Text

Final - RO & SRO Written
	ML023240459
Person / Time
Site:	Nine Mile Point
Issue date:	09/27/2002
From:	Pisano L; Constellation Nuclear
To:	Conte R; NRC/RGN-I/DRS/OSB
	Conte R
References
	50-220/02-303 50-220/02-303
	Download: ML023240459 (140)
	v • d • e

ES-401 NMP #1 9/02 EXAM Written Examination Form ES-401-9 Note: Resolutions to comments in bold and italics Review Worksheet

1. 2. 3. Psychometric Flaws 4. Job Content Flaws & 01iq N 6. 7.

Q# LOK LOD r Ti (F/H) (1-5) Stem Cues T/F Cred. IPartial Job- Minutia #/ Back- §, W S Explanation Focus j Dist. Link units ward W POI. __

1 F 2 E Add word "containment" to the stem- recommended changes made 2 H 2 S 3 H 3 E TYPO - fixed 4 H 2 E Bold, underline and cap the word "required" - - recommended changes made 5 H 2 S 6 F 2 S 17/ F 2 x U Explain why distractors"A&D" are credible also reword the stem.

S 10 Question and K/A replaced.

8 F 1-2 S Borderline level of difficulty okay in limited #s 9 H 2 E add the word "execute" to the "D"distractor - - recommended changes made 10 H 3 S Instructions

[Refer to Section D of ES-401 and Appendix B for additional information regarding each of the following concepts.] I

1. Enter the level of knowledge (LOK) of each question as either (F)undamental or (H)igher cognitive level.

2. Enter the level of difficulty (LOD) of each question using a 1 - 5 (easy - difficult) rating scale (questions in the 2 - 4 range are acceptable).

3. Check the appropriate box If a psychometric flaw Is Identified:

The stem lacks sufficient focus to elicit the correct answer (e.g., unclear Intent, more Information is needed, or too much needless Information).

The stem or distractors contain cues (i.e., clues, specific determiners, phrasing, length, etc).

The answer choices are a collection of unrelated true/false statements.

More than one distractor Is not credible.

One or more distractors is (are) partially correct (e.g., if the applicant can make unstated assumptions that are not contradicted by stem).

4. Check the appropriate box if a job content error Isidentified:

4 The question is not linked to the Job requ rements (i.e the quest on has a valid K/A but, as written, Is not operational in content).

The question question conta requires ns the datareca The with of anknowledge levelisofoaccuracy unrealistic that c icorforinconsistent the closed reference test mode (i.e., it is not required to be known from memory).

units (e.g., panel meter in percent with question in gallons).

The question requires reverse logic or application compared to the job requirements.

5. Check the appropriate box if the sampled question does not match the approved K/A or an SRO-only question is not at the SRO level.

6. Based on the reviewer's judgment, is the question as written (U)nacceptable (requiring repair or replacement), in need of (E)ditorial enhancement, or (S)atisfactory?

7. For any "U"ratings, at a minimum, explain how the Appendix B psychometric attributes are not being met.

ES-401 2 Form ES-401-9

1. 2. 3. Psychometric Flaws 4. Job Content Flaws J ;Ofhier 6. 7.

LOK LOD I S.

I I

. . I pI/w1 (1-.F 5Rt~m ICues T/F Cred. Partial Job- Minutia I #/ IBack- Q U/E/S Explanation

. I DiFt.' unitslwardW 11 F 3 E Change "may be" in the stem to "should be" - revised as recommended 12 H 2 S 13 F 1-2 S Borderline LOD also don't agree it is higher order - okay In limited number. Licensee Indicatedthat this was Intended to be a memory level question.

14 H 3 S 15 H 3 S 16 F 2 S 17 H 2 U Check to make sure battery rating is required knowledge from memory Replaced with a new question 18 F 2 X S check KA - replace K/A - Agreed K/A match Is okay 19 H 2 E Fig "B"- move title after Fig, "B"in the stem - reworded the stem to clarify 20 H 2 E reword stem to "may fall as a direct..." - recommended changes made 21 F 2 E revise wording of stem change "must be" to "shall be" in service to agree with procedure - recommended changesmade 22 F 2 E add to stem SOP-9.1 - recommended changes made 23 F 2 S 24 F 2 S 25 H 3 E #25 appears to have the wrong proposed answer - see justification section - Answer verified correctand correctedjustification.

26 F 2 S 27 H 3 S 28 H 3 S 29 H 2 S Make sure that valid RO learning objectives exist for all EOP application type questions on this exam. Licensee verified valid RO learning objectives exist 30 H 2 S .......... same comment as #29. Licensee verified valid RO learning objectives exist

1. 2. 3. Psychometric Flaws 4. Job Content Flaws . hi' 6. 7.

Q# (F/H)

LOK ILOD (1-5) Stem Cu 1ues.

, ro0. Paral Jo b- Mnutia #/ Back /E/ Explanation Focus _ Dist. Link units ward KA I___

31 H 2 E same comment as #29 &30. Licensee verified valid RO learning objectives exist Deleted commas In each distractorand reorded distractors.

32 H 2 S Give applicants entire reference not just Att 5 also don't tell appplicants which question reference will be used to answer.

33 F 2 E modify stem - stem modified change ? To:

34 H 2 S 35 H 2 E In stem add the word "temperature" after RWCU pump areaand HTX 36 F 2 S 37 F 2 E change verb in "A&B" from "may" to "will" to agree with "C&D" recommended changes made 38 F 2 S 39 H 2 S 40/ F 2 X S "A&B" distractors verify credible. Licensee verified distractors 45 credible.

41 F 2 E Add words In the stem "the core" in front of srpay pump min flow relief valve - recommended changes made 42/ F 2 X X X U 12 be used to fill torus? Tighten stem focus Tightened stem focus also

$7 revised distractors.

43 H 3 S 44 H 2 S 45 F 2 S #44 stem may cue #45, changed#44 SRM#12 withdrawn to maintain on scale 46 F 2 S 47 F 3 S 48 H 2 S 49 H 2 S 50 H 2 S stem 2 indications for valve open and 2 for valve closed - revised stem

ES-401 4 Form ES-401-9

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 1 l. 6. 7.

/ Back- Q R V U/E/S Explanation (F/H) Stem Cues T/F (1-5) Focus Linjk Minutia Iunits ward KZA.0-PA Dist.' Partial Job-Cred.

51 H 2 X E stem unclear - rewrote stem 52 F 2 S 53 H 3 E change stem: "may result" to "will result" . recommended changes made 54 H 3 E reorder bullets in stem - recommended changes made 55 H 2 S H 2 E Change "may" to "will" in each answer choice - recommended 56 changes made 57 H 2 S 58 H 2 S 59 F 2 S 60 H 2 S 61 H 2 S 62 F 2 E Add to stem RBEVS in a normal "standby" line-up - recommended changes made 63 H 2 E Modify the stem - recommended changes made 64 F 3 E Add "EDG 103" In front of "output breaker" In choices "A&B" recommended changes made 65 F 2 S 66 F 2 S 67 F 2 E Add the word "the" to choices "B,C, and D" 68/ F 3 X X U Delete"before next use" in answer col. 2 also "B&D" Implausible 71 distractors Revised stem and distractors 69 F 2 S 70 H 2 S NUREG-1021, Revision 8, Supplement I

ES-401 5 Form ES-401-9 Q# 1.

LOK 2.

LOD 3. Psychometric Flaws 4. Job Content Flaws 5.Othere 6. 7.

I (F/H) (1-5) Stem Cues T/F rCred. Partial JOb- Minutia Focu jDSt.

D Link

/ Back-Junits ward qi' IQ U/E/S Explanation 71 F 2 S 72 F 2 S 73 H 2 S 74 F 2 S 75 H 2 S 76 H 2 S 77 F 2 S 78 F 3 S 79 F 2 S 80 F 2 S 81 H 2 S 82 H 2 E change "dump" valve to" bypass" valve in stem - recommended changes made 83 H 2 S 84 H 2 S 85 H 2 S 86/ H 3 X U Make sure "B" Is incorrect - Distractor"B" was revised 33 87 H 2 X U "B&D" distractors implausible with cavity drained also revise wording in "A"to add "pump'. Added word and revised distractors.

88/ F 1-2 U LOD Borderline - tested OPs test - Replaced question same K/A used 36 89 F 2 S 90 H 3 X U Ability to locate not tested also better tested in Part C of the exam Modified question and distractors significantly.

NUREG-1021, Revision 8, Supplement 1

.AA1 S Form ES-401-9 1 2. 3. Psychometric Flaws 4. Job Content Flaws 5,.tber 6. 7.

Q# LOK LOD1i (F/H) (1-5) Stem Cues T/F Cred. Partial Job- Minutia #/ Back- Q S U/E/S Explanation Focus Dist. Link 1 units ward 1: I _

91/ H 2 X U Not ter 3 genetic (system type) replace question Replaced K/A and

)o question 92 H 2 S 93 H 2 X E Correct answer for #34 may be cued by reference (core map) - deleted reference as a handout 94/ H 2 X U Not tier 3 generic (system type) replace question Replaced K/A and

)6 question 95/ F 2 X U Distractors don't appear credible for a very high Rad area Distractors 35 were revised 96 F 1-2 X X X U Tighten stem and answer choices. As currently written, "first frisking" and "each area" doesn't make sense. Add minimum to frisking requirement. See proposed rewrite. Tightened stem and reworded choices.

97 H 2 X X U "D"isn't this a true statement regardless Not tier 3 generic (system type) replace question Replaced K/A and question 98 H 2 S 99 H 2 S 100 F 2 S 3RO 13 F 2 S 14 F 2 X X U "A, B & D"distractors credible? Stem cues Power ops to contiue Revised stem 15 H 1-2 X U low LOD. Direct lookup with the reference. Better tested in dynamic scenario or JPM Wrote a new question that matches the K/A 20 H 3 S 21 F 1-2 U low LOD. Direct lookup with the reference. Already tested on dynamic scenario Revised distractors 24 H 1-2 U low LOD. Direct lookup with the reference. Already tested on dynamic scenario Used same K/A and wrote new question

1. 2. 3. Psychometric Flaws 4. Job Content Flaws . Othf.. 6. 7.

(F/H)

_ (1-5) Stem Cues TIF Cred. Partial Job- Minutia #/ Back-"

Focus __ Dist. ~Link _ nits ward~

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U Explanation K/A mismatch, low LOD test more than one EOP leg or EOP at the time 27 F 2 for an SRO Direct lookup with the reference. Already tested on dynamic scenario Replaced K/A and question 33 F 1-2 E low LOD. Rewrote question to calculate time to boiling given RPV I II__ operatinghistory 36 H 1-2 U low LOD, test more than one EOP leg or EOP at the time for an SRO.

Rewrote question 39 H 2 S o0 H 2 S 12 H 2 E don't just give the one section 50.72 &73 give all 1 OCFR2-50. Add a comma to the fourth bullet, put "the" in front of NRC 18 H 3 S 55 H 1-2 U low LOD, test more than one EOP leg or EOP at the time for an SRO Already tested on dynamic scenario? Replaced K/A and question 56 H 1-2 X X U MSIVs Isolated given in the stem clues the correct answer both vessel and containment isolations required. Replaced question 57 H 3 U Already tested on dynamic scenario Replaced K/A and question 39 F 2 E Simplify answer choice wording 31 F 2 X X U Distractor"D" could be correct. If error discovered next day how could it distractor be reported any earlier? Revised stem and 34 H 3 S 35 H 3 E Add "a" before refueling outage 38 F 1-2 X U C&D don't appear credible also LOD 39 F 2 E Add "power" after "100%". 2 nd bullet: add ' for "valve's" ; remove ' for "its".

32 H 2 S Change stem "what individual is responsible for final authorization to dose limit to 1600 mrem" extend someone's annual Nine Mile

)3 H 2 S Need to reword stem - all answers could be correct (i.e securing "within" 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months could mean securing in 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ). Proposed re-write:

"Which one of the following describes how much longer Drywell venting can continue? Drywell venting can continue for another..."

4 H 2 S Distractor C and D too similar, come up with a replacement (maybe "raise RPV level to 100 inches"?)'. A bit detailed for a tier 3 generic.

Replaced "D" distractor

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NMP1 Licensed Written Examination Candidate Reference List Question Number Reference given RO-02 SRO-03 Power-To-Flow Maps RO-13 SRO-26 EOP Flowcharts RO-20 SRO-06 EOP Flowcharts RO-25 SRO-18 EOP Reference Power-To-Flow EOP given Flowcharts Flowcharts Maps RO-29 SRO-38 fOP Flowcharts RO-30 Nl -EOP- 1 Attachments 6, 15, 18 RO-31 SRO-25 FOP Flowcharts KU-32 RKU-43 N1-OP-21F with Attachments RO-35 SRO-41 EOP Flowcharts RO-49 EOP Flowcharts Nl-ARP-F4 (except F-4-4-1)

RO-71 SRO-58 EOP Flowcharts RO-92 SRO-91 Technical Specification Sections 3 & 4 (w/o Bases)

RO-93 NI-REP-20 RO-99 SRO-99 N1-SOP-9 Flowchart SWU-14 Technical Specification Sections 3 & 4 (w/o Bases)

SRO-15 Mi2Ar r Iowinarts SRO-21 EOP Flowcharts SRO-24 EOP Flowcharts SRO-33 NI-ODP-OPS-0108 SRO-40 EOP Flowcharts SRO-42 EOP Flowcharts 10CFR50 SRO-48 Technical Specification Sections 3 & 4 (w/o Bases)

SRO-55 EOP Flowcharts SRO-56 EOP Flowcharts SRO-57 EOP Flowcharts SRO-84 Technical Specification Sections 3 & 4 (w/o Bases)

SRO-85 Technical Specification Sections 3 & 4 (w/o Bases)

SRO-88 N1-FHP-25 SRO-94 EOP Flowcharts

U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: Region I Date: September 30, 2002 Facility/Unit: Nine Mile Point Unit 1 License Level RO Reactor Type: GE BWR-2 Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent.

Examination paper will be collected six hours after the examination starts.

Applicant Certification All work done on this examination is my own. I have neither nor received nor received aid.

Applicant's Signature Results Examination Value: 100 Points Applicant's Score Points Applicant's Grade Percent NUREG-1021, Revision 8, Supplement 1

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 1 1 KJA # 295007 295007 Importance Rating 2.9 3.1 Proposed Question: RO 001, SRO 002 The plant is shutdown, with the following:

Reactor Pressure is 130 psig

Shutdown Cooling (SDC) is being lined-up Which one of the following describes the SDC containment isolation valves that can be opened?

A. INSIDE or OUTSIDE but not both B. Only the INSIDE C. Only the OUTSIDE D. Neither the INSIDE nor OUTSIDE Proposed Answer: A Justification: A correct- high pressure (120 psig and above) will prevent both valves from valves from opening BIC incorrect - high pressure will prevent both valves from opening but is not specific to which valve D is incorrect- one valve can be opened with high pressure but not both Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-4, D.3 K/A: 295007 AK2.05. Knowledge of the interrelations between HIGH REACTOR REACTOR PRESSURE and the following: Shutdown cooling: Plant-Specific PRESSURE 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank # X12721 Modified Bank #

New History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group# 1 1 K/A # 295009 295009 Importance Rating 3.3 3.4 Proposed Question: RO 002, SRO 003 The plant is operating at 100% power, with the following:

A Feedwater malfunction causes RPV water level to drop to -80" before being stabilized by the operator

All rods fail to insert on the scram

Reactor power is 25%

Bypass valves are controlling pressure Which one of the following describes where the plant is operating on the Power to Flow Map?

The plant is operating with...

A. adequate forced circulation B. inadequate forced circulation C. adequate natural circulation D. inadequate circulation Proposed Answer: D Justification: D is correct - Recirc pumps tripped at +5 inches; no forced flow exists. Level is -80 inches; no natural circulation exists at this low level; natural circ line is valid only at normal operating level. AIB are incorrect - Recirc pumps tripped at +5 inches; no forced flow exists.

C is incorrect - Reasons as described for D.

Proposed references to be provided to applicants during examination: Power to Flow Map, Five Loop Operation Technical Reference(s): DDC IE00135 Power to Flow Map, Five Loop Operation NI-ODP-PRO-0305 K/A: 295009 LOW AK1.05. Knowledge of the operational implications of the following REACTOR concepts as they apply to LOW REACTOR WATER LEVEL: Natural WATER LEVEL circulation 10 CFR Part 55 Content: 55.41(5) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 1 1 KJA # 295009 295009 Importance Rating 3.9 3.9 Proposed Question: RO 003, SRO 004 The plant is operating at 100% power, with the following:

12 FW pump is running and 11 FW pump is in standby

A turbine trip occurs

"* Reactor Water level drops to 35 inches

"* Two minutes later, Reactor Water level is stable at 72 inches With no operator action, which one of the following describes the Feedwater pumps that are running to control Reactor Water level?

A. 11 and 12 running with 12 FCV closed B. 11 and 13 are both injecting to control level C. 11 and 12 running with 11 FCV closed D. 12 and 13 are both injecting to control level Proposed Answer: C Justification: C is correct- Turbine trip and reactor level of 53" both initiate HPCI, 12 FW will control at 72" while 11 FW controls at 65" BID are incorrect 13 FW is not available two minutes after a turbine trip A is incorrect 12 FW will raise level to 72" driving 11 FW FCV closed Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-1 6, sect B.2, page 10 K/A: 295009 LOW AA1.01 Ability to operate and/or monitor the following as they apply to REACTOR LOW REACTOR WATER LEVEL: Reactor feedwater WATER LEVEL 10 CFR Part 55 Content: 55.41(3) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 1 1 K/A# 295010 295010 Importance Rating 2.6 2.6 Proposed Question: RO 004, SRO 005 The plant is operating at 40% power, with the following:

Makeup to the drywell is in progress from the 11 nitrogen makeup line

A small steam leak develops within containment

Drywell pressure rises to 3.7 psig Which one of the following identifies the additional condition REQUIRED to cause automatic closure of the Drywell Nitrogen Makeup and Bleed Isolation Valves?

A. The CAD bypass switches in NORMAL.

B. The CAD bypass switches in BYPASS.

C. RPV level below 5 inches and the CAD bypass switches in NORMAL.

D. RPV level below 5 inches and the CAD bypass switches in BYPASS.

Proposed Answer: A Justification: A is correct- pressure above isolation with isolation circuit in enabled.

C incorrect- level input not required. D incorrect- level input not required and isolation is bypassed. B incorrect- isolation is bypassed Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-223-1-04 K/A: 295010 HIGH AA1.03 Ability to operate and/or monitor the following as they apply to DRYWELL HIGH DRYWELL PRESSURE: Nitrogen makeup: Plant-Specific PRESSURE 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank # X Modified Bank #

New History Susquehanna NRC '99 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 1 1 K/A # 295014 295014 Importance Rating 4.0 4.1 Proposed Question: RO 005, SRO 008 The plant is operating at 100% power, with the following:

A failure to scram occurs

All rods fail to insert

Liquid Poison is initiated

RPV water level is -94 inches and stable

RPV pressure is 1000 psig and stable

Reactor power is 10%

Which one of the following, by itself, will cause reactor power to rise?

A. Raising RPV water level to -41 inches with Feedwater B. Terminating and preventing all RPV injection, allowing RPV level to drop C. Terminating all injection, initiating ADS, and allowing RPV pressure to drop D. Starting a second CRD pump while holding RPV water level constant with Feedwater Proposed Answer: A Justification: A correct - level is lowered below -41'" to lower power. B incorrect this has already happened and yielded the opposite effect.

C incorrect - lowering pressure voids the core further, lowering power (void coeff still predominant at this power level). D incorrect Injection is below core plate, FW heating is gone, thus this source provides essentially no change in subcooling or reactivity insertion.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Ni-ODP-PRO-0305 K/A: 295014, AA1.07, Ability to operate and/or monitor the following as they apply INADVERTENT to INADVERTENT REACTIVITY ADDITION: Cold water injection REACTIVITY ADDITION 10 CFR Part 55 Content: 55.41(5) X 55.43 55.45 Question Source: Bank # X Modified Bank#

New History LaSalle 1, 1996 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 1 1 K/A# 295015 295015 Importance Rating 4.0 4.1 Proposed Question: RO 006, SRO 009 The plant is operating at 100% power, when a scram occurs, with the following:

RPS remains energized

All rods fail to insert

Scram fuses are now pulled

Scram Pilot Solenoid Valve lights are deenergized

Scram air header pressure is 0 psig

All rods remain at their same positions Without re-installing the RPS fuses, which one of the following methods would insert the rods?

A. Repeated manual scrams B. Manually driving rods C. Venting the over-piston areas D. Individual rod scram switches Proposed Answer: C Justification: C correct - this is the only available method to insert the hydraulically locked rods with re-pressurizing the scram air header. AIB/D incorrect - all require that fuses be reinstalled to re-pressurize Proposed references to be provided to applicants during examination:None Technical Reference(s): 01-OPS-006-344-1-11, NI-EOP-3.1 K/A: 295015 AK2.04, Knowledge of the interrelations between INCOMPLETE Incomplete SCRAM and the following: RPS Scram 10 CFR Part 55 Content: 55.41(10) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group# 1 1 K/A# 295015 295015 Importance Rating 3.8 3.8 Proposed Question: RO 007, SRO 010 A failure to scram is in progress, with the following:

"* Reactor power is oscillating between 15% and 35%

"* ERVs are cycling

"* ECs are in service

"* RO has been directed to manually open ERVs until RPV pressure drops to 965 psig Which one of the following describes the reason for controlling pressure at 965 psig?

Pressures higher than 965 psig...

A. increase the probability of ERV failure.

B. do not ensure adequate margin to the ERV lift setpoints.

C. exceed the 100% steam flow turbine bypass capacity.

D. will allow reactor power oscillations to continue.

Proposed Answer: B Justification: B is correct - Candidate must conclude that the MSIVs are closed, based on the range within which power is oscillating while ERVs are cycling in response. Per EOP Bases, page 136, with the MSIVs closed, the 965# value simply ensures "adequate margin to the ERV lift setpoints." Therefore, pressures higher than 965# provide less than adequate margin to the setpoints. C is incorrect This would be the 'classic' reason for choosing the 965# value IF the MSIVs were open. For these stem conditions, the MSIVs are closed; therefore, 100% steam flow turbine bypass capacity is irrelevant. A is incorrect - ERV failure probability is based on the rate of ERV cycling, not the given margin that might exist between actual RPV pressure and the lowest ERV lift setpoint.

D is incorrect - Oscillating reactor power is contingent upon RPV level and/or pressure oscillations.

The bottom-line purpose of taking manual ERV control is stop these oscillations. Just because pressure might be higher that 965 psig does not mean it is also oscillating.

I Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-EOP-3, EOP Bases N1-ODP-PRO-0305 K/A: 295015, Incomplete AK1.04, Knowledge of the operational implications of the following concepts as they Scram apply to INCOMPLETE SCRAM: Reactor pressure 10 CFR Part 55 Content: 55.41(10) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None 1-02-096

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 1 1 K/A # 295006 295006 Importance Rating 4.3 4.4 Proposed Question: RO 008, SRO 017 The plant has just scrammed from rated power.

Which one of the following identifies the control rod position indications on the F Panel before the scram is reset?

A. XX B. Blank C. -99 D. Red bars Proposed Answer: B Justification: B is correct - Both digit positions are blank. AIC are incorrect related to Process Computer RPIS, but not F Panel. D is incorrect Two red bars are present at position 48 (full-out); two green bars are present at the full-in position.

Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-201-1-02 K/A: 295006 SCRAM AA2.02, Ability to determine and/or interpret the following as they I apply to SCRAM: Control rod position 10 CFR Part 55 Content: 55.41(7) X T55:43 55.45 Question Source: Bank #

Modified Bank #

New NEW History

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 1 1 K/A # 295007 295007 Importance Rating 3.4 4.0 Proposed Question: RO 009, SRO 001 The plant is operating at 100% power, with the following:

The EPR begins to fail

Reactor power rises above 100% during the transient

CSO has lowered recirc flow to restore power to 100%

Reactor pressure has stabilized at 1080 psig

EPR is still in control Which one of the following describes the required operator action?

A. Lower the MPR setpoint until the MPR takes control.

B. Lower recirc flow until the MPR takes control.

C. Perform a normal plant shutdown within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

D. Scram the reactor and execute N1-SOP-1.

Proposed Answer: D Justification: D correct - RPV pressure has exceeded a T.S. scram setpoint without initiating an automatic scram. A/B incorrect - plausible for the candidate who does not have a firm understanding of pressure regulator operation and/or T.S. requirements. C incorrect - just a reasonable, plausible distractor in conformance with other T.S. action requirements.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Tech Spec 2.1.2.c K/A: 295007 HIGH 2.1.33: High Reactor Pressure -Ability to recognize indications for REACTOR system operating parameters which are entry level conditions for PRESSURE technical specifications 10 CFR Part 55 Content: 55.41(10) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 1 Group# 1 K/A # 295025 Importance Rating 4.3 Proposed Question: RO 010 The plant is operating at 100% power, with the following:

Generator load reject occurs

Reactor automatically scrams

All Turbine Bypass Valves fail 100% closed Which one of the following describes the impact on the Reactor Coolant System Safety Limit and its associated RPS scram setpoint?

SAFETY LIMIT SCRAM SETPOINT A. Exceeded Not exceeded B. Exceeded Exceeded C. Not exceeded Not exceeded D. Not exceeded Exceeded Proposed Answer: D Justification: D is correct - Candidate must know that the RCS S.L. is 1375 psig and that the associated LSSS/SCRAM setpoint is </= 1080 psig. The generator load reject analysis is not concerned with the RCS Si.; rather its concern is with the MCPR S.L. for Fuel Clad Integrity. MCPR is preserved by way of the high pressure and/or high flux scram that results from the TSV closure with failed TBVs. 1375 psig is not exceeded, but the LSSS (scram setpoint) for RPV pressure is exceeded. A/BIC are incorrect - different combinations that do not conform to the correct answer.

Proposed references to be provided to applicants during examination: None Technical Reference(s): TS 2.2.1, TS 2.1.2 (inc. bases)

K/A: 295025 HIGH EA2.01, Ability to determine and/or interpret the following as they apply to REACTOR HIGH REACTOR PRESSURE: Reactor pressure PRESSURE 10 CFR Part 55 Content: 55.41(5) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 1 1 K/A # 295037 295037 Importance Rating 3.4 4.1 Proposed Question: RO 011, SRO 022 The plant has experienced a seismic event, with the following:

Reactor Power is 40%

MSIVs are closed

ERVs are controlling pressure

RPV level is being controlled with Feedwater

Liquid Poison Tank drain valve has broken off, causing the tank to drain Which one of the following alternate boron injection methods should be used per N1-EOP-3.2?

A. Liquid Poison Hydro Pump B. Liquid Poison Pump from test tank C. RWCU system from cleanup filters D. Condensate system from demineralizers Proposed Answer: C Justification: C correct - per EOP-3.2. A incorrect - Connects to LP tank, which is drained. B incorrect - Not allowed by EOP-3.2, and no boron in that tank anyway. D incorrect - No connection to LP system.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-EOP-3.2 K/A: 295037 Scram EK2.13. Knowledge of the interrelations between SCRAM CONDITION Condition PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR Present and UNKNOWN and the following: Alternate boron injection methods: Plant Reactor Power Specific Above APRM Downscale or Unknown 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank # X Modified Bank #

New History Dresden 1996 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 1 1 KIA# 295037 295037 Importance Rating 4.6* 4.6*

Proposed Question: RO 012, SRO 023 A failure to scram has occurred, with the following:

Mode Switch is in SHUTDOWN

All RPS Rod Group Solenoid lights are OFF

Scram Valve indicator lights (BLUE) are OFF Which one of the following is required to insert the control rods per N 1-EOP-3.1?

A. Depress both RPS pushbuttons B. Repeated manual scrams C. Remove rod group fuses D. Vent the Scram Air Header Proposed Answer: D Justification: D is correct- SCRAM valves are not open, the SCRAM header is still pressurized. A/BIC incorrect- RPS has already been tried and failed.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1 -EOP-3 K/A: 295037 SCRAM EA1.01. Ability to operate and/or monitor the following as they apply CONDITION to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE PRESENT AND APRM DOWNSCALE OR UNKNOWN: Reactor Protection System REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank# X12817 New History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group# 1 1 K/A # 500000 500000 Importance Rating 3.3 3.9 Proposed Question: RO 013, SRO 026 Which one of the following identifies the post-LOCA H2 and 02 concentrations at which a loss of containment integrity may result from a deflagration event?

DRYWELL H2 DRYWELL 02 TORUS H2 TORUS 02 A. 6% 5% 0% 4%

B. 5% 6% 5% 5%

C. 0% 6% 5% 6%

D. 7% 4% 0% 4%

Proposed Answer: A Justification: A correct - >= 6% H2 and >= 5% 02 in same, or either, space. BICID incorrect - combination of concentrations < deflagration Proposed references to be provided to applicants during examination: N1-EOP-4.2, HYDROGEN CONTROL Technical Reference(s): 01-OPS-006-344-1-19 N1-EOP-4.2, HYDROGEN CONTROL K/A: 500000, High EK1.01, Knowledge of the operational implications of the following Containment concepts as they apply to HIGH CONTAINMENT HYDROGEN Hydrogen CONCENTRATIONS: Containment integrity Concentrations 10 CFR Part 55 Content: 55.41(10) X 55.43 55.45 Question Source: Bank # X 12849 Modified Bank #

New History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 3 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group# 1 2 K/A # 295001 295001 Importance Rating 3.3 3.3 Proposed Question: RO 014, SRO 028 The plant is operating at 55% power with the following:

4 Reactor Recirculation loops are operating

Reactor Recirculation Pump 11 is ready for restart The following events then occur:

An individual controller malfunction results in a speed decrease for Recirc Pump 14

Operation is now in the Restricted Zone

APRM power has stabilized

Reactor Period is infinite and steady

No LPRMs are in alarm Which one of the following describes the required operator action?

A. Start Recirc Pump 11.

B. Manually scram the reactor and enter SOP-1.

C. Use the RECIRC MASTER controller to raise recirc flow to exit the Restricted Zone.

D. Take manual control of Recirc Pump 14 to raise recirc flow to exit the Restricted Zone.

Proposed Answer: C Justification: C is correct- This is the precise action specified by SOP-2. A is incorrect Specifically prohibited by SOP-2. B is incorrect - There are no indications of thermal hydraulic instability requiring this action. D is incorrect - This action will probably work; however, SOP-2 specifically requires use of the RECIRC MASTER controller and provides the alternative of CRAM rod insertion.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-SOP-2 K/A: 295001 Partial or AA1.05, Ability to operate and/or monitor the following as they apply to Complete Loss of PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW Forced Core CIRCULATION: Recirculation Flow Control system Flow Circulation 10 CFR Part 55 Content: 55.41(10) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank# X457 New History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level RO Tier # 1 Group # 1 K/A # 295001 Importance Rating 3.1 Proposed Question: RO 015 The plant is operating at 100% power, with the following:

A malfunction results in a lowering speed on all five running Recirculation Pumps

Operator action has been taken to stop the lowering flow

Operation remains outside the Restricted Zone on the Power to Flow Map Which one of the following describes the effect of this flow reduction on the APRM margins to rod blocks and scrams?

A. The same margins exist now as before and the margins will remain constant.

B. Lesser margins exist now than before and these margins will remain constant.

C. Lesser margins exist now than before and these margins will become even less.

D. Greater margins exist now than before but these margins will return to the original margins.

Proposed Answer: C S/t1iy1 *'

Justification: C is correct - FCTR curves the setpoints downward to avert the scram zone and the lowering FW temperature will continue to raise power.

AIB/D are incorrect - wrong direction Proposed references to be provided to applicants during examination: None Technical Reference(s): DDC1 E00135, Five Loop Power Flow Map K/A: 295001 Partial or AA2.02, Ability to determine and/or interpret the following as they Complete Loss of apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW Forced Core CIRCULATION: neutron monitoring Flow Circulation 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 2 K/A # 295002 295002 Importance Rating 3.4 3.4 Proposed Question: RO 016, SRO 029 Which one of the following describes the reason for manually tripping the main turbine if exhaust pressure is above 5 inches HgA with generator load less than 190 MWe?

Protects the...

A. main turbine from blading damage.

B. main turbine exhaust hoods from excessive temperatures.

C. main condenser from over-pressurization.

D. main condenser shells from excessive stress due to unbalanced steam loads.

Proposed Answer: A Justification: A is correct- per OP-31, D.1.0. BIC/D are incorrect - not the basis for Ithis operator action.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Ni-OP-31, D.1.0 K/A: 295002 Loss of AK3.02, Knowledge of the reasons for the following responses as Main Condenser they apply to LOSS OF MAIN CONDENSER VACUUM: Turbine Trip Vacuum 10 CFR Part 55 Content: 55.41(10) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank#

New X History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 2 K/A # 295008 295008 Importance Rating 2.8 2.8 Proposed Question: RO 017, SRO 030 The plant is operating at 100% power when a seismic event occurs, with the following:

"* The EPR setpoint controller rapidly fails low

"* An automatic reactor scram results

"* HPCI initiates as designed With no operator action, which one of the following describes a consequence of this event?

Failure of...

A. EC piping and supports if used for pressure control B. ERVs to automatically open if needed C. main condenser shells D. main turbine blading Proposed Answer: A Justification: A is correct - When EPR fails, TCVs'JTBVs open wide, causing a huge swell of RPV water level which results in FWP and turbine trip. If turbine trip doesn't process the scram first, MSIV closure

(<850 psig in RUN) will. In any event, RPV level remains high with no path to steam down (MSIVs closed). Continued EC operation with level >95" will result in water-hammer related damage to EC piping and supports (per N1-OP-13, D.17.0). B is incorrect - Although the ERVs may suffer the effects of high water level, this distractor is wrong because the potential ERV threat is being able to re-close them after allowing water through them. C/D are incorrect - MSIVs are closed; turbine/condenser is isolated and protected from flooding the main steam lines.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-13, D.17.0 K/A: 295008 High AK1.02, Knowledge of the operational implications of the following Reactor Water concepts as they apply to HIGH REACTOR WATER LEVEL:

Level Component errosion/damage 10 CFR Part 55 Content: 55.41(7) X 55.43 Question Source: Bank #

Modified Bank #

New X History

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 2 K/A # 295004 295004 Importance Rating 3.1 3.1 Proposed Question: RO 018, SRO 031 N1-SOP-1 8, Station Blackout, directs the stripping of certain non-vital 125 VDC loads.

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

Without AC power...

A. the battery boards will not automatically transfer to the standby static battery chargers; stripping loads is necessary for a manual dead-bus transfer.

B. loads will not automatically transfer between Battery Boards 11 and 12; stripping loads is necessary for a manual dead-bus transfer.

C. the static battery chargers will not carry the battery boards; stripping loads preserves DC battery power for vital functions.

D. MG Set 167 is unavailable; stripping loads preserves battery board power for the process computer.

Proposed Answer: C Justification: C correct- SBC's are lost; preserving DC power for Rx Instrumentation, EC controls, and starting EDGs is critical in a blackout. AIB incorrect - there are no auto-transfer features for either BBl 1/12 loads, or in-service/standby SBCs. D incorrect - although MG set 167 is unavailable, DC power is preserved for vital functions (described above), not for the PPC (an AC-powered system).

Proposed references to be provided to applicants during examination: None Technical Reference(s): NI-OP-47A, NI-SOP-18 K/A: 295004 Partial or AK2.01 - Knowledge of the interrelations between Partial or Total Total Loss of DC Loss of DC Pwr and the following: Battery charger Pwr 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 2 K/A # 295028 295028 Importance Rating 3.5 3.7 Proposed Question: RO 019, SRO 032 With Drywell Temperature and Reactor Pressure on the "BAD" side of Figure B, RPV SATURATION TEMPERATURE, of the EOP's, which one of the following describes the relationship between INDICATED RPV water level and ACTUAL RPV water level?

All instruments...

A. will read higher than ACTUAL B. will read less than ACTUAL C. except Fuel Zone will read higher than ACTUAL D. except Fuel Zone will read less than ACTUAL Proposed Answer: C Justification: C is correct- With the instrument reference legs in saturation, flashing could occur resulting in less dense reference legs so the variable legs would appear heavier (more water) making the instrument read higher than actual for all instruments except Fuel Zone as the Fuel Zone is density compensated and can be used when in the "Bad" zone as long as it is NOT flashing BID are incorrect- INDICATED level will be higher than ACTUAL A is incorrect - right response but does not effect Fuel Zone directly Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-EOP-4 K/A: 295028 HIGH EK1.01 Knowledge of the operational implications of the following DRYWELL concepts as they apply to HIGH DRYWELL TEMPERATURE:

TEMPERATURE Reactor Water level measurement 10 CFR Part 55 Content: 55.41(10) X 55.43 55.45 Question Source: -Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 1 K/A# 295013 295013 Importance Rating 2.9 3.2 Proposed Question: RO 020, SRO 006 A loss of coolant accident has occurred, with the following:

Torus Water temperature is 99 0 F

Torus Water level is 11 feet

Torus pressure is 15 psig

Containment Sprays cannot be established Which one of the following components may fail as a direct result of these conditions?

A. ERV tailpieces B. Drywell vent valves C. Drywell closure bolts D. Downcomers Proposed Answer: D Justification: D is correct- Torus pressure is above the Suppression Chamber Spray Initiation Limit which requires containment sprays to prevent cyclic condensation and fatigue failure of the downcomers. AIBIC are incorrect- not the component referenced in EOP bases as likely to fail due to chugging.

Proposed references to be provided to applicants during examination: EOP Flowcharts Technical Reference(s): N1-ODP-PRO-0305, pg 467 K/A: 295013 HIGH AK1.04. Knowledge of the operational implications of the following SUPPRESSION concepts as they apply to HIGH SUPPRESSION POOL POOL TEMPERATURE: Complete condensation TEMPERATURE 10 CFR Part 55 Content: 55.41(7) X F55.4 3 55.45 (5) X Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 1 K/A # 295013 295013 Importance Rating 3.6 3.7 Proposed Question: RO 021, SRO 007 The EOP's have been entered due to high Torus Temperature, with the following:

At 11:01, Torus Temperature is 85 0 F

At 11:05, Torus Temperature is 86 0 F

All four Containment Spray loops are available Which one of the following states the time that Torus Cooling shall be placed in service by?

A. 11:16 B. 11:20 C. 11:31 D. 11:35 Proposed Answer: A Justification: A is correct- per N1-EOP-1, Aft. 16, step 2.1, Torus cooling shall be in service within 15 minutes of Torus temperature equal to or greater than 85F BIC/D incorrect- uses greater than 85F and/or substitutes 30 min for 15 min Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-EOP-1, Att 16, step 2.1 K/A: 295013 HIGH AK2.01. Knowledge of the interrelations between HIGH SUPPRESSION SUPPRESSION POOL TEMPERATURE and the following:

POOL Suppression pool cooling TEMPERATURE 10 CFR Part 55 Content: 55.41(5,10) X 55.43 (l) X 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 1 K/A # 295016 295016 Importance Rating 3.0* 3.1 Proposed Question: RO 022, SRO 012 The plant was initially at 100% power when the following occurred:

"* The Control Room has been evacuated

"* The reactor has just been scrammed from outside the Control Room Which one of the following describes how the reactor is to be confirmed shutdown per NI-SOP-9.1, Control Room Evacuation?

A. Locally verify that both MG 131 and MG 141 are coasting down.

B. Observe that all control rods are fully inserted on the process computer at the TSC.

C. Observe the CONTROL RODS IN lights are energized on the Remote Shutdown Panels.

D. Locally verify the scram air header is depressurized and confirm all scram valves are open.

Proposed Answer: C Justification: C is correct - per SOP-9.1. A is incorrect - Not per SOP-9.1, and confirms status of RPS bus power but not actual rod position. B is incorrect - Not per SOP-9.1. D is incorrect - Not per SOP-9.1, and does not account for a 'hydraulic' ATWS.

Proposed references to be provided to applicants during examination: None Technical Reference(s): O1-OPS-001-212-1-01, NI-SOP-9.1 K/A: 295016 AA1.03, Ability to operate and/or monitor the following as they apply Control Room to CONTROL ROOM ABANDONMENT: RPIS Abandonment 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank # X12755 Modified Bank#

New History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 2 K/A# 295018 295018 Importance Rating 3.3 3.4 Proposed Question: RO 023, SRO 034 A loss of Service Water has occurred, with the following:

"* N1 -SOP-7 has been entered

" 1H1-4-2, R BUILDING SW PRESS/SERV W PUMP HDR PRESS LOW is in alarm

"* Neither Emergency Service Water Pump can be started Which one of the following sources should be used to supply cooling to the RBCLC heat exchangers?

A. Domestic Water B. Fire Water C. Circulating Water D. TBCLC Water Proposed Answer: B Justification: B is correct- per N1-SOP-7 Fire Water is to be used A/CID are incorrect- not referenced by SOP-7.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-SOP-7, A K/A: 295018 PARTIAL AA1.01. Ability to operate and/or monitor the following as they apply OR COMPLETE to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING LOSS OF WATER: Backup Systems COMPONENT COOLING WATER 10 CFR Part 55 Content: 55.41(5,7) X 55.43 55.45 Question Source: -Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 2 K/A # 295022 295022 Importance Rating 3.4 3.6 Proposed Question: RO 024, SRO 035 A plant heatup is in progress, with the following:

RPV pressure is 850 psig

The only available Control Rod Drive (CRD) pump has just tripped

Three withdrawn control rods have accumulator pressures reading 940 psig Which one of the following describes the required operator action?

A. Place the Reactor Mode Switch in SHUTDOWN.

B. Attempt to insert at least one withdrawn control rod one notch.

C. If a CRD pump cannot be started within 20 minutes, scram the reactor.

D. Commence a normal orderly shutdown in accordance with N1-OP-43C within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

Proposed Answer: A Justification: A is correct - action specified in ARP F3-1-2. B is incorrect - only done if attempting to confirm that at least CRD pump is running. C is incorrect - An option only when RPV pressure > 900 psig. D is incorrect - not per any related procedure.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-ARP-F3-1-2 K/A: 295022 Loss of AK2.07, Knowledge of the interrelations between LOSS OF CRD CRD Pumps PUMPS and the following: Reactor pressure (SCRAM assist): Plant Specific 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank # X Modified Bank #

New History Lasalle 2000 NRC Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 1 K/A # 295026 295026 Importance Rating 3.2 3.6 Proposed Question: RO 025, SRO 018 A loss of coolant accident and loss of Powerboard 102 have occurred, with the following:

Containment Sprays have been utilized to lower Torus pressure

The RPV has been depressurized using EC's and ERV's

Torus Water temperature is 200OF

Torus Level is 9.5 feet

Torus pressure is 6 psig

Drywell pressure 7 psig Which one of the following states the maximum Core Spray flow (Ibm/hrx 10E4) that may be used for RPV injection?

A. 175 B. 240 C. 350 D. 480 Proposed Answer: C Justification: C is correct - 9.5 feet- 4.5 feet = 5 feet x .433=2 psig + 6 psig=8psig, so 5 psig curve is used. With two sub-systems available the max flow is 240 AIB/D are incorrect different combinations of one subsystem values and/or using the 10 psig curve which would be chosen if candidate forgets to subtract 4.5 feet from 9.5 feet.

Proposed references to be provided to applicants during examination: NI-EOP-4 Technical Reference(s): N1-ODP-PRO-0305, pg 48 and N1-EOP-2, Detail NI K/A: 295026 EK2.03: Knowledge of the interrelations between SUPPRESSION SUPPRESSION POOL HIGH WATER TEMPERATURE and the following:

POOL HIGH Suppression chamber pressure: Mark-I&ll WATER TEMPERATURE 10 CFR Part 55 Content: 55.41(5) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: Needs SRO Tech Verify number of subsystems/which curve to use

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 1 K/A # 295003 295003 Importance Rating 3.7 3.8 Proposed Question: RO 026, SRO 019 The plant is in a Station Blackout. The 115kV line from the South Oswego Substation has just been restored Which one of the following describes the response of breaker RIO and Powerboards 11 and 12, given the above conditions?

Breaker RI 0 PB 11 and 12 A. Will automatically close Will re-energize automatically B. Will automatically close Must be re-energized manually C. Must be manually closed Will re-energize automatically D. Must be manually closed Must be re-energized manually Proposed Answer: A Justification: A correct - auto re-closure feature and bus re-energizes B incorrect bus re-energizes C incorrect - auto re-closure feature D incorrect auto re-closure feature and bus re-energizes Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-262-1-01 N1-OP-30 H.3.1 K/A: 295003 Partial or AA1.01 - ability to operate and/or monitor thew following as they Complete Loss of apply to PARTIAL OR COMPLETE LOSS OF AC POWER: A.C.

AC Power electrical distribution system 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank # X Modified Bank#

New History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 2 K/A # 295028 295028 Importance Rating 3.6 3.8 Proposed Question: RO 027, SRO 037 The plant is operating at 100% power, with the following:

Drywell Cooler 16 is tagged out

The normal feeder breaker to Powerboard 17A malfunctions and trips open

"* Powerboard 17A has been checked by Electrical Maintenance and is fault free

"* Drywell temperature is rising Which one of the following states the maximum number of Drywell Coolers that can be placed in service?

A. 2 B. 3 C. 4 D. 5 Proposed Answer: D Justification: D is correct - Powerboard 17A can be cross tied and the 3 drywell coolers lost restarted for a total of 5 A is incorrect -would be the number if 17A could not be reeenergized B incorrect - would be the number if 17A could not be energized and if drywell cooler 16 was believed powered from 17A C incorrect - powerboard 17A can be reenergized and drywell coolers restarted Proposed references to be provided to applicants during examination: None Technical Reference(s): Ni -ARP-A5-3-6 K/A: 295028 HIGH EK3.04: Knowledge of the reasons for the following responses as DRYWELL they apply to HIGH DRYWELL TEMPERATURE: Increased drywell TEMPERATURE cooling 10 CFR Part 55 Content: 55.41(5,7,10) X 55.43 (5) 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 1 K/A# 295016 295016 Importance Rating 4.2 4.3*

Proposed Question: RO 028, SRO 011 The plant is operating at 100% power, with the following:

A fire occurs filling the control room with heavy smoke

Control Room Abandonment is directed

All control room actions were completed While transiting to RSP 12 to establish control of Reactor Water Level, which one of the following describes the initial Reactor Water Level response?

Reactor Water Level will initially...

A. Rise to the Main Steam Lines and higher B. Rise to >95 inches and then stabilize C. Lower to <53 inches then recover on HPCI D. Lower to triple low and reflood on ADS/Core Spray Proposed Answer: C Justification: C correct -before exiting the control room a scram and trip of the 13 FW pump will occur, this will lower water level to the HPCI initiation point A/B incorrect-on the scram from 100% power reactor water level will drop D incorrect - Feedwater is available from FW pumps 11 and 12, but if assumed not available would occur Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-SOP-9.1 K/A: 295016 Control AA2.02 - Ability to determine and interpret the following as they apply Room to Control Room Abandonment: Reactor water level Abandonment 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 2 K/A # 295033 295033 Importance Rating 3.9 4.2*

Proposed Question: RO 029, SRO 038 The plant is operating at 100% power, with the following:

1H11-4-8, AREA RADIATION MONITORS, alarms

REFUEL BRIDGE (HIGH RANGE) is in alarm

RX BLDG-EAST WALL, EL. 340 is in alarm

Affected areas have radiation levels between 8,500 and 10,000 mr/hr

NO area temperature or sump level alarms exist in the Reactor Building Which one of the following states the action required and the areas that should be evacuated?

ACTION EVACUATION AREA A. Perform RPV Blowdown (EOP-8) Refuel Floor (RB 340)

B. Shutdown the Reactor (OP-43) Refuel Floor (RB 340)

C. Perform RPV Blowdown (EOP-8) Reactor Building (All)

D. Shutdown the Reactor (OP-43) Reactor Building (All)

Proposed Answer: B Justification: B is correct - No indication of primary system discharging and the only affected area is the refuel floor A/C are incorrect- no primary system discharging D is incorrect- only the refuel floor is affected Proposed references to be provided to applicants during examination: EOP Flowcharts Technical Reference(s): NI-EOP-5, H1-4-8 K/A: 295033 High EK1.02 - Knowledge of the operational applications of the following Secondary concepts as they apply to the High Secondary Containment Area Containment Radiation Levels: Personnel protection Area Radiation Levels 10 CFR Part 55 Content: 55.41(12) X 55.43 (4) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 1 Group # 2 K/A # 295029 Importance Rating 3.9 Proposed Question: RO 030 EOP-4 has been entered due to Torus water level, with the following:

Torus water level is 11.3 feet

80-40 CONT SPRAY BYPASS BV 111 is stuck OPEN Which one of the following loops of Containment Spray should be used to lower Torus Level?

A. 111 B. 112 C. 121 D. 122 Proposed Answer: A Justification: A correct - per EOP-1 attachment 15 only loop 111 can be used if BV 80-40 cannot be closed BIClD incorrect- all these loops require BV 80-40 be closed Proposed references to be provided to applicants during examination: N1-EOP-1, Attachment 15 Technical Reference(s): N1-EOP-1, Attachment 15, step 2.2 K/A: 295029 High Generic 2.1.30 Ability to locate and operate components/including Suppression Pool local controls Water Level 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank#

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 1 K/A # 295038 295038 Importance Rating 2.5 3.4 Proposed Question: RO 031, SRO 025 A loss of coolant accident has occurred, with the following:

Primary containment water level is 30 feet

Torus pressure is 38 psig and rising

H2 and 02 concentrations are below their thresholds for PC venting

Estimates indicate that any PC venting would exceed General Emergency level dose rates Which one of the following identifies the required action and the proper vent path?

A. Vent the Drywell now using the vent and purge fan.

B. Do NOT vent until directed by EOP-4.2 and then use the vent and purge fan.

C. Vent the Torus now using the Post-LOCA Vent path.

D. Do NOT vent until directed by EOP-4.2 and then use the Post-LOCA Vent path.

Proposed Answer: A Justification: A correct - Per EOP-4, step PCP-8 and Figure D. Ok to exceed release rates; PC level too high to vent from torus. PC pressure too high to use Post-LOCA path thru RBEVS. BID incorrect-With the PCPL about to be reached, PC gas concentration is irrelevant; EOP-4 directs us to vent now. C incorrect - PC water level too high (>27') to use torus.

Proposed references to be provided to applicants during examination: EOP flowcharts Technical Reference(s): Ni -EOP-4, NI-EOP-4.1 K/A: 295038 High Off- 2.3.9 - Knowledge of the process for performing a containment purge site Release Rate 10 CFR Part 55 Content: 55.41(10,12 X 55.43 (4,5) X 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 2 2 K/A # 600000 600000 Importance Rating 3.0 3.1 Proposed Question: RO 032, SRO 043 The plant is at power, with the following:

A fire develops in Remote Shutdown Panel 11

The Fire Brigade leader reports the fire appears out but there is heavy smoke Which one of the following Smoke Removal Zones should be purged?

A. Zone 1 B. Zone 2 C. Zone 3 D. Zone 4 Proposed Answer: B Justification: B correct -the RSP-1 1 is located on TB 250 in zone 2 AMCID are incorrect per the OP-21 F zone map Proposed references to be provided to applicants during examination: N1-OP-21F, entire Aft. 5 Technical Reference(s): N1-OP-21F K/A: 600000 Plant Fire AA1.05 - Ability to operate and/or monitor the following as they apply On Site to Plant Fire On Site: Plant and control room ventilation systems 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: -Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 1 Group # 3 K/A # 295021 Importance Rating 3.3 Proposed Question: RO 033 The plant is in COLD SHUTDOWN, with the following:

Shutdown Cooling (SDC) is in service

No Recirc pumps are available Which one of the following describes how high RPV level should be raised and the reason for raising level?

RAISE LEVEL TO ABOVE THE... REASON A. Recirc pump return nozzles Promote Natural Circulation B. Main Steam Line nozzles Promote Natural Circulation C. Recirc pump return nozzles Prevent SDC flow short cycling D. Main Steam Line nozzles Prevent SDC flow short cycling Proposed Answer: B Justification: B is correct - Prescribed by N1-OP-4 for the condition of no forced circulation. AIC incorrect - wrong RPV nozzle used as reference height in OP D incorrect - right nozzle but reason is for maintaining SDC min flow valves closed with no recirc pumps Proposed references to be provided to applicants during examination: None Technical Reference(s): Ni-OP-04, D.8 K/A: 295021 LOSS AK3.01: Knowledge of the reasons for the following responses as OF SHUTDOWN they apply to LOSS OF SHUTDOWN COOLING: Raising reactor COOLING water level 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 3 1 K/A # 295023 295023 Importance Rating 3.3 3.4 Proposed Question: R0034, SRO 016 Fuel movements are in progress, with the following:

0 SRM Channel 14 fails downscale and is declared inoperable 0 Only SRM Channels 11 and 13 are operable 0 The next fuel movement is the removal of a bundle from core quadrant 12 Which one of the following describes the impact of these conditions on the next fuel movement, and the reason for that impact?

The next fuel movement .....

A. must NOT proceed because there is no operable SRM in quadrant 12.

B. must NOT proceed because the minimum number of operable SRMs requirement is not satisfied.

C. may proceed because the minimum number of operable SRMs requirement is satisfied.

D. may proceed because there is still an operable SRM in the quadrant adjacent to quadrant 12.

Proposed Answer: A Justification: A is correct- There is no operable SRM in this quadrant. B is incorrect - The minimum required SRMs are operable, but SRM 12 is not. C is incorrect - Although the minimum SRMs operable requirement is met, there is no operable SRM in this quadrant. D is incorrect - SRM 12 must be operable to proceed.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-34 Precaution 14, O1-OPS-001-234-1-01 K/A: 295023 Refueling AA1.06, Ability to operate and/or monitor the following as they apply IAccidents to REFUELING ACCIDENTS: Neutron monitoring 10 CFR Part 55 Content: 55.41(7,10) X 55.43 (2,7) X 55.45 Question Source: Bank #

Modified Bank# X12776 New History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2

Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Group # 3 2 K/A # 295032 295032 Importance Rating 3.5 3.8 Proposed Question: RO035, SRO 041 An unisolable steam leak has developed in the Reactor Building, with the following:

"* Reactor Building Ventilation Exhaust Radiation is 7 mr/hr

"* K3-3-4, CLEAN-UP SYSTEM LEAK AREA T HI, alarms

"* RWCU Pump Area temperature is in alarm

"* RWCU Heat Exchanger Room temperature is in alarm

"* Local temperatures are consistent with alarms Which one of the following describes the required action and the reason for that action?

ACTION REASON A. Commence a normal plant shutdown Minimize effects of Level instrument errors B. Scram and blowdown the RPV Allow RPV Flooding with Core Spray C. Commence a normal plant shutdown Minimize the threat to Secondary Containment D. Scram and blowdown the RPV Reduce driving head of steam leak Proposed Answer: D Justification: D correct - With two areas above max safe Blowdown is required to reduce the leakrate A/C incorrect - action if only one area above max safe B incorrect - right action but wrong reason Proposed references to be provided to applicants during examination: N1-EOP-5 Technical Reference(s): N1-EOP-5, NI-ODP-PRO-0305 K/A: 295034 HIGH EK3.01: Knowledge of the reasons for the following responses as SECONDARY they apply to HIGH SECONDARY CONTAINMENT AREA CONTAINMENT TEMPERATURE: Emergency/normal depressurization AREA TEMPERATURE 10 CFR Part 55 Content: 55.41(10) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History

Examination Outline Cross-reference: Level RO Tier # I Group # 2 K/A # 295036 Importance Rating 3.3 Proposed Question: RO 036 The plant is operating at 100% power, with the following:

H2-2-1, R BLDG FL DR SUMPS 11-16 AREA WTR LVL LEVEL HIGH, alarms

Computer printout confirms the alarm Which one of the following describes the required operator action?

A. Verify affected sump pump is running and direct Radwaste to determine cause of high level.

B. If affected sump rate of rise exceeds 4.2 gpm, manually scram and enter N1-SOP-1.

C. Direct Radwaste to pump down affected sump with temporary pump per N1-OP-10.

D. Enter N1-EOP-5, Secondary Containment Control, and determine cause of high level.

Proposed Answer: D Justification: D correct - per ARP-H2-2-1 actions. AIB/C incorrect - various actions from similar ARPs involving RB Eqpt Drains and Drywell sumps.

Proposed references to be provided to applicants during examination: None Technical Reference(s): ARP-H2-2-1 K/A: 295036 2.4.50 - Ability to verify system alarm setpoints and operate controls Secondary identified in the alarm response manual Containment High Sump/Area Water Level 10 CFR Part 55 Content: 55.41(10) X 55.43 55.45 Question Source: -Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 2 K/A# 201001 201001 Importance Rating 3.0 2.9 Proposed Question: RO 037, SRO 067 The plant is operating at 100% power, with the following:

F3-3-2, CRD CONTROL AIR PRESSURE HI-LO, alarms

CRD air header pressure is reading 85 psig on the F Panel Which one of the following describes the effect on the Control Rod Drive System?

A. Rods will begin to drift in.

B. CRD high temperatures will result.

C. The scram air header relief valve will soon lift causing all rods to scram.

D. Opening of the Scram Inlet and Outlet valves will be delayed on a scram.

Proposed Answer: D Justification: D correct - high pressure may result in excessive scram times. A is incorrect - No connection exists between high air pressure and the conditions that may result in rod drift. B is incorrect- The scram pilot regulator does not supply the CRD FCV, which supplies cooling flow.

C is incorrect -There is no such relief valve.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-5, P&L 2.0, N1-ARP-F3-3-2 K/A: 201001 K6.03. Knowledge of the effect that a loss or malfunction of the CONTROL ROD following will have on the CONTROL ROD DRIVE HYDRAULIC DRIVE System: Plant Air Systems HYDRAULIC 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 1 K/A # 202002 Importance Rating 2.6 Proposed Question: RO 038 The plant is operating at 100% power, with the following:

5 Recirc Loops are in service

The OPEN limit switch for Recirc Pump 14 discharge valve fails (contact no longer closed)

Which one of the following describes the Recirc Pump 14 response?

Recirc Pump 14...

A. will trip.

B. speed will lower to 20%.

C. scoop tube will lock up.

D. will remain running at its current speed.

Proposed Answer: D Justification: D correct - The speed interlock is off of full-close valve position.

A incorrect - Suction valve closure is a pump trip, not discharge valve. B incorrect - Speed is limited only when the valve is fully closed. C incorrect - Discharge valve position has no relationship to scoop tube lock up.

Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-202-1-01, NI-OP-01 K/A: 202002 K5.02. Knowledge of the operational implications of the following RECIRCULA- concepts as they apply to RECIRCULATION FLOW CONTROL TION FLOW SYSTEM: Feedback signals CONTROL SYSTEM 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A # 206000 206000 Importance Rating 3.5 3.8*

Proposed Question: RO 039, SRO 044 A reactor scram from rated power has occurred, with the following:

Feedwater Pump 12 is injecting to the RPV

A fault causes the loss of Battery Board 12.

Which one of the following describes the effect of this power loss on RPV injection, and the operator action necessary to compensate?

Feedwater Pump 12 will...

A. trip immediately and must be restarted.

B. trip immediately and will auto-start after manually transferring control power to BB 11.

C. not trip and must be manually tripped if RPV level rises to the high level trip setpoint.

D. not trip, but FCV 12 will lock up, Feedwater Pump 11 must be started to control RPV level.

Proposed Answer: C Justification: C is correct - Pump tripping power is lost. RPS control power remains available from the AC inverter source. A is incorrect - Pump tripping power is lost. B is incorrect - Pump tripping power is lost.

D is incorrect - FCV 12 does not lock up on this power loss.

Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-259-1-01 K/A: 206000 High A2.05, Ability to (a) predict the impacts of the following on the HIGH Pressure Coolant PRESSURE COOLANT INJECTION SYSTEM; and (b) based on Injection System those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: D.C.

failures: BWR-2,3,4 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A # 206000 206000 Importance Rating 3.9 3.8 Proposed Question: RO 040, SRO 045 The plant is operating at 100% power with Feedwater Pumps 12 and 13 in service, with the following:

HPCI CH 12 light on the 'E' Panel energizes

Plant remains at 100% power

All systems operate as designed Which one of the following identifies the status of HPCI and identifies the other alarms or indications that the operator will also report?

HPCI STATUS OTHER ALARMS I INDICATIONS TO REPORT A. Initiated HPCI CH 11 light on the 'E' Panel is also energized B. Initiated F4-4-1, HPCI MODE AUTO INITIATE, has alarmed C. NOT initiated Both Bypass Flow Control Valves 11 and 12 indicate closed D. NOT initiated FW Pump 12 flow is reading 1.9 x 106 Ibm/hr Proposed Answer: D Justification: D is correct - This indicates a shift of the 12 FW FCV to HPCI Mode due to high pump flow. However, HPCI has not initiated. AIB are incorrect - The plant is at 100% power and HPCI initiation signals are scram signals. Since HPCI has not initiated, no reason for HPCI CH 11 light to be energized. No reason for F4-4-1 to be alarming. C is incorrect - Since HPCI has not initiated, no reason for either Bypass Flow Control Valve to indicate closed.

Proposed references to be provided to applicants during examination: None Technical Reference(s): ARP-F4-4-1, 01-OPS-001-259-1-01 K/A: 206000 A3.07, Ability to monitor automatic operations of the HIGH High Pressure PRESSURE COOLANT INJECTION SYSTEM including: Lights and Coolant Injection alarms: BWR-2,3,4 System 10 CFR Part 55 Content: 55.41(7) X 55.43 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A # 209001 209001 Importance Rating 3.8 3.9 Proposed Question: RO 041,SRO 046 A loss of coolant and impaired injection has occurred, with the following:

RPV level drops and stabilizes at -20 inches

RPV pressure drops and stabilizes at 750 psig

Drywell pressure rises and stabilizes at 11 psig ONE minute later, with no operator action, which one of the following describes the Core Spray 111 pump flowrate and the status of the Core Spray 111 pump min flow relief valve?

CORE SPRAY 111 PUMP FLOW PUMP MIN FLOW RELIEF A. 500 GPM CLOSED B. 500 GPM OPEN C. 3400 GPM CLOSED D. 3400 GPM OPEN Proposed Answer: B Justification: B correct- 111 CS pump would be running but ADS would not yet initiated so the pump would be running on min flow AICID are incorrect - various configurations of incorrect BV position (CLOSED) and/or flowrate (3400=full flow)

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-2, section B, page 4 K/A: 209001 LOW A1.03. Ability to predict and/or monitor changes in parameters PRESSURE associated with operating the LOW PRESSURE CORE SPRAY CORE SPRAY SYSTEM controls including: Reactor water level SYSTEM 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group# 1 1 K/A # 209001 209001 Importance Rating 3.5 3.6 Proposed Question: RO 042, SRO 047 The plant is in Cold Shutdown, with the following:

Torus water level is low

Core Spray pump 111 is inoperable and clearance tagged out of service

Core Spray pumps 112, 121 and 122 are operable and in standby Which one of the following identifies the Core Spray System(s) that may be used to add water to the Torus, per plant procedures and/or Technical Specifications?

A. Either Core Spray System may be used B. Only Core Spray System 11 may be used C. Only Core Spray System 12 may be used D. Neither Core Spray System may be used Proposed Answer: B Justification: B is correct - Core Spray 11 keep-fill is available for Torus fill as both 121 and 122 are operable (i.e., use keep-fill from the already inop'd System 11, leaving System 12 fully operable). AIC are incorrect Core Spray 12 is not available for Torus fill due to Core Spray 111 being INOP. D is incorrect - Core Spray 11 is available (uses Keep Fill, not Core Spray pumps).

Proposed references to be provided to applicants during examination: None Technical Reference(s): NI-OP-02, H.1 K/A: 209001 LOW A4.09. Ability to manually operate and/or monitor in the control room:

PRESSURE Suppression pool level CORE SPRAY SYSTEM 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group# 1 K/A # 212000 Importance Rating 2.7 Proposed Question: RO 043 A failure to scram has occurred, with the following:

Reactor Water level is being controlled -41 inches to -84 inches

NI-EOP-3.1, Alternate Rod Insertions is in progress

RPS logic jumper 6, RELAY 11 K8 TO RELAY 11 K52B fails to work Which one of the following describes the Reactor Protection System (RPS) response when RPS reset is attempted?

A. A full scram remains in effect B. Only a half scram on 11 RPS remains in effect C. Only a half scram on 12 RPS remains in effect D. A full scram reset occurs Proposed Answer: B Justification: B is correct - an RPV level low scram signal would still be detected by RPS 11 A/CID are incorrect - RPS 12 will reset Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-EOP-3.1, Print C-1 9859-C, Sheet 4 K/A: 212000 A1.09: Ability to predict and/or monitor changes in parameters REACTOR associated with operating the REACTOR PROTECTION SYSTEM PROTECTION controls including: Individual relay status: Plant-Specific SYSTEM 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A# 215004 215004 Importance Rating 3.2 3.2 Proposed Question: RO 044, SRO 049 A plant heatup is in progress with the following:

All IRMs are reading mid-scale on Range 8

SRM 12 detector is being withdrawn to maintain SRM counts on scale Then, the following occurs:

A short circuit causes the SRM 12 detector to fully retract from the core

SRM 12 counts stabilize at 40 CPS Which one of the following describes the impact of this malfunction on the ability to continue the plant heatup?

A. SRM 12 must first be bypassed to clear the rod block.

B. SRM 12 must first be repaired and declared operable.

C. Control rod withdrawals may continue provided SRM 12 remains above its downscale.

D. Control rod withdrawals may continue provided all IRMs remain above the downscales on Range 8.

Proposed Answer: D Justification: D is correct - the rod block from SRM not full in below 100 CPS is bypassed with IRMs on range 8 or above. A is incorrect - a rod block does not exist with IRMs on range 8. SRM bypass will be necessary only if power is allowed to lower below range 8. B is incorrect - T.S. only requires 3 SRM channels for the heatup. C is incorrect-Although SRM 12 is < 100 CPS, the rod block does not exist with IRMs on Range 8.

Proposed references to be provided to applicants during examination: None Technical Reference(s): 01 -OPS-001-215-1-02 K/A: 215004 Source K4.06, Knowledge of SOURCE RANGE MONITOR (SRM) SYSTEM design Range Monitor feature(s) and/or interlocks which provide for the following: IRM/SRM interlock (SRM) System 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A # 215004 215004 Importance Rating 2.8 2.8 Proposed Question: RO 045, SRO 050 A reactor startup is in progress, with the following:

Reactor power is 5.5 x 10 4 cps on all SRMs

SRM 11 detector becomes stuck and will not withdraw Which one of the following describes when SRM 11 must be bypassed in order for control rod withdrawals to continue?

To continue rod withdrawal, SRM 11 must be bypassed when...

A. SRM 11 reaches 1 x 105 cps B. any SRM reaches 1 x 105 cps C. IRM 11, 12, 13, or 14 is on Range 7 D. all IRMs are on Range 7 Proposed Answer: A Justification: A is correct - any single SRM reaching 1E5 cps will impose a rod withdrawal block, but bypassing SRM 11, specifically, will remove the rod block if only SR2 11 is at 1E5 cps. B is incorrect SRIs at 1E5 cps will cause Although any of the other signal a rod block, bypassing SRM 11 will affect only the SRMV 11 upscale to RMVCS. ClD are incorrect - plausible only ifcandidate is confused about the feature that auto-bypasses the SRMV upscale rod block when IRIVs are on Range 7.

Proposed references to be provided to applicants during examination: None Technical Reference(s): O1 -OPS-001-215-1-02 K/A: 215004 SOURCE K5.03. Knowledge of the operational implications of the following RANGE concepts as they apply to SOURCE RANGE MONITOR (SRM)

MONITOR (SRM) SYSTEM: Changing detector position SYSTEM 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: -Bank #

Modified Bank #

New X History

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A# 215005 215005 Importance Rating 2.6 2.8 Proposed Question: RO 046, SRO 051 The plant is operating at 100% power, with the following:

"* The power supply leads to an LPRM detector associated with APRM 15 require repair

"* You have been assigned to work on the clearance Which of the following identifies the electrical bus you will have to tag out for this work?

A. 24 VDC Bus 11 B. 24 VDC Bus 12 C. 120 VAC RPS Bus 11 D. 120 VAC RPS Bus 12 Proposed Answer: D Justification: D correct - The LPRM detectors are powered from RPS 120 VAC, APRM 15 is associated with channel 12 A/B are incorrect- LPRM detectors are powered from RPS 120 VAC C is incorrect-APRM 15 is associated with channel 12 Proposed references to be provided to applicants during examination: None Technical Reference(s): N1 -OP-38C/sect C, 01 -OPS-001 -215-1-02/Table 25 K/A: 215005 K2.02. Knowledge of electrical power supplies to the following:

AVERAGE APRM channels POWER RANGE MONITOR/LOCA L POWER RANGE MONITOR

_ SYSTEM 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A# 215005 215005 Importance Rating 3.6 3.6 Proposed Question: RO 047, SRO 052 A plant startup is in progress, with the following:

The Mode Switch is in RUN

Reactor Power lowers due to steam load changes and feedwater input

APRM's 11, 12 and 16 have Downscale alarms Which one of the following describes the ability to individually drive control rods?

Control rods may be individually driven...

A. In both directions B. Inward direction only C. Outward direction only D. In neither direction.

Proposed Answer: B Justification: B is correct - with APRM downscale and Mode switch not in STARTUP/REFUEL a withdraw block is applied AICID are incorrect

-other possible effects but withdraw block will occur for given conditions Proposed references to be provided to applicants during examination: None Technical Reference(s): O1-OPS-001-215-1-02/Table 29 K/A: 215005 A1.03: Ability to predict and/or monitor changes in parameters AVERAGE associated with operating the AVERAGE POWER RANGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls MONITOR/LOCA including: Control rod block status.

L POWER RANGE MONITOR SYSTEM 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History

Examination Outline Cross-reference: Level RO Tier # 2 Group # 1 K/A # 216000 Importance Rating 3.2 Proposed Question: RO 048 The plant is operating in Four-loop mode, with the following:

Recirc pumps 11, 12, 13 and 15 are operating

One Channel 11 RPV level instrument is OOS for testing and is inputting a LOW signal

One Channel 11 RPV pressure instrument fails high Which one of the following describes the effect on the Reactor Recirculation System pumps?

All Recirc pumps...

A. will remain running at current flowrates B. will runback to minimum flow C. except 12 will trip D. will trip Proposed Answer: A Justification: A is correct - Two pressure or two level inputs are required on either channel 11 or 12 to trip all recirc pumps B is incorrect - runback would not occur, but is a recirc response to other conditions C incorrect - if trip is initiated all pumps would trip. D incorrect - two signal of one parameter required to initiate a trip Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-40, section B, page 9 K/A: 216000 K3.30: Knowledge of the effect that a loss or malfunction of the NUCLEAR NUCLEAR BOILER Instrumentation will have on following:

BOILER Recirculation system Instrumentation 10 CFR Part 55 Content: 55.41(7) X 55._43-:ý 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 1 K/A # 216000 Importance Rating 2.4.10 Proposed Question: RO 049 A Failure to Scram has occurred, with the following:

Reactor Power is 4%

Reactor Water Level is being controlled -20 to -30 inches 1F4-2-1, CORE LEVEL TORUS TEMP. MONITOR SYS. 12 TROUBLE alarms

H482, FZWL12 DRYWL TEMP is in alarm

Drywell Temperature is 220°F

A disparity exists between Fuel Zone instruments

A similar disparity exists between Lo-Lo-Lo instruments Which one of the following RPV Level instruments should be used to control RPV Level?

A. Lo-Lo-Lo Ch 11 B. Lo-Lo-Lo Ch 12 C. Fuel Zone Ch 11 D. Fuel Zone Ch 12 Proposed Answer: C Justification: C correct - Fuel Zone range -240 to +110, Ch 11 Density Compensation is working AIB incorrect - High Drywell temperature makes the minimum usable indication is -10 inches D incorrect- Fuel Zone Ch 12 is inoperable due to loss of density compensation Proposed references to be provided to applicants during examination: ARP F4-2-1, N1-EOP-3 Technical Reference(s): ARP-F4-2-1, Ni1-EOP-3 K/A: 216000 Nuclear 2.4.10 - Knowledge of annunciator response procedures Boiler Instrumentation 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A# 218000 218000 Importance Rating 3.7 3.8 Proposed Question: RO 050, SRO 053 The plant is operating at 100% power, with the following:

PRESS SAFETY / RELIEF VALVES FLOW Alarm is alarming

ERV 121 Acoustic Monitor reads high

Acoustic monitor channels for all ERVs have similar noise levels

Downstream tailpiece temperature for ERV 121 reads 138 OF Torus water temperature is 74 OF and steady Which one of the following identifies the appropriate operator response to these conditions?

A. Repair the ERV 121 acoustic monitor during the next cold shutdown.

B. Commence an emergency power reduction per N1-OP-43B.

C. Cycle the control switch for Emergency Relief Valve 121.

D. Pull the control power fuses for ERV 121 in "F" Panel.

Proposed Answer: A Justification: A correct - The remaining indicators confirm an acoustic monitor failure B incorrect - SRV is not open C incorrect - SRV is not open D incorrect - SRV is not open Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-218-1-01 K/A: 218000 A3.03: Ability to monitor automatic operations of the AUTOMATIC AUTOMATIC DEPRESSURIZATION SYSTEM including: ADS valve acoustical DEPRESSURIZA monitor noise: Plant-Specific.

TION SYSTEM 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank # X12450 Modified Bank #

New History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 1 K/A # 218000 Importance Rating 3.8 Proposed Question: RO 051 A loss of coolant accident has occurred, with the following:

The Powerboard 102 supply to ADS Channel 11 is lost.

Valid and sustained ADS initiation signals are received.

Which one of the following identifies the status of the Primary and Backup ADS valves 116 seconds after the initiation of the ADS signals?

PRIMARY VALVES BACKUP VALVES A. Closed Closed B. Closed Open C. Open Closed D. Open Open Proposed Answer: C Justification: C is correct - ADS Channel 12 is unaffected by the PB 102 loss; all of the ADS logic necessary to actuate ERVs will function normally; i.e.,

primary valves will open at Time = 111 sec and prevent the opening of the secondary valves. AIB/D are incorrect - plausible if examinee confuses relationship between PB 102 120 VAC power and the RPS ADS logic and is distracted by our choice of 116 sec in the stem.

Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-218-1-01 K/A: 218000 K5.01, Knowledge of the operational implications of the following Automatic concepts as they apply to the AUTOMATIC DEPRESSURIZATION Depressurization SYSTEM: ADS logic operation.

System 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: -Bank #

Modified Bank# X New History 2000 NMPI ROISRO

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group# 1 1 K/A # 223001 223001 Importance Rating 3.0 3.1 Proposed Question: RO 052, SRO 054 A loss of Instrument Air to the CAM and H 2-0 2 Monitors has occurred.

Which one of the following describes the response of the sample stream isolation valves, and the required alternate method for monitoring the primary containment atmosphere?

RESPONSE OF SAMPLE STREAM IVs ALTERNATE MONITORING METHOD A. Fail open Channel 12 H 2 -0 2 local indication B. Fail open Channel 12 H2-0 2 Control Room indication C. Fail closed Drywell PASS sample D. Fail closed Torus PASS sample

'Yýp uf_&Tt, A) 'b,-Lc4 Llezý Z-11 Proposed Answer: A - I It i.

Aý&L I Justification: A is correct - Per SOP-6. B is incorrect - cal gas IVs fail closed; therefore no control room indication is valid. COD are incorrect sample IVs fail open, and although SOP-6 suggests a PASS (if desired), it still requires monitoring Ch. 12 locally.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-SOP-6 K/A: 223001 Primary K1.10, Knowledge of the physical connections and/or cause- effect Containment relationships between PRIMARY CONTAINMENT SYSTEM AND System and AUXILIARIES and the following: Plant air systems Auxiliaries 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank# X12473 Modified Bank #

New History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 1 K/A # 223001 Importance Rating Proposed Question: RO 053 The plant is operating at 100% power, with the following:

RBCLC pump 12 is operating

RBCLC Temperature Controller 70-23B is operating in automatic

Controller 70-23B's output signal rapidly fails high (maximum output signal)

Without operator action, which one of the following control room annunciators will result?

A. K3-4-4, CLEANUP SYSTEM ISOLATION B. F2-1-1, REACT RECIRC PUMP - MOTOR 11 C. K1-4-6, TORUS - DW VAC RELIEF CHECK VALVE OPEN D. H1-2-1, REACTOR BLDG COOL PUMP 12 TRIP-V-SUCT Proposed Answer: C Justification: C is correct - Controller's max output signal acts to fully close RBCLC HX bypass valve and fully open the RBCLC HX Service Water outlet valve; results in maximum RBCLC cooling (minimum supply temperature). Sudden significant drop in DW cooler RBCLC temperature may cause rapid DW atmosphere cooldown, DW pressure drop and create the d/p (DW-Torus) necessary to momentarily open the Torus-DW Vacuum Breakers. AIB are incorrect - plausible if examinee believes that the max output signal results in minimum cooling (higher RBCLC supply temp). D is incorrect - RBCLC pump suction pressure would not lower if system temp lowered.

Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-223-1-02, N1-ARP-K1-4-6, H1-4-1 K/A: 223001 Primary K4.05, Knowledge of Primary CTMT and Auxiliaries design feature(s) and or CTMT and interlock(s) which provide for the following: Maintains proper suppression Auxiliaries pool to drywell differential pressure 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3

Examination Outline Cross-reference: Level RO Tier # 2 Group # 1 K/A # 223002 Importance Rating 3.6 Proposed Question: RO 054 The plant is operating at 100% power with the following:

Battery Board 11 is clearance tagged

Transferrable loads have been transferred to Battery Board 12 Then, the following occurs:

Loss of Powerboard 161B

Failure to Scram

Liquid Poison is initiated

RPV water level has been intentionally lowered to -41 inches Which one of the following describes the current status of the Reactor Water Cleanup System (RWCU) and Reactor Sample Return IVs (63-04 and 63-05)?

RWCU REACTOR SAMPLE RETURNS A. Isolated Closed B. Isolated Open C. Not isolated Closed D. Not isolated Open A

Proposed Answer:,WB Chal~e. C'0v'eJk c6vAt4&y TD A.QYCt"4~6Mam-sr1 Justification: B is correct - Although loss of PB 161 B prevents closure of RWCU Inboard IV(33-02R), the Outboard IV is a DC valve powered from Battery Board 12. The RWCU Return IV (33-01R) still has power from PB 171 B. When Liquid Poison is initiated, RWCU will successfully isolate. The Reactor Sample Return IVs are DC valves powered from Battery Board 11 and are not transferrable loads. When an RPV Lo-Lo level occurs (+5 inches) for the failure to scram actions (lowering level), these valves will not auto-close. A/CID are incorrect - for the reasons described above.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Ni1-SOP-17, N1-OP-47A, 01-OPS-001-204-1-01 K/A: 223002 Primary A4.01, Ability to manually operate and/or monitor in the control room: Valve closures.

Containment Isolation/Nuclear Steam Supply Shutoff 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 13 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A # 239002 239002 Importance Rating 3.7 3.8 Proposed Question: RO 055, SRO 059 The plant is operating at 100% power, with the following:

Reactor Water level is 72 inches

Then, an ERV inadvertently opens With no operator action, which one of the following describes Reactor Water level response?

Reactor Water level will...

A. Rise and remain above 72 inches B. Rise and then return to 72 inches C. Lower and remain below 72 inches D. Lower and then return to 72 inches Proposed Answer: B Justification: B correct - the ERV opening will result in a level swell which will be corrected for by FWLC A incorrect - correct initial response but ERV steam flow is measured as part of the steam flow signal used by FWLC C/D incorrect - Reactor Water level will not lower Proposed references to be provided to applicants during examination: None Technical Reference(s): Nl-OP-01 section B K/A: 239002 SRVs A1.05 - Ability to predict and/or monitor changes in parameters associated with operating the SRVs controls including: Reactor Water Level 10 CFR Part 55 Content: 55.41(5,14) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A # 239002 239002 Importance Rating 3.9 4.0 Proposed Question: RO 056, SRO 060 The plant is operating at 100% power, with the following:

Battery Board 12 is lost

A sustained loss of Turbine Building ventilation has occurred

No operator action is taken Which one of the following describes the plant response per Technical Specification bases?

A. Reactor Coolant Safety Valves will lift.

B. Reactor Coolant Pressure Safety Limit will be violated.

C. Peak cladding temperatures greater than 15000 F will occur.

D. Loss of RPV inventory due to steam flow will exceed the Core Spray capacity.

Proposed Answer: A Justification: A is correct - Per T.S. 2.2.2, all 9 Safety Valves may lift to prevent exceeding the 1375 psig Safety Limit in the event of an MSIV closure from rated power followed by a high flux scram. BB 12 loss leaves the plant with only 3 ERVs, with insufficient relief capacity to keep pressure below the Safety Valve setpoints. Sustained loss of TB Ventilation will ultimately (no operator action) allow Main Steam Pipe Tunnel temperature to reach 200 F.

MSIV's will auto-close as a result. B is incorrect - Purpose of the Safeties lifting is to prevent exceeding the Pressure Safety Limit. C is incorrect - Plausible only if examinee is distracted to think that the "Minimum Number of ERVs Required for ED" is something greater than 3. D is incorrect - Plausible only if examinee is distracted to believe that the 3 available ERVs, combined with the lifting Safeties, let down inventory at a rate that is greater than Core Spray's capacity to replace.

Proposed references to be provided to applicants during examination: None Technical Reference(s): TS 2.2.2 Bases K/A: 239002 K3.01, Knowledge of the effect that a loss or malfunction of the Relief/Safety RELIEF/SAFETY VALVES will have on following: Reactor pressure Valves control 10 CFR Part 55 Content: 55.41(7) X 55.43 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group# 1 1 K/A # 241000 241000 Importance Rating 3.8 3.9 Proposed Question: RO 057, SRO 061 The plant is operating at 90% power, with the following:

Power ascension is in progress

Turbine Load Limiter is set at 95%

As reactor power is raised, which one of the following effects will be seen when the Turbine Load Limiter setpoint is reached?

A. The intercept valves will throttle closed and the bypass valves will open.

B. The control valves will stop opening, causing a high pressure reactor scram.

C. The control valves will stop opening and the bypass valves will open to maintain pressure.

D. There will be no effect because the Governor and Speed Load Changer are in control of the turbine.

Proposed Answer: C Justification: C is correct - Pressure is controlled by control, then bypass valves.

Load is controlled by control valves. The Load Limiter may be employed to limit control valve opening (Load). A is incorrect - IV's are not affected by the Load controls. B is incorrect - Ignores operation of Bypass valves. D is incorrect - The Load Limiter can limit the oil available to the control system, limiting CV opening.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1i-OP-31, 01-OPS-001-248-1-01 K/A: 241000 K1.01, Knowledge of the physical connections and/or cause-effect Reactor/Turbine relationships between REACTOR/TURBINE PRESSURE Pressure REGULATING SYSTEM and the following: Reactor power Regulating System 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank# X12523 New History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 1 K/A # 241000 Importance Rating 3.1 Proposed Question: RO 058 The plant is operating at 40% power when the turbine Speed Control Unit output linkage to one Combined Intercept Reheat Stop Valve for Low Pressure Turbine 'A' (LP 'A') becomes disconnected from the valve.

Without operator action, which one of the following describes an effect of this malfunction?

A. Turbine trip and reactor scram B. Increased vibrations on the LP 'A' journal bearings C. Potential turbine overspeed on a loss of generator load D. 10% closure on at least one Turbine Control Valve (TCV)

Proposed Answer: B Justification: B is correct - SCU linkage holds open the IV portion of the CRV. A disconnected linkage results in that IV going closed, stopping steam flow into one of the two inlets to LP 'A'.

Unbalanced steam flow into LP 'A' (thru the unaffected CRV) results in increased vibrations. A is incorrect - TSV/TCV closure scram signals, yes; CRV closure scram signals, no. C is incorrect - Stem suggests the opposite; this LP 'A' steam inlet is already closed. D is incorrect - Plausible only if examinee confuses the relationship between the CRVs (IV portion) reaching fully open at about the same time when TCVs are approximately 10% open, on a turbine startup.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Ni-OP-31, 01-OPS-001-248-1-01 K/A: 241000 K1.09, Knowledge of the physical connections and/or cause-effect Reactor/Turbine relationships between REACTOR/TURBINE PRESSURE REGULATING Pressure SYSTEM and the following: Combined intermediate valves: Plant-Specific Regulating System 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: -Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A # 259002 259002 Importance Rating 3.5 3.6 Proposed Question: RO 059, SRO 062 The plant is operating at 100% power, with the following:

Feedwater Pumps 11 and 13 are running

A fuse blows in the FCV 13 control circuit (loss of control signal)

Which one of the following describes the FCV 13 valve response and the response of its position indication on the F Panel?

VALVE POSITION INDICATION A. Fails as is Fails as is B. Fails as is Fails upscale C. Fails open Fails as is D. Fails open Fails upscale 1/1

,4., t}/1 Proposed Answer: A Justification:

-- ' ' ". 1 I ,

A is correct - Loss of control signal results in FCV lockup (remains in its current position); M/A station position indication does the same. BID are incorrect - Although FCV (valve itself) locks up, the position indication fails upscale when AC power is loss to the FCV 13 positioner, itself (not a control signal type of failure). C is incorrect - FCV locks up in place.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Ni-ARP-H3-4-6 K/A: 259002 K4.13 Knowledge of REACTOR WATER LEVEL CONTROL SYSTEM REACTOR design feature(s) and/or interlocks which provide for the following:

WATER LEVEL FWRV lockup CONTROL SYSTEM 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A # 259002 259002 Importance Rating 3.6 3.6 Proposed Question: R0060, SRO 063 The plant is at 100% power, with the following:

12 and 13 FW pumps are in service

FWLC is in a normal configuration

The selected RPV Level input signal to FWLC fails low Which one of the following describes the expected RPV level response and the action that should be taken?

RPV LEVEL RESPONSE ACTION A. Rise Take 13 FCV to MANUAL B. Rise Select the standby level input C. Lower Take 13 FCV to MANUAL D. Lower Select the standby level input Proposed Answer: A Justification: A is correct - Level input low will cause a level error that will act to open 13 FCV and raise RPV level, taking 13 FCV to manual will correct the problem in a controlled fashion B is incorrect - not the right action per OP-16, F. 10- all input swapping should be done with FCV's in MANUAL CID are incorrect - water level will rise not lower Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-16, Sec B/pg 6 & F.10 K/A: 259002 A2.03: Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Loss of level inputs.

10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A # 261000 261000 Importance Rating 3.2 3.4 Proposed Question: RO 061, SRO 064 A small drywell steam leak has occurred, with the following:

Torus venting via RBEVS train #11 has been established using the POST-LOCA vent line

Drywell pressure is 3.1 psig and steady Which one of the following describes the effect on the containment venting evolution if RBEVS Exhaust Fan 11 were to trip?

Venting flow will...

A. stop and Drywell pressure will rise B. continue at a much lower rate, allowing Drywell pressure to rise slightly C. stop until RBEVS train #12 auto starts; Drywell pressure may rise slightly D. continue at the same rate, but will be untreated before reaching the main stack Proposed Answer: A Justification: A is correct - The fan trip isolates the train and the flowpath. B is incorrect - dampers close on fan trip. C is incorrect - There is no such auto start feature. D is incorrect - BV-202-36 is closed; no such path thru normal RBVS is available.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-09; N1-EOP-1, Aft 10 K/A: 261000 Standby K3.03, Knowledge of the effect that a loss or malfunction of the Gas Treatment STANDBY GAS TREATMENT SYSTEM will have on following:

System Primary containment pressure: Mark-I&ll 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 2 LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 1 K/A # 261000 Importance Rating 2.7 Proposed Question: RO 062 The plant is operating, with the following:

Reactor Building Emergency Ventilation (RBEVS) is in a normal standby line-up

Power is lost to Channel 11 Reactor Building Exhaust Radiation monitor Which one of the following identifies the status of the Reactor Building Emergency Ventilation System (RBEVS) Inlet Dampers?

RBEVS 11 INLET RBEVS 12 INLET A. Closed Closed B. Open Closed C. Closed Open D. Open Open Proposed Answer: D Justification: D is correct - RBEVS inlet dampers are normally closed and open upon initiation, both trains will initiate on loss of power to a rad monitor A/BIC are incorrect - combinations of incorrect responses should examinee not recognize the expected response Proposed references to be provided to applicants during examination: None Technical Reference(s): NI-OP-10 K/A: 261000 Standby A4.09, Ability to manually operate and/or monitor in the control room:

Gas Treatment Ventilation valves/dampers System 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 1 1 K/A # 264000 264000 Importance Rating 3.4 3.5 Proposed Question: RO 063, SRO 066 The following conditions exist:

Core Spray Topping Pump 111 is clearance tagged

A Loss of coolant accident occurs

Containment Sprays are running

RPV water level is -90 inches and steady

Drywell pressure is 21 psig and slowly lowering Then, the following occur:

A loss of offsite power

Both emergency diesel generators simultaneously start and load.

With Containment Spray initiation signals still present, which one of the following identifies the number of Containment Spray Pumps running 50 seconds after the emergency diesel generators energize their powerboards?

A. 0 B. 2 C. 3 D. 4 Proposed Answer: B Justification: B is correct - Powerboard 103 pumps are unaffected by the tagged CSTP 111; so both Cont Spray Pumps 121 and 122 are running within 30 seconds of the EDG 103 re energizing PB 103. Due to CSTP 111 be OOS, a 3rd Cont Spray Pump won't be running until 55 seconds after PB 102 is re-energized (50 sec b/up timer + 5 sec normal sequencing). AIC/D are incorrect - Not by design. All are plausible if examinee does not understand this starting sequence logic.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-14, drawing C-19410-C, Sh. 8 K/A: 264000 A3.05, Ability to monitor automatic operations of the EMERGENCY Emergency GENERATORS (DIESEL/JET) including: Load shedding and Generators sequencing (Diesel/Jet) 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 1 K/A # 264000 Importance Rating 3.4 Proposed Question: RO 064 The plant is operating at 100% power, with the following:

Loss of 115KV Line 4 and T-101S

EDG 103 starts and successfully loads on PB 103

A fire is then confirmed in the EDG 103 room Which one of the following describes how the RO would remove EDG 103 from service per N1-OP-45, Emergency Diesel Generators?

Place...

A. EDG 103 output breaker in PTL, then place EDG 103 control switch in EMERGENCY STOP.

B. EDG 103 output breaker in PTL, then place EDG 103 control switch in STOP.

C. PB 103 normal supply breaker in PTL, then place EDG 103 control switch in EMERGENCY STOP.

D. PB 103 normal supply breaker in PTL, then place EDG 103 control switch in STOP.

Proposed Answer: C Justification: C correct - Must first lock out normal supply power to the bus, then satisfy the remaining engine control logic by indicating there is no need for the EDG to re-power the bus, by positioning the engine control sw in EMERG STOP. A/B incorrect normal supply power is not locked out. D incorrect - normal supply power is locked out, but EDG will re-start on bus undervoltage using the normal STOP switch.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Ni-OP-45, 01-OPS-001-264-1-01 K/A: 264000 A2.06: Ability to (a) predict the impacts of the following on the EMERGENCY EMERGENCY GENERATORS (DIESEL/JET); and (b) based on those predictions, GENERATORS use procedures to correct, control, or mitigate the consequences of those abnormal (DIESEL/JET) conditions or operations: Opening normal and/or alternate power to emergency bus 10 CFR Part 55 Content: 55.41(7,10) X 55.43 f55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 2 3 K/A # 201003 201003 Importance Rating 3.6 3.7 Proposed Question: RO 065,SRO080 The plant is operating at 100% power, with the following:

A Reactor scram occurs

10 scram inlet valves on withdrawn control rods fail to open

All scram outlet valves open per design Which one of the following describes the response of the 10 malfunctioning control rods?

The 10 control rods will...

A. scram at normal scram speed, with reactor pressure alone.

B. scram at normal scram speed, with accumulator pressure alone.

C. scram more slowly than the others, with reactor pressure alone.

D. scram more slowly than the others, with accumulator pressure alone.

Proposed Answer: C Justification: C is correct - With the reactor at rated pressure, the accumulator provides the motive force to achieve proper scram time, but reactor pressure will fully insert the rods. A is incorrect - The rods scram due to reactor pressure but slower with accumulator/charging header isolated. B is incorrect - Accumulator is isolated from mech. D is incorrect - Accumulator is isolated from mech.

Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-201-1-01 K/A: 201003 Control K4.04: Knowledge of CONTROL ROD AND DRIVE MECHANISM Rod and Drive design feature(s) and/or interlocks which provide for the following:

Mechanism The use of either accumulator or reactor water to SCRAM the control rod 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group# 2 2 K/A# 201006 201006 Importance Rating , 3.0 3.0 Proposed Question: RO 066, SRO 068 A reactor startup is in progress, with the following:

The tenth rod of RWM Group 01 has just been withdrawn to position 48

RWM Group 01 contains 12 rods Which one of the following identifies the number of RWM errors displayed when the next rod in sequence is selected?

INSERT ERRORS WITHDRAW ERRORS A. 0 rods displayed 0 rods displayed B. 2 rods displayed 0 rods displayed C. 0 rods displayed 2 rods displayed D. 2 rods displayed 2 rods displayed Proposed Answer: B Justification: B is correct - The RWM latches the next group when the last 2 rods in a group withdraw sequence remain inserted past the insert limits.

AIC/D are incorrect - Plausible combinations to discriminate against an examinee who lacks an understanding of what generates RWM errors.

Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-201-1-03 K/A: 201006 RWM A4.03, Ability to manually operate and/or monitor in the control room:

Latched group indication: PSpec(Not-BWR6) 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group# 2 2 K/A # 202001 202001 Importance Rating 2.8 2.9 Proposed Question: RO 067, SRO 070 The plant is planning for an upcoming refueling outage, with the following:

Work on Recirc Loop 14 will require plugging both the suction and discharge nozzles

All other Recirc Loops will be available For the upcoming outage, which one of the following describes Shutdown Cooling (SDC) availability and the reason?

Shutdown Cooling would be...

A. unaffected due to Recirc Loop 15 being available B. unavailable due to the loss of only the suction path C. unavailable due to the loss of only the discharge path D. unavailable due to the loss of both the suction and discharge paths Proposed Answer: B Justification: B is correct - The SDC system takes a suction on Recirc Loop 14 and returns via Recirc Loop 15 A/ICD are incorrect - the suction path and only the suction path will be lost Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-04, B.1, page 3 K/A: 202001 K3.08: Knowledge of the effect that a loss or malfunction of the RECIRCULATIO RECIRCULATION SYSTEM will have on following: Shutdown N SYSTEM cooling system 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 2 2 K/A # 204000 204000 Importance Rating 3.4 3.8 Proposed Question: RO 068,SRO 071 The plant is operating at 100% power, with the following:

RWCU system is in a normal lineup

RWCU Filter 11 is in service

Then, a RWCU system isolation occurs Which one of the following identifies the filter(s) that would require backwashing to restore the RWCU system to a normal lineup?

A. Neither filter B. Only Filter 11 C. Only Filter 12 D. Both Filters Proposed Answer: B Justification: B is correct - Filter 12 is in standby when the system isolates. Filter 11 must be backwashed and precoated to be either placed in standby or be put back in service. NC are incorrect - Filter 11 must be backwashed. D is incorrect - Because Filter 12 was already in standby upon system isolation, it does not require backwashing.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-40, N1-OP-3 K/A: 204000 Reactor 2.1.32, Ability to explain and apply system limits and precautions.

Water Cleanup System 10 CFR Part 55 Content: 55.41(10) X 55.43 (5) 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 2 1 K/A # 219000 219000 Importance Rating 2.5 2.9 Proposed Question: R0069, SRO 072 The plant is operating, with the following:

Torus Cooling is desired

Powerboard 167 is not available Which one of the following states the Containment Spray loop(s) that can be placed in Torus Cooling from the Control Room?

A. 121 only B. 122 only C. 121 or 122 D. None Proposed Answer: D Justification: D correct - Without PB 167 (600 VAC), can't open BV-80-118, which is necessary to place any of the 4 loops in torus cooling. AIBIC incorrect - BV-80-118 is the only valve (common to placing any of the 4 loops in service) that restricts the ability to place at least one of the loops in service. All other valves to be operated in aligning for torus cooling are DC-solenoid operated, AOVs.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-14, PID C-18012-C, Sh.2 K/A: 219000 K2.01: Knowledge of electrical power supplies to the following:

RHR/LPCI: Valves Torus/Suppressio n Pool Cooling Mode 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank#

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group# 2 2 K/A# 219000 219000 Importance Rating 2.8 2.9 Proposed Question: R0070, SRO 073 Containment Spray Loop 121 is cooling the Torus when a tube ruptures in its heat exchanger.

Which one of the following describes the initial response of Containment Spray Pump 121 discharge pressure and Raw Water Pump 121 flow?

SPRAY PUMP DISCHARGE PRESSURE RAW WATER PUMP FLOW A. Lowers Rises B. Lowers Lowers C. Rises Rises D. Rises Lowers Proposed Answer: C Justification: C correct - RW pressure is > Cont Spray pressure. Ruture results in a higher Cont Spray pressure and increased RW flow (additional flow path made available). Both of these parameters are monitored in the control room. AIBID incorrect - for reasons as already described for A.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-14, B.0, D.3.0 K/A: 219000 A1.03: Ability to predict and/or monitor changes in parameters RHR/LPCI: associated with operating the RHR/LPCI: TORUS/SUPPRESSION TORUS/SUPPRE POOL COOLING MODE controls including: System pressure SSION POOL COOLING MODE 10 CFR Part 55 Content: 55.41(7,14) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group# 2 2 K/A # 226001 226001 Importance Rating 2.6 2.8 Proposed Question: RO 071, SRO 058 The plant is determined to be above the upper-most limit of Figure K, Containment Spray Initiation Limit.

Per the EOP Bases, which one of the following identifies the potential consequence of spraying the containment at this time, and why that consequence might occur?

POTENTIAL CONSEQUENCE WHY IT MIGHT OCCUR A. Vacuum Breakers operate Thermal transient B. Vacuum Breakers operate Rapid pressure reduction C. Containment fails Thermal transient D. Containment fails Rapid pressure reduction Proposed Answer: D Justification: D correct - Per EOP bases evaporative cooling could result in excessive negative pressure in the conatinment either before the vacuum breakers can respond or in excess of the relief capability/non-condensible content of the Torus AIBIC incorrect combination of response for being left of the curve and/or wrong reason.

Proposed references to be provided to applicants during examination: EOP flowcharts Technical Reference(s): Nl-ODP-PRO-0305 K/A: 226001 K5.06 - Knowledge of the operational implications of the following RHR/LPCI: concepts as they apply to the RHR/LPCI: CTMT Spray Mode:

CTMT Spray Vacuum breaker operation Mode 10 CFR Part 55 Content: 55.41(5) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 2 2 K/A # 245000 245000 Importance Rating 2.8 3.1 Proposed Question: R0072, SRO 074 Which one of the following describes a turbine vibration condition that would require the operator to immediately trip the turbine per N1-OP-31, "Tandem Compound Reheat Turbine"?

A. 7 mils for 2 minutes while at 400 rpm B. 9 mils for 1 minute while at 700 rpm C. 7 mils for 2 minutes while at 1000 rpm D. 9 mils for 1 minute while at 1300 rpm Proposed Answer: B Justification: B is correct - turbine trip is required @ 8 mils below 800 rpm.

A is incorrect - 7 mils is the limit for continued operation between 800-1400 rpm. C is incorrect - above the 800 rpm limitation. D is incorrect - vibration 14 mils would require immediate trip at this rpm.

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # X12529 New History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 2 K/A # 256000 Importance Rating 2.8 Proposed Question: RO 073 The plant is operating at 90% power, with the following:

Condensate Pumps 11 and 12 are running

FW Booster Pumps 12 and 13 are running

FW Pumps 12 and 13 are running

Condensate Pump 13 is tagged out

Condensate Pump 11 now trips Which one of the following describes the plant response and the required operator action?

A. FW Booster pumps trip on low suction pressure; immediately scram the reactor.

B. FW Booster pump suction pressure slowly lowers; perform Emergency Power Reduction.

C. Feedwater pumps trip on low suction or discharge pressure; immediately scram the reactor.

D. Feedwater pump suction pressure slowly lowers; perform Emergency Power Reduction.

Proposed Answer: C Justification: C is correct - FWP 12 trips on low suction pressure; FWP 13 trips on low discharge pressure; RPV water level rapidly lowers; scram the plant before +53". A is incorrect - There are no low suction pressure trips for the FWBPs. BID are incorrect - FWBP suction pressures rapidly lower with only 1 running Cond Pump @ 90%. No time for EPR to make a difference before RPV water level drops to +53".

Proposed references to be provided to applicants during examination: None Technical Reference(s): Nl-OP-15A K/A: 256000 A1.03: Ability to predict and/or monitor changes in parameters REACTOR associated with operating the REACTOR CONDENSATE SYSTEM CONDENSATE controls including: System Pressure SYSTEM 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 2 1 K/A # 262001 262001 Importance Rating 3.5 3.7 Proposed Question: RO 074,SRO 065 The plant is operating at 100% power, with the following:

Main generator output voltage is normal when the amplidyne begins to fail (Time = 0 seconds)

The following main generator output voltages result:

TIME OUTPUT VOLTS 20 seconds 26,000 30 seconds 27,000 40 seconds 28,000 50 seconds 29,000 60 seconds 30,000 Without operator action, which one of the following identifies the earliest time by which the house loads should have fast transferred to reserve power?

A. 30 seconds B. 40 seconds C. 50 seconds D. 60 seconds Proposed Answer: C Justification: C correct - Generator over-voltage trips are: 110% x normal (24,000) for 30 seconds, and 118% x normal for 2 seconds. Given time-line indicates output volts will reach the first trip point (26,400) @ Time = 24 seconds, and will reach the 2nd trip point (28,320) @ Time = 43 seconds. The 1 st trip's timer will trip the generator at Time = 54 seconds (i.e., 24 + 30).

The 2 nd trip's timer will trip the generator at Time = 45 seconds (43 + 2). By Time = 50 seconds, house loads will have auto-transferred to reserve transformers. The turbine will have tripped, and the reactor will have scrammed, as well, but are incidental to this question. A/B/D incorrect - See explanation above.

Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-262-1-01 K/A: 262001 A.C. K6.03 Knowledge of the effect that a loss or malfunction of the following will Electrical have on the A.C. Electrical Distribution; Generator trip.

SDistribution 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 2 K/A # 263000 Importance Rating 3.3 Proposed Question: RO 075 The plant is operating at 100% power, with the following:

Static Battery Chargers (SBC) 161A and 171A are in service

Powerboard 16A is inadvertently deenergized then re-energized TWO minutes after Powerboard 16J.is re-energized, and with no additional operator action, which one of the following identifies the Battery Board 11 and 12 voltages?

Battery Board 11 Battery Board 12 A. 125 VDC 125 VDC B. 125 VDC 135 VDC C. 135 VDC 125 VDC D. 135 VDC 135 VDC Proposed Answer: D Justification: D is correct- 100 seconds after AC power is restored to SBC161A (from powerboard 164the SBC will align itself to the Battery 11 and fk)Aý return voltage to the normal float voltage of 135VDC AIB are incorrect

- Battery 11 voltage will return to 135 VDC. C is incorrect - Battery 12 is unaffected as it's SBC is powered from PB-17A(B).

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-47A, B.4 K/A: 263000 D.C. K1.01, Knowledge of the physical connections and/or cause- effect Electrical relationships between D.C. ELECTRICAL DISTRIBUTION and the Distribution following: A.C. electrical distribution 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 2 2 K/A # 271000 271000 Importance Rating 3.6 3.3 Proposed Question: RO 076, SRO 075 The plant is operating at 100% power, with the following:

H1-2-6, AIR EJECTOR OFF GAS FLOW-PRESS TEMP HIGH alarms

C145 OFFGAS TEMP HIGH is in alarm

C135 OFFGAS PRESSURE HIGH is in alarm

H1-1-6, AIR EJECTOR HI-LO STM PR V CLOSED alarms Which one of the following is required?

A. Verify Offgas Stack BV closes B. Place a second Offgas vacuum pump in service C. Perform an Emergency Power Reduction D. Place the Mode Switch in Shutdown Proposed Answer: C Justification: C correct - Offgas conditions would result in closure of SJAE BV's, per N1-OP-25, H.4.1.1 Emer Power Red required A incorrect - Stack BV closes on high rad only B incorrect - OP-25 action, but not required for conditions D incorrect - SCRAM not immediately required on Offgas isolation Proposed references to be provided to applicants during examination: None Technical Reference(s): Ni-ARP-H 1-2-6, Ni-OP-25, H.4 K/A: 271000 Offgas K4.08 - Knowledge of Offgas design feature(s) and or interlock(s) which provide for the following: Automatic system isolation 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group# 2 2 K/A # 272000 272000 Importance Rating 3.6 3.5 Proposed Question: RO 077, SRO 076 Containment inerting is in progress, with the following:

Annunciator H11-1-8, STACK GAS MONITORS HIGH RADIATION, alarms

RN10A is above the HIGH-HIGH alarm setpoint

RN1 OB has just reached the HIGH alarm setpoint and is rising Which one of the following describes the required operator action?

Verify...

A. isolation of the inerting lineup has occurred B. Reactor Building Emergency Ventilation has started C. isolation of the inerting lineup when RN10B reaches the HIGH-HIGH setpoint D. Reactor Building Emergency Ventilation starts when RN 1OB reaches the HIGH-HIGH setpoint Proposed Answer: C Justification: C is correct out-of-2 logic (both RNI 0's > HIGH-HIGH) required to initiate isolation of PC vent/purge valves. A is incorrect - for the reasons described above. BID are incorrect- RN1OA/B have no relationship with Secondary ContainmentlRBEVS.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Ni-OP-50B, Ni1-ARP-H1-1I-8 K/A: 272000 Radiation A3.09, Ability to monitor automatic operations of the RADIATION Monitoring MONITORING SYSTEM including: Containment Isolation Indications.

System 10 CFR Part 55 Content: 55.41(7,11) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank# X New History 2000 NMP1 SRO audit, modified Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2

Examination Outline Cross-reference: Level RO Tier # 2 Group # 2 K/A # 286000 Importance Rating 2.9 Proposed Question: RO 078 Which one of the following describes how the fire protection system responds to a loss of Powerboard 176 when Fire Jockey Pump 12 is running?

Concerning the Fire Jockey pumps...

A. both lose power; the Motor Driven Fire Pump starts at 110 psig to restore header pressure.

B. both lose power; the Diesel Driven Fire Pump starts at 110 psig to restore header pressure.

C. 11 loses power; Fire Jockey Pump 12 maintains header pressure.

D. 12 loses power; Fire Jockey Pump 11 starts to maintain header pressure.

Proposed Answer: A Justification: A correct- Both jockey pumps are powered from PB 176. MDFP starts at 110 psig. B incorrect - DDFP starts at 100 psig. CID incorrect - Both jockey pumps are powered from the same source.

Proposed references to be provided to applicants during examination: None Technical Reference(s): 01-OPS-001-286-1-01 K/A: 286000 Fire K2.02 - Knowledge of electrical power supplies to the following:

I Protection Pumps 10 CFR Part 55 Content: 55.41(7) X F55.43 55.45 Question Source: Bank #

Modified Bank # X12651 New History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 2 K/A # 290001 Importance Rating 3.3 Proposed Question: RO 079 The plant is operating at 100% power, with the following:

Light can be seen from outside, along the train rails coming under the Reactor Building Outer Swing Door.

Inner Lift Door is closed and sealed

The SSS has been informed Which one of the following identifies the status of Secondary Containment Integrity and the required actions?

Secondary Containment Integrity is ...

A. intact and an ACR is required.

B. intact and RBEVS must be started.

C. not met and an ACR is required.

D. not met and RBEVS must be started.

Proposed Answer: A Justification: A correct - If the inner lift door is properly sealed, Secondary Containment is intact, however, the loss of the rail seals constitutes a degradation requiring the SSS to be notified such that compensatory actions can be taken. B incorrect - RBEVS is not required to be operated. CID incorrect - Secondary Containment Integrity is maintained by the inner lift door seals.

Proposed references to be provided to applicants during examination: None Technical Reference(s): NI-OP-52; 01-OPS-001-223-1-03 K/A: 290001 A2.01: Ability to (a) predict the impacts of the following on the SECONDARY SECONDARY CONTAINMENT; and (b) based on those predictions, CONTAINMENT use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: tPersonnel airlock failure 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 2 2 K/A # 290003 290003 Importance Rating 3.4 3.5 Proposed Question: RO 080, SRO 077 The plant is operating at 100% power, with the following:

Control Room Vent Rad Monitor Channel 11 fails

The radiation monitor is bypassed With no operator action, which one of the following describes the status of the Control Room Emergency Ventilation System 11 and 12 should a high radiation condition actually occur?

CREVS 11 CREVS 12 A. Standby Standby B. Running Standby C. Standby Running D. Running Running Proposed Answer: D Justification: D correct - Only one radiation monitor input is required for CREVS initiation A/BIC are incorrect- only one signal is required so both trains of CREVS will be running Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-49 section B K/A: 290003 K1.01 Knowledge of the physical connections and/or cause- effect CONTROL relationships between CONTROL ROOM HVAC and the following:

ROOM HVAC Radiation monitors 10 CFR Part 55 Content: 55.41(7,10) X 55.43 (2) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group# 2 2 K/A # 290003 290003 Importance Rating 2.7 2.9 Proposed Question: RO 081, SRO 078 The plant is operating at 100% power, with the following:

An after-cooler tube leak has developed in the LEAD Instrument Air Compressor #11

The BACKUP Instrument Air Compressor (IAC) #12 is unaffected and is maintaining air header pressure Assuming no loss of instrument air, which one of the following describes the potential effect of this condition on the Control Room Ventilation (CRV) System?

A. Chiller compressors may trip on RBCLC high temperature.

B. Chiller compressors may cycle off and on as RBCLC flow fluctuates.

C. CRV normal supply dampers may drift open and closed as IAC #12 loads and unloads.

D. CRV emergency supply dampers may drift open and closed as IAC #12 loads and unloads.

Proposed Answer: B Justification: B is correct - RBCLC supplies the IAC inter-after-coolers. A tube leak can cause RBCLC flow fluctuations. The CRV chiller compressors will cycle on/off (automatically) as RBCLC supply flow moves above and below 51 gpm at the compressors. A is incorrect - compressors are unaffected by RBCLC temps; no such high temp trip exists. CID are incorrect - Plausible distractors for the examinee who believes there is a relationship between lower than normal IAS pressures (while on the BACKUP IAC) and the I potential for CRV dampers to operate abnormally.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-SOP-8; Ni-OP-011; Ni-OP-49 K/A: 290003 Control K6.02, Knowledge of the effect that a loss or malfunction of the following will I Room HVAC have on the CONTROL ROOM HVAC: Component cooling water systems 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X SLOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 2 K/A # 300000 Importance Rating 3.3 Proposed Question: RO 082 An operator reports that the air supply line to BV 16-08, FW Heater 123 Extraction Steam bypass valve, broke off when he attempted to tighten a leaky fitting.

Which one of the following identifies the BV 16-08 bypass valve and FW temperature response?

BV 16-08 FW Temperature A. Opens Lowers B. Opens Rises C. Closes Lowers D. Closes Rises Proposed Answer: A Justification: A correct - The BV fails open on loss of air bypassing extraction steam to the main condenser effectively removing FW heating from the 123 heater and FW temperature lowers. B incorrect - BV fails open but FW temperature will lower not rise CID incorrect - BV fails open Proposed references to be provided to applicants during examination: None Technical Reference(s): C-18002-C sheet 3 K/A: 300000 K3.02: Knowledge of the effect that a loss or malfunction of the INSTRUMENT (INSTRUMENT AIR SYSTEM) will have on the following: Systems AIR SYSTEM having pneumatic valves and controls 10 CFR Part 55 Content: 55.41(7,) X 55.43 55.45 Question Source: Bank#

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 2 2 K/A # 400000 400000 Importance Rating 2.8 3.0 Proposed Question: RO 083, SRO 079 The plant is operating, with the following:

Annunciator H1-4-4, CLOSED LOOP COOL MAKEUP TANK LEVEL HIGH-LOW, alarms

Instrument Air to TBCLC LCV 71-127 is lost Which one of the following describes the expected TBCLC MU TANK level response and the action that should be taken?

MU TANK LEVEL ACTION A. Rising Isolate the MU Tank LCV by shutting its BV B. Rising Verify auto-start of the standby Condensate Transfer pump C. Lowering Take manual control of the MU Tank LCV D. Lowering Verify auto-start of the standby Condensate Transfer pump Proposed Answer: A Justification: A is correct -The LCV fails open on a loss of air and the ARP requires isolating the LCV B is incorrect - There is no auto-start feature for the standby Condensate Transfer pump. CID is incorrect With the LCV failed open the tank level will rise not lower Proposed references to be provided to applicants during examination: None Technical Reference(s): N1 -ARP-H 1-3-4, 4-4 K/A: 400000 CCWS A2.02: Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

High/low surge tank level 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New NEW History

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 2 2 K/A# 215001 215001 Importance Rating Proposed Question: RO 084, SRO 082 The plant is operating at 100% power, with the following:

A TIP detector is stationary at the top of the core for a trace Loss of coolant accident occurs S RPV water level is +25 inches and slowly lowering S Drywell pressure is 3.6 psig and slowly rising Which one of the following identifies the status of the TIP detector and describes the radiation level in the TIP Room?

TIP DETECTOR RADIATION LEVEL A. Stationary at top of core Elevated B. Stationary at top of core Normal background C. Retracting Elevated D. Retracting Normal background Proposed Answer: C Justification: C is correct - High DW pressure, alone, initiates a TIP isolation. TIP room rad levels were already higher than normal and will rise even higher as the detector automatically retracts, until it is in shield. NB are incorrect - Plausible only if examinee believes it takes both High DW pressure and RPV Lo-Lo water level to initiate TIP isolation (level is still above +5 inches). D is incorrect - while the detector was out-of-shield and the TIP tube un-isolated from the core, rad levels were already elevated, not normal.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-39 K/A: 215001 A1.01, Ability to predict and/or monitor changes in parameters associated Traversing In- with operating the Traversing In-core Probe controls including: Radiation core Probe levels 10 CFR Part 55 Content: 55.41(7,12) X 55.43 (4) X 55.45 Question Source: Bank#

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group# 3 K/A # 215001 Importance Rating 3.1 Proposed Question: RO 085 A full core calibration of the LPRMs using the TIP System is in progress, with the following:

The detector from machine number 3 is inserting

The inserting TIP has reached the core bottom limit

Loss of RPS bus 11 occurs Which one of the following describes the effect on the TIP system?

The TIP will...

A. Withdraw at slow speed B. Withdraw at fast speed C. Insert at slow speed D. Insert at fast speed Proposed Answer: C Justification: C correct - The half isolation signal generated from the RPS power loss has no affect on TIP operation and at the core limit the TIP will shift to slow speed. A/B incorrect - Half isolation will not cause TIP to retract D incorrect - At the core limit the TIP will shift to slow speed.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-40;N1-OP-39 K/A: 215001 K6.04: Knowledge of the effect that a loss or malfunction of the TRAVERSING following will have on the TRAVERSING IN-CORE PROBE: Primary IN-CORE containment isolation system: Mark-I&ll(Not-BWR1)

PROBE 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Group # 3 3 KJA # 233000 233000 Importance Rating 2.5 2.5 Proposed Question: RO 086, SRO 083 The plant is operating at power, with the following:

Fuel Pool filter #11 is scheduled to be backwashed and precoated

Valve 54-105, ROOT - LT 54 FP FLTR SLUDGE TANK, on RB 237N has been inadvertently closed rendering the level transmitter inoperative The tank indicates empty with actual level at 4.0 feet Which one of the following describes the consequences if this condition is not corrected before the backwash evolution is conducted?

A. The Reactor Building Ventilation System will be contaminated.

B. The sludge tank will over-flow back to the in service filter.

C. This is the normal level for conducting a backwash and there will be no consequences.

D. A Sludge Tank High Level alarm will alert the operator to the problem prior to other adverse consequences.

Proposed Answer: A Justification: A is correct - level will exceed 4.5' during the backwash flooding the RB vent system. B is incorrect - Backwash dump blocking valve is interlocked closed on an in-service filter, preventing back-flow to the in-service filter. C is incorrect - Tank level should be low at start

-1.3'. D is incorrect - This transmitter is the only one on the tank AND the high level alarm is at 5.2' which is above the level where RB ventilation floods.

Proposed references to be provided to applicants during examination: None Technical Reference(s): ARP-SFP-1-6; N1-OP-6, D.1.0 ;P&ID C-18008-C K/A: 233000 Fuel Pool K1.14, Knowledge of the physical connections and/or cause- effect Cooling and relationships between FUEL POOL COOLING AND CLEAN-UP and Cleanup the following: Reactor building ventilation 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level RO Tier # 2 Group # 3 K/A # 233000 Importance Rating 2.8 Proposed Question: RO 087 The plant is in an outage, with the following:

The core is offloaded and the cavity is drained

A loss of Instrument Air to the Spent Fuel Pool Cooling (SFP) system occurs Which one of the following identifies the effect on Spent Fuel Pool Temperature and identifies an action to mitigate?

Effect on temperature Action to mitigate A. Lowers Remove the operating SFP pump from service B. Lowers Throttle closed SFP heat exchanger RBCLC outlet valve C. Rises Feed and Bleed using fire hoses and the main condenser D. Rises Throttle open SFP heat exchanger RBCLC outlet valve Proposed Answer: C Justification: C correct - fuel pool cooling pumps trip on a loss of air which will cause pool temperature to rise, feed and bleed is available A incorrect - pool temperature will rise, ifair system effect on the RBCLC supply to SFP heat exchanger is not understood (fails as is) and pump trip is not recognized could be selected B incorrect - pool temp will rise. D incorrect - temp will rise but without instrument air and the SFP pumps tripped, the normal method of temperature control will be ineffective.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Ni-ODP-OPS-0108, N1-SOP-6 K/A: 233000 FUEL A2.15: Ability to (a) predict the impacts of the following on the FUEL POOL COOLING POOL COOLING AND CLEAN-UP; and (b) based on those AND CLEAN-UP predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High system temperature 10 CFR Part 55 Content: 55.41(7,10) X 55.43_

55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: NRC: B and D SDC distractors are implausible with cavity drained. Replaced B and D with normal temperature control methods.

Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Group #

K/A# 2.1.18 2.1.18 Importance Rating 2.9 3.0 Proposed Question: RO 088, SRO 086 A reactor startup is in progress, with the following:

CSO is determining the Reactor Period for the required critical data

At stopwatch time 00:00:00, SRM 'A' reads 1,000 cps

At stopwatch time 01:57:00, SRM 'A' reads 2,000 cps Which one of the following identifies the value for Reactor Period to be entered in the CSO Log?

A. 81 seconds B. 109 seconds C. 168 seconds D. 226 seconds Proposed Answer: C Justification: C correct - Counts have doubled in a lapse time of one minute+57 seconds; i.e., 117 seconds.

CSO then multiplies 117 x 1.44 to get a Period = 168 seconds. A incorrect - plausible if the candidate divides 117 seconds by 1.44, rather than multiply. B incorrect - plausible if candidate divides 157 seconds by 1.44, rather than multiply. D incorrect - the most likely CSO error. CSO misreads stop time as 157 seconds, rather than the actual stop time of 117 seconds. Multiplying 157 x 1.44 = 226 seconds.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-43A, Step 2.5.2 K/A: 1 GENERIC 2.1.18 2.1.18, Ability to make accurate, clear, and concise logs, records, status boards, and reports.

10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Group #

K/A # 2.1.9 2.1.9 Importance Rating 2.5 4.0 Proposed Question: RO 089, SRO 087 The plant is operating at power, with the following:

A rod sequence exchange is in progress

An RO candidate is present to perform a control rod reactivity manipulation for qualification

The candidate has just selected a rod to be moved

The ATC RO notices the wrong rod is displayed as selected on the F Panel, but has not yet been moved Which one of the following describes the ATC RO's required action?

Direct the candidate to...

A. stop and back away from the E Panel.

B. stop, self-check again, and re-select the correct rod.

C. verify the selected rod against the RMR form for the sequence exchange.

D. verify the rod selected display on the F Panel against the same display on the E Panel.

Proposed Answer: A Justification: A correct - per Ops Manual, Section 3.3.8.j.4. B/ICD incorrect - each of these lack the 'positive trainee control' expected anytime an evolution goes wrong.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Ops Manual 3.3.8.j.4 K/A: I Generic 2.1.9 2.1.9: Ability to direct personnel activities inside the Control Room 10 CFR Part 55 Content: 55.41(10) X 55.43 Fý55.4_5 Question Source: Bank#

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 3 Group #

K/A# 2.1.31 Importance Rating 4.2 Proposed Question: RO 090 The plant is operating at 100% power, with the following:

11 FW FCV is in MANUAL

13 FW FCV is in BAL

RPV level is 73 inches and stable

Then, power is reduced to 90%

After the power transient, RPV level is 73 inches and stable Which one of the following describes the response of the 13 FW FCV indication during the RPV level transient?

A. Deviation meter first swings right, then swings left, and stabilizes at the center.

B. Deviation meter first swings left, then swings right, and stabilizes to the left of center.

C. Demand meter first swings right, then swings left, and stabilizes at a value lower than before the transient.

D. Demand meter first swings left, then swings right, and stabilizes at a value greater than before the transient.

Proposed Answer: A Q1l4-, erIe-X,4/W P*P--Jo c8le-tA --,6vr'PL 1e/wu./ue-Justification: A is correct - During steady state (zero level error signal), the deviation meter is kept 'balanced' at the center of the meter scale (i.e., @ the 'zero deviation' or red dot position on the scale). When the downpower begins, the deviation meter swings to the right, because a positive level error exists, causing the FCV to be open more than is required for the now lowering steam flow and feed flow demands. As the transient comes to an end, the deviation meters swings left, back towards the center (zero level error position) and stabilizes there. RPV water level is back to its pre-transient value. B is incorrect - See 'A' justification. Additionally, with the 13 FCV in BAL (i.e.,

automatic mode), the controller deviation will return to zero (i.e., centered). CID are incorrect - 'C' is a very plausible distractor because the demand meter does indeed also move during the transient. However, with a lowering feed flow requirement during the downpower, the demand meter swings left of its pre-transient value and stablizes at some lower value (i.e., this lower value is a result of a demand signal to the FCV, itself, that positions the valve less open than it was before (to accommodate the lower feed flow that is now required)). The demand meter has no reason to swing back toward its pre-transient value, because, indeed, the controller stabilizes with there being a lower demand signal sent to the FCV. This same discussion (a mirror image) justifies why 'D' is incorrect.

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: 1-02-095

Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Group#

K/A# 2.2.11 2.2.11 Importance Rating 2.5 3.4 Proposed Question: RO 091, SRO 090 A Temporary Change Package has just been used to perform the following:

"* Annunciator H3-1-4, FEEDWATER HEATER 111 - 115 LEVEL HIGH, has had 3 of its inputs defeated

"* The work has been documented in the Defeated Annunciator Log

"* There are no previous entries for this annunciator in the Defeated Annunciator Log Which one of the following identifies the color of the sticker to be attached to the annunciator window, and describes the information to be included on that sticker, per GAP-OPS-01, Administration of Operations?

STICKER COLOR INFORMATION TO INCLUDE A. Yellow Defeated Annunciator Log Sequence Number B. Yellow Temporary Change Package number C. Red Date that the inputs were defeated D. Red Annunciator window number (H3-1-4)

Proposed Answer: B Justification: B is correct - Per GAP-OPS-01, 3.120.7, attach a yellow sticker whenever < all of the inputs are defeated. There are 5 inputs to this window. Information to include on the sticker includes: authorizing package ID (the TCP #, in this case), and affected computer points (when practicable). A is incorrect - No such Log number exists. C/D are incorrect- Red sticker is required only when defeating all of the inputs to the given annunciator.

Proposed references to be provided to applicants during examination: None Technical Reference(s): GAP-OPS-01 K/A: I Generic 2.2.11 2.2.11., Knowledge of the process for controlling temporary changes 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Group #

K/A # Generic Generic 2.2.24 2.2.24 Importance Rating 2.6 3.8 Proposed Question: RO 092, SRO 091 The plant is operating at 100% power, with the following:

TIME CONDITION 0100 Containment vent and purge valve High Drywell Pressure isolation logic Channel 1 passes its surveillance test and is returned to normal 0100 Channel 2 is placed in test 0200 Channel 2 fails its surveillance when its trip setpoint is found out of spec high 0300 Channels 3 and 4 have passed their surveillance and have been returned to normal Which one of the following identifies the time at which the Technical Specification LCO clock should be started for the inoperable channel?

A. 0100 B. 0200 C. 0700 D. 0800 Proposed Answer: A Justification: A is correct- Channel goes to inop status at the time it's placed in test. Since it failed its test, it remains in test, and therefore remains inoperable. Direction is provided at the beginning of STs to declare equipment inop when it is rendered inop by the ST (e.g., channel is placed in test). B is incorrect - By this time, the channel has already been inop for an hour. C is incorrect - This is plausible only if examinee believes the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ACT extends the operability status of the channel. D is incorrect - This plausible if the examinee believes the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ACT extends operability status beyond time where failed surveillance is discovered; presumably because the channel is still in test.

Proposed references to be provided to applicants during examination: T.S. 3.6.2 Technical Reference(s): TS 3.6.2 K/A: I Generic 2.2.24 2.2.24, Ability to analyze the effect of maintenance activities on LCO status 10 CFR Part 55 Content: 55.41(10) X 55.43 (3) X 55.45 Question Source: Bank #

Modified Bank # X New History NMP1 1998 NRC EXAM, modified Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 3 Group #

K/A # 2.2.33 Importance Rating 2.5 Proposed Question: RO 093 A control rod sequence exchange is in progress, with the following:

Reactor Power is 75%

Control Rod 26-27 is at position 14 Which one of the following Control Rods would be permissible to move from position 12 to position 18?

A. Control Rod 26-31 B. Control Rod 26-35 C. Control Rod 30-23 D. Control Rod 22-27 Proposed Answer: B Justification: B correct - Rod 26-35 is NOT adjacent to 26-27 therefore rod tips will NOT cross A/ICD incorrect - All these rods are either face or diagonally adjacent to rod 26-27 Proposed references to be provided to applicants during examination: N1-REP-20 Technical Reference(s): N1-REP-20 K/A: I Generic 2.2.33 2.2.33: Knowledge of control rod programming 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Group #

K/A # Generic Generic 2.4.47 2.4.47

__Importance Rating 3.4 3.7 Proposed Question: RO 094 ,SRO 096 A normal power reduction from 100% to 95% is in progress, with the following:

"* The power reduction is for the Control Rod Exercising and Operability Test, N1-ST-WI

"* A loss of annunciator power to Panels Al/B1 occurs Which one of the following describes the impact of this loss on continuing the power reduction and surveillance test?

CAN POWER REDUCTION CONTINUE? CAN SURVEILLANCE CONTINUE?

A. Yes Yes B. Yes No C. No Yes D. No No Proposed Answer: D Justification: D is correct - Per Ni-SOP-15, all unnecessary plant transients and testing shall be halted, no matter whether a partial loss, or full loss, of annunciator capability. The surveillance must be aborted; so, too, must any normal power change; i.e., an

,unnecessary plant transient'. AIBIC are incorrect - not allowed by SOP-15.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Nl-SOP-15 K/A: I Generic 2.4.32 2.4.32, Knowledge of operator response to loss of all annunciators 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Group# -

K/A # Generic Generic 2.3.2 2.3.2

__Importance Rating 2.5 2.9 Proposed Question: RO 095, SRO 095 According to GAP-RPP-08, Control of High Radiation/Locked High RadiationNery High Radiation Areas, which one of the following describes the minimum controls requirements for rooms with potential for very high radiation levels?

The room shall be locked at all times, its entrance conspicuously posted, and...

A. guarded when the room is left open B. fenced off (up to the overhead) when the room is left open C. guarded when the room is left open, and have a conspicuous flashing light nearby D. fenced off (up to the overhead) when the room is left open, and have a conspicuous flashing light nearby Proposed Answer: A Justification: A correct - per GAP-RPP-08, section 3.1. The stem focuses on VHR rooms, rather than open (non-lockable, non-enclosed) areas. BID incorrect - fencing off to the overhead is the GAP requirement for barricading a non-enclosed VHR (sections 3.1.2 and 3.1.7). C incorrect - conspicuous flashing warning lights are required near the entrance to non-enclosed VHRs, only.

Proposed references to be provided to applicants during examination: None Technical Reference(s): GAP-RPP-08 K/A: I Generic 2.3.2 2.3.2, Knowledge of facility radiation ALARA requirements.

10 CFR Part 55 Content: 55.41(12) X 55.43 55.45 Question Source: Bank #

Modified Bank#

New X History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO Tier # 3 Group #

K/A # 2.3.4 Importance Rating 2.5 Proposed Question: RO 096 As the In-Plant 'E', you are assigned to the Turbine Building rounds for your shift, with the following:

A contaminated area is established around each FW Booster pump

An inspection of each FW Booster pump is required early in the shift

The RWP requirement for tours in each FW Booster pump area is shoe covers and booties

No leaks exist in any of the areas

There are no other contaminated Turbine Building areas that you are required to enter Which one of the following describes when your EARLIEST frisk is required per GAP-RPP-01, Radiation Protection Program?

Your earliest frisk is required when you exit the...

A. first FW Booster pump area B. second FW Booster pump area C. third FW Booster pump area D. Turbine Building Proposed Answer: C I

C correct - Per GAP-RPP-01, Step 3.5.2, frisking is not required until exiting the last of the multiple contaminated areas. This is provided that full dress-out is not required, and that there is no reason to expect that the individual may become contaminated. A/B incorrect beyond the minimum (earliest) frisking requirements for the multiple area entries. D incorrect - Too late; does not meet the minimum (earliest) frisk requirements.

to be Proposed provided to applicants during examination: None references Proposed references to be provided to applicants during examination: None Technical Reference(s): GAP-RPP-01 K/A: Generic 2.3.4 2.3.4: Knowledge of radiation exposure limits and contamination control/including permissible levels in excess of those authorized 10 CFR Part 55 Content: 55.41(12) X 55.43 55.45 Question Source: Bank #

Modified Bank # X New History NMP1,2000 audit

Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Group #

K/A # Generic Generic 2.1.10 2.1.10 Importance Rating 2.7 3.9 Proposed Question: RO 097, SRO 097 A particular Technical Specification surveillance test (ST) is required to be performed once per month, with the following:

"* The ST was last completed on 9/1/02

"* The Next Due Date is 10/2/02 Which one of the following identifies when the ST will become 'overdue'?

The last day the ST can be completed before going overdue is...

A. 10/6/02 B. 10/9/02 C. 10/10/02 D. 10/13/02 Proposed Answer: B Justification: B is correct - Per TS 4.0.1, max allowable ST extension is not exceed 125% of the assigned interval (1.25 x 31 days = 38; i.e., grace period = 7 days). Per GAP-SAT-01, ST is 'overdue' if not completed by the end of the grace period. A is incorrect - plausible if candidate confuses 'overdue' with 'past due' (i.e., 115% of interval). C is incorrect - plausible if candidate thinks the 25% grace period for this monthly is 8 days, vice 7 days. D is incorrect - Just another date to balance out the intervals between each of the 4 choices; not really plausible in and of itself.

Proposed references to be provided to applicants during examination: None Technical Reference(s): TS 4.0.1, GAP-SAT-01 K/A: I Generic 2.1.10 2.1.10., Knowledge of conditions and limitations in the Facility License 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Group #

K/A # 2.4.5 2.4.5 Importance Rating 2.9 3.6 Proposed Question: RO 098, SRO 098 The plant is operating at 100% power, with the following:

ERV 113 sticks open

Efforts to close the ERV have failed

Reactor is manually scrammed

EOP-2 is entered on RPV low level

ERV 113 closes and is confirmed closed

Torus temperature is 800 F and rising slowly

CRS directs the Control Room 'E' to place torus cooling in service Which one of the following identifies the procedure used for placing a loop of torus cooling in service?

A. N1-EOP-4, Primary Containment Control B. N1-EOP-1, Attachment 16, Torus Cooling C. N1-ARP-L1-4-4, Drywell - Torus Temp High D. N1-OP-14, Containment Spray Proposed Answer: D Justification: D correct - Off-Normal Section H.0 must be used when EOP-4 has not been entered. Given these conditions, only EOP-2 has reason to be entered. No EOP-4 entry conditions exist. With ERV closed, and torus temp rising slowly thru 80 F, temp will turn before 85 F. If torus cooling is to begin, now, it must be aligned lAW the OP. NB incorrect - EOP-4 not yet entered;can't get to EOP-1, Att.16 except thru EOP-4. C incorrect- This ARP monitors DW/Torus air space temps, not torus water temp.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Ops Manual NMP1 EOP User's Guide K/A: Generic 2.4.5 2.4.5. Knowledge of the organization of the operating procedures network for normal, I abnormal, and emergency evolutions 10 CFR Part 55 Content: 55.41 (10) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Group #

K/A # 2.4.27 2.4.27 Importance Rating 3.0 3.5 Proposed Question: RO 099, SRO 099 A plant SCRAM has occurred, with the following:

An electrical fire is burning in Battery Room 12

Reactor Pressure is being controlled 700 to 800 psig on the EC's Which one of the following identifies the preferred instruments to monitor Reactor Pressure and EC Water Level?

REACTOR PRESSURE EC WATER LEVEL A.36-31A 121/122 B.36-31A 111/112 C.36-32A 121/122 D.36-32A 111/112 Proposed Answer: B Justification: B is correct- per table 9.3. A/C/D are incorrect-different combinations of either non-preferred reactor press instrument or EC water level instruments or both.

Proposed references to be provided to applicants during examination: SOP-9, FIRE IN PLANT, Flowchart I-Technical Reference(s):

K/A: I Generic 2.4.27 N1-SOP-9 2.4.27 Knowledge of fire in the plant Drocedure.

10 CFR Part 55 Content: 55.41(10) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Group #

K/A # 2.4.22 2.4.22 Importance Rating 3.0 4.0 Proposed Question: RO 100, SRO 100 An override in the level control leg of EOP-3 provides direction to terminate and prevent injection to the RPV except CRD and Boron.

Which one of the following is the basis for this action?

A. Reduce power to within bypass valve capacity B. Achieve faster reactor shutdown by concentrating the liquid poison C. Prevent instabilities and to reduce reactor power D. Ensure that HCTL will not be exceeded before liquid poison can be initiated Proposed Answer: C Justification: C correct - per the EOP bases A incorrect - Not a consideration (bypass valves may be unavailable) B incorrect - Not a consideration D incorrect - Not an assured response Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-ODP-PRO-0305 K/A: Generic 2.4.22 2.4.22: Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations 10 CFR Part 55 Content: 55.41(5) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: Region I Date: September 30, 2002 Facility/Unit: Nine Mile Point Unit 1 License Level SRO Reactor Type: GE BWR-2 Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent.

Examination paper will be collected six hours after the examination starts.

Applicant Certification All work done on this examination is my own. I have neither nor received nor received aid.

Applicant's Signature Results Examination Value: 100 Points Applicant's Score Points Applicant's Grade Percent NUREG-1021, Revision 8, Supplement 1

Examination Outline Cross-reference: Level SRO Tier # 1 Group# 1 K/A # 295017 Importance Rating 4.2*

Proposed Question: SRO 013 The plant is operating at 100% power, with the following:

H1-1-8, STACK GAS MONITORS HIGH RADIATION, alarms

C062 OGESMS RADIATION LEVEL HIGH is in alarm

The OGESMS monitor on J panel is reading just above the alarm setpoint.

Which one of the following describes the action required based on the Offsite Release?

A. E-plan entry is required B. Place RBEVS in service C. Offsite Release LCO entry is required D. Have Chemistry calculate the actual Offsite Release Proposed Answer: D Justification: D correct - LCO release rate may be being exceeded. Chemistry sample and calculation is required to make such determination A incorrect - Not a confirmed alarm AND not near E-plan level C incorrect- alarm is not confirmed B incorrect- referenced action in ARP but only if an initiating condition is met Proposed references to be provided to applicants during examination: None Technical Reference(s): ARP-HI-1-8 K/A: 295017 High Off- AA2.01 - Ability to determine and interpret the following as they apply site Release Rate to High Off-site Release Rate: Off-site release rate 10 CFR Part 55 Content: 55.41(7,12) X 55.43 (4)

Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 1 Group# 1 K/A # 295017 Importance Rating 3.7 Proposed Question: SRO 014 The plant is operating at 100% power, with the following:

Offgas radiation monitoring is unavailable

Recirc Pump #13 speed controller fails high

Recirc Pump #13 is manually tripped

The plant stabilizes at 80% power

Plans are to return to 100% power Which one of the following describes the potential risk of not restoring Offgas radiation monitoring operability?

A. Gross fuel failure detection and MSIV closure capability will be impaired B. Buildup of fission product activity in station systems could go undetected C. Child thyroid doses at the site boundary could exceed 10CFR limits during normal operations D. Child thyroid doses at the site boundary could exceed 10CFR limits on a steam line break Proposed Answer: B Justification: B correct - Main basis for Offgas Rad Monitor operability, specified on page 312 of tech specs. A incorrect- Done manually based on MSL Rad monitor reading 3.75 X NFPB. C incorrect - monitored by stack monitor, not Offgas; rates would be exceeded if offgas were bypassed. D incorrect - Basis for MSIV closure at 3.75 X NFPB on MSL RAD Monitor Proposed references to be provided to applicants during examination: TS 3.6.14 and 3.6.15 Technical Reference(s): Tech spec bases page 312, TS Table 3.6.14-2 K/A: 295017 High Off- 2.2.25 - Knowledge of bases in technical specifications for limiting I site Release Rate conditions for operations and safety limits 10 CFR Part 55 Content: 55.41(7,12) X 55.43 (2,4) 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 1 Group # 1 K/A # 295023 Importance Rating 4.6 Proposed Question: SRO 015 The plant is operating at 100% power, with the following:

RBEVS has automatically initiated

Valid high radiation signals exist Which one of the following describes the bases for making an emergency classification?

A. Spent Fuel Handling Accident B. Radwaste Handling Accident C. Loss of Reactor Coolant System D. Loss of Primary Containment Proposed Answer: A Justification: A correct - Classification bases for EAL 1.4.2. B incorrect - Does not result in RBEVS actuation. CID incorrect - Can result in increased radiation levels, but not for making an emergency declaration based on provided stem conditions.

Proposed references to be provided to applicants during examination: EAL Flowchart Technical Reference(s): EPMP-EPP-0101 K/A: 295023 Refueling AA2.05 - Ability to determine and interpret the following as they apply Accidents to Refueling Accidents: Entry conditions of emergency plan 10 CFR Part 55 Content: 55.41 55.43 (4,7) X 55.45 (11) X Question Source: Bank#

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None 1-02-094

Examination Outline Cross-reference: Level SRO Tier # 1 Group # 1 K/A # 295030 Importance Rating 3.8 Proposed Question: SRO 020 Technical Specifications requires a normal plant shutdown when all 3 of the following conditions exist for more than 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

Torus downcomer submergence is below the minimum allowable

Torus water temperature is above 850 F 0 Reactor coolant temperature is above 2150 F Which one of the following identifies the minimum allowable downcomer submergence and describes the basis for the 2150 F value?

MINIMUM SUBMERGENCE BASIS FOR 2150 F A. 3.5 feet An ERV discharge will not result in a damaged tailpiece B. 4.25 feet An ERV discharge will not result in a damaged tailpiece C. 3.5 feet A blowdown will not challenge the torus design pressure limit D. 4.25 feet A blowdown will not challenge the torus design pressure limit Proposed Answer: C Justification: C is correct - TS 3.3.2 LCO specifies the LCO and 24 action when > 215 F. Min allowable submergence is 3.5', max allowable is 4.25'. Basis for LCO justifies the 215 F value. BID are incorrect - 4.25 is max allowable submergence. A is incorrect - plausible if candidate confuses the min torus level required for ensuring adequate downcomer submergence with a high torus level that challenges the ERV Tailpipe Level Limit of the EOP Bases.

Proposed references to be provided to applicants during examination: None Technical Reference(s): Ts 3.3.2, with Bases; NI-OP-14, D.2.0 P/L K/A: 295030 Low 2.1.32 - Ability to explain and apply system limits and precautions Suppression Pool Water Level 10 CFR Part 55 Content: 55.41(5,10) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level SRO Tier # I Group # I K/A # 295003 Importance Rating 4.1 Proposed Question: SRO 021 The plant is operating at 100% power, with the following:

"* PB 103 is deenergized for bus repairs

"* A transformer fault on reserve transformer 101 N occurs After 15 minutes, which one of the following describes the EAL that should be declared?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Proposed Answer: B Justification: B correct- power is lost to T101N and T101S is not available with PB 103 OOS. No backfeed alignment has been made, only DG 102 is available [EAL 6.1.3]. A incorrect- conditions are met but higher classification (6.1.3) should be declared. CID incorrect - DG 102 is available and PB 102 should be energized Proposed references to be provided to applicants during examination: EAL Flowchart Technical Reference(s): EPIP-EPP-01 K/A: 295003 Partial or Generic 2.4.41 - Knowledge of the emergency action level threshold Complete Loss of and classification AC Pwr 10 CFR Part 55 Content: 55.410 55.43 (5) X 55.45()

Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295024 Importance Rating 4.0 Proposed Question: SRO 024 A loss of coolant accident has occurred, with the following:

Containment Spray Pump 111 is clearance tagged out of service

PB 103 is deenergized due to a fault

3 ERVs are open

All available systems are injecting

RPV water level is -70 inches and stable

Torus pressure is 18 psig and slowly lowering

Torus level is 12 feet and stable

Containment Spray Pump 112 has just tripped on over-current Which one of the following is required?

A. Vent the Torus per EOP-4.1, Section 2 B. Cool the Torus per EOP-1, Attachment 16 C. Supply raw water to Containment Spray Loop 11 sparger D. Secure the Containment Spray lineup and re-start all Drywell Cooling Fans Proposed Answer: C Justification: C is correct - Per EOP-4, Step PCP-4, can use external source (raw water) to spray as long as we stay below PCPL. A is incorrect - conditions to vent PC do not exist. B is incorrect - requires a Containment Spray pump; not available. D is incorrect - no direction for this given in EOPs, with these conditions.

Proposed references to be provided to applicants during examination: EOP Flowcharts Technical Reference(s): N1-EOP-4 K/A: 295024 HIGH 2.1.23. HIGH DRYWELL PRESSURE: Ability to perform specific DRYWELL system and integrated plant procedures during different modes of PRESSURE plant operation cooling: Plant-Specific 10 CFR Part 55 Content: 55.41 55.4L3 (5) X 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 1 Group# 2 K/A # 295001 Importance Rating 3.3 Proposed Question: SRO 027 The plant has just scrammed (spurious) from 100% power, with the following:

TIME CONDITION 0000 Complete loss of offsite power 0000 PB 101 deenergizes and locks out on a fault 0100 Line 4 is restored to in-house loads 0115 Commenced plant cooldown 0400 Reactor coolant temperature is 350 OF 0700 Reactor coolant temperature is 200 OF Which one of the following identifies the EARLIEST time a Recirc Pump can be started per plant procedures?

A. 0100 B. 0115 C. 0400 D. 0700 Proposed Answer: D Justification: D is correct - per Ni-OP-1, H.13.0. All 5 recirc pumps are lost at time = 0000. Power is not restored to any recirc pumps until 60 minutes later (time = 0100). To avoid thermal cycling of RPV lower head a recirc pump can only be re-started within 30 minutes of the total loss of forced flow, else, it must wait until cold shutdown (< 212 F); i.e., time = 0700. A is incorrect - although power is now available to start a pump, OP-1 does not allow it, for the reasons cited above. BIC are incorrect - plausible to the SRO candidate who either cannot recall the thermal cycling concern and restart criteria, or was never aware of it.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-t, H.13.0 K/A: 295001 Partial or AK2.01, Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS Complete Loss of OF FORCED CORE FLOW CIRCULATION and the following: Recirculation system Forced Core Flow Circulation 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 1 Group # 2 K/A # 295018 Importance Rating 3.9 Proposed Question: SRO 033 A core off-load is in progress, with the following:

A loss of all RBCLC pumps has occurred

Spent Fuel Pool temperature is 131°F and rising

Determination of Time to 140OF is required

Decay Heat Production (DHP) is 6.42 MBTU/Hr

Decay Heat Removal (DHR) is 0 MBTU/Hr Which one of the following identifies the Time to 140°F?

A. 3.15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months B. 3.67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months C. 8.49 hours5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months D. 9.53 hours6.134259e-4 days 0.0147 hours 8.763227e-5 weeks 2.01665e-5 months Proposed Answer: C Justification: C correct - per Attachment 2 of ODP-OPS-01 08 calculation. A incorrect plausible if examinee chooses "k"value for SFP only (gates installed). B incorrect - plausible if examinee chooses "k" value for Cavity only (gates installed). D incorrect -plausible if examinee chooses "k" value for both the Cavity and the SFP Time to 140F.

Proposed references to be provided to applicants during examination: Nl-ODP-OPS-0108 Technical Reference(s): Ni-ODP-OPS-0108 K/A: 295018 Partial or Generic 2.4.9 - Knowledge of low power/shutdown implications in Total Loss of accident (eg. LOCA or loss of RHR) mitigation strategies CCW 10 CFR Part 55 Content: 55.41(7,10) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 1 Group # 2 K/A # 295028 Importance Rating 4.2 Proposed Question: SRO 036 A seismic event has occurred, with the following:

"* Complete loss of offsite power

"* Steam leak inside the drywell

"* PB 103 is deenergized due to a fault

"* All SOP-5 actions are complete

"* RPV level is 44" and lowering at 2" per minute

"* RPV pressure is 720 psig and lowering

"* Drywell temperature is 250 OF and rising at 5 OF per minute

"* Drywell pressure is 9 psig and rising

"* Torus temperature is 104 OF 0 Torus pressure is 8 psig and rising Which one of the following describes the next required action?

A. Initiate Containment Spray B. Initiate Torus Cooling C. Install Core Spray IV jumpers D. Perform an RPV blowdown Proposed Answer: A A is correct - DW temp rate of rise will drive it beyond 300 F much sooner than level rate of drop will reach 84". SRO must give higher priority to spraying that to RPV level. B is incorrect - Although torus temp is high, spray lineup will be complicated by torus cooling, and will take precedence. C is incorrect - With RPV level not recovering, core spray is required for makeup. D is incorrect - Conditions do not yet exist to warrant a blowdown.

Proposed references to be provided to applicants during examination: EOP Flowcharts Technical Reference(s): Ni -EOP-4 K/A: 295028 High EA2.04 - Ability to determine and interpret the following as they apply to High Drywell Drywell Temperature: Drywell pressure Temperature 10 CFR Part 55 Content: 55.41(7) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 1 Group# 2 K/A # 295034 Importance Rating 4.2 Proposed Question: SRO 039 The plant is stable following a scram, with the following:

Several Fire alarms initiate in the Reactor Building

RBEVS initiates on high ventilation radiation

Both Reactor Building vent rad monitors are reading similar

Reactor pressure begins to lower Which one of the following describes the event, as it would be applied to emergency procedures?

A. A fire has occurred involving radioactive materials B. A fire has occurred along with radiation detector failures C. A non-primary system is discharging D. A primary system is discharging Proposed Answer: D Justification: D correct -high temp/humidity from steam will cause fire alarms, elevated vent rad caused by gaseous and particulate from reactor coolant AND only a primary system would cause a corresponding reduction in RPV pressure A/BIC are incorrect -none of these events would account for the lowering reactor pressure Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-ODP-PRO-0305, page 221 K/A: 295034 EA2.02 - Ability to determine and interpret the following as they apply Secondary to Secondary Containment Ventilation High Radiation: Cause of high Containment radiation levels Ventilation High Radiation 10 CFR Part 55 Content: 55.41(7,10) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Examination Outline Cross-reference: Level SRO Tier # I Group # 2 K/A # 295034 Importance Rating 3.6 Proposed Question: SRO 040 A plant transient has occurred, with the following:

"* Reactor Building (RB) Vent radiation rose to 8 mr/hr

"* Multiple temperature and fire alarms were received in the RB Initial actions are taken to stabilize the plant and limit leakage into the RB, resulting in:

"* Reactor Building Vent radiation lowers to 3 mr/hr

"* Further access into the RB is required to isolate the leaking components Which one of the following describes the required action and bases for that action?

ACTION BASES A. Verify RB isolated/EVS running Provide for filtered release B. Verify RB isolated/EVS running Maximize RB air flow C. Restart RB Ventilation/Shutdown EVS Provide for filtered release D. Restart RB Ventilation/Shutdown EVS Maximize RB air flow Proposed Answer: D Justification: D correct -with RB vent rad <5 mr/hr step SC-2 requires restart of RB vent to maximize RB air flow A incorrect - this is the action/reason for RB vent rad > 5 mr/hr BIC are incorrect -combinations of actions/reasons that one or both are not appropriate for the conditions Proposed references to be provided to applicants during examination: N1-EOP-5 Technical Reference(s): N1 -ODP-PRO-0305/ page 224, N1 -EOP-5/step SC-2 K/A: 295034 Generic 2.4.18 - Knowledge of the specific bases for EOPs Secondary Containment Ventilation High Radiation 10 CFR Part 55 Content: 55.41(7,10) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History I Question Cognitive Level: I Memory or Fundamental Knowledge I

FComprehension or Analysis X LOD: 3 Comments:

Examination Outline Cross-reference: Level SRO Tier # 1 Group # 2 K/A # 600000 Importance Rating 3.6 Proposed Question: SRO 042 The plant is at 100% power, with the following:

DG 102 is running for monthly testing

A loud bang is heard and heavy smoke fills the DG room

The DG trips and the smoke dissipates

10 minutes after the loud noise, the room is entered and the DG is observed to be damaged but no fire is observed Which one of the following identifies when the NRC is required to be notified of this event?

Notify the NRC within...

A. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months D. 30 Days Proposed Answer: A Justification: A correct - EAL 8.2.2 is met as DG 102 is unavailable as a result of a fire/explosion B/CID are incorrect-any emergency declaration requires NRC notification within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Proposed references to be provided to applicants during examination: EAL Flowchart, 10CFR50.72 and 73 [NOTE - CANDIDATE TO BE GIVEN ALL OF 10CFR PART 50]

Technical Reference(s): EPIP-EPP-01 K/A: 600000 Plant Fire 2.4.30 - Knowledge of which events related to system On Site operations/status should be reported to outside agencies 10 CFR Part 55 Content: 55.410 55.43 (5) X 55.45()

Question Source: Bank#

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 2 Group # 1 K/A # 212000 Importance Rating 2.4.21 Proposed Question: SRO 048 The plant is operating at 100% power, with the following:

Required surveillances for MSIV 112 position input to RPS are determined as due

All other MSIV position indication surveillances are up to date

The surveillances will take 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months to complete Which one of the following is required?

A. Complete the required surveillances B. Close MSIV 112 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months C. Trip RPS 11 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. Trip RPS 12 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Proposed Answer: A Justification: A correct - per TS table 3.6.2.a, note (o), required channels can be taken out of service for surveillances for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if redundant parameter monitoring is provided in the trip system. With MSIV 111 position indication operable and only one input is required to trip RPS 11 the requirement is met. B incorrect - required action if MSIV inoperable combined with time allowed to conduct surveillance CID are incorrect - actions required if redundant trip capability in RPS trip system not functional and/or wrong trip system Proposed references to be provided to applicants during examination: Table 3.6.2.a Technical Reference(s): Table 3.6.2.a, 01-OPS-001-212-1-01/page 16 K/A: 212000 RPS 2.4.21: Knowledge of the parameters and logic used to assess the status of safety functions including:

1. Reactivity control

2. Core cooling and heat removal

3. Reactor coolant system integrity

4. Containment conditions

5. Radioactivity release control.

10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3

Examination Outline Cross-reference: Level SRO Tier # 2 Group # 1 K/A # 207000 Importance Rating 4.0 Proposed Question: SRO 055 The plant is operating, with the following:

"* Differential pressure transmitter 36-06A has failed its surveillance calibration check

"* The failure is non-conservative Which one of the following describes the action required by Technical Specifications?

A. Restore operability of the affected system within 7 days and perform the required additional surveillances on the remaining operable system until then.

B. Trip the inoperable channel in one trip system within one hour and trip the inoperable channel in the other trip system within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. Trip the inoperable channel in both trip systems within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or be in cold shutdown within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months thereafter.

D. Commence a normal shutdown within one hour and be in cold shutdown within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

Proposed Answer: B Jusiif~atlon jusulTIcatlon: B is correct - per TS 3.6.2c(f). Since 36-06A inputs to both trip systems (EC System 11 and System 12), the second half of 3.6.2c(f) applies. A is incorrect - plausible if candidate goes directly to EC LCO 3.1.3 and assumes that 36-06A makes one of the EC's inop, with a resulting 7-day LCO and increased surveillances on the other EC. C is incorrect - plausible if candidate is uncertain of the EC high steam flow isolation logic arrangement, and the fact that the single failed d/p transmitter feeds both trip systems. Instead, candidate may determine that the first half of the note (3.6.2c(f)) applies. D is incorrect - plausible if the candidate is very unsure of himself with respect to TS operability for instrumentation and/or the ECs. The "one and ten shutdown" is a very common Unit 1 requirement, elsewhere; therefore, provides ciood distraction.

Proposed references to be provided to applicants during examination: Tech Specs and P&ID C-18017-C Technical Reference(s): TS 3.1.3; 3.6.2; PID C-18017-C K/A: 207000 Isolation 2.1.12, Ability to apply technical specifications for a system (Emergency)

Condenser 10 CFR Part 55 Content: 55.41 55.43 (5) X 55.45 Question Source: Bank#

Modified Bank #

New NEW History

Examination Outline Cross-reference: Level SRO Tier # 2 Group # I K/A # 223002 Importance Rating 4.0 Proposed Question: SRO 056 An unisolable RWCU system leak to the Reactor Building is in progress, with the following:

"* A failure to scram has occurred

"* Reactor power is 15%

"* RPV water level is being intentionally lowered

"* Multiple Reactor Building Area water levels are >5 feet

"* Only 2 ERVs can be opened and are open

"* Reactor pressure is 1080 psig and rising slowly

"* Torus pressure is 2 psig and steady Which one of the following identifies the MSIV isolations that should be defeated and the required pressure control band?

MSIV Isolations to defeat With TBV's, establish and maintain RPV pressure...

A. Lo-Lo RPV level, only 500 - 1080 psig B. Lo-Lo RPV level, only about 75 psig C. All isolations 500 - 1080 psig D. All isolations about 75 psig Proposed Answer: D Justification: D is correct - Crew is in EOP-3, 5 and 8. EOP-5 requires blowdown for RB area water levels. 15% power is generating more steam pressure than what 2 ERVs can relieve; therefore, RPV pressure is rising slowly. EOP 8 assumes pressure control. EOP-8, Step 15 directs use of alternate blowdown systems if <3 ERVs are open and RPV pressure >72 psig. EOP-8, Table 0, directs use of EOP-1, Att.1 to override all MSIV isolations. AIB are incorrect - Would override the Io-lo level isolation, only, if pressure control were still in EOP-3; it is not. C is incorrect - The only place in the EOPs that directs override of all MSIV isolations is in EOP-8, Table 0, and only for the purpose of supporting EOP-8, Step 15. The objective of that step is to maintain RPV pressure <72 psig above torus pressure.

Dr~n tm rafarenc, tn hb nrnvided to applicants durina examination: EOP Flowcharts Technical Reference(s): EOP-3, 5, 8 K/A: 223002 Generic 2.4.6 - Knowledge of symptom based EOP strategies PCIS/Nuclear Steam Supply Shutoff 10 CFR Part 55 Content: 55.41(10) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 12 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 2 Group# 1 K/A # 211000 Importance Rating 4.0 Proposed Question: SRO 057 The plant is operating, with the following:

"* The Liquid Poison (LP) system temperature indicating controller and alarm temperature switches were found to have a +30 OF error on their last calibration

"* LP tank volume is 1341 gallons

"* LP tank concentration is 15.9%

"* LP tank temperature is reading 90 OF Which one of the following identifies the operability status of the Liquid Poison system, and describes the required action?

A. Operable. Direct Chemistry to dilute the LP tank concentration to 12.5%.

B. Inoperable. Be shutdown within one hour and less than 110 psig within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

C. Operable. Restore tank temperature per Section H of N1-OP-12, Liquid Poison.

D. Inoperable. Commence a normal orderly shutdown within one hour.

Proposed Answer: D Justification: Dis correct - Per TS 3.1.2.d, LP is inop and an orderly shutdown must be commenced. A/C are incorrect - LP is inop. B is incorrect - A rapid shutdown and depressurization to less than saturation is not required.

Proposed references to be provided to applicants during examination: Tech Specs Technical Reference(s): T.S. 3.1.2, N1-OP-12 K/A: 211000 Standby 2.1.33, Ability to recognize indications for system operating Liquid Control parameters which are entry-level conditions for technical System (Liquid specifications.

Poison) 10 CFR Part 55 Content: 55.41(7,10) X F55.43 (5) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments: None

Proposed Question: SRO 069 The plant is operating, with the following:

Three Recirc loops are operating, with two isolated

Reactor power is 91%

power level?

Which one of the following describes the bases for not continuing operation at this A. LOCA analyses yield higher peak clad temperatures in this configuration.

is B. The margin between the MCPR Safety Limit and the Limiting Safety System Setting reduced in this configuration.

C. MCPR analyses show higher peaking due to flow unbalance in this configuration.

D. Cold start of a loop could exceed the MCPR limit.

Proposed Answer: A Justification: A correct - TS 3.1.7 bases identifies smaller coastdown inventory as reducing LOCA cooling and raising peak clad temperature B incorrect

- nonsense unless candidate doesn't understand interrelations C incorrect- nonsense unless candidate doesn't understand flow baffle configuration D incorrect - This is 4 loop unisolated basis.

Proposed references to be provided to applicants during examination: None Technical Reference(s): TS 3.1.7 bases (pg 73)

K/A: 202001 Generic 2.2.25 - Knowledge of the bases in technical specifications Recirculation for limiting conditions for operations and safety limits 10 CFR Part 55 Content: 55.41(7) X 55.43 (2) X 55.45 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 3 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 2 Group# 3 K/A # 201003 Importance Rating 3.3 Proposed Question: SRO 081 Control rod movements are in progress, with the following:

The third of ten rods on a control rod move sheet is being repositioned

The target position for the control rod is position 14 from position 18

The control rod is inserted to position 12 Which one of the following describes the situation where this control rod movement error would require immediate notification of both Reactor Engineering and the General Supervisor Operations, per N1-OP-5?

A. CRS discovers the error while the next rod is being moved B. Rod positioner (RO) discovers the error while moving the tenth rod C. Independent Verifier discovers the error just prior to the next rod being moved D. STA discovers the error while the first rod on the next move sheet is being moved Proposed Answer: D I

D is correct - discovery this late is a mis-positioned control rod event, requiring GSO notification and RE notification. AIBIC are incorrect- per OP-5, H.4.1, control rod mis-position is not reportable if caught during use of the control rod move sheet.

Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-5, H.4 K/A: 201003 Control Generic 2.1.14 - Knowledge of system status criteria which require Rod and Drive the notification of plant personnel Mechanism 10 CFR Part 55 Content: 55.41(10) X 55.43 (5) X 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

r. Level 1 r rLJ

,-.r-,r omu I I

Examination Outline Cross-reference: Level Tier # 3 Group #

K/A # Generic 2.1.11

_ Ai I I

Importance Rating I _

3.8 1

Proposed Question: SRO 084 A plant startup is in progress, with the following:

Reactor power is 85%

Core Flow is 77%

Recirc Flow control is in automatic

MCPR is determined to be 1.50 Which one of the following describes when the first action is required?

A. 15 Minutes B. 30 Minutes C. 2 Hours D. No action is required Proposed Answer: A Justification: A is correct- per the COLR the required value of MCPR at 77%

flow/AUTO is 1.45 x 1.1=1.59, with MCPR not in compliance action is required to be initiated within 15 minutes B is incorrect- not a referenced time limit of MCPR spec C is incorrect- is the second action time of MCPR spec D incorrect- action is required but if Kf is not applied would be the chosen answer Proposed references to be provided to applicants during examination: TS 3.1.7, COLR Technical Reference(s): TS 3.1.7.c, COLR 2.1/Fig 25 2.1.11: Knowledge of less than one hour technical specification action K/A: I Generic 2.1.11 statements for systems 10 CFR Part 55 Content: 55.41 55.43 (2) X 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 3 Group #

K/A # Generic 2.1.12

_Importance Rating 4.0 Proposed Question: SRO 085 The plant is operating at 100% power following a refueling outage, with the following:

A review of startup valve line-ups and logs is being conducted

Position of valve 05-10, "BV-EC VENT TO TORUS MANUAL AFTER (05-07)" is questioned

Investigation reveals that valve 05-10 is stuck closed Which one of the following describes the action required to support continued power operations?

Valve 05-10 must be opened within...

A. 1 Hour B. 7 Days C. 14 Days D. 30 Days Proposed Answer: C Justification: C is correct - the vent path to the Torus for both EC's is isolated requiring restoration of at least one within 14 days B incorrect - This would be chosen if one EC loop was believed INOP A is incorrect This would be required if both EC's were INOP or the valve could not be opened within 14 days D is incorrect- This would be the time allowed if only one EC were affected Proposed references to be provided to applicants during examination: TS 3.1.3 Technical Reference(s): TS 3.1.3, DWG C-18017-C K/A: I Generic 2.1.12 2.1.6: Ability to apply technical specifications for a system 10 CFR Part 55 Content: 55.41 55.43 (2,5) X 55.45 Question Source: Bank #

Modified Bank#

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 3 Group #

K/A # 2.2.28 Importance Rating 3.5 Proposed Question: SRO 088 A core shuffle is in progress, with the following:

A fuel bundle is inserted into the correct core location

The fuel bundle is mis-oriented Which one of the following identifies whose permission is required before re-orienting the fuel bundle?

A. SSS B. Operations Manager C. Plant General Manager D. NRC Proposed Answer: B Justification: B correct- per FHP-25, section 3.1, 41.11. AICID - not per procedure.

Proposed references to be provided to applicants during examination: N1-FHP-25 Technical Reference(s): N1-FHP-25/page 4/Att.4 K/A: I Generic 2.2.28 2.2.28 - Knowledge of new and spent fuel movement procedures 10 CFR Part 55 Content: 55.41 55.43 (7) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 3 Group #

K/A # Generic 2.2.5 Importance Rating 2.7 Proposed Question: SRO 089 The plant is operating at 100% power, with the following:

An extremely cold weather system has settled in the area

Engineering has determined that the operability of a safety related valve is threatened by the cold environment Engineering recommends, via an Engineering Reply (ER), installation of portable heating to maintain the required environmental conditions Which one of the following describes the appropriate controls to maintain the valve operable?

A. The ER provides the necessary approvals and the heater may be installed B. A Work Order/Job Traveler is all that is needed to approve heater installation C. A Temporary Change Package needs to be issued to install the heater D. A Design change must be made to increase the available heating in the room Proposed Answer: C Justification: C is correct- This is the process specified for making temporary changes to plant design A incorrect- Ref 3.2.2.a prohibits use of an ER in this manner B incorrect- This activity constitutes a necessary change to plant design to maintain equipment operability. While an ACR may be needed to accomplish installation, the ACR is not acceptable to document a design change which installs support equipment essential to maintaining operability of safety related plant equipment. D incorrect- Ultimately, a permanent design change may be appropriate, however a temporary change has been evaluated in the ER as acceptable Proposed references to be provided to applicants during examination: None Technical Reference(s): NIP-CON-01 K/A: Generic 2.2.5 2.2.5: Knowledge of the process report for making changes in the facility as described in the safety analysis 10 CFR Part 55 Content: 55.41 55.43 (3) X 55.45 Question Source: Bank #

Modified Bank #

New NEW History Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level SRO Tier# 3 Group #

K/A # Generic 2.3.1

__Importance Rating 3.0 Proposed Question: SRO 092 Per S-RAP-RPP-0703, Authorization to Exceed Administrative Dose Limits, which one of the following individuals provides the final authorization required to extend a worker's annual Nine Mile dose limit to 1600 mrem?

A. Vice President- Nuclear B. Plant Manager C. Unit RP Manager D. RP Supervisor Proposed Answer: B Justification: B is correct- The Plant Manager is designated as the final authority A/CID incorrect- not the referenced final approver for dose above 1500 but below 2000 Proposed references to be provided to applicants during examination: None Technical Reference(s): S-RAP-RPP-0703 K/A: Generic 2.3.1 2.3.1. Knowledge of 10 CFR 20 and related facility radiation control I requirements 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New History 2000 U2 NRC Exam: 2000 NMP1 Audit exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis LOD: 2 Comments: None

Examination Outline Cross-reference: Level SRO Tier # 3 Group #

K/A # Generic 2.3.9 Importance Rating 3.4 Proposed Question: SRO 093 The plant is operating at 100% power, with the following:

Drywell Pressure is 1.8 psig

Drywell venting using 11 RBEVS fan is in progress

The Drywell vent and purge valves had been open for 75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months before this venting

The current venting has been in progress for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Which one of the following identifies how long Drywell venting can continue?

Drywell venting can continue for another...

A. 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months B. 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months C. 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months D. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Prtr1 Answr A AA)eQ. 4- 7ksc%rA §cAtA ___

Justification: A is correct - OP-9, limits vent and O~urge valve operation to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per year under these conditions, with 75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months already used plus 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for the current vent that leaves 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months to go B is incorrect as it does not account for the current vent time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months CID are incorrect based on the RBEVS LCO entry for the vent but is not limiting Proposed references to be provided to applicants during examination: None Technical Reference(s): N1-OP-9/H.1 Note 1 K/A: I Generic 2.3.9 2.3.9: Knowledge of the process for performing a containment purge 10 CFR Part 55 Content: 55.41(7) X 55.43 55.45 Question Source: Bank #

Modified Bank #

New X History Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 3 Comments: None

Proposed Question: SRO 094 One hour ago, a reactor scram occurred, with the following:

RPV Level dropped to 35 inches and is now 70 inches and stable

RPV Pressure dropped to 900 and is now 970 psig and stable

Torus water level is 11 feet and stable

EC 11 CANNOT be isolated

EC 11 Vent radiation monitor is in alarm Chemistry reports that the off-site dose projection is 1080 mRem/hr TEDE.

Which one of the following describes the required actions?

A. Perform an RPV blowdown.

B. Bypass ADS C. Lower RPV pressure to 450 to 550 psig D. Open ERVs to lower RPV pressure to 965 psig Proposed Answer: A Justification: A is correct- Rad Release EOP for GE and primary system discharging outside containment. B incorrect- The vents are on the main steam line to the EC. Opening the vents will be ineffective in reducing the release. C incorrect- This is an appropriate action before the General Emergency level is reached. Once the General Emergency level is reached, the required action is to open 3 ERVs (RPV blowdown). D incorrect- An EOP-3 action; there is no ATWS in these stem conditions.

Proposed references to be provided to applicants during examination: EOP Flowcharts Technical Reference(s): N1-EOP-6, Radioactivity Release Control K/A: Generic 2.3.11 2.3.11. Ability to control radiation releases.

10 CFR Part 55 Content: 55.41(10,12) X 55.43 (5) X 55.45 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X LOD: 2 Comments:

ML023240459

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