following criteria require 4-hour notification:

1.

§50.72(b)(2)(i) - The initiation of any nuclear plant shutdown required by the

plant's Technical Specifications.

2.

§50.72(b)(2)(iv)(A) - Any event that results or should have resulted in Emergency

Core Cooling System (EGGS) discharge into the reactor coolant system as a

result of a valid signal except when the actuation results from and is part of a

pre-planned sequence during testing or reactor operation.

3.

§50.72(b)(2)(iv)(B) - Any event or condition that results in actuation of the reactor

protection system (RPS) when the reactor is critical except when the actuation

results from and is part of a pre-planned sequence during testing or reactor

" operatlon.

4.

§50.72(b)(2)(xi) - Any event or situation, related to the health and safety of the

public or onsite personnel, or protection of the environment, for which a news

release is planned or notification to other government agencies has been or will

be made. Such an event may include an onsite fatality or inadvertent release of

radioactive contaminated materials.

D.

The following criteria require 8-hour notification:

NOTE

The non-emergency events specified below are only reportable if they occurred

within three years of the date of discovery.

1.

§50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the

nuclear power plant, including its principal safety barriers, being seriously

degraded.

2.

§50.72(b)(3)(ii)(B) - Any event or condition that results in the nuclear power plant

being in an unanalyzed condition that significantly degrades plant safety.

3.

§50.72(b)(3)(iv)(A) - Any event or condition that results in valid actuation of any of

the systems listed in paragraph (b)(3)(iv)(B) [see list below], except when the

actuation results from and is part of a pre-planned sequence during testing or

reactor operation.

a.

Reactor protection system (RPS) including: Reactor scram and reactor trip.

(

TVAN Standard

Regulatory Reporting Requirements

SPP-3.5

Programs and

Rev. 0017

Processes

Pace 19 of 63

Appendix A

(Page 4 of 11)

3.1

Immedi ate Notification* NRC (continued)

NOTE

Actuation of the RPS when the reactor is critical is also reportable

under §50.72(b)(2)(iv)(B) above.

b.

General containment isolation signals affecting containment isolation valves

in more than one system or multiple main steam isolation valves (MSIVs).

c.

Emergency core cooling systems (ECCS) for pressurized water reactors

(PWRs) including : High-head. intermediate-head, and low-head injection

systems and the low pressure injection function of residual (decay) heat

removal systems .

d.

ECCS for boiling water reactors (BWRs ) including: core spray systems;

high-pressure coolant injection system; low pressure injection function of the

residual heat removal system.

e.

BWR reactor core isolation cooting system.

f.

PWR auxiliary or emergency feedwater system.

g.

Containment heat removal and depressurization systems, including

containment spray and fan cooler systems.

h.

Emergency ac electrical power systems, including: Emergency diesel

generators (EDGs).

4.

§50.72(b)(3)(v) - Any event or condition that at the time of discovery could have

prevented the fulfillment of the safety function of structures or systems that are

needed to:

a.

Shut down the reactor and maintain it in a safe shutdown condition;

b.

Remove residual heat;

c.

Control the release of radioacti ve material; or

d.

Mitigate the consequences of an accident.

(

TVAN Standard

Regulatory Reporting Requirements

SPP-3.5

Programs and

Rev. 0017

Processes

Pane 20 of 63

Appendix A

(Page50f11)

3.1

Imme diate Notlficatlon

> NRC (continued)

NOTE

According to §50.72 (b)(3)(vi) events covered by §50.72(b)(3)(v) may include

one or more procedural errors, equipment failures, and/or discovery of design,

analysis, fabrication, construction, and/or procedural inadequacies. Howeve r,

individual component failures need not be reported pursuant this paragraph if

redundant equipment in the same system was operable and available to

perform the required safety function.

5.

§50.72(b)(3)(xii) - Any event requiring the transp ort of a radioactively

contaminated person to an offsite medical facility for treatment.

6.

§50.72(b)(3)(xiii) - Any event that results in a major loss of emergency

assessment capability, offsite response capability, or offsite communications

capability (e.g., significant portion of control room indication, emergency

notification system, or offsite notification system),

E.

Follow-up Notification (§50,72(c))

With respect to the telephone notifications made under paragraphs (a) and (b) [§50.72

(a) and §50.72 (b), respectively] of this section [§50.72]. in addition to making the

required initial notification, during the course of the event:

a.

Immediately report (i) any further degradation in the level of safety of the

plant or other worsening plant conditions including those that require the

declaration of the Emergency Classes, if such a declaration has not been

previously made; or

(1)

Any change from one Emergency Class to another, or

(2)

A termination of the Emergency Class.

b.

Immediately report (i) the results of ensuing evaluations or assessments of

plant conditions,

(1)

The effectiveness of response or protective measures taken, and

(2)

Information related to plant behavior that is not understood.

c.

Maintain an open, continuous communication channel with the NRC

Operations Center upon request by the NRC.

(

TVAN Standard

Regulatory Reporting Requirements

SPP-3.5

Programs and

Rev. DOH

Processes

Paqe 21 of 63

Appe ndix A

(Pag e60f11 )

3.2

Twenty-Four Hour Notification - NRC

Any violation of the requirement contained in specific operating license conditions, shall be

reported to NRC in accordance with the license condition.

3.3

Two-Day Notificati on - NRC

§50.9(b) - The NRC shall be notified of incomplete or inaccurate information which contains

significant implications for the public health and safety or common defense and security.

Notification shall be provided to the administrator of the appropriate regional office within

two working days of identifying the information. Licensing is responsible for determining

reportability (with input from affected organizations) and notifying NRC in accordance with

§50.9.

3.4

Sixty-Day Verba l Report

§50.73(a)(2)(iv)(A) requires that any event or condition that resulted in manual or automatic

actuation of the specified systems be reported as a Licensee Event Report (LER [Refer to

Appendix 'A, Section 3.5]). This CFR section also allows that in the case of an invalid

actuation, other than actuation of the reactor protection system when the reactor is critical,

an optional telephone notification may be placed to the NRC Operations Center within 60

days after discovery of the event instead of submitting a written LER.

A.

Verbal Report Required Content:

If the verbal notification option is selected (NUREG 1022, Revision 2, Section 3.2.6.,

"System Actuation"), instead of a LER, the verbal report:

1.

Is not considered an LER.

2.

Should identify that the report is being made under §50.73(a)(2)(iv)(A).

3.

Should provide the following infomnation:

a.

The specific train(s) and system(s) that were actuated.

b.

Whether each train actuation was complete or partial.

c.

Whether or not the system started and functioned successfully.

NOTE

Licensing will ensure that the information that is provided to NRC during the

Sixty-Day Verbal Report is verified in accordance with BP-213.

(

TVAN Standard

Regulatory Reporting Requirements

SPP-3.5

Programs and

Rev. 0017

Processes

Paqe 22 of 63

Appendix A

(Page 7 of 11)

3.4

Sixty*Day Verbal Report (continued)

B.

Verbal Report Development and Review

Licensing will:

1.

Develop (with input from responsible organization) the response (i.e., report

summary) to address the required input.

2.

Ensure that the reporting details are reviewed by MRC.

C.

Telephone Report Timeliness

Licensing will make the 60-day telephone report promptly after the PER for the invalid

actuation event is reviewed by MRC.

3.5

Written Report. NRC

A.

A repert on a Safety Limit Vioiation shall be submitted to the NRC, the NSRB, and the

Site Vice President if required by Technical Specifications.

B.

Any violation of the requirements contained in the Operating license conditions in lieu

of other reporting requirements requires a written follow-up report if specified in the

license.

C.

Reporting Radiation Injuries

1.

§140.6(a) requires, as promptly as possible, submittal of a written notice [e.g.,

report] in the event of:

a.

Bodily injury or property damage arising out of or in connection with the

possession or use of the radioactive materiai at the licensee's facility

[location): or

b.

In the course of transportation: or

c.

In the event any radiation exposure claim is made. (Refer to RCDP*9,

"Radiological and Chemistry Control Radiological Exposure Inquiries")

2.

The written notice shall contain particulars sufficient to identify the licensee and

reasonably obtainable information with respect to time, place, and circumstances

thereof, or the nature of the claim.

(

TVAN Standard

Regulatory Reporting Requirements

SPP-3.5

Programs and

Rev. 0017

Processes

Pace 23 of 63

Appendix A

(Page 8 of 11)

3.5

Written Report- NRC (continued)

D.

Licensee Event Reports

A written report shall be prepared in accordance with §50.73(a)(i) for items in the

60-day report criteria or Technical Specifications. The report shall be complete and

accurate in accordance with the methods outlined in this procedure. The completed

forms shall be submitted to the USNRC, Document Control Desk, Was hington, DC

20555. NUREG 1022, Revision 2, contains the instructions for completion of the LER

form. Licensing is responsible for developing (with input from affected organizations)

and submitting the written reports (or optional telephone reports [refer to Appendix A,

Section 3.4]) required by §50.73.

NOTE

Unless otherwise specified in the reporting criteria below, an event shall be

reported if it occurred within three years of the date of discovery regardless of the

plant mode or power level, and regardless of the significance of the structure,

system , or component that initiated the event.

E.

Report Criteria

1.

§50.73(a)(2)(i)(A) - The completion of any nuclear plant shutdown required by the

plant's Technical Specifications.

2.

§50.73(a)(2)(i)(B) - Any operation or condition which was prohibi ted by the plant's

Technical Specifications, except when:

a.

The Technical Specification is administrative in nature;

b.

The event consisted solely of a case of a late surveillance test where the

oversight was corrected, the test was performed, and the equipment was

found to be capable of performing its specified safety functions; or

c.

The Technical Specification was revised prior to discovery of the event

such that the operation or condition was no longer prohibited at the time of

discovery of the event.

3.

§50.73(a)(2)(i)(C) - Any deviation from the plant's Technical Specifications

authorized pursuant to §50.54(x).

4.

§50.73(a)(2)(ii)(A) - Any event or condition that resulted in the condition of the

nuclear power plant, including its principal safety barriers, being seriously

degraded.

5.

§50.73(a)(2)(ii)(B) - Any event or condition that resulted in the nuclear power plant

being in an unanalyzed condition that significantly degraded plant safety.

(

TVAN Standard

Regulatory Reporting Requirements

SPP*3.5

Programs and

Rev. 0017

Processes

Pane 24 of 63

Appendix A

(Page 9 of 11)

3.5

Written Report * NRC (continued)

6.

§50.73(a)(2)(iii) - Any natural phenomenon or other external condition that posed

an actual threat to the safety of the nuclear power plant or significantly hampered

site personnel in the performance of duties necessary for the safe operation of the

nuclear power plant.

7.

§50.73(a)(2)(iv)(A) - Any event or condition that resulted in manual or automatic

actuation of any of the systems listed in paragraph (a)(2)(iv)(B) [see list in item

no. 3.5E.8 below]. except when

a.

The actuation resulted from and was part of a pre-planned sequence during

testing or reactor operation; or

b.

The actuation was invalid and (i) occurred while the system was properly

removed from service or (ii) occurred after the safety function had been

already completed.

NOTE

In the case of an invalid actuation, other than actuation of the reactor

protection system (RPS) when the reactor is critical, a telephone

notification to the NRC Operations Center within 60 days after discovery of

the event may be provided instead of submitting a written LER

(§50.73(a>> . [Refer to Appendix A, Section 3.4]

8.

§50.73(a)(2)(iv)(B) - The systems to which the requirements to paragraph

(a)(2}(iv)(A) of this section apply are:

a.

Reactor protection system (RPS) including: reactor scram or reactor trip.

b.

General containment isolation signals affecting containment isolation valves

in more than one system or multiple main steam isolation valves (MSIVs) .

c.

Emergency core cooling systems (ECCS) for pressurized water reactors

(PWRs) including: high-head, intermediate-head, and low-head injection

systems and the low pressure injection function of residual (decay) heat

removal systems.

d.

ECCS for boiling water reactors (BWRs) including: core spray systems;

high-pressure coolant injection system; low pressure injection function of the

residual heat removal system.

e.

BWR reactor core isolation cooling system.

f.

PWR auxiliary or emergency feedwater system.

(

TVAN Standard

Regulatory Reporting Requirements

SPP-3.5

Programs and

Rev. 0017

Processes

Paoe 25 of 63

Appendix A

(Page 10 of11)

3.5

Written Report - NRC (continued)

g.

Containment heat removal and depressurization systems, including

containment spray and fan cooler systems.

h.

Emergency ac electrical power systems. inciuding: emergency diesel

generators (EDGs).

i.

Emergency service water systems that do not normally run and that serve as

ultimate heat sinks.

9.

§50.73(a)(2)(v) - Any event or conditi on that could have prevented the fulfillment

of the safety function of structures or systems that are needed to:

a.

b.

7C.

(

d.

Shut down the reactor and maintain it in a safe shutdown condition;

Remove residual heat;

Control the release of radioacti ve material; or

Mitigate the consequences of an accident.

(

NOTE

Events reported above may include one or more procedural errors,

equipment failures, and/or discovery of design, analysis, fabrication,

construction, and/or procedurai inadequacies. However, individual

component failures need not be reported pursuant to this criterion if

redundant equipment in the same system was operable and available to

perform the required safety functi on [§50.73(a)(2)(vi)].

10.

§50.73(a)(2)(vii) - Any event where a single cause or condition caused at least

one independent train or channe l to become inoperable in multiple systems or two

independent trains or channels to become inoperable in a single system designed

to:

a.

Shut down the reactor and maintain it in a safe shutdown condition ;

b.

Remove residual heat;

c.

Control the release of radioacti ve material; or

d.

Mitigate the consequences of an accident.

11.

§50.73(a)(2)(viii)(A) - Any airborne radioactivity release that, when averaged over

a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , resulted in airborne radionuclide concentrations in an

unrestricted area that exceeded 20 times the applicable concentration limits

specified in Appendix B to Part 20, table 2, column 1.

(

TVAN Standard

Regulatory Reporti ng Requ irements

SPP*3.5

Programs and

Rev. 0017

Processes

Paqe 26 of 63

Appendix A

(Page 11 of 11)

3.5

Written Report - NRC (continued)

12.

§50.73(a)(2)(viii)(B) - Any liquid effluent release that, when averaged over a time

period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , exceeds 20 times the applicable concentrations specified in

Appendix B to Part 20, table 2, column 2, at the point of entry into the receiving

waters (i.e., unrestricted area) for all radionuclides except tritium and dissolved

noble gases.

13.

§50.73(a)(2)( ix)(A) - Any event or condition that as a result of a single cause could

have prevented the fulfillment of a safety function for two or more trains or

channels in different systems that are needed to:

a.

Shut down the reactor and maintain it in a safe shutdown condition;

b.

Remove residual heat;

c.

Control the release of radioactive material; or

tt.

Mitigate the consequences of an accident.

NOTE

Events covered above may include cases of procedural error, equipment

failure, and/or discovery of a design, analysis, fabrication, construction,

and/or procedural inadequacy. However, licensees are not required to report

an event pursuant to this criterion if the event results from a shared

dependency among trains or channels that is a natural or expected

consequence of the approved plant design or normal and expected wear or

degradation [§50.73(a)(2)(ix)(B)].

14.

§50.73(a)(2)(x) - Any event that posed an actual threat to the safety of the nuclear

power plant or significantly hampered site personnel in the performance of duties

necessary for the safe operation of the nuclear power plant including fires, toxic

gas releases, or radioactive releases.

(

OPL271C168

Revision 8

Page 3 of 8

I. PROGRAM:

OPERATOR TRAINING - LICENSED

II. COURSE:

LICENSE TRAINING

III. LESSON TITLE:

REPORTING REQUIREMENTS

IV. LENGTH OF LESSON/COURSE:

4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (s)

V. TRAINING OBJECTIVES:

A.

Terminal Objective:

Upon completion of this lesson, the student shall be able to perform the actions

necessary to comply with regulatory and plant reporting requirements. The class

participant will be required to demonstrate their knowledge of this material by

successfully completing a written examination at the end of the program in which this

lesson, and others, are presented. The passing grade will be established by training

program procedures.

B.

Enabling Objectives:

1. Perform a plant response assessment using the O-TI-QXX-OOO-001 .0," Event

Critique, Post Trip Report, Equipment Root Cause, and Outage Milestone PER

Evaluation."

a. State the responsibilities of each crew member. [C.1]

b.

Conduct a plant response assessment.

2.

For a given condition, determine the regulatory reporting requirements using

appropriate reference material.

a.

List the tools available to the operator for determining regulatory reporting

requirements.

b.

Define the key terms used to determine regulatory reporting requirements.

c.

State the criteria requiring one hour notification of the NRC.

d. State the criteria requiring four hour notification of the NRC.

e.

State the criteria requiring eight hour notification of the NRC.

f.

State the criteria requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification of the NRC.

g.

State the criteria requiring 2 day notification of the NRC.

h.

State the criteria requiring a written report or LER to the NRC.

i.

State the criteria requiring a telephone notification may be made in lieu of a

LER to the NRC.

3.

For a given condition, determine plant management reporting requirements

using SPP-3.5.

4.

For a given PER, complete a reportability determination per SPP-3.1.

c

-

- - ---

-

QUESTIONS REPORT

forSEQUOYAH 2007 - NRC EXAM REV DRAFT

88. G2.1.7 001

Given the following plant conditions:

- A Normal Plant cooldown is in progress

- The following table is a plot of the cooldown:

TIME

0800 -

0815

0830

0845

0900

0915

0930

RCS TCOLD

54rF

530°F

520°F

505°F

498°F

478 °F

44rF

TIME

0945

1000

1015

1030

1045

1100

1115

RCS TCOLD

425 °F

395°F

382°F

364°F

340°F

320°F

220°F

Incorrect. 52 deg Fin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , greater than 50 required by the procedure, <

Correct. This is 103 degrees in one hour.

Incorrect. Limits were exceeded at 1000 because c/d rate was 103 deg F for

Which ONE (1) of the following describes the first time that the Technical Specification

RCS Cooldown rate limit was exceeded?

A. 0915

B. 0930

(

C~ 1000

D. 1115

A.

T.S.

B.

Incorrect. 31 degrees in 15 minutes and 51 degrees in thirty minutes which

would indicate exceeding 100 degrees per hour however for one hour the rate was 73

degrees.

C.

D.

that hour

Monday, March 12,20072:35:37 PM

169

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior. and

instrumentinterpretation .

(-.

Question No.

Tier 3 Group 1

Importance Rating:

Technical Reference:

94

SRO 4.4

TS 3.4.9.1 , 0-SI-SXX-068-127, PTLR

(

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPT200.RCS Objective 6.a

Question Source:

Bank

Question History:

Robinson 2007 NRC

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.2

Comments:

Source:

Cognitive Level:

Job Position:

Date:

BANK

HIGHER

SURVEILLANCE REQUIREMENTS

4/2007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

ROBINSON 2007 NRC

SEQUOYAH

NO

Monday, March 12, 20072:35:37 PM

170

(

V.

OPT200.RCS

Rev. 2

Page 4 of 50

TRAINING OBJECTIVES (Cont'd):

B. Enabling Objectives (Cont'd):

5.

Describe the operation of the RCS as it relates to the following:

a. Precautions and limitations

b. Major steps performed while placing the RCS in service

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect RCS operation

f. How a instrument failure will affect system operation

6.

Describe the administrative controls and limits for the RCS as explained in this

lesson:

a. State Tech SpecsffRM LCOs that govern the RCS

b. State the TS/TRM LCOs that have a::: I hour action statement

c. Interpret applicable Tech Specs/TRM LCOs

d. Identify Limitations of the ODCM

c-: Interpret applicable ODCM limitations

f. Given the conditions/status of the RCS components and the appropriate

sections of the Tech Spec, determine if operability requirements are met and

what actions are required

7.

Discuss related Industry Events

a. Crystal River 3; Void formed in Loop

b. Calvert Cliffs2; ReactorCoolantPiping wastage

c. Quad Cities; ReactorCoolantchemistry

d. Fort Calhoun; SpentFuel Pool Chemistry

VI.

TRAINING AIDS:

A.

Classroom Computer and Local Area Network (LAN) Access

B.

Computer projector

C.

Simulator (if available)

(

Administrative Topics

.. State Tech SpecsfTRM LCOs that govern the Reactor

Coolant System

3.3.3.7 Accident Monitoring Instrumentation

3.4.1

Reactor Coolant Loops and Coolant Circ

3.4.2

Safety Valves - Shutdown

3.4.3

Safety and Relief Valves - Operating

3.4.6.2 RCS: Operational Leakage

3.4.7

Chemistry

3.4.8

Specific Activity

3.4.9.1 PressurefTemperature Limits

5.4.1

Design Pressure and Temperature

5.4.2

Volume

E06

X.

LESSON BODY:

NOTE:

Point out to students the T/S information on page 29 in the System

Description, and ODCM 112.0 & 1/2.1

S. Refer to a copy ofSQN Technical Specifications for the details ofthe LCO,

applicability, action(s), surveillance(s) and basis for each (cont'd)

OPT200.RCS

Rev. 2

Page 38 of 50

(

REACTOR COOLANT SYSTEM

3/4.4.9 RCS PRESSURE AND TEMPERATURE (pm LIMITS

LIMITIN G CONDITION FOR OPERATION "

3.4.9.1 RCS pressure, RCS temperature, and ReS heatup and cooldown rates shall be maintained

within the Iimils specified in the PTLR.

APPLICABILITY: At all times.

ACTIONS:

a. With the requirements of the LCO not met in MODE 1, 2. 3. or 4, restore the parameter(s) to

within limits in 30 minutes and determine RCS is acceptable' for continued operation within 72

hours. With the required action above not mel, be in MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in

MODE 5, with RCS pressure < 500 psig, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. 'With the requirements of the LCO not met any time other than MODE 1,2, 3, "or 4, immediately

iniliate actionto restore parameter(s) to within fimits and, prior to entering MODE 4, determine

RCS is acceptable' for continued operation.

4.4.9.1.1 Verily" RCS pressure, RCS temperature, "and RCS heatup and cooldown rates are within

the timits specified in the PTLR every 30 minutes.

The determination that the ReS is acceptable for continuedoperation must be completed for any

entry into Action (a) or (b).

Only required to be performed during RCS healup and cooldown operations and RCS inservice

leak and hydrostatic testing.

"

.'

. .- ....

SEQUOYAH - UNIT 1

3/44-23

November 9, 2004

Amendment Nos. 12, 87. 157,294.297

(

rRESSURE TEMrERATURE LIMITS RErORT

MATERIAL PROPERTY BASIS

LIMITING IvtATERIAL: LOWER SHELL FORGING 04

LIMITING ART VALUES AT32 EFPY:

1/4T,2J6°F

3/4T, 186°F

2500

2250

2000

(

1750

0-

iii

fO:. 1500

"

~

J

'"'"" 1250

~

D.

"0"

~

!J. 1000

J

U"

0

750

500

250

o

O pe rlim V ersion:5.t R un:t 5680

Minimum

B oltup Temp =

50.

o

50

100

150

200

250

300

350

400

450

500

550

Moderator Temperature (Deg. F)

Figure 2-2

Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown

Rates u p to 100°Flhr) Applicable for the First 32 EFPY (wJl\\fargins for

Instrumentation Error of 10°Fand 60 psigl (Plotted Data provided on Table 2-2)

r

1. .1., ")M1.

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

1. G2.1.13 001INEWIILOWERIISRO/SEQUOYAH/412007INO

The card reader is broken that is prohibiting access to a vital area.

Who, by title, will authorize use of, and issue, the Vital Area Access key?

A. Unit Supervisor

B. Shift Manager

C. Operations Manager

D. Plant Manager

A. Incorrect.

US does not control the key

B. Correct. The Shift Manager holds the Vital Area Access key, and issues it as

needed

C. Incorrect. SM, see above

D. Incorrect. SM, see above.

Memory level item

Knowledge of facility requirements forcontrolling vitalI controlled access.

Question No.

e

Tier 3 Group 1

Importance Rating:

Technical Reference:

95

SRO 2.9

SPP-1 .3

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPL271 SECURITYObjective 4

New

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

43.5

Comments:

Monday, March 12, 20074:50:07 PM

1

c

OPL271SECURITY

Revision I

Page 3 of 16

I.

PROGRAM:

OPERATOR TRAINING - LICENSED

II.

COURSE:

LICENSE TRAINING

Ill.

LESSON TITLE:

SECURITY

IV.

LENGTH OF LESSON/COURSE:

V.

TRAINING OBJECTIVES:

8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months lecture/discussion

(

A. Terminal Objective:

Upon completion of this lesson and others presented, the student shall demonstrate an

understanding of the TVAN security process by successfully completing a written

examination with a score of 80 percent or greater.

B. Enabling Objectives

1.

State the purpose and function ofTVAN Security. (KIA 2.1.27)

2.

Discuss the lines ofauthority regarding Security and Operations during normal and

emergency conditions. (KIA 2.4.37)

3.

"Given site procedures, initiate off-hours site access request. (KIA 2.1.13)

4.

Discuss the site access control requirements at TVAN. (KIA 2.1.13)

5.

Discuss the TVAN Security facilities. (KIA 2.4.42)

6.

Discuss the Security communications systems. (KIA 2.4.43)

7.

For a given situation, and given site procedures, determine the appropriate response. (KIA

2.4.28)

a.

Sabotage (KIA 2.4.28)

b.

Security event/emergency (KIA 2.4.29)

8.

Discuss the Security response during a declared emergency at TVAN. (KIA 2.4.29)

9.

Discuss the conduct and control of a local or site evacuation. (KIA 2.4.29)

10.

For a given security situation, and given site procedures, determine the appropriate

notificat ions required. (KIA 2.4.30,2.4.41)

(

X.

LESSON BOD Y:

A.

Discuss the following topics.

State the purpose and function of TVAN Security. (KIA 2.1.27)

Discuss the lines of authority regarding Security and Operations during

normal and emergency conditions. (KIA 2.4.37)

Given site procedures, initiate off-hours site access request. (KIA

2.1.13)

Discuss the site access control requirements at TVAN. (KIA 2.1.13)

Discuss the TVAN Sec urity facilities. (KIA 2.4.42)

Discuss the Security communications systems. (KIA 2.4.43)

For a given situation, and given site procedures, determine the

appropriate response. (KIA 2.4.28)

a.

Sabotage (KIA 2.4.28)

b.

Security event/emergency (KIA 2.4.29)

Discuss the Security response during a declared emergency at TV AN.

(KI A 2.4.29)

Discuss the conduct and control of a local or site evacuation. (KIA

2.4.29)

For a given security situation, and given site procedures, determine the

appropriate notifications required. (KIA 2.4.30)

OPL27 1SECURITY

Revision I

Page 6 of 16

INSTRUCTOR NOTES

Power Point OPL271Security

should be used to present

material.

Objective I

FSAR

Objective 2

Site org charts, REP

Objective 3

NSDP-3, SPP-1.3

Objective 4

NSDP-3, SPP-1.3

Objective 5

Objective 6

FSAR

Objective 7

If Security present, have them

present overview of Security

response to plant attack

(safeguards).

Site security event AOP,

EPIP-I

Site security event AOP,

EPIP-I

Objective 8

Site EPIP-II

Objective 9

Site EPIP-8

Objective 10

Site EPIP-I, SPP-3.5, NSDP-I

B.

If applicable, present any recent industry events

See attachments 10 & II & 12.

(

TVAN Standard

Plant Access and Security

SPP*1.3

Programs and

Rev. 0010

Processes

Paqe 29 of 52

3.7.1

Vital Area Access (continued)

Site Quality Manager

Site Security Manager

Site Support Manager

Systems Engineering Manager

The completed form SPP-1.3-8 is sent to PAS for processing.

C.

The SFA L contains the personnel with the authority for requesting a change in

clearance level status for access to VAs utilizing Form SPP-1.3-8.

D.

A review of personnel authorized for access to VAs and the clearance level assignment

shall be made monthly by the approving SFAL individual.

E.

Employee(s) are responsible not to attempt or enter VAs for which they do not have the

correct clearance level. Employee(s) shall ensure they do not permit other employees

(badged or visitor) access to VAs for which these employees do not have the correct

clearance level.

F.

A limited number of VA keys may be issued to plant Shift Manager (SM)/Unit

Supervisor (US) personnel for use when an emergency or abnormal operating

condition exists and immediate access is required to protect plant equipment public

health and safety. Use of these keys in routine situations is unauthorized and

considered a security violation.

3.7.2

Control of Vehicl es within th e Protected Area

A.

Vehicles. except under emergency conditions, are searched for firearms, incendiary

devices, and explosives which could be used for sabotage purposes prior to entry into

the Protected Areas. The vehicle search process is outlined in Security Site

Implementing procedures.

B.

All vehicles are logged in and out of the protected area.

C.

Vehicles carrying hazardous materials inside the protected area must be escorted by

an armed member of the security force.

D.

Vehicles not under the control of an individual with unescorted access are allowed

inside the Protected Area as long as the driver of the vehicle is escorted by an

individual with unescorted access , provided with training on the responsibilities of

vehicle escort functions by security, and maintains a means of communications with

the CAS or SAS.

3.7.3

Designated Vehicles

A.

Designated vehicles are owned or leased by TVA or TVA approved contractors.

Designated vehicles are generally limited in their use to onsite plant functions and

remain inside the protected area except for operational, maintenance, repair, security,

and/or emergency purposes.

--- - --- --

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

(

(,

92 . G2.2.5 001

Given the following plant conditions:

-

A plant design change request form is in the approval process.

-

The proposed modification will modify the control rod overlap setpoints in the

logic cabinets.

Which ONE (1) of the following describes who must approve the change prior to

implementation?

.

A'! Plant Manager

B. Site Vice President

C. Maintenance and Mods Manager

D. Engineering and Support Manager

A. Correct. Per SSP-9.3

B. The Site Vice President will be involved in the approval process, but the Plant

Manager has final authority for approval.

C. The MODs manager has responsibility for completion of the physical work, but does

not have final approval authority.

D. Incorrect. The engineering manager does handle the completion of the DeN and

MODs package, but does not have final approval authority.

Monday, March 12, 2007 2:35:37 PM

177

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Knowledge of the process for making changes inthefacility as described in thesafety analysis report.

(

Question No.

Tier 3 Group 2

Importance Rating:

Technical Reference: .

96

SR02.7

SPP-9.3

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPL271C209, Obj. 12

Question Source:

Bank

Question History:

Sequoyah Bank SPP-9.3-2

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

43.3

Comments :

(

Source:

Cognitive Level:

Job Position:

Date:

BANK

LOWER

4/2007

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH BANK

SEQUOYAH

NO

Monday, March 12, 2007 2:35:37 PM

178

(

'-..

TVAN STANDARD

PROGRAMS AND

PROCESSES

PLANT MODIFICATIONS

AND

ENGINEERING CHANGE CONTROL

APPENDIXB

Page 3 of4

SPP-9.3

Rev. 13

Page 59 of 126

GUIDELINES FOR THE COMPLETION OF A CHANGE REQUEST FORM

(

20

21 - 23

24

25

26 - 28

29

30

31

Enter the appropriate reference documents (e.g., WO

number). The document revision number or date should

also be included.

Sign and date to indicate verification performed (as

applicable) in accordance with NEDP-5 for change. This

verification must be performed for safety-and quality-

related changes .

Sign and date to indicate independent review of the

decision to use an Equivalent Change as stated in

Section 3.1.1.M.3 for Equivalent Changes. This signature

also attests that Section 3.1.4.8 and 3.2.4.8 items have

been considered for a change.

Principal Engineer/Manager may N/A non-affected

disciplines.

Other required signatures as applicable. Site VP for "AA"

parent DCNs designating approval of the oversight plan

as required in section 3.1.10

For PICs see Section 3.4.3.C.2.h.

Sign and date to indicate approval when the affected lead

engineer determines that the changes are satisfactory and

the documented justification for the review for safety is

commensurate with the safety significance/complexity of

the change . Lead discipline RLE may N/A nonaffected

disciplines.

Signature , as required, to meet NQAP requirement that

proposed modifications to plant nuclear safety-related

structures, systems, and components shall be approved

by the Plant Manager prior to implementation . N/A for

EDCs and PICs.

Sign and date to concur for DCNs and for PICs initiated

against DCNs by an organization other than the Imp. Org.

and for AA PICs that have had additions or changes from

the initial requests . N/A for EDCs, for PICs against

EDCs, and for PICs initiated by the Implementing

Organization unless additions or changes have been

made from the initial request.

Sign and date to indicate approval.

Enter the Issue RIMS # and Closure RIMS # if applicable

TE

Verifier/Independent

Reviewer

10CFR50.59172.48 Reviewer

Affected Lead Engineers

Plant Manager or designee

Implementing Organization

SE Manager

data entry organization

c

TVAN STANDARD

PROGRAMS AND

PROCESSES

PLANT MODIFICATIONS

AND

ENGINEERING CHANGE CONTROL

SPP-9.3

Rev. 13

Page 29 of 126

B.

For changes requiring PORC review as stated in 3.1.1.M.6, TE shall

schedule the PORC review and distribute copies to PORC members and

obtain PORC review. (Refer to Section 3.1.5.B.3.c)

C.

The TE shall obtain approval of the SEM on Form SPP-9.3-2 (Block 31)

and other applicable documents. Then submit the DCN and DIP

package, original comment sheets, and impact review forms (SPP-9.3-3,

SPP-9.3-4, SPP-9.3-5, SPP-9.3-6, SPP-9.3-19, as applicable) to the

responsible data entry organization.

D.

For design changes the Plant Manager or designee shall approve

implementation by signing Block 29 (Form SPP-9.3-2), as appiicable.

E.

The data entry organization shall:

1.

Input appropriate data into Document Tracking System (DTS).

If the DCN is staged, enter each stage as a package associated

with the DCN package in accordance with Appendix H, "DTS

Status Codes."

2.

Forward original comment sheets, and impact review forms

(SPP-9.3-3, SPP-9.3-4, SPP-9.3-5, SPP-9.3-6, SPP-9.3-19, as

applicable) to the appropriate coordinator.

(

3.

Assign a RIMS/EDM number to the DCN and DIP package and

distribute.

3.1.7

Implement and Test

NOTE

It may become necessary for equipmenUcomponents to be placed in

service in order to perform certain fieldwork (e.g., calibration of

instruments). This work may be accomplished in accordance with

standard work processes and without having to stage the DCN.

Coordination with Operations and Engineering must occur when this

option Is implemented and be documented in the work implementing

document.

A.

Work implementing documents shall be planned, scheduled,

implemented, and closed in accordance with appropriate work control

procedures.

B.

Post-modification testing shall be handled in accordance with

appropriate testing procedures.

C.

Requirements for special post-modification testing, if applicable, are

prepared by SE-DE as test scoping documents in accordance with

SPP-8.3 or test specifications and issued or referenced as part of the

DCN package. Refer to Appendix C, General, Number 14 for screening

on requirement to prepare a test scoping document. The requirement

for SE-DE review of PMT results will be stated when required.

D.

Upon completion of work, the implementing organization completes

Form SPP-9.3-10, "Work Completion Statement" in accordance with the

appropriate work control procedures and submits to Site Engineering.

(

c.

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

93 . 02.2.9 001

Given the following plant conditions:

Unit 2 is in Mode 3.

Engineering has requested that the 2A SI pump be started with the

discharge valve throttled to 75% open to determine starting current.

The Operations Manager has determined that an Urgent Procedure change is

required to support the outage critical path schedule.

The test is NOT described in the current test procedure or the Safety Analysis

Report.

The Shift Manager may approve the test procedure change

_

A. without any restrictions.

B. with concurrence from another SRO.

C. after licensing concurrence is obtained.

D~ after a written safety evaluation has been approved.

A.

Incorrect; Not described in FSAR, then the SM cannot approve by

him(her)self.

B.

Incorrect; 2 SROs can approve normal procedure changes.

C.

Incorrect; Licensing concurrence is not required, results of a review would be

sent through Licensing.

D.

Correct; See SPP-2.2 Sections 3.6 requires a screening review. The

screening review will result in a 50.59 review.

Monday, March 12,20072:35:37 PM

179

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Knowledgeof the process fordetermining if theproposed change, testor experiment increases the probability of occurrence or

consequences of an accident during the change. test or experiment.

(

Question No.

Tier 3 Group 2

Importance Rating:

Technical Reference:

97

SRO 3.3

SPP-2.2

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPL271 C209 Objective 12

Question Source:

Bank

Question History:

Sequoyah Bank 10CFR50.59-1

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.3

Comments:

(

I need the procedure to verify and justify. Need to beef up justification

c

Source:

Cognitive Level:

Job Position:

Date:

BANK

HIGHER

4/2007

Source IfBank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH BANK

SEQUOYAH

NO

Monday. March 12,20072:35:37 PM

180

(,

TVAN STANDARD

PROGRAMS AND

PROCESSES

1.0

PURPOSE

10CFR50.59 EVALUATIONS OF

CHANGES, TESTS, AND EXPERIMENTS

SPP-9.4

Rev. 7

Page 6 of 25

This Standard Program and Process (SPP) establishes requirements for the TVA Nuclear

(TVAN) process of reviewing and evaluating changes, tests, and experiments as required by

10CFR50.59, "Changes, Tests, and Experiments" for plants with operating licenses .

2.0

SCOPE

This procedure applies to all personnel involved in the preparation, review, and approval of

Screening Reviews and 10CFR50.59 Evaluations. Such evaluations are required for proposed

(1) changes to the facility as described in the Final Safety Analysis Report as updated (UFSAR) .

(2) changes to procedures described in the UFSAR, and (3) tests or experiments not described in

the UFSAR , to ensure that the changes , tests, or experiments do not require NRC approval prior

to implementation.

(

3.0

Not all activities are subject to the requirements of 50.59. Changes to the Physical

Security/Contingency Plan and implementing procedures, Radiological Emergency Plan and

implementing procedures , and the Nuclear Quality Assurance Plan (NQAP) are not subject to

10CFR50.59 requirements. Changes to the Physical Security/Contingency Plan, the

Radiological Emergency Plan, and NQAP are made in accordance with 10 CFR 50.54(p),

50.54(q), and 10 CFR 50.54(a)(3), respect ively. Appendix A provides a list of activities that are

controlle~by other regulations or are not in the scope of 50.59.

INSTRUCTIONS

The TVAN program for implementing the requirements of 10CFR50.59 consists of a screening

review thai is performed to determine if a technical specification change is required or if a 50.59

Evaluation is required and a 50.59 Evaluation to determine if a license amendment per 10 CFR 50.90 is required prior to making a change . The TVAN program is based on NEI 96-07 [Revision

1J, Guidelines for 10 CFR 50.59/mplementation. The current version of NE196-07 is provided in

the Business Support Library (BSL) folder with this procedure and shall be used as specified

below. In addition to NEI 96-07, a 10CFR 50.59 Resource Manual is also in the folder with

SPP-9.4 . The Resource Manual provides guidance and numerous examples of interpretations of

50.59. The Resource Manual also provides guidance on the information that needs to be

provided when preparing 50.59 documents as required by this procedure . \\I\\i11en the Resource

Manual is referenced in a form, the content of that section of the form shall be prepared in

accordance with the guidance provided in the Manual. If there are questions of interpretation

between the Resource Manual and NEI 96-07, NEI 96-07 shall be used. Wild card search %spp-

9.4% in BSL may be used to access NEI 96-07 and the USA Resource Manual.

3.1

Preparation of Screening Reviews and 50.59 Evaluations

3.1.1

All changes to the facility, changes to procedures , and tests or experiments not

described in the UFSAR shall be evaluated to determine if the activity is within

the scope of 10 CFR 50.59. Activities controlled by other regulations or change

processes do not require a Screening Review or 50.59 Evaluation. See NEI 96-

07 Section 4.1 and Appendix A of this procedure for more details on the scope

of 10 CFR 50.59. Procedures described in the UFSAR means those procedures

that contain information described in the UFSAR such as how structures,

systems , and components are operated and controlled (including assumed

operator actions and response times). Appendix A and NEI 96-07 Sections 3.11,

4.1.2, and 4.1.4 provide guidance on determining if procedures are within the

scope of 50.59. Minor/editorial procedure changes (as described in SPP-2.2,

QUESTIONS REPORT

forSEQUOYAH 2007 - NRC EXAM REV DRAFT

96. G2.3.4 00 1

(

Given the following plant conditions:

A General Emergency has been declared.

During accountability of site personnel, security reports that one (1) individual can not

be located. Security access records were reviewed and it was determined that the

individual is located in the Auxiliary Building.

A few minutes later, HP reports that an injured individual is in the Pipe Chase in a

radiation field of 50 RlHr.

Two individuals have volunteered for the rescue. BOTH of the volunteers have been

briefed on the risks involved:

Individual A is a 37 year old male.

Individual B is a 46 year old male.

In accordance with EPIP-15, Emergency Exposure Guidelines, which volunteer will be

selected for the rescue, and what is the maximum exposure he may receive?

A. Individual A;-25 Rem maximum.

(

B. Individual B; 25 Rem maximum.

C. Individual A; potentially greater than 25 Rem.

D~ Individual B; potentially greater than 25 Rem.

o is correct.

A and C are incorrect because the older employee will be selected due to a lower

chance of long term effects of radiation exposure affecting this person . B is incorrect

because the person may potentialiy receive >25 Rem.

c

Monday, March 12, 20072:35:38 PM

185

--_.._--

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Knowledge of radiationexposure limits and contamination control, including permissiblelevelsIn excess of those authorized.

(

Question No.

Tier 3 Group 3

Importance Rating:

Technical Reference:

98

SRO 3.1

EPIP 15

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPL271C198REP Objective 1.f

Question Source:

Modified

Question History:

Robinson 2007 NRC Exam

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

(

Comments:

Source:

Cognitive Level:

Job Position:

Date:

MODIFIED

HIGHER

4/200 7

Source IfBank:

Difficulty:

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday. March 12, 2007 2:35:38 PM

186

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

99. G2.4.49 001

(

Given the following plant conditions:

A transient has occurred on Unit 1 resulting in the following alarms:

OTOT RUNBACKIROO STOP ALERT

ROD CONTROL URGENT FAILURE

OPDT RUNBACK

Reactor power indicates the following :

N41-105.2%

N42 - 106.2%

N43-105.9%

N44-106.1%

Tavg is 571°F

Which ONE (1) of the following has occurred , and which procedure(s) islare required to

be implemented?

A. Uncontrolled Rod Withdrawal; E-O, Reactor Trip or Safety Injection.

(

B. Uncontrolled Rod Withdrawal; C.01, Rod Control Malfunctions.

C~ SG Safety Valve opened coincident with a rod control failure; E-O, Reactor Trip or

Safety Injection.

O. SG Safety Valve opened coincident with a rod control failure; AOP-S .05, Steam

Leak and AOP-C.01, Rod Control Malfunctions.

A and B are incorrect because Tave is lower than required for the power level. Tavg

should be at 578 for 100% power, so a rod withdrawal is likely not the initiator. Also,

rods will not withdraw if there is an urgent failure.

A does contain the correct action.

Action for B would be correct if a trip wasn't required and the event was a rod

withdrawal

o is incorrect because a setpoint for reactor trip has been exceeded as indicated by the

first out annunciator for OPOT

o is correct. Must recognize that there is a first out alarm and take action to trip the unit

Monday, March 12. 20072:35:38 PM

191

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Ability to perform without reference to procedures those actions that require immediate operation of system components and

controls.

(

Question No.

Tier 3 Group 4

Importance Rating:

Technical Reference:

99

SR04.0

ARPs, E-O

Proposed references to be provided to applicants during examination:

None

Learning Objective:

E-O, B.6.a

Question Source:

Bank

Question History:

WTSI

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments:

(

Source:

BANK

Cognitive Level:

HIGHER

Job Position:

Date:

4/2007

(,

Monday, March 12,20072:35:38 PM

Source IfBank:

Difficulty:

Plant:

Last 2 NRC?:

WT SI

SEQUOYAH

NO

192

(

OPL271E-0

Revision 0

Page 3 of 16

I.

PROGRAM:

OPERATOR TRAINING - LICENSED

II.

COURSE:

LICENSE TRAINING

III.

LESSON TITLE:

E-O, "Reactor Trip or Safety Injection"

IV.

LENGTH OF LESSON/COURSE:

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

V.

TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of HLC Procedures training, the participant shall be able to explain,

using classroom evaluations and/or simulator scenarios, the requirements of E-O,

"Reactor Trip or Safety Injection".

B. Enabling Objectives

Obiectives

O.

Demonstrate an understanding of NUREG 1122 knowledge's and abilities

associated with Reactor Trip or Safety Injection that are rated z 2.5 during Initial

License Training and z 3.0 during License Operator Requalification Training for

the appropriate position as identified in Appendix A

J.

State the purpose/goal of this E-O.

2.

Describe the E-O entry conditions.

3.

Summarize the mitigating strategy for the failure that initiated entry into E-O.

4.

Describe the bases for all limits, notes, cautions, and steps of E-O.

5.

Describe the conditions and reason for transitions within this procedure and

transitions to other procedures.

6.

Given a set of initial plant conditions use E-O to correctlv:

a.

Recoqnize entry conditions.

b.

Identifv required actions.

c.

Respond to Contingencies.

d.

Observe and Interpret Cautions and Notes.

7.

Apply GFE and system response concepts to the abnormal condition - prior to,

durlno and after the abnormal condition.

(

1.0

PURPOSE

REACTOR TRIP OR SAFETY INJECTION

E-O

Rev. 28

(

This procedure verifies proper response of the automatic protection systems

foilowing manual or automatic actuation of a reactor trip or safety injection,

assesses plant conditions, and identifies the appropriate recovery procedure.

2.0

SYMPTOMS AND ENTRY CONDITIONS

2.1

SYMPTOMS REQUIRING REACTOR TRIP

A.

Source range flux greater than 10

5 cps (blockable above P-6).

B.

Intermediate range flux greater than 25% power (blockabie above P-l 0).

C.

Power range flux greater than 25% power (blockable above P-l0).

D.

Power range flux greater than 109% power.

E.

Power range rate greater than +/- 5% in 2 seconds.

F.

Pressurizer level greater than 92% (biocked below P-7).

G.

Pressurizer pressure greater than 2385 psig.

H.

Pressurizer pressure less than 1970 psig (blocked below P-7).

I.

RCS flow less than 90% (2/4 loop low flow trip blocked below P-7;

y. loop low flow trip blocked below P-8).

J.

RCP voltage less than 5.022 kilovolts (blocked below P-7).

K.

RCP frequency less than 57 hertz (blocked below P-7).

L.

Overtemperature 1I.T greater than 115% (variable).

M.

Overpower 1I.T greater than 108.7% (variable).

N.

Turbine trip (stop valves closed or autostop oil pressure less than 45 psig)

(blocked below P-9).

O.

SIG level less than 10.7% [15% ADV] (variable time delay below 50% power).

P.

Safety injection.

Q.

SSPS general warning in both trains.

Page 2 of 21

(

SON

REACTOR TRIP OR SAFETY INJECTION

E-O

Rev. 28

2.2

SYMPTOMS OF REACTOR TRIP

A.

Any valid reactor trip signal [status panel M-5 or M-6].

B.

Any reactor trip alarm lit [M-4D].

C.

Rapid drop in neutron level indicated by nuclear instrumentation .

D.

Shutdown and control rods inserted.

E.

Rod bottom lights lit.

F.

Rod position indicators at zero.

2.3

SYMPTOMS REQUIRING SAFETY INJECTION

A.

Pressurizer pressure less than 1870 psig (blockable below P-11).

(

B.

SIG pressure less than 600 psig (Ieadflag) (blockable below P-11).

C.

Containment pressure greater than 1.5 psig.

2.4

SYMPTOMS OF SAFETY INJECTION

A.

Any valid SI signal [status panel M-6].

B.

Any SI alarm lit [M-4D].

C.

ECCS pumps running .

3.0

OPERATOR ACTIONS

Page 3 of 21

(

QUESTIONS REPORT

for SEaUOYAH 2007 - NRC EXAM REV DRAFT

98 . 02.4.30 001

Given the following plant conditions on Unit 1:

Condition

0803

0805

0808 .

0812

0829

PZR level dropping slowly

Crew enters AOP-R.05 , RCS Leak and Leak Source

Identification

Crew trips the reactor

ALERT is declared

SITE AREA EMERGENCY is declared

(

Which ONE (1) of the following is the LATEST time that the state and local authorities

must be first notified of the event in progress?

A. 0818

B. 0823

C~ 0827

D. 0843

A. Incorrect. 15 minutes from start of event

B. Incorrect. 15 minutes from reactor trip

C. Correct. 15 minutes maximum fol/owing declaration for notification of state and local

authorities

D. Incorrect. 15 minutes from SAE. Should have already notified of alert

Monday, March 12, 2007 2:35:38 PM

189

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV DRAFT

Knowledge of which events related to system operations/status should be reported to outside agencies.

(

Question No.

Tier 3 Group 4

Importance Rating:

Technical Reference: .

100

SRO 3.6

EPIP-1

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

Question Cognitive Level:

10 CFR Part 55 Content:

Comments:

OPL271C168 Objective 2.c

New

Higher

43.5

(

Source:

Cognitive Level:

Job Position:

Date:

NEW

HIGHER