ML053330522

Draft - RO & SRO Written Exam (Folder 2)

ML053330522
Person / Time
Site:	Hope Creek
Issue date:	10/21/2005
From:	Suzanne Dennis Operations Branch I
To:	Reid J Public Service Enterprise Group
Conte R
References
ES-201, ES-201-2 IR-05-301
Download: ML053330522 (117)

Text

ES-201 Examination Outline Quality Checklist Form ES-201-2

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Facility: Hope Creek - SRO Only Exam Date of Examination: 11/28/05 Initials Item Task Description a b c#

1. a. Verify that the outline(s) fit(s) the appropriate model, in accordance with ES-401.

w - 59 R b. Assess whether the outline was systematically and randomly prepared in accordance with I Section D.l of ES-401 and whether all WA categories are appropriately sampled. mt9 59

c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics. M&9

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] d. Assess whether the justifications for deselected or rejected WA statements are appropriate. b&] ] vfl 1 :q~

a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions, instrument and component failures, technical specifications, and major transients.

b. Assess whether there are enough scenario sets (and spares) to test the projected number and mix of applicants in accordance with the expected crew composition and rotation schedule without compromising exam integrity, and ensure that each applicant can be tested d using at least one new or significantly modified scenario, that no scenarios are duplicated from the applicants audit test(s), and that scenarios will not be repeated on subsequent days.

c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative and quantitative criteria specified on Form ES-301-4 and described in Appendix D. w

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3. a. Verify that the systems walk-through outline meets the criteria specified on Form ES-301-2:

(1) the outline(s) contain(s) the required number of control room and in-plant tasks w distributed among the safety functions as specified on the form t (2) task repetition from the last two NRC examinations is within the limits specified on the form 3v18 i 59 T (3) no tasks are duplicated from the applicants audit test(s)

(4) the number of new or modified tasks meets or exceeds the minimums specified on the form (5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria on the form.

b. Verify that the administrative outline meets the criteria specified on Form ES-301-1:

(1) the tasks are distributed among the topics as specified on the form (2) at least one task is new or significantly modified 7T& 50 (3) no more than one task is repeated from the last two NRC licensing examinations

c. Determine if there are enough different outlines to test the projected number and mix of applicants and ensure that no items are duplicated on subsequent days. q )9

4. a. Assess whether plant-specific priorities (including PRA and IPE insights) are covered in the appropriate exam sections.

G b. Assess whether the 10 CFR 55.41/43 and 55.45 sampling is appropriate.

E N c. Ensure that WA importance ratings (except for plant-specific priorities) are at least 2.5. %9 E

d. Check for duplication and overlap among exam sections. la A e. Check the entire exam for balance of coverage. SO L

f. Assess whether the exam fits the appropriate job level (RO or SRO). & Jt7 1 /VJ+*

a.

b.

c.

Author Facility Reviewer ()

NRC Chief Examiner (#)

Michael L. Brown I SnLnfJ _- P(d&J II

/a- .. - \

I *

d. NRC Supervisor 5 /A J W

ES-201 Examination Outline Quality Checklist Form ES-201-2 of normal evolutions, instrument and component failures, technical specifications, and major transients.

I M b. Assess whether there are enough scenario sets (and spares) to test the projected number U and mix of applicants in accordance with the expected crew composition and rotation L schedule without compromising exam integrity, and ensure that each applicant can be tested using at least one new or significantly modified scenario, that no scenarios are duplicated m@ SO A

T from the applicants' audit test(s), and that scenarios will not be repeated on subsequent days.

0 R

3. a. Verify that the systems walk-through outline meets the criteria specified on Form ES-301-2:

(1) the outline(s) contain@)the required number of control room and in-plant tasks w distributed among the safety functions as specified on the form I (2) task repetition from the last two NRC examinations is within the limits specified on the form T (3) no tasks are duplicated from the applicants' audit test(s)

(4) the number of new or modified tasks meets or exceeds the minimums specified on the form (5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria (1) the tasks are distributed among the topics as specified on the form at least one task is new or significantly modified 4.

G 1 E N

E R

A i~

1 a.

b.

Author Facility Reviewer (')

Michael L. Brown I

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c. NRC Chief Examiner (#) SnqL Og&d!& } /d 'zh

d. NRC Supervisor @. J L (c+ k- /h-y.v

ES-301 Simulator Scenario Quality Checklist Form ES-301-4

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Facility. Hope Creek Date of Exam: 11/28/05 Scenario Numbers: 1 12 / 3 Operating Test No.:

I QUALITATIVE ATTRIBUTES lnitia a b' I

1. The initial conditions are realistic, in that some equipment and/or instrumentation may be out of service, but it does not cue the operators into expected events.

2. The scenarios consist mostly of related events.

3. Each event description consists of the point in the scenario when it is to be initiated the malfunction(s)that are entered to initiate the event the symptoms/cues that will be visible to the crew the expected operator actions (by shift position) the event termination point (if applicable) incorporated into the scenario 59

ES-301 Transient and Event Checklist Form ES-301-5

2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis.

Malfunction Checklist

'Scenario #I Scenario #2 'Scenano #3 Inadvertent HPCl Initiation " A Core Spray pump otal Malfunctions (5-8) PT-N078B fails Low (TS) CrS) trips Loss of "6" MG Set Trip of 106130 "A APRM fails Failure of Standby MG Set Rod 22-35scrams Lube Oil pump to start Turbine Bldg chillers trip "A' CRD pump trips EHC Filter Clogging "6" Recirc pump high vibz RClC Steam Leak Electrical ATWS Loss of Offsite Power Small Break LOCA Failure of "A" EDG to star RHH pump being aligned "E' SRV fails to OPEN for Drywell Spray trips Total = 6 Total = 7 lalfunctions after EOP Entry RClC Isolation valves fail to RHR pump being aligned

' -2) close for Drywell Spray trips Failure of 'A' EDG to star

'E" SRV fails to OPEN Trip of '6' RHR pump Total = 2 Total = 1 Total = 2 Loss of "6' MG Set (AB-IC- Inadvertent HPCl Initiation " A APRM fails (AB.IC-bnormal Events (2-4) 0003) (TS)(AB.RPV-0001) 0004)

Turbine Bldg chillers trip (AB.CONT-000 1)

Rod 22-35scrams ( A B X - Lube Oil pump to start Recirc pump high vibs (AB.RPV-0003)

Total = 3 I

Loss of Offsite Power I

I Total = 2 Total = 1 EO41 01 AB-0000 EO-0102 Total = 1 OP contingencies requiring I ubstantive actions (0-2)

Crew initiates a manual

-t Total = 0 RRCS is manually actuate within 2 minutes of ITotal = 0 Crew removes Recirc pump from service within scram before RClC room reaching an Automatic 2 minutes of reaching the ritical Tasks (2-3) temp reaches 250°F Scram setpoint Danger setpoint Crew manually scrams Crew Emergency the reactor within 3 Depressurizes the plant Crew places Drywell spray minutes of entering within 5 minutes of when ED in service before ED criteri Region 1 with the conditions are reached. is reached OPRM's INOP ICrew places Suppression IT Pool Cooling in service Crew opens at least 5 SRV's prior to Supp Pool temp i to comply with ED criteria (Total = 3 ota =

I

, I 17°Fotal =

ES-301 Competencies Checklist Form ES-301-6 RX 1 er tin T st N .:

APPLICANTS I

Competencies InterpreVDiagnose Events and Conditions Comply With and Use Procedures (1)

Operate Control Boards (2)

Communicate (land Interact Demonstrate Supervisory Ability (3)

Comply With and Use Tech. Specs. (3)

Notes:

(1) Includes Technical Specification compliance for an RO.

(2) Optional for an SRO-U.

(3) Only applicable to SROs.

Instructions:

Circle the applicants license type and enter one or more event numbers that will allow the examiners to evaluate every applicab Author:

NRC Reviewer:

ES-301 Control Room / In-Plant Systems Form ES-301-2 Outline System / JPM Title Type Code* Safety Function

a. Perform an Accident Monitoring Instrumentation Channel Check D , S, A 7 (ZZ-025)

b. Place RACS in Service (ED-003) D, S, A 8

c. Recirc Flow Control System / Reset a Recirc MG Set Scoop Tube M,S, A 1 Lockup (AH. Path - Recirc speed inexplicably rises following reset)

(BB002)

d. Transfer 1E Power Supply to Backup N, S, A 6

e. PClS / Resetting Isolation Systems N, s, E 5 l f. Swap FW Level Control, Startup Level Controller to Single Element control

g. HPCl - Startup HPCl in the Full Flow Test mode (BJ002) M, s, A 2

4

h. Manually Start FRVS system (GU001) D, S 9

i. Main Steam/ Establish Control from Outside the Control Room N 3

j. A/C Electrical / Startup a 20KVA Inverter N 6

1) k. Control Rod Drive / Isolate a CRD HCU (BF006)

I D,R I 1 Type Codes Criteria for RO ISRO-I I SRO-U (A)lternate path (5) 4-6 14-6 12-3 (C)ontrol room (D)irect from bank (4) 5 91s 015 4 (E)mergency or abnormal in-plant (1) 1/1/1 (L)ow-Power (1) 2 1 I 2 112 1 (N)ew or (M)odifiedfrom bank including 1(A) (7) 2 2 1 2 212 1 (P)revious 2 exams (0) 5 3I 5 3 / 5 2 (randomly selected)

(WA (11 2 1 I 2 112 1 (S)imulator NUREG-1021, Revision 9

ES-301 Control Room / In-Plant Systems Form ES-301-2 Outline Facility: HoPe Creek Date of Examination: 11/28/05 Exam Level (circle one): SRO Operating Test No.:

1 Control Room Systems@(8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)

System / JPM Title Type Code* Safety Function

)I a. Perform an Accident Monitoring Instrumentation Channel Check i-02; ~

1 D , S, A 1I 7

b. Place RACS in Service (ED-003) ID,S,A 8

~ _ _ _ _ _ ~ ~ ~ ~ ~

c. Recirc Flow Control System / Reset a Recirc MG Set Scoop Tube M, S, A 1 Lockup (Alt. Path - Recirc speed inexplicably rises following reset)

(BB002)

d. Transfer 1E Power Supply to Backup N, S, A 6

e. PClS / Resetting Isolation Systems N, S, E 5

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f. Swap FW Level Control, Startup Level Controller to Single N, S, L 2 Element control

g. HPCl - Startup HPCl in the Full Flow Test mode (BJ002) M, S, A 4

h. Main Steam/ Establish Control from Outside the Control Room N 3

i. A/C Electrical / Startup a 20KVA Inverter N 6 1 j. Control Rod Drive / Isolate a CRD HCU (BF006) I D,R I 1

• Type Codes Criteria for RO I SRO-I I SRO-U (A)lternatepath (5) 4-6 I 4-6I 2-3 (C)ontrol room (D)irect from bank (3) 91sais4 (E)mergency or abnormal in-plant (1) 1/1/1 (L)ow-Power (1) 2 112 112 1 (N)ew or (M)odifiedfrom bank including 1(A) (7) 22/22/21 (P)revious 2 exams (0) s 3 I < 3I 5 2 (randomly selected)

(WCA (1 1 2 112 112 1 (S)imulator NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 Administrative Topic Type Describe activity to be performed (see Note) Code*

Conduct of Operations S, A, D Conduct Weekly Power Distribution Lineup I Equipment Control I s, D Conduct Reactor Recirculation Single Loop

)I Operation Complete an On The Spot Change (OTSC) l Conduct of Operations R, N None Radiation Control S, D Complete a Major Equipment and Electrical Status (MEES) Form N0TE:AII items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 5 3 for ROs; 5 4 for SROs & RO retakes)

(N)ew or (M)odified from bank ( 2 l ) , (A)lternate Path (P)revious 2 exams ( c 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Facility: HoDe Creek Date of Examination: 11/28/05 Examination Level : SRO Operating Test Number:

Administrative Topic Type Describe activity to be performed (see Note) Code*

Conduct of Operations Equipment Control ws,

• D Conduct Reactor Recirculation Single Loop

\ Operation Conduct of Operations TpF R, Complete an On The Spot Change (OTSC)

Radiation Control Rv I Review a Containment Purge Form Emergency Plan s, M Classify an Emergency Event - 2.4.41 - May be done after a scenario using the simulator.

I NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (I 3 for ROs; I 4 for SROs & RO retakes)

(N)ew or (M)odified from bank ( 2 l ) , (A)lternate Path (P)revious 2 exams (I 1; randomly selected)

DU & \ d L 51% u1.sr-r I 7 d I+ /of

Amendix D Scenario 0ut Iine Form ES-D-1 Facility: Hope Creek Scenario No.: 3 (Spare) Op-Test No.:

Examiners: Operators:

Initial Conditions: Plant is at 80% Dower, middle of life, returnina to power after a mini-outaqe, the operators are at step 5.4.29 of IOP-3 with the 3rdRFP havinq iust been Dlaced in service.

Turnover:

Plant at 80% power with the 3rdRFP havinq just been placed in service. Need to perform Core Spray Loop A operabilitv test. OPRM Svstem is INOP due to an existinq 10CFR21 issue. The OPRM svstem is still functional but is considered INOP per Tech Specs. Severe weather is forecast for the area.

Event Malf. Event Event No. No. Type* Description 1 N (BOP) Perform HC.OP-ST.BE-0002, Core Spray Pump Loop A Full flow C (BOP) test. Core Spray Loop A pump will trip when test valve is CfCRSI oDened. fTSI 2 R (RO) Raise Load using Recirc flow or Rods R (CRS) 3 I (RO) A APRM fails (TS)

I (CRS) 4 C (BOP) Turbine Building Chillers trip C (CRS) 5 C (RO) B Recirc Pump high vibration.

6 M (ALL) Loss of off-site power due to storm 7 C (BOP) A EDG fails to start 8 C(R0) B RHR pump trips

Appendix D Scenario Outline Form ES-D-1 Facility: Hope Creek Scenario NO.: 2 Op-Test No.:

Examiners: Operators:

Initial Conditions: 80% power, middle of cycle. A control rod sequence exchanae has iust been performed and Operators are preparina to raise power to 80% load and return the 3rdRFP back to service.

Turnover :

SLC pump AP-208 has been taaqed out for a motor replacement and is expected back in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

No other equipment is Out of Service. Raise power to -80% electrical load and place the 3rdRFP in service (currently runnina on recirc)

I

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Facility: Hope Creek Scenario NO.: 1  % Op-Test No.:

Examiners: Operators:

Initial Conditions: 4% power, Reactor Startup in prowess, B EHC pump blocked for maintenance Turnover:

A Reactor Startup is in prowess with IOP-3 completed up to step 5.3.21. Reactor power is approximately 4%. 1BP116 EHC pump is taqqed out for maintenance and will be out of service until a new pressure compensator arrives tomorrow.

Event Description 1 R (RO) Withdraw Group 7 control rods to position 8 R (CRS) ~

2 I (RO) PT-N078B, Steam Dome Pressure Transmitter fails LOW (TS)

I (CRS) 3 I (ALL) Loss of 6 MG Set Control Rod 22-35 inadvertently scrams (TS)

Steam Leak from RClC piping RClC isolation valves fail to close E SRV fails to close

(Question# 1 1 -

Hope Creek RO Exam Nov 2005 rTiey# 1 Gram# 1 I

'ISRO I I Importance 3.6 I

Ld3001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 3

AK1.03 Knowledge of the operational implications of the following concepts as they apply to the Partial or Complete Loss of Forced Core Flow Circulation Thermal Limits :(CFR: 41.8 to41.10/45.3) 3

'3 Question Given: Hope Creek was at 100% power when the " 6 Recirc pump developed excessive vibration and needed to be tripped.

Y WHICH ONE of the following actions is REQUIRED to be taken in accordance with HC.OP-IO.ZZ-0006, Power Changes during Operation?

MAPLHGR limit must be reduced and MCPR safety limit must be reduced.

A 7

B MAPLHGR limit must be reduced and MCPR limits must be raised.

L//

MAPLHGR limit must be raised and MCPR limits must be reduced.

MAPLHGR limit must be raised and MCPR limits must be raised.

x x"

Answer B References HC.OP-AB.RPV-0003 (Q), Rev. 9.Recirculation System HC.OP-IO.ZZ-0006,Rev. 33. Power Changes during Operation Justz@ation References during Exam None A. INCORRECT per IOP-6, step 5.3.7 MAPLHGR limit must be reduced and MCPR safety limit must be raised Alba (z" g b L Z L B. CORRECT - see "A" Above C. is INCORRECT per IOP-6. step 5.3.7 which states that the MCPR safety limit must be raised 3v D.is INCORRECT per IOP-6, step 5.3.7 MAPLHGR limit must be reduced IOP006E006 - (R)Assess plant conditions and determine if the requirements for entering SINGLE LOOP have been met.

Question Source Mod Memo y Level Comprehension Level Question History:

SXD review - 7/21/05 LOD 1.75 perhaps rewrite to make more difficult, removed "initially" from in front of at 100%

power in stem.

AF - 8/23 NOT fair to ask subsequent actions questions from memory, feels everybody will jump on "C" M0 - 9/26 - SXD to resolve SXD - OK RJC - tie to lesson plan objection MB - added a lesson plan tie in - IOP Lesson Plan OBJ 6 ) Assess plant conditions and determine if the requirements for entering SINGLE LOOP have been met.

AF - still doesn't think they will know this from memory, Perhaps look for MCPR questions in bank -, maybe give them IOP-6

.'9 - 10125 Re-wrote question to only ask what happens to Thermal Limits on Trip of Recirc pump IAW IOP-6

OR0 per# 1 Group# I 3 7SRO Importance 3.9 1

./

Partial or Complete Loss of AC 1 6 AA2.05 Ability to determine and interpret the following as they apply to Whether a partial or complete loss of A.C. Power 3

Partial or Complete Loss of AC has occurred:(CFR: 41.10 143.51 45.13) s Question Given the following conditions:

- The plant is in Operational Condition 5 with the Electrical Distribution System aligned in the Normal lineup.

- An internal short on Transformer 1BX-501 causes a sudden pressure fault on the transformer.

Which one of the following describes the resulting availability of power for the Safe Shutdown Systems?

A Power is lost permanently to both 4.16KV switchgear 10A401 and 10A403.

13 KV breakef0SX+md BS 1-2 staslosed.

B and D Dies I Generators start but their output breakers DO NOT CLOSE.

x B Power is lost momentarily to both 4.16KV switchgear 1OA402 and 1OA404.

13 KV breaker$B45-3aRBBS 1-2 tripfopen.

Power is restored when the B and D Diesel generators output breakers close.

Power to both 4.16KV switchgear 10A402 and 10A404 fast transfers to Transformer 1AX501.

?

13 KVbreake ES4Xhnd BS 1-2 t i e p e n .

B and D diese generators START but their output breakers DO NOT CLOSE.

x Power to both 4.16KV switchgear 10A402 and 10A404 fast transfers to Transformer lAX501.

13 K V b r e a k e r & H a n d -BS 1-2 t i h p e n . 07 B and D diesefgenerators DO NOT START. k - -

. ;L;t.>

&qNb aQ-c/..LI 3 III Answer D References -

Hope Creek Question (276871 Modified Drawing E-0001 and 066-01: Class 1E AC Power Distribution NOHOlEAC00-02 -CLASS 1E AC POWER DISTRIBUTION, page 32 of 93

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Justification References during Exam Drawing E-0001 Justification:

Correct answer. 13 Kv Breakers BS 2-3 and BS 1-2 trip open. Bus section 2 is de-energized, Bus section 1 remains energized. The bus infeed breaker swap to the AX501 feed. The loads remain energized. Because one infeed is always available, the Diesels do NOT start.

A - INCORRECT Power is NOT permanently lost to both 4.16KV switchgears. Power is restored when the bus infeed breaker swaps to the AX501 feed.

B INCORRECT Power is NOT restored from the B 8 D Diesel Generators C INCORRECT The B 8 D Diesel Generators DO NOT START Question Source Mod Memory Level Comprehension Level Question Histo ry:

SXD Review 7/21/05 - OK AF - 8/23 will do more research on - mav have Droblem with "D" SXD OK -

&F - Possible WA mismatch, reworded question, SXD to look at

Hope Creek RO Exam Nov 2005 -

I # 1 Group# 1 1 7, 3004 Partial or Total Loss of DC Pwr / 6 z AK3.01 Knowledge of the reasons for the following responses as they apply to Partial or Total Loss of DC Pwr Load shedding Plant Specific:(CFR: 41.5/41.10 I 2

45.6 145.13)

Question 3

With the plant at 100% power, the plant loses power to 125V DC Class 1E switchgear 1OD410.

If the plant were to experience a LOCA, how will Load shedding and control of non-lE loads be affected:

Load shedding of Non-1E loads that get control power from 10D410 ...

A will occur and these loads can be be operated from the Control Room (ie. Load shedding and control will NOT be affected) will occur, however, these loads can NOT be operated from the Control Room.

B

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x wilt NOT occur, however, these loads can be operated from the Control Room x

will NOT occur and these loads can NOT be operated from the Control Room.

J

/

Answer D References INPO Question 23597 (somewhat)

Hope Creek Lesson Plan NOHOlEAC00-02. CLASS 1E AC POWER DISTRIBUTION p34 talks about load shedding of non-1E loads on a LOCA NOHOlDCELEC-00, DC ELECTRICAL DISTRIBUTION p.22 talks about 125V DC supplying breaker control power Justification References during Exam None A. INCORRECT - Load shedding will occur.

B. - INCORRECT - load will NOT auto trip on a Load Shed Signal and CANNOT be operated from the Control Room.

C. - INCORRECT load will NOT auto trip on a Load Shed signal

0. - CORRECT - Load shedding will NOT occur and load CANNOT be operated from the Control Room Question Source Mod @I Memory Level 0 Comprehension Level Question History:

SXD Review - 7/27 Question Stem confusing -

7/27 - Rewrote Question Stem re-submitted 812 JD Weak Question, doesn't address K/A - WA Q about DC load manual shedding to conserve battery life 8t3 rewrote question again.

AF - 8/23 -word search make Not's all caps. Possible KIA mismatch.

MB - 9126 SXD to resolve SXD - OK as is AF - still thinks it's a KIA mismatch

[Question# 4 I Hope Creek RO Exam Nov 2005 -

rTier# 1 Group # I xi005 Importance 3 Main Turbine Generator Trip / 3 3

AG2.1.2 Knowledge of operator responsibilities during all modes of 3 plant operation (CFR: 41.10 / 45.13)

Question Due to a main turbine vibration problem with a generator load of 110 MWe, a manual turbine trip is performed.

Which of the following describes the Maximum Time Limit permitted to open the generator Output Breakers for the given conditions in accordance with procedure HC.OP-SO.AC-0001, Main Turbine? (Assume they have NOT already tripped on reverse power.)

A Immediately B Within 15 seconds of the turbine trip

• /-

x Within 60 seconds of the turbine trip Within 90 seconds of the turbine trip A l Answer References -

Hope Creek Question Q53470 HC.OP-SO.AGOOOl(Q) Rev. 48, MAIN TURBINE OPERATION P8L 3.1.1 5 Justipcation References during Exam None Correct Answer: "15 seconds" -Procedure caution calls for operator actions within 15 seconds of the turbine trip at low power.

The following distractors are incorrect as follows:

"immediately' Procedure caution calls for operator actions within 15 seconds "60 seconds" - Procedure caution calls for operator actions within 15 seconds of the turbine trip at low power.

"90 seconds"-Procedure caution calls for operator actions within 15 seconds of the turbine trip at low power. Only when above 150 MWe is the time extended to 90 seconds.

Question Source Bank Memory Level 0 Comprelzension Level Question History:

SXD Review 7/21 Had question about lower power -

7/27 verified power level ok per IOP-4 p.15 AF 8/23 - A and B essentially the same. Suggests making A longer than 15 seconds.

MB 9/26 - SXD to resolve SXD - Minor change to stem MB -1013 Made changes as requested AF-OK

aRO [Tier# 1 Group # 1 I 'L

'iSRO Importance 3.7 3

3 irects the operator to RESET the scram (SB) if conditions permit

~ 1 56b" 4 I23 'c'

1. 4 B

1 A

-To reduce the potential for CRD pump runout and reduce the amount of time for the HCU accumulators to recharge.

To restore the CRD hydraulic system to normal for insert and withdrawal capability if rods are found at the 02 A5.r

&&de5 To reestablish the normal primary vessel boundaries by isolating the CRD HCU from the scram discharge volume (SDV) and closing the SDV vent and drain valves.

'>(

x 6

To prevent excessive discharge of hot radioactive water to the Reactor Building Equipment Drain Sump.

J Answer B References -

Hope Creek Question Q56128 NOHOlABOOOM)l, Reactor Scram AB-0000 p.14 Justification References during Exam / As-m Justification:

C - INCORRECT To reestablish the normal primary vessel boundaries by isolating the CRD HCU from the scram discharge volume (SDV) and dosing the SDV vent and drain valves. Incorrect - the Scram reset will open the vents and drains B - CORRECT - To restore the CRD hydraulic system to normal for insert and withdrawal capability if rods are found at the 02 or beyond position. Correct.

A - INCORRECT -To reduce the potential for CRD pump runout and reduce the amount of time for the HCU accumulators to recharge. Incorrect - system flow restricting orifice limit pump runout to 200 gpm D INCORRECT To prevent excessive discharge of hot radioactive water to the Reactor Building Equipment Drain Sump. Incorrect - resetting scram will send water to the Rx Bldg Equipment Drain Sump Question Source Bank Memory Level 0 Comprehension Level Question History:

Submitted 7/22 SXD Reviewed 7/23 - for Distactor C asked is this verified?

AF- 8/23 swapped A & C justification.

MB - Made changes as requested AF-OK

[Question # 6 Hope Creek RO Exam Nov 2005

~~

,5016 Control Room Abandonment / 7 AG2.1.30 Ability to locate and operate components, including local controls. (CFR: 41.7 / 45.7)

Question Remote Shutdown Panel Transfer Switch "B" has been placed in the EMERGENCY position.

Which of the following lists the SRVs that can be operated at the Remote Shutdown Panel (100399) AND describes the status of their controls in the Control Room (CR)?

A A, B, C , D. 8 E. R controls still function normally Y

B

?

A, B, C, D, & E. R controls are disabled R controls still function normally C

I, R controls are disabled.

J Answer D References Hope Creek Question - (262205, HC.OP-IO.U-0008, Section 5.1. Attachment #I . 8.2.9 Step NOHOIMSTEAMC-02, MAIN STEAM SYSTEM, Obj R3d Justification References during Exam None D - CORRECT - F, H 8 M. CR controls are disabled. Only SRVs M. F & H can be controlled from the RSP and when the transfer switches are in EMERGENCY, the CR functions are disabled.

A - INCORRECT A, B,C, D & E are the ADS valves NOT the valves that can be controlled from the RSP.

B - INCORRECT - A, B, C, D & E CANNOT be controlled from the RSP.

C - INCORRECT CR controls are disabled when RSP transfer switch B has been placed in the EMERGENCY position.

Question Source Bank Memory Level 0 Comprehension Level Question History:

SXD Review 7/21 LOD 1.75 evaluate Revising AF-OK

OR0 [Tier# 1 Group # I 3 rlSRO Importance 2.9 3

,5018 AA2.04 Partial or Total Loss of CCW / 8 Ability to determine and interpret the following as they apply to System Flow 2

Partial or Total Loss of CCW :(CFR: 41.10/43.5/ 45.13) 3 Question Hope Creek is at 100% power with the following SACS lineup: 3

- "A" 8 " C SACS pumps running supplying TACS and the " A SACS loads.

- "D" SACS pump running supplying the "B" SACS loads.

When a Small Break LOCA occurs outside the drywell causing a SCRAM. Reactor Water level drops to minus 50" as HPCl auto starts and recovers level.

Which of the following correctly describes the SACS lineup and how flow is being supplied to the TACS loads?

B A "A" 8 "C" SACS pumps running supplying TACS and the "A" SACS loads "0"SACS pumps running supplying the "B" SACS loads "A" 8 "C" SACS pumps running supplying TACS and the "A" SACS loads x

"B" 8 "D" SACS pumps running supplying the "B"SACS loads

- ~~~ ~ ~~

"A" 8 "C" SACS pumps running supplying the "A" SACS loads "D" SACS pump running supplying the "8"SACS loads TACS loads are isolated.

"A" 8 "C" SACS pumps running supplying the "A" SACS loads L, "B" 8 "D" SACS pumps running supplying the "B" SACS loads TACS loads are isolated.

Answer B References Hope Creek Procedure HC.OP-AB.COOL-0002, SAFETY/TURBINE AUXILIARIES COOLING SYSTEM, p. 9-1 3 NOHOlSTACSO-02, SAFETY AND TURBINE AUXILIARY COOLING WATER SYSTEM, p.49-50 Justification References during Exam None A INCORRECT - "5"SACS pump will AUTO START on a Level 2 LOCA B CORRECT - "B" SACS pump will AUTO Start on LOCA Level 2, TACS does NOT isolate C - INCORRECT TACS does NOT isolate on a Level 2 LOCA only on a Level 1 LOCA D - INCORRECT - See "C" Question Source New Memory Level Comprehension Level Question History:

SXD Review 7/21 - Maybe SRO level question, maybe a direct lookup 7/27 I don't think it's a direct lookup Look up Lesson Plan Objective AF 8/23 normally operating with 3 SAC'S pumps, NOT an RO Question, replace.

MB - 9/27 - rewrote question SXD - OK AF - minor editorial change - added outside the drywell MB - Incorporated changes

7SRO RO I Tier#

Importance 3 Group# 1 ]

L3g Hope Creek RO Exam Nov 2005 -

3 Y

s as they apply to Instrument Air Compressor L

Power supplies:(CFR:

41.7145.5/45.6) 4 Question Y Given the following conditions:

Hope Creek is starting up from a Refueling outage, the plant is currently in OPCON 3 with temperature at 240°F and with the Instrument Air pressure at 105 psig and the InstrumenVService Air Compressors are aligned as follows:

Compressor Control Mode Status 00K107 MAN Running 10K107 MAN OFF 1OK100 AUTO OFF A Maintenance Worker accidentally bumps into 7.2KV Bus 10A120 causing it's input breaker to open and the bus to de-energize.

Assuming NO operator actions, which of the following correctly states the expected response of the Instrument/ Service Air systems?

A Service Air compressor 10K107 de-energizes, Instrument Air header pressure remains at 105 psig.

Y-B Service Air compressor e-energizes, Instrument Air header pressure drops to 92 psig, when Service Air Compressor 1OK10 Service Air compres A

returns pressure to -95 psig.

e-energizes, Instrument Air header pressure drops to 85 psig when x

Emergency Air Corn

- \

00 starts and returns pressure to -1 05 psig.

x Service Air compress r 00K107 eenergizes, Instrument Air header pressure drops to 85 psig when Emergency Air C o m p KlOO u starts and returns pressure to -95 psig.

Answer D References NOHOISERAIR-01, SERVICE AIR SYSTEM, p.4748 NOH01INSAIR-01, INSTRUMENT AIR SYSTEM, p l 5 , 4 2 Justification References during Exam None &$g 7

• z ILd A. INCORRECT - Power to SAC 10K107 is from 7.2 KV bus 10A110, NOT 10AI20 B. INCORRECT - SAC 10K107 will NOT start at 92 psig because it's in M A N Control. --

C. INCORRECT ElAC 10K100 will auto start at 85 psig, however, it unloads at 100 psig, thereby making in NOT capable of raise pressure to 105 psig.

D.CORRECT Loss of Power to 10A120 causes a loss of Power to SAC 00Kl07, Instrument Air header pressure drops to 85 psig, when ElAC 10K100 starts and brings pressure back to some value c 100 psig.

-~ ~ ~

Question Source New 0Memory Level PI Comprelzension Level Question History:

SXD reviewed 7/25 - minor editorial changes to stem and distractor B changed 105 psig to 95 psig.

AF - 8/23 - 3 Instrument air questions - 1 question contradics this one. 105 psig isn't normal for instrument air.

Changed to OK AF - 10/13 minor change MB - incorporated change

Hope Creek RO Exam Nov 2005 -

aRO [Tier # 1 Group# 1 I 7 1SRO Importance 3.4 7-

,JO21 A42.05 Loss of Shutdown Cooling / 4 Ability to determine and interpret the following as they apply to Loss of Shutdown Cooling Reactor Vessel Metal Temperature (CFR: 41.10 z

l43.5f45.13) 2 Question Given the following conditions:

- The reactor has been shutdown for 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> following 1000 EFPD of operation. Ab9

- The plant is in Cold Shutdown with RPV metal temperature of 140°F.

- A total loss of Shutdown Cooling occurred at 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br />.

- All efforts to restore heat removal from the RPV have failed.

- Both Recirculation pumps have been secured.

Assuming NO additional operator action, when will the plant reach OPCON 3?

A 1245 x

g 1307 1330 c

1352 J

Answer E References -

Hope Creek Question Q61328. HC.OP-AB.RPV-0009, Figure land Technical Specification Table 1.2 Justification References during Exam Figure 1 of HC.OP-AB.RPV-0009 Justification 1307- correct- Operational Condition 3 is achieved when the Reactor temperature reaches 200°F. The 140'F curve of Figure 1 intersects the 90-hour line between the 1.OOO and 1.250 hour0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> lines. 1307 is the only option that is between 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and fifteen minutes following the loss of SDC.

1245. incorrect- Value obtained by using t h e m c u r v e .

1330. -incorrect- Value obtained by using the 120°F curve.

J 0 clc?,

1352. -incorrect- Value obtained by using the 100°F curve.

Question Source Bank 0Memory Level E] Comprehension Level Question History:

SXD Review 7/21 OK -

AF - 8/23 - WA mismatch - make NO caps SXD put metal temp's in stem MB made changes as requested.

AF-OK

lQuestion# 10 I Hope Creek RO Exam N O2005 - ~

pier# Group# 1 1 3 Importance 3.4 7 A023 Refueling ACC/ 8 2 AK2.03 Knowledge of the interrelations between Refueling Accidents and the following Radiation Monitoring equipment (CFR41.7 /45.7/

45.8)

Y Question 5bPtc W S ~ Ir Scp Given the following conditions:

- The plant is in a refueling outage with a fuel move in progress.

- The 'A' Refuel Floor Radiation Monitor has failed downscale. NO actions have been taken to address this failure.

- At time 0000 a fuel bundle is dropped and radiation levels on the refuel floor start to slowly rise.

- At time 0005 the B Refuel Floor Radiation Monitor reaches its Hi Trip Setpoint.

- At time 0010 the C Refuel Floor Radiation Monitor reaches its Hi Trip Setpoint.

Under these conditions, an automatic trip of the Reactor Building Ventilation Exhaust (RBVE) fans due to Hi Refuel Floor Radiation levels:

A will occur at time 0010.

/ L R , T W L7 1 B will NOT occur due to the 'A' Refuel Floor Radiation Monitor being failed downscale will occur at time 0005.

will NOT occur until at least 1 Reactor Building Exhaust radiation monitor senses high radiation.

J Answer A References INPO Question 25978 NOH04000221C-01, RADIATION MONITORING SYSTEM p. 29 HC.OP-SO.SM-OOO1 Justification References during Exam None k - CORRECT - Per lesson plan p.29 item g. Automatic actions on a Refuel Floor Exhaust RM-23A HIGH radiation intensity level (any two of the three) RBVE fans trip.

B INCORRECT - still have 2/3 monitors available C INCORRECT - since A channel is failed downscale, need 213 to get actuation. Therefore won't get actuation when B channel gets high signal.

D INCORRECT - will get a trip of RBVE fans on either Hi Refuel Floor Rad levels and RBVE rad levels.

Question Source Mod Memory Level Comprehension Level Question History:

SXD Review 7/21 - Changed Distractor D to make it clearer AF - 8/23 all caps NO in 2nd bullet, minor editorial changes MB - 81 24 Made changes as requested AF-OK

\

- -* \\

Hope Creek -

RO Exam Nov 2005

@RO ]r# 1 ~roup# 1 1 1SRO Importance 4 7-EA1.03 Ability to operate and/ or monitor the following as they apply to LPCS High Drywell Pressure 7.5-Question The A Core Spray pump is in full flow test mode in accordance with HC-0P.IS.BE-0001, Core Spray Pumps A and C Inservice Test. A steam leak in the drywell has caused the following conditions:

- Reactor scrammed and all rods inserted.

- RPV level lowered to -60 inches and is now rising with HPCl

- Drywell pressure is 3.0 psig rising.

- RPV pressure is 800 psig lowering.

- Offsite power remains available to the 4KV buses.

Based on the above conditions. which one of the following is the correct response of the Core Spray system?

B j, "A" Core Spray pump continues to run in full flow test, all others are operating in min flow.

ALL Core Spray pumps are operating on min flow.

x ALL Core Spray pumps are tripped and ALL pumps will start when RPV pressure lowers to 461 psig.

C ALL Core Spray pumps are injecting.

J Answer 13 References INPO Question 24762 NOHOICSSYSO-01, CORE SPRAY SYSTEM

~

Justification References during Exam None A - INCORRECT - Core spray full flow test valve doses upon Receipt of a CSS initiation signal.

B CORRECT Core Spray received a start signal at DW pressure > 1.68 psig. This caused all Core Spray pumps to start, however, RPV pressure is > 461 psig so upstream injection valves are closed and pumps are operating on their mini-flow valves. Core Spray test valve auto closed upon receipt of a CSS initiation signal.

C INCORRECT - Core Spray pumps receive a start signal with pressure > 1.68 psig.

D - INCORRECT - Core Spray pumps upstream injection valves don't open until RPV pressure is < 461 psig.

"initiation" pump start signal is reached., A Core Spray running, NO trip signal to any CS pumps and NO loss of power., NO Core Spray "initiation" pump start signal is reached., Correct, > 2 psig signal closes full flow test valve Question Source Mod 0Memory Level kd Comprehension Level Question History:

SXD review 7/21 - OK 812 JD - Minor editorial change to "A" distractor - Incorporated AF - 8/23 bulletize laundry list MB - 81 24 Made changes as requested AF Minor changes "3 - Incorporated changes

[Question # 12 I Hope Creek RO Exam Nov 2005 -

rTier # 1 roup # 1 I Importance 3.8 A025 High Reactor Pressure I 3

'2 EA1.02 Ability to operate and I or monitor the following as they apply to High Reactor Pressure Reactorflurbine pressure regulating system :(CFR:

41.7145.51 45.6) 3 Question 3 Hope Creek was operating at 75% power when a loss of feedwater healing OCCUE.

Assuming NO operator action, which of the following describes the effect on Reactor pressure and Main Tuhine Pressure regulating system response:

Reactor Pressure will I , which will cause the Main Turbine Pressure regulating system to send a signal to the Control valves to II .

A I. Increase

11. Close Y B 1. Increase II. Open x

1. Decrease

11. Close I. Decrease J 11. Open Y

Answer 0 References NOHOlEHCLOG-02, EHC CONTROL LOGIC, p.8 Justifcation References during Exam None A - INCORRECT - when a Loss of FW heating occurs, colder FW will be sent to the reactor. This colder feedwater will cause a collapse in voids and a decrease in moderation, causing Reactor Power to increase. As reactor power increases, Reactor Pressure will increase. The increase in Reactor Pressure will cause the EHC system to OPEN the control valves to stabilize Reactor pressure.

B CORRECT C - INCORRECT see "A" above D INCORRECT see "Aabove Question Source New Memo y Level Comprehension Level Question History:

New Question 9/7 SXD - OK AF Possible KIA mismatch, doesn't address W A abnormal, SXD to resolve

@RO [Tier# 1 Group# 1 1 r!SRO Importance 3.9

,JO26 Suppression Pool High Water Temp. I 5 EG2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 45.2 /

45.6)

Question Given the following conditions:

- An ATWS is in progress

- APRM's read 10%

- Manual rod insertion is in progress

- MSlVs are closed

- Pressure is being maintained at 850 psig using SRVs

- Suppression Pool temperature is 195°F and rising at 1"F/5 min.

- Suppression Pool level is 70" and lowering at 1"/20 min.

- Suppression Pool pressure is 22 psig and rising at 1 psi/l5 min.

Based on the conditions above, which of the following describes the appropriate action and the reason for that action?

c r w '. +Lh",QL5P  ?& I _

B Emergency Depressurize to prevent exceeding the HCTL.

x Emergency Depressurize to prevent exceeding the PSPL. ~ - r G 4 ? O ~ B/ ML E~O - L b d . 1 4 -

Reduce RPV pressure to prevent exceeding the HCTL.

J Answer D References Hope Creek Question 462056 HC.OP-EO.=-101, Reactor Pressure Vessel Control HC.OP-EO.ZZ-010 BWR Owners Group EPGs/SAG Appendix B Section 5 -

Cautions Justifcation References during Exam EOP-102 without entry conditions A INCORRECT - Initiate Suppression Chamber spray to reduce SC pressure below 9.5 psig. SC Sprays are NOT initiated once SC pressure exceeds 9.5 psig.

B - INCORRECT With RPV pressure at 850 psig and SP temperature at 195 "F and rising at 5"F/min, the HCTL will be exceeded in 35 min. IAW Step SPTT-9, a pressure reduction prior to an ED is wamnted.

C - INCORRECT -SP level is at 22" and rising at 1"/15 min. and SP level is at 70 " and lowering at 1"/20 min. Since at these rates of change, it will be about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> before the PSPL is exceeded, an ED is NOT yet appropriate.

D - CORRECT - With RPV pressure at 850 psig and SP temperature at 195 O F , rising at 5"F/min. the HCTL will be exceeded in 35 min. and with power still at 10%. IAW Step SPTT-9 a pressure reduction is appropriate.

Question Source Bank 0Memo y Level Comprehension Level Question History:

SXD review 7/21 - OK AF 8/23 - removed comment, couple of NPSH questions, need EOP Caution 2 MB 8/24 Made changes as requested MB - 9/27 - after looking at NPSH questions decided this question asked essentially the same information as (21.5, changed out this question.]

SXD - OK 4F - added References to EOP-102, need to look at graph to get answer.

' - added references

13 RO [Tier # Group # I 3 1SRO Importance 3.9

,5028 High Drywell Temperature / 5 Y EG2.1.30 Ability to locate and operate components, including local controls. (CFR: 41.7 / 45.7) 4 Question Given the following conditions:

- A Large Break LOCA has occurred in the Drywell.

- " B RHR has been aligned for Drywell Spray.

- Subsequently, the Control Room needs to be evacuated and the Remote Shutdown Panel (RSP) manned.

- After taking local control at the RSP, ALL Transfer switches are placed in EMER.

What is the EXPECTED status of Drywell Spray and what Controlllndication does the Operator have over "6"Drywell Spray valves at the RSP?

II.

A I. Drywell spray continues. %

II. Operator has both Control and Indication over the "8" Drywell Spray valves at the RSP.

B I. Drywell spray continues. X II. Operator has only Indication over the "8"Drywell Spray valves at the RSP.

I. Drywell spray is terminated.

II. Operator has both Control and Indication over the " B Drywell Spray valves at the RSP.

f

~~ ~ ~~~ ~ ~ -~ ~

1. Drywell spray is terminated.

II. Operator has only Indication over the "B"Drywell Spray valves at the RSP. 1/

Answer References Hope Creek Question - (256161, EOP- Caution 1, LP 0302-000.00H-00134-13 Obj 8 HC.OP-I0.Z-0008, p. 8 NOHOlREMS/D-Ol, P.1421 HC.OP-EO.U-O1O1, RPV Control Justification References during Exam None A- INCORRECT - Drywell spray is terminated when control is transferred to the RSP B - INCORRECT - Drywell spray is teminated when control is transferred to the RSP C INCORRECf - Operator has ONLY indication over the "B" Drywell Spray valves at the RSP D - CORRECT - Drywell spray is terminated and the operator only has indication of the "B" Drywell spray valves at the RSP.

-~

Question Source New Memory Level Comprehension Level Question Hktory:

SXD review 71 21 OK -

JD 8/2 - K/A - Locate & Operate - asked to write question to J. Munro about Locate & Operate question AF - 8/23 weak K/A mismatch NOT -

MB - SXD to resolve SXD use pump and valve numbers in distractors. perhaps re-sample MB - 10/3 - Changed question stem to ask for location of equipment in addition to reason for operating.

SXD - OK AF - added HV-FO214 possible WA mismatch ME added HV-F021A, SXD to resolve KIA 7 - 11/01 - re-wrote question to address AF concerns

&? RO pier# 1 Group# 1 1 SRO Importance 3.5 I

~~ ~

Question The plant has experienced a transient and the following is observed:

- Suppression ChamberfWqressure: 9 psig

&xo c .a

- Suppression Pool temper ure: 240 degrees F

- Suppression Pool level at 0" 2-

- Reactor pressure: 100 psi

- RHR " A pump flow: 1 000 gpm

- Core Spray "B" pump Flow: 1500 gpm

- All other low pressure ECCS pump are NOT in service.

AW

• f e d A There is sufficient NPSH for the "E" Core Spray Pump ONLY.

& t/

B There is sufficient NPSH for the "A" RHR pump ONLY.

x There is sufficient NPSH for both the "A"RHR pump and the "B" Core Spray Pump.

C There is NOT suffiuent NPSH for any pump.

J Answer A References INPO Question 14383 EOP CAUTION 2 Justification References during Exam EOP Caution 2 Using EOP Caution 2 and realizing that being above the curve is the area of Unacceptable operation:

The limiting temperature for CS pump at 5 psig and 1500 gpm = 232°F The limiting temperature for CS pump at 10 psig and 1500 gpm = 244°F Interpolating for 9 psig gives a Temperature limit of -242°F for 9 psig. Since given temperature = 240°F this puts the B CS pump in the area of ACCEPTABLE operation.

The limiting temperature for RHR pump at 10 psig is 235"F, since Given temperature is 240°F this puts the pump in the region of UNACCEPTABLE operation.

This makes ONLY Answer A CORRECT.

Question Source Mod 0Memory ~ e v e l ComprelzensionLevel Question History:

SXD reviewed 7/22 - OK AF 8123 another NPSH question, 74.5" drives them to AB.155, 2 correct answers as written, changed "D"to any pump. Perhaps change question 13 MB - 81 24 Made changes as requested AF - t o check Caution

[Question# 16 I Hope Creek RO Exam Nov 2005 - 2,

[r# 1 Croup# I 'L 3

,5031 EK2.10 Reactor Low Water Level I 2 3 Knowledge of the interrelations between Reactor Low Water Level and the foliowing Redundant reactivity control 7

Question Given the following:

- The plant is operating at 100% power.

- A transient results in a scram setpoint being exceeded.

- The Reactor Protection System fails to automatically scram the reactor.

Without operator action, which of the following describes how the Control Rods will be automatically inserted to shutdown the reactor via the ARI system? 7 A An RPV level of minus 50 (-50) inches will air header.

e . ENERGIZE the ARI valves to depressurize the scram

/

B An RPV level of minus 50 (-50) inches will -i DE-ENERGIZE the ARI valves to depressurize the scram air header.

An RPV pressure of 1050 psig will +mm&Bk& ' ENERGIZE the ARI valves to depressurize the scram air header.

c An RPV pressure of 1050 psig will DE-ENERGIZE the ARI valves to depressurize the scram air header.

Answer A References INPO Question 22776 NOH01RRCSOO-00. REDUNDANT REACTIVITY CONTROL SYSTEM (RRCS). p.8 Justification References during Exam None A CORRECT -with RPV level c -38" the ARI valves are energized to depressurize the scram air header resulting in rod insertion.

B. INCORRECT - valves are Energized to actuate, NOT de-energized.

C. INCORRECT - ARI pressure setpoint is 1071 psig, NOT 1037 psig D - INCORRECT - ARI pressure setpoint is 1071 psig, NOT 1037 psig.

Question Source Mod @lMemory Level 0 Comprehension Level Question History:

SXD review 7/21 - Add (via the ARI system) to the end of the stem. Removed "control rod insertion will begin within 15 ...) from all distractors AF - 8/23 2 correct answers - feels like 1071 psig will energize ARI valves. Changed all distractor and correct answer to a value vs. > than a number or less than a number.

ME - 8/24 Made changes as requested AF - OK

[Question# 17 Hope Creek RO Exam Nov 2005 RO I Tier# 1 Group# 1 I 2

-A037 Importance 3.8 z SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown /

1 3

EA1.02 Ability to operate and I or monitor the following as they apply to SCRAM Condition Present and Reactor Power Above APRM RRCS 3 Downscale or Unknown Question The plant was operating at 98% power when a transient occurred. Following the transient 3 SRVs opened. 2 minutes later, reactor pressure is stable with 1 SRV open. NO operator actions have been taken.

Which of the following is correct for these conditions?

Both Recirculation Pumps A have tripped.

B are running normally.

x are running at minimum speed x

are currently running but will trip in 1.9 minutes when a time delay times out.

&A bt5ll;cb~T*

7- chsct Answer A References INPO Question 23485 NOH01RRCSOO-00, REDUNDANT REACTIVlTY CONTROL SYSTEM (RRCS). p.13 Justification References during Exam None A CORRECT Following the transient, all SRVs opened. Reactor pressure has to be greater than 1071 psig for all valves to open. Reactor pressure greater than 1071 psig causes both Rearc Pumps to trip. A is the only correct answer.

6.INCORRECT plausible because may NOT have hit a trip condition.

C. INCORRECT plausible because recirc pumps have runbacks, operator may incorrectly believe a runback condition has been met.

D - INCORRECT plausible because a 3.9 minute timer does exist on RRCS. however it is for SLC initiation, NOT Recirc pump trip.

Question Source Mod 0Memory Level 0 Comprehension Level Question History:

SXD Review 7/21 - removed # of SRVs from stem. Removed "off from Distractor A AF - 8/23- changed all SRVs to 3 SRVs In stem MB - 81 24 Made changes as requested AF-OK

kRO [r# 1 Group# 1 I 3 1SRO Importance 3.9 3

3 A systems operated for RPV control are given a higher priority than stopping a rad release.

d-B isolation of a EOP support system requires an upgrade of the Emergency Classification.

C they are required to support alternate reactor depressurization methods.

x J

additional radiological consequences from them are unlikely.

X Answer A References INPO Question 25837 BWROG, EPGdSAGs Appendix B. section 9 Radioactivity Release control HC.OP-EO.U-103/4. Reactor Building & Rad Release Control Bases Document - p. 13 8 14 Justifcation References during Exam None Per EOP Bases document 103/104:

The objectives of RPV Control, Primary Containment Control, and the EPG contingencies are given higher priority than the obiectives of Radioactivity Release Control. Systems that must be operated to perform other steps of the EPGs are thkefore NOT isolated in this step.

A - CORRECT matches bases document B - INCORRECT NOT in accordance with bases document C - INCORRECT - NOT in accordance with bases document D - INCORRECT - NOT in accordance with bases document Question Source Bank Memory Level 0 Comprehension Level Question History:

SXD review 7/22 - Minor editorial changes (added procedure)

AF - 8/23 - NO comments

[Question# 19 1 Hope Creek RO Exam Nov 2005 -

%I# 1 Group# 1 1 ImDortance 2.5

,3000 Plant Fire On Site / 8 AK1.O1 Knowledge of the operational implications of the following Fire Classifications by type concepts as they apply to the Plant Fire On Site (CFR: 41.8 to 41.10/45.3),

-~ ~

Question A fire occurs in the Upper Cable Spreading Room (Control Equipment Mezzanine Room 5403).

- The installed fire protection system automatically actuates.

- The room must be entered to determine if the tire has been extinguished.

(1) What is the classification of the fire that is expected in this area?

AND (2) What safety hazard, from the automatic system actuation, shall be considered prior to operators entering the Cable Spreading Room?"

A (1)ClaSS c (2) Suffocation from oxygen depletion due to the discharge of C02 in the area B (1)Class B (2) Suffocation from oxygen depletion due to the discharge of halon in the area (1) Class C (2) Suffocation from oxygen depletion due to the discharge of halon in the area (1) Class B (2) Suffocation from oxygen depletion due to the discharge of C 0 2 in the area Answer A References INPO Question 24855 NOHOIFIRPRO-02, FIRE PROTECTION, p.55, p. 63 and p.85 Justification References during Exam None A- CORRECT - Class C fire due to electrical equipment in area, Suffocation due to discharge of C 0 2 B - INCORRECT - NOT a Class B fire and NO halon in that room C INCORRECT - NOT expecting to get a halon discharge in that room D - INCORRECT NOT a class B fire Question Source Mod Memory Level 0 Comprehension Level Question History:

SXD review - 7/21 - Changed water to Halon AF - 8/23 - OK

(Question # 20 I Hope Creek RO Exam Nov 2005 - 'z

(# 7 Group# 1 I IT? SRO I Importance 3.3

,5005 Main Turbine Generator Trip / 3 A D . 04 Knowledge of the interrelations between MAIN TURBINE Main generator protection GENERATOR TRIP and the following: (CFR: 41.7/45.8)

Question Given the following conditions:

- The plant is operating at 20% power

- A main generator load reject has just occurred

- A fault in the control arcuit causes a power/load unbalance trip during the load reject Which of the following is the immediate expected response of the Turbine Control Valves (TCVs) and the Reactor Protection System (RPS)?

A TCVs throttle dose, RPS trips

>(

g TCVs throttle close, RPS does NOT trip

>(

C TCVs fast close, RPS trips x

TCVs fast close. RPS does NOT trip d

Answer D References Hope Creek Question - (261307, HC.OP-AB.BOP-0002 Additional Information /Automatic actions and notes NOH01MNTURB-02, MAIN TURBINE CONSTRUCTION AND COMPONENTS, p. 66 Justification References during Exam None CORRECT - TCVs fast dose, RPS does NOT trip. The load reject causes the TCVs to fast dose. The fast dosure does NOT initiate a RPS trip because turbine load is <30%. Since power is within the capacity of the BPVs, NO pressure transient will trip RPS.

INCORRECT - TCVs throttle dose, RPS does trip. The load reject causes the TCVs to fast close. The fast closure does NOT initiate a RPS trip because turbine load is <30%. Since power is within the capaaty of the BPVs, NO pressure transient will trip RPS.

INCORRECT TCVs fast close, RPS does trip. The fast closure does NOT initiate a RPS trip because turbine load is

<30%. Since power is within the capacity of the BPVs, NO pressure transient will trip RPS.

INCORRECT - TCVs throttle close, RPS does NOT trip. The load reject causes the TCVs to fast close Question Source Bank c3 Memoty Level 0 Comprehension Level Question History:

SXD review - 7/21 - OK AF 8/23 - OK AF 10/13 possible WA mismatch

[Question # 21 Hope Creek -

RO Exam Nov 2005 ITier # 1 Group # 2 I Importance 3.2

~25002 Loss of Main Condenser Vac 1 3 7 AA2.02 Ability to determine and interpret the following as they apply to Loss of Main Condenser Vacuum's Reactor Power Plant Specific:(CFR:

41.10/43.5/ 45.13) 3

.k,?d*--

fi Question cc Given the following:

- All four Circulating Water Pumps are in operation

- Plant is operating at 100% power

- Circulating Water System Inlet temperature is 80°F

- Indicated Main Condenser pressure is 2.75 in HgA c+n=

Then, AP501 is removed from service.

Assume the remaining Circulating Pumps' Discharge Valves are reopened fully, NO rise in basin temperature and NO other operator actions are taken.

What is the expected condenser backpressure and what is the expected change in reactor power (if any) following the removal of Circulating Water Pump AP501 from service?

-~ ~~~

A 3.5 in HgA, reactor power increases (ie. Greater than 2%)

B 3.5 in HgA, reactor power stays the Same (ie. Doesn't change more than 2%)

.J 4.15 in HgA, reactor power increases (le. Greater than 2%)

>(

4.15 in H g 4 reactor power stays the same (ie. Doesn't change more than 2%)

3 Answer References -

Hope Creek Question Q55132 HC.OP-SO.DA-0001, Rev. 35, Attachment 5 Justification References during Exam Attachment 5 from HC.OP-SO.DA-0001 A INCORRECT - Reactor power should NOT change with a decrease in vacuum. If anything reactor power may go down a little bit due to increased condenser temperature and reduced condenser subcooling B- CORRECT- 3.5 inHgA If CW inlet temp does NOT change, then the condenser vacuum rises vertically on the graph until it reaches the line for three pump operation @ 80 degF. Since the inital back-pressure of 2.75 indicates 100 percent CF. Reactor power should remain the same C - INCORRECT 4.15 - 3 pump ops at 70 percent CF.

D - INCORRECT 4.15 - 3 pumps ops at 70% CF Question Source Bank Memory Level 0 Compeltension Level Question History:

SXD review 7/21 - OK JD 812 K/A asking for Reactor Power 813 initially was going to change question to add reactor power change, decided to ask Steve on Monday 814 Rewrote questions AF - 8/23 change G to capitalize in "A" MB 8/ 24 Made changes as requested AF minor changes MB - Incorporated changes

Hope Creek RO Exam Nov 2005 -

a RO per# 1 Group# 2 1 T-] SRO Importance 3.4

,5008 AK3.06 High Reactor Water Level / 2 RClC Turbine Trip 5

Knowledge of the reasons for the following responses as they apply to High Reactor Water Level 3 Question During a transient, the RO started the RClC system for reactor water level control using the appropriate operating procedure. Level rose above the High Reactor Water level at 54" after which it lowered below the Low Reactor Water level at -38".

Which of the following describes the reason for, and expected response of RClC during the reactor water level transient?

A The RClC Trip and Throttle Valve (HV-4282) will close on High Water Level and RClC will automatically restart on Low Reactor Water Level.

B The RClC Trip and Throttle Valve (HV-4282) will close on High Water Level and RClC will have to be reset and manually started on Low Reactor Water Level.

The RClC Steam Supply Valve (F045)will dose on High Water Level and RClC will automatically restart on Low Reactor Water Level.

J The RClC Steam Supply Valve (F045)will close on High Water Level and RClC will have to be reset and manually started on Low Reactor Water Level.

Answer References NOHDlRCICOM)2, REACTOR CORE ISOLATION COOLING SYSTEM, p22-23 Jusa$cation References during Exam None A INCORRECT - Trip and Throttle valve does NOT close on Level 8 B INCORRECT - Trip and Throttle valve does NOT close on Level 8 C - CORRECT Steam supply valve will dose and RClC will auto restart at Level 2 D - INCORRECT - RClC will auto restart at Level 2

~~ ~

@ Memory Level Comprehenswn Level Question History:

SXD review 7/22 - LOD = 1 - rewrite question 8/3 - rewrote question AF - 8/23 - changed 58" to 54" MB - 81 24 Made changes as requested AF-OK

t Question #

6d RO 23 pier#

1 1 Group # 2 I Hope Creek RO Exam Nov 2005 -

'Lsp 3

SRO Importance 3 Lf 3

Question The plant is currently at 27% power. Plans for the shift are to continue the startup and power ascension. A 3

malfunction in the Feedwater Control System has resulted in the following:

- RPV level is 25 inches and trending down

-Total Feedwater flow is 2.5 mlb/hr and steady

- 3 Circ Water pumps are running

- Condenser Vacuum is 3.8" HgA and degrading Assume NO operator actions have been taken. Which of the following statements is correct regarding the Reactor Recirculation system response based on these CURRENT plant conditions?

A Speed Limiter 1 (30% flow) is actuated to ensure Recirculation Pump net positive suction head protection based on feedwater flow.

B Speed Limiter 2 (45% flow) is actuated to ensure Recirculation Pump net positive suction head protection based on RPV level.

Speed Limiter 2 (45% flow) is actuated to bring Condenser Vacuum back to normal.

C 3

Speed Limiter 1 (30% flow) is actuated to bring Condenser Vacuum back to normal.

m /: w x Answer A References New Question NOHOlRECIRC02. Reactor Recirculation System. P. 53-55 Jusascation References during Exam None A CORRECT -Total FW flow is -17% which is 20%, this causes a Speed Limiter #I runback to ensure Recirc Pump NPSH B - INCORRECT - Speed Limiter 1 is actuated, NOT Speed Limiter 2 C INCORRECT - Speed Limiter 1 is actuated, NOT Speed Limiter 2 D INCORRECT - Condenser vacuum is rising but still within normal limits. Must be 4.5" to cause a Recirc pump runback.

Question Source New 0Memory Level Comprehension Level Question History:

SXD Review 7/21 minor editoral comments AF - 8/23 changed "A" to FW flow vs. RPV level, also AF to check numbers MB - 81 24 Made changes as requested - AF still to vefiQ Numbers SXD OK-AF-OK

Hope Creek RO Exam Nov 2005 -

97Q

@ RO F r # 1 Group# 2 1 Imvortance 3.1

,5029 High Suppression Pool Wtr Lvl I 5 EK2.07 Knowledge of the interrelations High Suppression Pool Wtr Lvl and the following Drywell/ containment water level:(CFR: 41.7 /45.7/45.8) 5 Question 3

An Override step in HC.OP-EO.=-0202, Emergency Depressurization, directs the operator to open the Inboard MSL Drain Valve (AB-HV-FO16) when Containment water level is expected to exceed 48 feet.

Which one of the following describes the reason for this action?

Opening the Inboard Main Steamline Drain Valve A maintains the availability of the Main Steamline drain path for reactor vessel pressure control if required.

B ensures as much heat energy as possible is rejected to the Main Condenser to minimize the dynamic loading on Containment.

x x

maintains Containment water level below the SRV solenoids by establishing a drain path from the reactor vessel to the Main Condenser.

ensures the SRV Tail Pipe Level Limit is NOT exceeded prior to emergency depressurization.

J Answer A References INPO Question 21944 BWROG EPGISAG'S App. B - P 326 HC.OP-EO.U-0202 flowchart HC.OP-EO.U-0202. Emergency Depressurization Bases, p.5 Justification References during Exam None A CORRECT - per the BWROG guidelines If primary containment water level rises above the elevation of the SRV solenoids, the SRVs may NO longer be operable. Other methods must then be used to control RPV pressure and prevent repressurization. Opening the inboard main steam line drain valve preserves the main steam line drains for future use.

B INCORRECT but plausible, while Opening AEHV-FO16 does NOT reject any heat to the Main Condenser it could reject heat to the condenser if the FO19 and F021 were open.

C INCORRECT but plausible, while opening AB-HV-FO16 does NOT necessarily maintain CNMT water level below the SRV solenoids. it may open a drain path to the main condenser.

D INCORRECT but plausible, while opening AB-HV-FO16 does NOT drain water from the steam lines, it could if both FO19 and F021 were open.

Question Source Mod Memory Level 0Compreliension Level Question History:

S X D reviewed 7/22 OK AF-OK

IQuestion # 25 Hope Creek RO Exam Nov 2005 -

I Tier# Group # 2 1 Importance 3.8 SO34 Secondary Containment Ventilation High Radiation I 9 EA1.01 Ability to operate and1 or monitor the following as they apply to Area radiation monitoring Secondary Containment Ventilation High Radiation system:(CFR41.7/45.5/45.6)

Question Refueling Floor radiation levels are rising as indicated by rising readings on the 3 refuel floor ARMS (New fuel storage vault channel A & B LRE-4813 A & B] and Spent fuel Pool ARM [RE-6607).

What automatic actions will occur if ALL of these 3 area radiation monitors reach their "HIGH" setpoint?

I. Control Room Annunciator "NEW FUEL CRITICAL RAD HIGH" (E6-A4) will ALARM II. Refuel Floor Evacuation Alarm is actuated 111. Reactor Building Ventilation System will Isolate IV. Filtration Recirculation Ventilation System will Auto Start.

Note - Question is looking for ONLY systemdalarms that receive inputs from these AREA Radiation Monitors.

A I I Only B I and II Only 11, 111 and IV C

I, II, 111 and IV

~

Answer B References NOH04000221C01, RADIATION MONITORING SYSTEM p. 14 Justification References during Exam None A - INCORRECT - In addition to the Evacuation alarm being sounded, CR will also receive the E6-A4. New Fuel Critical Rad High B - CORRECT -per lesson plan will receive both the Refuel floor evacuation alarm and the new fuel critical alarm.

C - INCORRECT - RBVS and FRVS don't receive an input from Area Rad alarms D - INCORRECT - RBVS and FRVS don't receive an input from Area Rad alarms Question Source New Memory Level 0 Conlprehenswn Level Question History:

New Question 9/3 SXD OK -

AF - minor changes MB - Incorporated changes

Hope Creek RO Exam Nov 2005 - Y RO 1 Tier # 1 Group # 1 SRO Importance 2-9 Y

,5036 EK1.01 Secondary Containment High Sump/Area Water Level / 5 Knowledge of the operational implications of the following concepts as they apply to the Secondary Containment High Radiation releases(CFR:41.8 to

'I5 Sump/ Area Water Level 41.10/45.3)

Question Given the following:

Tm M-M., I l l

- T h e RClC turbine is on fire and the Fire Brigade has been actively spraying water on RCIC.

- The Fire Brigade reports steam coming out of the RClC steam supply line.

- The Fire Brigade has just reported that the fire is under control and they should be securing shortly,

- RClC pump room (41 10) Floor level is 6" ( 4

- RHR Pump room "B" (4109) Floor level is 5" (y fZdfLJT\J"c5 7R=I.Cw

- RHR Pump room "D" (4107) Floor level is 5" ( F 4

- Core Spray Pump room "8" (4104) Floor level is 3" (

- Core Spray Pump Pump mom "D" (4105) Floor level is 3" (

7 m 1 p ( . A d d 5 pnau4o47

- "D" South Reactor Building Sump pump (DP-265) is tagged out for motor replacement

- Reactor Building HVAC Exhaust Rad level is 1.5 x 10-3 microcuries/ ml

- Refueling Floor HVAC Exhaust Rad level is 1.O x 10-4 microcuries/ml

- RBVS is running In addition to restoring floor levels to normal using all available sump pumps, which of the following correctly states the proper operator actions to be taken and the reasons for those actions:

I. Isolate all water discharging into the RHR pump rooms in order to terminate level challenges to RHR pump rooms.

II. Runback Redrc and initiate a manual scram.

111. Emergency Depressurize the Reactor in order to place primary in it's lowest possible energy state.

IV. Initiate FRVS and Isolate RBVS in order to prevenffminimize off-site releases due to gw5 WJ high radiation levels.

lQBJ i s /S*kr&

A I -Only B I I and IV d

II, 111 and IV I, II and 111 D

Answer B References HC.OP-EO.ZZ-0103/4, BASES pages 1, 3, 7.8, 10 Justification References during Exam HC.OP-EO.ZZ-O103/4 with entry conditions blacked out.

A - INCORRECT Would NOT want to isolate Fire Protection water discharging to the RHR pump room.

B CORRECT Max Safe OP limit has been exceeded in 1 areas, the RClC Room. Due to High radiation on in the reactor building you must assume RCS is discharging to Rx building from RClC steam line. A manual Scram needs to be initiated. Due to High rad, need to start FRVS C - INCORRECT - Max Safe OP limit has NOT been exceed in 2 or more areas. Therefore you don't want to Emergency Depressurize.

D - INCORRECT don't want to stop Fire Protection, don't need to Emergency Depressurize.

~~~~~~~~ ~

Question Source New 0Memory Level Comprehension Level

Question History:

New 9/20 SXD - Minor comments to read better MB - 9/26 - Made changes as requested

'F - suggests removing 8,C, D RHR pump sumps as they won't get water from Fire Protection unless really bad, NOT

'e where Rad is coming from

,d -talk to SXD perbaps, Re-sample MB - 10/27- rewrote question to have ALL affected equipment come from the South Reactor Building sump. Since all this equipment is located on the bottom of the Reactor Building it seems plausible that with Fire Water coming into RClC pump room, sump pump could be overloaded and water could backup into other connected rooms through the sump. Rad is corning from RClC steam line.

Hope Creek RO Exam Nov 2005 -

aRO F# 1 Group# 2 I Importance 3.1

,JOOO High CTMT Hydrogen Conc. 1 5 3 EK2.02 Knowledge of the interrelations between High CTMT Hydrogen Conc. And the following containment oxygen monitoring systems(CFR:

a-41.7 I 45.7 145.8)

Question Given the following conditions:

Hope Creek has experienced a transient and the following conditions are present:

- Drywell H2 concentration is reading 1.5% by volume

- Drywell 0 2 concentration is reading 5.5% by volume

- Drywell Pressure is 1.5 psig and stable

- Reactor water level is +IO" and rising slow Assuming NO other operator actions have occurred, what is the status of the H2102 monitors?

Assuming the above readings are correct and Containment venting CANNOT be performed, what actions shall be taken with regards to the H2 Recombiners in accordance with HC.OP-EO.U-0102, Primary Containment Control?

A H2/02 monitors are in-service and the ti2 Recombiners shall be placed in service.

~~

/-

~~

B H2102 monitors are isolated because of Containment Isolation. however the H2 Recombiners shall be placed in service.

H2/02 monitors are isolated because of a Containment Isolation. however the H2 Recombiners shall NOT be placed in service 3(

H2102 monitors are in-service and the H2 Recombiners shall NOT be placed in service.

J Answer A References NOH01H202AN-01, Hydrogen Oxygen Analyzer System - p. 17 NOHOlH2RECM-00, CONTAINMENT HYDROGEN RECOMBINER SYSTEM, p.8 HC.OP-EO.ZZ-OlOP(Q)-FC, PRIMARY CONTAINMENT CONTROL, step PCIH-1 Justification References during Exam EOP-102 A. CORRECT - H2 Recombiners shall be placed in service due to High H2 concentration per EOP 102, mncenhtion

> 0.5% and e 2%

B. INCORRECT - Never received Containment Isolation, CNMT Isolation pressure is 1.68 psig C. INCORRECT - H2 Recombiners shall be placed in service due to High H2 Concentration per EOP 102 D. INCORRECT - H2 Recombiners shall be placed in service due to High H2 Concentration per EOP 102 Question Source New Memory Level Comprehension Level Question History:

SXD revlew 7/21 - OK AF - 8/23 - feels like "A" should be correct answer, for 0 2 monitors to be reading anything they would have had to been overriden and placed back in service. Normally monitors are NOT in service and they would be reading. Changed stem to Drywell pressure of 1 psig and made "A" the correct answer. AF-to relook at question. Have SD look at question again. Made all monitors H2102 monitors SD - Change justification MB 9/26 - Made changes as requested.

SXD AF to relook at question AF- change OPERABLE to In-service added EOP-102, operators will already have procedure anyway

\AB Incorporated changes

~~

piestion # 28 I Hope Creek RO Exam Nov 2005-per# 2 Group# 1 nSRO A3000 Importance 3.6 RHWLPCI: Injection Mode 3

A I .04 Ability to predict and/or monitor changes in System pressure Question parameters associated with operating the RHWLPCI:

INJECTION MODE (PLANT SPECIFIC) controls including:

(CFR: 41.5 / 45.5) 7 Hope Creek was at 100% when a Small Break LOCA occurred concurrent with a loss of ALL High pressure injection.

The Reactor is being depressurized using the SRVs due to level NOT being able to be maintained above TAF. ALL LPCl and Core Spray pumps have started as required. Reactor Pressure is 400 psig at this time.

Concerning the "A" RHR system ONLY, which of the following correctly describes the EXPECTED system parameters and configuration.

A LPCl "A" Injection valve is OPEN, "A" system flow indicates 10,000 gpm, "A" pump discharge pressure is approximately 400 psig.

B LPCl "A" Injection valve is OPEN, "A" system flow indicates 0 gpm, " A pump discharge pressure is approximately 340 psig.

/-

LPCl *A" Injection valve is CLOSED, "A" system flow indicates 2,300 gpm, "A" pump discharge pressure is approximately 340 psig.

LPCl "A" Injection valve is CLOSED, "A" system flow indicates 0 gpm, "A" pump discharge pressure is D appoximately 340 psig.

Answer B References Brunswick NRC Exam 2003, Q. 10 NOH01RHRSYSC-03, RESIDUAL HEAT REMOVAL SYSTEM, p.410, 53 Justification References during Exam None A - INCORRECT - System pressure is > RHR shutoff head, RHR flow should read 0 B - CORRECT - Injection valves open when system pressure c 450 psig. however Reactor pressure is still > shutoff head of pump, therefore indicated flow = 0 gprn.

C - INCORRECT - LPCl Injection valves OPEN when Rx pressure c 450 psig.

D - INCORRECT - LPCl Injection valves OPEN when Rx pressure < 450 psig.

Question Source Mod Memory Level @I Comprehension LeveI Question History:

9i7 - New - had to re-sample KIA, initial WA was NOT an RO level. RO's NOT required to know bases of Tech Specs for RHR.

SXD - Check Hope Creek References for correct pressures MB - Per HC lesson plan LPCl pump shutoff head = 366 psig, normal pressure 171 psig wl10,OOO gpm flow. Also per lesson plan LPCl Injection valves OPEN when reactor pressure lowers to 450 psig.

SXD - OK AF - changed pressure to 340 psig, changed valves to valve MB - Incorporated change

IQuestion # 29 I Hope Creek RO Exam Nov 2005 -

I #2 Group# 1 7 7SRO ImDortance 3.5

-.~5000 Shutdown Cooling A3.03 Ability to monitor automatic operations of the Shutdown Cooling System(RHR Shutdown Cooling Mode) including lights and alarms (CFR:41.7/45.5) 3 Question 3

Given the following Plant conditions:

Hope Creek is in OPCON 3 Cooling down for a Refueling Outage, "A" Shutdown Cooling is being placed in service and is currently in the following status:

- "A" RHR fill and vent ha ompleted. However, the F007A - RHR Pum valve's breaker wa inadverntantly

- " A RHR Loop has bee

- Both Reactor Recirc Pumps have been secured.

The RO is lining up "A" RHR system for Shutdown cooling and valves are currently lined up as follows:

- FOO9 - Shutdown Cooling INBD ISLN MOV - Open

- F008 - Shutdown Cooling OUTBD ISLN MOV - Open

- AP202 RHR PUMP - Running

- F015A - RHR Loop A Ret to Recirc - Throttled Open

- F007A - "A" RHR pump mini-flow - Closed

- F024A - "A" RHR Full Flow test valve - Closed

- F027A - "A" Torus Spray lnj valve - Closed "RHR A S/D CLG & MIN FL VLV OPEN" alarm is received in the control room.

• ssuming NO Operator actions are taken, which of the following conditions will result:

Y\

A FOO8 and FOO9 will Auto Close when the ni-flo? valve FOO7A begins to open.

~~~ ~

B FOO8 and FOO9 will Auto Close on Low RPV level 3 (+12.5")

C NO Auto Actions will occur, this is an expected alarm for the above conditions.

x FOO8 and FOO9 will Auto Close on Low RPV level 1 (-129")

D Answer B References NOH01RHRSYSC-03, RESIDUAL HEAT REMOVAL SYSTEM,

p. 30 Justification References during Exam None A - INCORRECT FOO8 and 9 will NOT Auto Close based on mini-flow valve position B - CORRECT - Having the Mini-flow valve open and taking suction from Reactor vessel will cause Reactor Vessel to lower, when vessel level reaches Low RPV Level 3. F008 and 009 will Auto Close.

C INCORRECT Reactor vessel will lower due to Mini-flow open and taking suction Reactor vessel.

D INCORRECT FOO8 and FOO9 will auto close on Low RPV level 3 and level should NOT get to Low RPV level 1.

Question Source New 0Memory Level PI omp prehension eve^

Question History:

SXD review 7/27 - OK

- 8/23 - made all bullets, changed "A" to when F007A starts to open vs. gets full open. AF feels this could be a rect answer because 3 minutes after valve gets full open, the valves will close on low level.

1318- 8/24 - Made changes as requested SXD - OK AF - OK

T a RO I Tier# 2 Group# I 3 7SRO Importance 2.8 3

3 7

~

A TACS will transfer to SACS loop "B". Later TACS ammchwn ' will l&utomaticaIIy isolate 1

B

~~

TACS will remain aligned to SACS loop "A".

x TACS will transfer to SACS loop " B . TACS is only isolated manually.

C

~~~~ ~ ~

TACS to SACS connections will immediately isolate on Low-Low-Low level in "A" SACS expansion tank.

D Answer A References HC.OP-AB.COOL-0002 Hope Creek Bank a56926 Justification References during Exam None A - CORRECT Low-Low-Low level will cause TACS to transfer. The leak will not be isolated so on a low-low-low level in the B SACS loop expansion tank. TACS will isolate.

B - INCORRECT -Will transfer to SACS loop "B" C - INCORRECT TACS will automatically isolate on low low low level in the "B" expansion tank.

D - INCORRECT - Will transfer to SACS loop "B".

~~~~~~ ~ ~

Question Source Bank Memory Level Comprehension Level Question History:

New 9/7 SXD - OK RJC - 10/6 - SRO rather than RO ( are they required to Know bases)

SXD - wait on Archie, find part in NUREG where only match 1 part of 2 part WA MB - Found part - ES-401 pages 5 8 6 AF - similar question on audit, SXD to resolve SXD - Re-sample MB - 10/27 - re-sampled - bank Question

louestion # 31 Hope Creek RO Exam Nov 2005 2

z 2

- JdOOO K5.05 HPCl z Knowledge of the operational implications of the following Turbine speed control concepts as they apply to the HPCl 3

Question Given the following conditions:

-The HPCl system running in automatic at rated flow.

-The flow element providing feedback to the flow controller begins to fail downscale, slowly.

How will actual HPCl turbine speed and system flow respond?

A Turbine speed will increase and flow will increase I/

B

~ ~~~

Turbine speed will decrease and flow will decrease

~ ~

Turbine speed will decrease and flow will remain at rated x

V A

Turbine speed will increase and flow will remain at rated JJ Answer A References Hope Creek Question Q56448 NOH01HPC10042. HIGH PRESSURE COOLANT INJECTION SYSTEM, p.30 Justification References during Exam None Correct answer:turbine speed will increase and flow will increase The following distractors are incorrect as follows:

turbine speed will increase and flow will remain at rated-Incorrect- As flow feedback lowers, controller will raise turbine speed and, with it actual flow rate will raise turbine speed will decrease and flow will decreaselncorrect- As flow feedback lowers, controller will raise turbine speed and, with it actual flow rate will raise turbine speed will decrease and flow will remain at rated-Incorrect- As feedback lowers, controller will raise turbine speed and, with it actual flow rate will raise Question Source Bank 0Memory Level L? Comprehension Level Question History:

SXD review - 7/21 OK -

AF - OK AF - K/A mismatch

Question # 32 Hope Creek RO Exam Nov 2005-1 Tier# 2 Group # 1 I SRO Importance 3

-dd001 LPCS K2.01 Knowledge of electrical power supplies to the following Pump power (CFR41.7)

Question Hope Creek has experienced a transient and a partial loss of Offsite power.

Current conditions are as follows:

- Red Lion Transmission Line (61x50) is DE-ENERGIZED with a ground fault on it

- 500KV Circuit Breaker BS1-3(61x) FAILED TO OPEN

/ c y GejLT5

- 13.8KV Circuit Breaker BS1-2 FAILED TO OPEN 6 B-J5

- Reactor has SCRAMMED and all rods are INSERTED

- Reactor water level is -135" and rising slowly AN'> reis

- Drywell Pressure is 1.334 and lowering slowly (Max. Pressure -1.W)

- "C" CS pump NORMWEMERGENCY TAKEOVER switch is(was) in the EMERGENCY position

- A and B Diesel Generators FAILED TO START Based on the above conditions. what is the status of the Core Spray Pumps?

A All Core Spray Pumps are running x

B A, 8, and D Core Spray Pumps are running d

Only C Core Spray Pump is running Only D Core Spray Pump is running

/Y Answer B References NOHOlCSSYS0-01, CORE SPRAY SYSTEM p.16 NOHOlEAC00-02. CLASS 1E AC POWER DISTRIBUTION 066-01: Class 1E AC Power Distribution (Training drawing) 027-01: Core Spray System (Training Drawing)

Justification References during Exam E-0001 A: INCORRECT "C" CS pump will NOT have started because it's Takeover switch is in the EMERGENCY Position B: CORRECT - The Loss of the Red Lion Line and the Circuit breaker faults will have caused a loss of Bus Section 1OX and Station Service XFMR 1BX501. however 1AX501 will still be energized from Offsite power, therefore power to 10A402 and 10A404 will auto transfer to 1AX501 causing all of the 4.16KV buses to be energized. As stated above "C"CS pump will NOT have started, leaving A, B and D CS pumps running.

C: INCORRECT - "C" CS pump will NOT have started because it's Takeover switch is in the EMERGENCY Position D: INCORRECT - A and B Diesel Generators failing to start will NOT cause their respective buses to be de-energized because they will have received power from 1Ax501 Question Source New 0Memory Level Comprehension Level Question History SXD reviewed 7/22 - give students 500KV switchyard print AF 8/23 - add bullets, made failed to open all caps. Made "B" correct answer, changed reference to be handed out to E-0001 to be given out.

MB - 8124 - Made changes as requested AF-OK

Hope Creek RO Exam Nov 2005 -

2,

- a 1000 SLC K4.04 Knowledge of SLC desiqn featurefs) and or interlock(s) which provide fG the following-Indication of fault in explosive valve firing circuits

'J (CFR41.7) 3 Question Hope Creek was operating at full power when an instrument air line break caused the outboard MSlVs to go closed. The following then occurred:

- The reactor failed to scram and attempts to drive rods were unsuccessful.

- The CRS ordered SLC injection.

- Both SLC pump AP208 and BP208 START pushbuttons have been depressed

- SLC pump control bezel start pushbuttons are backlit RED.

- The squib valve continuity lights are LIT.

- Pump discharge pressure is 1395 psig.

- Reactor Pressure is currently 1025 psig.

Based on these indications which of the following correctly describes the status of the SLC system?

d, with SLC pumps running therefore, SLC is NOT injecting.

SQUIB valves are OPEN, with SLC pumps running, therefore SLC is injecting.

L/

B

)f SQUIB valves are OPEN, however, the SLC pumps are NOT running, therefore SLC is NOT injecting.

c U

ed AND SLC pumps are NOT running, therefore, SLC is NOT injecting.

J

)c Answer

• References INPO Question 20790 NOH01SLCSYS-00. STANDBY LIQUID CONTROL SYSTEMS, p.27-29 Just$cation References during Exnm A - CORRECT - the pump control bezel start pushbuttons backlit RED, along with pump discharge pressure of 1395 psig indicate the pumps are running. Squib valve continuity lights being lit, indicate valves are dosed, therefore NO injection is occurring.

B - INCORRECT - Squib valves are closed C - INCORRECT Squib valves are closed D INCORRECT - SLC pumps are running Question Source Mod [3 Memory Level @ Comprehension Level Question History:

SXD review 7/21 Minor editorial changes AF - 2nd bullet - CRS NOT SS, added give figure of Control Bezels - AF to find figure number. Made LIT all caps.

MB - 8/24 - Made changes as requested SXD - Don't need figure MB - Removed figure from references SXD -OK AF-OK

louestion I# 341 Hope Creek RO Exam Nov 2005 -

I Tier # 2 Group# 1 I 3 Imvortance

- I2000 RPS 3

Y K3.11 Knowledge of the effect that a loss or malfunction of the RPS will have on the following Recirculation system (CFR41.7145.6)

L1 Question 3

Given the following: peg 5+T

- The Reactor is initially At@ power

-The Main Turbine is synchronized to the grid and loaded

-The RX RECIRC PUMPS RPS TRIP BYP alarm (Cl-E3) is NOT illuminated

- A loss of 8RPS Bus has occurred What is the operational effect of a fast closure of all Turbine Control Valves during this condition?

A

~~

EOC-RPT trip of Recirculation Pump A and NO trip of Recirculation Pump B x

J B EOC-RPT trip of both Recirculation Pumps EOC-RPT trip of Recirculation Pump B and NO trip of Recirculation Pump A C

Both Rearculation Pumps running with half-scram inserted J

Answer References Hope Creek Question - Q61263, HC.OP-AB.ZZ.IC-0003 discussion section step 2 NOHOlRECIRGOZ. Reactor Recirculation System, p.37 and p. 69 Justification References during Exam None Justification:

EOC-RPT trip of both Recirculation Pumps Correct, loss of RPS bus power, at any reactor power level, in conjunction with the cited Turbine Control Valve fast closure will result in EOC-RPT trip of both Recirculation Pumps.

This occurs due to a loss of the automatic bypass for EOC-RPT when less than about 30% power (first stage pressure less than 135.7 psig). The keylock bypass of the EOC-RPT trip is removed with the Main Turbine loaded. The RX RECIRC PUMPS RPS TRIP BYP alarm is deared when the RECIRC PUMP TRIP NB SYSTEM DISABLE switch is placed in the NORM position. This defeats the bypass of the RPT trips.

EOC-RPT t i p of Recirculation Pump A and NO t i p of Recirculation Pump B - Incorrect, both pumps will trip.

EOC-RPT t i p of Recirculation Pump B and NO trip of Recirculation Pump A Incorrect, both pumps will trip.

Both Recirculation Pumps running with half-scram inserted - Incorrect, both pumps will tip.

Question Source Bank 0Memo y Level Comprehension Level Question Histoy :

SXD Review 7/21 - OK AF-OK

Y Question # 35 Hope Creek RO Exam N O2005 - ~ 7 I Tier # 2 Group # 1 1 3 1SRO Importance 2.9 3

- 5003 7

I IRM K4.04 Knowledge of the IRM design feature(s) and or interlock(s) Varying system sensitivity which provide for the following levels using range switches (CFR41.7) Y Question Which of the following correctly explains how IRM system sensitivity level is varied using the range switches:

Placing the Range switch from Range 6 to Range 7 A changes which pulse height discriminators are placed in service.

>(

B changes which input resistors and which attenuators are placed in service.

>( a \

changes which pulse preamps are placed in service.

changes which voltage preamps and which attenuators are placed in service.

3 Answer D References NOHOlIRMSYS-01, INTERMEDIATE RANGE MONITORING SYSTEM - p. 8-9 JustiJication References during Exam None A - INCORRECT - Pulse height discriminators are used in the SRM detectors NOT the IRMs B - INCORRECT - Input resistors are used in the APRMs NOT the IRMs C - INCORRECT - Pulse preamps are used in the SRM detectors D - CORRECT - changing the range switch from 6 to 7 will change both which voltage preamp is placed in service and which attenuator is placed in service Question Source New @I Memory Level 0Comprehension Level Question History:

New 9/20 SXD OK-AF OK-

lQuestion # 36 I Hope Creek RO Exam Nov 2005 -

m# *2.5 Group# 1 I SRO Importance 2003 IRM K2.01 Knowledge of electrical power supplies to the following IRM Channels/ detectors (CFR41.7)

Question A Loss of 24VDC occurs to 1AD307 DC Distribution Panel.

& 73 Which of the following describes the effect on Nl's:

SRM IRM APRM A NOchange fails low NO change B fails low NO change NO change fails low fails low NO change Answer c References NOHOlDCELEC-00, DC ELECTRICAL DISTRIBUTION, p.38 NOHOIIRMSYS-01, Intermediate Range Monitoring System, p26 Simplified Training prints for SRM, IRM and APRMs Justification References during Exam None A INCORRECT - SRM's are powered from 24VDC and would fail downscale B INCORRECT - IRM's are powered from 24VDC and would fail downscale C - CORRECT SRM's and IRM's are powered from 24 VDC and would fail downscale, APRM's are powered from 120 VAC panels and would remain unchanged D INCORRECT - APRM's are powered from 120 VAC and would NOT fail downscale Question Source New 0Memory Level comprehension Level Quesnbn Histoy:

SXD Review - 7/21 - LOD 1.O - rewrite question to make it more difficult 8/3 Rewrote question AF - is it memory or comprehension. Changed to comprehension MB - 8/24 Made changes as requested AF-OK

!Question # 37 1 -

Hope Creek RO Exam Nov 2005 3 I # I Imvortance 3.4 2 Group# 1 3

K1.02

.SO04 Source Range Monitor 3

~~~

Question Knowledge of the physical connections and/or cause-effect relationships between Source Range Monitor and the following:

Reactor Manual Control 7

The following plant conditions exist:

- Reactor Mode Switch is in STARTUPKTANDBY

- Intermediate Range Monitors (IRM) A, C, D, E, and G are on Range 3;

- All other IRMs are on Range 2

- Source Range Monitor (SRM) A is reading 0.5 cps

- SRMs B and C are reading 8.3 x 10E4

- SRM D mode switch is in STANDBY

- A rod block signal has been generated.

Which one of the following has caused the rod block?

B A SRM Detector Wrong Position SRM Downscale Y

SRMUpscale

)I Answer References INPO Question 21837 NOH01SRMSYS-OI. SOURCE RANGE MONITORING (SRM)

SYSTEM, p32 Justijication References during Exam None A - INCORRECT - Detector Wrong Position does NOT generate a Rod Block B - INCORRECT SRM Downscale bypassed with Associated IRM range 3 C - INCORRECT SRM Upscale doesn't corne in until 1E5 cps D - CORRECT With Reactor Mode Switch NOT in RUN and SRM detector channel switch out of operate a Rod Block on SRM INOP will be generated.

Question Source New Memory Level Comprehension Level Question History:

New SI7 SXD - OK AF - Minor changes MB Incorporated changes

lQuestion# 38 I Hope Creek RO Exam Nov 2005 -

F~IGroup# 2 1 1 I

I J -6s Y 5005 G2.1.28 APRM I LPRM Y+

Knowledge of the purposes and function of major system components and controls (CFR: 41.7) 5-Question With the plant at 100% power, APRM A is indicating 99% and has the following LPRM input signals:

&sdL

- 5 LPRMs reading between 95 and 100 QAJ$Uf U f 4 b T h

- 7 LPRMs reading between 80 and 95

- 5 LPRMs reading between 50 and 80

- 4 LPRMs reading between 35 and 50 If the HIGHEST reading LPRM is BYPASSED, which one of the following describes the immediate effect on APRM A indicated power and the difference between the APRM A indicated power and the calculated (heat balance) core thermal power? n A APRM output is higher and the absolute difference is higher.

APRM output is lower and the absolute difference is lower.

x $

B 3c APRM output is lower and the absolute difference is higher.

C

/-

NO effect on either, the averaging amplifier adjusts the output for the bypassed LPRM.

J Answer c References INPO Question 24521 NOHOlAPRM00-01, AVERAGE POWER RANGE MONITORING (APRM) SYSTEM, p. 10 NOHOlLPRM00-01, LOCAL POWER RANGE MONITORING (LPRM) SYSTEM Justifcation References during Exam None Question is asking function of the averaging amplifier in the APRM circuit.

A - INCORRECT - By removing the highest, the average of all remaining LPRMs will be lower.

B - INCORRECT - If APRM output lowers, since initial power given is 100% the absolute difference must rise.

C - CORRECT - APRM output will lower and absolute difference will be higher.

D INCORRECT - While the averaging amplifier will adjust for removing LPRM input. it continues to average remaining LPRMs which will mathematically be a lower value.

Question Source Mod 0Memory Level Comprehension Level Question History:

New 917 SXD - OK AF changed number of LPRM strings MB made changes as requested

7 K1.01 IO00 RClC Knowledge of the physical connections and/or cause-effect relationships between RClC and the following Condensate storage and transfer system

'I Question Given the following

- Hope Creek is operating at 100% power.

- The RClC system is in standby with a suction from the CST.

- The Quarterly HPCl flow rate test is in progress and taking longer than expected.

Then, Suppression Pool High Level alarm has just been received.

What i s the expected response of RClC to the Suppression Pool High Level alarm?

A NO effect since RClC suction valves do NOT transfer on High Suppression Pool Level

/-

B RClC Suppression Pool Suction Valve HV-F031 receives an OPEN signal, when it gets FULL OPEN, the RClC CST Suction Valve HV-F010 will go CLOSED x

RClC Suppresson Pool Suction Valve HV-F031 receives an OPEN signal AND RClC CST Suction Valve HV-C F Oo~receives a CLOSED signal.

~

x RClC Suppression Pool Suction Valve an OPEN signal, RClC CST suction valve HV-FO10 does NOT receive any signal.

Answer A References NOHOlRCIC00-02. REACTOR CORE ISOLATION COOLING SYSTEM, p46 Brunswick Exam 2003, Q27 modified Justification References during Exam None A CORRECT - Per lesson plan neither HV-FO10 or HV-F031 receive a signal on High Suppression Pool level B INCORRECT - HV-FO31 does NOT receive an OPEN signal on High Suppression Pool level C - INCORRECT - see "3" above D - INCORRECT see "B" above Question Source Mod @ Memory Level Comprehension Level Question History:

9/7 - New Had to re-sample initial question concerned RClClRHR interconnect which has been removed.

SXD - K/A mismatch MB -wrote new question SXD -OK AF Minor changes, reword search looking for NO and NOT, replace bullet with - bullet MB Made changes as requested

2 2

4 RO I Tier# 2 Group# 1 I 3 7SRO Importance 3-2 7

3 Question With all systems operable and the station at 100% power, a seismic event causes a blackout condition (loss of offsite power and all EDGs fail to close on their respective busses) and a small break LOCA.

Conditions are as follows after the seismic event:

- Drywell pressure 1.09 psig and stable

- Reactor level lowering slowly and just crossing minu 1s Based on this information, which of the following is true?

A 105 seconds from T=O the ADS valves will automatically open.

B 300 seconds from T=O the ADS valves will automatically open.

2 405 seconds from T=O the ADS valves will automatically open.

3(

The ADS valves will NOT automatically open unless conditions change.

J Answer D References -

Hope Creek Question Q56457 - modified HC.OP-SOSN-0001 section 3.3.1 Justification References during Exam None Per HC.OP-SO.SN-0001 section 3.3.1 A - INCORRECT - NO RHR or Core Spray pumps will be running, ADS will NOT initiate B - INCORRECT NO RHR or Core Spray pumps will be running, ADS will NOT initiate C - INCORRECT - NO RHR or Core Spray pumps will be running, ADS will NOT initiate D- CORRECT - until power is restored to either the RHR or Core Spray pumps, ADS will NOT initiate.

Question Source Mod 0Memory Level @I Comprehension Level Question History:

Modified Hope Creek Question (256457on 9/20 SXD - Minor comments ME - 9/27 - Made changes as requested AF - WA mismatch

2 a RO rTier# 2 Group # I 7 1SRO Importance 3.9 3

Question Select the action(s) that will ONLY dose all the NS4 outboard isolation valves other than the MSIVs.

3 B

A "B" and "C" NS4 logic channels are deenergized.

x "B" NS4 logic manual initiation collar is armed and pushbutton is depressed.

"A" and "D"NS4 logic channels are deenergized.

C X

"D" NS4 logic manual initiation collar is armed and pushbutton is depressed.

J Answer D References Hope Creek Question - 053931 NOHOlNSSSSO-00, NUCLEAR STEAM SUPPLY SHUTOFF SYSTEM (NSSSS) - D.10.1.1.13 Training Print 045:Ol: Nudear Steam Supply Shutoff System

~~~~~~~ ~ ~ ~~ ~

Justification References during Exam None IAW B21-1090-0062 and HC.OP-SO.SM-0001 -

A - INCORRECT - this will cause a full group one [MSIV] isolation [e.g. MSlVs will dose]

B - INCORRECT - this will cause NO isolation C INCORRECT - this will cause a full NS4 isolation and the MSlVs will close D - CORRECT - ' D NSSSS logic manual initiation collar is armed and push-button is depressed.-Correct Question Source Bank Memory Level Comprehension Level Question History:

SXD review 7/21 - Minor editorial change - LOD 1.5 evaluate making question more difficult AF 8/23 - 2 correct answers C and D, added ONLY to stern.

MB - 8/24 - Made changes as requested AF-OK

3 4RO per# 2 Group# 1 I z

'7 SRO Importance 3.9 z

A4.06 3

Ability to manually operate andlor monitor in the control room Reactor water level (CFR:

41.71455 to 45.8) 3 Question The plant is operating at 100% power, with the following:

- Reactor water level is 35 inches

- An SRV inadvertently opens With NO operator action, which one of the following describes Reactor Water level response?

Reactor Water level will:

A lower and then return to 35 inches B lower and remain below 35 inches rise and then return to 35 inches

,/

rise and remain above 35 inches J

Answer C References -

Hope Creek Question ID 22077 NOHOlFWCONTC-02. FEEDWATER CONTROL SYSTEM, p.11 Justification References during Exam None A - INCORRECT - lower and then return to 35 inches (see answer C)

B - INCORRECT lower and remain below 35 inches (see answer C)

C - CORRECT - rise and then return to 35 inches. RPV Swells up on the RPV pressure reduction when the SRV initially opens. RPV level returns to 35 inches due to DFCS setpoint of 35 inches.

D - INCORRECT - rise and remain above 35 inches Question Source Bank 0Memory Level 2 Comprehension Levei Question Hktory:

SXD review 7/21 Minor Editorial changes AF - 8123 OK

3

@RO I Tier# 2 Group# I T? SRO Importance 2.8 3

3 Question The plant is operating at 70% reactor power with 2 Reactor Feed Pumps (RFPs) running in automatic with the Master Level PDS level set at 35 inches.

- A Narrow Range level is reading 36"

- B Narrow Range level is reading 35" and is the Median Controlling channel

- C Narrow Range level is reading 34" When B Narrow Range level fails to 33" Assuming NO operator action, which of the following describes the initial plant response?

Y A RClC initiatesaefreactor water level lowers to Level 2.

B Reactor water level will remain at 35 inches.

h RrLf*L Answer D References INPO Question 20357 NOH01FWCONTC-02, FEEDWATER CONTROL SYSTEM, p l 1 Justipcation References during Exam None A - INCORRECT -with Indicated level programmed level, actual level will rise B - INCORRECT -with indicated level c programmed level, actual level will rise C - INCORRECT - Level does NOT rise to Level 8 D - CORRECT - When B fails it is NO longer the Median value, B Narrow range channel will NO longer be Median Select, that will transfer to "C" because it's setpoint is 1" low actual level will rise one inch until "C" Channel is reading 35" Question Source Mod Memory Level Comprehension Level Question History:

New 917 SXD - add more justification MB added more justification SXD -OK AF - had me change correct answer to more correct answer.

MB - made changes as requested

. . .. ~ .

IQuestion # 44 Hope Creek RO Exam Nov 2005 -

1 Tier# 2 Group # 1 I Importance 3.6

-.I1000 SGTS K3.02 Knowledge of the effect that a loss or malfunction of the SGTS Off-site release rate (CFR:

will have on the following 41.7145.6)

Question Given the following conditions:

- Small Break LOCA in progress

- Drywell pressure 2 psig and rising slowly

- RPV Level had dropped to -10" and is now at +15" and stable

- Reactor Building Exhaust Radiation had been reading 1 E-4 microcuries/cc prior to being isolated

- RBVS has isolated 1-c' E-Y

- FRVS is in service

- HV-4951, PRI CNTMT VENT TO CPCS BYPASS has FAILED OPEN The CRS has directed venting of the drywell be performed per CONTROL SYSTEM OPERATION.

In order to oDen GU-HD-9372A. DRYWELL PURGE D R W E L L VENT EXH DAMPER. HV-4951, PRI CNTMT VENT TO CPCS BYPASS and HV-4952, PRI CNTMT TO CPCS INBD ISLN DMPR you isolation override push buttonsonlOC650E.

Drywell pressure is lowering and off-site release rate rising when the Reactor Building UmWetim Exhaust Hi-Hi radiation alarm is received.

Based on these conditions what will happen to:

- HD-9372A, D R W E L L PURGE DRYWELL VENT EXH DAMPER

- HV-4952, PRI CNTMT TO CPCS INBD ISLN DMPR

- Off-site release rate A HD-9372A - Closes HV-4952 Closes Off-site release rate - Decreases B HD-9372A - Remains Open HV-4952 Closes Off-site release rate - Decreases HD-9372A - Closes HV-4952 Remains Open Off-site release rate Decreases HD-9372A - Remains Open HV-4952 - Remains Open Off-site release rate Increases Answer D References HC.OP-SO.GS0001 (a),CONTAINMENT ATMOSPHERE CONTROL SYSTEM OPERATION, p. 44 NOHOIRBVENT-01, REACTOR BUILDING VENTILATION, p. 49 Justi$cation References during Exam None A INCORRECT - HV-4952 does NOT close on a Hi Hi Radiation Signal B - INCORRECT HD-9372A does NOT remain open on a Hi Hi Radiation Signal it recloses C - INCORRECT - HD-9372A closes and HV-4952 remains Open on a Hi Hi Radiation Signal if they had been previously over-ridden open.

D - CORRECT - If overrides have been depressed these valves will NOT dose on a Hi Hi Radiation Signal.

Ouestion Source New 0Memory Level Comprehension Level

Question History:

New 9/21

- Archie to verify that HV-4952 remains OPEN if it had been over-ridden open.

^YD - OK (Archie to look at)

' - K/A mismatch, NOT touching KIA, try to chop out extra words

,& added a failure of HV-4951 in and changed answer to D AF - OK

2 Hope Creek RO Exam Nov 2005 -

230 [Tier# 2 Group # 1 T? SRO Importance 3.1 2,

,2001 K4.03 AC Electrical Distribution Knowledge of AC Electrical distribution design feature(s) and Interlocks between z

or interlock(s) which provide for the following automatic bus transfer and breakers (CFR:41.7)

'2 Question With the plant in a normal electrical lineup for 100% power, the TRIP pushbutton is pressed for breaker 5240201, Normal Feed Breaker for 1OA402 on Control Room panel 1OC651E.

Which choice below describes the response of the 10A402 Bus and "B" EDG?

A The Atemate Feed Breaker, 52-40208 will close energizing Bus 10A402."B" EDG will NOT be running.

K B Bus 10A402 will be de-energized. The "B" EDG will NOT be running.

C Bus 10A402 will be de-energized. The "B" EDG will be running with its output breaker open.

x The "B" EDG will be running and its output breaker will dose energizing Bus 10A402.

J Answer References -

Hope Creek Question (253557, NOHOlEAC00-02, CLASS 1E AC POWER DISTRIBUTION, p.27 Justification References during Exam None CORRECT Bus 10A402 will be de-energized. The "B" EDG will NOT be running. The automatic transfer to the alternate feed and the start of the Diesel will NOT occur if the normal breaker is manually tripped.

INCORRECT - The Alternate Feed Breaker, 52-40208 will close energizing Bus 10A402."B" EDG Lockout will prevent the EDG start and output breaker dosure. The automatic transfer to the alternate feed will NOT occur if the normal breaker is manually tripped.

INCORRECT - Bus 10A402 will be de-energized.The "B" EDG will be tunning with its output breaker open. The automatic start of the Diesel will NOT occur if the normal breaker is manually tripped.

INCORRECT The "B" EDG will start and its output breaker will dose energizing Bus 10A402.The automatic start of the Diesel will NOT occur if the normal breaker is manually opened Question Source Bank Memory Level 0Comprehension Level Question History:

S X D review 7/21 minor editorial changes AF - 8/23 made NOT all caps in "A" MB - 8/24 Made changes as requested AF - minor changes, WA mismatch MB - made changes as requested, SXD to resolve K/A

Hope Creek RO Exam Nov 2005 -

3 RO (r# 2 Group# 1 I 7SRO Importance 2.8 3

K6.02 Knowledge of the effect that a loss or malfunction of the DC electrical power following will have on the UPS (ACIDC) (CFR:41.7/45.7)

Question 3 Hope Creek is at 100% power with the following lineup on 120V Class 1E Cyberex 20KVA Inverter 1AD481:

- CB 125V DC Power Breaker Closed

- CB-201 - 480V AC Normal Power Breaker Closed

- CB-301 - 480V AC Backup Power Breaker Open

- Auctioneered Bypass Switch is in the BYPASSfisition By @a b pes '

- Manual Bypass Switch is in the NORMAL Position An Operator inadvertently opens the CB-21 (Battery Output from Auctioneered Circuit).

What effect will that have on Class 1E Instrument Distribution Panel lAJ481?

Class 1E Panel 1AJ481 will be ...

A de-energized due to Auctioneered Bypass Switch being in the BYPASS 1 Position.

B energized from 480V AC Backup Power.

energized from 480V AC Normal Power.

C de-energized due to CB-301 - 480V AC Backup Power Breaker being Open.

3 Answer References NOHOIEACOO-02, CLASS 1E AC POWER DISTRIBUTION, p.

60-62 Justification References during Exam Figures 5, 6 and 8 of NOHOlEAC00-02.

AV2114D.vsd and AV2114F.vsd. AV-2114C A - INCORRECT - Auctioneer Bypass - Allows bypassing of one of the two Auctioneer Diodes (either diode can perform the design function) since either diode can perform the design function, bypassing diode 1 will have NO EFFECT.

8. INCORRECT -Breaker CE301 is given as OPEN and there are NO auto closures for this breaker.

C. - CORRECT - Power is normally supplied to 12OV AC Distribution Panels from the Normal AC Power source -5 Rectified to DC and then inverted back to AC. Since backup DC Power is lost, normal AC Power will still be available and the Distribution Panel will be powered as it normally is.

0. INCORRECT Panel 1AJ481 is NOT de-energized.

Question Source New 0Memory Level 13Comprehension Level Question History:

SXD review - 7/22 - OK AF - bullets, added figure 5 have SD to look at it.

SXD - Get him a copy of figure SXD - No figure due to direct lookup AF - minor changes MB - Made changes as requested

buestion # 47 Hope Creek RO Exam Nov 2005-3

-7

,3000 DC Electrical Distribution I

Al.01 Ability to predict and/or monitor changes in parameters associated with operating the DC Electrical distribution controls including Battery chargingldischarging rate (CFR:41S145.5)

YL4 Question Control Room annunciator D3-F2 "125VDC SYSTEM TROUBLE" is alarming. Upon investigation the Operator determines that Digital Point D4631 "125VDC BATTERY CHARGER 1AD413" is in alarm and Battery Charger 1AD414 Q U ~ ~ G $ ~ , ' ~ ~

is TNQf'. On panel 1OC650 the Operator reports the following:

125vDC Switchgear 10D410:

- Bus Voltage is reading 125 VDC

- Bus Current is reading 220 Amps The following is indicated on the 125VDC Battery Charger, 1AD413, control panel:

- DC Voltmeter is reading 125 VDC

- DC Ammeter is reading 200 Amps

- Timer switch is at 0

- FLOAT light is lit

- AC PWR ON light is lit

- DC Under Voltage light if off

- DC Over Voltage light is off

- Hi Voltage Shutdown light is off

- Insufficient Charging Current light is OFF WITH NO OPERATOR ACTION, which one of the following describes the expected 10D410 bus voltage trend and the reason for that trend?

The bus voltage will . . .

A lower because the bus load exceeds the charget's capacity.

\

B rise because an equalizing charge is being provided.

x

~~~

rise because a malfunction of the float charge is indicated.

~~~ ~~

D lower because AC power is NOT being supplied to the charger.

Y Answer A References INPO Question 24538 NOH01DCELEC-00, DC ELECTRICAL DISTRIBUTION, ~ 2 5 - 2 6 ,

p. 19-20 Justification References during Exam None A - CORRECT with Switchgear Load > Charger Output voltage will lower over time B INCORRECT - Equalizing Charge is NOT being provided with Timer switch at 0.

C INCORRECT - Float charge is malfunctioning because charge voltage should be > bus voltage, however this will cause voltage to lower, NOT rise over time.

D - INCORRECT - AC on and float equalize lights indicate charger has AC power Question Source Mod 0Memory Level Comprehension Level Q u es ti0 n History :

-VD review - 7122 OK -

- changed Bus current to 220 amps and DC ammeter to 200 amps, deleted Equalizing light is off, and changed

..sufficient charging current light to OFF.

MB - 8/24 - Made changes as requested AF - minor changes MB - Made changes as requested

@RO [Tier# 2 Group # 1 1

'? SRO Importance 2.9 4

Which one of the following actions are REQUIRED in accordance with HC.OP-SO.KJ-0001, EMERGENCY DIESEL GENERATORS OPERATION, to restored generator parameters within acceptable limits and the reason for this action?

A Lower reactive load using the GOVERNOR DECREASE P o prevent generator overvoltage.

B

~~

Lower reactive load using the VOLTAGE CONTROL LOWER PB to prevent generator winding Overheating.

~ ~

Raise real load using the VOLTAGE CONTROL RAISE PB to prevent generator winding overheating.

x X

Raise real load using the GOVERNOR INCREASE PB to prevent reverse power.

D Answer D References HC.OP-SO.KJ-0001, EMERGENCY DIESEL GENERATORS OPERATION, p. 33 and 34 NOH01EDG000-02. EMERGENCY DIESEL GENERATORS (EDG)

Limerick 2005 Exam Question 11 Just@cation References during Exam None A - INCORRECT with generator sync'd to grid, GOV PB changes Real Load NOT reactive load B - INCORRECT - Local control procedure has you ADJUST KiloVar loading to approx. 100 to 500 KvARs using VOLTAGE CONTROL RAlSElLOWER Control Handle, since KVAR loading is already 200 WARS. this does NOT need to be done.

C - INCORRECT - with the generator sync'd to the grid, VOLTAGE CONTROL RAISWLOWER Control Handle will change Reactive load, NOT real load.

D - CORRECT per procedure precaution 3.1.3.

Question Source Mod 0Memory Level ConipreIiension LEVCI Question History:

New 9/20 SXD - Check location MB - Verified location and terminology taken directly from Procedure SXD -OK AF - Changed to Control Room vs. Remote shutdown

- Made changes as requested

- 1 . -

Iouestion # 49 I Hope Creek RO Exam Nov 2005 - 3

,0000 Instrument Air 3

A3.02

~

Ability to monitor automatic operations of the Instrument Air including Air temperature (CFR 41.71455) 3 Question Hope Creek is at 100%.

Instrument k r status is as follows:

- OOKl07, Service Air Compressor - Disassembled for Compressor work

- 10K107, Service Air Compressor - Tripped due to Low Lube Oil Pressure - currently being investigated

- 1OK100, Emergency Instrument Air Compressor - Running

- Instrument Air Pressure - 90 psig stable A SACSnACS AUTO ISOLATION alarm is received on low pressure.

The Operators take the Mode Switch to shutdown and stablize the plant at a Reactor level of +35"(lowest level = +lo").

Assuming NO operator actions are taken and Instrument Air loads after the trip equal Instrument Air loads before the trip, what effect will this have on the Instrument Air system.

A It will have NO effect on the Instrument Air System, instrument air pressure shall be - equal to pretrip value.

I f B Discharge air temperature will increase until the Air Compressor trips on Discharge Air Temperature high, instrument air pressure will be lower than pretrip value.

Cooling water supply flow will decrease until the Air Compressor trips on Low Cooling Water Supply pressure, instrument air pressure will be lower than pretrip value.

~~~

Reactor water level dropping to 10" causes the Air Compressor to ow RPV Level, instrument air pressure will be lower than pretrip value.

Answer A References NOHOllNSAIR-01, INSTRUMENT AIR SYSTEM. P.13-14 Justification References during Exam None A CORRECT Since ElAC is running and it is cooled by RACS and trips on low RPV level of -38", a loss of TACS should have NO effect on EAlC and instrument air pressure should remain constant.

B - INCORRECT ElAC is cooled by TACS, plausible distractor, if candidate thinks cooling water is isolated to compressor, discharge air temperature would increase and may cause compressor trip.

C - INCORRECT ElAC is cooled by TACS, plausible distractor, if candidate thinks cooling water is isolated to compressor, cooling water supply flow would decrease and may cause compressor trip.

D - INCORRECT RPV level must drop to -38" to cause ElAC to trip.

Question Source New 0M e m o y Level Comprehension Level Question History :

SXD Review 7/21 - LOD - 1.0 - re-write to make more difficult 8/4 - Re-wrote question.

AF - another Instrument air question similar lo number 8. Weak WA match. Look at changing question to better match WA MB - I think K/A is ok based on the fact that it is testing student's knowledge of what is cooling the EIAC, if question stem was changed from a loss of TACS to a loss of RACS, Instrument air temp would increase and air compressor could trip on high instrument air temp.

SXD - to resolve

'D -OK

- bullets

,d3 - Made changes as requested

)Question # 50 I -

1 Hope Creek RO Exam Nov 2005 2 2 RO rTier# Group # 1 I Imaortance 2.6

-,LOO1 K5.02 A.C. Electrical Distribution Knowledge of the operational implications of the following Breaker control 2

concepts as they apply to A.C. ELECTRICAL DISTRIBUTION:

(CFR: 41.5 J 45.3) 2-switch o e 500KV breaker Emergency Trip handle accomplish?

./'

B Removes control power for the breaker.

v-Transfers control to the blockhouse.

x Activates the breaker counter.

J Answer B References Hope Creek Question - Q68164 NOHOlMNPWRO-04, MAIN POWER SYSTEM, p21 Justification References during Exam None A INCORRECT - Kirk Key can be removed with toggle in either position.

B CORRECT Interrupts breaker control drcuit to prevent electrical operation.

C - INCORRECT - No local control at the breaker except for test.

D INCORRECT - The breaker counter is mechanical.

Question Source Bank @I Memory Level [7 Comprehension Level Question History:

SXD review 7/21 OK-MB - 7/28 - Need to add references AF - 8/23 - 3rd Instrument air question. Double jeopardy with number 8 and 49. Perhaps throw out and get a different W A Maybe change out number 8.

MB - Revisit 8, 49, 50 SXD OK Questions MB - 9/27 - Re-sampled due to AF comments, New Question SXD -OK AF-OK

-2 Hope Creek -

RO Exam Nov 2005 2

RO I Tier# 2 Group# I 1SRO Importance 2.7 2 dcl000 K6.01 Component Cooling Water Knowledge of the effect that a loss or malfunction of the following will have on the Component Cooling Water Valves (CFR:41.5/45.5) 7 Question 3

With the plant operating at 100% power, 1-ED-TE-2617, RACS HX Outlet temperature fails downscale.

How will the following be affected?

I ED-TV-2617, "RACS HX 1AEIBE 217 BYPASS" valve II HV-2537A, "HX AE 217 INLET" valve 111 - Actual RACS temperature A I - goes closed I I - fails open 111 - goes down B I-goesopen I1 - fails closed 111 -goes up I - goes closed x

II - unaffected 111 goes down I - goes open D II - unaffected 111 goes up Answer D References NOHOIRACSOOC-02. Reactor Auxiliary Cooling System - p.15 Justijkation References during Exam None A - INCORRECT TV-2617 goes full open on a failure of it's input signal, causing RACS temperature to increase B - INCORRECT - HV-2537A does NOT get a temperature signal and therefore will remain as is.

C - tNCORRECT - TV-2617 fails open on a failure of it's input signal causing RACS temperature to increase D - CORRECT - N - 2 6 1 7 will Open, HV-2537A remains unaffected and RACS temperature will increase because less flow is going through the HX.

~ ~

Question Source New 0Memory Level 0 Comprehension Level Question History:

8/24 - new SXD - OK AF- changed fails open to goes open MB - made changes as requested

Hope Creek RO Exam Nov 2005

@RO [Tier# 2 Group# 1 I 1SRO Importance 3.4

,5004 A I .03 Source Range Monitor Ability to predict and/or monitor changes in parameters associated with operating the SOURCE RANGE MONITOR (SRM) SYSTEM controls including RPS status z Question Which set of conditions within the Source Range Monitoring System, would generate a Reactor Protection system SCRAM signal:

A -

Shorting links REMOVED short period trip in 2 channels B -

Shorting links REMOVED upscale trip in 1 channel C Shorting links - INSTALLED upscale t i p in 2 channels Shorting links - INSTALLED x

~~

short period trip in 1 channel Answer References NOHOISRMSYS-01, Source Range Monitoring (SRM) System -

p. 24 INPO Question 20334 Justification References during Exam None A - INCORRECT - short period trip will NOT generate a RPS w a r n signal - only get ALARM B - CORRECT - Upscale trip on 1 channel with the shorting links removed will generate a SCRAM C - INCORRECT -with shortling links installed a trip will NOT be generated on Upscale trip D - INCORRECT - never get a scram on short period trip, only get ALARM Question Source Bank Memory Level 0 Comprehenswn Level Question History:

8/24 New SXD - OK AF-OK

Hope Creek RO Exam Nov 2005 -

ERO I Tier# 2 Group # I r] SRO A002 Importance 3.3 PClSlNudear Steam Supply Shutoff 3

K6.04 Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT ISOLATION SYSTEM/ NUCLEAR STEAM SUPPLY SHUT-OFF Nuclear boiler instrumentation (CFR: 41.7 / 7 3

45.7)

Question While operating RHR in shutdown cooling, reactor water level transmitter LT-NOBOA fails downscale.

SELECT the response of the RHR shutdown cooling supply valves, HV-FOOB and HV-FOOS.

A Both RHR shutdown cooling supply valves will automatically close.

x B Only one of the RHR shutdown cooling supply valves automatically close cooling supply valve will dose if low level is sensed by LT-NOBOB.

Only one of the RHR shutdown cooling supply valves automatically second RHR shutdown cooling supply valve will close if LT-N080C fails downscale.

Neither RHR shutdown cooling supply valve will change position automatically.

J Answer References Hope Creek Question - Q53932 NOHOlRHRSYSC-03, RESIDUAL HEAT REMOVAL SYSTEM, P.30 Justification References during Exam None

- - Both RHR shutdown cooling supply valves will automatically close. -Incorrect - the tip must occur in both channels "a" and "b"/"c" and 'd" to cause any isolation

- - Neither RHR shutdown cooling supply valve will change position automatically. Correct - the trip must occur in both channel "A" and "B" to cause an isolation

- - Only one of the RHR shutdown cooling supply valves automatically close and the second RHR shutdown cooling supply valve will dose if Level 3 is sensed in the "8" NSSSS logic. -Incorrect - the trip must occur in both channels to cause any isolation. Only one would dose and only when the second signal is received.

- - Only one of the RHR shutdown cooling supply valves automatically close and the second RHR shutdown cooling supply valve will close if Level 3 is sensed in the "c" NSSSS logic. -Incorrect - the trip must occur in both channels to cause any isolation Question Source Bank Memory Level 0CompreIiensionLevel Question History:

SXD review 7/21 - OK AF - OK

IQuestinn # 54 Hope Creek RO Exam Nov 2005 -

I Tier# 2 Group# 2 I 7 a 1006 Importance 2.9 RWM z

K6.03 Knowledge of the effect that a loss or malfunction of the following will have on the RWM Rod Position indication 3

Question There is a Control Rod with an inoperable notch position reed switch. When looking at the Rod Worth Minimizer display 3

screen for that rod, how would it's position be indicated?

A RWM would display a suggested substitute position.

B RWM would display a default value of "-"

• ',  % U W D (hd&td 345u x

C RWM would display the last known good position.

x RWM would display a default value of "00" J

Answer A References INPO Question 1885 NOHOlRODMIN-01, ROD WORTH MINIMIZER p.15 Justification References during Exam None A - CORRECT per Lesson Plan - If a control rod is moved to a position with a failed reed switch, the RWM program wil1:a)Allow a single notch insert or withdraw permissive to allow the control rod to be moved to verify its actual position. b)Suggest to the operator a substitute position, which is its calculated inferred position.

8 - INCORRECT - See "A" C - INCORRECT See "A" D - INCORRECT See "A" Question Source Mod Memory Level Compre~iensionLevel Question History:

SXD review - 7/22 Had questions talk to Archie about what would be displayed. Perhaps change inop notch position to a given position (ie. 12). If you pull rod from 10 to 12 and position 12's reed switch is INOP is 12 displayed.

AF-OK

lQu,fi.n # 55 Hope Creek RO Exam Nov 2005-I Tier# 2 Group # 2 I SRO Importance 3.3

-dL002 Recirculation Flow Control A2.07 Ability to (a) predict the impacts of the following on the Loss of feedwater signal Recirculation flow control and (b) based on those predications, inputs use procedures to correct, control, or mitigate the conseqences of those abnormal operation Question Given the following conditions:

- Unit startup is in progress to 100%

- Reactor power is 47%

- Reactor water level is 35

- FW control is in 3 Element control

- 3 Primary condensate pumps are running

- 3 Secondary condensate pumps are running

- The 2nd RFP has just been place in service

- A Loop Feed flow indicates - 3.1 E6 lbslhr

- B Loop Feed flow indicates - 3.1 E6 Ibdhr

- Both Recirc pumps are running in Master Manual control with recirc pump speed and total m e flow at -65%

When an event occurs causing both Primary Condensate Pump BP-102 and Secondary Condensate Pump BP-137 to trip.

The operators are able to stabilize the plant at 47% power.

Shortly thereafter the A Loop Feed flow transmitter upscale fails causing its output to go to 8.5 E6 lbslhr What effect if any will this feedwater signal failure have on the Recirculation Flow Control circuit?

g C

Both Recirc pumps scoop tubes will lockup at their current position x

Both Rearc pumps will runback to their Speed Limit #2 (45%) speed and stablize there.

I.

Both Recirc pumps will runback to their Speed Limit #1 (30%) speed and stablize there.

Answer A References HC.OP-I0.U-0003, Startup from Cold Shutdown to Rated Power - p.4245 NOH01RECCON-02, Reactor Recirculation Flow Control System - p.30-32 Justification References during Exam Power to Flow Map A - CORRECT - given the above conditions, the operators have both a Primary Condensate pump trip and a secondary condensate pump trip. Initially these will have NO effect on the Recirc pumps because FW flow is < 75%.

However, when the FW flow transmitter fails, FW flow will raise to 87% causing both a Speed Limit 1 and Speed Limit 2 runback. since Speed Limit 1 is < Speed Limit 2, the pumps will run all the way back to their Speed Limit 1 (30%)speed setting. FW flow must be armed prior to pumps trip in order for runback.

B - INCORRECT - No Scoop tube lockup signal has been generated.

C - INCORRECT -While the pumps get a Speed Limit 2 signal, they will run all the way back to their Speed Limit 1 position since it is a lower speed.

D - JNCORRECT - see A aboves justification Question Source New 0Menlory Level Conlpreliension Level

Question History:

8/24 - New SXD - OK RJC - 2 part WA

• # B- Per NUREG 1021 ES-401 oaae 5-6 "When selectina r writina auestions for WAS that test oupled knowledge or

'lities (e.g.. the A2 K/A stateme& in Tiers 1 and 2 a n i a numb& of generic WA statements, such as 2.4.1, in Tier

-), try to test both aspects of the K/A statement. If that is NOT possible without expending an inordinate amount of resources, limit the scope of the question to that aspect of the K/A statement requiring the highest cognitive level (e.g.,

the (b) portion of the A.2 K/A statements) or substitute another randomly selected KIA."

AF - Flow must be =. 75% before pump trip in order to get runback. will be a hard question

[Question ## 56 Hope Creek RO Exam Nov 2005-I@ RO rl SRO I Tier#

Importance 2

3.8 Group # 2 I 3

- ,9000 K4.03 RHWLPCI: ToruslPool Cooling Mode Knowledge of RHRlLPCl ToruslPool Cooling Mode design Unintentional reduction in 3

Question feature(s) and or interlocks which provide for the following vessel injection flow during accident conditions 7

Given the following plant conditions:

- Drywell pressure 3.2 psig

- Drywell temperature f 70°F

- Suppression Pool pressure 1.8 psig

- Suppression Pool temperature 96°F

- Reactor water level + 25 inches

- RPV pressure 400 psig The plant has scrammed on high Drywell pressure and the actions of both Primary Containment Control and RPV Control are being carried out.

The RHR system was in a normal lineup at the beginning of the transient and all automatic actions occurred as designed.

The CRS orders Suppression Pooling Cooling started on the " ARHR Loop. Which of the following switch manipulations will have to be performed in order to start Suppression Pool Cooling on the "A" RHR Loop IAW HC.OP-S0.BC-0001, RHR System Operation?

A AUTO OP OVRD must be pressed on BC-HV-FO174 RHR LOOP A LPCl INJ MOV before valve can be dosed. Once valve is closed then BC-HV-F0244 RHR LOOP A TEST RET MOV can be opened by B BC-HV-FO17A RHR LOOP A LPCl INJ MOV must be dosed by depressing it's dosed pushbutton.

FO17A is closed then BC-HV-F024A, RHR LOOP A TEST RET MOV can be opened by depressing pushbutton.

AUTO OP OVRD must be pressed for BCHV-FO174 RHR LOOP A LPCl INJ MOV prior to depressing it's CLOSED pushbutton. Once F017A is closed then AUTO CL OVRD must be pressed for BCHV-F024A, RHR LOOP A TEST RET MOV prior to depressing it's pushbutton.

1&Le AUTO CL OVRD must be pressed on BCHV-F0T7A, RHR LOOP A LPCl INJ MOV before valve can be dosed. Once valve is dosed then AUTO OP OVRD must be pressed on BC-HV-F024A, RHR LOOP A TEST RET MOV prior to opening F024A.

Answer C References INPO Question 2069 HC.0P-S0.BC-OOOl(Q) - Rev. 40. RESIDUAL HEAT REMOVAL SYSTEM OPERATION, p. 23, Note 5.5.5 Justification References during Exam None A - INCORRECT - AUTO CL OVRD must be pressed on F024A before valve can be opened with LPCl hifiation signal present.

B. INCORRECT must depress AUTO OP OVRD for F017A prior to closing F017A with LPCl signal present C. CORRECT - per Procedure Note 5.5.5 If a LPCl Initiation signal is present, the AUTO OP OVRD must be pressed on BC-HV-F017A(B) RHR LOOP A(B,C,D) LPCl INJ MOV, before the valve can be closed. The AUTO CL OVRD must be pressed on BC-HV-F024A(B) RHR LOOP A(B) TEST RET MOV, and BC-HV-F017A(B) must be closed before BC-HV-F024A(B) can be opened.

D. INCORRECT Must Depress AUTO OP OVRD on F017A NOT AUTO CL OVRD Question Source Mod 0Memory Level Conipreiiension Level Question History:

SXD Review 7/22 - verify pushbutton labels are correct

- funky bullets - added RPV pressure = 400 psig in Stem.

j - 8/24 - Made changes as requested

7-RO [Tier# 2 Group # 2 1 Hope Creek RO Exam N O2005 - ~

z 7SRO Importance 4.2 3

3 Question 3 The plant is shutting down for a refueling outage.

T ) $bg1e.?*~-2 Current plant conditions are as follows:

- Mode Switch - STARTUP

- Reactor Power - 4% @+

A

- Reactor Pressure - 1000 psig

- Reactor Level - 35

- Condenser vacuum - 3.5 in HgA

- All MSlVs open An event occurs:

3 Minutes later plant conditions are as follows:

- Mode Switch - SHUTDOWN

- Reactor Power - All Rods inserted

- Reactor pressure - 700 psig decreasing

- Reactor Level - (-50 lowering)

- Condenser Vacuum - 23 in HgA Degrading Based on the above conditions and assuming NO operator actions, what is the status of the MSlVs and explain the reason for that status.

A MSlVs all OPEN - NO automatic dosure signal exists 7

MSlVs all CLOSED - due to 1 Automatic Closure signal - Low Reactor Pressure MSlVs all CLOSED - due to 1 Automatic Closure signal - Low Condenser Vacuum C

D MSlVs all CLOSED - due to 2 Automatic Closure signals Low Reactor Pressure and Low Condenser Vacuum Answer c References NOHOlMSTEAMC-02, MAIN STEAM SYSTEM p.24 1

Justification References during Exam A - INCORRECT - Condenser Vacuum of 21.5 will cause MSlVs to d o s k - P l a u s i b l e ~ this

~ ~ isolation f l ~ can be bypass with a keylock switch.

B - INCORRECT - Low Reactor Pressure MSlV dosure signal is bypassed when Mode Switch is NOT in RUN C - CORRECT - Low Condenser vacuum setpoint of 21.5 has been reached and limit has NOT been bypassed.

D - INCORRECT - Low Reactor Pressure MSlV closure signal is bypassed when Mode Switch is NOT in RUN Question Source New 0Memo ry Level 8 Comprehension Level Question History:

SXD Review 7/21 LOD 1.O re-write question 8/4 - Wrote new question AF - lot of bullets, changed vacuum to HgA and made NO all Caps.

MB - 8/24 - Made changes as requested AF - asked to delete some bullets MB - Made changes as requested

2 Hope Creek -

RO Exam Nov 2005 7+

a RO ~

[Tier# 2 Group# 2 I 2 SRO Imp0 rtance 2.5

.ilooo Main Turbine Gen. / Aux.

2 K1.02 Knowledge of the physical connections and/or cause effect relationships between Main Turbine Generator I Aux and the Condensate system (CFR:41.2 to 41.9 / 45.7 to 2

~~

following 45.8)

/

Question Hope Creek is operating at 75% power with all controls in automatic when a leak develops in Feedwater Heater 4A -

ME-104 causing level to rise.

The Operator observes A7-E2, Feedwater Heater Trip annunciator illuminates and FWH 4A level rises to 30" before stablizing at 30".

Assuming NO operator actions which of the following correctly describes the expected response of Main Turbine Generator MW and Reactor Power to this event:

Main Turbine Generator MW I Reactor Power II A I-Godown pi54 II - Goes down (D1LIu5 B I-Godown II - Goes up x

~ ~~

I - GOUP II - Goes down x

I GO UP II-Goesup cl Answer References HC.OP-AB.BOP-0001, Feedwater Heating, P. 1 NOHOlMNCOND-01. CONDENSATE SYSTEM, p.34 NOHOIFWHEAT-00. FEEDWATER HEATER EXTRACTION, VENT AND DRAIN SYSTEM. p. 34-37 JustiiJication References during Exam None A - INCORRECT - Expected response to this transient is that FW level will rise until the Hi Hi level is reached. At this point all turbine inputs to the FWH are isolated. Since extraction flow being removed from the turbine goes down, MW go up. In addition, FW heating of the 4A FWH goes away. therefore FW temperature goes down. Since FW temperature goes down, but steam pressure remains constant. reactor power rises.

B - INCORRECT - see A C INCORRECT - see A D - CORRECT - see A Question Source New 0Memory Level Comprehension Level Question History:

New 8/24 SXD - OK AF-OK

[Question # 59 I Hope Creek RO Exam Nov 2005-pier# Group # 2 I 5

-.,a1100 Radwaste A2.01 Ability to (a) predict the impacts of the following on the Radwaste and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those abnomal operation System rupture (CFR:41.51 4 3 3 45.31 45.13) Y Question Hope Creek is returning to service following a refueling outage.

The plant is currently in OPCON 4, when the Radwaste Operator reports, the Equipment and Floor Drain system needs to be removed from service for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> due to a rupture in the system.

How will the loss of the Equipment and Floor Drain System affect the reactor startup?

With the Equipment and Floor Drain System u P or3 Q I wt

/~ -.- \ -F.

A the plant will be unable to place RHR in shutdown codling. This could lead to reactor vessel temperature problems. x B the plant will be unable to rinse or place in service new condensate polishers. This could lead to reactor vessel chemistry problems.

C W G !

the plant will be unable to blowdown from the RWCU system. This could cause reactor vessel chemistry problems.

Y plant will be unable to get rid of excess water in condenser caused by reactor heatup. This could lead to reactor vessel level and/or condenser hotwell level problems.

Answer B References NOH01RWOVER-01, RADWASTE SYSTEM OVERVIEW p. 15 Justification References during Exam None A - INCORRECT - RHR is already in shutdown cooling, therefore the RHR line is already preheated. This would be a problem is the plant were in OPCON 2 getting ready to go to OPCON 4.

B - CORRECT from Radwaste System Overview lesson plan, p.15 - with the Equipment and Floor Drain system INOP, the plant will be unable to rinse or place in service new condensate polishers. This could lead to reactor vessel chemistry problems.

C INCORRECT - the plant can still blowdown from RWCU system to the condensate system.

D INCORRECT - excess water in condenser is dump to CST NOT the Equipment Drain system.

Question Source New Memory Level Comprehension Level Question History:

New 8/24 SXD - Bad distractors MB - 9/27 - changed distractors SXD may need to change 1st distractor, check lesson plan MB - 1013.- changed 1st &tractor.

SXD - OK AF - making decisions to change OPCON's is NOT the ROs job, need a better distractor A MB 10125 - changed distractor A

Hope Creek RO Exam Nov 2005- 2 RO I Tier# 2 Group#

• I 3 SRO Importance 3.2 4 K5.01 Knowledge of the operational implications of the following concepts as they apply to the Radiation Monitoring Hydrogen injection operation's effect on process 3

Question radiation indications 3

The plant was operating at full power with indicated H2 injection flow at 10 SCFM, when FE-601 (Flow Input to Hydrogen Flow Controller - FIC-601) fails LOW (ie. A LOW flow is INPUT into FIC-601).

Which of the following describes the expected result?

The Hydrogen Flow Control Valve Main Steam Line Radiation Levels will (11)

A I -open II - rise g I -open II - lower I - dose 3 II - lower Answer A References INPO Question 8753 NOHOlHWClOM)I, HYDROGEN WATER CHEMISTRY INJECTION SYSTEM, p. 12 M-101-0 Sht 1 & 2 Justification References during Exam 1 N/. +- I 70 A - CORRECT - FIC-601 attempts to maintain a certain H2 flow to the Secondary Condensate Pumps, when this flow 5 t4.r I input fails LOW - FIC-601 will attempt to raise H2 flow by opening the H2 Flow Control valves, opening these valves will result in Rising Main Steam Line Radiation Levels.

B - INCORRECT - H2 FCVs will open C - INCORRECT - While the FCVs will open rapidly, there is NO Low Recirc Dissolved Oxygen Level alarm.

D - INCORRECT - FCVs will open.

Question Source Mod Memory Level L? Comprehension Level Question History:

SXD review - 7/27 - Minor editorial changes AF -question asking for what happens on a local Chemistry panel. Changed answer key. To rise, lower MB - 8/24 - Made changes as requested

Hope Creek RO Exam Nov 2005 -

RO p# 2 Group# 2 1

-,.3001 Primary Containment System and Auxiliaries 4

K2.09 Knowledge of electrical power supplies to the following: (CFR: Specific Drywell cooling fans: Plant-41.7)

Question Hope Creek is at OPCON 4 with the Drywell Cooling Fans aligned as follows:

- A1 fan in AUTO - NOT Running

- A2 fan in MAN - Running in HIGH

- B1 fan in MAN - NOT Running

- 82 fan in MAN - NOT Running

- C1 fan in AUTO - NOT Running

- C2 fan in MAN - Running in HIGH

- D1 fan in MAN - Running in HIGH

- D2 fan in AUTO - NOT Running

-The Alternate Incoming Feeder Breaker to 10A401 is tagged out for Maintenance

- All of the E l , EZ,F1, F2. G1, G2, H I , H2 fans are tagged out for Maintenance l&C was working on the 4.16KV bus 10A401 and caused the normal feeder breaker to trip open and a LOP signal to be sent to the " A EDG. The "A" EDG came up to speed and restored power to the bus.

Assuming NO operator actions which of the following describes the DW Cooling Fans status following the transient:

A Running Fans - A I , B1, C1, D1, A2,C2, D2 NOT Running Fans - B2 B Running Fans - A I , D1, A2, C2, D2 NOT Running Fans - B1, C1, 82 Running Fans - A2,C2, D2 NOT Running Fans - AI, BI, C1, D1. B2

)(

Running Fans A2.82, C2, D2, A l , C1, D1 NOT Running Fans - B1 Answer A References INPO Question 20343 NOHOIDWVENT-02, Drywell Ventilation System, p15 HC.OP-SO.GT-0001, DRYWELL VENTILATION SYSTEM OPERATION, p.5 Justification

- . I References during Exam None A - CORRECT - Bus 10A401 powers MCC 10252, on a LOP all of the # I fans lose power A l , B1. etc. Any #1 fan that was running will have a Low flow condition on it. Since the #2 fans did NOT lose power, if they sense a Low flow and are in AUTO they will start. This causes fan D2 to start. In addition, all #2 fans that were running will continue to run. this leaves fans A2 and C2 running. When bus 10A401 is repowered ALL #1 fans get a start signal, this causes all # I fans to start.

B - INCORRECT - this would be correct, if you thought only fans that were running previously received a start signal.

C - INCORRECT - this would be correct. if you thought that the bus was stripped on a LOP (it is stripped on a LOCA)

D - INCORRECT- this would be correct if bus 10A401 powered all the #2 fans.

Question Source Mod Memory Level 2l omp prehension Level Question History :

Modified Question 9/20 SXD - OK AF - changed to OPCON 4. changed to I&C problem, still NOT sure about Question

[Question# 62 Hope Creek RO Exam Nov 2005-2

-.~6000 Reactor Condensate System 2 Al.07 Ability to predict and/or monitor changes in parameters System lineup Question associated with operating the REACTOR CONDENSATE SYSTEM controls including: (CFR: 41.5 145.5) 3 Given the following:

- 100% power operation with all Circulating Water pumps in service

- Loss of the 10A502 4.16KV Switchgear (Loss of power to the Circulating Water pump discharge valve hydraulic power units)

Which of the following describes the response of the Circulating Water pump discharge valves (HV-2152A,B,C, and D):

[Refer to attached figure.]

___ ~~ ~

A All failing full closed B HV-2152A and C failing full closed, and the HV-21525 and D failing as is HV-2152A, and C failing as is, and the HV-2152B and D failing full closed x

C All failing full open J

Justification References during Exam A - INCORRECT - All failing full dosed. A and C fail as is which is open. L- Figure showing Circ Water valves B - INCORRECT - HV-2152A and C failing full closed, and the HV-21526 and D failing as is. Opposite of actual.

C - CORRECT - HV-2152A, and C failing as is, and the HV-21525 and D failing full dosed. The HCU valves are aligned per HC.OP-SO.DA-0001.

D - INCORRECT - All failing full open. B and D fail dosed on the 502 bus loss.

Question Source Bank G

! I! Memo y Level Comprehension Level

'Question History:

New 918 SXD - OK AF look at rewriting question to CREF question and use Hi rad signal with damper closed. Put standby train in Manual, other one won't auto start.

M E 10127 - re-sampled - new bank question

I 3

(Question# 63 Hope Creek RO Exam Nov 2005 7 RO pier# Group# J 3

2 2 Importance 3.2

-.loo02 K1.20 Reactor Vessel lnternals Knowledge of the physical connections andlor cause effect relationships between Reactor Vessel lntemals and the following Nuclear instrumentation (CFR:41.2 to 41.91 45.7 to 45.8) 3 Question Which of the following correctly states how the LPRM strings are mounted in the Reactor Vessel: 3 A The LPRM's are mounted in dry tubes that B The LPRM's are mounted in wet tubes tha I

from below the core.

X The LPRM's are mounted in dry tubes tha enter m the above the core.

The LPRM's are mounted in wet tubes tha 3

Answer 3(c/ References NOHOIRXVESS-02, Reactor Vessel and lntemals - p.16 Justification References during Exam None A - CORRECT - the SRM's and IRMs are mounted in dry tubes that enter from below the core.

up B*TL B - INCORRECT - Control rod guide tubes are preforated with 4 holes to cool the nuclear instrumentation, the nudear instrumentation is housed in dry tubes.

C - INCORRECT -the LPRM's strings are contained in a dry tube housing and the assembly is installed and removed from above the core but the dry tubes enter below the Core.

D - INCORRECT the LPRM's are NOT mounted in wet tubes.

Question Source New @ Memory Level 0 Comprehension Level Question History:

New 8/29 SXD - OK RJC - LOD 1 SXD - Leave in AF - change answer to "A"

hestion # 641 -

Hope Creek RO Exam Nov 2005 7

2 I

Importance 3.8 2 4000 A4.02 Rod Position Information System Ability to manually operate and/or monitor in the control room: Control rod position 3 Question (CFR: 41.7 145.5 to 45.8)

Given the following conditions:

7

- control rod withdrawal signal is present

- control rod 46-35 has a Data Fault indicated on the Rod Select Module This indicates control rod 46-35 has ...

A an odd reed switch closed.

more than one reed switch is closed.

x B

/

an even reed switch is closed.

J Answer B References Hope Creek Question - Q55925 HC.OP-SOSF-0001, Rev 9, Attachment 1 Justification References during Exam None A - INCORRECT - an odd reed switch closed. This causes a "-" on the four rod display and a Rod Drift if no rod motion signal is present but not a Data Fault.

B - CORRECT - more than one reed switch is dosed. More than one position reed switch closed will cause the Data Fault light to illuminate.

C INCORRECT - drifted. A drifting rod will cause a Rod Drift alarm when an odd reed switch is passed but will not cause a Data Fault.

D- INCORRECT - an even reed switch is closed. This is normal for a stationary control rod.

Question Source Bank Memory Level 0 Comprehension Level Question History:

Re-sampled 9/14 SXD - WA mismatch MB - 9/27 inserted correct KfA based on re-sample SXD -OK AF-OK MB - 10/27 - Re-sampled due to over sampling of 223001

L Hope Creek RO Exam Nov 2005 - 2 2

,3000 K3.01 Fuel Pool Cooling and Clean-up Knowledge of the effect that a loss or malfunction of the FUEL Fuel pool temperature 7

Question POOL COOLING AND CLEAN-UP will have on following:

(CFR: 41.7 /45.6)

Given the following conditions:

3

- The plant is in Operational Condition 1, two weeks after a refueling outage

- The Fuel Pool Cooling system is operating with one pump and heat exchanger in service

- The Fuel Pool to Reactor Cavity Gates are installed

- No makeup water sources are available Which of the following is the effect on Spent Fuel Pool water temperature and level if a leak develops on the common FPCC Pump Suction?

A Fuel pool temperature will remain stable and water level will lower slightly then stabilize.

B Fuel pool temperature will rise and water level will continuously lower.

x Fuel pool temperature will rise and water level will lower slightly and stabilize.

L/

Fuel pool temperature will remain stable and water level will continuously lower.

x Answer References Ref: M-53-1 Hope Creek Bank - Q56203 Justification References during Exam None A INCORRECT - FPCC is lost - Fuel Pool temp will rise.

B - INCORRECT Level will lower to the bottom of the weir overflow pipe and then stop.

C - CORRECT: rise and water level will lower slightly and stabilize. The skimmer surge tank will drain and the FPCC pumps will trip. Fuel pool level will drain to the bottom of the weir overflow pipe then stop. Water temp will increase because FPCC is lost. Temperature rise causes water to expand, level maintained at the weir.

D INCORRECT - FPCC is lost - Fuel Pool temp will rise Question Source Bank 0Memory Level Comprehension Level Question History:

New 9/21 SXD - OK AF - minor comments MB made changes as requested MB - 10131 - re-sampled due to oversampling of 223001

. . ~...

L-bP Hope Creek RO Exam N O 2005- ~ 3

@RO I #

1SRO Importance 3.1 3 Group#

7 Ability to obtain and verify controlled procedure copy (CFR:

c) 45.10 145.13)

Question 3

Which one of the following identifies procedures considered "current revision" without DCRMS verification prior to use?

3 A Only Department Implementing Procedures (DIPs) in the Control Room B Any procedures stamped "Controlled C o p y in RED.

v-C DIPs in the Control Room and at operations field locations.

v-Nuclear Administrative Procedures and Department Administrative Procedures stamped "Controlled Copy" in RED.

Answer c References INPO Question 23092, Question taken from Salem NRC Exam 1 1/02 Justification References during Exam s*ef 10 9 C-T~OC C9-1 NC.DM-AP.U-0005, Step 5.1.2 l k \ L .4&

Question Source Bank 0 Memory Level 0 Comprehension Level Question History:

New 8/29 SXD - OK AF - NOT sure if we have objective that they need to know this.

Hope Creek RO Exam Nov 2005 7

_. 1.14 Genetic Knowledge of system status criteria which require the notification of plant personnel (CFR: 43.5 145.12) 3 Question 3

The plant is in an outage, with the following:

- RHR pump AP202 was started for Shutdown Cooling (Event 1)

- RHR pump BP202 is being removed from Shutdown Cooling (Event 2) 3 Which one of the following identifies the event(s), if any, that REQUIRE a Plant Page announcement per NC.NA-AP.ZZ-0005, STATION OPERATING PRACTICES EVENT1 EVENT2 A EVENT1-YES EVENT 2 - YES B EVENT1-YES EVENT 2 - NO EVENT 1 - NO EVENT2-YES EVENT 1 - NO EVENT2-NO Answer / References NC.NA-AP.U-0005, STATION OPERATING PRACTICES, p 19 Justification References during Exam None A - INCORRECT - per NAP-5 only have to announce Start of rotating equipment, NOT stopping of rotating equipment B - CORRECT - see "A" above C - fNCORRECT - see "A" above c.h*c J c

e SlrMQS D - INCORRECT - see "A" above Question Source New IZ Memory Level 0 Comprehension Level Question History:

New 918 SXD OK-AF - OK

Hope Creek RO Exam Nov 2005-ai0 [Tier # 3 Group # I 3 1SRO Importance 3.4 3

Ability to recognize indications for system operating parameters which are entry level condition for Technical Specifications (CFR: 43.2 / 43.3 /45.3) 3 Question During Plant startup the following conditions are observed:

TIME RPV Pressure 0700 172 psig 0715 191 psig I f 0730 205 psig 0745 233psig 0800 373psig Which one of the following is the latest time at which heatup must be secured in order to prevent exceeding the Technical Specification limit for heatup at the CURRENT heat up rate?

3 C 0830 0845 x

Answer B References Hope Creek Question - Q56983 Steam Tables Tech Spec Justification References during Exam Steam Tables Justification 172 psig = 186.7psia=376F 191psig=205.7 psia = 384°F 205 psig - 219.7 psia = 390F 233 psig = 247.7 psia = 400F 373 psig = 387.7 psia = 442F-This gives a 42F change in 15 rnins. Current heatup rate is 42F every 15 min (168 degreeslhr). 0815 - Correct- At this rate we must terminate the HIU by 0815 to keep from exceeding the allowable heatup, we would be at 484°F (this would be 100 degreedhr).

~~-

Question Source Bank [7 Memory Level a Comprehension Level Question History:

SXD review 7/21 - OK AF 2 answers. Changed 7:30 pressure to 205 psig to make only 7 correct answer M B - 8/24 - Made changes as requested

buestion # 69 Hope Creek RO Exam Nov 2005- -2 2 RO [ Tier # 3 Group # I SRO Importance 3.7

-.L.1 Generic Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

Question The plant is shutdown with 'B'RHR in shutdown cooling, OPCON 4. Inservice stroke time testing needs to be performed on the discharge valve of the ' A recirculation pump prior to commencing startup.

What precautionsllimitiations exist to allow/prevent this evolution to take place?

A As long as RPV vessel level is offscale high on all Narrow Range instruments, Shutdown cooling may be secured and the recirculation discharge valve stroked without loss of decay heat removal and vessel stratification.

B System Operating procedures for both Recirculation system and RHR system prohibit the opening of Recirculation pump discharge valves while RHR is in Shutdown Cooling, to prevent core bypass flow and vessel stratification.

~ ~~

This evolution can only be performed after the ' B Recirc pump is placed in service and establishment of forced circulation through the vessel is assured.

Prior to stroking the discharge valve on ' A Recirculation pump the suction valve must be verified dosed, and the suction valve's power supply breaker open. . /

V Answer References -

Hope Creek Question Q56375 HC.OP-I0.U-0002 section 3.2.5 JustiJication References during Exam None Justification:IAW HC.OP-IO.ZZ-0002 section 3.2.5 "Prior to stroking the discharge valve on ' ARecirculation pump, the suction valve must be verified closed. and the supply breaker opened."Correct "This evolution can only be performed after the '6' Recirc pump is placed in service and establishment of forced circulation through the vessel is assured."- Incorrect- The 'can only' distractor is wrong because the word "only " is used, along with the combination of RHR and Recirc pump combinations would still require the suction valve closed while stoking the valve "System Operating procedures for both Recirculation system and RHR system prohibit the opening of Recirculation pump discharge valves while RHR is in Shutdown Cooling, to prevent potential core bypass flow and vessel stratification." - Incorrect- The 'SOPdistractor is wrong because the IO allows this condition and applicable exception to the SO guidance "As long as RPV vessel level is pegged high on all Narrow Range instruments, Shutdown cooling may be secured and the recirculation discharge valve stroked without potential problem of loss of decay heat removal and vessel stratification." Incorrect- The 'RPV vessel level' is wrong because minimum level for natural circulation is +80" which is well above the Narrow Range detector capability to read, and does NOT assure the appropriate level.

Question Source Bank Memory Level Comprehension Level Question History:

' W Dreview 7/21 - OK

- OK

- WA mismatch

IQuestion # 70 Tier# 3 Group# I Hope Creek RO Exam Nov 2005 Y

CI

-.34 Generic Knowledge of the process for determining the internal and external effects on core reactivity (CFR: 43.6) 3 3

Question A Reactor startup from Cold Shutdown is in progress.

The ECP was calculated based upon the following:

- Reactor Coolant temperature at 140 "F

- Total Core Flow at 30 X 10E6 lbmlhr

- At time of criticality, Reactor has been shutdown for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />

- Feedwater temperature 120 "F Which of the below will result in criticality later in the rod pull sequence than the Predicted ECP?

~~

A Total Core Flow is reduced to 25 x 10E6 lbmlhr B Feedwater temperature drops to 100°F d

Criticality occurs 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after shutdown C

~~~ ~ ~~ ~

Reactor Coolant temperature drops to 125°F d

Answer c References INPO Question - 25685 GFES Justification References during Exam None A - INCORRECT - Change in Core Flow has NO effect on criticality until voiding occurs. Criticality as predicted.

B - A drop in FW temperature of 5°F is a net positive reactivity effect if FW is injecting. If not, there is NO effect.

Criticality will occur earlier or as predicted.

C - CORRECT - 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> vs. 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> results in criticality occurring at a higher Xe concentration requiring rods to be withdrawn more, therefore later than predicted.

D - INCORRECT - A reactor coolant drop in temperature is a net positive reactivity effect. Criticality earlier than predicted.

Question Source Bank 0Memory Level Comprehension Level Question History:

SXD OK -

AF - minor changes MB - made changes as requested.

IQuestion # 71 I -

Hope Creek RO Exam Nov 2005 pier# 3 Group# 1 Importance 2.6

.J . 1 Generic Knowledge of 10 CFR 20 and related facility radiation control requirements (CFR: 41.12 / 43.4. 45.9 I45.10).

Question Radiation Protection technicians have surveyed the Refuel Noor Reactor Head Laydown Area during an outage and obtained the following results:

- Highest Area Dose Rate one foot from any source in the room: 72 mr/hr

-Airborne Concentration: 0.15 DAC

- Smear Results: 750 dpmll00 cm2 gamma Based on these results the area shall be posted as a:

1. Radiation Area

11. High Radiation Area 111. Very High Radiation Area IV. Contaminated Area V. Airborne Radioactivity Area A 1,andV g I, IV,andV I and IV C

I1 and IV fl Answer A References -

Hope Creek Question Q76884 (Modified slightly)

NC.NA-AP.ZZ-0024, rev 13, p.23 Justification References during Exam None A - CORRECT - Airborne rad area > 10% or .10 DAC B - INCORRECT - NOT a Contaminated Area must be > 1000 dpm/lOcm2 C - INCORRECT - NOT a Contaminated area and it is an Airborne Area D INCORRECT - NOT a High Radiation Area - must be > 100mr/hr Question Source Bank 0 Memory Level 0 Compreliension Level Question History:

SXD review 7/21 - LOD 1.5 - evaluate writing a more difficult question Changed question out with another HC bank question that seems more difficult AF-OK RJC - Remove should, change to shall MB - Made changes as requested - 1016

Hope Creek RO Exam Nov 2005 - 3 1SRO RO Importance 3

2.9 Group# I 3 Ability to perform procedures to reduce excessive levels of 3

Question radiation and guard against personnel exposure (CFR: 43.4 /

45.10) 3 Given:

-The Shift Manager declared a Site Area Emergency thirty (30) minutes ago.

-The TSC is NOT activated.

3

- The OSC is activated.

- EOP actions outside the control room are necessary to vent the scram air header.

- The maximum expected exposure is 500 mRem

-The task is NOT going to require entry into a Harsh Environment Area

- Acts of sabotage are NOT suspected

- Area Radiation Monitors (ARMS) on the Reactor Building 102' elevation are alarming.

Which one of the following describes the requirements to perform the directed actions of venting the scram air header?

A The operator may NOT enter Reactor Building until the TSC is activated.

g Theoperato I be assigned to an OSC team of at least 2 people.

~~~~~~~

The operator is NOT required to be a qualified emergency response member as long as at least ONE member of his team is.

The operator may perform actions independently as a single person OSC team.

J \

Answer D References INPO Question 267 NC.EP-EP.ZZ-0202, P15

~

JustiJication References during Exam None A - INCORRECT there are NO requirements that prohibit Reactor Building entry until the TSC is activated.

B - INCORRECT You can use use a single person OSC team as long as 3 criteria are met - expected exposure is less than 1000 mR. Task does NOT require entry into a Harsh Environment Area, and Acts of Sabotage are NOT suspected.

C - INCORRECT ALL personnel who are selected for OSC teams are qualified emergency response members.

D - CORRECT - see "B"above. Task meets criteria for single member OSC team.

Question Source Bank Memory Level 0 Comprehension Level Question History:

New 8/29 SXD - KIA mismatch??

MB - 9/27 - I don't think so, this question is asking for what procedures need to be followed to reduce excessive personnel exposure during an emergency

- Need assistance from Archie to determine proper Hope Creek Procedure that gives this guidance -

AF - No Procedure reference. Archie to see if we have procedure guidance MB - 1111 - changed question to match EP-EP 202 criteria.

/Question#

~

73 I Hope Creek RO Exam Nov 2005 -

TTier # 3 Group #

Imvortance 4.0

-.4.49 Generic 'L Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

Question Given the following condtions:

7-

- Power ascension is in progress following a refuel outage

- Reactor power is 97%

- PSV-FO13P opens inadvertently and does NOT reclose Select the immediate operator action.

A Depress the "Reset Logic Armed" pushbutton for "B" Low-Low set logic.

x B Lock the mode switch in SHUTDOWN.

-K Reduce reactor power to 95%.

C Dispatch the operator to remove the SRV fuses J

Answer c References HC.OP-AB.RPV-0006 Hope Creek Bank - '377604

~~~~ ~ ~ ~

Justification References during Exam None A - INCORRECT - Subsequent action B - INCORRECT - Retainment override if unable to close the valve.

C - CORRECT - Immediate action.

D - INCORRECT - Subsequent action.

Question Source Bank E] Memory Level 0 Comprehension Level Question History:

New 8/29 SXD - OK AF - Job link, don't think anybody will know this, perhaps make more operationally oriented, what conditions do they need to start fire pump manually - activate fire systems from the control room, ar.qk-0002 - give flowchart.

MB - 10127 - re-sampled

IQuestion # 74 Hope Creek RO Exam Nov 2005 ERO Tier# 3 Group ## I

-4.39 Generic Knowledge of the ROs responsibilities in emergency plan 7

implementation (CFR: 45.11)

Question You are a licensed Reactor Operator assigned to the WIN Team, in the WIN Team ofice. You do NOT have assigned responsibilities in the Emergency Response Organization (ERO).

A transient occurs that results in the declaration of an ALERT Emergency and Accountability.

To which of the following locations do you report?

A The Control Room

>(

B TheTSC X

J Nuclear Administration Building (TB2)

Answer c References INPO Question 25692 Lesson Name OVERVIEW - NEPOVERVIEWC, p.15 Justification References during Exam None A - INCORRECT per Overview lesson plan All Ops personnel report to OSC for accountability.

B - INCORRECT see A above.

C - CORRECT see A above D - INCORRECT see A above Question Source Mod l&dMemoy Level 0 Comprehension Level Question History:

New 918 SXD - Add justification - low level of difficulty MB - 9/27 - Need assistance from Archie to determine which procedure provides guidance to Licensed Operators NOT on shifl as to where they report during an emergency. My answer was based on GET info. -

SXD - Very GET, may need to resample, Archie to provide procedure guidance if possible RJC - Make sure it is a requirement MB - Archie to check AF - Got procedure reference

Hope Creek RO Exam Nov 2005 -

aR0 1Tier # 3 Group # 1

'1SRO Importance 3.3 Knowledge of annunciators alarms and indications / and use of the response instructions. (CFR: 41.10 / 45.3)

Overhead Annunciator Window Box h4- 10C320 has 2 pieces of red tape diagonally placed across the annunciator window in the shape of an "X.

Which one of the following describes the significance with SH.OP-AP.ZZ-0030, Operator Burden Program?

A The entire annunciator window is inoperable.

B One the annunciator window J

Indicates that a design change request notification has been submitted against the annunciator window.

d Answer A References INPO Question 818 SH.OP-AP.ZZ-0030, Operator Burden Program, p.4 Justiflcation References during Exam %me ftCq741L 2%-3 ~ + W H W A - CORRECT -per SHOP-30 p.4 - if the entire annunciator window is inoperable then 2 pieces of red tape diagonally placed across the annunciator window in the shape of an "X B - INCORRECT - per SHOP-30 p.4 - if one or more inputs of a multiple input annunciator are inop then red tape should be placed diagonally across the annunciator window.

C - INCORRECT - While a notification should have been written against the annunciator window, the red "X indicates that the annunciator is INOP.

D - INCORRECT per SHOP-30 p.4 - if a design change request notification has been written against an instrument a piece of red tape should be placed across the instrument to alert the operator that the instrument is NOT reliable. NOT 2 pieces of red tape in an X" Question Source Mod Memory Level 0 Comprehension Level Question History:

SXD review 7/27 - NOT SRO level - rewrite Rewrote question - 7/29 - somewhat based on INPO Question 22362 JD - 8/2 - W h y are C,D plausible ME - 8/3 - Changed AB.CONT-0005, Irradiated Fuel damage to EO.U-O103/4 since Radiation levels in the Reactor Building are rising and operator may be concerned about reactor building release.

AF - OPCON 5 vs. Mode 5, changed Fuel Pool to LPCl and changed RHR to Fire water.

MB 8/30 Realized Question should be RO -went back to original question.

SXD - OK

-OK

Hope Creek SRO Exam Nov 2005 - Y RO 1 4 SRO rTier#

Importance 1 Group #

3.8 1

I AG2.1.32 Ability to explain and apply system limits and precautions (CFR 41.10/ 43.2/ 45.12)

Question Hope Creek was operating at 30% power when a Station Blackout (loss of all onsite and offsite power) occurred causing a Reactor Scram.

Current plant conditions are as follows:

- Drywell temperature - 300°F decreasing slowly

- RPV pressure - 273 psig decreasing slowly

- Reactor Power - all rods fully inserted

- Reactor level - (-100" decreasing)

- RCIC - tagged out and disassembled

- HPCl - tripped on overspeed and will NOT restart

- "A" EDG -tagged out for maintenance

- "B" EDG - running unloaded - output breaker failed open on anti-pump circuitry

- "C" EDG -tripped on Bus differential overcurrent

- "D" EDG - failure to start - low air pressure -20 psig Based on these conditions, the Control Room Supervisor shall:

A direct the NE0 to reset the Bus differential overcurrent on the "C" EDG and restart the "C"EDG.

'>(

B direct the RO to depress the TRIP pushbutton on the "B" EDG output breaker and verify output breaker closes.

I f enter procedure HCOP-EO.U-0202. Emergency Depressurization based on high Drywell temperature.

Y enter procedure HC.OP-EO.=-0202, Emergency Depressurization before Reactor Water Level decreases to -

D 129".

Answer B References HC.OP-ABZ-0135, Station Blackout// Loss of Offsite Power//

Diesel Generator Malfunction p. 2 Justification References during Exam EOP Flowchart - 202 and 101, 102 with NO entry conditions A - INCORRECT - bus differential current should NOT be reset without electrical maintenance determining and correcting the cause.

B - CORRECT - per HC.0P.AB.Z-0135, Station Blackout p. 18 step 5.16 -The Anti-pump circuitry on the D/G output breaker could cause the output breaker to fail open. To load the D/G under this condition the operator must depress the TRIP push-button (even though the breaker is already tripped) to reset the logic. When the TRIP push-button is released, then the breaker will close and the DIG will load.

C - INCORRECT - Emergency Depressurization procedure should NOT be entered until DW temperature exceeds 340°F and current drywell temperature is decreasing.

D - INCORRECT - Emergency Depressurization procedure should NOT be entered until is less than -129" but before level decreases to -185" Question Source New 0Memory Level 0 Comprehension Level Question History:

SXD review 7/27 - NOT SRO level - re-write 812 - re-wrote question AF - bullets. give the Operator the EOP flowchart. Look at possible double jeopardy.

"B 8124 - Made changes as requested

Hope Creek SRO Exam Nov 2005 -

DRO [Tier # Group # 1 7SRO Importance 3.3

,5004 Partial or Total Loss of DC Pwr 1 6 AA2.04 Ability to determine and interpret the following as they apply to System Lineups Partial or Total loss of DC power:(CFR: 41.10 J 43.5 /45.13)

Question Given the following condtions:

The Reactor is in Operational Condition 4.

WQ The NEO's are performing a system lineup on 24 VDC.

Plant startup operations are in progress.

The negative battery charger for the "A" f24 VDC System is found to be out of service.

The positive battery charger for the "6" G'4 VDC System is placed on an equalizing charge.

All other equipment was found to be aligned for normal operation.

Which of the following will occur if these conditions remain for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />?

An RPS trip will occur due to:

A IRMs upscale1lNOP (112 scram)

B SRMs upscale/lNOP (Full Scram)

LPRMs upscale/lNOP (Full Scram)

C APRMs upscale1lNOP (112 scram)

J Answer A References Hope Creek Bank - (261702 HC.OP-AB.ZZ-0151. Sections 2.1, 4.5 & 5.1 0301-000.00H-000069-13, Sections V1I.A. 1, Vll.B.1-3. & Figures 32 8 33 HCGS Incident Report 86-067 Justification References during Exam The negative charger only charges the negative battery while the positive charger only charges the posrtive battery.

Even with the positive charger operating in the Equalizer mode, the negative battery will be discharged resulting in the loss of the DC bus.

A - CORRECT - IRMs upscale (112 scram). The loss of the -24VDC from the A f 24VDC System will cause IRM indications to rise (upscale). This will insert a 112 scram from RPS Channel A.

B - INCORRECT - SRMs upscale (Full Scram). SRM indications to lower (downscale)

C - INCORRECT LPRMs upscale (Full Scram). LPRMs and APRMs are unaffected by the loss of -24VDC.

D - INCORRECT - W R M s upscale (1/2 scram). LPRMs and APRMs are unaffected by the loss of -24VDC Question Source Bank Memory Level 0Comprehension Level Question History:

New 912 1 SXD - OK AF -OK MB 1111 - changed out question with Bank question 061702

RO I Tier # 1 Group# 2 7SRO Importance 3.8

-,5006 AG2.1.32 SCRAM / 1 Ability to explain and apply system limits and precautions (CFR 3

41.10/ 43.2145.12)

Question Hope Creek has just scrammed from 100% power The Control Room Supervisor has entered HC.OP-AB.22-0000, Reactor Scram and has the following plant conditions:

- RPV Level - (+33" stable)

- RPV Pressure - 1000 psig stable

- Mode Switch - Locked in Shutdown position

- All Control Rods fully inserted You have reached step S-8:

"IF Conditions permit THEN RESET the Scram AND INSERT a Half Scram (if required) you to INSERT a Half Scram?

1. 2 APRM deteEtars INOP II. 2 IRM deteetars INOP Ill. 1 Reactor Vessel Steam Dome Pressure High Transmitter INOP B II Only x

c I, I I Only x

I, II,and 111 D

Answer A References NOHOlAB0000-01, Reactor Scram AB-0000 - p. 17-18 Tech Spec 3.3.1 Justification References during Exam 3.3.1 A - CORRECT Per TS 3.3.1 b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least 1 trip system in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1 For the APRM's in OPCON 3 - Minimum OPERABLE Channels per Trip System is 2, therefore if 2 APRM were INOP you would only have 1 in that Trip System OPERABLE and would need to insert a Half-scram B - INCORRECT - Per TS 3.3.1-1 in OPCON 3 you are only required to have 2 IRM's OPERABLE per trip system, since you have 3 available having 1 INOP still leaves 2 that are OPERABLE and so you would NOT have to insert a Half-Scram C - INCORRECT - see B above D - INCORRECT see B above, also Reactor Steam Dome Pressure High transmitter is only required in OPCON 1 or 2, since you are in OPCON 3 this would be N/A Question Source New Memory Level @I Comprehension Level Question History:

New 8/30

'D - borderline SRO. write something to address objective Obj 6 in AB0000-01

- 10/3 wrote new question to address OBJ 6 (SRO ONLY) objective of Lesson Plan in ABOOOO-01

=AD - OK AF - WA mismatch. want 3.3.1 MB - talk to SXD, I think they should know what puts them in a Tech Spec without having to have Tech Specs.

RO I Tier# 1 Group # 1 1 7SRO Importance 3.7 2

2-Question Hope Creek is operating at 100% power when an Instrument Air line in the Turbine Building ruptures. The air compressors are unable to keep up with the loss of air and Instrument Air pressure is lowering.

What will the long term Reactor Pressure Vessel level control and pressure control strategy be for the loss of Instrument Air in accordance with HC.OP-AB.ZZ-0000, Reactor SCRAM?

A Bypass valves for pressure control, Maximize CRD for level control.

Y-B SRVs for pressure control, Maximize CRD for level control.

SRVs for pressure control, HPCIIRCIC for level control.

x

./

Bypass Valves for pressure control, HPCIRCIC for level control.

3 Answer C References INPO Question 25895 NOHOlMSTEAMC-02, MAIN STEAM SYSTEM, p. 46 Jusapcation References during Exam None A - INCORRECT - Condenser is NOT available and NO condensate line up is possible due to level control valves fail dosed on a loss of air.

B - INCORRECT - CRD flow control valves fail closed on a loss of air C - CORRECT - Outboard MSlVs will go dosed on a loss of air, therefore NO steam for feedpumps or use of the main condenser for decay heat. Condensate will be unavailable due to NO feedpath on a loss of air.

D - INCORRECT - Condenser is NOT available for pressure control Question Source Mod Memory Level Comprehension Level Question History:

New 918 SXD - Look for procedure tie in.

MB - 9/27 -Added per AB-SCRAM SXD -OK AF -Another Inst. Air question, minor questions MB - Made changes as requested.

c 3

0RO I Tier# 1 Group# 1 I q SRO Importan ce 3.3 d'qL *4.,

-d5028 EG2.4.50 High Drywell Temperature / 5 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR 45.3)

U Y

Question Given the following conditions:

- A small steam leak has occured in the drywell causing a reactor scram

- Two control rods are at position 06

- RPV level +30 inches

- RPV pressure 920 psig

- Suppression pool level 75 inches

- Suppression pool temperature 80 "F

- Drywell pressure 3 psig

- Average drywell temperature 330 "F and rising at 1°F per minute

- Suppression chamber pressure 3 psig Which of the following describes the next operator action(s) in accordance with the Emergency Operating Procedures?

A Shutdown the Reactor Recirculation Pumps and Drywell Cooling Fans and initiate one loop of drywell spray.

B Verify all injection into the RPV except SLC, CRD and RClC is and then emergency depressurize the reactor.

Rapidly depressurize the reactor using the main turbine bypass valves.

Initiate suppression chamber sprays and commence a normal reactor cooldown. (Less than 90 F per hour) 3 Answer References Hope Creek Question Q56045 HC.OP-EO.ZZ-0102 Bases, step DWn-5 Just9cation References during Exam EOP 101a. 102 A - INCORRECT Shutdown the Reactor Recirculation Pumps and Drywell Cooling Fans and initiate one loop of drywell spray.-incorrect- CANNOT DW Spray since outside of DWT-P curve.

B - CORRECT - Verify all injection into the RPV except SLC, CRD and RClC is terminated and prevented and then emergency depressurize the reactor.-correct- EOP-0202 step ED-3 C - INCORRECT - Rapidly depressurize the reactor using the main turbine bypass valves.-incorrect- EOP-1O1A prevents use of BPVs in this situation D - INCORRECT Initiate suppression chamber sprays and commence a normal reactor cooldown. (Less than 90 F per hour)-incorrect- must stabilize pressure until SID under all conditions without Boron Question Source Bank 0Memory Level Comprehension Level Question History:

SXD review - 7/29 - OK AF give EOP flowcharts during exam MB - 8/24 - Made changes as requested

[Question # ai I Hope Creek SRO Exam Nov 2005 - 2 p r #

ImDortance 1 Group#

4.2 1 1 3

-~5030 EA2.01 Low Suppression Pool Wtr Lvl / 5 Ability to determine and interpret the following as they apply to Suppression Pool level 3

Low Suppression Pool Water level (CFR:41.10/ 43.5/ 45.13)

Question With the plant operating at 100% power the RO reports to you that Suppression Pool Level has drifted out of the allowable Technical Specification value.

Investigation reveals that a small leak has developed on the Instrument line for Suppression Pool Level transmitter LT-4805-1 just downstream of valve V9982.

Using the attached figure, how will the reading on LT-4805-1 compare to ACTUAL Suppression Pool level and what is the Technical Specification bases for maintaining ACTUAL level at the proper level.

A - LT-4805-1 will read HIGHER than ACTUAL level

- Bases for maintaining level is to ensure adequate NPSH exists for ALL pumps (HPCI. RCIC, LPCl and CSS) to inject following a Design Basis LOCA B - LT-4805-1 will read LOWER than ACTUAL level.

- Bases for maintaining level is to ensure adequate NPSH exists for ALL pumps (HPCI. RCIC, LPCl and CSS) to inject following a Design Basis LOCA

- LT-4805-1 will read HIGHER than ACTUAL LEVEL

- Bases for maintaining level is to ensure primary containment pressure will NOT exceed design pressure during a primary system blowdown.

- LT-48051 will read LOWER than ACTUAL LEVEL k 1 D - Bases for maintaining level is to ensure primary containment pressure will NOT exceed design pressure during a primary system blowdown.

Answer D References NOH01PRICON-02, Primary Containment Structure - p.21 Tech Spec - bases 3.5.3 and 3.6.2 Justif cation References during Exam Figure showing LT-4805-1 A - INCORRECT - leak on high pressure side of tap will cause indicated level to read LOWER than actual.

B - INCORRECT - Bases for Suppression Pool level is either: Ensure a sufficient supply of water is available to the HPCI. CSS and LPCl systems - NOT the RCIC pump. OR to ensure primary containment pressure will NOT exceed design pressure during a primary system blowdown.

C - INCORRECT - leak on high pressure side of tap will cause indicated level to read LOWER than actual D - CORRECT Question Source New 0Memory Level Comprehension Level Question History:

New 8/30 SXD - minor comments MB - 9/24 Made changes as requested AF-OK

Hope Creek SRO Exam Nov 2005 -

URO E r # Group# I SRO

,5018 Importance 3.6 Partial or Complete loss of CCW 3 G2.4.30 Question Knowledge of which events related to system operationdstatus should be reported to outside agencies 3

Hope Creek is operating at 100% when a partial loss of Reactor Auxiliary Cooling System (RACS) flow to the Reactor Water Cleanup (RWCU) System Non-regenerative Heat Exchanger resulted in an automatic isolation of RWCU Inlet Outboard Isolation Valve HV-FO04, due to RWCU Non-Regenerative Heat Exchanger discharge high temperature isolation signal. NO other isolation valves were actuated. Plant remains stable at 100% power.

Which of the following identifies a proper assessment of 10CFR50.72 , Notifications?

The event is:

A reportable per 10CFR50.72 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

g reportable per 10CFR50.72 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

reportable per 10CFR50.72 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

x NOT reportable per 10CFR50.72.

J Answer D References 10CFR50.72 Hope Creek Event Classification Guide - Section 11 Justification References during Exam ECG Section 11 Wants Tech Bases for 11.3.3 A - INCORRECT - Did NOT cause a deviation from Tech Specs.

B - INCORRECT -An RPS actuation was NOT initiated as a result of this signal.

C - INCORRECT - Only 1 system affected, NOT required to be reported.

D - CORRECT - This event is NOT reportable per 10CFR50.72 as item (b)(3)(iv)(B)(2) requires containment isolation signals affecting more than 1 system. This signal only affects 1 system.

Question Source New 0Memory Level @ Compreltension Level Question History:

New 9/21 SXD - OK AF - minor comments, wants tech bases for ECG 11.3.3 MB Made minor editorial changes, SXD to resolve giving the tech bases

RO I Tier # 1 Group # 2 1 SRO Importance 3.7

,5009 Low Reactor Water Level I 2 AA2.02 Ability to determine and interpret the following as they apply to Steam flow/ feed flow Low Reactor Water Level (CFR: 41.1 01 43.5 145.13) mismatch

/ \

Question Hope Creek is operating at 75% power problem on the Bleeder trip valve with the following conditions:

- Feedwater control is in 3 element control

- A Steam Flow indicates - 2.5 E6 Ibdhr

- B Steam Flow indicates - 2.5 E6 lbslhr

- C Steam Flow indicates - 2.5 E6 Ibdhr

- D Steam Flow indicates - 2.5 E6 Ibdhr

- FW flow (NOOIA) indicates - 5.0 E6 Ibdhr

- FW flow (NOOIB) indicates - 5.0 E6 Ibs/hr

- Reactor Water level - Normal at 35" stable

- Reactor Pressure - 1000 psig stable

- Generator MW - 750 MW An event occurs.

1 Minute after event initiation the following conditions are observed:

- A Steam flow indicates - 1.7 E6 IbsIhr

- B Steam flow indicates - 2.5 E6 Ibs/hr

- C Steam flow indicates - 2.5 E6 Ibdhr

- D Steam flow indicates - 2.5 E6 Ibdhr

- FW flow (NOOIA) indicates - 5.0 E6 IbsIhr

- FW flow (N001B) indicates - 5.0 E6 Ibs/hr

- Reactor Water level is 38"and lowering slowly

- Reactor Pressure - 990 psig stable

%erator MW - 670 MW Based on the above conditions, what event has happened and what procedure shall you direct the operaton to respond to the event?

A "A" Steam line's input to Total Steam flow has partially failed causing Steam RowIFeed flow mismatch, go to procedure HC.OP-AR.ZZ-0007 window F-1. "DFCS ALARMiTRBL" d

B "A"Main Turbine Stop Valve has failed dosed, go to procedure HC.OP-AB.BOP-0002, MAIN TURBINE An SRV has opened on the "A" steam line, go to procedure HC.OP-AB.RPV-0006, SAFETY RELIEF VALVE C f The 6A Feedwater heater bleeder trip valve has failed, go to procedure HC.OP-AB.ZZ-0001, TRANSIENT PLANT CONDITIONS Answer C References HC.OP-AB.RPV-0006, Safety Relief Valve p.1 NOHOlMSTEAMC-02. MAIN STEAM SYSTEM.

~ ~~~ ~

Justification References during Exam None A - INCORRECT - While "A" steam linels input to Total Steam flow could cause the difference in indicated Steam Flow, it would NOT cause Generator MW to decrease.

B - INCORRECT - While "A" Main stop valve failing closed would cause a decrease in MW, it would NOT cause Reactor pressure to decrease, it would increase.

C - CORRECT - A safety on "A" steam line would cause, "An's steam tine flow to decrease, MW to decrease and Reactor Pressure to decrease.

D - INCORRECT - 6A's bleeder trip valve going closed would cause MW to go up NOT down.

estion Source New c3 Memory Level @I Compeltension Level

Question History:

SXD review 7/27 - LOD 1 - rewrite 811 rewrote question - MB AF - added bullets, changed values to HC numbers. on Cchanged safety to SRV.

. 6- 8/24 - Made changes as requested

- Lots of Info, may want to take info out, minor comments

Question # 84 Hope Creek SRO Exam Nov 2005 - Li

[ Tier# 1 Group# 2 I SRO Importance 4 A 5 0 10 High Drywell Pressure I 5 AG2.4.6 Knowledge of symptom based EOP mitigation strategies (CFR:

41.10143.5/45.13)

Question Following a station blackout event the STA reports the following parameters to the CRS:

- RPV Level minus 35 inches

- Drywell temperature3SLF

- Drywell pressure o w s i g A'3o&'I"") s O

\ F?

/ ;u S- h 42 Which of the following ACTIONS SHALL be taken and what is the REASON for the action?

A ACTION - Spray the Drywell REASON - Convection cooling Drywell is needed to prevent over pressure condition in the drywell.

~~ ~~

g ACTION - Spray the Drywell REASON - Evaporative cooling of Drywell is needed to prevent over pressure condition in the drywell.

ACTION - Emergency Depressurize REASON - Evaporative cooling would result in rapid Dryw-ell pressure reduction to less than atmospheric and possible implosjon of the Dryw%ll.

ACTION - Emergency Depressurize 46J REASON Convection cooling would result Drywell pressure reduction to less than atmospheric and possible implosion of the Drywell.

Answer c References HC.OP-E0.ZZ-0102, flowchart and bases p. 9 NOHOlE0102P-00. HC.OP-EO.=-0102 PRIMARY CONTAINMENT CONTROL DRYWELL (TEMPERATURE I PRESSURE AND HYDROGEN )

INPO Question 21160 Jusrification References during Exam Flowchart E-0102 (minus entry conditions)

A INCORRECT - Per the curve D W - P the plant is in the UNSAFE region, therefore you do NOT want to initiate Drywell spray. If SRO miss calculates or mis-readings the Curve D W - P they may think they are in the Safe region.

B INCORRECT see "A" above.

Depressurize, since DW temperature is approaching 340°F and you cannot use Drywell$ray.

f C - CORRECT - Since the operator cannot Spray the drywell, the only other option is to Blowdown o Emergency D INCORRECT - while the operator does wish to Blowdown, the Reason per the bases is that Evaporative cooling could result in Drywell pressure reducing to 2 psig and causing a Drywell implosion.

Question Source Bank c]Memory Level CompreJzension Level Question History :

New 8/30 SXD - minor comments MB - 8/24 - Made changes as requested AF - minor comments MB Incorporated changes as requested.

3 buestion # 05 Hope Creek SRO Exam Nov 2005 - 5

-,ti012 High Drywell Temperature AA2.01 Ability to determine and/or interpret the following as they apply Drywell temperature to HIGH DRYWELL TEMPERATURE : (CFR: 41 . I O 143.5 /

45.13)

Question Which one of the following identifies the bases for the Drywell Average Air Temperature Limiting Condition for Operation (LCO)?

In the event of a DBA. initial drywell average air temperature is assumed to be less than or equal to:

A 135°F so that the resultant peak accident temperature is maintained below 300°F during main steam line break conditions and is consistent with the safety analysis.

B 135"F,so that the containment peak air temperature does NOT exceed the design temperature of 340°F during LOCA conditions and is consistent with the safety analysis.

150°F so that the resultant peak accident temperature is maintained below 300°F during main steam line break conditions and is consistent with the safety analysis.

x 150°F so that the containment peak air temperature does NOT exceed the design erature of 340°F during LOCA conditions and is consistent with the safety analysis.

Answer )($ References Tech Specs 3.6.1.7 and bases Justification References during Exam None A - INCORRECT - maintain peak temperature c 340°F NOT 300°F and accident is LOCA vs. Main Steam line break.

6 - C O R R E T per bases of 3.6.1.7 C - INCORRECT - temperature is 135°F vs. 150°F and see "A" above D - INCORRECT - temperature is 135°F vs. 150°F Question Source New Memory Level Comprehension Level Question History:

New 9/21 SXD OK -

AF-OK

[Question # 86

~

Hope Creek SRO Exam Nov 2005 4 SRO I Tier#

Importance 2 Group#

3.3 I Y

-~6000 G2.1.14 HPCl Knowledge of system status criteria which require the notification of plant personnel. (CFR: 43.5 / 45.12) 3 Question Hope Creek is in a Startup, you are the Operations Field Supervisor performing a Secondary Containment inspection when you discover a TMOD tag on some temporary instrumentation connected to the steam piping on the HPCl turbine. The TMOD tag is inside a contaminated area and is damaged and unreadable.

Who are you REQUIRED to notify in accordance with NC.DE-AP.ZZ-O030(Q), Control of Temporary Modifications?

A A The Control Room Supervisor B The Shifi Manager Answer c References NC.DE-AP.ZZ-O030(Q) - CONTROL OF TEMPORARY MODIFICATIONS, p. 16 Justification References during Exam None A - INCORRECT - Procedure requires you to notify the Responsible Engineer B - INCORRECT - See "A" C - CORRECT - See " A D - INCORRECT - See "A" Question Source New Memory Level 0 Comprehension Level Question History:

New 8/31 SXD - WA mismatch - NOT notifying outside agencies, notify Plant personnel MB - 9/27 -Wrote new Question SXD - OK AF-OK

Hope Creek SRO Exam Nov 2005 -

OR0 rTier# 2 Group # I 'I 9SRO Importance 3.2 3

A2.02 Ability to (a) predict the impacts of the following on the LPCS Valve closures and (b) based on those predictions, use procedures to correct control or mitigate the consequences of those abnormal operation (CFR: 41.V 4 3 3 45.3/ 45.13)

Question A transient has occurred on Hope Creek.

/

- Drywell pressure peaked at 4 psig

- Drywell pressure is now 1 psig and steady

- RPV level is 1 8 and dropping

- RPV pressure is 230 psig and lowering

- "A" Core Spray pump is running

- "C"Core Spray pump has tripped

- HV-F005A CORE SPRAY INBOARD ISOLATION MOV was CLOSED to terminate injection from the "A" Core Spray System

- HV-F004A CSS Loop Upstream Injection valve is OPEN In accordance with the guidance provided in HC.OP-SO.BE-0001. Core Spray System Operation, to raise RPV water level using the "A" Core Spray pump, the CRS shall direct the RO to:

A Open HV-FOOSA and once HV-FOOSA is full OPEN, throttle HV-FOO5A from the control room to control RPV level B -riceP , M V - F 0 0 5 A and throttle HV-F005A locally to control RPV level U d - f B b7(Lcwc -

C C W V - F 0 0 4 A , a~c&l&F004 room to control RPV level -

is full CLOSED, OPEN HV-FOOSA and throttle HV-FOOSA from the control L -

c _

N-w Qb"- f%c(A OPEN HV-FOOSA and once HV-FOOSA is full OPEN, throttle HV-F004A locally to control RPV level D

Answer B References NOHOlCSSYSO-01, Core Spray System p. 18 HC.OP-SO.BE-0001, Core Spray System Operation. p. 3 and 5 Justification References during Exam None A - INCORRECT - per Procedure SO.BE-0001. Interlock 3.3.4, HV-F005A can NOT be opened unless HV-F004A is f#'wgb*

5'4 Fully closed. Therefore must dose HV-F004A prior to opening HV-FOOSA B - CORRECT - must close HV-F004A, then OPEN HV-FOOSA and per limitation 3.1.4 with only 1 CS pump running HV-F005A should be throttled to limit pump flow and Throttling of HV-F005A can only be performed locally.

C - INCORRECT - per Procedure SO.BE-0001, thoffling of HV-FOOSA can only be performed locally.

D - INCORRECT - see "A" above Question Source New 0Memory Level Comprehension Level Question History:

New 8MO SXD - check to see if answer is required by procedure.

MB -Wrote new question SXD - OK RJC - are actions being asked direct the RO to, put in accordance with "add procedure" MB -Added procedure guidance AF - WA mismatch

Hope Creek SRO Exam Nov 2005 7

aRO I Tier # 2 Group # 1 I

'qSRO Importance 3.7 3

I 5005 APRM I LPRM A2.02 Ability to (a) predict the impacts of the following on the APRMI Upscale or downscale trips.

LPRM and (b) based on those predictions, use procedures to correct control or mitigate the consequences of those abnormal operation (CFR: 41.51 43.51 45.31 45.1 3)

Question A reactor startup is in progress at 9% power, OPCON 2, when the Startup Level Control Valve LV-1785 fails FULL OPEN with the following results:

- FULL SCRAM

- RO reports 4 rods at positions between 04 and 08

- APRMs are DOWNSCALE

- RPV level fell to at a low of 15" and is now slowing rising

- BOP is able to control the Startup Level Control Valve LV-1785 in MANUAL

- RPV Pressure is 900 psig and slowly trending down.

Which of the following is the CAUSE of the SCRAM?

What DIRECTION is required?

A CAUSE - RPV High Water Level 3(

DIRECTION - RESET the SCRAM and insert control rods per HC.OP-AB.ZZ-0000 x B CAUSE - APRM Upscale rt r/

DIRECTION - RESET the SCRAM and insert control rods per HC.OP-AB.ZZ-0000 CAUSE - RPV High Water Level )c DIRECTION - ENTER HC.OP-EO.U-OlOlA, determine a success path and insert control rods L/

CAUSE - APRM Upscale DIRECTION - ENTER HC.OP-EO.ZZ-OlOlA, determine a success path and insert control rods J Answer D References INPO Question 25648 HC.OP-AB.=-0000, step S-8 HC.OP-E0.Z-0101 RPV control Entry conditions NOH01APRMOO-01, Average Power Range Monitoring System -

p. 40 Justification References during Exam None A - INCORRECT - RPV High Water Level does NOT give a SCRAM, also based on RPV level of 15" and rising, RPV High Water level setpoint was NEVER reached.

B - INCORRECT APRM upscale trip was received due to insertion of cold water causing APRM's to rise to 12-15%

and generate a SCRAM, since NO E-101 parameters have been reached the proper course is to enter AB-000.

C - INCORRECT - Should NOT enter EO-1OlA since NO EO-101 parameters have been met.

D - CORRECT Should enter EO-1O1A since it is not certain that the reactor is shutdown from all conditions.

Question Source Bank 0Memory Level @lComprehension Level Question History:

New 8131 SXD - circled OPCON 2, make sure 9% is still OPCON 2 MB - 9/27 - Per 10-0003, Operator is to withdraw control rods to 7-10% PRIOR To placing Mode Switch to RUN, therefore Plant could be at 9% and NOT be in RUN.

SXD - OK AF - correct answer D MB - SXD to look at.

IQuestion # a9 I Hope Creek SRO Exam Nov 2005 - -3

[Tier# 2 Group#

4 1

Y

-29002 G2.1.33 Reactor Water Level Control Ability to recognize indications for system operating parameters which are entry level conditions for technical 3

specifications. (CFR: 43.2 t 43.3 145.3)

Question With the plant at 100% Power, I&C reports to you that LT-NO8OA, RPV Low Level to NS4 ISLN and RPS Trip Logic has failed it's Quarterly surveillance.

LT-NO8OC is also out of service and in the tripped condition.

Given Tech Spec section 3.3, Instrumentation.

Assuming NO other instruments are out of service and that LT-NOBOA CANNOT be repaired.

When is the plant REQUIRED to be in HOT SHUTDOWN?

A Within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> (you are in 3.0.3).

B Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

x Within 7 days + 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C J

HOT SHUTDOWN is NOT REQUIRED, provided "A"channel is placed in tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. J Answer D References Tech Spec 3.3 M-42 sht 2 Justification References during Exam Tech Spec section 3.3 A - INCORRECT - since both channels are on 1 trip system you just need to place the trip system in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B - INCORRECT - not in 3.0.3 C INCORRECT see "A" above D - CORRECT - see "A" above Question Source New 0Memory Level Comprehension Level Question History:

New 8/31 SXD - add justification and references, check on Instrument #

M B - 9/27- -Archie - How does Hope Creek determine which Instruments satisfy which tech Specs -, logs S U N .

Requirements for surveillances AF NO instrument will require you to shutdown, ask steve what his thought process was.

SXD - AF to provide 2 instruments that will cause a problem in tech Specs.

!Question-# 90 Hope Creek SRO Exam Nov 2005 -

Y 3 SRO I Tier #

Importance 2 Group #

3.7 1 ] 4

,4000 EDGs A2.08 Ability to (a) predict the impacts of the following on the EDGs Initiation of emergency and (b) based on those predictions, use procedures to correct generator room fire control or mitigate the consequences of those abnormal protection system.

operation (CFR: 41.5143.5145.3145.13)

Question At 1000 on November 28th, with Hope Creek operating at loo%, you are performing a Post-Maintenance Run on the

" ADiesel Generator.

1 NEO, 1 Mechanical Supervisor and a Vendor Representative are at the Diesel observing the run.

At 1010,you observe the "A" Diesel Generator trip and receive a FIRE alarm from the " A Diesel Generator room. The N E 0 calls you and informs you:

- he heard an explosion coming from the " A Diesel Generator and it appears that the #8 cylinder has a hole in it.

- the Vendor Representative has been hit by a piece of metal and is bleeding on the floor

- the room is filling up with smoke You P dispatch the fire brigade and a medical team to the scene.

The NE0 and the Mechanical Supervisor pull the injured Vendor Representative from the mom and wait for medical At 1020,the Fire Brigade reports back that there is NO fire on the scene, however, the Vendor Representative has died from the injury.

The Mechanical Supervisor estimates that it will take 1 week to repair the " A Diesel Generator The plant has remained at 100% power throughout this event.

What REQUIREMENTS (if ANY) are there to notify the NRC and if any, what is the LATEST that you are REQUIRED to notify them?

There are NO requirements to notify the NRC.

B Yes, you are required to notify the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

~ ~ ~~ ~~~

Yes, you are required to notify the NRC within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C Yes, you are required to notify the NRC within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

D Answer B References ECG section 9.3.9.2 Justification References during Exam Event Classification Guide - section 9, section 1 1 A - INCORRECT - There are requirements to notify the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for an explosion on site.

B CORRECT based on ECG 9.3.1,for an explosion in the protected area, you must declare an Unusual event and notify the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C - INCORRECT - Unit Shutdown to comply with Tech Specs and Reporting a Fatality are 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reports, however you were required to notify them within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the explosion.

D - INCORRECT - This would be required if the plant were in a seriously degraded condition. however the explosion on site is more restrictive.

Question Source New 0Memory Level 0 Comprehension Level

'vestion History:

,w 9/22 SXD - OK AF - get out tech bases and ask Steve is a Loud bang from an Piece of equipment destroying itself is considered an explosion or should I just write explosion in stem MB - changed to explosion. - .

=

- -A-

Question # 91 -

Hope Creek SRO Exam Nov 2005 7

OR0 I Tier# Group# 2 I 7

' SRO Importance 3.3 3

~

Question Hope Creek is at 80% power when a single eventhalfunction occurred affecting the CRD system.

NO operator actions have been performed.

The operator observed the following indications BEFORE the event and AFTER the CRD System stabili BEFORE AFTER CRD flow controller - flow indications CRD flow controller - setpoint 63 gpm Cooling Water flow Cooling Water Pressure Drive Water Pressure 270 psid PJTpsid Charging Pressure 1475 psig 1650 psig Which of the following is the cause of this event I Assuming NO Operator actions, what actions are REQUIRED by Tech Specs? II A I. The flow control valve failed dosed.

II. Be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B 1. The Stabilizing valves have failed closed.

II. Be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

~~~ ~ ~~

fl I. The flow control valve failed closed.

II. NO Actions are required by Tech Specs because the Control Rods are still OPERABLE.

I. The Stabilizing valves have failed closed.

11. NO Actions are required by Tech Specs because the Control Rods are still OPERABLE.

Answer c References Tech Spec 3.1.3.1 NOHOlCRDHYD-01, CONTROL ROD DRIVE HYDRAULICS, p.11 Justification References during Exam Tech Spec 3.1.3.1 A - INCORRECT - Flow mntrof valve failing dosed will give the above indications. Flow Control valve failing closed will cause the Control Rods to be trippable but inop for causes other than being mechanically bound, however because more than 8 control rods are INOP need to be in HS within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per action c.

B - INCORRECT Stablizing valves failing dosed will NOT cause the above indications.

C CORRECT - While the control valve is still failed close, the Control Rods are still considered OPERABLE because the operator can NOT move them.

D INCORRECT see "C"above Question Source New 0Mentory Level Comprehension Level Question History:

SXD Not SRO - re-write MB - 9/27 - re-wrote question to make more SRO level SXD - OK AF - WA mismatch

Hope Creek SRO Exam Nov 2005 F# 2 Group# 2 I q SRO A2001 Importance Recirculation 4

3 G2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 45.2 145.6)  ?

Question While operating at 100% power, the "A' Recirc pump trips on Bus differential.

Reactor power decreases to 55%. Core and Recirculation Loop flows are as follows:

- A Recirc Loop Flow is 0 gpm

- A Recirc Jet Pump flow is 7.OE 6 lbmlhr

- B Recirc Loop Flow is

- B Recirc Jet Pump flow is 31.OE 6 lbmlhr

- OPRM's are OPERABLE ake credit for 1 I s. 1

- _ J A ACTION - Insert Control Rods to exit the Power-Flow Operating Map REGION I.

MECHANISM - The Fixed Upscale trips B ACTION - Insert Control Rods to exit the Power-Flow Operating Map REGION I.

MECHANISM -The OPRM SCRAM ACTION - Insert Control Rods to exit the Power-Flow Operating Map REGION II.

MECHANISM -The Fixed APRM Upscale trips.

ACTION - Insert Control Rods to exit the Power-Flow Operating Map REGION II.

MECHANISM -The OPRM SCRAM.

Answer A References INPO Question - 22668 Tech Spec Section 2 HC.OP-SO.B50002(Q), REACTOR RECIRCULATION SYSTEM OPERATION, Attachment 2 Justification References during Exam HC.OP-SO.BB-O002(Q), Attachment 2 A - CORRECT - Based on the Core Flow and power given, the plant is in REGION I, since OPRM's are OPERABLE, Operator needs to insert rods to reduce power to dear APRM upscale alarms and exit Region I per RPV-0003.

B - INCORRECT - Tech Specs does NOT take credit for the OPRM trip setpoint C - INCORRECT - Core flow and Power given place unit in REGION I, NOT REGION I1 D - INCORRECT - Tech Specs does NOT take credit for the OPRM trip setpoint Question Source Mod 0Memory Level Comprehension Level Question Histoy:

New 8/31 SXD add references -

MB Added references, modified question somewhat.

SXD - OK AF - A, B both correct MB - Changed distractors A and C

3 RO [Tier# 2 Group # 2 1 7 SRO

- ,5000 Importance Main Turbine Gen. I Aux.

3.8 7

A2.05 Ability to (a) predict the impacts of the following on the Main Generator trip Turbine Gen. / Aux and (b) based on those predictions, use procedures to correct control or mitigate the consequences of those abnormal operation (CFR: 41.5/ 4 3 3 45.31 45.1 3)

Question Hope Creek is shutting down to repair a the Main Turbine from S e e enerator load to 5%, but it remains in the depressed e main g e n A. What is the expected plant response assuming NO operator actions?

8 . What actions shall you direct?

~~

A A. Generator Trips on Reverse Power, Main Turbine trips from Generator Trip.

Generator Output Breakers open B. Ensure Generator and Main Turbine trips, go to AB.BOP-0002 B A Generator Output breaker opens, Turbine Speed increases and is controlled by Speed Regulation circuit.

B. Ensure Main Turbine does NOT overspeed, continue shutdown per 10.Z-0004.

A Generator Trips on Reverse Power, Reactor Scrams on either High Flux or High Pressure.

B. Ensure Generator and Main Turbine trips, go to AB.=-0000 or EO.=-0101 as appropriate. x A Generator trips on Reverse Power, Generator Output Breaker Opens, Turbine trips on Overspeed.

B. Ensure Generator and Main Turbine trips, go to AB.BOP-0002  %

Answer A References HC.OP-AB.BOP-0002. Main Turbine, p. 1 NOH01MNGENO-02, MAIN GENERATOR SYSTEM, Table 2 Justification References during Exam None A - CORRECT - with power 25% NO automatic Reactor Scram will Occur, Generator will trip on Reverse Power due to Power = 0, when Generator Trips, it will cause a Main Turbine Trip and Generator Output breakers open. AB.BOP-0002 give operator actions for a Turbine Generator Trip.

B - INCORRECT - Turbine will trip on Generator trip C - tNCORRECT with Power 25%. Generator/ Turbine Trip will NOT cause a Reactor Scram D - INCORRECT Turbine will trip on Generator Trip, NOT on Overspeed Question Source New 0Memory Level @I Comprehension Level Question History:

New 9/22 SXD - OK RJC should/ shall MB Made changes as requested

Hope Creek SRO Exam Nov 2005 -

L2 SRO p r #

Importance 3 Group#

4.4 1

_.1.7 Generic Ability to evaluate plant performance and make operational judgments based on operating characteristics / reactor behavior I and instrument interpretation. (CFR: 43.5 / 45.12 / 45.13)

Question Given the following

- A Reactor Scram occurred.

- There are still 20 rods at Position 48.

The following sequence of events takes place:

- Scram is reset

- ARI is reset.

Then, there is a break in the scram air header.

Which of the following methods shall you direct the RO to pursue in order to insert control rods?

- -\

mming of Control Rods with SRI Switches locally. rN L - -

,,, q!#

- . L..

vFn 1-B \ Direct attempt to manually drive Control Rods. ---

d Attempt an additional manual scram.

C Direct de-energizing scram solenoids by removing the RPS fuses.

J Answer References Susquehanna Exam August 2003 NOHOICRDHYD-01, CONTROL ROD DRIVE HYDRAULICS EO-0101A Justification References during Exam EOP Flowcharts with entry conditions blacked out A - INCORRECT - Without air, scram inlet and outlet valves should already be open.

B CORRECT -

C INCORRECT - Without air, scram cannot be reset since the discharge vent and drain valves remain dosed.

0 INCORRECT - Scram inlet/outlet valves already open on loss of air.

Question Source Bank 0Memory Level 0 Comprehension Level Question History:

New 918 SXD Need to add justification, seems to be from another plant, make Hope Creek MB 9/26 changed to a different question SXD -OK RJC - Should/shall, add procedure MB added shall, can't find procedure reference because procedures don't seem to give specifics, procedure only says, perhaps Archie can provide specific procedure guidance

Hope Creek SRO Exam Nov 2005 - Y F l [Tier # 3 Group # Y 2.9

_.1.34 Generic Ability to maintain primary and secondary plant chemistry within Y

allowable limits (CFR: 41.10 I43.5 / 45.12)

Question The plant was operating at 20% power. Plant Chemistry reported to the Main Control Room the following chemistry parameters:

- Reactor pH 8.8

- Reactor Water conductivity 11 microrn hos/cm

- Reactor Water chlorides 150 ppb Six hours later with the plant in OPCON 2, Chemistry reports the following:

- Reactor pH 6.5

- Reactor Water conductivity 0.9 micromhos/cm

- Reactor Water chlorides 150 ppb Which one of the following actions is appropriate for these plant conditions?

A Be in OPCON 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and OPCON 4 within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

B Return to OPCON 1 where chemistry would be back in spec.

Stay in OPCON 2 and restore chlorides to within limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in OPCON 3 within the next 12 q, hours and OPCON 4 within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Restore Chlorides to within spec within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> or perform an engineering evaluation.

P Answer p( r;f References INPO Question 24577 UFSAR 5.2.3.2.2.2and UFSAR Table 5.2-8 Justifcation References during Exam UFSAR 5.2.3.2.2.2 and Table 5.2-8 A - CORRECT per ACTION a. -with conductivity exceeding 10mmho/cm be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B - INCORRECT - plausible because based on given conditions for OPCON 2, plant chemistry would be in spec if plant returned to OPCON 1.

C - INCORRECT - plausible if only look at Action b.

D - INCORRECT - plausible if only look at Action c.2 Question Source Mod 0Memory Level @I Comprehension Level Question History:

S X D review 7/27 - Talk to licensee ensure correct answer is correct and once conductivity is < limit, they exit the condition and can return to power.

AF-OK

. I

Hope Creek SRO Exam Nov 2005 RO I Tier # 3 Group # I 7SRO Importance 3.3 Knowledge of the process for managing troubleshooting activities (CFR: 43.5 145.13) r" Question Which of the following condition(s) would REQUIRE Field Engineering to review a Troubleshooting Plan developed in accordance with SH.OP-AP.ZZ-0008. OPERATIONS TROUBLESHOOTING AND EVOLUTIONS PLAN DEVELOPMENT:

I. Equipment is NOT removed from service or tagged and presents a risk of tripping the plant either directly or as a result of causing a major plant transient.

II. Equipment is NOT removed from service or tagged. Could result in an unexpected load reduction, a plant transient, or a reportable event. Should NOT result in a reactor, turbine, or generator trip.

111. Equipment is NOT removed from service or tagged. Could have an effect on plant equipment but shall NOT present a risk of causing an unexpected load reduction, plant transient or reportable event.

IV. Equipment is removed from service or tagged such that troubleshooting or testing activities shall NOT adversely affect the operation or safety of the plant.

A I only B I and II only c I, 11, 111 and IV D

Answer 0 References SH.OP-AP.ZZ-0008, OPERATIONS TROUBLESHOOTING AND EVOLUTIONS PLAN DEVELOPMENT, p. 6 and 9 Justiflcation References during Exam None SH.OP-AP.ZZ-0008. OPERATIONS TROUBLESHOOTING AND EVOLUTIONS PLAN DEVELOPMENT states that Field Engineering SHALL review a troubleshooting plan if the plan is determined to be either HIGH RISK or VERY HIGH Risk. The 4 conditions presented are the 4 conditions outlined in SHOP-8, I= Very High Risk, II=High Risk, Ill=

Medium Risk, IV = Low Risk A - INCORRECT - both High Risk and VERY High Risk must be evaluated B CORRECT C INCORRECT - Medium Risk does NOT need to be evaluated D - INCORRECT - Medium Risk and Low Risk do NOT need to be evaluated Question Source New Memory Levet Compreliension Level Question History:

New 919 SXD - confusing - put in High Risk, Very High Risk, etc.

MB Changed 9126 SXD - Change back to old way to make more difficult MB - 1013 Changed back to old way SXD - OK

'-OK

gR0 I%# 3 Group# I 3 SRO Importance 3.5 3

You are the CRS of Hope Creek on Saturday night. &

M 1 . m r

Maintenance has just completed adjusting the switch on valve BC-HV-F024B RHR LOOP B TEST RET MOV.

Before declaring the valve OPERABLE, which of the following Test activities needs to be performed on BCHV-F024A RHR LOOP A TEST RET MOV in accordance with NC.MD-AP.ZZ-0050, Maintenance Testing Program Matrix:

B A Valve Interlock Test External Leak Check x

Stroke Time Test C

~~

Response Time Test 3

Answer References NC.NA-AP.ZZ-0050, Station Post Maintenace Testing, P.8 NC.MD-AP.ZZ-0050. Maintenance Testing Program Matrix, p 86-89 Justification References during Exam NC.MD-AP.ZZ-0050 A INCORRECT - Valve Interlock test is only applicable to A & B RHR Shutdown cooling valves B - INCORRECT - External Leak Check (p90) is to be performed on an Air-Operated Valve for PMT for packing adjustment.

C. CORRECT - for Limit Switch adjustment for an MOV in the OPEN direction, the following tests need to be performed:

1. Functional Stroke

5. Stroke Time Test

9. Thermal Overload Bypass Surveillance D - INCORRECT, Response Time Test is a Test to determine the time interval from when a specified setpoint or condition is reached until a specified activity occurs. This is needed for an RT (Re-test) on a TRIP UNIT, Replacement (P.31)

Question Source New cl Memory Level Comprehension Level Question History:

New SXD No comment 9/23 AF - Not important, check if you have procedure, maybe rewrite. Valve is broke what is retest requirements. Give them procedure.

MB - 10/27 - re-wrote question

RO I Tier # 3 Group # I 3 SRO Importance 3.1 Question An NE0 has been assigned to enter the Condenser Bay at power to investigate a steam leak. His current radiation history is as follows:

- Annual Exposure to date: 3280 mR TEDE

- Expected dose for this entry: 300 mR

- Highest Expected Dose Rate for the area: 6OOmWhr

- NE0 will be provided with continuous RP coverage during his entry Which ONE of the following describes the REQUIRED action needed to complete the steam leak investigation per NAP 24, Radiation Protection Program, if any, based upon the above conditions:

A Dose Level Extension must be obtained prior to entry.

B Planned Special Exposure must be obtained.

J A Special RWP must be written.

NO additional action required.

7

- x Answer D References INPO Question 19298 NC.NA-AP.ZZ-0024, RADIATION PROTECTION PROGRAM p. -

27 Justzpcation References during Exam None A - INCORRECT - since operator has already exceeded 3000 TEDE, the next extension is NOT required until he will exceed 4000 TEDE B - INCORRECT - Planned Special Exposure is only required if dose is to exceed 10CFR20 limits, which this will NOT C - INCORRECT - Special RWP is only required to be written for entry into a VHRA ( ~ 5 0 raddhr).

0 per 5.8.1. In addition, Section 5.1 1.3 of NAP-24, For work situations requiring immediate access, RP may substitute continuous coverage in lieu of an RWP..

D. - CORRECT -Already extended.

Question Source New 0Memory Level Compreliension Level Question History:

SXD review 7/27 - too easy - LOD 1 AF-OK SXD - Beef up MB - 1013 -Wrote new question SXD - OK AF-OK

IQuestion # 99 Hope Creek SRO Exam Nov 2005 -

z I Tier# 3 Group # 1 Importance 4

-2

+.22 Generic Knowledge of the bases for prioritizing safety functions during abnormallemergency operations (CFR: 43.5 / 45.12)

Question Hope Creek is experiencing an ATWS You are the CRS and you just directed the RO to inhibit the automatic initiation of the Automatic Depressurization System (ADS).

Which of the following is the reason why you directed the RO to inhibit the automatic initiation of ADS?

To prevent

~

A A power excursion due to low pressure ECCS injection l/

B Large irregular neutron flux oscillations x

Exceeding 110°F Suppression Pool Temperature before boron injection Causing a Brittle fracture of the Reactor Vessel J

Answer A References INPO Question 24595 HC.OP-EO.ZZ-OlOlA, ATWS - RPV CONTROL, P. 18 Justification References during Exam None A - CORRECT - Per EOP 101A bases - Further, rapid and uncontrolled injection of large amounts of relatively cold.

unborated water from low pressure injection systems may occur as RPV pressure decreases to and below the shutoff heads of these pumps. Such an occurrence would quickly dilute in-core boron concentration and reduce reactor coolant temperature. When the reactor is NOT shutdown, or when the shutdown margin is small, sufficient positive reactivity might be added in this way to cause a reactor power excursion large enough to severely damage the core.

B INCORRECT - ADS initiation would NOT cause flux oscillation but rather a rapid reduction in core power due to voids C - INCORRECT This may or may NOT be true but it is NOT the reason for inhibiting ADS D - INCORRECT - While an ADS actuation will cause a Thermal Shock to the vessel, the vessel will be de-pressurized so you will NOT have a PTS concern Question Source Mod k4 Memory Level 13 Compreliension Level Question History:

SXD review 7/27 - OK AF - on "D"changed to Brittle fracture.

MB - 8/24 - Made changes as requested RJC - borderline SRO, get Archie to offer suggestions to make more SRO.

OR0

7 SRO 1

Importance 2.8 1 5-

-.4.36 Generic Knowledge of chemistry/health physics tasks during emergency operations (CFR: 43.5) 3-Question Given the following conditions:

The plant is operating at 100% power The results from the last Rx water sample are:

- pH = 7.0

- Conductivity = 1 mmho/un

- Chlorides = 0.5 ppm

- An ATWS-LOCA event occurs

- The core is flooded using only Torus and CST water

- RPV level is then stabilized as SLC injects into the RPV.

Which of the followina Chemistrv conditions would result from this event?

A Conductivity = 12 mmholcm, pH = 6.5 & Chlorides = 1.O pp B Conductivity = 12 mmholcm. pH = 8.0 & Chlorides = 1.0 ppm Conductivity = 1 mmholcm. pH = 6.5 & Chlorides = 1.0 pprn C

A Conductivity = 12 mmho/cm, pH = 6.5 8 Chlorides = 0.5 pprn D

Answer c References Hope Creek Bank - Q62251 M-38-0 sht 1 8 2 A

Justification References during Exam None - w A - CORRECT - Conductivity = 12 mmholcm, pH = 6.5 8 Chlorides = 1.O ppm. Conductivity should rise above the original value and will be higher than 10 mmho/un due to poor quality Torus water and soluble material transported from the Drywell. pH should remained only slightly changed as Torus and CST water is neutral and SLC injection should drop the pH slightly but will NOT cause pH to lower below 5.3. Chlorides should rise slightly and may exceed 1.O ppm.

B - INCORRECT - Conductivity = 12 mmho/cm, pH = 8.0 & Chlorides = 1.0 ppm. pH should remained only slightly changed as Torus and CST water is neutral and SLC injection should drop the pH slightly but will NOT cause pH to lower below 5.3.

C - INCORRECT - Conductivity = 1 mmho/cm. pH = 6.5 & Chlorides = 1.0 ppm. Conductivity should rise above the original value and will be higher than 10 mmho/cm due to poor quality Torus water and soluble material transported from the Drywell.

D - INCORRECT - Conductivity = 12 mmho/cm, pH = 6.5 & Chlorides = 0.5 ppm. Chlorides should rise slightly and may exceed 1.O

~~

Question Source Bank 0Memory Level Comprehension Level Question History:

New 9/9 SXD - add references MB - 9/26 - Need references from Archie AF - rewrite - NOT linked to job - OSC, epep202, OSC duties 3 - rewrote questions, need Archie to assist in verifying Question is OK.

j - swapped out with bank question.