Text

September 2002 Exam 50-395/2002-301 Administrative Documents
	ML023100219
Person / Time
Site:	Summer
Issue date:	09/17/2002
From:	Ernstes M; Division Reactor Projects II
To:	Byrne S; South Carolina Electric & Gas Co
References
	50-395/02-301 50-395/02-301
	Download: ML023100219 (130)
	v • d • e

VIRGIL C. SUMMER NUCLEAR STATION - EXAM 2002-301 50-395 SEPTEMBER 9 - 17, 2002 Administrative Documents (Yellow Paper) Exam Preparation Checklist...... Exam Outline Quality Checklist... Exam Security Agreement....... Administrative Topics Outline (Final)

..... ES-201-1->

.....

ES-201-2

.....

ES-201-3 I/

.....

ES-301-1 / Control Room Systems and Facility Walk-through Test Outline (Final)......................................

ES-301-2 Operating Test Quality Check Sheet... Simulator Scenario Quality Check Sheet Transient and Event Checklist........ Competencies Checklist.............

1 Written Exam Quality Check Sheet....

1 Written Exam Review Worksheet......

1 Written Exam Grading Quality Checklist 1 Post-Exam Check Sheet............

ES-301-3.

ES-301-4 /

.ES-301-5 /

.......... ES-301-6 /

. ES-401-7 --

. ES-401-9 /

ES-403-1 /

ES-501-1

ES-201 Examination Preparation Checklist Form Facility:

Summer Date of Examination: 9/9-13/02 Examinations Developed by:

Facility

/

NRC (circle one)

Chief Target Examiner's Date*

Task Description / Reference Initials-180 1. Examination administration date confirmed (C.1.a; C.2.a & b)

LM-120 2. NRC examiners and facility contact assigned (C.1.d; C.2.e)

LM-120 3. Facility contact briefed on security & other requirements (C.2.c)

LM-120 4. Corporate notification letter sent (C.2.d)

LM

[-90]

[5. Reference material due (C.1.e; C.3.c)]

LM-75 6. Integrated examination outline(s) due (C.1.e & f; C.3.d)

LM-70 7. Examination outline(s) reviewed by NRC and feedback provided LM to facility licensee (C.2.h; C.3.e)

-45 8. Proposed examinations, supporting documentation, and LM reference materials due (C.1.e, f, g & h; C.3.d)

-30 9. Preliminary license applications due (C.1.1; C.2.g; ES-202)

LM-14 10. Final license applications due and assignment sheet prepared LM (C.1.1; C.2.g; ES202)

-14 11. Examinations aroved by NRC supervisor for facility licensee LMV review (C.2.h; WSJ*.)

-14 12. Examinations reviewed with facility licensee (C.1.j; C.2.f & h; C.3.g)

LM-7 13. Written examinations and operating tests approved by LM NRC supervisor (C.2.i; C.3.h)

-7 14. Final applications reviewed; assignment sheet updated; waiver LM letters sent (C.2.g, ES-204)

15. Proctoring/written exam administration guidelines reviewed with LM-7 facility licensee and authorization granted to give written exams (if applicable) (C.3.k)

-7 16. Approved scenarios, job performance measures, and questions LM distributed to NRC examiners (C.3.i)

Target dates are keyed to the examination date identified in the corporate notification lette Toe are for planning purposes and may be adjusted on a case-by-case basis in coordination with the facility license Applies only to examinations prepared by the NR ES-201 Examination Outline Form ES-201-2 (R8,S1)

Quality Checklist Facili Summer Date of Examination:

9/17/02 Initials Item Task Description b*

a C a. Verify that the outline(s) fit(s) the appropriate model per ES-40 A 44ill c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topicsr /

"

E N

d. Assess whether the justifications for deselected or rejected K/A statements are appropriate*

.* t a. Using Form ES-30itm e serio, that tho scenarios deuplicts over the required number of tts normal evolutions, insteruent atd ompove tfare and maor transient SI b. Assess whether there are enough scenario sets (and spares) to test the projected number and M

mix of applicants in accordance With the expected crew composition and rotation schedule without compromising exam intaithe equre e

r oft can be tested using at least one new or significantly modified scenario, that no scenatos are duplicated from the applicants' audit test(s)*,

and scenarios wlnoberpated over successive day T. To the extent possible, assess whether the outake rconform(s) wth lee qualitative and Lks quantitative criteria specified on Form ES-301-4 an desrbed in Appendix.

a. Verify that:

(1) the aouline(s) contain(s) the required number of control room and in-plant tasks, W

(2) no more than 30% of the test material is repeated from the last NRC examination, (3)

o tasks are duplicated from the applicants' audit test(s), and T

no more tha 80% of any operating test 's taken directly from the licensee's exam bank b. Verify that:

(1) the tasks are distributed among the safety function groupings as specified in ES-301, (2) one task is conducted in a low-power or shutdown condition, Lot,

'

(3) 40% of the tasks require the applicant to implement an alternate path procedure, (4) one in-plant task tests the applicant's response to an emergency or abnormal condition, and (5) the in-plant walk-through requires the a plicant to enter the RC c. Verify that the required administrative topics are covered, with emphasis on performance-LI based activitie d. Determine if there are enough different outlines to test the projected number and mix of 0.4'

,

applicants and ensure that no items are duplicated on successive day.

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

,

approprite exam sectio E b. Assess whether the 10 CFR 55.41/43 and 55.45 sam ling is appropriat L N

E c. Ensure that K/A im ortance ratings (except for plant-specific priorities) are at least R L

A d. Check for du lication and overlap among exam section L e. Check the entire exam for balance of coverae LAIU f. Assess whether the exam fits the a, ro riate job level RO or SRO.

eledNte 12_)ýreDate a. Author b. Facility Reviewer (*)

c. NRC Chief Examiner (#)

d. NRC Supervisor M " A" -i

. +*

Note:

Not applicable for NRC-developed examinations Independent NRC reviewer initia items in Column c;" chief examiner concurrence require ES-201 Examination Securit Agreement Form ES-201-3 Pre-ExaminatLon 971/02 I acknowledge that I have acquired specialized knowledge about the NRC licensing examinations scheduled for theweek(s) of

as of the date of my signature. I agree that I will not knowingly divulge any Information about these examinations to any persons who have not been authorized by the NRC chief examiner. I understand that I am not to instruct, evaluate, or provide performance feedback to those applicants scheduled to be administered these licensing examinations from this date until completion of examination administration except as specifically notln authorized by the NRC.Furthermore, I am aware of the physical security measures and requirements (as documented in the facility licensee's procedures) and understand that violation of the conditions of this agreement may result in cancellation of the examinations and/ar an enforcement action against me or the facility licensee. I will immediately report to facility management or the NRC chief examiner any indications or suggestions that examination security may have been compromise.

Post-Examination To the best of my knowv$40. I did not divulge to any unauthorized persons any information concerning the NRC licensing examinations administered during the week(s) of know e

. From the date that I entered into this security agreement until the completion of examination administration, I did not instruct, evaluate, or pfvidb performance feedback to those applicants who were administered these licensing examinations, except as specifically noted below and authorized by the NR SIGNdATIIUR (2 DATE NOTE 24 of 24 NUREG-1021, Revision 8 UMI SIGNATURE (1)

ES-201 Examination Securit Aeement Form ES-201 -3 Pre-Examination 9/t2 I acknowledge that I have acquired specialized knowledge about the NRC licensing examinations scheduled for the week(s) of

/0 2 as of the date of my signature. I agree that I will not knowingly divulge any information about these examinations to any persons who have not been authorized by the NRC chief examiner. I understand that I am not to instruct, evaluate, or provide performance feedback to those applicants scheduled to be administered these licensing examinations from this date until completion of examination administration, except as specifically nottd below and authorized by the NRC.Furthermore, I am aware of the physical security measures and requirements (as documented in the facility licensee's procedures) and understand that violation of the conditions of this agreement may result in cancellation of the examinations and/or an enforcement action against me or the facility licensee. I will immediately report to facility management or the NRC chief examiner any indications or suggestions that examination security may have been compromised. Post-Examination To the best of my knowle, I did not divulge to any unauthorized persons any information concerning-the NRC licensing examinations administered during the week(s) of ZSS/ From the date that I entered Into this security agreement until the completion of examination administration, I did not instruct, evaluate, or provide performance feedback to those applicants who were administered these licensing examinations, except as specifically noted below and authorized by the NRC.

PRINTED NAME 1. J0e

--, A M" f_-e,

.....

1.

Cf a.Set -!! /"SA-NOTES:

24 of 24 NUREG-1021, Revision 8 DATE NOTE SIGNATURE (2)

ES-301 Administrative Topics Outline Form ES-301-1 (R8, S1)

Facility: _-_Summer r

Date of Examination: __9/9-13/02__

Examination Level (circle one):

RO / SRO Operating Test Number:

Administrative Describe method of evaluation:

Topic/Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions Conduct of Review of License Operator Status Report to determine Operations current active licenses. (Shift Turnover)

Conduct of Determine adequate shift manning Operations Equipment Evaluation of Surveillance Test results Control Radiation Determine personnel exposure limit for non-essential personnel Control Emergency Plan Classify an Emergency Plan Event

Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2 (R8, SI)

System / JPM Title a. Transfer to Cold Leg Reciriculation JPS-5 b. Loss of Intermediate Range Instrumentation JPS-029 Type Code*

DS LDS c. Stuck Rod JPS-043 DS d. Identify and isolate RCS leak to CCWS JPS-042 e. Response to imminent pressurized thermal shock JPS-93 f. Manually initiate Reactor Building Spray JPSF-019 DS Safety Function

7

8 NS 4P AS

g. Transfer in-service charging pump (NRC) JPSF-046 DAS2 a. Locally start an Emergency DIG during a loss of offsite power (with Failure of field to flash) JPPF-012 b. Control Room evacuation (duties of BOP operator)

(Modified JPPF-049)

c. Establish Demineralizer Water Alternate cooling to Charging Pumps (Failure of Chilled Water Supply)

(New JPPF-NRC)

DA

M

DAR

Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA ES-301

ES-301 Operating Test Quality Checklist Form ES-301-3 (R8, Si)

Facili Summer Date of Examination: 9/9-13/02 0 eratin Test Number:l 1. GENERAL CRITERIA a

b*

cs The operating test conforms with the previously approved outline; changes are consistent with LM sampling requirements (e.g., 10 CFR 55.45, operational importance, safety function There is no day-to-day repetition between this and other operating tests to be administered LM distribution

.r The operating test shall not duplicate items from the applicants' audit test(s)(see Section N/A Overlap with the written examination and between operating test categories is within acceptable LM It appears that the operating test will differentiate between competent and less-than-competent LM i

aplicants at the desi nated license leve. WALK-THROUGH (CATEGORY A & B) CRITERIA Each JPM includes the following, as applicable:

EM

initial conditions initiating cues

references and tools, including associated procedures reasonable and validated time limits (average time allowed for completion) and specific designation if deemed time critical by the facility licensee

specific performance criteria that include:

-detailed expected actions with exact criteria and nomenclature

- system response and other examiner cues statem ents describing im portant obse rvations to be m ade by the applicant-criteria for successful completion ot the task-identification ot critical steps and their associated performance standards

- restrictions on the se uence of steps, if applicable Theprescit edqetionsfo oea ing Cte gory Ase aurigte pre doi nantyioensreferexmn ceatin d mee wthen EM criteria in Attachment 1 of ES-30 a c c esta b le lim its 3 0 % fo r th e w a lk -th ro u h a n d d o n o t c or nro m is e te s t in te rit At least 20 percent of the JPMs on each test are new or significantly modifie LM The associated simulator operating tests (scenario sets) have been reviewed in accordance with Form ES-LM It eth e fclrtyignatues wil s nicableore NR*Cevn eloped test ri t

nde dent e

v li inted Columhef eig neture c

u inFa iti al conditions N//

[a.TEac

inldeThe f

ollowyignatur asno applicable:

LM R-eelpdtss rnefeenesanndtoolsreincludingitsoial tems pr o

eumn";re efease cnurneeurd

ES-301 Simulator Scenario Quality Checklist Form ES-301-4 (R8, Sl)

-Faii :

M E-i~ -

Date of Exam:*9 *9/f o -'Scenario Numbers: /I 2-/

Operatin Test No.: /

Faciliy J21,1 QUALITATIVE ATTRIBUTES Initials a

b*

c# The initial conditions are realistic, in that some equipment and/or instrumentation may be out of LA

service, but it does not cue the operators into expected event.

The scenarios consist mostly of related event.

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

the event termination oint (if applicable) No more than one non-mechanistic failure (e.g., pipe break) is incorporated into the scenario without a credible preceding incident such as a seismic even.

The events are valid with regard to physics and thermodynamic.

Sequencing and timing of events is reasonable, and allows the examination team to obtain complete evaluation results commensurate with the scenario ob ective.

If time compression techniques are used, the scenario summary clearly so indicates. Operators A have sufficient time to carry out expected activities without undue time constraints. Cues are N

ive.

The simulator modeling is not altere.

The scenarios have been validated. Any open simulator performance deficiencies have been evaluated to ensure that functional fideli is maintained while runnin the lanned scenario.

Every operator will be evaluated using at least one new or significantly modified scenario. All other scenarios have been altered in accordance with Section D.4 of ES-30.

All individual operator competencies can be evaluated, as verified using Form ES-301-6 (submit the form alon with the simulator scenarios.

Each applicant will be significantly involved in the minimum number of transients and events specified on Form ES-301-5 submit the form with the simulator scenarios).

1 The level of difficulty is appropriate to support licensing decisions for each crew positio TARGET QUANTITATIVE ATTRIBUTES JPER SCENARIO; SEE SECTION DA.4.)

Actual Attributes Total malfunctions (5-8)

/ Malfunctions after EOP entry (1-2) Abnormal events (2-4)

31 Major transients I1-2)

/

/ EOPs entered/requiring substantive actions (1-2)

2- /1./ EOP contingencies requiring substantive actions (0-2)

/ Critical tasks ý2-3

S /

ES-301 Transient and Event Checklist Form ES-301-5 (R8, S1)

OPERATING TEST NO.:

Evolution Type Reactivity Normal Major lMnirmur um er

1

1 Reactivity

Normal

istrument /

3omponent Major

Scenario Number S are

Instructions:

Author:

(1)

Enter the operating test number and Form ES-D-1 event numbers for each evolution typ (2)

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.4.d) but must be significant per Section C.2.a of Appendix (3)

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicant's competence count toward the minimum requirement.

NRC Reviewer:

Applican t Si ype RO

Applicant #1 Applicant #2 Applicant #3 RO/SRO-I/SRO-U RO/SRO-I/SRO-U RO/SRO-I/SRO-U Competencies SCENARIO SCENARIO SCENARIO

2

1

34

2

4 Understand and Interpret 2, 4 2,3, 3 Annunciators and Alarms 5,6 4,5 5, 6 4, 5 Diagnose Events 2, 4 2, 3 2,4 1,2 and Conditions 5,6 4,5 5, 6 4, 5 Understand Plant 1,2 2,3 1,2 1,2 4,5 4,5 4,5 3,4 and System Response

6

Comply With and 1,2 1,2 1,2 1,2 3,4 3,4 3,4 4,5 Use Procedures (1)

5 6

5,6 Operate Control 1,2 1,2 3,5 4,5 Boards (2)

Communicate and 1,2 1,2 1,2 1,2 Interact With the Crew 3,4 3,4 3,4 3,4 5,6

5,6

Demonstrate Supervisory 1,2 1,2 3,4 3,4 Ability (3)

5,6 Comply With and 2,4 1,2 Use Tech. Specs. (3)

5,6

_____

-

~5,6

_

-

--

-

Notes:

(1) Includes Technical Specification compliance for an R (2) Optional for an SRO- (3) Only applicable to SRO Instructions:

Circle the applicant's license type and enter one or more event numbers that will allow the examiners to evaluate every applicable competency for every applican Author:

L UW-bk

"A I

NRC Reviewer:

-

gvý 4ýý r&0 Z-L Form ES-301-6 (R8, S1)

I

Written Examination Quality Checklist Form ES-401-7 (R8, Si)

Facility: Summer Date of Exam: 9/17/02 Exam Level: RO/SRO Initial Item Description a

b*

Ic Questions and answers technically accurate and applicable to facility LM a. NRC K/As referenced for all questions LM b. Facility learning objectives referenced as available R

- eve-'--FA no Me

SRO questions are N/A appropriate per Section D.2.d of ES-401 Question selection and duplication from the last two NRC licensing exams T, i appears consistent with a systematic sampling process A Question duplication from the license screening/audit exam was controlled as LM indicated below (check the item that applies) and appears appropriate:

-the audit exam was systematically and randomly developed; or

- the audit exam was completed before the license exam was started; or

_X the examinations were developed independently; or

- the licensee certifies that there is no duplication; or other (explain) Bank use meets limits (no more than 75 Bank Modified New LM percent from the bank at least 10 percent new, and the rest modified); enter the actual question

24

-

distribution at rightI Between 50 and 60 percent of the questions on Memory C/A LM the exam (including 10 new questions) are written at the comprehension/analysis level;

58 enter the actual question distribution at rig ht References/handouts provided do not give away answers LM Question content conforms with specific K/A statements in the previously LM approved examination outline and is appropriate for the Tier to which they are assigned; deviations are justified 1 Question psychometric quality and format meet ES, Appendix B, quidelines LM 1 The exam contains 100, one-point, multiple choice items; the total is correct and LM agrees with value on cover sheet dte Nae / Signature Date a. Author-Lee R. Miller lf,*

--*

8/13/02 b. Facility Reviewer (*)

N/A c. NRC Chief Examiner (#) 411

,-j-$,e d. NRC Regional Supervisor

'0 C,

Note:

The facility reviewers initials/signature are not applicable for NRC-developed examination # Independent NRC reviewer initial items in Column "c;" chief examiner concurrence required.

ES-401

ES-401 Written Examination Form ES-401-9 Review Worksheet 1.

3. Psychometric Flaw 4Job Content Faws 5.Ohr7 0#LKLOD 3e

//

(/) (1-5)

Cred. Partial Job-Minutia

/

Back-Q= SRO UESExplanation FocusDist, Linkunits ward K/A Oni

001AK1.05 1. Label change to conform to facility terms 2. Change initial condition power level from 17% to 22% to get further above C-5 auto rod interloc. Made label chang. Made change initial condition power level 25%

Accepted Changes

No comment

001 K5.38 Plant design does not have feedwater pump trip that results in a runbac Changed initial condition to reflect plant desig Accepted Changes

002A3.01 1. Label changes to conform to facility term. PZR programed level at 100% power is 60%.

1. Made label change. Made change initial condition 3. Changed stem to reflect "PORV has reseated" Accepted Changes

003A2.01 Label change to conform to facility terms Made label change Accepted Changes

3. Psychometric Flaws 4. Job Content Flaws I5.Other I I

I I. t e 7.

"I (F/H)

(1-5)

Stem Cues T/F Cred. Partial Job-Minutia

/

Back-Q= SRO U/E/S Explanatlon Focus Dis Link units ward K/A OnI

003AA1.01 Label change to conform to facility terms Made label change Accepted Changes

004K6.05 Label change to conform to facility terms Made label change Accepted Changes Acceptedn

No comment

No comment flM 1 I nfl 7.

I I

I I.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

Q#

LOK LOD Se us

/

(F/H)

(1-5) StemiCues T/F Cred. Partial Job-Minutia

/

Back-0= SRO U/E/S Explanation Focus Dst Link units ward I/A Oy

_

X 006K2.04 Operations Management does not expect operators to commit power supplies to memor Through their training, operators must learn set points, immediate actions, system designs and interrelationships, administrative procedures, and applications of knowledge to the job. The knowledge that is learned is expected to be demonstrated through the NRC examination format that measures recognition and recall of safety-significant knowledge without relying on references. This approach is consistent with the timely retrieval of information that may be required during the licensed operators' job and that might otherwise not be possible if the applicants prepared only for open-reference examinations. If too many open-reference questions are allowed on the initial licensing examination, the need and ability to learn and retrieve a broad body of knowledge would be lessene Similarly, the confidence that the baseline body of knowledge had been truly established could be questioned. (Taken from Operator Licensing FAQ # 42)

The K/A is valid and has a value of 3.6/ The question remain Facility felt that NRC was asking the operator to recall from memory something that was not realistic for NRC to ask at this level of detail and that this would be a post exam comment. That the level of knowledge was beyond what was expected by Operations Managemen The question remains as a valid question for the exa.. Psychometric Flaws 4. Job Content Flaws 5. Other.

01#

LOK LOD (F/H)

(1-5)

Stem Cues T/F Cred. Partial Job-Minutia

/

Back-Q=

SRO UiE/S Explanation IFocusj Djs I Link I units ward K/A Ony

007A4.10 1. Label change to conform to facility terms 2. Change RM-A2 to RM-A4 in distracter A 1. Made label change. Changed RM-A2 to RM-A4 in distracter A and Accepted Changes

007EA2.02 Format problem Format problem exists only in exported copy facility worked fro Added reading for N44 - 101% and broke out readings for each SIG Accepted Changes 008AK1.01 Request wider span for temperatures in the answer and distracters No change necessar Accepted No comment No comment 011EK1.01 1. Label change to conform to facility terms 1. Made label change Accepted Changes 13. Psychometric Flaws 4. Job Content Flaws 5. Other.

Q#

LO T O Ste Explanatiortia cH)

s-5)

S es Cre i Job-Minutia

/

Back-Q=

SRO U/E/S Explanation PFocus Dis ik nt ward K/A On

011 K6.04 1. Suggest changes to remove effect of rod control 2. Clarify normal GP1 B/U heater status as "on".

3. Change failure from level reference failure to Tavg Median 1. Effect of rod control is not contained in answer or distracters. No additional information is need in stated initial conditions concerning rod contro. Label changes made to facility specific on PZR heater group. Changing failure would result in question not matching KA. No change made in failur Accepted Changes

No comment

013A4.02 Label change to conform to facility terms Made label changes Accepted Changes

013K4.16 1. Label change to conform to facility terms 2. Change distracter A to avoid possible additional correct answer 1. Made label changes 2. Changed distracter A Accepted Changes

3. Psychometric Flaws Stem Cues Focus T/F I. 4. Job Content Flaws Cred. Partial Job-Minutia Dist. I I Link I

units Back ward 5. Other Q=

K/A Q#

Accepted Changes

01 5AK3.03 Format problem Format problem exists only in exported copy facility worked fro Accepted LOK (F/H) LOD (1-5)

21 Explanation 014A1.03 Correct answer A to Summer design 1. Answer A changed to median select Delta T and distracter C changed to median select Tavg Explanation 015K3.06 Procedurally, the only viable answer is to have rod control remain in manual. Per the AOP, rod control would NOT be restored to automatic until step 17, and only after the NI had been restored and bistables reset. Although choice C is true insofar as testing how the rod control system is actually designed, an operational exam such as this tests knowledge of procedures and their usage, whereas the question is written as if testing the construction of the system at a technician leve The alternative questions provided offer knowledge on the operational level of the interface between rod control and nuclear instrumentation, as well as the effects of nuclear instrumentation failures on rod control system operatio Through their training, operators must learn set points, immediate actions, system designs and interrelationships, administrative procedures, and applications of knowledge to the job. The knowledge that is learned is expected to be demonstrated through the NRC examination format that measures recognition and recall of safety-significant knowledge without relying on references. This approach is consistent with the timely retrieval of information that may be required during the licensed operators' job and that might otherwise not be possible if the applicants prepared only for open-reference examinations. If too many open-reference questions are allowed on the initial licensing examination, the need and ability to learn and retrieve a broad body of knowledge would be lessene Similarly, the confidence that the baseline body of knowledge had been truly established could be questioned. (Taken from Operator Licensing FAQ # 42)

Recommended replacement questions do not match the KA. The question will be modified but will keep the intent.

Added "with no operator actions". Accepted Changes I

I I

1. Psychometric Flaws 4. Job Content: Flaws 5. Other.

0#

LOK LOD rd ata o-Mnta#

Back-9=

SOUET/S_

Explanation

01_Al.01 1. Label change to conform to facility terms 2, Change reference to 200 degrees and 0 degrees to "much higher" and "much lower" 1. Label changes mad. Changes made in what core subcooling monitor should read (Maximum subcooling and superheat)

Accepted Changes

022A1.04 Format problem Format problem exists only in exported copy facility worked fro Accepted

022AA2.03 1. Label change to conform to facility terms 1. Label changes mad Accepted Changes

024AA2.02 1. Label change to conform to facility terms 1. Question will be replaced. Question can be answered using question 2 Accepted Changes.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

0*LOK LOD

__

f~

__

_

_________________

(F/H)

(1-5 Stem Cues T/F Cred. Partial Job-Minutia

/

Back-C= SRO U/E/S Explanation Focus(1-5)Focu Dis Link units ward K/A Only

026AA2.06 Want change to stem to clarify questio Changed stem to clarify questio Accepted Changes

026AK3.03 Editorial change Editorial change mad Accepted Changes

No comment

027AA2.16 1. Editorial change suggeste. Delete initial condition addressing PZR heater status Made editorial changes and deleted initial condition addressing PZR heater status Accepted Changes

No comment.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

Ot#

LOI(

LOD-MiuiExlnto (F/H)

(1-_5) tem Cues T/FI cred. Partial Job-Minutia

!

Back-Q=

SRO U/E/S Explanation Focus is Link units ward K/9 On y

028K2.01 Ops Management would not expect this level of detail knowledge of whether the feed is 1 DB1 OR 1 DB2 to be committed to memor Through their training, operators must learn set points, immediate actions, system designs and interrelationships, administrative procedures, and applications of knowledge to the job. The knowledge that is learned is expected to be demonstrated through the NRC examination format that measures recognition and recall of safety-significant knowledge without relying on references. This approach is consistent with the timely retrieval of information that may be required during the licensed operators' job and that might otherwise not be possible if the applicants prepared only for open-reference examinations. If too many open-reference questions are allowed on the initial licensing examination, the need and ability to learn and retrieve a broad body of knowledge would be lessene Similarly, the confidence that the baseline body of knowledge had been truly established could be questioned. (Taken from Operator Licensing FAQ # 42)

Recommended replacement questions do not match the KA. The question will be not be change Facility felt that NRC was asking the operator to recall from memory something that was not realistic for NRC to ask at this level of detail. That the level of knowledge was beyond what was expected by Operations Managemen The question remains as a valid question for the exa.. Psychometric Flaws 4. Job content Flaws 5. Other.

Q9 LOK LOD

]

C arilJ-Mnua

/

Back-'

O= SRO U/E/S Explanation (F/H)

(1-5)

Stem Cues T/F rd PatlJo-Mnia

Focus

]

Dis Link units ward K/A Onlv

029EK3.01 Remove specific reference to -0.33 dpm and change to is negativ No change made to -0.33 dpm. Instructor's lesson plan specifies this knowledge as WOG background document required knowledg Accepted changes

032AA2.06 Summer uses the Gammametric designed SR/IR nuclear instruments and therefore any questions with regards to the automatic removal of high voltage are non-applicable to the station. The alternate questions are in the same section of the K/A catalog; 4.2 Generic APE' Changed to a new question and added Group A from 100 to 150 step Accepted changes

033A2.01 Facility design of fuel storage racks ensures < 0.95 Keff Question was modified using facility comment Accepted changes

X 033AA2.02 Question is not operational in content and requires detailed knowledge of settings and circuitry The question modifie Accepted changes Explanation 034K4.01 Replace question because FH personnel indicated all choices are justifiable to a degre. Who are the FH personnel asked and are they on the security agreement?

2. How do each of the choices protect a fuel assembly from binding while being loaded into the core?

No additional FH personnel discussed this question, only asked personnel already on the security agreemen Added "automatically protect" and "entering the vessel" to the question. Accepted changes

X 035K3.01 1. Operational experience and scenarios run on simulator showed STM PRESS LO SI will occur over varying time frames with no change in PZR leve. Changed answers to reflect STM PRESS LO SI and no change in PZR leve Accepted changes

036AK1.01 Format problem Format problem exists only in exported copy facility worked fro Accepted

No comment

038EK3.08 Editorial change Editorial change mad Accepted changes.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

Q#

LOK LOD S

eP (F/H)

(1-5)

Stem Cues TIF Cred. Partial Job-Minutia

/

Back-Q=

SRO U/E/S Explanation Focus Dis Link units ward K/A O1 _

039K1.02 1. Delete initial condition concerning Steam Dump Bypass interlock switche. Add "B" to stem for steamline PORV 3. Add automatically to will not open in answer A and change distracter C reason for relief not openin. Deleted initial condition concerning Steam Dump Bypass interlock switche. Added "B" to stem and modified distracter Accepted changes

041A2.02 Editorial change Editorial change mad Accepted changes

No comment

054AA1.01 1. Spelled out MFP to avoid confusio. Changed choices B and C 1. Spelled out MFP 2. Proposed changes altered question focus Changed 'B' will remain running. Accepted changes

055EG2.4.16 Editorial change Editorial change mad Accepted changes

4. Job Content Flaws

1 5. OtherI Back-Q= SRO I

I.

U/E/S Explanation 055G2.4.11 1. Clarification of initial conditions which are too close to trip criteri. Make Answer D match trip criteria in AOP20. Initial conditions are above the trip set points and should not lead to "A" as the correct selectio. No change is made to Answer D. It is sufficient as i Changed power to 35% and 350 MWe and modified answer D. Accepted changes

056K1.03 Editorial change Editorial change mad Accepted changes

No comment

058AK3.02 Editorial change Editorial change mad Accepted changes

lNo comment

No comment

061 K4.02 Editorial change Editorial change mad Accepted changes

No comment un LS I We V

I rd I K/A I ON

IIJ.

3. Psychometric Flaws 4.Job Content Flaws 5. Other.

(F/H)

(1-5)

Stem Cues T/F Cred. Partial Job-Minutia

/

Back-Q= SRO U/_/S Explanation Focus I

Dis Link units ward K/A ON y

No comment

063K3.01 1. Requires specific, detailed knowledge of ARP-001-636, window 4-3. Request copy of ARP and informed as to all the alarms and indications that have occurred. Expectation for Operators is far more general that the question requires and requires more knowledge that an operator is trained and expected to kno Providing a copy of ARP-O01 -636, window 4-3 makes the question a direct look-up. Therefore, a copy of ARP-001 636, window 4-3 is not permitted. Just because material covered by a KIA is not trained on does not mean that the question is not valid. See question 1 The question will not chang Facility felt that NRC was asking the operator to recall from memory something that was not realistic for NRC to ask at this level of detail and the Facility representative indicated that this would be a post exam comment. That the level of knowledge was beyond what was expected by Operations Managemen The question remains as a valid question for the exa A4.02 Format problem Format problem exists only in exported copy facility worked fro Accepted changes

No comment.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

Q#

(1-5)

Stem Cues T/F Cred. Partial Job-DMinutia

1 Back-

= SRO U/EIS Explanation F

(-5)

Focus Ds Link units ward IJA Ony

068AK3.17 Provide a CREP panel drawing as a reference Through their training, operators must learn set points, immediate actions, system designs and interrelationships, administrative procedures, and applications of knowledge to the job. The knowledge that is learned is expected to be demonstrated through the NRC examination format that measures recognition and recall of safety-significant knowledge without relying on references. This approach is consistent with the timely retrieval of information that may be required during the licensed operators' job and that might otherwise not be possible if the applicants prepared only for open-reference examinations. If too many open-reference questions are allowed on the initial licensing examination, the need and ability to learn and retrieve a broad body of knowledge would be lessene Similarly, the confidence that the baseline body of knowledge had been truly established could be questioned. (Taken from Operator Licensing FAQ # 42)

No reference will be provide The facility representative stated that if one was to look al the individual question that one could not disagree with the question but looking at the entire exam, a high majority of the exam is being affected by obscure questions. Accepted

No comment

071 K3.05 Change "has failed to upscale" to fails low or as i Changed stem to "fails AS-IS" Accepted changes No comment I.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

Q#

(-LO5 Stem Cues T/F Cred. Partial Job-Minutia

/

Back-Q= SRO u/E/S Explanation Focus Dis I Link units ward K/A Oni

X 075G2.1.32 The question implies that precautions in all SOP's should be committed to memory, which is an unrealistic expectatio However, some items concerning system operation may be committed to memory and these have been captured in the precautions of many SOP's. We recommend a reference be provided for this question (copy of the precautions for SOP 207), but we believe one of the alternate questions should be use Providing a copy of SOP-207 makes the question a direct look-up. Therefore, a copy of SOP-207 is not permitte The recommended questions for replacement (two questions were provided) do not conform to the KA. The recommended questions are best suited for T2G2 System:

075 Circulating Water Syste Facility representative considered this the worst question and did not consider this question as acceptable. He stated that the operators were not trained to this level The facility representative was reminded that just because the operators were not trained on material did not exclude it from being tested. Changed distracter B to

"0830" to make a wider span between distracter B and answer A2.02 1. Added initial condition for clarification 2. Change distracter A and answer C Will make changes to initial condition and changes to reflect true procedure flow pat Change reading on P14402 to 5#. Accepted changes.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

0Q# LOK LODff (

I (F/H)L j (1-5)

Stem Cues T/F Cred. Partial Job-Minutia

/

Back-Q=

SRO U/E/S Explanation

~~Foc s

___

DEr Link units wadK/A Inx...._______________________

076AK2.01 Editorial change Editorial change mad Accepted changes

No comment

G2.1.10 Editorial change Editorial change mad Accepted changes

No comment

G2.1.29 1. Delete one initial condition 2. Change distracter A to eliminate it as a viable alternativ. Initial condition delete. Distracter A is changed to eliminate it as a viable alternative Accepted changes.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

Q#LOK LOD-f

-

i

[--

(F/H)

(1-5)

Stem Cues TIE Cred. Partial Job-Minutia

/

Back-Q= SRO U/E/S Explanation Focus Dist."

Link units ward K/A Only

G2.. Badging is outside protected area, changing intended criteria, moved reason to vacate CR "inside the fence".

2. Removed references to Station log. CRS has discretion to log temp. relief if so desire. SAP-200 (page 8) 6.2.7F...when the relieved operator remains on site and is expected to return to his assigned duties, does not require a Station Log Book entr. The question asks for the MINIMUM items an unexpected or temporary relief turnover should includ No change to the question is require Based on on-shift experience and particular supervisor discretion rather than the procedure, the applicants may not choose the correct answer. There were additional comments that the requirement of "remain on site" was interpreted as "remain inside the protected area."

Individual relieved will remain inside the protected are Added procedural reference to what is the MINIMUM required b No comment

No comment

G2.2.19 Change distracter A to eliminate it as a viable alternativ Distracter A has been change Accepted changes

No comment

No comment.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

Q#

LOK LOD S

(F/H)

(1-5)

Stem Cues T/F Cred. Partial Job-Minutia

./

Back-Q-

SRO U/E/S Explanation tFocus I

Ds Link unitsjward K

On

G2. Add the words "annual" and "limit" to Occupational Dose for clarificatio The words "annual" and "limit" have been added to Occupational Dose for clarificatio Accepted changes

No comment

G2.4.10 Amplify initial conditions to provide realistic means of receiving RCP oil information Change was made to provide amplifying informatio Accepted changes

G2.4.11 1. Add to initial conditions to specify operating in GOP-9 and change RHR flow and amps from "are oscillating" to" begin to oscillate" 2. Change answer C from 900 gpm for 33 minutes to 450 gpm to stabilize amps and flo. GOP-9 reference has been added to initial conditions 2. Additional information of maintaining level 3" above half-pipe was not added. This information could affect the distracter. Distracter C has been changed from 33 minutes to 40 minute Accepted changes

No comment

No comment.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

0#

LOK LCD f

-

[l (F/H)

(1-5)

Stem Cues T/F Cred. Partial Job-Minutia

/

Back-Q=

SRO U/i./S Explanation

No comment

i Lin u

w K

n IIINo comment

No comment

W/E05EK. Change the initial conditions to delete the following:

The unit is in an Emergency conditio The operators verified that a secondary heat sink is required All RCPs are tripped 2. Make clear that the RCP's are tripped in step 17 of EOP 1. Initial conditions were changed to delete "The unit is in an Emergency condition".

2. Initial conditions were changed to indicate that all RCP's were tripped per EOP-15.0 step Accepted changes.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

a#t LOK LD rI (F/H)

(1-5)

Stem CuesiTIE Cred. Partial Job-Minutia

/

Back-Q=

SRO U/EIS Explanation Focus Dis Link units ward K/A Only

W/E06EK Provide EOP1 2.0 as a reference. Not memory ite Through their training, operators must learn set points, immediate actions, system designs and interrelationships, administrative procedures, and applications of knowledge to the job. The knowledge that is learned is expected to be demonstrated through the NRC examination format that measures recognition and recall of safety-significant knowledge without relying on references. This approach is consistent with the timely retrieval of information that may be required during the licensed operators' job and that might otherwise not be possible if the applicants prepared only for open-reference examinations. If too many open-reference questions are allowed on the initial licensing examination, the need and ability to learn and retrieve a broad body of knowledge would be lessene Similarly, the confidence that the baseline body of knowledge had been truly established could be questioned. (Taken from Operator Licensing FAQ # 42)

Providing a copy of EOP-1 2.0 makes the question a direct look-up. Therefore, a copy of EOP-12.0 is not permitte No reference will be provide The facility representative stated that this was a function of the STA, Shift Engineer, and they did not expect the SRO to know all entry conditions for all the EOP's (Red/Orange path).

W/E08EK Editorial change Editorial change mad Accepted changes No comment I.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

O#

LOK LCD

[xpanaio (F/H)

(1-5)

Stem Cues T/F Cred. Partial Job-Minutia

/

Back-Q=

SRO UiEiS Explanation

- t

-

Focus

_

Dist.'

Lik I

units Iward K/A

__

_

__

_

__

_

__

_

W/E09EA Question requires memory knowledge of EOP1.3 in its entiret No criteria given which would force EOP11.4 provide Initial condition directed at CST will be expanded indicating CST level is not adequate to support at Natural Circ cooldown rate of <50 F/hr. The question does not require memory knowledge of EOP-11.3 in its entirety. It does require an overall knowledge of the EOP1.3 and 1.4, the relation between the two, and when transition from on to the other should be don The facility representative thought that EOP-1.3 and EOP-1. 4 are obscure and that they were sensitive to the failures from the last exam. They did state that the operators were trained on all EOP' Will expand Answer D to remove reference to "perform the first 9 step W/E11 G2. Can answer D be discounted? FCV-605A opening fully could push flow to runou The initial conditions has been corrected FCV-605 should be HCV-603. Given the RCS Hot Leg Level at 16 inches, vortexing will occur before the RHR pump would trip from being in runou Accepted changes, but would like answer D to be more wron No comment.

3. Psychometric Flaws 4. Job Content Flaws 5. Other.

O#

LOK LOD Explanatrtn (F/H)

(1-5)

Stem Cues T/F Cred. Partial Job-Minutia

/

Back-Q=

SRO U/E/S Explanation Focus Dis Link units ward K/A Ony

W/E13EA EOP 15.1 is very obscure procedure and is therefore rarely used. If the question in it's current form is used, then a copy of the EOP should be provided and the answer changed as indicated to allow for the TDEFW pump to simply be started if

"B" or "C" is the affected S/G. No reference is made to the oniy steam supply being from the affected S/G and therefore there is no correct answer as written. Our recommendation would be to make the question more broad in scope as exampled by the questions provide Providing a copy of EOP-15.1 makes the question a direct look-up. Therefore, a copy of EOP-1 5.1 is not permitte The correct answer will be modified to delete the reference

"using the 'B' S/G as the steam supply" because an assumption must be made that the highest S/G pressure for those S/Gs that can supply steam is the "B" SIG. The recommended question for replacement (only one provided) does not conform to the KA. The recommended question is best suited for T2G2 System: 035 Steam Generator System (S/GS).

The question was changed to S/G The facility representative stated that the changes to the questions represents what is in the procedure, but that if one was to look at the individual question that one could not disagree with the question but looking at the entire exam, a high majority of the exam is being affected by obscure questions. The facility representative stated that they had a problem with about 10% of the exam in that the applicants will not have the required detailed level of knowledge.

No comment

! :2 3. Psychometric Flaws 4 Job Content Flaws 5. Other.

(F/H)LOK (1-5)

Stem Cues T/F Cred. Partial Job-_i Minutia

1 BaCk-w =

SRO U/E/S Explanation

_________ Focus

__

j Dis Link units ward[/lOl____________________________________________

W/E15EA Provide EOP-12.0 as referenc Through their training, operators must learn set points, immediate actions, system designs and interrelationships, administrative procedures, and applications of knowledge to the job. The knowledge that is learned is expected to be demonstrated through the NRC examination format that measures recognition and recall of safety-significant knowledge without relying on references. This approach is consistent with the timely retrieval of information that may be required during the licensed operators' job and that might otherwise not be possible if the applicants prepared only for open-reference examinations. If too many open-reference questions are allowed on the initial licensing examination, the need and ability to learn and retrieve a broad body of knowledge would be lessene Similarly, the confidence that the baseline body of knowledge had been truly established could be questioned. (Taken from Operator Licensing FAQ # 42)

Providing a copy of EOP-12.0 makes the question a direct look-up. Therefore, a copy of EOP-12.0 is not permitted. If provided for this question, it could be used to answer question 9 No reference will be provide The facility representative stated that if one was to look al the individual question that one could not disagree with the question but looking at the entire exam, a high majority of the exam is being affected by obscure questions.

100 No comment

Monday, August 26, 2002 Mr. Miller and Mr. Rose:

After a detailed review of all 100 Written Exam items, we have concurred that there are 37 questions, which require no further changes and are ready-as-is for the 9/17/02 SRO Exam at V.C. Summer. These exam questions numbers are:

8

14

18

32

45

52

55

57

62

68

71

75

78

83

86

88

93

98 100 Additionally, we have identified another 45 items, which we are classifying as requiring "Minor Changes." These

"Minor Changes" include such items as: re-formatting text, correcting spelling or capitalization, correction of non plant-specific terms, and also providing clearer information and correcting minor stem and distractor errors. They do not change the intent of the question or of the KJA item they are testing. We are re-checking these items one last time this morning, and I will: (1) fax you the marked-up version of these questions AND: (2) E-mail you our proposed corrected version, shortly after lunch today, August 26, 200 I will also E-mail you early this afternoon the Simulator and In-Plant Walkthrough JPM's with your comments from last week incorporated. Attached also to this E-mail you will find the Third or Backup Scenario, both the D-1 attachment as well as our write-up of expected actions and procedural flowpat The thirty-seven (37) "as-is" items plus the forty-five (45) "Minor Change" items you will have after lunch, coupled with the nine (9) items we submitted comments on last Friday, brings the total to ninety-one (91) items. We are reviewing the remaining nine (9) items today with Operations Management to ensure they meet management expectations, for there are some questions in our minds as to their technical accuracy or whether the information they test is appropriately tested in the Closed-Reference mode as currently writte We look forward to discussing the Exam with you on Thursday, August 29 at the Region II Offices. We anticipate arriving at your offices around 0830 on Thursday. Please feel free to contact me at (803) 931-5162 should you have any additional questions concerning exam material G. Lippard

Written Examination Form ES-401-9 Review Worksheet Instructions

[Rder to Appendix B for additional Intormation regarding sadh of the ftllowing c*ncepte.]

I Enter the level ol knowledge (LOK) of each question as either FundslUmenl or (H*igher cognitive leve.

Enter the level of difficulty (LOD) of each question using 81 -5 (easy - d.efctLt) rating scale (queattionflan the 2-4 range re acceptable). Check the appropriate ox If a psychonietric flaw Is Identified:

The stem lac*s sufficient tfous to slict tie correct anwer (e.g.. unclear Intent, more Informniolen Is neded, or 1oo much needless Inlformat*o*)

The stem or dstracors contain cues Os.. dlues, apecIfic deterthteri. phrasing, lntwo, sic).

The answer choices are a collection of u*deated truefelase statement More than one distractor Is not credibl One or more detractors Is (are) partlly correct (e.g., i the applicant can make unstated assumptions that are not contradicted by stem). Chec the appropriate box If a ob content error Is Identified:

note ratinalIn ontent)

The question Is not Irked to the job reqren questionhasvaldAbaswrtten.In conte The question requires the recall of k*o that too specfic for the dosed referenc test mode ee.,itnotr dto be known frm meoy).

The question contains data with an unreeblla level of accuracy or Inconsistent units (e.g.. panel moatr in perowith question gall").

The question requires reverse Wi or spplicatlcn compared to the job requtrement.

Based on the reviwers judgment Is the question as written (L)nacteptale (requiring r*piror replaement), It need of (E)d~tOrll enhannceent, or (S)atblactwry? Forn amtl o

a aminimum, e how the bdixB cs.donei ariutesarenol me N...

...

..

I, f lUVIW o044 of 45 ES-401 II NUREG','-1U2",

Re[vlision 8

97. WlE13EA1.1 I

-"8" S/G pressure is 1250 psi "8" Narrow range level is 82%

-EOP-i6.1, "Response to Steam Generator Overpressure has been entere The condenser is not availabl The "B" PORV is stuck dose Which ONE of the following describes the preferred method to reduce "B" S/G pressure in accordance with EOP-1 5.1?

A. Commence an RCS cooldown to below 565 O B. Start the TD EFW pump using the "B" SIG as the steam suppl C. Isolate EFW to the "B" Steam Generato D. Establish Blowdown from the "B" Steam Generato Lesson Plan EOP-15.1 Response to Steam Generator Overpressure. Objective 210 Bank Question 392 A. Incorrect, this would be done if starting the TD EFW did not wor B. Correct, the procedure directs the operator to lower pressure by starting the TO EFW to lower S/G-r*essur C. Incorrect, this would be done after D. Incorrect, this would be done to lower leve Comments on questions we to consider for replacement 97. EOP 15.1 is a yellow path entry procedure. The answer to this question is step 4 alternative-action and it is not an immediate operator action. This is a very obscure procedure and is therefore rarely used. If the question in it's current form is used, then a copy of the EOP should be provided and the answer changed as indicated to allow for the TDEFW pump to simply be started if "B" or "C" is the affected S/G. No reference is made to the only steam supply being from the affected S/G and therefore there is no correct answer as written. Our recommendation would be to make the question more broad in scope as exampled by the questions provide Recommended Question for Replacement 97. W/E13EA1.1 1

-"B" SIG Pressure is 1250 psi "B" Narrow range level is 82%

-EOP-1 5.1, "Response to Steam Generator Overpressure has been entere The condenser is not availabl The "B" PORV is stuck close Which ONE of the following describes the preferred method to reduce "B" SIG pressure in accordance with EOP-15.1?

A. Commence an RCS cooldown to below 565 O B. Start the TD EFW pum C. Isolate EFW to the "B" Steam Generato D. Establish Blowdown from the "B" Steam Generato K/A catalog page W/E Westinghouse

Steam Generator Overpressure EA1.1 Ability to operate and/or monitor the following as they apply to the S/G overpressure; Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual feature. W/E13EA1.1 1 The plant is operating at 100% power. The steam dump system in its normal alignment. A failure occurs which results in an Indicated pressure of 1150 PSIG on IPI 2010. Which of the following describes the operation of the "B" steamlnne PORV, IPV 2010 due to this condition; A. Since the steam dump system is in the TAVG mode, the PORV will ONLY open if the required difference exists between Indicated average temperature and reference temperature and the steam dumps receive an arming signa B. The PORV will automatically swap to the PWR RLF mode when pressure reaches 1133 PSIG and will open to relieve pressure. The valve will be controlled by the M/A station and will close at approximately 1090 PSI C. As long as the condenser available interlock Is present, the PORV is blocked from opening. The only way to allow the PORV to open would be to place the steam dump mode selector switch to the RESET positio D. The PORV will not open because its automatic opening in this condition would result in an uncontrolled cooldown. In order for the PORV to open, both steam dump interlock switches must be taken to the BYP INTLK positio. 075G2.1.32 1-The Unit is starting up after a refueling outag The Circulating water system is being started u At 0800 the 'A' CW pump was started and secured due to a water box cover leakin At 0810 the 'A' pump was started again, but tripped immediatel Which one of the following describes the earliest time that the "A' Circulating water pump could be restarted?

A. 'A' circulating pump can be started at any time, six pump starts per day are allowe B. 0840 is the earliest that the 'A' circulating water pump could be starte C. 0850 is the earliest that the 'A' circulating water pump could be starte D. 0910 is the earliest that the 'A' circulating water pump could be starte SOP-207, precautions and limitations # Lesson Plan TB-08 Circulating Water System, objective TB-8-0 A. Incorrect, the pump is limited to six starts per day, but should have 40 minutes after two attempts of starting from cold condition B. Incorrect, this would 1r6 *

the applicant choose 30 minutes as the cooldown tim C. Correct, 0810 + 40 minutes would be 0850, this would be the earliest tim D. Incorrect, this wouldl

%t e applicant choose 60 minutes as the cooldown tim Comments on questions we to consider for replacement 65. SOP 207, Circulating Water System, is a multi-level use procedure. However, section III A, startup of the circ water system, is a continuous use procedure. Operations management expectation is that all precautions are reviewed prior to starting any component or system. This makes the review of precautions mandatory and should not be committed to memory. The starting duties of all pumps are listed in the specific SOP precautions and should always be reviewed prior to equipment operation. The question implies that precautions in all SOP's should be conmmitted to memory, which is an unrealistic expectation. However, some items concerning system operation may be committed to memory and these have been captured in the precautions of many SOP's. We believe our questions are more in-keeping with that standard and require prompt operator attentiona We recommend a reference be provided for this question (copy of the precautions for SOP 207) if the question is used in its original form, but we believe one of the alternative questions should be used.

Recommended Question for Replacement 65. 075G2.1.32 1 The plant Is operating at 100% power during the winter months and Is using two circulating pump operation with "A" and "B" circulating water pumps running. The "C" circulating water pump is Idle and available for use. What would be the expected plant response to a trip of the "B" circulating water pump?

A. The "B" circulating water pump discharge valve will close and the "A circulating water pump continues to run unaffecte B. The "B" circulating water pump discharge valve will close and the "A" circulating water pump discharge valve will close to the 30% open positio C. The "B circulating water pump discharge valve will not close and must be closed by the operator to prevent backflow and the "A" circulating water pump continues to run unaffecte D. The "B circulating water pump discharge valve will not close and must be dosed by the operator to prevent backflow and the "A" circulating water pump discharge valve will close to the 30% open positio Answer: B K/A catalog page 075 Circulating Water System G2.1.32 1-Generic Knowledges and Ability: ability to explain and apply all system limits and precaution Change to the CCW system Question: 73 65. 075G2.1.32 1 Normally, if the running Component Cooling Water (CCW) pump in the ACTIVE train were to trip, the non-essential loads would be:

A. Protected since the standby pump would star B. Protected since the INACTIVE loop CCW pump would star C. Lost until the swing pump could be properly racked into the ACTIVE loo D. Lost until the INACTIVE loop could be valved into the non-essential loo Answer; A

58. 063K3.01 I P

Diesel Generator 'A is running for a surveillance when DPNI HA is de-energize What is the immediate effect on Diesel Generator 'A' by the loss of DPN1 HA?

A. Diesel will continue to run, the diesel can be stopped from local STOP PB but not the MCB Test switch, and only the emergency engine protective trips are enable B. Diesel will continue to run, the diesel can not be stopped from the local STOP PB or the MCB TEST switch, and the engine protective trips are disable C. Diesel will continue to run, Diesel speed control is locked at current speed, the diesel can be stopped from the local STOP PB but not from the MCB TEST switch, and the emergency engine protective trips are disable D. Diesel will immediately trip due to the engine protective trips being actuate REF: Summer Exam Bank #555 ARP-001 4-3 DG A Loss of DC Distracter A - Incorrect, DIG can not be stopped from the local Stop PB and no protective trips are enable Answer B - correct, the ability to shutdown the DIG by placing the Test switch In stop (MCB) or by depressing the-STOP PB (Local) Is lost. The DIG engine protective trips are disabled due to the Inability to energize XVX10998A-D.G, Air toýFuel:Rack)VSDOCyl solenoid v-alye,,.-,...

Distracter C - incorrect, speed control is not affected, the DIG can not be stopped from the local Stop PB, and the engine protective trips are disable Distracter D - incorrect, the DIG will not Immediately tri Recommended Question for Replacement 58. 063K3.01 I What will be the response of D/G A" if a Safety Injection occurs while DPN-1 HA is deengerglzed?

A. Diesel will not start because DC fuel oil pump is deengergize B. Diesel will not start because DC air start solenoids are deenergize C. Diesel will start and remain in standby ready for loadin D. Diesel will start, but exciter will fail to field flas K/A catalog page 3.6-6 063 DC Electrical Distribution K3.01 1 Knowledge of the effect that a loss or malfunction of the DC electrical system will have on the following; the EDG S

Comments on questions we to consider for replacement 58. Requires the operator to have specific, detailed knowledge of ARP-001-636, window 4-in order to answer a question to this level of detail, the operator needs to be provided with a copy of the ARP for this specific window and be informed as to all the alarms and indications that have occurred. The expectation for operators is far more general and the original question is to the level that an operator is both trained and expected to know from memory. If the reference is not provided, then we believe the original form of the question should be use. 0351<3.01 1 Given the following plant conditions;

- The unit is operating at 75% steady state power

- All systems are in automatic control

- MSIV PVM-2801 B slowly goes full closed Which one of the following indicates the initial RCS response to the closure of MSIV PVM-2801B?

(Assume no operator actions)

A. RCS loops "A, B, C's" Tavg decrease together. PZR level decreases as expected for a Reactor Tri The Reactor trips on low Tavg and High Steam Flow on S/G "B".

B. RCS loops "A and C's" Tavg increases while RCS loop "B" Tavg increase is greater than the increase for RCS Loops "A and C". PZR level increases. No Reactor trip occur C. RCS loops "A and C's" Tavg increases while RCS loop "B" Tavg decreases. PZR level decrease Reactor trips on Turbine tri D. RCS loops "A and C's" Tavg decreases while RCS loop "B" Tovg increase is greater than the increase for RCS Loops "A and C". PZR level decreases. No Reactor trip occur REF: TB-2 Main Steam System TB-5 Turbine Control and Protection System IC-9 Reactor Protection and Safeguards Actuation System AB-2 Reactor Coolant System Distracter A - incorrect, RCS loop Tavgs will not decrease together and the reactor will not tri Distracter B - incorrect, RCS loop A and C Tavgs will decrease not increase. PZR decreases not increas Distracter C - incorrect, RCS loop A and C Tavgs will decrease not increase. The reactor does not tri Answer D - correct, with no reactor trip, RCS Loop A and C Tavg decreases and Loop B Tavg increases, and PZR level decrease Comments on questions we to consider for replacement 39. V.C. Summer Station has operational experience with this very issue. We ran this scenario on the simulator over varying timeframes and each produced the same result. The possibility exists that the rate of closure could be slow enough to prevent a STM PRESS LO SI, so it is necessary to remove any discussion of a slowly failing MSIV to preclude confusion. Based on operational experience, a STM PRESS LO SI will occur, and the simulator provided reiteratio There was no perceptible change in pressurizer level, so all reference to pressurizer level change has been removed. The effects observed on the simulator have been captured in answer B, and the other distractors modified accordingly based on the original distractors provided and the corrected data from the simulato Recommended Question for Replacement 39. 035K3.01 1 K/A catalog page 3.4-14 035 Heat Removal From Reactor Core; Primary System; Steam Generator System K3.01 1 Knowledge of the effect that a loss or malfunction of the S/Gs will have on the following: RCS 39. 035K3.01 1 Given the following plant conditions:

- The unit is operating at 75% steady state power

- All systems are in automatic control

- The IB operator is performing STP 121.002, Main Steam Valve Operability Test on "B" main steam isolation valve (PVM 2801B). A failure results in the full closure of PVM 2801 Which one of the following indicates the initial RCS response to the closure of MSIV PVM-2801 B?

(Assume no operator actions) RCS loops "A" and "C" TAVG begin to decrease; "B" loop TAVG begins to increase. No reactor trip or safety injection occur RCS loops "A" and "C" TAVG begin to decrease; "B" loop TAVG begins to increase. Reactor trip

-.and safety injection occurs due to STM PRESS LO S S RCS loops "A" and "C" TAVG begin to increase; "B" loop TAVG begins to decrease. No reactor trip or safety injection occur RCS loops 8A" and "G" TAVG begin to decrease; "B" loop TAVG begins to increase. Reactor trip and safety injection occurs due to STM PRESS LO S Correct answer:

B S

Comments on questions we to consider for replacement 37. This question is not linked to the job requirement of any licensed operator. The question has a valid K/A, but is not operational in content and requires detailed knowledge of settings and circuitry which are the responsibility of the I&C technician. Operators should be tested on IR channel response, automatic actions, and subsequent power limitations. The effects of changing to the Cambelling mode of operation are observable by the board operator, but the variations are too great to capture in the static situation that exists in a written examination. The indications listed do not necessarily constitute improper calibration and could only be determined by a technician, not an operator. The alternative questions clearly test the K/A listed and provide operationally sound alternatives to the indications give S

Recommended Question for Replacement

1700 ANSWER: #1700 COMMENTS:

a. SR trip blocked by P-6 b. trip occurs since 1/2 IR Instruments exceeds 25% and no block was accomplished at P-10 d. Hi flux rate trip is in PR instruments QUESTION: 1701 37. 033AA2.02 I A plant startup is in progress. Reactor power is 3%, control rods are in MANUAL, and Te,, is being maintained on the steam dumps when N-35B, Intermediate Range Instrument fails LOW. Which ONE (1) of the following describes the expected plant response, assuming no operator action? The reactor will trip on SR high flu The reactor will remain at 3% powe The reactor will trip on negative flux rat All Intermediate Range high flux trips will be disable S

1701 ANSWER: #1701 COMMENTS:

a. SR High flux trip is unblocked when BOTH IR <7.5 10'6%.

c. No longer trip signal d. Still below P-IO, cannot block N-36 tri (Rev. 1, Revised per MRF 90007. Distractor d. clarified - N36 IR trip Is operable, 9/14/94).

S

35. 032AA2.03 1

-Refueling operations are in progres SR N-31 and 32 read 15 cp Both IR Nis indicate off-scale LO PR N-41 is out of service, all appropriate blstables are trippe All other power range channels are reading 0%.

A failure has occurred on PR N-43 causing it to drift high to about 25%powe Which ONE of the following actions are required?

A. Immediately terminate all fuel movement In progress and emergency borate per AOP-1 0 "Emergency Boration."

B. Immediately terminate all fuel movement in progress and determine the boron concentration of the RCS at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Notify Maintenance to investigate the power range instrument failure and continue with the refuelin D. Place the Rod Stop Bypass switch for the failed PR channel to bypass and continue with the refuelin Bank Question From Farley Exam Ban Lesson plan IC-8 Nuclear Instrumentation, objectives IC-8-32,37, and 3 A. Incorrect, Terminating all fuel movement is correct, but AOP-106.1 Emergency Boration Is not require W B. Correct, with a loss of both source range detectors (Auto de-energized due to 2/4 >

P-1 0 technical specifications direct the these action C. Incorrect, lAW TS 3.9.2 refueling may not continue until the source ranges are returned to servic D. Incorrect, this would be the actions that would be taken if the plant was at power, and this will not allow the refueling to be continue p

Comments on questions we to consider for replacement 35. V.C. Summer uses the Gammametric designed SR/TR nuclear instruments and therefore any questions with regards to the automatic removal of high voltage are nonapplicable to the statio The alternative questions are in the same section of the K/A catalog; 4.2 GENERIC APE'.

.

......................

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Recommended Queston for Replacement 35. 032AA2.03 1 K/A catalog page 4.2-25 032 Generic Abnormal Plant Evolutions; Loss of source range nuclear instrumentation AA2.03 1 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation; expected values of Source range indication when high voltage is automatically remove Possible replacement questions - new K/A item requireds LORT 139 (modified)

35. New K/A number required: 032AA2.06 I confirmation of reactor tri A reactor startup Is in progress. The NROATC is in the process of verifying that the ALL RODS ON BOTTOM annunciator has cleared. A voltage spike on N-32 causes the Source Range Instrument to peg high and then fail low. The rods remain at their previous position and N-31 indicates normally. Given these conditions, the NROATC should: Insert a manual reactor trip and enter EOP 1.0, REACTOR TRIP/SAFETY INJECTION ACTUATION Bypass N-32 and continue with the reactor startup since Tech Specs only requires one of two sources ranges to be operable and N-31 indicates normall * Stop withdrawing rods since Tech Specs requires both SR channels to be operable and, with one channel inoperable, all positive reactivity additions must be suspende Shutdown the reactor per GOP 5, REACTOR SHUTDOWN FROM STARTUP TO HOT STANDBY Correct answer, A QUESTION: 1874 35. New K/A number required: 036AA2.02 I Fuel Handling Incidents: ability to determine and the following as they apply to the Fuel Handling Incidents; occurrence of a fuel handling inciden The following plant conditions exist:

The Unit is in MODE One (1) source range neutron flux monitor is out of servic The Source Range Audio Count Rate Drawer is out of servic Core alterations are in progres Which ONE (1) of the following Technical Specification Action Statements should be implemented? Suspend ALL operations involving positive reactivity change Recommended Question for Replacement Emergency borate the RCS until a boron concentration of 2150 ppm Is establishe Immediately evacuate the refueling area until the Audio Count Rate is returned to servic Replace the Reactor Vessel Head until both neutron flux monitor and neutron flux alarm are returned to servic #1874 ANSWER: S

Comments on questions we to consider for replacement 23. Procedurally, the only viable answer is to have rod control remain in manual. This an immediate operator action per AOP 401.10, POWER RANGE CHANNEL FAILURE. Per the AOP, rod control would NOT be restored to automatic until step 17, and only after the NI had been restored and bistables reset. Although choice C is true insofar as testing how the rod control system is actually designed, an operational exam such as this tests knowledge of procedures and their usage, whereas the question is written as if testing the construction of the system at a technician level. Additionally, with no rate of change in answers B and D, there would be no demanded rod motion and therefore no THEORETICAL reason to prevent restoration of rod control to automatic, although any of these choices would be in direct violation of current procedures. These choices therefore make the question confusing as three answers offer distractors which emphasize rate-of-change circuitry. Reference IC-5, Rod Control, page 23, specifically states that operation of JS408 is not covered by procedures. The alternative questions provided offer knowledge on the operational level of the interface between rod control and nuclear instrumentation, as well as the effects of nuclear instrumentation failures on rod control system operatio S

Recommended Question for Replacement 23. 015K3.06 1

K/A catalog page 3.7-5 015

-

Nuclear Instrumentation System K3.06 1 Knowledge of the effect that a loss or malfunction of the NIS will have on The following; Reactor regulating system Possible replacement questions QUESTION: 1153 The following plant conditions exist:

-

ROD CNTRL BANK SEL is In AUT Control bank D is at 220 steps

-

Power according to First Stage pressure is an equivalent g9 percen Power range instrument N-44 has been slowly drifting high at one quarter percent per hour for the past 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Which ONE (1) of the following describes the effect this failure will have on the rod control system? Rods move IN due to the power rate mismatc Rods move OUT due to the power rate mismatc Rods do NOT mov Rods move IN due to the T3.4T, mismatc #1153 ANSWER: QUESTION: 4103 (modified)

Which one of the following parameters determines the magnitude of the gain imposed by the Variable Gain Unit in the Reactor (Rod) Control Unit? Median select Tav Ni channel N-44 powe Turbine impulse pressur Power mismatch change rat #4103 ANSWER: Recommended Question for Replacement W

QUESTION: 2018 (modified)

Which ONE (1) of the following failures will cause automatic rod insertion? Median select T,, signal fails LO Rod Control Urgent Failur Power range N44 fails HIG............--

Firststag turbinel fpes-re fails HIG #2018 ANSWER: (1.00)

I

Recommended Question for Replacement 23. 015K3.06 1

K/A catalog page 3.7-5 015 Nuclear Instrumentation System K3.06 1 Reactor regulating system Possible replacement questions QUESTION: 1153 The following plant conditions exist:

-

ROD CNTRL BANK SEL is in AUT Control bank D is at 220 steps

-

Power according to First Stage pressure is an equivalent 90 percent

-

Power range instrument N-44 has been slowly drifting high at one quarter percent per hour for the past 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Which ONE (1) of the following describes the effect this failure will have on the rod control system? Rods move IN due to the power rate mismatc Rods move OUT due to the power rate mismatc * Rods do NOT mov Rods move IN due to the Ta%,Tr mismatc #1153 ANSWER: QUESTION: 4103 Which one of the following parameters determines the magnitude of the gain imposed by the Nonlinear Gain Unit in the Reactor (Rod) Control Unit? Median select Tav NI channel N-44 powe Turbine impulse pressur Power mismatch change rat #4103 ANSWER:. 006K2.04 1 O-Unit is at 100% powe The Electric Plant is in a normal full power line-u Which ONE of the following is the power supply to MVG-8808C ACCUM DISCH ISO A. IDA2Y B. IDA2X C. 1DB2Y D. 1DB2X Question from previous Farley Exam. (1999) Modified for Summe Lesson Plan AB-10, Emergency Core Cooling System. Objective AB-10-1 A. Incorrect, wrong train of powe B. Correct, this is the power supply stated in the lesson materia C. Incorrect, power for this valve does not come from 1D D. Incorrect, power for this valve does not come from 1D S

Comments on questions we to consider for replacement 10. Operations Management at V.C. Summer Station does not expect operators to commit S

power supplies to memory to this detailcd level. Although an operator should recognize that a ESF related valve would be on an MCC supplied from a class Ie bus, it has never been expected nor desired to have an operator commit each power supply to memory. Our procedures, specifically EOP's, capture the power supply and breaker location when the procedure calls for local operation. Furthermore, the operation of these components is so critical that the procedure should be references each and every time the components are operated. Any questioning that implies that power supplies should be memorized to this level of detail invites a mistake should an operator fail to have perfect memory and the implication be that their memory is the established requirement. It is considered acceptable for operators to understand the operation of the power lockouts and their operation is considered "skill of the crafr" and is therefore a required operator knowledge. We believe our replacements are more in-keeping with both Summer Station operations management as well as the NRCs philosophies and expectation p

Recommended Question for Replacement 10. 006K2.04 1 K/A catalog page 3.2-17 006 Emergency Core Cooling System K2.04 1 Knowledge of power supplies; ESFAS-operated valves Possible, replacement question 10. 006K2.04 1 The plant is operating at 100% power. All ESF electrical equipment is in its normal configuration. Which of the following "A" train valves can be energized by taking the TRN A PWR LCKOUT switch to the ON position?

A. MVG 8884 CHG LP A TO HOT LEG MVG 8706A RHR LP A TO CHG PP C. MVG 8808A A (Accumulator) DISCH ISOL D. MVG 8701A RCS LP A TO PUMP A Correct answer: A

Possible replacement question 10. 006K2.04 1 The operating crew is performing EOP 2.3, TRANSFER TO HOT LEG RECIRCULATION. All ESF electrical equipment is in its normal configuration. What steps must be taken to open MVG 8886 CHG LP B TO HOT LEGS?

A. Place MVG 8886 CHG LP B TO HOT LEGS to the open position B. Locally unlock and close the breaker for MVG 8886, then place MVG 8886 CHG LP B TO HOT LEGS to the open position C. Place TRN B PWR LCKOUT to the ON position, then place MVG 8886 CHG LP B TO HOT LEGS to the open position D. Locally open the breaker for MVG 8886, reinstall the control power fuses, close the breaker for MVG 8886, then place MVG 8886 CHG LP B TO HOT LEGS to the open position Correct answer: C S

Recommended Question for Replacement Possible replacement question - new K/A item 10. New K/A number required: 006K2.01 1 Knowledge of bus power supplies for the following; ECCS pumps The plant is operating at 100% power. A problem develops with XFMR 1 DAI & I DA2 FEED and the breaker trips open. Which of the following loads will be lost due to this fault:

A. "AX Charging Pump B. "A"RHR Pump C. "A" MDEFW Pump 0. "A* RB Spray Pump Correct answer: B S

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95. W/E1IG2.4.9 1 Given the following conditions:

- The plant is in Cold Shutdown with RCS temperature at 110 deg. RHR pump A and RHR heat exchanger A are in operatio RCS Hot leg level Is at 16 inches (mid-loop operations).

- RHR Heat Exchanger A Outlet Hlow Control Valve (FCV-605A) has Just stroked from 20%

open to full open due to a circuit faul If NO operator action is taken, which ONE of the following will occur to cause a loss of RHR cooling?

A. RHR pump overspeed trip from runout due to low discharge pressur B. RHR pump loss of suction due to vortexing at the RCS loop suctio C. RHR pump overcurrent trip due to high discharge pressur D. RHR pump overcurrent trip caused by pump runout due to low discharge pressur REF: Indian Point Exam 1996 AB-7 RHR SOP-115 RHR AOP-115.5 Loss of RHR with RCS not Intact (Mode 6)

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99. W/E15EA2.1 I-A Large Break LOCA has Occurre EOP-2.2 "Transfer to Cold Leg Recirculation" has just been complete The STA reports the following conditions:

-Reactor Building Pressure 2.0 psi Reactor Building Radiation 10 RIH RHR Sump Level 420 f Which ONE of the following describes the immediate containment concern and the correct procedure to enter?

A. Inadequate suction to the RHR pumps, transition to EOP-2.4 "Loss of Emergency Coolant Recirculation."

B. Erroneous instrumentation readings, transition to EOP-17.2 "Response to High Reactor Building Radiation Level," when desire C. Reactor Building structural integrity; transition to EOP-1 7.0 "Response to High Reactor Building Pressure."

D. Flooding vital equipment In the Reactor Building; transition to EOP-1 7,1 "Response to Reactor Building Flooding."

Modified from Diablo Canyon 99 exa Lesson Plan EOP-17.1 "Response to Reactor Building Flooding, objective 218 A. Incorrect, RHR sump level is adequate, Loss of emergency coolant recirculation is not the procedure that is required to be entered with these condition B. Incorrect Radiation levels are high, but EOP-1 7.2 is entered on operator discretion and sump level is a higher priorit C. Incorrect, Pressure is somewhat high, however it does not meet the threshold for entry (12psig). in a lorge break LOCA this procedure would have already been performed, and re entry is not require D. Correct, Reactor Building sump level is high and flooding is a concern and level has reached the threshold value to enter bOP-1 gr eolaso ;uap~sad jawwns Ll'd LeSS St'S £08 diý,ItTa 23 Le 2nu

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The following conditions exist

- Reactor power

- Control Bank D is at 137 steps withdrawn

- Rod control is in AUTO

-;q44 l4tubine ImpulsePr. nr okeuted-fo.rmyulk' h-rl,.o.J.tut"u ytm rRCP.tr SrSC Wcss 'Aw*40-4 t"A e-rII If PT-448 fails HIGH, how will the rods in Control Bank D respond?

A. Move inward at 48 steps per minut B. Move inwa*td at 72 steps per minut C. Move outward at 72 steps per minut D. Move outward at 48 steps per minut REF: Kowaunco Exam 19097 IC-5 Rod Control TB-5 Turbine Control and Protection System Distracter A - 48 SPM is the speed for manual operation of control banks and wrong directio Distracter B - inward movement is a misapplication of PT-485 failing hig Answer C - correct maximum speed of 72 SPM in the outward direction Distracter D - 48 SPM Is the speed for manual operation of control bank d1i:1o El so 2 nU c naOzO luapTsae JaWfns a'*

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Tonhours after the runback, i d i

tain Tavg on program and reactor powir constan Which ONE of the following describes rod motion requirements over the next TWO HOURS?

(Assume boron concentration is maintained constant.)

A. Rods will have to be periodically withdrawn since xenon concentration will follow its post-runb ack build-i4n rat B. Rods will have to be periodically Inserted since xenon concentration will be decreasing due to deca G. Rods will have to be withdrawn since the new power level will result in a high rate of xenon btild-i D. Rods will have to be Inserted since the new power level will cause a high rate of xenon burnou REF: Kewaunee Exam 1997 rAl W

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4. 002A3.01 1 The following conditions exist:

- Reactor power 100%

- RCS activity is elevated but below Technical Specification levels

- Pzr pressutreC 2 5 psig-Pzr levek*6e&

- An attempt has been made to reseat PORV "A", PCV-445A

- When conditions stabilize

- Reactor power 100%

- Pzr pressur, 28 psig Pzr level O Z'7 How would the operator be able to tell if the PORV has closed?

A. The PORV tailpipe temperature should first decrease and then begin to increase to alarm setboin Position lights for PCV-44A showing CLOSE indicatio C. Level change in RCDT 0. Lowerreadings for containment radiation monitors ne-u MOM2A

,lM-A2 REF: Sraidwood Exam1998 IC-3, Pressurizer Pressure and Level Control SOP-101, Reactor Coolant System

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p.1 Summer Resident Office 5. 003A2,01 1 With the plant operating at 100% power the crew took the following actions:

"- "RCP VLO" alarm annrm

-,ia P-617 2-4 was entered

- "Reactor Coolant Pump Se

,

-

1.2 was entered

- "Operatinal.

.

e, STP-114.002 determin a

ýI for 'A' RCP has faile Which one of the following actions should the crew take on determination of #2 seal for 'A' RCP has failed?

A. Immediately trip the reactor and secure 'A' RC B. Reduce reactor power to < 38% within thirty (30) minutes and then secure 'A' RC C. Isolate seal injection flow to the "A' RCP and continue normal plant operation provided RCP bearing temperatures are not exceede D. Continue normal plant operation provided 'A' RCP total #1 seal flow is between 0.8 gpm and gp REF: Summer Exam Bank Question #3657 Modified ARP XCP-617 2-4 RCP Standpipe HI/LO Alarm AOP-1 01.2 Reactor Coolant Pump Seal Failure Distracter A incorrect, action not required for #2 seal failur Distracter B incorrect, action not required for #2 seal failur Distracter C incorrect, Seal injection flow should not be isolated to any running RCP_

Re changed distracter "c" to "isolate seal injection flow" because as it was previously worded, an argument could be made for two correct answers.

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6. 003AA1.01 I Given theolwn 9c ant conditions:

-Reactor Power ies'/

- Bank D rods are at 55 steps a

.e(St),XControl Bank D rod was dropped and recovered

- TheflilseConverter was NOT reset per Step 14 of AOP-403.6, Dropped Control FR What effect will these events have on continued rod control system operation?

As control rods are..

o-r A. withdrawn, 49z: Tempcratur Delta T Will NOT stop Control Bank ) withdrawal when require Rajp,N~jM-tnol iMAn i r to - B. inserted, the RodJ I.srit 0z-it Alarm will be received at a lower actual rod positio ýA C. inserted, Bank C control rods will begin insertion at a lower value of Control Bank 0 actual positio L,,I-S'(C-IJ)

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D. withdrawn, the Bank D Rod Withdrawal 44i t

is wilt NOT elan before Control Bank 0 is fully withdraw REF: Prairie Island Exam 1997 Summer Cycle 13 COLR AOP-403.6 Dropped Control Rod IC-5 Rod Control Distracter A - Incorrect, the Pulse-to-Analog Converter does not input to the OTOT Answer B - Correct Distracter C - incorrect, Bank C control rods will begin to insert at a higher value of Control Bank D positio Distracter D - Incorrect, the Pulse-to-Analog Converter will think the Bank ) rods am higher than actua ZŽ"

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7. 004K6.05 1 Given the following plant conditions:

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at 100% power, with CVCS aligned for normal operatio " *

rifice isolation valve is in servic,-p$

VCT*

-,levetrlis 32%. nautomatcLT 1, VCT le e*l transmitter, fails hig VOT level is 32%.

.

-All controls are in auoaiLT11,VTeattnsterfilh Which one of the following describes the final actual VCT level?

(Assume no operator action.)

A. Increases to 71% and stabilize B. Increases to 100% (full).

C. Cycles between 20% and 40% due to auto-makeu D. Decreases to 0% (empty).

REF: Farley Exam 1998 AB-3 Chemical and Volume Control System Distracter A - Level will not reach 71% because auto makeup will stop at 40% and with LCV-115A open level will begin decreasin Distracter B - Level will initially decrease as LCV-115A diverts. Level will decrease until auto

Makeup starts when VCT level reaches 20% but will stop increasing when auto makeup stops at 40%.

Answer C - Level will cycle between 20% and 40% because makeup flow is greater than flow through LCV-1 1 5 Distracter D - Level will decrease because LVC-1 15A Is open until 20% level when auto makeup begins and causes level to increase.

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11. 007A4101 at 100% powe op 1 Arfstrument failure occurred an

..ORV opene The operator has taken the POR* to the dosed position, AO.CP 4'

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ý do AMflWS Which ONE of the following describes the initial indications the operator could use to verify thathe PORV was not fully closed after the PORV was placed indeiose.? -Me4 A. Tailpipe temperature would rise to between 200-300 OF, PRT pressure would rise, RM-wculd alar B. Tailpipe temperature would rise to between 500-600 OF, PRT temperature would rise, RM-A2 would alar C. Tailpipe temperature would rise to between 200-300 OF, PRT pressure would rise, VOT leve would lowe D. Tailpipe temperature would rise to between 500-600 OF. PRT temperature would rise, Reactor Building Pressure would increas Modified from bank question 54 Lesson Plan AB-2 Reactor Coolant System, Objective AB-2-1 A. Incorrect, RM-A2 would not alarm until the PRT ruptured.

B. Incorrect, tailpipe temperature will not rise to 500-600 degrees F, and RM-A2 will not until the PRT rupture C. Correct, this is the correct tailpipe temperatures, PRT pressure will rise and VCT will lower. (Could put charging system flow would increase)

D. Incorrect, RB pressure will not increase until the PRT ruptures.

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12. 007EA2.02 1 Given the following conditions:

Reactor power is at 100% steady state

2

Power Range NIS 102%

103%

102%

PZR pressure 1880 psig 1910 psig 2500 psig (455)

(456)

(457)

instrument numbers PZR level 72%

92%

90%

(459)

(460)

(462) instrument numbers Tave 584F 585F 582F S/G levels 43% (A) 34% (B) 89% (C)

(all S/G instruments for a SIG read the same level)

What is the FIRST required action for these conditions?

A. Trip the reactor and initiate actions of EOP-1.0, Reactor Trip/Safety Injection Actuatio B. Verify a turbine runback is initiate C. Reduce power to LESS THAN 100% indicated to ensure 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months average does NOT exceed 100%

powe D. Initiate a MANUAL Safety Injection and initiate actions of EOP-1.0, Reactor Trip/Safety Injecton Actuatio REF: Braidwood Exam 2000 EOP-1.0 Reactor Trip/Safety Injection Actuation IC-9 Reactor Protection and Safeguards Actuation System Trip setpoints Logic Low PZR Press 1870 2/3 High PZR Press 2380 2/3 High PZR Lvi

213 Low S/G Lvi

213 High S/G Lvi

2/3 on any SIG Answer A - Reactor Trip required by turbine trip on high SIG level in (0) S/G Distracter B - No indications provided that would require a turbine runbac Distracter C - Not the first action required, but could be focused on if Rx trip required not realize Distracter D - Not required by indication provide LEBS 91t'1 coo asojo 4uapTsaH jawwns cSd

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A /0 13. O08AKI.01 1 Which one of the following represents the conditions of the steam entering the PRT from a leaking PORV if pressurizer pressure is 1385 psig and PRT pressure is 5 psig? (Assume an ideal thermodynamic process)

A. Superheated steam 260-270 degrees F B. Superheated steam 240-250 degrees F C. Saturated steam 225-235 degrees F D. Saturated steam 275-285 degrees F REF: Mollier Diagram SOURCE: 2000 Summer Exam and Watts Bar RO exam 1998 Reference provided - Steam tables and Mollier Diagram If the Mollier Diagram Is used to determine the temperature, the applicant could start with the intersection of 1400 psia and the saturation line. A reasonable range or enthalpy at this intersection would be 1172-1175 btu/Ibm. Following a constant enthalpy of 1172 btulbm to the point where it intersects 20 psia (given PRT pressure of 5 psig plus 15 psia) will yield a temperature almost exactly half-way between 240°F and 280 0 Using the Steam Tables would yield a calculated value for an enthalpy of 1172 btu/Ib.

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16. 011EKI.01 1 GtGivepa llowing plant conditions:

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-7Je g *.,-.-Xew-e-e confirming that Natural Circulation exists Which one of the following conditions provides indication that natural circulation exists?

A. RCS Hot leg temperatures are trending to saturation temperature for steam pressur B. S/G pressures are slowly increasin C. RCS subcoling based on core exit TC's is 40 degrees F and slowly increasin D. The delta-T across the SIG's are 10 degrees F and slowly decreasin REF: Robinson Exam 1996 EOP-2.1, Post-LOCA Cooldown and Depressurization Distracter A - RCS Hot leg temperature should be stable or decreasing for natural circulation indicatio Distractor B - S/G pressure should be stable or decreasing for natural circulation indicatio Answer C - correct Distracter D - a 10 degree delta-T across the S/Gs would not indicate natural circulation

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- PZR level is at programmed level of 55% for current stable plant conditions-ALL systems are operating correctly in automatic wL+A

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What is the initial plant response

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PZR backup heaters energize and the proportional heaters arepff, the "PZR LOS DEV HI/La" annunciator actuates, and charging flow decrease S..PZR backup heaters enargize and proportional heaters are on. the "PZR LOS DEV HI/LO" annunciator actuates, and the chargingflow increase C. Charging flow increases to the new program level of 60% and there is no change in PZR heater status (proportional heaters are on and the packup heater

"D. Charging flow decreases to the new program level of 50% and there is no change in PZR heater status (proportional heaters are on and thebackup heaterw&epfl RER NEW IC-3 Pressurizer Pressure and Level Control ARP-001 XCP-616 PZR LOS DEV HI/LO

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Given the following plant conditions:

gleam Ian-Two minutes ago an MSIV inadvertently osed causing secondary safeties to lift and a reactor trip and safety injection due to lowp pressure signal to be generate The reactor trip breakers failed to ope A,64,etrh tsAt~rsW~'.

- The operators tripped the reactor by opening C rG2 t

ond fector bro, effi An the Red DrG?, MG Ccrtzl zaz- ý er EOP-1Kesponse to Abnormal Nuclear Power Generatio It is now desired to reset SI and secure SI equipmen RCS pressure is 1800 psi. Which one of the following will prevent resetting SI from the Main Control Board under these conditions?

A. RCS pressure is less than SI setpoin B. The Si timing retsys C. Permissive P-11 has actuate D. Permissive P-4 has not actuate REF: Farley Exam 1998 IC-9 Reactor Protection and Safeguards Actuation System EOP-1 3,0, Response to Abnormal Nuclear Power Generation EOP-1.0, Reactor Trip/Safety Injection Actuation Distracter A - P-11 is the set point above which a blocked SI signal will auto unblock. Being

below P-1I will not prevent resetting S Distracter B - SI will not reset if the 60 second timer is active, but the timer timed out 1 minute ag Distracter C - Pressurizer pressure below the SI setpoint will initiate an SI signal but will not prevent resetting S Answer D - With the reactor trip breakers closed the required P-4 signal will not be generated anc SI cannot be reset from the MC.4 5"S/,VC/oA -e 7 4 5:,t dl:1o1 E3 sa 2nuj Vcud LEGS st's COB ojotupS~

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20 013K4 I161 YEOP-16.0, Response to Imminent Pressurized Thermal Shock,wth3RCP runnin Which one of the following actions is correct in order to avoid, or limit, thermal shock or pressurized thermal shock to the reactor pressure vessel?

A. -StabiiG RCS pressure a

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B. Cooldown at maximum rate using the steam generator C. Isolate the accumulator D. Stop all reactor coolant pump REF: Cook Exam 2001 EOP 16.0, Response to Imminent Pressurized Thermal Shock EOP-16.1, Response to Anticipated Imminent Pressurized Thermal Shock To offset thermal stress caused by cooldown, the cooldown must be stopped, temperature stabilized and pressure reduced, Distracter A - pressure should be reduce Distracter B - cooldown should be minimize Answer C - Isolate all SI Accumulators to prevent injection of cold water that could cause additional thermal stresses Distracter D - RCP are restarted if stopped and restart criteria me _

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21. 014A*1.03 1 Which one of the following is used as the reactor power input to theAsertioAimit (RIL)

computer'

A. Median Selectedt-9e,11z r7 B. First stage impulse pressure C. Calculated Thermal Power D. Calculated Steam Flow RFF: Cnnk Exam 2001 IC-6 RCS Temperature Indication System The median selected I Tavg is sent to the rod insertion limit programmer aoL330 4uapTsaN jawwns jtj:jO a) ge 2nu T'd'

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22. 015AK3.03 1 LGiven the following conditions:

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Reactor power is at 100% steady stat ROE'"B" VIBR HIalarm ha lit

- Excessive vibration is confirmed on RCP "B" with 20 mils shaft vibration and eased

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.""-WlhONEo h olwn statements is correct regarding the course of action required?

A. Reduce Reactor power to less than 38% (P-8 permissive is illuminated) and secure RCP "B".

B. RCP "B" should be tripped prior to a reactor trip to minimize pump damag C. Trip the reactor prior to tripping RCP "B" to prevent an automatic trip and unnecessary challenge to a safety syste D. RCP "B" should be tripped per SOP-101, Reactor Coolant System, and proceed to Hot Standby per GOP-4, Power Operation, and GOP-5, Reactor Shutdown from Startup to Hot Standb REF: ARP-001 Panel XCP-618 Annunciator Point 1-3 and SOP-101 Distracter A - incorrect, reflects the method for removing one reactor coolant pump from service per SOP-1 01 Rev 22, Reactor Coolant Syste Distracter B - incorrect, with reactor power >38% and Shaft vibration > 20 mils and Frame vibration increased the reactor should be tripped first then RCP "B".

Answer C - correct, per ARP-001 Panel XCP-618 Annunciator Point 1-3 Distracter D - incorrect, ARP-001 Panel XCP-618 Annunciator Point 1-3 actions for Reactor Power <38%.

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An OPEN has developed in a thermocouple used by the Subooling Monitor. What impact will the failed thermocouple have on the Subcooling Monitor after steady state conditions are reached?

A. The-bgnolepq mny-w9!indicate >20,-degteep subcoolin B. S

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25. 022A1.04 1 Given the following:

- The plant is operating at 75% reactor powe The Reactor Building sump was pumped down to the Floor Drain Tank twent*.

minuteE ag Which one of the following would provide an alarm for a 0.7 GPM leak from the reactor coolant system to the Reactor building?

A. Reactor Building Sump level B. Reactor Building Radiation level C. Reactor Building Temperature C

U4 D. Reactor Building/zoolingynit condensate drain flow REF: GS-7 Leak Detection Distracter A - incorrect, sump level would provide indications of leaks >1 OGPM Distracter B - incorrect, radiation level will cause an alarm when leakage exceeds I gpm Distracter C - incorrect, temperature may not increase until leakage is excessiv Correct D - condensate drain flow will alarm when leakage exceeds 0.5 GPM dsz:xo 2) 8a 2nu Lt'd tCBS St'E £08 aoljjo 4uap~saN Jawwns

26. 022AA2.03 1-Tlj-Plant Is at 100% powe e3I Makeup System is in AUT "od Control is in Manua /

-The Reactor Operator notices that Tavg has decreased 2 O Which ONE of the following could contribute to decrease in RCS temperatur A. The mixed bed demineralizer is depleted and is no longer removing boron ion B. A newly replaced CVCS mixed bed demineralizer was put in servic C. The Boric acid filter is clogged preventing boric acid from mixing in the blende tb ICflq-'

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D. FCV-113A, Btrvctd has failrcpntrctV

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27. 024AA2,02 1 A plant transient has resulted in a condition that requires rapid emergency boration. The NROATC begins emergency boration vi tE:

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Fl-t10, EMERG BORATION FLOW, read 0 gp Investigation reveals MVG-8104, EMERG BORATE, will not ope Charging flow is 50 gp What operator actions are required next to supply boric acid flow to the RCS?

A. Transfer charging pump suction to the RWS B. Open, Fcr..,A0 t~'R aS C43,Uh F-CV-0

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JP9 VLV D_ Set lewn.-.'.... i to maximum to automatically borat FeV 113 ASSB, B.4 FioW ccn4roJer REF: Summer Exam Bank #504 Distracter A - Incorrect; Insufficient charging flow rate at RWST boron concentration (<2,500 ppm).

Answer B - Correct Distracter C - Incorrect; 168B should be verified CLOSED and opening 8439 will not borate through the blende Distracter D - Incorrect; Ineffective unless auto makeup in progres Rev. 1, (04/12193) Reformatted stem and given information to eliminate "lookup"; ARP was originally given in stem. Revised COMMENT because original comment was not justifie Rev. 2, (05/12/93) Replaced "Transfer charging pump suction to the RWST" with "Align gravity drain from the BATs to the charging pump suction" due to Ops. Rep. concerns that original choice a was too defendable as an alternative; therefore, two answers were potentially correc Rev. 3, (06118193) Restored original choice b. from Rev 1. Reduced charging flow to 50 gpm to ensure that flowrate would be insufficient to achieve an equivalent boric acid fiowrate of 30 gpm at 7000 pp Rev.4, (dow 01121/02) changed wording in stem to "a condition that requires rapid emergency boration" to agree with the latest rev of AOP-106.1.

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28. 026AA2.06 1

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Givendthe following conditions exist:

- A small break LOCA to containment has occurred from 100% power

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,e A. Reactor Coolant Pump B. Charging Pump C. Residual Heat Removal Pump D. Reactor Building Spray Pump REF: Summer LO 4783, AOP-118.1 Rev 2 CFR 55-41(7,8,10) apply Source: Kewaunee 1997, correct answer and distracter d changed to be specific to Summer Distracter A Any running RCP should be stopped within 10 minutes or if a temperature limit associated with the RCP is reache Answer B Any running Charging pump should be stopped within 1 minute Distracter C RHR pumps should not be run longer than 90 minutes without CCW flow Distracter D Not supplied by CCW 02 ci Less sile £08 eoiaso quep~sad uawwns dsto

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been directed to enter EOP-14.0. Response to Inadequate Core Cooling, due to a red path on-a'Ce Core Cooling critical safety function status tre RCS pressure is approximately 1100 psi ROPs are secured with the "A" charging pump,/providing seal injectio Component Cooling Water (COW) flow has bben lost to the containmen In Step 3 of EOP-14.0, it is noted that CCW is not available to the RCP Which one of the following actions should the Control Room staff undertake?

A. Disregard the lack of CCW flow to containment, and continue with EOP-1 B. Stop at this point in EOP-1 4.0; do not continue until an RCP is runnin C. Continue with EOP1 4.0 while attempting to reestablish CCW flow to the RCPs as personnel resources permi. Complete actions to depressurize the steam generators to 140 psig, and then attempt to reestab ish CCW flow to the RCP A REF: Summer Exam Bank #387 CCW should be restored If possible to preserve RCP availabilit Procedure may restart pumps w/o support conditions as a 'last ditch" effor Consistent w/general guidance when unable to complete a ste Inconsistent "w/sequence of procedur Requires examinee to interpret and apply afternative action listed after determining support conditions are not met per SOP-10, 027AZ1 I1 rPT-444, CNTRL CAN PRESS PSIG, indication fails lo What ar e ime diate actions that should be taken per AP-401.5, PtSSUri7er Pressure Control Channel Failure?

A. Compare P1-444 and PI-445 control channel pressure indications normw lVerify the PZR PORVs are closelEnsure Rod Control Bank select switch is in Aut B. Ensure Rod Control Bank select switch is in AutVaintain RCS pressure between 2220 and 2260 C. Verify the PZR PORVS are closed, Compare PZR control channel with protection channel indicatior6Pheck PI-444 control channel pressure indication norma D. Maintain RCS pressure between 2220 and 2250 ps ompare PZR control channel with protec:ion channel indications REF: AOP-401-5 rev 3 Distracter A PI-445 control channel pressure indications normal and Ensure Rod Control Bank select switch Is in Auto are supplemental step Distracter B Ensure Rod Control Bank select switch Is in Auto and Maintain RCS pressure between 2220 and 2250 psig are supplemental step Answer C Correct, three immediate actions Distracter D Maintain RCS pressure between 2220 and 2250 psig is a supplemental ste LE s ts Cos ao,,,ij=O,U~

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34. 029EK3.01 1-The (ýis at 100% powe A tofal loss of feed has occurre Steam generator 1o-0o level alarms have come i An Automatic Reactor Trip did not occu A Manual Reaotor Trip is initiate Which ONE of the following describes a correct method of verifying that the reactor is tripped, and the reason for tripping the reacto A. Verify Rod all bottom lights lit, OR RCS Temperature trending down; to ensure an RCS over pressurization event will not occur, B. Verify all reactor trip AND bypass breakers open, AND SUR e; to ensure only decay and RCP heat are being added to the RC C. Verify Reactor Power trending down. AND RCS Temperature trending down; to ensure an RCS over pressurization event will not occu D. Verify Reactor Power trending down OR All rod bottom lights lit; to ensure only decay heat and F CP heat is being added to the RC EOP 1.0 "Reactor Trip and Safety Injection", and EOP-1 3.0 "RESPONSE TO ABNORMAL NUCLEAR POWER GENERATION". Lesson Plan EOP-13.0 objective 204 A. Incorrect RCS temperature trending down is not an indication of a Reactor trip, and this is the wrong reason according to the WOG and lesson pla B. Correct, these are indications that a reactor trip has occurred, and this is the correct reason for performing the trip lAW the WOG. and Lesson Pla C. Incorrect, Reactor Power trending down is one indication that a trip may have occurred, but is also an indication of just a down power condition, and temperature can be Indications of the sarrie thing, and this is not the correct reason for verifying the reactor trippe D. Incorrect, the procedure requires both of these actions to be performed to verify that the reactor is trippe r Ae 9, 3~ 3 LCOSSt'SCOB aotSJO ¶jUapTsad.40wwfl tyi s

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40. 036AK1.0 1 The following plant conditions exist:

K - MODE 6 with CORE ALTERATIONS in progres The REFUEL CAV LVL H11LO annunciator is actuate RMG-17A & B (RB Manipulator Crane monitors) have high radiation alarm The SFP gate is installe Which one (1) of the following would require immediate evacuation of the Reactor Building per AOP 123.1, 'Decreasing Level in the Spent Fuel Pool or Refueling Cavity during Refueling'?

A. Low pressure alarm on the SFP gate boot seal &. Leaking of the SF C. Readings on RMG-17A

h D. Actuation of the SFP LVL HI/LO annunciato REF: Summer Exam Bank #1758 Oistracter A incorrect, since corrective actions can be taken to repressurize the seal without evacuation of R Distracter B incorrect, incorrect because SFP can be isolated from RB even if it is leakin ANSWER C Distracter D Incorrect, incorrect because SFP can be isolated from RB even if it is leakin CORE ALTERATIONS was used vs. fuel shuffle to eliminate questions about the credibility of the SFP gate being installed during fuel movement. An initial condition of "SFP gate installed" makes choice"a" a viable destructo AOP-123.1 Caution states that RB should be evacuated if dose rates > 20 R/hr.

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42. 038EK3.08 I-A SGTR is in progress on the 'B' S/ e a Crew has implemented EOP-4.0 "Steam Generator Tube Rupture."

(.J-943 indicates 200 gp,(ZIIE>

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" he crew is at the step for determining the required core exit thermocouples-The Reactor Operator roporte that RCS procsure h2s resoherd 1340 psi Which ONE of the following describes what action should be taken next and why?

A. Trip all RCP',fCP's should be tripped anytime during EOP-4.0 if the trip criteria is me B. Do not trip RCP'1Trip criteria does not apply and a controlled cooldown is imminen C. Trip all RCP',he trip criteria has been met and injection flow has been verifie D. Do not trip RCPJCP trip criteria only applies prior to isolation of the steam generato Modified Bank Question # 292 open reference ban Lesson Plan EOP-4.0 objective # 191 A. Incorrect, RCP's should be tripped but the trip criteria only applies prior to a operator controlled cooldow B. Incorrect, the trip criteria is met and a cooldown has not be commence C. Correct, the trip criteria is met and Injection flow has been verified, and an operator controlled cooldown has not been starte D. Incorrect, the RCP trip criteria applies up until the point that a controlled cooldown has be started.

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43. 039KI.02 1 Given th.followiflQ conditions:

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- TStheam Dump Mode Selector switch is selected to'-q'avg"

/ - PORV Manual/Auto stations are selected t ll~orn*oatir M*erle-

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Answer A - Correct, PT-2010 provides the pressure input to PrurJv IovUUi ctil-,,.

Distracter 8 - "B" SIG power relief will not open in relief mod Distracter C - "B" SIG power relief will open in Tavg control mod Distracter D - "B" SIG power relief will not open in relief mode.

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'Ainitfls at 80% power kA.ilure in the steam dump control circuitry causes the bank one steam dumps to ope The operator immediately takes the Train A and B Steam Dump interlock bypass switch to OFF-RESET to close the valve One of the valves falls to clos Which one of the following describes the approximate power level that the plant will reach, and what action(s) will mitigate the event?

A. Power will rise to 92% and stabilize; an emergency boration should be commenced to reduce power, until valve can be isolate B. Power will rise to 86% and stabilize; an emergency boration should be commenced to reduce power, until valve can be isolate C. Power will rise to 92% and stabilize; turbine load must be reduced to lower power, until valve can be isolate D. Power will rise to 86% and stabilize; turbine load must be reduced to lower power, until valve can be isolate Modified from Summer Bank question #257 Lesson Plan, IC-1 Steam Dumps, objective # IC-1-2 A. Incorrect, with only one valve open power should rise about 6%, and turbine load would

have to be reduced to reduce powe B. Incorrect, this would be the correct power rise, however turbine load should be reduced to tower power until the valve can be isolate C. Incorrect, this would be the power rise if both valves were ope D. Correct, this is the correct power rise and action to take to reduce power.

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46. 054AA1.01 1-A plant startup is in progres MDEFW pumps are being used to control S/G level The 'A' MFI 3start-

.

-IA MDEFW pump has been secure 'r-.

Just prior to securing the 'B' MDEFW pump the i

trip, -Immediatele following annunciators illuminat " Y-?-"EFP SPOT HOR PRESS LO XFER TO SW"

-"MD EFP A (B) SUCT PRESS LO" Which ONE of the following describes the correct status of the EFW system based on the above conditions?

A. Both MDEFW pumps willruto starel the EFW FCV's will get a full open signal, and suction will transfer to SW immediat l B. setfrMDEFW pumpS willrunning, all the EFW FCV's will get a full open signal, and suction will transfer to SW after a 5 second time dela C, 4M~eh MDEFW pumpS willining, all the EFW FCV's will remain as is, and suction will transfer to SW after a 5 second timd dela D. Both MDEFW pumps wilIauto start all the EFW FCVs will remain as is, and suction will transfer to SW immediatel A Lesson Plan 18-3 Emergency Feedwater, objective # IB-3-13 and 1 Modified from a bank question from the Summer Bank, and Watts Bar Ban A. Incorrect, The B MDEFW pump is already running, the A pump will auto start, the FCV's will not get a full open signal in this condition, and the transfer is delayed 5 second B. Incorrect, the FCV's will not get a full open signa C. Correct, both pumps will be running, the FCV's will be as is and the transfer has a 5 second time delay, D. Incorrect, B pump is already running, and the transfer is delayed 5 second a c

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47. 055EG2.4.16 1 The following plant conditions exist:

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- Operators are performing immediate Operator Actions0!

")of EOP-1.O,

Reactor TriP/

a Safety Injectio A RED path condition exists on HEAT SIN Which ONE of the following actions should be taken if ALL power is lost to the AC emergency busses?

A. Immediately, transition to EOP-6.0, Loss of All ESF AC Powe B. Complete f of EOP-t.0 and then transition to EOP-1 5.0 Response to Loss of Secondary Heat Sink C. Immediately, transition to EOP-15.0 Response to Loss of Secondary Heat Sink D. Complete l; of EOP-1.0 and then transition to EOP-6.0. Loss of All ESF AC Powe REF: Summer Exam Bank #1894 d61:TO

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48. 055G2.4.11 I-Nis incdicate*Z31/,p -per. *

4,0-Turbine Load IlZ-1t15NW CVP A/BIC TR'lP-aihnunciator is li Condenser Vacuum is 4.5 inches Hg absolute,,

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Which ONE of the following describes the correct actions to be taken to mitigate this event?

A. Trip the Turbine and go to AOP-214-1 "Turbine Trip."

B. Start the standby Main Condenser and Auxiliary Vacuum pump and reduce turbine load to 20% E t 5% per minut A Aa C. Trip the reactor, trip the turbine, and enter EOP-1.0, "Reactor Trip/Safety Injection Actuation."

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P D. Start the standby Main Condenser Vacuum pump andauxiliaryyrcuum.limP, irnaarrduntr

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ne /ods.3,,o Alle.,p 7%&- 754-rune Modified from Bank question #436 Lesson Plan AOP-206.1 "Decreasing Main Condenser Vacuum", objective # 302 AOP-20 A. Incorrect. conditions for tripping the turbine is < 300MWe and main condenser pressure > 5 inches absolut B. Incorrect, Starting the vacuum pumps are correct however a caution prior to step 4 instructs the operator not to reduce load to 30%.

C. Incorrect, Conditions are not met for a reactor trip at this time.(< 50% power)

D. Correct, lAW AOP-206.1 the operator should start the vacuum pumps if pressure does not decrease then the RNO would have the operator trip the turbin C25,.o A

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- A failure of a card in the process racks causes the deaerator startup drain valve too go fu, open)

Which one ofthe Followng descries the affect on the main feedwater system? (Assumefloperator action is taken.)

A. Feedwater gooster and Feedwater pumps tri B. Feedwater eooster and Feedwater pumps will not tri C. Feedwater gooster pumps do not trip and Feedwater pumps will tri D. Feedwater pumps "C" and "B" will trp and Feedwater pump "A" will not tri REF: TB-6 Condensate System TB-7 Feedwater System The deaerator level is drained to the condenser. Deaerator storage tank Lo-Lo-LO Level (2 of 3)

trips the feedwater booster pumps and feedwater pumps.

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- The plant tripped from MODE 1

- Voltage on Bu4i DA and IDB is zero

- EOP-6.0, Loss of All ESF AC Power, has been entere An SI signal has been generate Attempts to restore ESF power have been unsuccessfu All ESF equipment has been placed in pull-to-loc IF DO power supplies start degrading~hich EOP provides direction to meet the conditio2fnd why are the actions, if any, necessary?

A A. EOP-6.2, tLoss of All ESF AC Power Recovery with SWequired'; no specific actionsrequired for DC power suppliesjreqifed Auxiliary Building batteries are designed for this conditio B. EOP-1.5, 'Rediagnosis'; actions serve to maintain DC voltage above 103 VD C. EOP-6.0, 'Loss of All ESF AC Power'; actions serve to maintain DC voltage above 108 VO D. SAM~eenditions are outside design bases and actions will be dictated by TS REF: EOP-6.0,'Loss of All ESF AC Power'

GS-3, DC Power Distracter A - incorrect, Operators would remain in EOP-6.0 and EOP-6.0 provides direction to maintain DC voltage above 1.8 VD Distracter B - incorrect. Operators would remain in EOP-Answer C - correct EOP-6.0 provides direction to maintain DC voltage above 1.8 VD Distracter D - incorrect, direction to minimize DC loads is provided in EOP-6.0.

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'/Given the following plant conditions:

- Thplant as operating steady-state at 1 00% powe "- 'B'ieseljenerator was being load tested for periodic survell*ana-estle§.

- A reactor trip coincident with a complete loss of the grid has just occurred, Safety Injection was NOT actuate * All offeita power to the cefeguards buses has bppn los The Train A ESF loading sequencer (ESFLS) malfunctioned and initiated NO action NO operator actions have been taken in response to the Loss of Offsite Powe deleerator is still running with its output breaker still close Which one of the following states the operating status of the Emergency Feedwater (EFW) pumps one minute after the plant trip and loss of offsite power occurred?

A. Only TDEFP running B. Only "A" MDEFW and TDEFP running C. Only "B" MDEFW and TDEFP running D. Only "B" MDEFW running REF: Summer 1998 Audit Exam RO98001 Distracter A - incorrect, 'A MDEFW will not start Distracter B - incorrect, 'A' MDEFW will not start Answer C - correct, TDEFW pump start occurs due to monetary loss of IDA and ID 'B' MDEFW starts undervoltage on the associated bus (XSW1 DB).

Distracter D - incorrect, TDEFW will start also.

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59. 064A4.02 I The following plant conditions exist:

fA loss of all ESF AC power has occurre *

Bus 1 DA Normal Feed breaker and Bus 1DA ALT FEED breaker open, but DIG 'A' fails to start automaticall DIG 'A' is locally started, but the Bus 1 DA DG FEED breaker fails to clos The local 86 lockout relay has not actuated and the condition of the 'A' Diesel Local Control Panel Status lights is as follows:

""READY FOR LOAD" - Not Lit

"READY FOR AUTO START" - Not Lit

"EMERG. START" - Bright

The IB AO reports that D/G speed is 508 RP The IB AO reports DIG voltage Is "470 volts Which ONE of the following conditions could be preventing the Bus 1DA DIG FEED breaker Yorn utomatically closing and what action would correct this condition?

Diesel Generator relays not reset, reset generator relays locall B. Diesel Speed below minimum, adjust DIG speed using local speed adjus C. Diesel Control Mode switch in local, place mode control switch in remot D. Diesel Voltage below minimum, raise D/G voltage using local voltage adjus Modified from open reference bank question 56 Lesson plan 18-5 Diesel Generator System, objective # lb-S-1 A. Incorrect, only the 86 lockout relay will prevent the breaker from closing on emergency start and the 86 relay has not picked u B. Incorrect, speed is greater than 504 rpm, therefore this will not prevent the breaker from closin C. Incorrect, the position of the mode switch has no effect on the breaker on an emergency star D. Correct, Voltage < 90% will prevent the output breaker from closing.

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63. 071 K3.05 1-A waste gas release

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-RM-A10 is In servibut has failed tVduring te releas Which ONE of the following describes the effect of this failure on the release in progress?

A. If its setpoint is exceed._G10 Auxiliary Building Waste Gas Decay Tank Area monitor will al rrm and close HCV-014 antinate the releas B. The release will be monitored by RM'GI0 Auxiliary Building Waste Gas Decay Tank Area. but nc automatic actions will occu C. If its setpoint is exceede-A3 Main Plant Vent Exhaust monitor will alarm and close HCV-014 and terminate the releas D. The release will be monitored by RM-A3 Main Plant Vent Exhaust monitor, but RM-A3 has no automatic function Lesson Plan GS-9 Radiation Monitoring Systems, objective GS-9-1 A. Incorrect, RM-G1 0 monitors the waste gas decay tank area and would not indicate upscale conditions unless a leak was in progress, and has no automatic action B. Incorrect, RM-GI0 monitors the area but should not upscale unless a leak occur C. Correct, if its setpoint is exceeded RM-3A will close HCV-01 D. The release will be monitored by RM-3A, however it does have automatic function *,

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66. 076A2.02 1Ji-XCP-604 1-2 "SWP A/C TRIP" alarm Annunciator XCP-604 1-4 "SWP NO DISMPRESS LOW" alarm Annunciators XCP-605 1-4 "SWP B/C DfSi PRESS LOW" alarm "

-PI-4402, Service Water Pump "A" Discharge Pressure, indicates JW'psi PI-4422, Service Water Pump "B" Discharge Pressure, indicates 48 psi "A" SW pump will not restart Which ONE of the following describes the actions that should be taken t",Wate this event?

A. Enter SOP-1 17 "Service Water System", and zzvaza any runn*in. dit

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B. Enter AOP-117.1 "Total Loss of Service Water", and trip all RCP C. Reference

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D. Roference AOP-1B.11"Total Loss of Component Cooling Water", and isolate all CCW load Modified from Bank question# 310 (NEED TO CHECK SW PRESSURE VALUES)

Reference: AOP-1 17.1 "Total Loss of Service Water."

A. Incorrect, SOP-1 17 will be referenced, however if diesel generators are required, fire service water would be aligned to suppl B. Incorrect, AOP-1 17.1 is a procedure to enter however the procedure allows stopping up to 2 RCPs when plant conditions permi C. Correct, AOP-117.1 directs the operator to refer to SOP-117 and to start the spare service water pump, this would mitigate this even D. Incorrect, AOP-1 18.1 may be referenced, but only unnecessary CCW loads would be isolate e-ardC

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67. 076AK2.01 What type of detectors are used fog)A-L1-Primary Coolant ILetdown Monito-npe-for monitoring high reactor coolant activzpd where are the sensinog'fcfation for RM-1?

A. A scinf-aton and a Geiger-Mueller detector, located npdr the letdown line upstream of the BTRe B. Two scintillation detectors, located O r the letdown line upstream of the BTRS Derw)

C. A scinrtflhItlon and a Geiger-Mueller detector, located nprr the letdown line downstream of the B! RS De V

)-rt e ed w in o ntea ft eB 0. Two Geiger-Muelle( detectors, located 4r the letdown line downstream of the BTRS De REF: Kewaunee exam 2000 General Systems GS-9 Radiation Monitoring System Rev 6 Answer A - correct Distracter B - Incorrect, Use of two different detectors with overlapping range Distracter C - Incorrect, location is upstream of the BTRS Demin5 Distracter D - Incorrect, Use of two different detectors with overlapping ranges and location is upstream of the BTRS Demin 422:10 a:) 82 2nu o08 330 juapTsad Jewwns LESS Si, 608

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r-The TD EFW pump is Tagged out for bearing replace en p_ raveust been dp ed inoperable due to a cornnon9 proble Which ONE of the following describes the actions that must be taken?

A. Restore at least one EFW pump to operable status or be inAiotsthutdown within 1 hou B. Immediately trip the reactor and initiate safety injection, using the Main Feedwater pumps to maintain S/G level C. Restore at least one EFW pump to operable status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and a second EFW pump within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , or be in Hot Standby in the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months D. A'nitiate actions to restore at least one EFW pump to operable status as soon as possibl Bank Question, from Farley exam ban T _T.8-3.7.1-2 A. Incorrect, this is not the action called for in the B. Incorrect, a reactor trip Is not required by the C. Incorrect, TS does not allow for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore one EFW pump to servic D. Correct, the TS directs to initiate actions to restore one EFW pump to operable status as soon as possibl '-s%

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72. G2.1.29 I-The plant Is MODE 6 preparing to startu AValve lineup is being conducted on the CVC Several aalves are reported to need verification in the open position at the 41' lev Reactor

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Buildin None of these valves are designated "IV Exempt."

-Local Radiation levels are about 50 mr/h The initial positioner received 25 mrem performing his portio Which ONE of the following courses of action is preferred'?

B. Order verification be performed by Health Physics to ensure ALARA complianc C. Waive the independent verification to reduce personnel exposures by verifying correct position by alternate mean D. Have the verifier with the most experience perform the independent verifications in the shortest amount of time possibl Bank question from 1992 Summer NRC Exa SAP-1 5 A. Incorrect. SAP-1 53 directs the shift supervisor to waive the IV if the dose received will be greater than l Omr, and should consider verifying position by alternate mean B. Incorrect, SAP-153 directs the shift supervisor to waive the IV if the dose received will be greater than I Omr, and should consider verifying position by alternate mean C. Correct, this is the action that i directed i SAP-15 incorrect, SAP-AI* directsthe jift supervisor to waiive the IV if the dose received will be greater than 1 Omr, and should naider verifying position by alternate mean jU

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76. G2.2.191 Which ONE of the following descrihbe the type of work that would be ranked as a PRIORITY 1 Maintenance Work Request (MWR)?

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io B. Major plant modification item C. Compliance items to ensure that a 72-hour action statement time limit is not exceede D. Improvement items for plant efficienc *REFERENCE 1992 Summer NRC exa. SNS SAP-300, Rev. 8 2. SNS SAP-601, pp. 3, & A. Incorrect, lAW SAP-300 priority 1 MWRs are those that must start Immediately and be worked through to completion including the call out of maintenance personnel and the establishment of shift work if necessar. Incorrect, lAW SAP-300 priority I MWRs are those that must start immediately and be worked through to completion Including the call out of maintenance personnel and the establishment of shift work if necessar C. Correct, this meets the definition in SAP-30 D. Incorrect, lAW SAP-300 priority I MWRs are those that must start immediately and be worked through to completion including the call out of maintenance personnel and the establishment of shift work if necessary.

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81. G2.4.10 1-Unit is at 33% reactor power following a start-u Annunciator XCP-617 point 1-5 "RCP A LOW OIL RESVR LVL HI/LO" is i /a4 is-Electrical Maintenance has determined that the oikel i

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t, Which ONE of the following describes the appropriate action(s) to be taken?

A. Immediately Trip the Reactor and secure RCP "A' in accordance with SOP-1 01 8. Secure RCP "A" in accordance with SOP-101, and be in Hot standby per GOP-4 and GOP-5 within one hou C. Monitor bearing temperature, if bearing temperature exceeds 1950F then secure RCP "A" in accordance with SOP-i01, and be in Hot Standby per GOP-4 and GOP-5 within one hou D. Monitor bearing temperature, if bearing temperature exceeds 195°F then immediately trip the recictor and secure RCP "A" in accordance with SOP-10 ARP XCP-617 annunciator point 1-AB-4 Reactor Coolant Pump Lesson Plan enabling objective AB-4-2 A. Incorrect, this would be the correct action if power was greater than 38%, and there was no level In the sight glas B. Incorrect this would be the correct action if there was no level in the sight glass with current plant conditions, C. Correct, with the level high but still in the sight glass the ARP has the crew monitor the lower radial bearing temperature and If It exceeds 195 degrees F then secure the pump and proceed to hot standb D. Incorrect, this would be the correct action if the reactor was greater than 38%.

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gweP 82. G2.4.11 1-The operating RHR pump flow and amps aee....

-AOP-1 15.1 "RHR Pump Vortexing' has been entered.

Mor Which ONE of the following would require tripping the running RHR pump?

A. RHR temperature rises to 215 0 B. RCS Hot Leg Level decreases to 15 inche C. RHR flow is reduced to.....

-,o,2,,n, D. RCS Pressure decreases to less than 50 psi Modified from VCS bank questions 2355 and 187 VCS AOP 115.1 Lesson Plan objective 227 A. Incorrect, the procedure directs the crew to implement STP-1 03.001 and to monitor Hot leg temperature, but does not direct the tripping of a RHR pump for these conditions.

B. Incorrect, the procedure directs the crew to trip the running RHR pump if level drops less than 14 inche C. Correct, the caution prior to step one states that the RHR pump should be limited to than 30 minutes with less than 1000 gpm flow rate.

to less D. Incorrect, at mid loop pressure In the RCS should already be less than 60 psi I'0 C

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90. W/E05EK3.2 I-EOP-l 5.0 "Response to loss of Secondary Heat Sink" has been entere Attempts to establish EFW Flow have faile.

.

Which ONE of the following is the primary reason for securing the RCP's at this point in the procedura?

A. This will establish natural circulation conditions and will tend to mitigate the transien B. They are secured to prevent the heat added by the RCPs from advetsely affecting indications us ad to determine whether or not RCS bleed and feed will be require C. This will reduce RCS pressure to ensure subsequent SI flow is adequate for ECCS requirement D. They are secured to reduce the heat input from the RCPs, thereby delaying the need for bleed and feed and gaining time to establish a means of supplying FW to a S/ Bank Question Modified some what from a Farley version of same questio Lesson Plan EOP-15.0 Response to Loss of Secondary Heat Sink. Objective # 209 A. Incorrect, natural circulation conditions will not mitigate this transient with out water in the Steam Generator B. Incorrect, The heat added to the indications by the RCPs will not have an effect on whether bleed or feed will be require C. Incorrect, RCS pressure may be reduced some but by itself this will not ensure that SI flow will be adequat D. Correct, Securing RCPs will reduce the heat input to the RCS and delay the need for going to feed and blee /iV-7S

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92. W/EO8EKI.1 I Given the following:

- The plant Is In an emergency conditio An excessive RCS cooldown has taken place in combination with an increase in RCS pressur The control room operators identify a RED path on the integrity status tree and start implementing EOP-16.0, Response to Imminent Pressurized Thermal Shock Conditio As per EOP-16.0, they allow the RCS to heat up, and they reduce RCS pressur When ROS subcooling has been reduced to 40°F, the operators notice that the integrity status tree has changed from a RED path to a YELLOW path conditio At the same time, they identify an ORANGE path on the containment status tre What actions should the operators take?

A. Go immediately to Step I of EOP-17.0 because the containment status tree has a higher priority than the integrity status tre B. Go immediately to Step I of EOP-I 7.0, Response to High Containment Pressure, because ORANGE path has a higher priority than a YELLOW pat S"5-IeLs C. Complete the actions of EOP-1 6.0, regardless of conditions on the other CSFI because EOP -1 was entered due to RED path conditio D. Complete the actions of EOP-16.0 because, once entered, EOP-1 6.0 should be performed to completion, unless pre-empted by a higher priority conditio REF: EOP-16.0, Response to Imminent Pressurized Thermal Shock Conditio SOURCE: Summer Exam Bank 2961 Less S'

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33. 028K2.01 I Initial Conditions:

-Unit was at 100% powe A D/G tagged out for maintenanc A LOCA is in progress in conjunction with a loss of off-site powe EOP's erc being performed and the crew is at the step for placing the H2 Recombiners in servic Which ONE of the following correctly describes the available recombiner and the source of power?

A. "A" recombiner from IDAI B. "B" recombiner from IDB1 C. "A"recombiner from 1DA2 D. "T" recombiner from 1DB2 Lesson Plan GS 2 "Safeguards Power', Objective GS-2-20, and 2 A. Incorrect, with the A DIG tagged out the A recombiner will not be available with an B. Incorrect, this rccombiner will be available, but from bus 16D C. Incorrect, this recombiner will not be availabl D. Correct, this recombiner will be available and is powered from 1 BD2.

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Recommended Question for Replacement 3 K5.04 1 The operating crew is performing FOP 2.0, LOSS OF REACTOR OR SECOTNDARY COOLANT, in response to a LOCA inside containment. Both hydrogen analyzers have been placed in service per SOP 122 and H2 concentration is currently 3% on both indicators. EO£1 2.0 directs that one post accident hydrogen recombiner be started and placed in service for this condition. After the local operator reports that "A" post accident H2 recombiner has been placed in service at the required power settinig, the NROATC reports that RB H2 concentration has increased to 3.6%. The correct course of action would be; Increase the power setting by 4 KW above the previous setting for the "'A" post accident H-I2 recombine Secure the "A" post accident H2 recombiner to eliminate the source of heat input to the buildin Place "B3" post accident H2 recombiner in service to provide an additional means of H2 removal from containmen Secure the "A" post accident H2 recombiner as it has failed and place the "B" H2 recombiler in service per SOP 12 Correct answer: A Reference: SOP 122, page 4 of 14, step 2.18 3 W5.01 1 Which of the following Reactor Building hydrogen concentrations would be considered excessive and prevent the operating crew from procedurally starting a post accident H2 recombiner;.5% % % %

Correct answer: C Reference: AB-15, Post Accident Hydrogen Removal, page 29 of 34

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Recommended Question for Replacement 3 K5.0 3 1 Which of the following will not add to the expected hydrogen generation duriing post accident conditions in the reactor building; zirconiull-water reactions radiolytic decomposition of water corrosion of metals within containment fission product gas release Correct answer: D Reference: T/S Bases, page B 3/4 6-5 dTi:rO 2:) La Snu

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36. 033A2.01 I Given the folgowi-og:

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- All core alterations have stoppe The Spent Fuel pool Is isolated from the Transfer Cana Decreasing boron concentration has been verified by sample analysis in the Spent Fuel Pool What impact would this have, if any, on parameters associated with the Spent Fuel Pool and what actions and/or procedure(s) would be used to correct or mitigate the consequences of this situation?

A. B thefu rack esin Keff would remain Is than 0.95. Use SOP-123 to transfer water frorr the transfer canal and the reactor cavit B. By the fuel rack design, Keff would remain less than 0.95. Use SOP-I 23 to drop the spent fuel pool level 5 feet and make up to the pool from the RWS C, Tj s ntTfu.el could reach criticalit Use AOP-1 23.2 and SOP-1 23 to transfer boric acid from ' he B

d Holdup Tank and then pump to the Spent Fuel Poo D. The spent fuel could reach criticality, the situation is not covered by the AOPs or EOPs, but woild be covered by the SAMG REF: AOP-123.2 DECREASING BORON CONCENTRATION IN THE SPENT FUEL POOL OR REFUEL CAVITY SOP-123 Spent Fuel Pool System SOP-I23 is not used to transfer water from the trainsfer canal or

,reactor cavitv y123 Distracter B 'lf pd r.n.hl. SOP-i23 is not used to transfer water from the RVVS p/rv/ Aw 2wai CU A~ UcS 12-42.2-M-r~ -SOP 012,CZ Cr /,f e0-ro 61P-wf A,112pý97Ž Distracter D - Actions not covered by SAMGs gC* P6.; e--j 17*-

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Modified 36. 033A2.01 1 Given the following:

- Core Offlood is complet All core alterations have stoppe The Spent Fuel pool is isolated from the Transfer Cana Decreasing boron concentration has been verified by sample analysis in the Spent Fuel Pool What impact would this have, if any, on parameters associated with the Spent Fuel Pool and what actions and/or procedure(s) would be used to correct or mitigate the consequences of this situation?

A. By the fuel rack design, Keff would remain less than 0.95. Use AOP-123.2 and SOP-123 to transfer boric acid from the Boric Acid Tanks to the Recycle Holdup Tank and then pump to the Spent F uel Poo B. By the fuel rack design, Keff would remain less than 0.95. Use SOP-123 to drop the spent fuel pool level 5 feet and make up to the pool from the RWS C. The spent fuel could reach criticality, Use SOP-123 to transfer water from the transfer canal an.J the reactor cavit D. The spent fucl could reach criticality, the situation is not covered by the AOPs or EOPs. but wo Ad be covered by the SAMGs.

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38. 034K4.01 I Which ONE of the following helps protect a fuel assemblyfrom binding while being loaded into the core" A. The Gripper being fully engage B. Using slow speed when the fuel assembly is entering the cor C. Hoist Overload interloc D. Slack Cable interloc Lesson Plan GS4 Fuel Handling System. Objective GS-4-1 Modified from Bank question # 201 A. Incorreot, the gripper being fully engaged does not prevent or alert the operator to a binding conditio B. Incorrect, the use of slow speed will not prevent the fuel assembly to bin C. Incorrect, an underload interlock would protect the fuel, but an overload would not occur while placing the fuel into the cor D. Correct, the slack cable would stop fuel descent into the core if the fuel began to bind.

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Recommended Question for Replacement. 034K4.02 1 The refueling upending machine uses

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as a hydraulic fluid because denineraliZed water; it prevents reactor cavity contamination and/or visibility loss in the event of a leak demineralized water; it provides both a cooling medium as well as a source of makeup due to controlled leakage which effectively offsets evaporatio C, DTE oil light; it is clear and will not impact visibility in the event of a hydraulic fluid lea DTE oil light; it is a water-soluble lubricant and has been proven to be a non contaminant when coming in contact with both fuel assemblies and RCS component Correct answer: A Reference: GS-4 Fuel Handling, page 40 of 82 38. 034K5.03 1 Technical Specification 3.9.3 requires that the reactor have been subcritical for a minimum of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the movement of irradiated fuel in the pressure vesse What is the basis for this T/S? Allow sufficient time for decay heat to subside to an acceptable level prior to changes in core geometr Ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission product Allow adequate time for radiation levels to subside to minimize personnel exposure prior to performing reactor vessel head remova This time is based solely on the 120 minimum time limit prior to draining to mid loop and is not applicable unless mid-loop operation is required prior to refuelin Correct answer: B Reference: T/S Bases, page B 3/4 9-1 Provide reference copy of Tech. Spec. 3.9.3 d*T:r0 Eo La 12n 896 0o o 0uap;sQŽ jawwns a CI LEBS SirC 608

61. 068AK3.17 1 Given the following plant conditions:

- Conditions exist that warrant a Control Room Evacuatio Offsite power is not availabl AOP-600.1, Control Room Evacuation, actions are complete to step 21 Alternative Action - The Control Room can not be re-entere Plant Management directs initiation of plant cooldown and entry to GOP-8, Plant Sritutdown from Hot Standby to Cold Shutdown with Control Room Inaccessible (Mode 3 to Mode 5)

After borating the ROS to cold, xenon-free shutdown concentration per GOP-8, how is tb-we Pressurizer boron concentration equalized with the RCS?

A. By increasing normal PZR spray flow with PZR heaters in manual to maintain PZR pressur B. By increasing Auxiliary Spray to the PZR with PZR heaters in manual to maintain PZ:R pressur C. By raising and lowering PZR level using PZR heaters to maintain PZR pressur D. By using normal boration flowpath and Auxiliary spray flowpath at the same time using PZR heaten; to maintain PZR pressur REF: GOP-8, Plant Shutdown from Hot Standby to Cold Shutdown with Control Room Inaccessible (Mode 3 to Mode 5)

AOP-600.t. Control Room Evacuation Distracter A - RCPs have been tripped as initial conditions of GOP-8. Normal spray not availabl Distracter B - Use of Auxiliary spray not directed by procedure GOP-$.

Answer C - Step 3.5 GOP-8 equalize PZR boron concentration by in-surge and out-surge Distracter D - Use of Auxiliary spray not directed by procedure GOP- ~-nt~rnl fflTi.Io 4uapTsam jawwns-12i:To 2a)

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73. G2.1.3 I The NROATC desires to leave the control room for approximately 30 minutes to get-picture Ir Which ONE of the following describes the MINIMUM items that a unexpected or tempora ry relief should include?

All A. Discuss existing plant conditions anticipated evolutionsd,

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1er s:..1a-en Bank Question 1992 Summer NRC exam modified slightl Reference: SAP-200, pages 7 and A. Incorrect, Logging the turnover is only required if the NROATC is leaving the sit B. Correct, lAW SAP-2 0 0 these are the minimum actions required for a temporary relie C. Incorrect, Logging is not required and all the actions are not liste D. Incorrect, a turnover sheet is not required and the turnover does not have to be logge,Ž*.....

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Modified 73. G2.1.31-The NROATC desires to leave the control room for approximately 30 minutes to vie,%v a training tape in tho Operations Msnaoer's Offic Which ONE of the following describes the MINIMUM items that a unexpected or tern porary relief should include?

A. Discuss existing plant conditions and anticipated evolution B. Discuss existing plant conditions, anticipated evolutions and review the Main Cointrol Board controls, instrumentation and annunciator C. Review the Main Control Board controls, instrumentation and annunciators and complete a turnover shee D. Discuss existing plant conditions, and anticipated evolutions, and complete a turnrover shee i C.4P Cos aOTJDo 4uapTsaNI jawwnS dEI:To 2:) 1.

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A. Total Effective Dose Equivalent and Committed Effective Dose Equivalen B. Deep Dose Equivalent and Committed Effective Dose Equivalen C. Total Effective Dose Equivalent and Planned Special Exposure D. Committed Effective Dose Equivalent and Planned Special Exposures, onl Bank Question from Surry NRC Exam 200 A. Incorrect, the components that make up a Radiation Worker's Occupational Dose is DDE and CED DDE + CEDE = TEDE B. Correct, DDE and CED C. Incorrect, the com ponents that make up a Radiation Worker's Occupational Dose is DDE and CED D. Incorrect, the components that make up a Radiation Worker's Occupational Dose is DDE and CED Sl./4XtW ctC17rVOWc-d,46~Cdf2,v

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91. W/EO6EK1.3 1 A LOCA is in progress with all RCPs secured, and the control room operators are attempting to stabilize plant conditions. An operator who is monitoring plant parameters observes the following:

RVLIS Narrow range:

50%

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Core exit TCs:

780F RCS Pressure 885 psig Which one of the followling describes current core cooling conditions and operational requirements?

A. Subcooled. Operator action is not required because core cooling is satisfactor B. Saturated. At their discretion, the operators can take action to restore subcooled coreý cooling per EOP-14.2, "Response to Saturated Core Cooling."

C. Degraded. Prompt action must be taken per EOP-14.1, "Response to Degraded Core Cooling," or conditions could degrad D. Inadequate. Prompt action must be taken per EOP-14.0 "Response to Inadequate Core Cooling, or core uncovery and fuel damage could occur.

Modified from a Bank Question # 42 Lesson Plan EOP-14.1 Response to Degraded Core Cooling, objective # 2070 and 20T1 A. Incorrect, the conditions given indicate that the RCS is in a superheat conditio B. Incorrect, the conditions given indicate that the RCS is In a superheat conditio C. Correct, the conditions given indicate that the RCS is in a degraded core cooling co this is the correct remedial action to tak D. Incorrect, the conditions given indicate a degraded core cooling conditio Lrea S'1, 1 08 EDOT.+J0 lueptlsaJ jawwns di

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94. W/E09EA2.1 i-A Loss of Off Site povver has occurred due to a seismic even a

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-Diesel Generators have started and are supplying electrical powe The CST has developed a leak and it has been determined that CST level is not 9 adequat EOP-1.1 has been 0 omplete RVLIS is availabl Which ONE of the following describes the correct procedure transition?

A. Transition to EOP-1.3 "Natural Circulation Cooldown'3l rz lu'

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B. Transition to GOP-6) 0 "Plant Shutdown From Hot Standby To Hot Shutdownhn4 zi tz C. Transition directly to EOP-1.4 "Natural Circulation Cooldown With Steam Void in Vessel" and continue to cold shutdow E O P

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180 A. Incorrect, the team will transition to EOP-1 3.0, but will not continue to cold shutdown using this procedure with CST level not adequat B. Incorrect, RCPs are not available therefore a forced cooldown can not be performe C. Incorrect, EOP-1.4 is the correct procedure to transition to, but transition must be performed after the first 0 stops of EOP1.3 is complet D. Correct, EOP-1.3 should be entered, the first 9 steps completed and then a transition to EOP 1.4 should be don......

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Modified 94. WIE09EA2.1 I-A Loss of Off Site power has occurred due to a seismic even Diesel Generators nave started and are supplying electrical powe The CST has developed a leak and it has been determined that CST level is not-EOP-1.1 has been complete..RVLIS is availabl Which ONE of the following describes the correct procedure transition?

A. Transition to EOP-1.3 "Natural Circulation Cooldown'.

B. Transition to GOP-6.0 "Plant Shutdown From Hot Standby To Hot Shutdown" adequate.

C. Transition directly to EOP-1.4 "Natural Circulation Cooldown With Steam Void in Vessel"and continue to cold shutdown.

D. Transition to EOP-l 5.0, "Response to Loss of Secondary Heat Sink." Without aOequate CST le jel, EFW is in jeopardy of being los...

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95. W/E1IG2.4.9 1 Given the following conditions:

- The plant is in Cold Shutdown with RCS temperature at 110 deg. RHR pump A and RHR heat exchanger A are in operatio SRCS [lot leg levol Is at 16 inches (mid-loop operations).

RHR Heat g

A Outlet Flow Control Valve (FCV-605A) has just stroked from 20%

open to full open due to a circuit faul If NO operator action is taken, which ONE of the following will occur to cause a loss of RI-IR cooling?

A. RHR pump overspeed trip from runout due to low discharge pressur B. RHR pump loss of suction due to vortexing at the RCS loop suctio C. RHR pump overcurrent trip due to high discharge pressur D. RHR pump overcurrent trip caused by pump runout due to low discharge pressur REF: Indian Point Exam 1996 AB-7 RHR SOP-71 RHR AOP-11 5.5 Loss of RHR with RCS not Intact (Mode 5)

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99. W/E15EA2.1 i-A Large Break LOCA has Occurre "-EOP-2.2 "Transfer to Cold Leg Recirculation" has just been cornplete The STA reports the following conditions:

-Reactor Building pressure 2.0 psi Reactor Building Radiation 10 RJH RHR Sump Level 420 f Which ONE of the following describes the immediate containment concern and the correct procedure to enter?t A. Inadequate suction to the RHR pumps, transition to EOP-2.4 "Loss of Emergency Coolant Recirculatlon."

B. Erroneous instrumentation readings, transition to EOP-17.2 "Resp~nse to High Reactor Building Radiation Level," when desire C. Reactor Building structural integrity; transition to EOP-17.0 "Response to High Reactor Building Pressure."

D. Flooding vital equipment in the Reactor Building; transition to EOP-17.1 "Response to Reactor Building Flooding-"

Modified from Diablo Canyon 99 exa Lesson Plan EOP-17-1 "Response to Reactor Building Flooding," objective 218 A. Incorrect, RHR sump level is adequate, Loss of emergency coolant recirculation is not the procedure that is required to be entered with these condition B. Incorrect, Radiation levels are high, but EOP-17.2 is entered on operator discretion and sump level is a higher priorit C. Incorrect Pressure is somewhat high, however it does not meet the threshold for entry (12psig). in a large breaK LOCA this procedure would have already been perrormed, and re entry is not require D. Correct, Reactor Building sump level is high and flooding is a concern and level has reached the threshold value to enter EOP-1 T-d(

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Written Examination Grading Form ES-403-1 (R8, SI)

Quality Checklist Facility:

Date of Exam:

Item Description Clean answer sheets copied before grading Answer key changes and question deletions justified and documented Applicants' scores checked for addition errors (reviewers spot check > 25% of examinations) Grading for all borderline cases (80% +/- 2%) reviewed in detail All other failing examinations checked to ensure that grades are justified Performance on missed questions checked for training deficiencies and wording problems; evaluate validity of Exam Level: RO/SRO Initials a

4D b

C

"A questions missed by half or more of the applicants Printed Name / Signature Date a. Grader b. Facility Reviewer(*)

c. NRC Chief Examiner (*)

d. NRC Supervisor (*)

Srsu1A~.~vb AIt/A

)a/4 (*)

The facility reviewer's signature is not applicable for examinatipns graded by the NRC; two independent NRC reviews are required.

ES-403 7/sil 2 9J-7-ti Z

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S-501 Post-Examination Check Sheet F

Task Description Date Complete Facility written exam comments or graded exams received and 09/26/02 verified complete Facility written exam comments reviewed and incorporated and 09/26/02 NRC grading completed, if necessary Operating tests graded by NRC examiners 09/30/02 NRC Chief examiner review of written exam and operating test 10/01/02 grading completed Responsible supervisor review completed 10/04/02 Management (licensing official) review completed 10/07/02 License and denial letters mailed 10/10/02 Facility notified of results 10/10/02 Examination report issued (refer to NRC MC 0610)

10/23/02 1 Reference material returned after final resolution of any N/A appeals Form ES-501-1 (R8, $1)

i::*-Knl

IR 05000395/2002301

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2002 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) Security Operator Exams 301(0) Emergency Planning Other

IR 05000395/2002301

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