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Text

Initial Exam 2011-301 Final Administrative JPMs
	ML112660354
Person / Time
Site:	Harris
Issue date:	09/22/2011
From:	; NRC/RGN-II
To:	;
References
	05-400/11-301
	Download: ML112660354 (141)
	v • d • e

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Harris Nuclear Plant Date of Examination: July 11 2011 Examination Level: RO I SRO Operating Test Number: 05000400/2011301 Administrative Topic Type Describe activity to be performed (see Note) Code*

Determine Rod Height Misalignment Using Conduct of Operations Thermocouples (JPM CR-i 39) Common P, R K/A G2.1.7 2011 NRCROAI-1 Determine the Target Rod Height and the Boron Conduct of Operations Required for a Rapid Power Reduction AW AOP-038 M, R K/A G2.1.25 2011 NRC RO A1-2 Review the Completed OST for Auxiliary Feedwater Equipment Control Pump 1 B-SB N, R (JPM ADM-i03) k/A G2.2.12 2011 NRC RO A2 Using Survey Maps determine stay times while Radiation Control performing a clearance activity.

M, R (JPM ADM-100) Common K/A G2.3.4 2011 NRC ROA3 NOT SELECTED FOR RO Emergency Procedures/Plan N/A 2011 NRC RO A4 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (4)

(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes) (0)

(N)ew or (M)odified from bank ( 1) (3)

(P)revious 2 exams ( 1; randomly selected) (1) 6/2812011 Rev. Final

2011 NRC ROAdmin JPM Sumrnary 2011 NRC RO Al-I Determine Rod Misalignment Using Thermocouples Previous 2009A NRC Exam JPM *randomly selected from bank K/A G2. 1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13) RO 4.4 SRO 4.7 The plant is at 90% power with a load decrease in progress when a control rod is observed indicating 12 steps higher than group demand. The candidate must perform Attachment 2 of AOP-001, Malfunction of Rod Control and Indication System, to calculate the temperature difference between the affected thermocouple and its symmetric thermocouples.

NOTE: Two thermocouple temperatures were changed with the resulting calculation now indicating a difference of greater than 10°F, indicating that the rod is misaligned. The 2009a JPM thermocouple temperatures resulted in a calculation of <10°F. During the 2009a exam the <10°F difference resulted was a rod position indication problem. For the 2011 exam the temperature difference of >10°F will have a concluding result of a rod misalignment.

2011 NRC RO Al -2 Determine the Target Rod Height and the Boron Concentration Change Required for a Rapid Power Reduction lAW AOP-038 MODIFIED K/A G2. 1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

(CFR: 41.10/43.5/45.12) RO 3.9 SRO 4.2 With plant conditions requiring a rapid power reduction to 65% power the candidate will be required to determine the target rod height, the time in core life and the amount of boric acid required for the power reduction.

This JPM was modified by changing the initial power level conditions and final power level.

2011 NRC RO A2 - Review the Completed OST for Auxiliary Feedwater Pump 1 B-SB NEW -

(JPM ADM-103)

K/A G2. 2.12 Knowledge of surveillance procedures.

(CFR: 41.10/45.13) RO 3.7 SRO 4.1 The candidate will be supplied a completed copy of OST-1 076, Auxiliary Feedwater Pump 1 B-SB Operability Test Quarterly Interval Modes 1-4 and be assigned the task of performing a peer check of the procedure prior to approval from the CRS.

2 6/28/2011 Rev. Final

2011 NRC RO Admin JPM Summary (continued) 2011 NRC RO A3 (Common) Using Survey Maps, SimpHfied Drawings, Plant Maps and valve lists, determine stay times while performing a clearance activity. MODIFIED K/A G2. 3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR; 41.12/43.4/45.10) RO 3.2 SRO 3.7 The candidate will be supplied a survey map of a location in the RAB and a clearance mission to complete in this radioactive area. The location also contains one or more hot spots. They must determine the individual stay times for two Auxiliary Operators (AO) without exceeding the annual administrative dose limits. They will be provided Survey Maps, Simplified plant drawings to locate valves, Plant Maps of the area and a plant valve list to determine the location of the valves they will be hanging a clearance on. The given information will supply the accumulated annual whole body doses for the two AOs, one of which recently worked for another utility. They must perform their calculations based on Progress Energy Administrative Dose Limits.

This JPM was modified by changing the location of the clearance and values of radiation areas.

2011 NRC RO A4 Not selected 3

6/28/2011 Rev. Final

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Harris Nuclear Plant Date of Examination: July 11, 2011 Examination Level: RO SRO Operating Test Number: 05000400/2011301 Administrative Topic Type Describe activity to be performed (see Note) Code*

Determine Rod Height Misalignment Using Conduct of Operations Thermocouples (JPM CR-i 39) Common P, R K/A G2.1.7 2011 NRC SRO Al-I Determine Subcooling with the Subcooling Margin Monitor Unavailable Conduct of Operations (JPM ADM-031)

M,R K/AG2.1.23 2011 NRC SRO AI-2 Review (for approval) a completed surveillance D, R procedure for PORV block valves.

Equipment Control (JPM ADM-035 SRO)

K/A G2.2.12 2011 NRC SROA2 Using Survey Maps, Simplified Drawings, Plant Maps M, R and Valve Lists, determine stay times while performing Radiation Control a clearance activity.

(JPM ADM-100) Common K/A G2.3.4 2011 NRC SRO A3 Given a Set of Plant Conditions Classify An Event and N, R manually complete an Emergency Notification Form.

Emergency Procedures/Plan K/A G2.4.41 2011 NRC SRO A4 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (5)

(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes) (1)

(N)ew or (M)odified from bank ( 1) (3)

(P)revious 2 exams ( 1: randomly selected) (1) 1 1I)Q/)fl1 1 P,

2011 NRC SRO Admin JPM Summary 2011 NRC SRO Al-I Determine Rod Misalignment Using Thermocouples Previous 2009A NRC Exam JPM *randomly selected from bank K/A G2. 1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12/45.13) RO 4.4 SRO 4.7 The plant is at 90% power with a load decrease in progress when a control rod is observed indicating 12 steps higher than group demand. The candidate must perform Attachment 2 of AOP-001, Malfunction of Rod Control and Indication System, to calculate the temperature difference between the affected thermocouple and its symmetric thermocouples.

NOTE: Two thermocouple temperatures were changed with the resulting calculation now indicating a difference of greater than 10°F, indicating that the rod is misaligned. The 2009a JPM thermocouple temperatures resulted in a calculation of <10°F. With <10°F difference the result was a rod position indication problem. With the temperature difference of >10°F the result is a rod misalignment. In the current JPM the SRO will need to determine Tech Spec requirements for a rod misalignment.

2011 NRC SRO A1-2 Determine Subcooling with the Subcooling Margin Monitor Unavailable (JPMADM-031) Bank- MODIFIED K/A G2. 1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(CFR: 41.10 / 43.5/45.2/45.6) RO 4.3 SRO 4.4 The applicant will be informed thata Small Break LOCA has occurred with SI actuated.

They will be provided with copies of the EOP Users Guide and multiple plant parameters.

They will be required to determine the RCS Subcooling margin lAW the EOP Users Guide directions.

This JPM was modified by changing the initial conditions to where the Containment pressure will be> 3 psig requiring the candidate to use adverse Containment values. In addition to this change the ERFIS computer will not be available. These two changes will require using different indicators and the results will be completely different values.

2011 NRC SRO A2 Review (for approval) a completed surveillance procedure for PORV block valves. (JPM ADM-035 SRO) Direct K/A G2. 2.12- Knowledge of surveillance procedures.

(CFR: 41.10/45.13) RO 3.7 SRO 4.1 The applicant will be provided with a handout of a completed copy of a PORV Block Valve full stroke quarterly surveillance. The procedure contains three (3) errors that the candidate must identify.

2 6/28/2011 Rev. Final

211-NRC SRO Admin-JPM Summary (continued) 2011 NRC SRO A3 (Common) Using Survey Maps, Simplified Drawings, Plant Maps and valve lists, determine stay times while performing a clearance activity.

(2009B NRC Admin JPM) MODIFIED K/A G2. 3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12/43.4/45.10) RO 3.2 SRO 3.7 The applicant will be supplied a survey map of a location in the RAB and a clearance mission to complete in this radioactive area. The location also contains one or more hot spots. They must determine the individual stay times for two Auxiliary Operators (AO) without exceeding the annual administrative dose limits. They will be provided Survey Maps, Simplified plant drawings to locate valves, Plant Maps of the area and a plant valve list to determine the location of the valves they will be hanging a clearance on. The given information will supply the accumulated annual whole body doses for the two AOs, one of which recently worked for another utility. They must perform their calculations based on Progress Energy Administrative Dose Limits.

This JPM was modified by changing the location of the clearance and radiation area intensities.

2011 NRC SRO A4 Classify an Event (NEW)

K/A G2. 4.41 Knowledge of the emergency action level thresholds and classifications (CFR: 41.10/43.5/45.11) RO 2.9 SRO 4.6 Given a set of initial conditions and the EAL Flow Path, the candidate must classify the appropriate Emergency Action Level for the event in progress. After completing the EAL classification the candidate will then manually complete an Emergency Notification Form (ENF).

3 6/28/2011 Rev. Final

Appendix C Job Performance Measure Form ES-C-i Worksheet Facility: Shearon Harris Task No.: 301 005H401 Task

Title:

Determine Rod Misalignment Using JPM No.: 2011 NRC JPM Thermocouples RO Al-i K/A

Reference:

G2.17 4.4/4.7 Examinee: NRC Examiner:

Facility Evaluator: Date:

Method of testinci:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

Initial Conditions: The plant was at 90 percent power, with a load decrease in progress, when the CRS noticed that the DRPI indication for rod H02 was reading 24 steps higher than the group demand. The load decrease has been stopped and AOP-001 entered.

Initiating Cue: The CRS has directed you to determine if Rod H02 is misaligned using Attachment 2 of AOP-001 and the provided T/C Core Maps.

2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Job Performance Measure Form ES-C-i Workshe8t- -

Task Standard: All calculations within +/- 20 of actual.

Required Materials: Calculator General

References:

AOP-00i, Attachment 1 and Attachment 2, Rev. 35 Handouts: JPM Cue Sheets Pages 11,12 AOP-001, Attachments 1 and 2, (Rev. 35)

Time Critical Task: No Validation Time: 12 minutes Critical Task Justification It the wrong thermocouples are used then none of the results will be St ep 2 correct Step 3 If the wrong values are selected then none of the results will be correct If the calculation for the averages were incorrect the results will be Step 4 incorrect If the differences are calculated incorrectly then the candidate will come St ep 5 to the wrong conclusion for Tech Specs 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 3 of 16 Form ES-C-i PRFQRMANGE iNFORMATION -

Start Time:

Performance Step: 1 OBTAIN PROCEDURE (provided with handout)

Standard: Obtains AOP-001 and refers to Attachments 1 and 2.

Comment:

V Performance Step: 2 DETERMINE THERMOCOUPLE LOCATION(S) ADJACENT TO THE MISALIGNED ROD USING THE CORE GRID MAP (SHEET 1), AND CIRCLE LOCATION(S) IN TABLE ABOVE.

THESE THERMOCOUPLES(S) ARE AFFECTED.

Standard:

Using the core grid map (Attachment 2, page 1 of 2),

Determines affected thermocouples to be G02, J02, and H03.

Circles G02, J02, and H03 on the table (Attachment 2, page 2 of 2).

Comment:

V Denotes Critical Steps

- 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 4 of 16 Form ES-C-i

- PERFORMANCE INFORMATION V Performance Step: 3 RECORD VALUES FOR ALL OPERABLE AFFECTED AND SYMMETRIC THERMOCOUPLES USING THE RVLIS CONSOLE OR ERFIS. SYMMETRIC THERMOCOUPLES ARE THOSE IN THE SAME ROW.

Standard: Locates RVLIS Console and accesses T/C CORE MAP for Train A and Train B. (Printout of RVLIS core map provided in handout)

Records value for Affected TC G02 (652° F) and Symmetric TC P07 (636° F).

Records value for Affected TC J02 (652° F) and Symmetric TC P07 (636°F).

Records value for Affected TC H03 (649°F) and Symmetric TCs C08 (636°F), H13 (640°F), and N08 (644°F).

Comment:

V Performance Step: 4 DETERMINE THE AVERAGE OF SYMMETRIC THERMO COUPLES, FOR EACH AFFECTED THERMOCOUPLE.

Standard:

Determines (636°F) for G02s Symmetric TC Determines (636°F) for J02s Symmetric TC Determines (640°F) for H03s Symmetric TC Comment:

If the candidate includes the adjacent TCs with the EXAMINERS NOTE: Symmetric TC numbers the averages will be wrong and the end result will be that a wrong final difference will be given:

Determines (644°F) for G02s Symmetric TCs Determines (644°F) for J02s Symmetric TCs.

Determines (640°F) for H03s Symmetric TCs.

V - Denotes Critical Steps 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 5 of 16 Form ES-C-i

- PERFORMANCE 1FORMAT[ON -

V Performance Step: 5 CALCULATE THE DIFFERENCE BETWEEN EACH AFFECTED THERMOCOUPLE AND THE AVERAGE OF ITS SYMMETRIC THERMOCOUPLES.

Standard: Calculates difference of 16°F 16°F for TC G02 16°F forTCJO2 9°F for TC H03 Determines that temperature difference is greater than 10°F and that Rod H02 is misaligned.

Evaluator Cue: CRS acknowledges report.

END OF JPM Comment:

Stop Time:

Terminating Cue: Difference between each affected thermocouple and its symmetric thermocouples has been calculated.

If the candidate included the adjacent TCs with the Symmetric TC numbers the averages would have been wrong and the calculated difference would be 8°F. The EXAMINERS NOTE: report would be that the temperature difference is less than 10°F. This report would provide the CRS with inaccurate information.

V - Denotes Critical Steps 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 6 of 16 Form ES-C-i PE.RFORMANCEII4FORMATION -

KEY I MALFUNCTION OF ROD CONTROL AND 1NDlCATON SYSTEM I Attachment 2 Adjicent and Symmetric Thermocouple Locations Sheet 1 of 3 THERMOCOUPLE LOCATIONS 1 2 3 4 5 6 7 8 9 10 ii 12 13 14 15 I I I I I ! I I I I A -* -* *....- :

i B - -*- H T ft ft ftT I

I I

I I -- I I C --

ft T ft ft T I - I I p p 0 T ft ft ft ft E T T T ft F T T R ft T G ft T ft T ft ft H \---,

ft 7 ft 1 T F J ft ft F T a K ft T ft T ft ftI ft ft L ft F ft F ft r M T ft R T ft F ft N T F R F R T p ft F ftT ft R T R - Control Rod T - Thermocouple T Thermocouples abandoned by EC 47997 AOP-GOI Rev. 35 Page 41 of 47 V - Denotes Critical Steps 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 7 of 16 Form ES-C-i PERFORMANCE INEORMAT KEY I MALFUNCTiON OF ROD CONTROL AND INDICATION SYSTEM I Attachment 2 Adjacent and Symmetric Thermocouple Locations Sheet 2 of 3 NOTE

. 610. E07, HOE. K08. and PO have no symmetric locations

. Symmetric thermocouples are those in the same row.

SYMMETRIC LOCM1OS GRID I_______ II LII IV TRAIN A B A B A B A B A08 HIS GO1* G15 R07 S L B05 E14 L14*

Y 0 COB Hi 3 NOB HO3)

M C D03 012 N04 M03 M A E04 005 E12 Mu L12 E T Hil E08 LOB H05 T I F05 Eli ElO KIV KOS LOS R 0 FO3 F13 NiO NOB K03 I N GOS FO9 J1O C S 008 HO$

co2) (Jo P07 MOO Ji2 Thermocouples abandoned by EC 4T997

1. DETERMINE thermocouple location(s) adjacent to the misaligned rod using care grid map (Sheet 1), AND CIRCLE location(s) in Table above.

AOP-OOi I Rev. 35 Page 42 of 47 V - Denotes Critical Steps 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 8 of 16 Form ES-C-i

- P-ERFO-RMANGE iNFORMAT4ON - -

KEY TIC CORE MAP (DATA SHEET 1) TRAIN B MSG # 245 2 3 4 __ 6 7 8 9 10 11 12 13 14 15 A I I A B 632 B C I 0 1636 650 D E I 644 50 640 6321 E 644 I F 644 632 644 IH 640 50 I K 644 640 644 50 I L M 1636 I M N I 640 N P I I P R R 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 TRAIN A T/C CORE MAP MSG # 245 1 2 3 4 12 13 14 15 A 50 A B 636 B C 632 C 0 D E 636 640 E F 644 648 644 F G 640 G H 640 644 652 H J J K 640 644 50 K L L M 640 644 M N 636 640 644 N P 640 P R 636 R 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 KEY V - Denotes Critical Steps 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 9 of 16 Form ES-C-i PERFORMANCE INFORMATION

[ MAISUNCTION OF ROD CONTROL AND INDICATION SYSTEM I Attachment 2 - Adjacent and Symmetric Thermocou pie [ocations Sheet 3 of 3

2. RECORD the following in the table below:

Adjacent TC number

Adjacent TC value using the RVLIS Console, ERFIS, or OSI-PI Symrnetri IC numbers (not including adjacent TCs)

Symmetric TC values for all OPERABIE TCs using the RVUS Console, ERHS, or OSI-PI

3. DETERM INE the average of symmetric thermocoup[es, for each adjacent thermocouple.

Adjacent it Symmetric TC Symmetric TC Number Value Number Value Average 302 NA P07 G02 652 6G COS 63 ll1 H03 649 N08 644 640 C02 N/A P07 J02 652 636

4. COMPARE each adjacent thermocouple vatue listed to its symmetric thermocouple average for indication of a misaligned rod. (REFER TO Attachment I )

--END OF ATTACHMENT 2 AOP-OO 1 I ev. 5 Page 43 of 47 V - Denotes Critical Steps 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 10 of 16 Form ES-C-i VERIFIQAT1QN QF CQMfLETIQN Job Performance Measure No.: 2011 NRC JPM ROISRO Al-i DETERMINE ROD MISALIGNMENT USING THERMOCOUPLES Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2011 NRC Admin Exam ROA1-l Rev. Final

Appendix C Page ii of 16 Form ES-C-i

. JPitQU.S1tEET Initial Conditions: The plant was at 90 percent power, with a load decrease in progress, when the CRS noticed that the DRPI indication for rod H02 was reading 24 steps higher than the group demand. The load decrease has been stopped and AOP-001 entered.

Initiating Cue: The CRS has directed you to determine if Rod H02 is misaligned using Attachment 2 of AOP-001 and the provided TIC Core Maps.

NAME_________________________ DATE___________________

CIRCLE THE CORRECT ANSWER FOR ROD H02:

ALIGNED MISALIGNED 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 12 of 16 Form ES-C-i JPMCUESHEET TIC CORE MAP (DATA SHEET 1) TRAIN B MSG # 245 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 A I A B 632 B C 636 I c D 1636 650 E I 644 50 640 6321 E F 50 644 F G 150 644 632 IG H I 649 644 I 640 50 K 640 I K L 644 640 644 50 1 L M 1636 N 640 N 1 636 1 R I R 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 TRAIN A TIC CORE MAP MSG#245 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 A I 50 1 A B

1636 B C 632 C 0 I D E I 636 640 1 E F 644 648 644 I F G 1 652 640 H I 640 644 652 640 640 I H I 652 K 1 640 644 50 I K L

M 640 644 I M N I 636 640 644 N P 640 I P R I 636 R 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 13 of 16 Form ES-C-i

-. JPMCUESI-IEET I MALFUNCTION OF ROD CONTROL AND lNCATION SYSTEM Attachment I Incicaion5 of Mi5aIiJned Rod Sht 1 of 1 Th tabig bGlcw indicates th siation in plant paramr which may bG indicative of roc misaliçnment. This varialion refers to reative changes n irdication from a reference conditDn at which the suspect rods poson was known to e properly aIiçnec. I he reference case miy be taken from prior opcating recoitlo, r it may be upatcd each time the proper rod positioning is verifiec by in-core measurements. Ir general, greater misalicnment will ause larger variafna Variations in Ni charnel indication are also affected by the core location f the s3pect rod For example, a msaligned rod that is closest to the \j_$$ deteclor shoiid indicate that N-.i4 flux raraneters are abnormal vhen compared with flux parameters of the other Power Range NI dannels. If the parameters bebv xhibi wi dbrIunnI vaTid1iuIL with an inividul DRPI diFrii y Irurni is up step counter demand ositIon by more than 12 steps, it is probaoly a rod position indication prcblem PLANT PAPAMETER VAI.UE lNDCAT1VE OF ROD MISAliGNMENT Quadrant Power Tilt Ratio (QPTR) Greater than I 02 Puier Rarrye lustiuuenaatiun Grtet thin 2% dilf ne Leten any Iwn thnnLs (PFFFR TO Attchm,t 4)

Delta Flux lnid:tc Gnater (1km 2% di11eii:e Letenaiw twn thanncIs (RFFFR TO Attthm,t 4)

Cuie Outlet Theinitucoupies Gra(et tlwi 1 QF dilTeiene between thermocouples adjacent 10 tie misaligned rod and fre average of symmetric tflermccoLples (PERFORM Atlacbnt 2 Axial Flux Traces (in-core movate CON5ULT Reactor Engineeing dctor) AND EVALUATE uirig in-corg rnovbl detecbrs cer EST-922, Conlroi Rod Position Determination Va lrcore lrtruniontation

--END OF A1TACHMENT 1-AOF-O0i I R&L 35 I Page 40 of 47 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 14 of 16 Form ES-C-i

. . . JPMCUESI-1.EET I F1ALFUNCTlON OF ROD CONTROL AND INDICATION SYSTEM I Attachment 2- Adjacent nd Symmetric ThermocoupLe Locations S1ieet : 3 TI IERMOCOU PIE LOCATIONS 1 2 3 $ 5 6 7 8 9 10 ii 12 13 14 16 I I I A

I I -L - - --

I II II I

B I I I C ,. ft 7 ft ft 7 j j E ft 1 ft r r ft 1 r F T T P R r i tI c I-I ft 7 ft 7 r ft 7 J r ft ft r T ft K ft 7 ft 7 ft RT R ft L R r T P T M

N 7 ft 7 ft 1 ft 7 p ft 7 ft R r R Control Rod I Thrnocniip 1* Thermocouples abaidoned by CC 47997 AOP-OO1 Rev. 35 Page $1 of 47 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Page 15 of 16 Form ES-C-i

. JPMCIJESHEET I MAISUNCUON OF ROD CONTROL AND INDICATION SYSTEM I Attachment 2 dJ]cent rnd SmnieuIc Thermocouple LocatIons lie2of 3 NOTE

. ElO, E07.. HOft K08, and P03 have no smetric iocations.

. Symmetric thermocouples are those in the same row.

SYMMETRIC LOCATIONS t3RlD Ii 111 IV TRAiN A B A B A B A B AOW H15 c301 31 R07 S L BOG E14 Y 0 DO 41 NOG 1403 M C 012 N04 M03 M A [04 E12 MN L12 C 1 411 [00 LOC 1405 T I P05 FlI 1O K11* 1<05 LOS R 0 FO3 F13 NW N06 KO I l4 G436 FOG JIC C S SO HOG G02 J02 P07 MOO J12 Theimocouple3 abandored by EC 47997

1. DETERMINE ermocDue loca:ors) adjacent to the nisagned rod using core grid map (Sheet 1), AND CIRCLE location(s) in TaDle above.

AO[--JUT I I wage 42 ot 4!

2011 NRC Admin Exam ROA1-1 Rev. Final

Appendix C Page 16 of 16 Form ES-C-i JRtvl CUE SHEET I MALFUNCTION OF ROD CONTROL AND INDiCATiON SYSTEM I Attachment 2 Adjacent and Symmetric Thermocouple Locations Sheet 3 of 3

2. RECORD the foowing in the table below:

Adjacent TC number

Adjacent TC value using the RVLIS Console, ERFIS, or OS [-P1

Symmetric IC numbers (not including adjacent TCs)

Symmetric IC values for all OPERABLE TCs using the RVLIS Console, ERFIS, or OSI-PI

3. DETERMINE the average of symmetric thermocouples, for each adjacent thermocouple.

Adjacent TC Symmetric TC Symmetric TC Number Value Number Value Average

4. COMPARE each adjacent thermocouple value listed to its symmetric thermocouple average for indicaon of a misaligned rod (REFER TO Attachment I )

--END OF ATTACHMENT 2--

AOP-OO1 I Rev. 35 Page 43 of 47 2011 NRC Admin Exam RO Al-i Rev. Final

Appendix C Job Performance Measure Form ES-C-i Worksheet -

Facility: HARRIS Task No.:

Task

Title:

Determine the Target Rod Height JPM No.: 2011 NRC Admin and the Boron Acid Addition JPM RO Ai-2 Required for a Rapid Power Reduction lAW AOP-038 K/A

Reference:

2.1.25 (2.8/3.1)

Examinee: Examiner:

Facility Evaluator: Date:

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

Initial Conditions:

The plant is operating at 90% power, with RCS Boron at 250 PPM.

Core Burnup is at 425 EFPD.

The Load Dispatcher just reported that a tornado has taken out a High Voltage Transmission tower on the grid.

The Load Dispatcher has requested HNP to rapidly reduce power to 65% and to reduce power at 25 DEH Units/minute.

The crew has entered AOP-038, RAPID DOWNPOWER.

Initiating Cue: The CRS has directed you to perform AOP-038, Step 3 Determine the required boric acid addition for desired power reduction - -.

Determine the final power target rod height, time in core life and the amount of boric acid that will be required. List each value on the space provided.

2011 NRC Admin Exam JPM RO A1-2 Rev. Final

Appendix C Job Performance Measure Form ES-C-i rkset Task Standard: Target Rod Height and Boration calculation within specified bands.

Required Materials: Calculator General

References:

AOP-038, RAPID DOWNPOWER, Revision 26 Handout: AOP-038, Attachment 2, Rev. 26 Time Critical Task: No Validation Time: 10 minutes Critical Task Justification Determine Target If the wrong rod height is determined the boric acid addition for rod Rod Height insertion will not be adequate Determine Time In If the wrong time in core life were used the amount of boric acid addition Core Life would be incorrect. Since this is EOL using either MOL or BOL the determination would yield significantly less boric acid than required.

Determine Required If the wrong amount of boric acid addition is calculated the rod height will Boric Acid Addition not be correct for the final plant power level 2011 NRC Admin Exam JPM RO Ai-2 Rev. Final

Appendix C Page 3 of 8 Form ES-C-i

- PERFORMANCE INFORMATION START TIME:

AOP-038, Step 3 Performance Step: 1 Reviews Step 3 Standard: Determines RNO applies and proceeds to Attachment 2.

Evaluator Cue: Provide handout or, if in the simulator, the applicant can pull the procedure.

Comment:

AOP-038, Attachment 2 V Performance Step: 2 Determine Target Rod Height Standard: Target Rod Height at 65% power 7060% Power: CB U @171-183 steps.

183171 =12/2=6 171 +6=177 steps Comment:

AOP-038, Attachment 2 V Performance Step: 3 Determine Point in Core Life Standard: Compares current RCS Boron/Core Burnup to table and determines EOL. 425 EFPD = EOL Comment:

V - Denotes Critical Steps 2011 NRC Admin Exam JPM RO Al -2 Rev. Final

Appendix C Page 4 of 8 Form ES-C-i

- PERFORMANCE INFORMATION AOP-038, Attachment 2 V Performance Step: 4 Calculate Boron Change.

Standard: Applies divide by two NOTE for the 5% increment (70% to 60%)

and the 10% increment totals in the EOL Column:

202 + 203 + (180/2) = 495 ( 490 500 gallons)

Comment:

AOP-038, Step 3 Performance Step: 5 Reports Step 3 completion to CRS.

Standard: Reports results of calculation.

Evaluator Cue: Acknowledge report.

Comment:

Terminating Cue: After the Step 3 results report: Evaluation on this JPM is complete.

STOP TIME:

V - Denotes Critical Steps 2011 NRC Admin Exam JPM RO A1-2 Rev. Final

Appendix C Page 5 of 8 Form ES-C-i PERFORMANCE INFORMATION KEY Power Reduction lAW AOP-038 90% 65%

RAPID DOWNPOWER Attachment 2 Sheet 1 of 1 Gallons of Boric Acidfrarget Rod Height Required for Power Reduction NOTE

. This Attachment serves as a reactivity plan. [C.3]

. This table applies to Cycle 17 on[y

. Target rod heights correspond to the lower (target) power level in each row.

. Gallons of boric acid in the table are for 10% power reduction wtcrements These are additiEve for power reductions of greater than 10%.

Example: A power reduction from 90% to 60% at BOL would require f181 gal ÷ 163 gal +155 gal = 499 gal

. For purposes of this procedure, 5% increments can be obtained by dividing bytwo

. As used in this table, the folIoing times in core life are assumed:

BOL (0 EFPD 161) (3000 MWD/MTU)

MOL (161 < EFPD 333) Cl0,000 MWDIMTU)

EOL (333< EFPD 507)i17,000MWD1MTU) 1(425 EFPD)

Power Target Gallons of Boric Acid Required for Power Reduction Level Rod

(% Height BOL MOL EOL (D Bank 0 s EFPD 161 161 < EFPD 333 333< EFPD 507 lOOto9O 206 216 2& 3.17 90o80 194 181 214 202 80 to 70 183 163 189 203 70 to 60 171 155 173 180/ =

60 to 50 159 146 173 189 50 to 40 147 155 157 205 40 to 30 135 156 166 206 30to20 124 166 166 255 20 to 10 1 12 157 175 312

-END 0l ATIACF{MENT 2-AOP-038 I Rev. I Page 21 of 24 V Denotes Critical Steps

- 2011 NRC Admin Exam JPM RO A1-2 Rev. Final

Appendix C Page 6 of 8 Form ES-C-i

- VER[FICATION OF COMPLION - -

Job Performance Measure No.: 2011 NRC Admin JPM RO A1-2 Determine the Target Rod Height and the Boron Concentration Change Required for a Rapid Power Reduction lAW AOP-038 Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2011 NRC Admin Exam JPM RO Ai-2 Rev. Final

Appendix C Page 7 of 8 Form ES-C-i JPMCIJES[-[EET Initial Conditions: . The plant is operating at 90% power, with RCS Boron at 250 PPM.

Core Burnup is at 425 EFPD

. The Load Dispatcher just reported that a tornado has taken out a High Voltage Transmission tower on the grid.

. The Load Dispatcher has requested HNP to rapidly reduce power to 65% and to reduce power at 25 DEH Units/minute.

. The crew has entered AOP-038, RAPID DOWNPOWER.

Initiating Cue: The CRS has directed you to perform AOP-038, Step 3 Determine the required boric acid addition for desired power reduction -

Determine the final power target rod height, time in core life and the amount of boric acid that will be required. List each value on the space provided.

NAME DATE_

Target rod height CBD at steps Time in core life used =

Required boric acid addition = gallons 2011 NRC Admin Exam JPM RO Al -2 Rev. Final

Appendix C Page 8 of 8 Form ES-C-i JPM CUE SHEET RAPID DOWN POWER Attachment 2 Sheet I of 1 Gallons of oric AcidlTarget Rod Height Required for Power Reduction NOTE a This Attachment serves as a reactivity plan. tG]

This table applies to Cycle 17 only.

Target rod heights correspond to the lower (target) power level in each row.

Gallons of boric acid in the table are for 10% power reduction increments. These are additive for power reductions of greater than 10%.

Example: A power reduction from 90% to 60% at BOL would require [181 gal +

I 63 gal +155 gal 49 ga

. For purposes of this procedure, 5% increments can be obtained by dividing by two.

a As used in this table, the following times in core life are assuniect:

BOL (0 EFPD 16i)(3000 MWD/MTU)

MOL (161 < EFPD 333) (10,000 MWDIMTU)

OL (333 < EFPD 507) (17,000 MWDMTU)

Power Target Gallons of Boric Acid Required for Power Reduction Level Rod

(%J Height BOL MOL EOL (D Bank) ci EFPD 161 161 < EFPD 333 333 < EFPD 507 100 to 90 206 216 264 317 90 to 80 194 181 214 202 80 to 70 183 163 189 203 70to60 171 155 173 180 60to60 169 146 173 189 50 to 40 147 155 157 205 40 to 30 135 156 166 206 30 to 20 124 166 166 255 20 to 10 112 157 175 312

-END OP ATTACHMENT 2 AOP-038 Re. 26 Page 21 of 24 2011 NRC Admin Exam JPM RO Al -2 Rev. Final

Appendix C Job Performance Measure Form ES-C-i Worktieet Facility: HARRIS Task No.:

Task

Title:

Review the Completed OST for JPM No.: 2011 NRC RO A2 Auxiliary Feedwater Pump 1 B-SB K/A

Reference:

2.2.12 (3.4)

Examinee: NRC Examiner:

Facility Evaluator: Date:

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

Initial Conditions:

The plant is operating at 100% power

OST-1076, AUXILIARY FEEDWATER PUMP lB-SB OPERABILITY TEST QUARTERLY INTERVAL MODES 1-4, has just been completed Initiating Cue: You are the BOP and have been asked to peer check the just completed copy of OST-1 076. Completely review the procedure and note any problems in the spaces provided.

2011 NRC Admin Exam RO A2 Rev. Final

Appendix C Job Performance Measure Form ES-C-i

- Wrt(Shet Task Standard: All errors (3) identified Required Materials: None General

References:

OST-1076, AUXILIARY FEEDWATER PUMP 1A-SA OPERABILITY TEST QUARTERLY INTERVAL MODES 1-4, Rev. 29 Handout: Completed OST-1 076 with the errors incorporated.

Time Critical Task: No Validation Time: 12 minutes Critical Task Justification Item 1 Identification of stopwatch beyond calibration date all timing data collected with out of cal device is non usable therefore the test is invalid.

Item 2 Outboard Horizontal Vibration reading not documented with the reading in the ALERT range the pump would still be operable but this could prevent corrective action or the required increase in surveillance frequency.

Item 3 Shut time for valve 1AF-51 exceeding CODE CRITERIA. This valve should have been retested immediately but was not.

2011 NRC Admin Exam RO A2 Rev. Final

Appendix C Page 3 of 6 Form ES-C-i PERFORMANCE INFORMAT1ON START TIME:

Performance Step: 1 Reviews the completed OST-1 076 for the 1 B-SB test.

Standard: Ensures proper conditions, signatures/initials, and may verify the current revision of the procedure Comment:

J Performance Step: 2 Review the completed OST Standard: Attachment 1 (Page 20) Identifies that the stopwatch that was used is past calibration date Evaluator Note: This makes the OST invalid andlor any timed elements would have to be repeated.

Comment:

J Performance Step: 3 Review the completed OST Standard:

Attachment 3 (Page 24) Outboard Horizontal Vibration reading is in the ALERT range but there is no documentation in accordance with Instructions l.a and 1 .b.

Evaluator Note: The pump would still be operable but this could prevent corrective action or the required increase in surveillance frequency.

Comment:

V Denotes a Critical Step

- 2011 NRC Admin Exam RO A2 Rev. Final

Appendix C Page 4 of 6 Form ES-C-i

- PERORMANCEiNFORMATION q Performance Step: 4 Review the completed OST Standard: Attachment 4 (Page 25) 1AF-51 SHUT time exceeds CODE CRITERIA and, in accordance with Step 7.0.1, should have been retested immediately.

Comment:

Performance Step: 6 Review the completed OST Standard: Returns the procedure unsigned and has identified three errors during the peer check review.

Comment:

Terminating Cue: When the procedure is returned: Evaluation on this JPM is complete.

STOP TIME:

- Denotes a Critical Step 2011 NRC Admin Exam RO A2 Rev. Final

Appendix C Page 5 of 6 Form ES-C-i

- VERIPFCATION OF COMPLETION Job Performance Measure No.: 2011 NRC RO A2 Review the Completed OST for Auxiliary Feedwater Pump 1 B-SB Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2011 NRC Admin Exam RO A2 Rev. Final

Appendix C Page 6 of 6 Form ES-C-i JPM CUE SHEET INITIAL CONDITIONS: . The plant is operating at 100% power

. OST-1076, AUXILIARY FEEDWATER PUMP lB-SB OPERABILITY TEST QUARTERLY INTERVAL MODES 1-4, has just been completed INITIATING CUE: You are the BOP and have been asked to peer check the just completed copy of OST-1 076. Completely review the procedure and note any problems in the spaces provided.

NAME DATE IF any discrepancies were identified in the review of OST-i 076 list them on the lines below 2011 NRC Admin Exam RO A2 Rev. Final

CON1iUOUS j Progress Eneijy HARRIS NUCLEAR PLANT PLANT OPERATING MANUAL VOLUME 3 PART9 PROCEDURE TYPE: OPERATiONS SURVEILLANCE TEST NUMBER: OST-1 076 TITLE: AUXILIARY FEEDWATER PUMP lB-SB OPERABILITY TEST QUARTERLY INTERVAL MODES 1-4 N : This procedure has been screened per PLP-1 00 Criteria and determined to be CASE III. No additional management involvement is required.

OST-1076 Rev, 29 I Page 1 of 31

Table of Contents Section, Page 1.0 PURPOSE 3

2.0 REFERENCES

4 2.1. Plant Operating Manual Procedures 4 2.2. Technical Specifications 4 2.3. Final Safety Analysis Report 4 2.4. Drawings 4 2.5. TechnicalManuals 4 2.6. Corrective Action Program (CAP) Items 5 2.7. Other 5 3.0 PREREQUISITES 5 4.0 PRECAUTIONS AND LIMITATIONS 6 5.0 TOOLS AND EQUIPMENT 7 6.0 ACCEPTANCE CRITERIA 7 7.0 PROCEDURE 9 7.1. Auxiliary Feedwater Pump I B-SB 9 7.2. Valve Testing 18 7.3. Test Completion 19 8.0 DIAGRAMS/ATTACHMENTS 19 Attachment I Calibration Data Sheet

- 20 Attachment 2- Performance Data 21 Attachment 3- Vibration Data 24 Attachment 4- Valve Test Data 25 Attachment 5- Valve Retest Data Sheet 26 Attachment 6 Certifications and Reviews

- 28 J OST-I 076 Rev. 29 I Page 2 of 31

1.0 PURPOSE This test demonstrates the operability of Auxiliary Feedwater Pump 1 B-SB at least once every 92 days on a Staggered Test Basis by satisfying the following:

a. Technical Specification Surveillance Requirements:

4.7.1.2.1.a.1.a

4.3,2.1 Table 4.3-2 Item 6a

b. Inservice Testing Program testing for the following components:

IAF-23, AFWP I B Recirc to CST Ck Valve

IAF-4, AFWP IA Recirc to CST Ck Valve

IAF-1 10, Aux Feed Pump IX Recirc to CST Ck Valve

1 CE-46, CST Suction Check Valve to 1 B-SB AFW Pump

IAF-49, AFWP IA FCV to A S/G

IAF-50, AFWP IA and 1 B Common FCV to C SIG

IAF-51, Aux FW B Regulator

IAF-93, Aux FW B Isolation

1AF-201, Aux Feed Pump 1A Line to A SIG Ck Valve IAF-54

1AF-54, Aux Feed Pump IA Line to A SIG Ck Valve

1AF-202, Aux Feed Pump lB-SB Line to B S/C Ck Vlv IAF-92

IAF-92, Aux Feed Pump 1 B-SB Line to B SIC Ck Vlv

IAF-203, AFWP IA & I B Common Line to C S/C Ck Valve 1 AF-73

IAF-73, AFWP IA & lB Common Line to C S/S Ck Valve

Auxiliary Feedwater Pump I B-SB

2. This test satisfies Technical Specification Surveillance Requirement 4.6.3.1 for Containment Isolation Valve IAF-93 operability following maintenance.

OST-1 076 Rev. 29 Page 3 ofJ

2.0 REFERENCES

2.1. Plant Operating Manual Procedures

1. OP-137

2. 1SI-ili

3. 151-800

4. ISI-.801 2.2. Technical Specifications

1. 3.7.1.2

2. 6.8.4.m (Inservice Testing Program)

3. 4.6.3.1

4. 4.7.1.2.1.a.1

5. 4.3.2.1 Table 4,3-2 item 6a 2.3. Final Safety Analysis Report

1. 7.3.2

2. 10.4.9 3, 10.4.9a

4. TM1 Appendix 2.4. Drawings

1. 5-S-0544

2. 5-S-0545

3. 6-B-401 1942 2.5. Technical Manuals

1. VM-BJ H-VOl, Pumps, Mtr Drvn Aux Feedwater OST-l 076 Rev. 29 Page 4 of 31

2.6. Corrective Action Program (CAP) Items

1. 93H0166 2.7. Other

1. PCR-6502

2. ESR95-01041

3. ESR 97-00475

4. HNP-IST-003, HNP 1ST Program Plan 3 Interval

5. HNP Mechanical Calculation AF-0047 3.0 PREREQUiSITES

1. VERIFY that the Auxiliary Feedwater System is aligned in a manner that will support the performance of this test.

2. COORDINATE the performance of this OST with other plant evolutions such that the minimum equipment operating requirements of Tech Specs are met.

3. OBTAIN any tools and equipment required per Section 5.0.

4. VERIFY instrumentation needed for the performance of this test is free of deficiencies that may affect instrument indication.

5. DIRECT Maintenance to install the following:

a. I B-SB AFW Pump suction pressure test gauge (0-50 psig) at 1 CE-54-V2, P1-2251 B Instrument Valve.

b. I B-SB AFW Pump discharge pressure test gauge (0-2000 psig) at IAF-26-HV2, P1-21 SIB Instrument Valve.

c. 0 to 100 inwc differential pressure test gauge, with an accuracy of plus or minus 1.0 inch water, at IAF-7-HVI, FI-IAF-2172S Instrument Valve, and IAF-7-LVI, Fl-IAF-21 72S Instrument Valve, to measure AFW Pump Local Recirculation Flow.

6. VERIFY AND RECORD calibration due dates for all instrumentation listed on the Calibration Data Sheet, including the identification number.

OST-1 076 I Rev. 29 Page 5 of 31

3.0 PREREQUISITES (continued)

7. VERIFY personnel taking vibration measurements are qualified per Reference 2.1.2.

8. INFORM the CRS that the performance of this OST during MODE 1, 2, and 3 initiates an LCO per Technical Specification 3.7.1.2.

9. OBTAIN CRS permission to perform this OST.

c ignatu&

4.0 PRECAUTIONS AND LIMITATIONS

1. This test should not be performed when a motor driven AFW pump is being used for SG level control, due to the MDAFW FCV auto open feature.

2. Only one Auxiliary Feedwater Pump shall be tested at a time.

3. Auxiliary Feedwater Pump 1 B-SB should not be operated at less than 50 gpm and should not be operated at 50 gpm for more than 300 hours0.00347 days 0.0833 hours 4.960317e-4 weeks 1.1415e-4 months .

(This is normally provided by the recirc line) (Reference 2.7.1)

4. If an accident initiation signal is received during the performance of this test, stop the test.

5. If any valve stroke time falls outside its Code Criteria, the valve will be immediately retested per the retest instructions or declared inoperable.

OST-1076 I Rev. 29 I Page 6 oTj

5.0 TOOLS AND EQUIPMENT

1. Two way radios or sound powered phones

2. Hand held pyrometer, if ERFIS is not to be used for temperature monitoring

3. Fluke Model 80228 Digital Multimeter (DMM) or Equivalent, if ERFIS is not to be used for discharge pressure monitoring

4. Calibrated Vibrometer per ISI-1 11 with an accuracy of +/- 5 percent

5. Calibrated Stopwatches Three (3 needed if timing all three FCVs together>

6. Pressure Test Gauge, 0 to 50 psig, with an accuracy of 0.5% of span or better

7. Pressure Test Gauge, 0 to 2000 psig, with an accuracy of 0.5% of span or better

8. 0 to 100 inwc test gauge, with an accuracy of plus or minus 1.0 inch water, ITT Barton or equivalent 6.0 ACCEPTANCE CRITERIA

1. This test will be completed satisfactorily if the following conditions are verified:

a. IAF-23, AFWP I B Recirc to CST Ck Valve, full strokes open as demonstrated on Attachment 2.

b. I CE-46, CST Suction Check Valve to 18-SB AFW Pump, strokes open as demonstrated on Attachment 2.

c. The following valves Stroke Close as demonstrated on Attachment 2:

IAF-201, Aux Feed Pump IA Line to A S/G Ck Valve IAF-54, and IAF-54, Aux Feed Pump IA Line to A S/G Ck Valve

1AF-202, Aux Feed Pump I B-SB Line to B S/G Ck Vlv, and IAF-92 Aux Feed Pump lB-SB Line to B SIG Ck Vlv

IAF-203, AFWP IA & 18 Common Line to C S/G CkValve 1AF-73, and IAF-73, AFWP IA & lB Common Line to C SIG CkValve

IAF-4, AFWP IA Recirc to CST Ck Valve

IAF-IlO, Aux Feed Pump IX Recirc to CST Ck Valve OST-1 076 I Rev. 29 I Page 7 of 31

6.0 ACCEPTANCE CRITERIA (continued)

d. The following valves achieve their fail safe position as demonstrated in Attachment 4.

1AF-49, AFWP 1A FCV to A SIC

IAF-51, Aux FW B Regulator

IAF-50, AFWP IA and I B Common FCV to C S/C

e. The following valves are full stroked with the stroke times within the Acceptance Criteria listed on Attachment 4:

IAF-49, AFWP 1A FCV to A SIG

IAF-51, Aux FW B Regulator

IAF-50, AFWP IA and I B Common FCV to C SIC

1AF-93, Aux FW B Isolation

f. Auxiliary Feedwater Pump I B-SB (1) Starts from the MCB and develops a differential pressure greater than or equal to 1530.3 psid (TS4.7.1.2.1.a.l of 1514 psid plus 16.3 psid per CaIc AF-0047) at a recirculation flow of greater than or equal to 50 gprn as demonstrated by pump differential pressure calculation from Attachment 2.

(2) Performance Data is within the Acceptance Criteria listed on Attachment 2.

(3) Vibration Data is within the Acceptance Criteria listed on Attachment 3.

j OST-1076 Rev. 29 Page 8 of 31

7.0 PROCEDURE The following Steps should be initialed after this test is completed.

h

1. IF during the performance of this test, a valve stroke time exceeds its Code Criteria, THEN IMMEDIATELY RETEST the valve per Attachment 5.

2. IF during the performance of this test, a valve exhibits abnormal or erratic action, THEN DOCUMENT the condition in the comments section of Attachment 6.

7.1. Auxiliary Feedwater Pump lB-SB

1. VERIFY I B-SB AFW Pump suction pressure test gauge (0-50 psig) at ICE-54-V2, P1-22518 Instrument Valve, is installed and valved in.

INSTALLED VALVED IN

2. VERIFY lB-SB AFW Pump discharge pressure test gauge (0-2000 psig) at IAF-26-HV2, P1-21518 Instrument Valve, is installed and valved in.

INSTALLED VALVED IN

3. VERIFY the 0 to 100 inwo test gauge is installed at IAF-7-[IV1, Fl-IAF-2172S Instrument Valve, and IAF-7-LVI, FI-IAF-2172S Instrument Valve.

4. RECORD I B-SB Pump idle suction pressure from test gauge installed in Step 7.1.1 on Attachment 2.

5. SHUT I CE-54-V2, P1-2251 B Instrument Valve.

OST-1 076 1 Rev. 29 I Page 9 of 31 j

7.1 Auxiliary Feedwater Pump I B-SB (continued)

6. CHECK the following computer points are less than 135°F AND DOCUMENT proper Stroke Close testing for indicated check valves on Attachment 2.

TAF2007A, Motor Driven AFW to SG A (IAF-201 and 1AF-54) Ia

TAF2007C, Motor Driven AFW to SG B (1AF-202 and IAF-92)

TAF2007E, Motor Driven AFW to SG C (IAF-203 and 1AF-73)

7. DIRECT Maintenance to valve in and vent the recirculation flow test gauge installed at IAF-7-HV1, Fl-IAF-21 72S Instrument Valve, and I AF-7-LVI, Fl-I AF-21 725 Instrument Valve.

VALVED IN VENTED

8. PERFORM prestart checks per OP-i 37 on Auxiliary Feedwater Pump I B-SB. i7-.

9. ESTABLISH communications between the Main Control Room and Auxiliary Feedwater Pump I B-SB.

Performance of Step 7.1.12 will initiate an LCO for Auxiliary Feedwater Pump ir lB-SB if in MODE 1,2, or3 per Technical Specification 3.7.1.2.

10. INFORM the CRS that Auxiliary Feedwater Pump I B-SB will be made inoperable.

11. RECORD the time and date that Auxiliary Feedwater Pump 1 B-SB was made inoperable on Attachment 6.

r-N9TE:) Performance of Step 7.1.12 will isolate Auxiliary Feedwater Pump lB-SB

,K..- from the Steam Generators.

12. SHUT IAF-208, AFWP lB Discharge Isolation, OST-i076 I Rev. 29 I Page 10 of 31

7.1 Auxiliary Feedwater Pump lB-SB (continued)

R 13. At the MCB, START Auxiliary Feedwater Pump lB-SB. (Reference 2.3.1)

Time started: fc) c)

CAU N Auxiliary Feedwater Pump I B-SB shouJ operated at less than 50 gpm and should not be operated at 50 gpm for more than 300 hours0.00347 days 0.0833 hours 4.960317e-4 weeks 1.1415e-4 months .

14, VERIFY that the recirculation flow on local indicator Fl-2172S is greater thanorequalto5Ogpm.

15. VERIFY the following FCVs indicate open and initial for the PRETEST ALIGNMENT on Attachment 4.

FCV-2051A (1AF-49), SG A

FCV-2051B (IAF-51), SG B

FCV-2051C (1AF-50), SG C The demand output signal indicator accuracy is 2% for the following controllers:

FK-2051A1 SA, AUX FW A REGULATOR IAF-49

FK-2051B1 SA, AUX FW B REGULATOR IAF-51

FK-2051C1 SA, AUX FW C REGULATOR IAF-50 an auto open signal is present the controller demand output will not remain

/at0%.

Steps 7.1.16 and 7.1.17 will time the MDAFW FCVs shut. It is recommended that only one valve at a time be timed. However, it is permissible to time all three FCVs simultaneously.

16. SIMULTANEOUSLY START the stopwatch(es) AND SET the demand output signals for the following controllers to 0%:

FK-2051A1 SA, AUX FW A REGULATOR 1AF-49

FK-2051 BI SA, AUX FW B REGULATOR IAF-51

FK-2051C1 SA, AUX FW C REGULATOR 1AF-50 OST-1 076 Rev. 29 Page 11 of 31

7.1 Auxiliary Feedwater Pump lB-SB (continued)

17. WHEN the following FCVs indicate shut, THEN STOP the stopwatch(es):

FCV-2051A (1AF49), SG A

-ía

FCV-2051B (1AF-51), SO B

FCV-2051C (IAF-50), SG C

18. RECORD valve stroke times from Step 7.1.17 on Attachment 4.

NTE:) The demand output signal indicator accuracy is 2% for the following controllers:

FK-2051A1 SA, AUX FW A REGULATOR IAF-49

FK-2051B1 SA, AUX FW B REGULATOR IAF-51

FK-2051C1 SA, AUX FW C REGULATOR IAF-50 N E: Steps 7.1.19 and 7.1.20 will time the MDAFW FCVs open. It is recommended that only one valve at a time be timed. However, it is permissible to time all three FCVs simultaneously.

19. SIMULTANEOUSLY START the stopwatch(es) AND SET the demand output signals for the following controllers to 100%:

FK-2051A1 SA, AUX FW A REGULATOR IAF-49

FK-2051 81 SA, AUX FW B REGULATOR IAF-51

FK-2051C1 SA, AUX FW C REGULATOR IAF-50

20. STOP the stopwatch(es) when the following FCVs indicate open:

FCV-2051A (IAF-49), SG A

FCV-2051B (1AF-51), SO B

FCV-2051C (1AF-50), SO C

21. RECORD valve stroke times from Step 7.1.20 on Attachment 4.

(j OST-1076 Rev. 29 I Page 12 of 31

7.1 Auxiliary Feedwater Pump I B-SB (continued)

N9I The demand output signal indicator accuracy is 2% for the following controllers:

. FK-2051A1 SA, AUX FWA REGULATOR 1AF-49

. FK-2051B1 SA, AUX FW B REGULATOR IAF-51

. FK-2051C1 SA, AUX FW C REGULATOR 1AF-50 f-N E If there is an auto open signal present the controller demand output will not remain at 0%.

22. SET the demand output signals for the following controllers to 0%:

. FK-2051A1 SA, AUX FW A REGULATOR IAF-49

-á-

. FK-2051B1 SA, AUX FW B REGULATOR IAF-51

. FK-2051 Cl SA, AUX FW C REGULATOR IAF-50

23. VERIFY the following FCVs indicate shut:

. FCV-2051A (IAF-49), SO A

. FCV-2051B (IAF-51), SG B

. FCV-2051C (IAF-50), SG C

/

Performance of the following Step will also initiate an LCO for Auxiliary Feedwater Pump I B-SB if in MODE 1, 2, or3 per Technical Specification 3.7.1.2.

24. OBTAIN a recirculation flow of 50 to 54 inches water column is indicated on the local test gauge (corresponds to 50 to 52 gpm) as follows:

(1i

a. Locally UNLOCK 1AF-22, AFWP I B Reciro to CST Isolation.

b. White counting turns from full open, THROTTLE IAF-22 to obtain 50 to 54 inches water column. )

1 j

25. RECORD the number of turns IAF-22 was throttled on Attachment 2.

26. ALLOW pump to run for at least 2 minutes after conditions are as stable as the system permits before recording data.

OST-1076 I Rev, 29 I Page 13 of 31

7.1 Auxiliary Feedwater Pump lB-SB (continued)

27. OPEN 1 CE-54-V2, P1-2251 B Instrument Valve.

28. RECORD I B-SB Pump operating suction pressure from test gauge installed in Step 7.1.1 on Attachment 2.

29. SHUT ICE-54-V2, Pl-2251B Instrument Valve.

30. OPEN IAF-26-HV2, P1-2151 B Instrument Valve.

31. RECORD I B-SB Pump operating discharge pressure from test gauge installed in Step 3.O.3.5.b on Attachment 2.

32. RECORD suction temperature from ERFIS point TCE9OIO or hand held pyrometer on Attachment 2.

33. RECORD Recirculation Flow differential pressure (inwc) from the local test gauge, on Attachment 2.

34. CHECK the following Recirculation Flow differential pressure (inwc) is within the acceptance criteria and document valve testing on Attachment 2.

I CE-46, CST Suction Check Valve to I B-SB AFW Pump

1AF-23, AFWP I B Recirc to CST Ck Valve

IAF-4, AFWP 1A Recirc to CST Ck Valve 1]

IAF-IlO, Aux Feed Pump IX Recirc to CST Ck Valve

35. CALCULATE AND RECORD Differential Pressure per NOTE 1 on Attachment 2.

2 cf Verified

36. CALCULATE AND RECORD Corrected Differential Pressure per NOTE 2 on Attachment 2.

oP Verified OST-1076 Rev. 29 I Page 14 of 3J

7.1 Auxiliary Feedwater Pump lB-SB (continued)

N The 1ST Program requires j readings to be taken in units of velocity (inches/second).

NQE: To ensure consistent readings are being recorded, the vibration data 7...4ecorded on Attachment 3 should be the values which the qualified operator stóred for each point in the Vibrometer.

NO E: /lf this test is only being run to obtain pump vibration data, the proper test flow and differential pressure conditions should be established to ensure consistent data is obtained for the 1ST program.

37. MEASURE vibration AND COMPLETE Attachment 3 using the values that were stored in the vibrometer.

38. LOCK OPEN IAF-22, AFWP 1 B Recirc to CST Isolation.

Verifled

39. At the MCB, VERIFY the following AUX FW MOTOR PUMP FLOW CONTROL VLVS shut:

FCV-2051A1 SA (1AF-49), SC A

FCV-2051B1 SA (IAF-51), SG B

FCV-2051C1 SA (IAF-50), SC C Stopping the AFW pump in the following Step will deenergize the MDAFW flow control valves and cause them to go to their Fail Safe Position.

40. At the MCB, STOP Auxiliary Feedwater Pump I B-SB.

Time Stopped: / 3.3

41. At the MCB, verify the following AUX FW MOTOR PUMP FLOW CONTROL VLVS open and initial for FAIL SAFE TEST on Attachment 4:

FCV-2051A1 SA (1AF-49), SC A

FCV-2051B1 SA (IAF-51), SC B

FCV-2051C1 SA (IAF-50), SC C j OST-1076 Rev, 29 I Page 15 of 31

7.1 Auxiliary Feedwater Pump I B-SB (continued)

The demand output signal indicator accuracy is 2% for the following controllers:

FK-2051A1 SA, AUX FW A REGULATOR IAF-49

FK-2051B1 SA, AUX FW B REGULATOR IAF-51

FK-2051C1 SA, AUX FW C REGULATOR IAF-50

42. At the MCB, SET the demand signal for the following controllers at 100%:

FK-2051A1 SA, AUX FW A REGULATOR 1AF-49 Verifled

FK-2051 BI SA, AUX FW B REGULATOR 1AF-51 Verified

FK-2051C1 SA, AUX FW C REGULATOR IAF-50 r2 U

Verified

43. INITIAL for the FULL STROKE TEST on Attachment 4 as verification of satisfactory valve operation for 1AF-49, IAF-51, and IAF-50. (As performed in Steps 7.1.15 through 7.1.21 and 7.1.41)

44. VERIFY IAF-49, IAF-51, and IAF-50 are open AND INITIAL for the Posttest Position Pos mit on Attachment 4.

45. INDEPENDENTLY VERIFY IAF-49, 1AF-51, and 1AF-50 are open AND p INITIAL for the Posttest Position Veil Init on Attachment 4.

The following step will restore the I B-SB AFW Pump to operable status.

46. LOCK OPEN IAF-208, AFWP lB Discharge Isolation.

7 L)

Verified

47. INFORM the CR8 that Auxiliary Feedwater Pump I B-SB is operable.

OST-1076 Rev. 29 Page 16 of 31

7.1 Auxiliary Feedwater Pump lB-SB (continued)

48. RECORD the time and date that Auxiliary Feedwater Pump I B-SB is operable on Attachment 6.

49. DIRECT Maintenance to remove the suction pressure test gauge at 1CE-54-V2, PI-2251B Instrument Valve and install pipe plug.

50. VERIFY ICE-54-V2 shut and plugged.

Verified

51. ISOLATE the discharge pressure test gauge by shutting IAF-26-HV2, Pl-2151B Instrument Valve.

52. DIRECT Maintenance to remove the discharge pressure test gauge at IAF-26-HV2, P1-21 SiB Instrument Valve and install pipe plug.

53. VERIFY 1AF-26-HV2 shut and plugged.

Verif1ed

54. DIRECT Maintenance to isolate the recirculation flow test gauge by shutting IAF-7-HV1, Fl-IAF-2172S Instrument Valve, and 1AF-7-LV1, Fl-1AF-2172S Instrument Valve.

55. DIRECT Maintenance to remove the recirulation flow test gauge at IAF-7-HVI, Fl-IAF-2172S Instrument Valve, and 1AF-7-LV1, Fl-IAF-2172S Instrument Valve and install pipe plugs.

(1

56. VERIFY the following are shut and plugged.

a. IAF-7-HVI, FI-1AF-2172S Instrument Valve Verified

b. I 7, Fl-i AF-21 726 Instrument Valve

/

Verified OST-1 076 Rev. 29 Page 17 of 31

7.2. Valve Testing REFER to Attachment 4 and test 1AF-93, AUX FW B ISOLATION, per the following instructions:

a. OBTAIN a calibrated stopwatch.

b. VERIFY IAF-93 is aligned to the Pretest Position AND INITIAL on Attach ment 4.

yPeormance of Step 7.2.1 .e will initiate an LCO for Auxiliary Feedwater Pumps IA-SA and I B-SB if in MODE 1, 2, or 3 per Technical Specification 3.7.1.2.

c. INFORM the CRS that Auxiliary Feedwater Pump IA-SA and 1 B-SB will be made inoperable.

d. RECORD the date and time that Auxiliary Feedwater Pumps IA-SA and 18-SB were made inoperable on Attachment 6.

e. SIMULTANEOUSLY START the stopwatch AND PLACE the control switch for 1 AF-93 to the position opposite the Pretest Position,

f. WHEN the valve has completed travel as indicated by a singular position indicating light for the demanded position (no dual indication),

THEN STOP the stopwatch.

g. RECORD the valve stroke time on Attachment 4.

h. SIMULTANEOUSLY START the stopwatch AND PLACE the control switch for IAF-93 to the pretest position shown on Attachment 4.

i. WHEN the valve has completed travel as indicated by a singular position indicating light for the demanded position (no dual indication),

THEN STOP the stopwatch.

j. RECORD the valve stroke time on Attachment 4.

k. VERIFY that 1AF93 has traveled to the Pretest Position as indicated by a singular position indicating light for the demanded position (no dual indication). /

j OST-1 076 Rev. 29 I Page 18 of 31

7.2 Valve Testing (continued)

I. INITIAL for the FULL STROKE TEST on Attachment 4 as verification of satisfactory valve operation (as performed in Substeps 7.2.1 .e through 7,2.1 .lc).

m. VERIFY IAF-93 is in the Posttest Position listed on Attachment 4 and initial in the space provided.

n. PERFORM independent verification of IAF-93, as required on 1)

Attach ment 4.

o. INFORM the CRS that Auxiliary Feedwater Pump IA-SA and I B-SB are operable.

p. RECORD the date and time that Auxiliary Feedwater Pumps 1A-SA and I B-SB are operable on Attachment 6.

7.3. Test Completion

1. REVIEW the Data on Attachments 2, 3, and 4 for applicable Acceptance Criteria being met. INFORM the CRS of any Acceptance Criteria not being met.

2. COMPLETE applicable portions of Attachment 6, Certifications and Reviews, AND INFORM the CRS that this OST is completed.

8.0 DIAGRAMSIATTACHMENTS Attachment I - Calibration Data Sheet Attachment 2- Performance Data Attachment 3- Vibration Data Attachment 4 Valve Test Data Attachment 5- Valve Retest Data Sheet Attachment 6 Certifications and Reviews OST-1 076 I Rev. 29 Page 19 of 31

Attachment I - Calibration Data Sheet Sheet I of I Inst/Model Description [

Inst ID Number Cal Due Date Vibrometer 0 to 50 PSIG Test Gauge J ZO/ 75 0 to 2000 PSIG Test Gauge J 5i 3 5 i/i 3/12 Stopwatch / iS Z2 4//6 / /

Stopwatch Stopwatch Hand_held_Pyrometer Differential Pressure Gauge, 0 to 100 inwc accuracy of +/- 1.0 inch

/i /, z j OST-1076 Rev. 29 I Page 20 of 31

Attachment 2 Performance Data Sheet I of 3 Auxiliary Feedwater Pump I B-SB Calculate Differential Pressure (P

) as follows:

1 q psid

/ - = )

Step7.1.31 Step7.128 Step7.t35 OT : Satisfactory Acceptance Criteria is determined by using the following equation and verifying the corrected differential pressure is within the acceptance criteria.

3 = 62.31 X /, )Y x o. o) oS tPcorr APind tPcerr = Corrected Differential Pressure

= Indicated Discharge Pressure minus Indicated Suction Pressure Specific Volume (ft Ilbm) for saturated liquid water (Ui) for suction temperature 3

recorded in Step 7.1.32. This number can be determined from Sheet 3 of this Attachment or Standard Steam Tables.

Step Component Tested Acceptance Criteria SatlUnsat (circle 7.1.6 IAF-201 and IAF-54 stroke close. TAF2007A < 135°F }UNSAT 7.1.6 IAF-202 and IAF-92 stroke close. TAF2007C < 135°F UNSAT 7.1.6 IAF-203 and 1AF-73 stroke close, TAF2007E < 135°F ijbUNSAT 7.1.34 ICE-46 Stroke Open 50 to 54 inwc Ai)I UNSAT 7.1.34 IAF-23 Full Stroke Open 50 inwc UNSAT 7,1.34 IAF-IlO Stroke Closed 50 inwc UNSAT 7.1.34 IAF-4 Stroke Closed 50 inwc (s) UNSAT OST-1076 Rev. 29 Page 21 of sij

Attachment 2 Performance Data Sheet2of3 INSTRUCTIONS

1. If pump hydraulic data does not meet acceptance crfteria:

a. Declare the pump INOPERABLE.

b. lriitiateaCR.

Auxiliary Feedwater Pump I B-SB (AFI B-SB)

STEP ACCEPTANCE NO, PARAMETER INSTRUMENT READING CRITERIA Test Gauge 7.1.4 Idle Suction Pressure 15 psig atlCE-54-V2 }

Numberofturns N/A 7125 S N/A I AF-22 throttled c?L Operating Suction Test Gauge 7.1.28 15 psig Pressure at 1CE54V2 Test Gauge 7.1.31 Discharge Pressure psig at IAF-26-HV2 )t (Q 7

N/A rw 7.1.32 Suction Temperature -73 °F N/A Pyrometer Recirculation Flow Test Gauge 7.1.33 50 to 54 inwc **

(Local Flow Gauge) at IAF-7 inwc Differential Pressure CaIcuted per N/A 7.1.35 ,/j1...PSlD 1530.3 psid Calculated per NOTE 2 an d 7.1.36 Differential Pressure )/ , PSID (Corrected (APcorr) 1775.4 psid c Circle instrument used.

(f 50 to 54 inches water column is equivalent to 50 to 52 gpm

Differential Pressure Reference Value = 1614 psid OST-1 076 I Rev. 29 Page 22 of 3J

Attachment 2- Performance Data Sheet3of3 Specific Volume (ft

/ibm) for Saturated Liquid Water (Uf) 3 Temperature Specific Volume Temperature Specific Volume Temperature Specific Volume 59 0.01603 80 0.01607 101 0.01613 60 0.01603 81 0.01607 102 0.01613 61 0.01603 82 0.01607 103 0.01614 62 0.01603 83 0.01608 104 0.01614 63 0.01603 84 0.01608 105 0.01614 64 0.01604 85 0.01608 106 0.01615 65 0.01604 86 0.01608 107 0.01615 66 0.01604 87 0,01609 108 0.01615 67 0.01604 88 0.01609 109 0.01616 68 0.01604 89 0.01609 110 0.01616 69 0.01605 90 0.01610 111 0.01617 70 0.01605 91 0.01610 112 0.01617 71 0.01605 92 0.01610 113 0.01617 72 0.01605 93 0.01610 114 0O1618 94 0.01611 115 0.01618 74 0.01606 95 0.01611 116 0.01618 75 0.01606 96 0,01611 117 0.01619 76 0.01606 97 0.01612 118 0.01619 77 0.01606 98 0,01612 119 0.01620 78 0.01606 90 0.01612 120 0.01620 79 0.01607 100 0.01613 121 0.01620 OST-1076 I Rev. 29 Page 23 of 31J

Attachment 3 Vibration Data Sheet I of I Auxiliary Feedwater Pump I B-SB (AFI B-SB)

I NSTRUCTIONS:

1. IF pump vibration data is outside the Acceptable Range but within the Alert Range, THEN PERFORM the following:

a. INITIATE an STR to direct the Surveillance Testing Scheduling Coordinator to double the testing frequency of the pump.

b. INITIATE an NCR.

c. RECORD the STR and NCR numbers in the Comments Section.

2. IF pump vibration data meets the Required Action Criteria, THEN PERFORM the following:

a DECLARE the pump INOPERABLE

b. INITIATE an NCR.

Step 7.1.37 Reference ACCEPTANCE CRITERIA Value Velocity Vibration Location in/sec (in/sec peak)

Acceptable Range 1 Alert Range Required Action Axial 0.135 0.325 > 0.325 to 0.70 > 0.70 Outboard 0.113 0.2825 > 0.2825 to 0.678 > 0.678 Horizontal O Outboard 0.152 0.325 > 0.325 to 0.70 > 0.70 Vertical o Inboard 0.14 0.325 > 0.325 to 0.70 > 0.70 Horizontal p Inboard Vertical 0.17 0.325 > 0.325 to 0.70 > 0.70 j OST-1 076 Rev. 29 I Page 24 of 31

Attachment 4- Valve Test Data Sheet I of 1 The shut time limiting values come from a mathematical multiple of the reference value or from ESR 97-00475 (19 seconds),

c7 whichever is less.

The open time limiting values come from a mathematical multiple of the reference value or from PCR-06502 and ESR 97-00475 (20 seconds), whichever is less.

PRETEST POSHEST REF ALIGNMENT FULL STROKE TEST FAlL SAFE TEST ALiGNMENT ACCEPTANCE CRITERIA (SEC)

(sec.)

Verification Stroke Time Fail Safe Position Posttest Pos Verf CODE CRITERIA LIMITING VALUE Valve Pretest mit of Travel (sec) Position Verified Position mit hit SHUT OPEN SHUT Number Position by md l Lights (lnit)IOPEN SHUT OPEN LOW HIGH LOW HIGH OPEN_[ SHUT 17.15 14.58 19.72 23.00 IAF-93 OPEN N/A N/A OPEN

4. -711_ 3 17.33 15.26 14.73 19.92 11.45 19.00 25.99 19.00*

1AF-49 OPEN j1 OPEN ., OPEN 1180 8.85 14.75 20.O0**

(1 14.86 11.15 1&57 19.00*

IAF-51 OPEN OPEN . OPEN 20.00**

14.13 10.60 17.66 15.20 11.40 19.00 19.00 1AF-50 OPEN

, i OPEN OPEN c 9.80 4.90 14.70 20.00**

Comments:

OST-1 076 I Rev. 29 Page 25 of 31

Attachment 5 Valve Retest Data Sheet Sheet I of 2 /

It N E This entire Attachment is N/A if no valve is retested due to exceeding t,h Code Criteria.

Determine if the stroke time exceeds the Limiting Value.

I. If the stroke time exceeds the Limiting Value, declaWihe valve inoperable and initiate a CR. (N/A if stroke time is less than thefrifniting Value>

2. If the stroke time is less than the Limiting Val but outside the Code Criteria limits, perform the following Steps:

1

a. If the cause is known to be mecical failure, or if a retest cannot be performed expeditiously, dec10 the valve inoperable and initiate an NCR (Except PMTRs). //

b. If retesting the valve is,red, perform the following:

NOTE: If necessary, separate markeA up sheets of this OST may be used to document necessary manipdlations. These sheets would be attached to this procedure and noted in comments Section of Attachment 6.

(I) D ermine which Steps need to be performed to set up conditions for sting the valve. CRS concurrence must be obtained and

/documented in the Comments section of Attachment 6.

(2/ Perform the Steps determined in the previous Step and document

/ stroke times/valve positioning on Sheet 2.

iJ

/ (3) If retest results are still outsIde the Code Criteria, declare the valve

/ inoperable and initiate an NCR (except PMTs).

(4) If retest results are within the Code Criteria, perform the following:

// (a) Declare the valve operable.

/

(b) Initiate an NCR identifying test findings for the first and second tests.

(c) Send test results to Responsible Engineer (1ST) for evaluation and documentation on the CR.

TOST-1076 Rev. 29 Page 26 of 31

Attachment 5 Valve Retest Data Sheet Sheet 2 of 2 (1) Fill out PRETEST ALIGNMENT, POSHEST ALIGNMENT, and ACCEPTANCE Cfifl values for the valve(s) being tested using the values in the initial test Attachment.

PRETEST POSTTEST ALIGNMENT FULL STROKE TEST ALIGNMENT ACCEPTANCE CRITERIA (SEC)

(1) (1) (1)

Z CODE CRITERIA Stroke Time (SEC) OPEN SHUT LIMITING VALUE Valve Pretest Posttest Number Position Init OPEN SHUT Positi9W Pos mit Verf Init Low High Low High OPEN SHUT z

z V

77

/

V z

OST1 076 Rev. 29 I Page 27 of 31

Attachment 6 Certifications and Reviews Sheet 1 of 2 This OST was performed as a: Periodic Surveillance Requirement:

Postmaintenance Operabi Nty Test:

Redundant Subsystem Test:

Plant Conditions: J ,uoP7 Mode: __)

OST Completed By:

Date:

Time:

OST Performed By:

Initials Name (Print) Initials Name (Print) 7-A A-)

aP f1 4 Le I..

fJ Re1/e>-

General Comments/Recommendation/Corrective Actions/Exceptions:

OST-1 076 Rev. 29 Page 28 of 31

Attachment 6 Certifications and Reviews Sheet 2 of 2 Time Date Step 7.1.11 I B-SB AFW Pump Inoperable 5C 2)/c //(

Step 7.1.48 1 B-SB AFW Pump Operable 9

/0 Step 7,2.1.d IA-SA and lB-SB AFW Pump Inoperable /c)/ 7

///OJf(

Step 7.2.1.p 1A-SA and I B-SB AFW Pump Operable Pages Used: it (-

OST Completed with NO EXCEPTIONS/EXCEPTIONS:

Date:

CRS Reviewed By: Date:

Responsible Engineer (1ST)

After receiving the final review signature, this OST becomes a QA RECORD and should be submitted to Document Services.

OST-1076 Rev. 29 I Page 29 of 31

Revision Summary General (Revision 29)

This revision incorporates the following PRRs:

PRR 411380: From HNP 1ST Formal Self-Assessment 376263 Ri HNP Is considered an outlier from other fleet 1ST programs as well as the industry with regards to having the reference values in the surveillance test procedure. Reference values inserted into the applicable surveillance test will provide Operations an instant comparison of the value used to develop the current code acceptance criteria (expected value) to that of the as-found. Should test anomalies be identified, a comparison of 1ST component trends available from the Operations Central Database can be utilized for quick analysis and evaluation. The ASME CM Code appears to require this information.

PRR 442814: Editorial correction to Step 7.1.2 which says test gauge test gauge. Recommend the following changes per l&C input:

Reword Step 3.0.5 to Direct Maintenance to install the following:

Change Steps 7.1.1 and 7.1.2 to verify gauges installed.

Move Step 7.1.6 up to be new Step 7.1.3 to verify gauge installed.

Step 7.1.1 should also be split into multiple steps due to two actions.

The Valve in 1CE-54-V2 portion of Step 7.1.1 should be a new step prior to recording idle suction pressure.

Step 7.1.7 should be changed to Verify Maintenance has valved in and vented the recirculation flow test gauge....

Step 7.1.49 is redundant to step 7.1.29.

PRR 444379: Incorporate re-baseline data for IAF-49, IAF-50, and IAF-5i per lSl-801 dated 1/26/2011. This is the result of maintenance that replaced the NTD driver cards and these valves have been stroke timed twice, allowing a re-baseline based on the results obtained.

This revision will remove the procedure from ADMIN HOLD.

Description of Changes Page Section, Change Description 5 3.0.5 Changed this Prerequisite from verifying Maintenance support is available to install instrumentation to directing Maintenance to install instrumentation. (PRR 442814) 9 7.1.1 Changed these steps from directing Maintenance to install and valve 7.1.2 in 1 B-SB AFW Pump suction and discharge pressure gauges to verify these instruments are installed and valved in.

Split out the signoff lines into two signoffs for each step; one for installed and one for valved in. (PRR 442814)

OST-1 076 Rev. 29 I Page 30 of 31

Revision Summary Page Section Change Description 9 7.1.3 Moved the following step from 7.1.6 to 7.1.3 to be closer to other instrument installation verification steps: (PRR 442814)

VERIFY the 0 to 100 inwc test gauge is installed at 1AF 7 HV1, Fl IAF 2172S Instrument Valve, and IAF 7 LVI, Fl 1AF 2172S Instrument Valve.

10 7.1.7 Changed the step from verifying Maintenance has installed the recirculation flow test gauge to directing Maintenance to valve in and vent the flow test gauge. (PRR 442814) 17 7.1.49 Deleted the step since it is redundant to Step 7.1.29: (PRR 442814)

ISOLATE the suction pressure test gauge by shutting I CE-54-V2, P1-22518 Instrument Valve.

22 Attachment 2 Added a reference for Auxiliary Feedwater Pump 1 B-SB differential pressure. (PRR 411380) 24 Attachment 3 Added a reference for Auxiliary Feedwater Pump I B-SB vibration parameters. (PRR 41 1380) 25 Attachment 4 Updated baseline stroke time values for IAF-49, 1AF-50, and IAF-51. (PRR 444379)

Added Reference Values for stroke time tests. (PRR 411380) j OST-1076 I Rev. 29 Page 31 of 31 I

Appendix C Page 1 of 11 Form ES-C-i

- Worksheet Facility: Shearon Harris Task No.:

Task

Title:

Using survey maps determine stay JPM No.: 201 1 NRC Exam times Admin JPM RO A-3 K/A

Reference:

G.2.3.4 RO 3.2 SRO 3.7 Examinee: NRC Examiner:

Facility Evaluator: Date:

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

Initia Conditions: Two individuals are tasked to hang a clearance on 1 CS-38, Letdown PCV sol vlv. The clearance includes the following valves:

1CS-35

1CS-36

1CS-37

1CS-38

1CS-39

1CS-40

1CS-43 Initiating Cue: Two Operators are being assigned to hang a clearance and perform work in a radiological area. Operator 1 has an accumulated annual Whole Body dose of 1750 mrem (Progress Energy). Operator 2 has an accumulated annual Whole Body dose of 700 mrem (Progress Energy) and worked at Nine Mile Point earlier this year where he has accumulated 2550 mrem.

The clearance and work activity is for 1CS-38, Letdown PCV Isol vlv. The ALARA group has determined that additional shielding is not warranted for this work.

Using the supplied survey map, determine the maximum allowable individual stay times for each Operator that would prevent exceeding the Progress Energy Annual Administrative dose limit while performing these activities.

Do not consider dose received during transit. The calculated dose should be ONLY what they would receive while working at the valves for the clearance.

Complete the information below and return to the evaluator when complete.

2011 NRC Exam Admin JPM RO A3 Rev. Final

Appendix C Page 2 of 1 1 Form ES-C-i WorRstet Task Standard: Calculation of stay times based on survey maps, two hours for Operator 1, six hours for Operator 2.

Required Materials: Survey map A45 RAB 236 LETDOWN & LETDOWN REHEAT HX &VG Map 21 SFD-5-S-1 304 General

References:

DOS-NGGC-0004 Progress Energy Annual Administrative Dose Limits Section 9.3 (Rev.12)

LIMIT = 2 rem Progress Energy dose not to exceed 4 rem total dose if non- Progress Energy dose for the current year has been determined.

Time Critical Task: No Validation Time: 10 minutes Critical Task Justification Step 1 Must determine dose rates in order to calculate stay time Step 2 Must determine available dose to determine stay time.

IF incorrect calculation of stay time is made the individuals could exceed St ep 3 their dose limits.

2011 NRC Exam Admin JPM RO A3 Rev. Final

Appendix C Page 3 of 1 1 Form ES-C-i PERFORMA1iCE INFORMAT1OFSI START TIME:

Evaluator Note: The order of performance does not matter IF THE APPLICANT ASKS FOR IT: Provide a copy of DOS NGGC-0004, Rev. 12 v Performance Step: 1 Using Radiological Survey Record Map A45 and RAB 236 LETDOWN & LETDOWN REHEAT HX &VG Map 21, determines dose rates in the area where the clearance will be applied Standard: Identifies that General Area Dose Rates are 125 mrem/hr Comment:

V Performance Step: 2 Determine the remaining dose for the year for each individual Standard: Operator 1: 250 mrem 2000 mrem 1750 mrem

- = 250 mrem Operator 2: 750 mrem 4000 mrem 700 mrem (PE)

- - 2550 mrem (NMP) = 750 mrem Comment:

V - Denotes Critical Steps 2011 NRC Exam Admin JPM RO A3 Rev. Final

Appendix C Page 4 of ii Form ES-C-i PERFORMANCE INFORMATIOI\1 v Performance Step: 3 Determine stay time for each operator (based on 1 St Operator reaching 2 Rem and the 2nd Operator reaching 4 Rem for the year)

Standard: Operator 1: 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 250 mrem ÷ 125 mrem/hr = 2 hrs Operator 2: 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 750 mrem ÷ 125 mrem/hr = 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Comment:

Terminating Cue: After the stay time has been calculated, this JPM is complete.

END OF JPM STOP TIME:

V - Denotes Critical Steps 2011 NRC Exam Admin JPM RO A3 Rev. Final

Appendix C Page 5 of 1 1 Form ES-C-i VERIFICATION OF COMPLETION Job Performance Measure No.: 2011 NRC Exam Admin JPM RD A-3 Using survey maps determine stay times. (DOS-NGGC-0004 Rev. 12)

DOS-NGGC-0004 Progress Energy Annual Administrative Dose Limits Section 9.3.

LIMIT = 2 rem Progress Energy dose not to exceed 4 rem total dose if non- Progress Energy dose for the current year has been determined.

Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2011 NRC Exam Admin JPM RD A3 Rev. Final

Appendix C Form ES-C-i JPMCUESHEE1 Initial Conditions: Two individuals are tasked to hang a clearance on 1 CS-38, Letdown PCV Isol vlv. The clearance includes the following valves:

. 1CS-35

. 1CS-36

1CS-37

1CS-38

1CS-39

. 1CS-40

. 1CS-43 Initiating Cue: Two Operators are being assigned to hang a clearance and perform work in a radiological area. Operator 1 has an accumulated annual Whole Body dose of 1750 mrem (Progress Energy). Operator 2 has an accumulated annual Whole Body dose of 700 mrem (Progress Energy) and worked at Nine Mile Point earlier this year where he has accumulated 2550 mrem.

The clearance and work activity is for i CS-38, Letdown PCV lsol vlv.

The ALARA group has determined that additional shielding is not warranted for this work..

Using the supplied survey map, determine the maximum allowable individual stay times for each Operator that would prevent exceeding the Progress Energy Annual Administrative dose limit while performing these activities.

Do not consider dose received during transit. The calculated dose should be ONLY what they would receive while working at the valves for the clearance.

Complete the information below and return to the evaluator when complete.

Name Record the maximum allowable stay time calculations below to the nearest hour and minute.

Operator 1: Operator 2:

2011 NRC Exam Admin JPM RO A3 Rev. Final

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Appendix C Form ES-C-i JPM CUE SHEET RAB 236 LETDOWN & LETDOWN REHEAT MAP 2.1 HEAT EXCHANGERS/VALVE GALLERY c 9 I i L

Li iw

Jti A -1O W jc 4)

LETDOWN UX

<PHOTOGRAPHS V

l. II II AREA RAD MONITOR

[j 2011 NRC Exam Admin JPM RO A3 Rev. Final

Appendix C Form ES-C-i JPM CUE SHEET MapA4S RADIOLOGICAL SURVEY RECORD Page of I Unitt j i34dgE1ev: RAB23O Area: LetdownHX&VG Dat: 07/01111 lT 0735W jKw?# 0992-011 NIA Clearance and work on ICS-3S Surveyor Tech Sxgnature [iwie e4 by B Bossmann No ently require$ into letdown HIX ro no shielding tddi for job # iiioom Locatios I c1!

2 .1K +/-lc.or 3 6K flor(hydr&n)

--- 4 lOOK floor

- 5 lK +/-o uide va1ve) 6 :1K piping JTz 30K I

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N 9 1K IcS-43 C_-:

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I:MI - /1 ri 1

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ARA Aeborne RaAxe C) SLocit % MttWy

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AirSampkLocauon No DAC uCtlcc i CA CootaounaootArea -

RMA Rathoactwe MatenabArea K 1000 1aeRatein mreaVhr un1as otheiwe noted N -

Ad uci/e QA Record 2011 NRC Exam Admin JPM RO A3 Rev. Final

C L1 c) 0 0

C x

LU 0

( z C/)

LU C

c\J 0

LL I

LU LU I

(I)

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Appendix C Form ES-C-i

. JPM CUE SHEET L

&FM1 Jt flLbrL1L r2 r -

2011 NRC Exam Admin JPM RO A3 Rev. Final

Appendix C Job Performance Measure Form ES-C-i Worksheet Facility: Shearon Harris Task No.: 301 005H401 Task

Title:

Determine Rod Misalignment Using JPM No.: 2011 NRC JPM Thermocouples and Evaluate Tech SRO Al-i Specs K/A

Reference:

G2.17 4.4/4.7 Examinee: NRC Examiner:

Facility Evaluator: Date:

Method of testinci:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAM INEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

The plant was at 90 percent power, with a load decrease in progress, when the CRS noticed that the DRPI indication for rod H02 was reading 24 steps higher than the group demand.

The load decrease has been stopped and AOP-OOi entered.

Initial Conditions:

ALB-13/8-5, Computer Alarm Rod Dev/Seq NIS Power Range Tilts alarm is the only MCB alarm received.

I&C investigated and found no obvious electrical problems.

The CRS has directed you to calculate the temperature difference between thermocouple(s) adjacent to the misaligned rod and the Initiatin g Cue average of symmetric thermocouple(s), using Attachment 2 of AOP-OOi and the provided T/C Core Maps.

2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Job Performance Measure Form ES-C-i Worlcshoet -.

Task Standard: All calculations within +/- 2° of actual.

Required Materials: Calculator General

References:

AOP-OOi, Attachment 1 and Attachment 2, Rev. 35 and Tech Specs Handouts: JPM Cue Sheets Pages 10-11 AOP-00i, Attachments 1 and 2, (Rev. 35)

Time Critical Task: No Validation Time: 15 minutes Critical Task Justification If the wrong thermocouples are used then none of the results will be St ep 2 correct Step 3 If the wrong values are selected then none of the results will be correct If the calculation for the averages were incorrect the results will be Step 4 incorrect If the differences are calculated incorrectly then the candidate will come ep to the wrong conclusion for Tech Specs If the wrong Tech Spec Action is selected an LCO action could be St ep 6 exceeded 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 3 of 21 Form ES-C-i

- PERFORMANCE INFORMATIQN Start Time:

Performance Step: 1 OBTAIN PROCEDURE (provided with handout)

Standard: Obtains AOP-0Ol and refers to Attachments 1 and 2.

Comment:

V Performance Step: 2 DETERMINE THERMOCOUPLE LOCATION(S) ADJACENT TO THE MISALIGNED ROD USING THE CORE GRID MAP (SHEET 1), AND CIRCLE LOCATION(S) IN TABLE ABOVE.

THESE THERMOCOUPLES(S) ARE AFFECTED.

Standard:

Using the core grid map (Attachment 2, page 1 of 2),

Determines affected thermocouples to be G02, J02, and H03.

Circles G02, J02, and H03 on the table (Attachment 2, page 2 of 2).

Comment:

V Performance Step: 3 RECORD VALUES FOR ALL OPERABLE AFFECTED AND SYMMETRIC THERMOCOUPLES USING THE RVLIS CONSOLE OR ERFIS. SYMMETRIC THERMOCOUPLES ARE THOSE IN THE SAME ROW.

Standard: Locates RVLIS Console and accesses T/C CORE MAP for Train A and Train B. (Printout of RVLIS core map provided in handout)

Records value for Affected TC G02 (652° F) and Symmetric TO P07 (636°F).

Records value for Affected TO J02 (652°F) and Symmetric TO P07 (636°F).

Records value for Affected TO H03 (649°F) and Symmetric TCs 008 (636°F), H13 (640°F), and N08 (644°F).

Comment:

V - Denotes a Critical Step 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 4 of 21 Form ES-C-i PERFORMANCE It1FQRMATIQN V Performance Step: 4 DETERMINE THE AVERAGE OF SYMMETRIC THERMO COUPLES, FOR EACH AFFECTED THERMOCOUPLE.

Standard:

Determines (636°F) for G02s Symmetric TC Determines (640°F +/- 2°F) for H03s Symmetric TC Determines (636°F) for J02s Symmetric TC Comment:

If the candidate includes the adjacent TCs with the EXAMINERS NOTE: Symmetric TC numbers the averages will be wrong and the end result will be that a wrong final difference will be given:

Determines (644°F) for G02s Symmetric TCs Determines (640°F) for H03s Symmetric TCs.

Determines (644° F) for J02s Symmetric TCs.

V - Denotes a Critical Step 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 5 of 21 Form ES-C-i PERFORMANCE INFQRMAT1Qt4 V Performance Step: 5 CALCULATE THE DIFFERENCE BETWEEN EACH AFFECTED THERMOCOUPLE AND THE AVERAGE OF ITS SYMMETRIC THERMOCOUPLES.

Standard: Calculates difference of 16°F 16°F for TC G02 9°F for TC H03 16°F forTCJO2

(+/- 2°F between all affected TCs and their symmetric TCs.)

Reports that temperature difference is greater than 10°F.

Evaluator Cue: CRS acknowledges calculations and report.

(IF calculation difference is >10°F then)

Cue: Evaluate Tech Specs for compliance.

Comment:

EXAMINERS NOTE: If the candidate included the adjacent TCs with the Symmetric TC numbers the averages would have been wrong and the calculated difference would be 8°F.

The report would be that the temperature difference is less than 10°F. This report would provide the CRS with inaccurate information.

V - Denotes a Critical Step 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 6 of 21 Form ES-C-i PRFQRMANQEINfORMATIQN Performance Step: 6 OBTAIN AND EVALUATE TECH SPECS Standard: Obtains Tech Specs and refers to LCO 3.1.3.1 Determines that ACTION a. is applicable. (See Attached)

(IF candidate incorrectly calculated the temperature difference the Tech Spec reported will be 3.1.3.2 ACTION a)

Evaluator Cue: CRS acknowledges Tech Spec call.

ENDOFJPM Comment:

Stop Time:

V - Denotes a Critical Step 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 7 of 21 Form ES-C-i

. PERFORMANCE INFQRMATIQN KEY I MALFUNCTION OF ROD CONTROL AND [NOICATION SYSTEM I Attachment 2 Adjacent and Synimetrc Thermocoupte Locations Sheet 1 of 3 THERMOCOUPLE LOCATIONS 1 2 3 4 5 6 7 8 9.101112131415 A *: ..

i iI j_........._________________L_____L__j i4 I, I B

I ft I I I I I I I C ...

T ft ft T i -

-a D

E ft T ft T T T ft T T F ft T ft ft ft T ft T ft 1 T ft 7 ft 7 ft G ft H (T ft 7 T ft 7 T ft T J 7 R R T T R K ft L ft 7 ft 7 ft M 7 ft T 7 ft N T T 7 ft T P I ftT R T R Control Rod T ThernocoupIe T* Thermocouples abandoned by EC 47997 AOP-ODI Rev. 35 Page 41 of 47 V - Denotes a Critical Step 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 8 of 21 Form ES-C-i

. PERFORMANQ LNEQRMATIQN KEY I MALFUNCTION OF ROD CONTROL AND INDICATION SYSTEM I Attachment 2 Adjacatit an.d Syrnnetric Thecmocou pie Locations Sheet2of3 NOTE

. 510. E07, Ff08. K08. and P08 have no symmetric iocations.

. Symmetric thermocouples are those in the same row.

SYMMETRIC LOCATIONS GRID IV TRAIN A B A B A B A B A08 1-115 G01 015 R07 S L B05 214 L14 Y 0 COB H13 NOB HO3)

M C D03 C12 N04 M03 M A E04 D05 E12 Mu L12 E T Hil 208 L08 H05 T I F05 Eli 210 Kil* K05 L06 R 0 F03* F13 NW N06 K03 I N GOB F09 J1O C S 008 H09 (9O) (JO2) P07 MOO Ji2 Thermocouples abandoned by EC 47997 I. DETERMINE thermocouple location(s) adjacent to the misaligned rod using core grid map (Sheet 1), AND CIRCLE location(s) in Table above AOP-COi Rev. 35 Page 4.2 of 47 V - Denotes a Critical Step 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 9 of 21 Form ES-C-i PERFQRMANCEIIWQRMATIQN KEY TIC CORE MAP (DATA SHEET 1) TRAIN B MSG#245 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 A I A B 632 B C I 7 C D 636 650 E 644 50 640 632 1 E F 50 644 I F G 150 644 632 HI 644 IH J I 640 50 K I 640 I K L I 644 640 644 o I L M 636 M N I 640 N P I P R R 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 TRAIN A TIC CORE MAP MSG # 245 1 2 3 4 12 13 14 15 A 50 A B 636 B C 632 C D D E 636 640 E F 644 648 644 F G 640 G H 640 644 652 H J J K 640 644 50 K L L M 640 644 M N 640 644 N P 640 P R 636 R 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 V - Denotes a Critical Step 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 10 of 21 Form ES-C-i PERFORMANCE LNFQRMAT1QN KEY I MALFUNCTION OF ROD CONTROL AND INDICATION SYSTEM I Attachment 2 Adjacent and Symmetric Thermocouple Locations Sheet 3 of 3

2. RECORD the following n the table below:

Adjacent TC number

Adjacent TC value using the RVLIS Console, ERFIS, or OSI-PI

Symmetric TC numbers (not including adjacent TCs)

Symmetric TG values for all OPERABLE TCs us:ing the RVLIS Consoler ERFIS, or OSI-PI

3. DETERM INE the average of mmetric thermocouples, for each adjacent thermocouple.

Adjacent TC Symmetric TC Symmetric TC Number Value Number Vahe Average J02 N/A P07 G02 52 C0.9 636

{13 G40 Mc G4 N0 44 Gd)

C02 N/A P07 3G J02 G52 636

4. COMPARE each adjacent thermocouple value listed to its symmetric thermocouple average for indication of a misaligned rod. (REFER TO Attachment 1.)

--END OF ATTACHMENT 2--

AOP-OO I Rev. I Pa!Je 43 of 47 V Denotes a Critical Step

- 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page ii of 21 Form ES-C-i PEREQRMANCE INFORMATION KEY FM 3RL 1.. 3. 1OVAi3LE CORJL S$E*L GROUP HGHT LlMfTflG CONDITFOU FOR OP1RATlCl

3. 1 .3 1 All shutdcn arid control rods shall be OPERABLE ad positioned wi thin L2 steps (in cated cioslticn) f their group step courter deniar4 postiri, APPLICABILOTY: 92OES 1 arid 2,

a. with one or iiore rods inoperable due to being irrtinovable as a result of excessive friction or iechanical interference or knctn to be uritrippable, determine t[mat the 5HUTUOUN M.G1N reulrenient of Specification 31.11 ts satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months amd be in NOT 5TANGt within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With noro than one rod miasaligned frccn the group stOp Counter deiriand position by riore than .t 12 steps (indicated position), be in HOT STANOY within S hours.

c. with more trian one rod Inoperable, due to a rod control urgent faiure alarn or obvious electrical probien ifl the rod coitrol systE1 existing for greater than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , be in NOT STANOBY within the follcing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. with one rod trippable but inoperable due to causes other than addressed by ACIION a... above. or misaligned from its group step counter denand height by irre than +/- 12 steps (indicated position), POSER OPERATION nay continue provided that within

1. hour I The rod is restored to OPF3M%LF status within the above alignment requirenents, or

2. The rod is declared inoperable and the remainder of the rods in the uroup with the innperable rod are aligned to within

+/- 12 steps of the inoperable rod while maintaining th rod sequence and insertion 1 nutS of Specification 3,1,3,6, The ThERNP& POWER 1eiel shall be restricted pursuant to Specification 3.1.36 during subsequent operation, or

3. The rod is declared inoperable and the SHUTO<JN MAGRN requirement of Specification 3.1.1,1 is satisfied. PJiE.R CPERAiIO nay then continue provided that:

A reevaluation of each accident analysis of Table 3.ii it performed ithiru 5 days: tins reevaluation shalT confirai that the previously atalyzed results f these acci.dents See Special Test Exceptions Specifications 3.10.2 and 3,10.3.

SHEARON HARRIS UNIT 1 314 1-14 .Aniendnuent Wo, 2 V - Denotes a Critical Step 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 12 of 21 Form ES-C-i PRFQRMANCINfQRMAT1QN KEY ftLACI tV1T CLP L M[tiNG Ccor : ON F OR OPTO ACTION ICcnt lnueO):

reoain val id Tm the duration of operation utder these conditions:

b) The SF4UTIJU tARIN renuirerient of pecticotion 3.1. 1,1 s determined at least mice per 12 tours:

0) A pcer distribution riiap s obtarrmeb froom the roabie mcore detectors and 0 F

( Z) and F are icr ified to be within their !iniits within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s: and dl The THFRt1 P0FR lcel s reduced to lcs. than or equal to 75 of RATIO FHERWmL (NER vdthin the next hour and vehin the follow ng 4 hcurs the Fip1 Neutron Flux Trip Sctpofnt s reduced to less than or equal to 5t of RkIED rNERtAL, PQWR.

FLCANCE RElRE1EN1S 4.1.3. 1.1 Th rc.iL:on or edch rod snal I c detrms;nd t e tvithin :he nroup dorrnd l:nit by verifying the individual rod postamis at least once per 12 hourS exct during te nteraIs ri the rod posttioi d1atiri mri tar is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.OJ,2 Each rod rot fully inserted n the core shall be deterniined to e OP[RBLE by bOvOt Of at 1m?St 10 steps 10 any one direCtiOrT 3t least On per 92 days.

H.4h0 irmrdmcnt ü. 93 V Denotes a Critical Step

- 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 13 of 21 Form ES-C-i PEFQRMkNQ [NfQRMATlQN KEY TABLE 3I1 ACCtEN1 MMUSS REUttNG EVALUAflON IN ThE VT OF N INOP RO Rc Custer Cntro1 AssmD1y Ths.rton Charact4rlstics od C1 ust Cntro1 A*s ely 4i sa iss f Reactoi Ca*lnt fa Sai1 Ruptured Pips

lros Cracks l.are P1.s Which tiatas th rçncy Cri Cool ing Singla Rod C1urtr Control Assal Wittdrawa1 at u1l Power ajo aaet Coolant Syst RIps uptur.s (L3sa-fCoolart Ac1 øit) 4aor crdar1t Coo ant Syst Pipe Ruur, Rupture of a Cotro1 Rd Drive echartfs Kcsiq (Rod Custer Control Aetly ectan)

SfEARC?4 iLARRIS UNfl 1. 3/4 116 V - Denotes a Critical Step 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 14 of 21 Form ES-C-i

- PERFORMANCE INFQRMATIQN -

KEY REPCTTVITY C0N120L SYSTt4S POSZflON rNQtATtON S(iEi5 OPERATt4G I

LflING CCNDITtaN FOR OPERATION 3.LZ The Dlgttal Rod PsItton tn4lctf on Systan and the Demand Po*ttlcn Indication System shall be OPERABJ.E and capahle of tarolntng the shutdcwn and control rod positioni within +/- U steps.

APPUCABILrr(: IGES l and L ACTTON:

a With a eaziia: of on, digital rod position indicator per bank inoperable

1 ther

1. Determine the position of the noniridicating rod(s) lnolractly by the savable Incoro detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and liemedi.ately after any notion of the nonindicating rd which exceeds Z4 steps in one drect1on since th* lest eterinatIon f the rodLs position, or

2. Raduci TNERMAI. POWER to less than 5 of RATED ThERMAL POWER witMn 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ..

b. With a *axizs of one demand position Indicator per bank Inoperable aitheri

1. Verify that all digital rod position indicators far the affected bank are QPERA8l.La that the sast wi thdrr.m rod and the least withdrmms rod of the bank as-withln a ziz.e of U steps of each other at l.ast once per 8 or Z. Reduce ThERMAl. POWER to iss than 5 of RATED ThERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REGUIREMENTS 4.]L2 Each di4ltal rod poitiø Indicator shall be determined to be OPERASLI by verifying that the Dand Position Indication System and the Digital Rod Position Indication system agree writhi a U steps at taut once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intvials when the rod pGsitian de.taticn monitor is inoperable, then coare the Demand Position Indication Syst and the Diltai Rod Position tndication System at least once per 4 hourS.

SHEARON MAR15 - UNIT 1 3/4 1.-lI V Denotes a Critical Step

- 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 15 of 21 Form ES-C-i VERJFLCAT1QN QF QQMPLTION -

Job Performance Measure No.: 2011 NRC JPM SRO Al-i DETERMINE ROD MISALIGNMENT USING THERMOCOUPLES Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 16 of 21 Form ES-C-i

. - JPM CUE SHEET -

Name:

Date:

The plant was at 90 percent power, with a load decrease in progress, when the CRS noticed that the DRPI indication for rod H02 was reading 24 steps higher than the group demand. The Initial Conditions: load decrease has been stopped and AOP-001 entered. ALB 13/8-5, Computer Alarm Rod Dev/Seq NIS Power Range Tilts alarm is the only MCB alarm received. l&C investigated and found no obvious electrical problems.

The CRS has directed you to calculate the temperature difference between thermocouple(s) adjacent to the misaligned rod and the 1nf i ia fing cUe average of symmetric thermocouple(s), using Attachment 2 of AOP-001 and the provided T/C Core Maps.

2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 17 of 21 Form ES-C-i

. - JPMCUESKEET TIC CORE MAP (DATA SHEET 1) TRAIN B MSG #245 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 A I 1 A B I 632 I B C 636 C D 1636 650 I E I 644 50 640 6321 E F 50 644 I F G 644 632 H I 649 644 I 640 50 jj K 1 640 K L 644 640 644 50 I L M 1636 I N 640 N P 1 636 I P R

R 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 TRAIN A T/C CORE MAP MSG # 245 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 A 50 1 A B ]636 I B C I 632 C D I D E I 636 640 E F 644 648 644 F G I 652 640 H I 640 644 652 640 640 I H I____ 652 K I 640 644 50 I K L I M 1 640 644 I M N I 636 640 644 N P 1 640 I P R I 636 R 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 18 of 21 Form ES-C-i

. JPMCUESHEET MALFuNCTION OF ROD CONTROL AND INDICATION SYSTEM I

Attachment I Incicaion5 of Mi5aIijned Rod Sheet 1 of 1 The table belcw indicates the variation in plant pararneter which may be indicative of roc misalknrnent. This varialion refers to reative changes n irdication from a reference condltDn at which the suspect rods posLtion was krwn to e properly aluçneo. he roforonco *ae may bo takon from prior pcating recoidc r it may be upkitcd cach time the proper rbd positioning is verifiec by in-core rneasurament. Ir general, greater misahcnment will ause larger variatbns. Variations in NI charnel indication are also affected by the core location f the J3pact rod. For example, maIigned rod that ia closest to the t-44 deteclor shoJd indicate that N-4 flux rarameters are abrmia1 when compared with flux parameters of the other Power Range NI clannels. If the parameters bebw xIiibi1 nu db1[urrrtl v ia[bri. with art irtividul DRPI diffeilny irutri ib ruup step counter demand position by more than 12 steps, it is probaj1y a rod position indication prcblem.

PLA4T PARAMETER VALUE INtCATIVE O ROD MISALlG1MENT Quadrant Poer Tilt Ratio (QPTR) Greater than 1 A]2 Povei Rauye lit urrieritatiuri Gra[et than 2 % dilleien;e Leeri any Iwn chnnLc (RFFFF TO Aththm,t 4)

Delta Flux hid catots Gwater lIwi 2% di1fereiie Leteri any Iwo channels (RFFFR TO Attchmc,t 4)

Cure Outlet 1TheiiiiucuupIe Greater thari 10F difference betueert thermocouples adjacent 10 tte misaligned rod ard ti-e a/el-age of symmetric themxcoLples (PERFORM Atlachnent 2 Axial Flux Traces (in-core rncvate CON5ULT ReactDr Engineering delector) AND EVALUATE using in-core movable detecbrs per EST-922, Conirol Rod Position Deterniiraticn Va Ircore I ntrurncntation

--END OF ATTACHMENT 1--

AOP-OO1 I Rei. I Page 40 cf 47 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 19 of 21 Form ES-C-i

.JPMCUESHEET.

j 4ALFUNCTION OF ROD CONTROL AND INDICATION SYSTEM Attachment 2- Adjacent nd Symmetric Thermocoi.pte Locations Sheet I : 3 TI i[RMDCOUPLE LOCATIONS 1 2 3 4 5 6 T 8 9 10 ii 12 *13 14 15 I

I I I I I A a a a*

i i I a a a B a a C ***

ft F ft ft F 0 F ft F ft ft ft E ft r ft r F 1 ft F F B 1* r (3

r I N; I H ft ft F ft F 1 ft F j N; ft r. 1 F ft K ft F ft F ft RT R T N; L F F F M N.

N T ft F ft F ft T p r ft7 ft R r R Control Rod T Thnnomiip P Thermocouples abaidonad by [C 4199?

AOP -001 Re. 35 Page 41 of 47 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 20 of 21 Form ES-C-i JPMCUESEtEET I MAL.F1JNCflON OF ROD CONTROL AND INDICATION SYSTEM I Attachment 2 cijcEnt imd SrnmetrIc Thermocouple Lccarlons tie{2u13 NOTE

. iO. E0T, H08, K08. and P03 have no simetrc ocaticns.

e Symmetric :hetniocouples ac those in the same row.

SYMMETRIC LOCAIIONS 3R1D 1 III IV TRAIN A B A B A 8 A B A08 HIS C30I 315 RO?

S B05 E14 Y 0 DO MD NOfi 1i03 M C DOD Z12 NOL MOD.

M [04 DO [12 Mu L12 C T MI COt LOC HOS T I FOE Ff1 K1I* K05 LG6 R 0 FO3 Ff3 NW N06 KOD.

I M G06 FOS JIC C S 1-109 G02 JOD P07 M&9 J12 Thermocoup1e abanored by EC 4Th97

1. DETERMINE frerrnocupe locaiors) adjacent to the nIsagned rod using core grid map (Sheet 1), AND CIRCLE location(s) in TaIe above.

I Rev. I iage42ot4f 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 21 of 21 Form ES-C-i JPM CUE SHEET I MALFUNCTION OF ROD CONTROL AND INDICATION SYSTEM 1 Attachment 2 Adjacent and Symmetric Thermocouple Locations Sheet 3 of 3

2. RECORD the following in the table below:

Adjacent TC number Adjacent IC value using the RVLLS Console, ERRS, or OS.l-P1

Syirmetric TC numbers (not including adjacent TCs)

Symmetric TC values for all OPERABLE TCs using the RVLIS Console, ERRS, or 051-Pt

3. DETERMINE the average of symmetric thermocouples, for each adjacent thermocouple.

Adjacent TC Symmetric IC Symmetric IC Number Value Number Value Average

4. COMPARE each adjacent thermocouple value listed to its symmetric thermocouple average for indication of a misaligned rod. (REFER TO Attachment 1 )

--END OF A1TACHMENT 2--

AOP-OO1 Rev. 35 I Page 43 of 47 2011 NRC Admin Exam SRO Al-i Rev. Final

Appendix C Page 1 of 10 Form ES-C-i WORKSHEET INFORMATION Facility: Shearon Harris Task No.: 301 1 35H601 Task

Title:

Determine Subcoolinci with the JPM No.: 201 1 NRC Exam Subcooling Margin Monitor Admin JPM SRO A1-2 Unavailable K/A

Reference:

G 2.1.23 RO 4.3 SRO 4.4 Alternate path: No Examinee: NRC Examiner:

Facility Evaluator: Date:

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

Initial Conditions:

A LOCA has occurred

SI is actuated

PATH-i is in progress

Containment pressure is 3.3 psig

All RCPs are running

The Subcooling Margin Monitor is not available

The hottest CET is reading 565° F

ERFIS Pt. TRC9300 is not available

Loop Thots are reading as follows:

TI-413 = 559°F TI-423 = 558°F Tl-433 = 557°F

All Loop Tavg are reading 557°F

ERFIS Point PRC9455L has failed

Pressurizer Pressure Channels are reading as follows:

P1-457 = 1740 psig P1-456 = 1730 psig P1-455 = i735psig

RCS Wide Range Pressure Channels are reading as follows:

P1-402 = 1690 psig P1-403 = 1695 psig The CRS has directed you to determine and report the Subcooling Initiating Cue:

2011 NRC Admin Exam JPM SRO Al -2 Rev. Final

Appendix C Page 2 of 10 Form ES-C-i WORKSHEET INFQRMATIQN.

Task Standard: Correct indications utilized and Subcooling margin determined within specified range.

Required Materials:

Calculator

Steam Tables General

References:

EOP USERS GUIDE Rev. 30 Handouts:

EOP USERS GUIDE, Table of Contents

EOP USERS GUIDE, Section 6.2 (pgs. 33 and 34)

Time Critical Task: No Validation Time: 12 minutes Critical Task Justification Must choose correct temperature to use based on provided plant Step 2 conditions. IF the wrong temperature is selected all calculations will be incorrect.

IF Containment pressure is not identified as ADVERSE then all Step 3 calculations will be using the normal Containment values and the calculations will be incorrect.

IF the saturation temperature were incorrectly calculated the final St ep 4 calculation will be incorrect.

Final Subcooling calculation is what the candidates task was. If the Step 5 reported values are incorrect the CRS may make incorrect choices with procedure steps based on the value given.

2011 NRC Admin Exam JPM SRO A1-2 Rev. Final

Appendix C Page 3 of 10 Form ES-C-i PERFORMANCE INFORMAT1QN Start Time:

Performance Step: 1 Refers to EOP Users Guide.

Comment: The candidate is not required to complete the task in discrete steps. Guidance for performing the task is provided in the EOP USERs GUIDE but procedure steps are not specified. The specified parameters must be selected for the subcooling calculation to be correct.

J Performance Step: 2 Evaluate plant conditions:

RCS temperature Standard: Determines the hottest CET reading is the preference and chooses:

565°F Comment:

J Performance Step: 3 Evaluate plant conditions:

RCS Pressure Standard: Determines containment conditions are adverse:

PI-402=169OPSIG Comment:

2011 NRC Admin Exam JPM SRO A1-2 Rev. Final

Appendix C Page 4 of 10 Form ES-C-i

. PERFQRMANCE INFOftMAT1QN.

,J Performance Step: 4 Determine saturation temperature.

Standard: Convert to PSIA: 1690 + 15 = 1705 PSIA Interpolates between 1700 and 1740 PSIA and determines saturation temperature to be 61 3°F but 614°F:

[5/40(616.385 61 3.189)] = 0.4

61 3.189 + 0.4 = 61 3.6°F

6i3614 Comment: Candidate can use the Steam Tables.

q Performance Step: 5 Determine subcooling.

Standard: Calculates and reports subcooling to be 48°F but 49°F

61 3.5°F 565°F = 48.5°F Comment:

Terminating Cue: After the candidate has reported the calculated Subcooling margin: Evaluation on this JPM is complete.

Stop Time:

2011 NRC Admin Exam JPM SRO Al -2 Rev. Final

Appendix C Page 5 of 10 Form ES-C-i VREIQATlQr\1OF QQMPLETIQN Job Performance Measure No.: 2011 NRC Admin Exam JPM SRO Al -2 Determine Subcooling with the Subcooling Margin Monitor Unavailable Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2011 NRC Admin Exam JPM SRO Al -2 Rev. Final

Appendix C Page 6 of 10 Form ES-C-i JPM CUE S1IEET Name:

Date:

INITIAL CONDITIONS: e A Small Break LOCA has occurred

Slis actuated

PATH-i is in progress

Containment pressure is 3.3 PSIG

All RCPs are running

The Subcooling Margin Monitor is not available

The hottest CET is reading 565°F

ERFIS Pt. TRC9300 is not available

Loop Thots are reading as follows:

Tl-413 = 559°F Tl-423 = 558°F Tl-433 = 557° F

All Loop Tavg are reading 557°F.

ERFIS Point PRC9455L has failed

Pressurizer Pressure Channels are reading as follows:

P1-457 = 1740 psig P1-456 = 1730 psig P1-455 = 1735 psig

RCS Wide Range Pressure Channels are reading as follows:

P 1-402 = 1690 psig P 1-403 = 1695 psig The CRS has directed you to determine and report the Subcooling INITIATING CUE:

Subcooling margin = °F 2011 NRC Admin Exam JPM SRO A1-2 Rev. Final

Appendix C Page 7 of 10 Form ES-C-i JPM CUE SHEET Information Progress tnergy Use HARRIS NUCLEAR PLANT PLANT OPERATING MANUAL VOLUME 3 PART 4 PROCEDURE TYPE: EMERGENCY OPERATING PROCEDURE (EOP)

NUMBER:

EOP-USERS GUIDE TITLE: USERS GUIDE EOP-USERS GUIDE Rev. 30 I Page 1 of 2011 NRC Admin Exam JPM SRO Ai-2 Rev. Final

Appendix C Page 8 of 10 Form ES-C-i

. PMQUSHT GUIDE USERS Table of Contents Section 1.0 Purpose 4 2.0 References 5 3.0 Responsibillties 8 4.0 DeflnitionslAbbreviations 9 5.0 Procedure 10 5.1 Control Room Usage of the EOP Network 10 5.2 Control Room Usage of Status Trees 19 5.3 Control Room Usage of EPPs, Foldouts. and FRPs 24 6.0 General Information 33 6.1 Background and Basis for EOP Network 33 6.2 RCS Subcooling 33 6.3 Resetting SI, Phase A, Phase B and FW Isolation 35 6.4 Sequencer Operation 35 6.5 Stable Pressure/Temperature 35 6.6 Reahgnrnent of Plant Equipment Following SI Termination 36 6.7 Definitions of Verify, Check andActuate 36 6.8 Use of RCS Wide Range Pressure Indication 38 6.9 Dump Steam at Maximum Rate 38 6.10 Control of Ruptured Steam Generator Level 39 6.11 Use of-NORMAL 40 6.12 PATh GUIDES 41) 6.13 Use of Operating Procedures 41 6.14 Plant Instrumentation and Plant Computer 42 6.15 Use of PRZ Spray During RCS Depressuazation 43 6.16 Use of Charging to Control RCS Inventory 44 6.17 SI Reinitiation 45 618 TDAFW Pump Speed Controller Setpoint 45 6.19 Tech Spec cooldown and Pressure/Temperature limits .. 46 6.20 Determining RCS Temperature and Cooldown Rate 46 6.21 Using Core Exit TCs to determine Inadequate Core Cooling Conditions 47 6.22 Consulting the Plant Operations Staff 47 6.23 Shutdown Margin Evaluation 40 6.24 CSIP Alternate Miniflow Valves 48 6.25 Safeguards Verification During FRP Implementation .... 50 6.26 Restoration of ROP Cooling and RCP Restaxt 51) 6.27 Use of SBANK SELECT During an ATWS 52 6.28 RCP Restart Following a SGTR 52 6.29 Loss of SI Flow Indication 53 6.30 Contingencies for Termination of SI Flow 54 6.31 Torque Switch and Overload Protection 54 6.32 Conservation of the 250 Volt Station Battery 55 6.33 Potential Gas Intrusion Into the CSIP Suctions 57 6.34 TransitIon to Severe Accident Management Guidance -. . .58 6.35 Immediate Actions During an ATWS 59 EOP-USERS GUIDE I Rev. 30 I Page 2 of 69 2011 NRC Admin Exam JPM SRO A1-2 Rev. Final

Appendix C Page 9 of 10 Form ES-C-i

. . JFM CUE SHT GUIDE USERS 60 GENERAL INFORMATION 6.1 Background and Basis for EOP Network During the verification and validation of the EOP network, and during Operator simulator training, many questions are frequently asked concerning the EOP network. The background and basis For the SHNPP EOP network is found in detail in the background documents of the WOG ERGs and the EOP Step Deviation Documents.

The following sections contain clarification of certain procedural requirements that are frequently questioned by the operators.

6.2 RCS Subcooling The ERFIS plant computer functions as the plant subcooling monitor RCS subcooling will normally be obtained from the top level Safety Parameter Display System (SPDS) screen (Turn-on-code:

SPTOP, Parameter: SUBCOOL (DEGF), ERFIS point TRCS400). IF for some reason the subcoohng monitor is not available, the operators will manually determine subcooling using one of the following (Reference 2.2.2.2):

Graph provided on the CSFSTs

Subcooling Margin CaIc. Prograrir Version 1.0

Steam Tables Subcooling values are generally presented in the following format 1O F [4Q F] C-20 F [50 F] M The top set of values is normally used when the subcooling monitor is available (designated by C>. The bottom set of numbers is used only when the subcooling monitor is not available (designated by M for manual) The subcoollng values used in the procedure were determined based on specific instrument inaccuracies. Should it be necessary to manually determine subtooling, the following conventions apply:

1. Primary temperature is obtained using one of the Following based on availability of the indications (listed in order of preference):

Core exit TC reading on SPOS (Turn-on-code: SPTOP, Parametec T EXIT (DEGFy, ERFIS point TRCS300).

This reading is the average of the five hottest core exit TCs and is the input used for the Subcooling Monitor.

Highest core exit TC reading from the Inadequate Core Cooling Monitor (ICCM).6.2 RCS Subcooling (continued)

EOP-USERS GUIDE I Rev. 30 I Page 33 of 69 2011 NRC Admin Exam JPM SRO Al -2 Rev. Final

Appendix C Page 10 of 10 Form ES-C-i JPM CUE SHEET GUIDE USERS

Highest active loop wide range T-hot (11-413, 423, 433). An active loop is defined as one that has forced or natural circulation flow. If any RCPs are running, all loops will be active (backilow is available in loops where RCPs are not running). A classic example of a non-active loop would be a loop that has a SGTR since it is isolated and natural circulation flow in this loop would not be available.

2. Primary pressure is obtained using one of the following based on the range and availability of RCS and PRZ pressure indication:

If ERFIS is available, then use the RCS pressure reading on SPDS (Turn-on-code: SPTOP, Parameter:

PRZ PRES (PSIGy, ERFIS point PRC9455L). If PRZ pressure is above 1700 PSIG, this reading is the lowest of the three PRZ pressure channels (PT-457, PT-456, and PT-455). IF PRZ pressure is below 1700 PSIG, this reading is the lowest of the two RCS wide range pressure channels: PT-402 (ERFIS point PRCO4O2) and PT-403 (ERFIS po4nt PR00403).

It PRZ pressure is greater than 1700 PSIG and CNMT conditions are normal, then use the lowest PRZ pressure indication (P1-457, P1-456, or P1-4551).

If PRZ pressure is off scale low or adverse CNMT conditions exist, then use the lowest of the two RGS wide-range pressure indications P1-402.1 or P1-403.

OnFy PT-402 and PT-403 are used since these transmitters are located outside containment

When RCS pressure is less than 700 PSIG, Pl-4C12A should be used. Pl-402A receives input from qualified instrument PT-402 and its narrow range scale provides a more precise indication of pressure.

EOP-USERS GUIDE I Rev. 30 I Page 34 of 69 2011 NRC Admin Exam JPM SRO A1-2 Rev. Final

Appendix C Page 1 of 6 Form ES-C-i

- PERFORMANCE irtFORMATtON Facility: Shearon-Harris Task No.: 002001 H20i Task

Title:

Review (for approval) a completed JPM No.: 2011 NRC Admin surveillance procedure for PORV SRO JPM A2 block valves.

K/A

Reference:

G2.2.12 (3.4)

Examinee: NRC Examiner:

Facility Evaluator: Date:

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

Initial Conditions:

The unit is operating at 100% power.

PRZ PQRV PCV-445B has a failure in the SHUT circuit. RC-1 15 has been closed and power is removed.

TS 3.4.4 Action b is in effect. EIR 20110285 has been initiated.

The control room crew has completed OST-1017, PRESSURIZER PORV BLOCK VALVE FULL STROKE TEST QUARTERLY INTERVAL MODES 1-2-3-4.

Initiating Cue: You are the CRS. Review the completed OST for approval.

2011 NRC Admin Exam JPM SRO A2 Rev. Final

Appendix C Page 2 of 6 Form ES-C-i ERFORMANCE INFORMATION Task Standard: All errors identified.

Required Materials: None General

References:

OST-i 01 7, PRESSURIZER PORV BLOCK VALVE FULL STROKE TEST QUARTERLY INTERVAL MODES 1-2-3-4, Revision 21 Handout: Completed OST-1 017 with errors that align with the JPM content.

Time Critical Task: No Validation Time: 15 Minutes Critical Task Justification Identification of stopwatch beyond calibration date all timing data Item 1 collected with out of cal device is non usable therefore the test is invalid.

Missing initial and date for step marked N/A. lAW PRO-NGGC-0200 Section 9.2.26.a.2) If the section or step is N/Ad the respective work group supervisor shall initial/date and document justification unless this Item 2 is allowed by the implementing procedure. In the case of the step that was N/Ad it is not allowed by the procedure and therefore should have been initialed and dated.

Shut time for 1 RC-1 13 exceeding limit if not identified an inoperable Item 3 component could be relied upon and fail when needed to perform its intended action.

2011 NRC Admin Exam JPM SRO A2 Rev. Final

Appendix C Page 3 of 6 Form ES-C-i

- PERFORMANCE INFORMATION START TIME:

Performance Step: 1 Obtain procedure.

Standard: Reviews Sections 3.0, 4.0, 5.0, 6.0.

Evaluator Cue: Provide handout for AUDIT JPM SRO A2.

Evaluator Note:

The steps of reviewing the procedure can be completed in any order.

There are three errors in the procedure. Only the errors are documented in the JPM.

Comment:

Performance Step: 2 Review the completed OST-i 017.

V Standard: Identifies Stopwatch beyond calibration date per Prerequisite

3.0.4. Comment

Performance Step: 3 Review the completed OST-1017.

V Standard: OST-i 017 attachment 2, N/A for 1 RC-1 15 with comment 1 was not initialed and dated per PRO-NGGC-0200 Section 9.2.26.a.2).

Comment:

V - Denotes Critical Steps 2011 NRC Admin Exam JPM SRO A2 Rev. Final

Appendix C Page 4 of 6 Form ES-C-i ERFORMANGE INFORMATION Performance Step: 4 Review the completed OST-i 017.

V Standard: SHUT time for 1 RC-1 13 exceeds LIMITING VALUE.

Comment:

Terminating Cue:

STOP TIME:

V - Denotes Critical Steps 2011 NRC Admin Exam JPM SRO A2 Rev. Final

Appendix C Page 5 of 6 Form ES-C-i VERIFCATIONOP COMPLETION Job Performance Measure No.: 2011 NRC Admin SRO JPM A2 Review (for approval) a completed surveillance procedure for PORV block valves. OST-1017 Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2011 NRC Admin Exam JPM SRO A2 Rev. Final

Appendix C Page 6 of 6 Form ES-C-i

JPMCUESHET

The unit is operating at 100% power.

PRZ PORV PCV-445B has a failure in the SHUT circuit. RC 115 has been closed and power is removed.

Initial Conditions:

TS 3.4.4 Action b is in effect. EIR 20110285

The control room crew has completed OST-1017, PRESSURIZER PORV BLOCK VALVE FULL STROKE TEST QUARTERLY INTERVAL MODES 1-2-3-4.

Initiating Cue: You are the CRS. Review the completed OST for approval.

NAME:

DATE:

IF any discrepancies were identified in the review of OST-1017 list them on the lines below 2011 NRC Admin Exam JPM SRO A2 Rev. Final

PrQ.res. CONTINUOUS HARRIS NUCLEAR PLANT PLANT OPERATiNG MANUAL VOLUME 3 PART9 PROCEDURE TYPE: OPERATiON SURVEILLANCE TEST NUMBER: OST-1 017 PRESSURIZER PORV BLOCK VALVE FULL STROKE TEST QUARTERLY INTERVAL MODES 1-2-3-4 NOTE: This procedure has been screened per PLP-100 Criteria and determined to be CASE Iii, No additional management involvement is required.

OST-1017 Rev, 21 Page 1 of 13

1.0 PURPOSE This OST demonstrates the operabity of each PORV block valve by operating the valve through one complete cycle of 1u11 travel unless the block valve is closed with power removed in order to meet the requirements of action b. or c. in Tech Spec 3.4.4, per Tech Spec Surveillance 4.4,4.2. This OST also partially satisfies the inservice Testing Program requirements.

2.0 REFERENCES

2.1. Plant Operating Manual Procedures

1. OP1OO

2. ISI-801 2.2. Technical Specifications

1. 3.4.4, Relief \Jalves

2. 6.8.4.m (Inservice Testing Program)

3. 4.4.4.2, Relief Valves Surveillance Requirements 2.3. Final Safety Analysis Report 1, 5.4.13, Safety and Relief Valves 2.4. Drawings

1. 5-S-I 301, Reactor Coolant System Sheet 2 2.5. Others

1. HNP-IST-003, HNP 1ST Program Plan 3 interval rosT-1o17 Rev, 21 Page 2 of 13

3.0 PREREQUISITES I Verify that the Reactor Coolant System is aligned in a manner that will support the performance of this OST.

2. Coordinate the performance of this OST with other plant evolutions such that the minimum equipment operating requirements of Tech Specs are met, 4

3. Obtain any tools and equipment required per Section 5.0. V

4. Complete the Calibration Data Sheet and verify instrumentation is within calibration.

5. Verify instrumentation needed for the performance of this test is free of deficiencies that may affect instrument indication.

6. Verify all prerequisites are met, then obtain CR8 permission to perform this OST.

-1i (1 Signature Date 4M PRECAUTIONS AND LIMITATIONS 1 Test only one PORV block valve at a time.

2. If testing is suspended for any reason retUrn the block valve being tested to the as found position.

3. Do not test block valves which are closed with power removed in order to meet requirements of ACTION b. or C. in Tech Spec 34.4,

4. If any valve stroke time falls outside its Code Criteria, the valve will be immediately retested per the retest instructions or declared inoperable.

OST-1 017 Rev. 21 Page 3 of 131

5.0 TOOLS AND EQUIPMENT

1. Calibrated Stopwatch 6.0 ACCEPTANCE CRITERIA This OST will be completed satisfactorily when the following are verified;

1. Stroke times of PORV block valves tested are within the stated acceptance criteria as listed on Attachment 2, Valve Test Data.

2. Full open and full closed position indication observed by position indication lights is proper for each PORV block valve tested,

3. The PORV block valves complete a full cycle of travel.

CST-i017 -

Re21 - -

Page4of13

7.0 PROCEDURE 9 The foflowing two steps should he signed off at the completion of the test.

NO

1. If, during the performance of this test, a valve stroke time exceeds its Code Criteria, immediately retest the valve per Attachment 3. Otherwise \ Ji this step is N/A.

2. If, during the performance of this test, a valve exhibits abnormal or erratic action, document the condition in the comments section of Attachment 4, Otherwise this step is N/A. 3,49

3. Complete the As Found positions in Step 7.05.

(CAUIiON PORV block valves closed with power reroved in order to meet the requirements of ACTION b or c. of Tech Spec 3.4.4 are not required to be tested. The Comments Section of Attachment 2 must reference the applicable E1R number for PORV block valves not tested.

4. Refer to Attachment 2 and test all valves listed per the following:

Steps 70.4.a through 70.4.1 are for testing of all valves listed on Attachment 2, Initialing for Steps is done when all valves on Attachment 2 are tested.

7

a. Verify the valve to be tested is aligned to the Pretest Position and initial the space provided on Attachment 2. / /

b. Simultaneously start the stopwatch and place the control switch for the valve to be tested to the position opposite the pretest position.

c. Stop the stopwatch when the valve has completed its travel as indicated by a singular position indicating light for the demanded position (no dual indication), 1 j OST-1 017 Rev. 21 Page 5 of 13

7O PROCEDURE (continued)

d. Record valve stroke time in space provided on Attachment 2.

e. Place the control switch for the valve in test to the Posttest Position shown on Attachment 2 and start the stopwatch.

Stop the stopwatch when the valve has completed its travel as indicated by a singular position indicating light for the demanded position (no dual indication). VI

g. Record valve stroke time in space provided on Attachment 2.

V

h. Verify that the valve has traveled to its Posttest Position as indicated by a singular position indicating light for the demanded position (no dual indication) and initial in the space provided on Attachment 2.

Initial for the Full Stroke Test on Attachment 2 as verification of satisfactory valve operation (as previously performed per

/

Steps 7.O.4.b through 7.O.4.h above).

j. Repeat Steps 7.O4.a through 7.041 above to test all required remaining valves on Attachment 2.

k. Perform independent verification of valves as required per Attachment 2.

Review all data taken on Attachment 2 and ensure all stroke times are within the stated Acceptance Criteria. Inform the CRS of any out of tolerance reading and take appropriate action.

5. Restore components to their as found condition as follows:

As Found Restored Verified a.

RC-13 IL

b. IRC-115

c. IRC-117

6. Complete Attachment 4, Certification and Reviews and inform the CRS when this test is completed or found to be unsatisfactory.

)

[ST-1017 Rev. 21 Page6of13

8.0 DIAGRAMS/ATTACHMENTS Attachment I - Caiibration Data Attachment 2 Valve Test Data Attachment 3 Valve Retest Data Sheet Attachment 4 Certifications and Reviews OST-1 017 Rev. 21 I Page 7 oij

rTi ci)

C Cu 1

Cu C

0 Cu cJ 0

6 o 2

/)

0 Cu I

r

Attachment 2 - Valve Test Data Sheet 1 of 1 Au spaces next to valve number shall be filled in with initials, data or N/A as applicable.

PRETEST FAIL SAFE POSHEST FULL STROKE TEST TEST ALIGNMENT ACCEPTANCE CRITERIA (SEC)

ALIGNMENT Verification of CODE CRITERIA Travel by md LIMITING VALUE Lights Stroke Time (SEC) OPEN SHUT (INIT)

Valve Pretest md Fail Safe Position Posttest Low Low High OPEN SHUT Number Position Irt Stem Lights OPEN SHUT Position Verified Position Pos Init Verf mit lRc-113 opjm T N/A 13. 21/ N/A N/A OPEN i ]Th. 11.67 15.77 13A9 18.25 20.58 2380 N/A OPEN 12.00 16.22 156 18.34 21.16 23.92 1RC-115 OPEN N/A N/A N/A OPEtjj / Y 11.94 16.141 14.26 1928 21.06 25.15 N/A I N/A -

iRC..117jNJ Comments: if.T,< ,Z C/f C V / ,a.-

Attachment 3 Valve Retest Data Sheet Sheet I of 2 This entire Attachment is N/A if no valve is retested due to exceeding the Code Criteria.

Determine if the stroke time exceeds the Umiting Value.

1. If the stroke time exceeds the Limiting Value, declare the valve inoperable and initiate an AR. (N/A if stroke time is less than the Limiting Value)

2. if the stroke time is less than the Limiting Value, but outside the Code Criteria limits, perform the following Steps:

a. If the cause is known to be mechanical failure, or if a retest cannot be performed expeditiously, declare the valve inoperable and initiate an AR (except for PMTRs).

b. if retesting the valve is desired, perform the following:

NOTE: If necessary, separate marked up sheets of this OST may be used to document necessary manipulations. These sheets would be attached to this procedure and noted in the comments Section of Attachment 4.

(Certifications and Reviews)

(I) Determine which Steps need to be performed to set up conditions for testing the valve. CR5 concurrence must be obtained and documented in the Comments section of Attachment 4. (Certifications and Reviews)

(2) Perform the Steps determined in the previous Step and document stroke times/valve positioning on Sheet 2.

(3) If retest results are still outside the Code Criteria, declare the valve inoperable and initiate an AR (except for PMTRs).

(4) If retest results are within the Code Criteria, perform the following:

(a) Declare the valve operabLe.

(b) Initiate a CR identifying test findings for the first and second tests.

(c) Send test results to Responsible Engineer (1ST) for evaluation and documentation on the AR.

I, OSI-1017 Rev. 21 Page 13 of 13j

c..)

a 4.,

0 0

(I) 4-a U) 04-L 1*

C U) 0

Attachment 4 Certifications and Reviews Sheet 1 of I This OST was performed as a: Periodic Surv&Uance Requirement:

Postmaintenance Operability Test:

Redundant Subsystem Test:

Plant Conditions: / )Cc) ? Mode: /

OST Completed By: -j_ *Q; Date:

Time:

OST Performed By:

Initials Name (Print) Initials Name (Print)

)fj (A.) JLtj. .Q((.,

ljt.; j/)/1 General Comme nts/Reco mme ndationfCo rrective Actions/Exceptions:

/ c *// 1 c /e. ,.i f /?u.- veM cve c /o 41 3 Pages Used: ,9 t /_

OST Completed with NO EXCEPTIONS/EXCEPTIONS:

Date:

CRS Reviewed By: Date:

Responsible Engineer (TST)

Reviewed By: Date:

AN I I After receiving the final review signature, this OST becomes a QA RECORD and should be submitted to Records Management.

OST-1017 Rev. 21 I Page i2ofJ

Revision Summary General (Revision 9 - Editorial Correction)

This revision replaces the reference to Tech Spec SurveWance Requirement 4.0.5 with the Inservice Testing Program. Tech Spec Amendment 127 deleted this surveillance requirement and moved the Inservice Testing Program to Tech Spec 6.84,m. (PRR 288110)

Descriøtion of Changes Section Channe Desc.tpion 2 1.0 Changed reference to Tech Spec Surveillance Requirement 4.0,5 to the 2.2.2 Inservice Testing Program.

General Rev 20 Editorial correction to deLete all references to system description.

Description of Changes Section Change Description All Updated revision level.

2 Step 2.5.2 Deleted Step 2.5,2 reference SD-100.03. System descriptions are no longer active documents.

Attachment 1-4 Formatted Attachments headings to new standards.

General Rev 21 Editorial correction to correct reference and make title changes.

Description of Changes Section Change Description 3 Step 3.0.6 Changed Unit SCO to CRS lAW OPS-NGGC-1000.

5 Step 7.0.3 Corrected Step reference from 7.0.0.5 to 7.0.5.

6 Step 7.0.4.1 & Changed Unit SCO to CRS lAW OPS-NGGC-1 000.

Step 7.0.6 10 Att 3 Shi Changed Unit SCO to CRS lAW OPS-NGGC-1000.

Step 2.b.(1) 12 Att 4 Sh 1 Changed Unit SCO to CRS lAW OPS-NGGC-1000. Also changed Document Services to Records Management.

OST-1 017 Rev. 21 Page 13 of 13

Appendix C Job Performance Measure Form ES-C-i WQrksheet -

Facility: Shearon Harris Task No.: 345001 H602 Task

Title:

Classify an event and complete JPM No.: 2011 NRC Exam Emergency Notification Form Admin JPM SRO A4 K/A

Reference:

G2.4.41 RO 2.9 SRO 4.6 Examinee: NRC Examiner Facility Evaluator: Date:

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

This is a two (2) part TIME CRITICAL JPM where each part will be timed separately.

Initial Conditions: A Reactor trip and Safety Injection have just occurred from 100% power due to low steam line pressure.

WEB EOC is not available.

Part 1: Using the attached information sheet and the EAL Matrix, classify the event. Mark up the EAL Matrix appropriately.

Initiating Cue:

Part 2: Based on your EAL classification, your next task will be filling out an ENF form.

2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Job Performance Measure Form ES-C-i

.. Workshaet Task Standard: Event classified as an Alert (HA2.1). ENF correctly filled out.

Required Materials: None General

References:

PEP-i 10 EAL Matrix; PEP-31 0; PEP 230 EP-EAL (allowed reference)

Handouts: JPM Cue Sheets Pages 7 and 8 PEP-lb EAL Matrix EP-EAL; PEP-230; PEP-310 (allowed references)

Time Critical Task: Yes Validation Time: 15 minutes for classification and 15 minutes to complete ENF Critical Step Justification Classification of the event is critical for determining State and County Step 2 notifications, public information notices, site information notices, and event reportability to the Nuclear Regulatory Commission.

Step 4 Attachment 9 has been completed correctly.

Form must be filled out correctly. The critical lines are identified on Page 8 of this JPM. Each line must be filled out correctly and are critical ENF Form based on HNP EP and NEI 99-02 Rev. 6 Standards.

2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Page 3 of 13 Form ES-C-i PRFQRMANCE INFORMATION Evaluator Cue: Start Time for this portion of JPM begins when the individual has been briefed START TIME:

Performance Step: 1 OBTAIN PEP-3i0 attachment 9, Nuclear Power Plant ENF.

Standard: Obtains attachment 9 of PEP-310. (EP-EAL is an allowed reference)

Evaluator Cue: WEB EOC is not available. Manual method must be used.

V Denotes a Critical Step

- 2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Page 4 of 13 Form ES-C-i PERFORMANCE INFQRMATIQN v Performance Step: 2 CLASSIFY EVENT Standard: Completed attachment 9 and identify alert due to Explosion in Table H.1 area damaging Safe S/D equipment (HA2.1)

Initiating Condition: Fire or explosion affecting the operability of plant safety systems required to establish or maintain safe shutdown HA2.1 Alert Fire or explosion resulting in EITHER:

Visible damage (Note 4) to any Table H-i structure OR system/component required for safe shutdown of the plant OR Control Room indication of degraded performance of any safe shutdown structure, system, or component within any Table H-i area STOP TIME:

Terminating Cue: Event classification sheet provided to evaluator.

Examiners NOTE: After the candidate returns this JPM classification page start the time for completing the ENF form.

V - Denotes a Critical Step 2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Page 5 of 13 Form ES-C-i P ERFQRMAF4Q INFQRMATIQN I Emergency Action Levels I Attachment I Emergency Action Level Technical Bases CatGery: H Hazards Subeategory: 2 Fire or Explosion Initiating Condition: Fire or explosion afrecting the operability of plant safety systems required to establish or maintain safe shutdown EAL:

HA2.1 Alert Fire or explosion resulting in EITHER:

Visible damage (Note 4) to ny Table F-I-i structure OR system/component required for safe shutdown of the plant OR Control Room indication of degraded performance of any safe shutdown structure, system, oi component within any Table H-i area Note 4: Visib D aage is:

Dmg tc rnnt r atructur that s rathy Qb rvabl witho ni mrt5, ta4ing, c alyi Damage s sufficient to cauae con cern regarin the ccninu ed operabt ity or relibili+/-y of affected s.sfety tructr, aystm, Qrmpnnt Example danage inciudes: deformation due o heat or impact. denting, penetradcn, wpture, crackng, pnt I tnrg Visbie Damage does not ndude svrfac demhs (e.g., partchpping, scratches)

Table H-I Structures Containing Safe Shutdown Equipment

Containment

Reactor Auxiliary Building

Fuel Haroing Buiding

Waste rocessng LiIding

Turbine ui1ding (ircludir Transmir Araa)

Emergency Diese4 Generator uding

Diesel Fel QI Storage Budirrg (DFOST)

ESW lntake Structure

- Auxiliary Reseneoir Inrake Strschre

- NSW Struotu re Sjitchyard Yard 21 Duct Sanks (underground raceways onainng Saf Situt.dorn pcwr, control at-rd insfrumeM cblgs) rJiflg any of the thoie areas Mode Applicability:

All EP-EAL Rev. S Page 158 of 296 V Denotes a Critical Step

- 2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Page 6 of 13 Form ES-C-i JRFQRMANQ INFQRMATION Emergency Action Levels Attachment I Emergency Action Level Technical Bases Basis:

Generic VISIBLE DAMAGE is used to identify the magnitude of the FIRE or EXPLOSION and to discriminate against minor FIRES and EXPLOSIONS.

The reference to structures containing safety systems or components is included to discriminate against FIRES or EXPLOSIONS in areas having a low probability of affecting safe operation. The significance here is not that a safety system was degraded but the fact that the FIRE or EXPLOSION was large enough to cause damage to these systems.

The use of VISIBLE DAMAGE should not be interpreted as mandating a lengthy damage assessment prior to classification. The declaration of an Alert and the activation of the Technical Support Center will provide the SEC with the resources needed to perform detailed damage assessments.

The SEC also needs to consider any security aspects of the EXPLOSION.

Escalation of this emergency classication level, if appropriate, will be based on EALs in category S, category F or Category R.

Plant-Specttlc The Table H-I vital area structures were obtained from the HNP safe shutdown analysis (ref. I).

A steam line break or steam explosion that damages permanent structures or equipment would be classified under this EAL The method of damage is not as important as the degradation of plant structures or equipment. The need to classify the steam iine break itself is considered in fission product barrier degradation monitoring (EAL Category F).

HNP Basis Reference(s):

I, AOP-036, Safe Shutdown Following A Fire

2. FSAR 7.4 I EP-EAL Rev. 9 Page 159 of I V - Denotes a Critical Step 2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Page 7 of 13 Form ES-C-i PERFORMANCE INFQRMATION Start Time for this portion of JPM begins when the Evaluator Cue: individual has completed the EAL classification and is ready to begin inputting data in the ENF notification form.

START TIME:

Performance Step: 3 Obtained PEP-310 attachment 9, Nuclear Power Plant ENF Standard: Obtained PEP-310 attachment 9.

Comments:

v Performance Step: 4 Completes PEP-310 attachment 9 (see key for critical steps)

Standard: PEP-3i0 attachment 9 filled out correctly Comments:

STOP TIME:

Terminating Cue: ENF form completed and provided to evaluator.

V - Denotes a Critical Step 2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Page 8 of 13 Form ES-C-i

. PRFORMANC INFORMATION Justification for critical steps of the ENF notification form can be found in the LOCT Performance Evaluation grading guidance. This guidance has been approved by EP and the latest revision is dated 7/4/10.

The standards for timely and accurate Emergency Action level (EAL) Classification development, Protective Action Recommendations (PARs) development and EAL and PARs Notifications are set forth in NEI 99-02 (Revision 6), Regulatory Assessment Performance Indicator Guideline.

NOTE: The following (line numbers) are applicable to the Emergency Notification Form (ENF)

(For the 2011 NRC Exam Admin JPM SRO A4 the following ENF lines must be correctly filled out and are therefore critical)

Initial Emergency Notification Form (ENF) completed apropriate to the event to include:

(Line 1) Whether the event is a drill or actual event

(Line 4) Class of emerency

(Line 4) EAL number

(Line 4) Description of emergency (EAL description)

(Line 5) Whether offsite protective measures (PARs) are necessary

(Line 5) Potentially affected population and areaas (Correct EPZ subzone information)

(Line 6) Whether a release is taking place

(Line 9) Wind direction and speed

(Line 10) Date and time of declartion of emergency

(Line 11 or Line 3) Plant and/or unit as applicable V - Denotes a Critical Step 2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Page 9 of 13 Form ES-C-i PERFORMANCE INFORMATION ENF FORM KEY Ahmait 9 NUCLFAR POWER PLANT EMFRGECY NOTTFiCTL(1N FQRLt Swt I I I 1tLL VZTUPA. £.vNr 5A01 4

2 MTR1 flDLLOWUP OTATON TrEICAT 3 H.t4

4. M1BECY JNUSUAL EVENT ALEFT SITE AREA EMEROENOY GE ERAL EMERGENCY cLS5Fc411cN; BASED Of GAL SAL GESCERTION: ALE $Lia

. 24 1

1 t / L; I

5. PROTECTUE ACTION RECOMMENDATtONS:

EVAGLRTE Caq mt w tEl rji.

OTHEP B. EMERCECY RELEASE: f4 Us OwnU 7 iELESE GNCNGB Ll pkp I nnr Jrdr B EVENT P0GNOS:S. plr B METEORCLOGCAL GATA Wn5 Im ses Vc1 Sp PrsF,Salivn IC ud ft DECiPRAT1ON

&tkri)

TERMWT1ON SSbllity CNv J tii lIme ,j4aW_

51 AVFEGTETJ UMT(S) 1U 12 UNiT STATUS J1 Pmvr Stutewn N n J(X)(,_ IsIs CJPi ilN 4i 1 12 lYmsr Shjte, sI 1Bp ,,,,.,, Dsl

.13 PMr 5 t

4 thn Si .1sls .?

13 REMAR5 NotlfJatonsj EEENCY RELEASE BATA. NOT RUINED LUE A IS SELECTED, 1i RCLSECNARAsDTER1!AS1ON. TYPC IE54*B Gaurcj UNITS Ci AGNTUDE: Nobi Ge.:_______ Othr FORM SUit BnE Date Tii D5 I_

1uqjte SM1T,rc ,., Cte Da:

15. PROJELrON FS EYti R zis, D[1cc IS, PROJECTED DOSE: STAC EDE (rrreml Aei.1t Tvtmd COG Irpfl 1? APPROSED BY:

NO ii FlED flr TIETCEIVED TIle .Jft&&.:, S ,_

BY, ,,. __________

Rev, 26 V Denotes a Critical Step

- 2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Page 10 of 13 Form ES-C-i YERIFICATIQN OF COMPLETION Job Performance Measure No.: 2011 NRC Exam Admin JPM SRO A4 Classify an event Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Page 11 of 13 Form ES-C-i JPM CUE SHEET This is a two (2) part TIME CRITICAL JPM where each part will be timed separately.

Initial Conditions: A Reactor trip and Safety Injection have just occurred from 100% power due to low steamline pressure.

WEB EOC is not available.

Part 1: Using the attached information sheet and the EAL Matrix, classify the event. Mark up the EAL Matrix appropriately.

Initiating Cue:

Part 2: Based on your EAL classification, your next task will be filling out an ENF form.

2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Page 12 of 13 Form ES-C-i JPM CUE SHT CLASSIFICATION INFORMATION SHEET A Reactor trip and Safety Injection have occurred due to low steamline pressure B Train has lost off-site power On start, the 1 B-SB EDG experienced a crankcase explosion resulting in Diesel Engine damage and damage in the engine room 1 B-SB bus remains de-energized Steam Generator Pressures are:

A SG 900 psi 9

B SG 350 psig and lowering

C SG 900 psig The Turbine Driven AFW pump has started and tripped on overspeed Containment pressure has reached ii PSIG and is slowly increasing The A-SA Containment Spray Pump started and tripped on overcurrent All radiation monitor readings are normal Wind direction is from 1550 with a speed of 9 mph CSFSTs:

Heat Sink is YELLOW

Inventory is YELLOW

All other CSFSTs are green Classify the event Name: Date:

Record your classification here and return to the examiner.

EAL Classification:

2011 NRC Exam Admin JPM SRO A4 Rev. Final

Appendix C Page 13 of 13 Form ES-C-i JPM CUE SHEET Name: Date:

Attchme.it 9- NUCLEAR POWER PLANL EMERGENCY NOTifICATION FORM Sh.et] f I 1 R L ACTUAL EVENT MSAG ______

2 4 AL FCLLOW LiP NOTC!.ON: TM_________ JATEéI UT-INTlCATC* #

!TF .nntir

4. EMERGENCV NUUAL j4T RGMDY CLSSPICAT1ON:.

EASED ON EAL ___________ EAL DECRP ON.

i PDTETE ACTtOW RECOMMENDAT1O4&

lwDuA-r SL Csa LE CF :i (PCTP5UI1 CCI 4 c1A.E M- TATa PI.t AC lOTHER

8. EtERGENCY RELEASE: None O&rdng Rs Ocurre

7. EEE 3 3NFCAtICE. NJ. jpwL riial eraIr rnrra oeatn

9. VNT CGNO1R to J rzd MFTFPDI fl(Al fl4T Wind flrini rnm Alirti rNot Rquad r liltial UotIIos

10. DECLAATI0N TERMNAION Precpiai The S4abty Ca De_ I fJ J 1L AFFECTEr N1T(S:

11. UNT TATUG:

]J ] J Ur St1u5 Nc4 Rur?i ilIIa El rowr c U i___ Py.r hiI1rThz fl

-.__% Poar Da: _j__

FOLLOW UP INFORMA1ION (Jns 14 throuih 16 Not R9guird for hitiI Notifisation)

ELER3ENCY RELEASE DArA. ?401 REQUIRED IF LIHE GA IS SELECTED

4. RELEASE CHRAJTERlZATOF: TYPE4 Eead Mced und *NTS: Ci Csec[ i.tDi/sec MM3NflUE: NoUe Gs odnes: ricues:

Fc1:fArbonie Start Thr,e Date _i_J_Sop ime Dnte Lqd Start Th-ie [te _? J__Sop Tme De _J__J

5. PRQiEC1ONPARAMETE2S: jecthn pericd: -Icurs tratec R!ate Durn Prcir peiforr9e Tirte Date

6. PROJECTED JOSE: DISTANCE TEDEir,rnI ,dit Thwoid DDE iinrrn Sebuiday 2 Hes 5 iIes 10 Ves

17. APPROVED 3Y Inc RECEIVED PE?-310 Rev.25 2011 NRC Exam Admin JPM SRO A4 Rev. Final

@@ Line 483: / Line 483: @@
 : 1. This test should not be performed when a motor driven AFW pump is being used for SG level control, due to the MDAFW FCV auto open feature.
 : 2. Only one Auxiliary Feedwater Pump shall be tested at a time.
-: 3. Auxiliary Feedwater Pump 1 B-SB should not be operated at less than 50 gpm and should not be operated at 50 gpm for more than [[estimated NRC review hours::300 hours]].
+: 3. Auxiliary Feedwater Pump 1 B-SB should not be operated at less than 50 gpm and should not be operated at 50 gpm for more than 300 hours.
 (This is normally provided by the recirc line) (Reference 2.7.1)
 : 4. If an accident initiation signal is received during the performance of this test, stop the test.
@@ Line 556: / Line 556: @@
 R      13.      At the MCB, START Auxiliary Feedwater Pump lB-SB. (Reference 2.3.1)
 Time started:     fc) c)
-CAU      N Auxiliary Feedwater Pump I B-SB shouJ                 operated at less than 50 gpm and should not be operated at 50 gpm for more than [[estimated NRC review hours::300 hours]].
+CAU      N Auxiliary Feedwater Pump I B-SB shouJ                 operated at less than 50 gpm and should not be operated at 50 gpm for more than 300 hours.
 ,     VERIFY that the recirculation flow on local indicator Fl-2172S is greater thanorequalto5Ogpm.
 : 15. VERIFY the following FCVs indicate open and initial for the PRETEST ALIGNMENT on Attachment 4.
@@ Line 906: / Line 906: @@
 Appendix C                                 Page 4 of ii                        Form ES-C-i PERFORMANCE INFORMATIOI\1 v   Performance Step: 3    Determine stay time for each operator (based on 1 St Operator reaching 2 Rem and the 2nd Operator reaching 4 Rem for the year)
-Standard:              Operator 1: [[estimated NRC review hours::2 hours]] 250 mrem &#xf7; 125 mrem/hr    = 2 hrs Operator 2: [[estimated NRC review hours::6 hours]] 750 mrem &#xf7; 125 mrem/hr    = [[estimated NRC review hours::6 hours]] Comment:
+Standard:              Operator 1: 2 hours 250 mrem &#xf7; 125 mrem/hr    = 2 hrs Operator 2: 6 hours 750 mrem &#xf7; 125 mrem/hr    = 6 hours Comment:
 Terminating Cue:       After the stay time has been calculated, this JPM is complete.
 END OF JPM STOP TIME:
@@ Line 1,133: / Line 1,133: @@
 Appendix C                                          Page ii of 21                              Form ES-C-i PEREQRMANCE INFORMATION KEY FM 3RL 1.. 3. 1OVAi3LE CORJL S$E*L GROUP HGHT LlMfTflG CONDITFOU FOR OP1RATlCl
 : 3. 1 .3 1 All shutdcn arid control rods shall be OPERABLE ad positioned wi thin L2 steps (in cated cioslticn) f their group step courter deniar4 postiri, APPLICABILOTY:       92OES 1 arid 2,
-: a. with one or iiore rods inoperable due to being irrtinovable as a result of excessive friction or iechanical interference or knctn to be uritrippable, determine t[mat the 5HUTUOUN M.G1N reulrenient of Specification 31.11 ts satisfied within 1 hour amd be in NOT 5TANGt within [[estimated NRC review hours::6 hours]].
+: a. with one or iiore rods inoperable due to being irrtinovable as a result of excessive friction or iechanical interference or knctn to be uritrippable, determine t[mat the 5HUTUOUN M.G1N reulrenient of Specification 31.11 ts satisfied within 1 hour amd be in NOT 5TANGt within 6 hours.
 : b. With noro than one rod miasaligned frccn the group stOp Counter deiriand position by riore than .t 12 steps (indicated position), be in HOT STANOY within S hours.
-: c. with more trian one rod Inoperable, due to a rod control urgent faiure alarn or obvious electrical probien ifl the rod coitrol systE1 existing for greater than [[estimated NRC review hours::36 hours]], be in NOT STANOBY within the follcing [[estimated NRC review hours::6 hours]].
+: c. with more trian one rod Inoperable, due to a rod control urgent faiure alarn or obvious electrical probien ifl the rod coitrol systE1 existing for greater than 36 hours, be in NOT STANOBY within the follcing 6 hours.
 : d. with one rod trippable but inoperable due to causes other than addressed by ACIION a... above. or misaligned from its group step counter denand height by irre than +/- 12 steps (indicated position), POSER OPERATION nay continue provided that within
 : 1. hour I       The rod is restored to OPF3M%LF status within the above alignment requirenents, or
@@ Line 1,148: / Line 1,148: @@
 b)  The SF4UTIJU tARIN renuirerient of pecticotion 3.1. 1,1 s determined at least mice per 12 tours:
 : 0)  A pcer distribution riiap s obtarrmeb froom the roabie mcore detectors and 0 F
-( Z) and    F   are  icr ified to be within their !iniits within [[estimated NRC review hours::2 hours]]: and dl The THFRt1 P0FR lcel s reduced to lcs. than or equal to 75 of RATIO FHERWmL (NER vdthin the next hour and vehin the follow ng 4 hcurs the Fip1 Neutron Flux Trip Sctpofnt s reduced to less than or equal to 5t of RkIED         rNERtAL, PQWR.
+( Z) and    F   are  icr ified to be within their !iniits within 2 hours: and dl The THFRt1 P0FR lcel s reduced to lcs. than or equal to 75 of RATIO FHERWmL (NER vdthin the next hour and vehin the follow ng 4 hcurs the Fip1 Neutron Flux Trip Sctpofnt s reduced to less than or equal to 5t of RkIED         rNERtAL, PQWR.
-FLCANCE RElRE1EN1S 4.1.3. 1.1 Th rc.iL:on or edch rod snal I c detrms;nd t e tvithin :he nroup dorrnd l:nit by verifying the individual rod postamis at least once per 12 hourS exct during te nteraIs             ri the rod posttioi d1atiri mri tar is inoperable, then verify the group positions at least once per [[estimated NRC review hours::4 hours]].
+FLCANCE RElRE1EN1S 4.1.3. 1.1 Th rc.iL:on or edch rod snal I c detrms;nd t e tvithin :he nroup dorrnd l:nit by verifying the individual rod postamis at least once per 12 hourS exct during te nteraIs             ri the rod posttioi d1atiri mri tar is inoperable, then verify the group positions at least once per 4 hours.
 .1.OJ,2 Each rod rot fully inserted n the core shall be deterniined to e OP[RBLE by bOvOt Of at 1m?St 10 steps 10 any one direCtiOrT 3t least On per 92 days.
 H.4h0                                                     irmrdmcnt  &#xfc;. 93 V Denotes a Critical Step
@@ Line 1,165: / Line 1,165: @@
 a    With a eaziia: of on, digital rod position indicator per bank inoperable
                       *1 ther
-: 1. Determine the position of the noniridicating rod(s) lnolractly by the savable Incoro detectors at least once per [[estimated NRC review hours::8 hours]] and liemedi.ately after any notion of the nonindicating rd which exceeds Z4 steps in one drect1on since th* lest eterinatIon f the rodLs position, or
+: 1. Determine the position of the noniridicating rod(s) lnolractly by the savable Incoro detectors at least once per 8 hours and liemedi.ately after any notion of the nonindicating rd which exceeds Z4 steps in one drect1on since th* lest eterinatIon f the rodLs position, or
-: 2. Raduci TNERMAI. POWER to less than 5 of RATED ThERMAL POWER witMn [[estimated NRC review hours::8 hours]]..
+: 2. Raduci TNERMAI. POWER to less than 5 of RATED ThERMAL POWER witMn 8 hours..
 : b. With a *axizs of one demand position Indicator per bank Inoperable aitheri
-: 1. Verify that all digital rod position indicators far the affected bank are QPERA8l.La that the sast wi thdrr.m rod and the least withdrmms rod of the bank as-withln a ziz.e of U steps of each other at l.ast once per 8          or Z. Reduce ThERMAl. POWER to iss than 5 of RATED ThERMAL POWER within [[estimated NRC review hours::8 hours]].
+: 1. Verify that all digital rod position indicators far the affected bank are QPERA8l.La that the sast wi thdrr.m rod and the least withdrmms rod of the bank as-withln a ziz.e of U steps of each other at l.ast once per 8          or Z. Reduce ThERMAl. POWER to iss than 5 of RATED ThERMAL POWER within 8 hours.
-SURVEILLANCE REGUIREMENTS 4.]L2 Each di4ltal rod poiti&#xf8; Indicator shall be determined to be OPERASLI by verifying that the Dand Position Indication System and the Digital Rod Position Indication system agree writhi a U steps at taut once per [[estimated NRC review hours::12 hours]] except during time intvials when the rod pGsitian de.taticn monitor is inoperable, then coare the Demand Position Indication Syst and the Diltai Rod Position tndication System at least once per 4 hourS.
+SURVEILLANCE REGUIREMENTS 4.]L2 Each di4ltal rod poiti&#xf8; Indicator shall be determined to be OPERASLI by verifying that the Dand Position Indication System and the Digital Rod Position Indication system agree writhi a U steps at taut once per 12 hours except during time intvials when the rod pGsitian de.taticn monitor is inoperable, then coare the Demand Position Indication Syst and the Diltai Rod Position tndication System at least once per 4 hourS.
 SHEARON MAR15   -  UNIT 1             3/4 1.-lI V Denotes a Critical Step
    -                                                   2011 NRC Admin Exam SRO Al-i Rev. Final

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