ML112790100

Initial Exam 2011-302 Draft Administrative JPMs

ML112790100
Person / Time
Site:	Browns Ferry
Issue date:	10/04/2011
From:	NRC/RGN-II
To:	Tennessee Valley Authority
References
50-259/11-302, 50-260/11-302, 50-296/11-302
Download: ML112790100 (82)

Text

Adniin RO Ala PAGE 1

OF 4

OPERATOR:

RO___

SRO___

DATE:________

JPM NUMBER:

Admin RO Ala TASK NUMBER:

U-000-AD-17 TASK TITLE:

Determine Adequate Performance of License Reactivation K/A NUMBER: 2.1.4 K/A RATING: RO 3.3 TASK STANDARD: Determine which of the reactivating personnel have correctly completed the reactivation requirements.

LOCATION OF PERFORMANCE:

Class Room REFERENCES/PROCEDURES NEEDED: OPDP-l0 VALDATION TIME: 10 minutes MAX. TIME ALLOWED: (Completed for Time Critical JPMs only)

PERFORMANCE TIME:

COMMENTS:

Additional comment sheets attached? YES NO RESULTS:

SATISFACTORY___

UNSATISFACTORY SIGNATURE:__________________

DATE:

EXAMINER

INITIAL CONDITIONS:

3 off-shift licensed personnel are returning to shift from rotating assignments and are reactivating their licenses. The following table gives information as to hours worked under direction of an activated licensee, tours performed, etc.

Pre-activation License Meeting Shift 1 Shift 2 Shift 3 Shift 4 Shift 5 Shift 6 Plant Tour Ops Training 12 hrs 12 hrs 12 hrs 12 hrs 12 hrs 12 hrs Per SM Manager U-3 RO U-3 RO U-i RO tagging U-2 RO U-2 RO instructions ROl Shift Manager UO RO conducted complete plant tour with STA (SRO)

Ops Training 12 hrs 12 hrs 12 hrs 12 hrs 1 2hrs Per SM Manager U-2 RO U-2 RO U-I RO U-3 RO U-3 RO instructions Ops Called conducted Superintendent for complete plant Random tour with extra R02 Drug test RO assigned to during that shift crew shift

missed end of shift turnover Ops Training 12 hrs 12 hrs 12 hrs 12 hrs 12 hrs Conducted Manager U-2 RO U-2 RO U-3 RO U-3 RO U-I RO complete plant R03 Ops tour with U-2 Superintendent RB AUO INITIATING CUE:

The Shift Manager has tasked you to determine which of the personnel, if any, have completed the requirements for license reactivation. If any personnel do not meet the requirements for license reactivation state the reason(s) why.

Admin RO Ala PAGE 3

OF 4

Class Room INITIAL CONDITIONS:

3 licensed personnel are returning to shift from rotating assignments and are reactivating their licenses. The following table gives information as to hours worked under direction of an activated licensee, tours performed, etc.

Pre activation Shift Shift Shift Shift Performance License Meeting 1

Shift 2 3

Shift 4 5

6 Plant Tour Step Ops Training 12 his 12 his 12 his 12 his 12 his 12 his Per SM instructions Does Not ROl Manager U-3 U-3 RU U-I tagging U-2 U-2 conducted complete Meet Shift Manager RU RU UU RU RU RU plant tour with STA (SRO)

Ups Training 12 his 12 his 12 his 12 hi-s l2his Per SM instructions Meets Manager U-2 U-2 RU U-I U-3 RU U-3 conducted complete requirements Ups RU Called for RU RU plant tour with extra R02 Superintendent Random Drug RU assigned to that test during shift shift crew missed end of shift turnover Ups Training 12 his 12 hrs 12 his 12 his 12 his Conducted complete Does Not Manager U-2 U-2 RU U-3 U-3 RU U-I plant tour with U-2 RB Meet R03 Ups RU RU RU AUU Superintendent IMTIATING CUE:

The Shift Manager has tasked you to determine which of the personnel, if any, have completed the requirements for license reactivation. If any personnel do not meet the requirements for license reactivation state the reason(s) why.

Adrn+/-n RO Ala PAGE 4

OF 4

START TIME____

Performance Step 1:

Critical X Not Critical Analyzes information provided to determine which personnel meet the requirements for license reactivation.

Standard:

Determines RO 2 meets the requirements for license reactivation.

SAT UNSAT N/A COMMENTS:

Performance Step 2:

Critical X Not Critical States the reason why RO 1 does not meet the requirements for license reactivation.

Standard:

ROl did not interview with Ops Superintendent contrary to OPDP-1O.

SAT UNSAT N/A COMMENTS:

Performance Step 3:

Critical Not Critical States the reason why R03 does not meet the requirements for license reactivation.

Standard:

RO3 performed plant tour with a non-licensed person contrary to OPDP-1O SAT UNSAT N/A COMMENTS:

END OF TASK STOP TIME

Adinin SRO Ala PAGE 1

OF 5

OPERATOR:

RO SRO___

DATE:

JPM NUMBER:

Admin SRO Ala TASK NUMBER:

U-000-AD-17 TASK TITLE:

Determine Adequate Performance of License Reactivation K/A NUMBER: 2.1.4 K/A RATING:

SRO 3.8 TASK STANDARD: Determine which of the reactivating personnel have correctly completed the reactivation requirements.

LOCATION OF PERFORMANCE:

Class Room REFERENCES/PROCEDURES NEEDED: OPDP-lO VALDATION TIME: 15 minutes MAX. TIME ALLOWED: (Completed for Time Critical JPMs only)

PERFORMANCE TIME:

COMMENTS:

Additional comment sheets attached? YES NO RESULTS:

SATISFACTORY UNSATISFACTORY___

SIGNATURE:

DATE:_______

EXAMINER

INITIAL CONDITIONS:

6 off-shift licensed personnel are returning to shift from rotating assignments and are reactivating their licenses. The following table gives information as to hours worked under direction of an activated licensee, tours performed, etc.

Pre-activation License Meeting Shift 1 Shift 2 Shift 3 Shift 4 Shift 5 Shift 6 Plant Tour Ops Training 12 hrs 12 hrs 12 hrs 12 hrs Conducted Manager U-I U-2 SRO U-2 SRO U-2 SRO complete plant SRO1 Ops SRO tour with SM Superintendent Ops Training 12 hrs 12 hrs 12 hrs 12 hrs Per SM Manager U-2 U-2 SRO U-3 SRO STA instructions SRO2 Ops SRO conducted Superintendent complete plant tour with Outside US (SRO)

Per SM Ops Training 12 hrs 12 hrs 12 hrs 12 hrs 12 rs instructions SRO 3 Manager U-i U-I U-2 WCC conducted 9P SRO SRO SRO SRO SRO complete plant Supenntendent tour with STA (SRO)

Ops Training 12 hrs 12 hrs 12 hrs 12 hrs 12 hrs 12 hrs Per SM Manager U-3 RO U-3 RO U-l RO tagging U-2 RO U-2 RO instructions Rol Shift Manager UO RO conducted complete plant tour with STA (SRO)

Ops Training 12 hrs 12 hrs 12 hrs 12 hrs l2hrs Per SM Manager U-2 RO U-2 RO U-I RO U-3 RO U-3 RO instructions Ops Called for conducted Superintendent Random complete plant Drug test tour with extra R02 during RO assigned to shift

that shift crew missed end of shift turnover Ops Training 12 hrs 12 hrs 12 hrs 12 hrs 12 hrs Conducted Manager U-2 RO U-2 RO U-3 RO U-3 RO U-i RO complete plant R03 Ops tour with U-2 Superintendent RB AUO IMTIAT1NG CUE:

The Shift Manager has tasked you to determine which of the personnel, if any, have completed the requirements for license reactivation. If any personnel do not meet the requirements for license reactivation state the reason(s) why.

Admin SRO Ala PAGE 3

OF 5

Class Room INITIAL CONDITIONS:

6 licensed personnel are returning to shift from rotating assignments and are reactivating their licenses. The following table gives information as to hours worked under direction of an activated licensee, tours performed, etc.

Pre activation Shift Shift Shift Shift Performance License Meeting 1

Shift 2 3

Shift 4 5

6 Plant Tour Step Ops Training 12 his 12 his 12 his 12 his Conducted complete Meets Manager U-I U-2 SRO U-2 SRO U-2 plant tour with SM requirements SRO1 Ops SRO SRO Superintendent Ops Training 12 his 12 his 12 his 12 his Per SM instructions Does Not Manager U-2 U-2 SRO U-3 STA conducted complete Meet 5R02 Ops SRO SRO plant tour with Outside Superintendent US (SRO)

Meets Ops Training 12 his 12 his 12 his 12 his 12 Per SM instructions requirements SRO 3 Manager U-i U-i U-i U-2 WCC conducted complete OpS SRO SRO SRO SRO SRO f) ant tour with S Superintendent (SRO)

Ops Training 12 his 12 his 12 his 12 his 12 his 12 his Per SM instnictions Does Not ROl Manager U-3 U-3 RO U-I tagging U-2 U-2 conducted complete Meet Shift Manager RO RO UO RO RO RO plant tour with STA (SRO)

Ops Training 12 his 12 his 12 his 12 his l2his Per SM instructions Meets Manager U-2 U-2 RO U-I U-3 RO U-3 conducted complete requirements Ops RO Called for RO RO plant tour with extra R02 Superintendent Random Drug RO assigned to that test during shift

shift missed end of shift turnover Ops Training 12 his 12 his 12 his 12 his 12 his Conducted complete Does Not Manager U-2 U-2R0 U-3 U-3 RO U-I planttourwithU-2 Meet R03 Ops RO RO RO RB AUO Superintendent INITIATING CUE:

The Shift Manager has tasked you to determine which of the personnel, if any, have completed the requirements for license reactivation. If any personnel do not meet the requirements for license reactivation state the reason(s) why.

Admin SRO Ala PAGE 4

OF 5

START TIME____

Performance Step 1:

Critical X Not Critical Analyzes information provided to determine which personnel meet the requirements for license reactivation.

Standard:

Determines SRO 1, 3 and RO 2 meet the requirements for license reactivation.

SAT UNSAT N/A COMMENTS:

Performance Step 2:

Critical X Not Critical States the reason why RO 1 does not meet the requirements for license reactivation.

Standard:

ROl did not interview with Ops Superintendent contrary to OPDP-1O.

SAT UNSAT N/A COMMENTS:

Performance Step 3:

Critical Not Critical States the reason why R03 does not meet the requirements for license reactivation.

Standard:

RO3 performed plant tour with a non-licensed person contrary to OPDP-1O SAT UNSAT N/A COMMENTS:

Adinin SRO Ala PAGE 5

OF 5

Performance Step 4:

Critical X Not Critical States the reason why SRO2 does not meet the requirements for license reactivation.

Standard:

SRO 2 did not complete shift 5 under the supervision of an active licensed individual in the position (Shift Manager or Unit Supervisor as applicable for SROs), therefore he did not meet his 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> requirement SAT UNSAT N/A COMMENTS:

END OF TASK STOP TIME

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Adinin RD Alb PAGE 1

OF 10 OPERATOR:

RO SRO___

DATE:

JPM NUMBER:

Admin RO Aib TASK NUMBER:

Conduct of Operations TASK TITLE:

2-SR-2 ICS Computer points K/A NUMBER: 2.1.19 K/A RATING: RO 3.9 TASK STANDARD: Perform Operator logs using ICS screens in accordance with 2-SR-2 Instrument Checks and Observations for log tables 1.1, 1.6, 1.25, and 1.30. Verifies acceptance criteria is satisfied in accordance with notes.

LOCATION OF PERFORMANCE:

Unit 2 Simulator (ICS computer terminal)

REFERENCES/PROCEDURES NEEDED: 2-SR-2 Rev 71 VALIDATION TIME: 20 minutes MAX. TIME ALLOWED: (Completed for Time Critical JPMs only)

PERFORMANCE TIME:

COMMENTS:

Additional comment sheets attached? YES NO RESULTS:

SATISFACTORY UNSATISFACTORY SIGNATURE:

DATE:_______

EXAMINER

INITIAL CONDITIONS: You are a Unit Operator assigned to Unit 2, and it is Friday morning at 0800. 2-SR--2, Instrument Checks and Observations, is being performed. All 2-SR-2 instrument checks and observations are complete with the exception of table 1.1, 1.6, 1.25, and 1.30.

INITIATING CUE: The Unit Supervisor directs you as the Unit Operator to complete 2-SR-2 for tables 1.1, 1.6, 1.25 and 1.30.

Admin RO Aib PAGE 3

OF 10 Simulator iNITIAL CONDITIONS: You are a Unit Operator assigned to Unit 2, and it is Friday morning at 0800. 2-SR-2, Instrument Checks and Observations, is being performed. All 2-SR-2 instrument checks and observations are complete with the exception of table 1.1, 1.6, 1.25, and 1.30.

INITIATING CUE: The Unit Supervisor directs you as the Unit Operator to complete 2-SR-2 for tables 1.1, 1.6, 1.25 and 1.30.

START TIME Adinin RD Alb PAGE 4

OF 10 Performance Step 1:

Critical X Not Critical Refers to 2-SR-2, Instrument Checks and Observations, table 1.1 AELE Li CORE ThERMAL POWER 4ND CORE POWER D>STRIEUTION DAY SHIFT Notes

3. 4 & 5 Standard:

Completes table 1.1 Data for Friday at 0800, Records 3456 for MWt, 100% for RTP,.899 for MFLCPR,.672 for MAPRAT and.769 for MFDLRX 0500 Fnday APPUC5B>L1TY:

Mode 1.yhen 3 25% SW Record the readings as soon as ssthte after the generator breaker has been closed.

Criteiia Source:

3.21.1;_3.2.21. 3.2.3.1;_UEFINIflONS SECTION 1.1_-FSAR 3.7.7 LOCAflON:

tCSCoreputer (Case SummarY

- CEUM)_______

Review blEats Core Percent TIME Thems Power L00T MFLCPR MAPRAT MFDLRX UMIT Unit DAY Note 2 Power (MS)

(% RIP>

(AC)

(Note 3)

INote 3)

>tote 3)

IAC)

Operator Unit Supw 1200 WEEK:

1400 Satrrday 1200 1400 Sunday Notes i&2 1600 0600 Morvdai 1200 1400 1600 1600 NOTES ARE FOLLOWING THE TABLE!

SAT UNSAT N/A COMMENTS:

Admin RO Alb PAGE 5

OF 10 Performance Step 2:

Critical X Not Critical (1)

Compliance with the Licensed Power Limit (LPL) (3458 Mwt) is demonstrated by the following process:

A.

No actions are allowed that would intentionally raise core thermal power above 3458 Mwt for any period of time. Small, short-term fluctuations in power that are not under the direct control of the unit operator are not considered intentional.

B.

Closely monitor the thermal power during steady-state power operation with the goal of maintaining the two-hour average at or below 3458 Mwt. If the core thermal power average for a 2-hour period is found to exceed 3458 Mwt, Operations take timely action to ensure that thermal power is less than or equal to 3458 MWt.

(This isimplemented by taking action when any running average less than or equal to the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average exceeds 3458 Mwt.)

C.

The core thermal power for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period ( 8 hr average) is not to exceed 3458 Mwt.

D.

If an evolution is expected to cause a transient increase in reactor power that could exceed 3458 Mwt, action should be taken to lower core power prior to performing the evolution.

E.

IF power is > 3463, REDUCE power.

F.

IF power is 3458 to 3463 MWt after allowing time for recent perturbations to settle, REDUCE power and EVALUATE the trend.

0.

IF any running 30 mm

Avg, 1 hr average, or 2 hr average is> 3458 MWt. REDUCE power.

(2)

Core Thermal Power is normally recorded every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when required. However, these readings may be marked N/A during TIP trace runs, control rod pattern adjustments, or anytime Core Monitoring System is blocked and/or < 25%

power. The Reactor Engineer is responsible for monitoring Core Thermal Limits. Monitoring of Core Thermal Power and other Core Thermal Limits is recommended following completion of planned rise in power and following any unexpected power change. If core monitoring software becomes unavailable, the Unit Supervisor/SRO and Reactor Engineer shall determine the appropriate frequency for monitoring Core Thermal Power but should not exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, using backup core monitoring computer, and taking into consideration current core conditions and margin to thermal limits. Power changes should not normally be made without the core monitoring software being available.

(3)

A. Consult the Reactor Engineer when value 0.955. Refer to 0-TI-248 for Administrative Limits.

B. Consult the Reactor Engineer when value ? 0.83 5. Refer to 0-TI-248 for Administrative Limits.

C. Consult the Reactor Engineer when value 0.985. Refer to 0-TI-248 for Administrative Limits.

(4)

If any Turbine Bypass valve(s) are inoperable or a Recirculation Loop is out of service, contact the Reactor Engineer and refer to the COLR for Turbine Bypass Out of Service (TBOOS) or Single Loop Operation (SLO) limits which must be applied.

(5)

MAPRAT within limits is used to verify that all APLHGRs are within the limits specified within the COLR, and <.850.

MFDLRX within limits is used to verify that all LHGRs are within the limits specified within the COLR.

MFLCPR within limits is used to verify that all MCPRs are within the limits specified within the COLR, and <.970 when core thermal power is > 90% RTP.

Standard:

Initial for Unit Operator for Friday at 0800 when acceptance criteria is verified in accordance with above notes.

SAT UNSAT N/A COMMENTS:_____________________________

Adinin RO Aib PAGE 6

OF 10 Performance Step 3:

Critical X Not Critical Refers to 2-SR-2, Instrument Checks and Observations, table 1.6 EAT BPL%NCE LAED i:

tcILsrc:

% wrP rr L Lrc 04Y SHIFT Wk LOCAflc#

G P1

!1 T1 E.

I82r Tzfdd ;4

-4 F)

-49 rF 17 <fl t2 ATJ NAT IAF My

Thy

TNiiy

Standard:

Completes table 1.6 Data for Friday. Records 377.2 for listed ICS points SAT UNSAT N/A _COMMENTS:

Admin RO Aib PAGE 7

OF 10 Performance Step 4:

Critical X Not Critical (1)

The computer points listed in Table 1.B. 1 and 1.B.2 are inputs to the ICS Core Thermal Power Heat Balance calculations. The points are monitored to ensure the inputs are in agreement and to ensure the license limits for thermal power are maintained. In addition to the above, these points should be monitored any time reactor power changes are performed.

(2)

A difference between Feedwater temperature points 3-48A, 3-48B, 3-50A, 3-50B, and NSSOO17 of greater than 2 degrees will require the notification of Site Engineering and suspending any rise in power until the discrepancy is resolved.

(3)

An alarm setpoint being exceeded will require notifying the Unit Supervisor immediately and, if action cannot be taken immediately to return the value to within limits, Site Engineering will be notified for assistance.

103 POINT DESCRIPTION HI ALARM H H ALARM CALCOSO Rx Power 30 Mm Avg.

3458 3483 CALCO2I Rx Power 1 Hr Avg.

3458 3481 CALC033 Rx Power 2 Hr. Avg 3458 345 CALC08 GeneratcrPower 1135 CALGO28 Efnoency 30 36 CALCO27 Load Le N)A I I3 CALCO24 Rx Power%

1002 100.5 TABL3 1.1 Standard:

TASLE 15.2 CS POINT 2ESCRIPTION HI ALARM HI HI ALARM 3-45A FW Temp 382 385 3-455 FW Temp 352 386 3-SOA FWTCmp 382 386 3-555 FSV Temp 302 385 NSSQ0I7 Aug. FWTernp.

332 385 C0N510G ToI RWCU Po Documents Sat and initials for Unit Operator for Friday when Maximum Deviation between Feedwater temperature computer points are within 2 degrees (Note 2) and the conditions of Note 3 are satisfied lAW with tables l.B.l and 1.B.2.

SAT_ UNSAT N/A COMMENTS:

Adinin RO Alb PAGE 8

OF 10 Performance Step 5:

Critical X Not Critical Refers to 2-SR-2, Instrument Checks and Observations, table 1.25 TA&E 125 ITUENTT53N DAY S34IFT APPUCASIUTY:

1.10555 52 R5551,5 25 1555Usd 5t5 teo.

Ce23 CTurce:

ecbUc{ 5eu5es,5l15 M552Si TTA

3. 2

.CCATlON:

Psne 25-14 aid CS CcloSater 5eew 122s5

1. LP2Y5ASTED el1

F5M0 15521150 APSU

.Ald A0P%4 LPSM APM LPPM OpSj5 LPP\\4 5iaoos5 2% 25 UT 1.IAX DES S1 2215 DAY TillS

=2 53 55 53 53

1 51 SUe2 0215 2 AC1 TAJNDAT JO Url Tlr55r 1I25Y 3U2 SODISOT TurSoy

!2i55y 502

TSs535 COST UArlssd3y

522

11ulss3 5522 Standard:

Completes table 1.25 Data for Friday, #LPRMs reading 3% on ICS. Records ZERO SAT UNSAT N/A COMMENTS:______________________________

Performance Step 6:

Critical X Not Critical (1)

Record number of LPRMs bypassed in the four APRM and LPRM cabinets as observed at Panel 2-9-14.

add these values together and record as Total # LPRMs Bypassed.

(2)

Less than 20 LPRMs in OPERATE or Less than 3 per level for any APRM will result in a Rod Block and a trouble alarm on the display panel. This does not yield an automatic APRM trip, but does, however, make the associated APRM INOP.

(3)

Record number of LPRMs reading less than 3% on the LPRM printout or display on ICS.

(4)

MAX DEV is not required to be met when the APRMs are downscale; however, unexpected inconsistencies should be reported to the Reactor Engineer. The total number of LPRMs bypassed shall equal the number of LPRMs reading less than 3% on ICS.

Standard:

Documents Sat and initials for Unit Operator for Friday when the conditions of Note 4 are satisfied.

SAT_ UNSAT N/A _COMMENTS:______________________________

Admin RO Aib PAGE 9

OF 10 Performance Step 7:

Critical X Not Critical Refers to 2-SR-2, Instrument Checks and Observations, table 1.30 TteLE

.!3 3ACTOte VE3CEL 3TEAI.1 OME EC3L1RE ITUVENTA1CN OAY EHFT

.&°FCAElUtY.

I 2

Rs 010 0ql.

Oil Curve. apse Ire1rur1o.

!.3.3.I.I)12, 3.3.3.11,3.4131 LOCATION:

CC (NOte 1 1 4) 2-345 2-340 2-3-33 28-13 evlee. 1P11C 14A.1I 2

C C

A Reterense TRIO 2EV 2EV VAX Ail 2212 Lil.

(135104) 3-74A 3-745 AC) 2-851-3-220 28-2-3-222 28-24-2255 2-OI2-3-22.Mv

,AC)

LIEIIT SAT.UNCAT LID 7rISoy 2-302 215111303 1105 23132y 1155 V 1033 05 oil Oolç 4

To0u-133

105 ieOPeod 03 1-105 70015333 1-103 Standard:

Completes table 1.30 Data for Friday. Records 1035 for ICS points 3-74A and B.

SAT UNSAT N/A _COMMENTS:________________________

Adni+/-n RO Aib PAGE 10 OF 10 Performance Step 8:

Critical X Not Critical (1)

These readings may be obtained from ICS using the Single Value Display or from the ATU output voltage translated into a PRESSURE Signal for the specific instruments. For ICS, type in SVD for Single Value Display, enter the point desired as 3-74A, record reading, select F4, enter 3-74W, record the second reading.

(2) 3-74A and 3-74B have a Maximum allowable deviation of 40 psig, AND 2-PIS-3-22D, -PIS-3-22C, 2-PIS-3-22BB, & 2-PIS-3-22AA, have a Maximum allowable deviation of 60 psig. No comparison is required between the 3-74A(B) and 2-PIS-3-22D(C)(BB)(AA).

(3) 3-74A and 3-74B SHALL be 1050 psig. 2-PIS-3-22D, 2-PIS-3-22C, 2-PIS-3-22BB, & 2-PIS-3-22AA SHALL be 1090 psig.

(4) 3-74A and 3-74B are to be recorded at 0800. The Auxiliary Instrument Room readings are not required to be taken at precisely 0800.

(5)

Following a change to Reactor Power and/or Pressure, verify the Steam Dome Limits are within the 0-TI-248, Administrative Limits and Design Analysis Limits (Appendix S)

Standard:

Documents Sat and initials for Unit Operator for Friday when the conditions of Notes 2 and 3are satisfied.

SAT_ UNSAT N/A COMMENTS:_______

END OF TASK STOP TIME

• • • • • • • • • • • • • • • • • • • Student Handout *********************

Admin RO Aib Unit 2 RFN Instrument Checks and Observations 2-SR-2 Unit 2 Rev. 0072 Page 21 of 149 (Page 1 of 87)

Surveillance Procedure Data Package

- Modes 1, 2, & 3 CORE IERMAL POWER AND CORE PtYNER DiSTRDLlTlON Thswee NeAtwe&

AFPLCAIL1TY:

Mode 1 when 25% RTP lecord the readings as socn as sible after The generator breaker has been closed.

Cr(teria Source:

3.2t1: &22.1: 3231: DEFINITIONS SECTION 11

- FSAR 37T LOCATlO1:

(CS Conputer (Case Summary

- CSUM Reuw Inials Core Percent TIME rnerrnai Power LIMIT MLCPR MAPRAT MDLR)(

LIMIT Unit DAY Note 2 Power (MWt)

(% Ri?)

(AC)

(Note 3A)

(Note &8)

(Note 3(D) iAC)

Operator Unit Supvr 0830 1030 1230 FrId3 1430 i3o 1LT3O 0830 1030 1230 Sa:urchly 1430 1630 1830 Notc Notcs 0830 1&2 3,4&5 1030 S

d 1230 Ufl3V 1830 0030 1030 Monday 1630 1wn TABLE 1.1 CAY SHIFT WSEK:

NOTES ARE FOLLOWINC THE TABLE!

• • • • • • • • • • • • • • • • • • • Student Handout Admin RO Aib Unit 2 BFN instrument Checks and Observations 2-SR-2 Unit 2 Rev. 0072 Pa9e 23 o1 149 (Page 3 of 87)

Surveillance Procedure Data Package - Modes 1, 2, & 3 (t)

Cdmpftance with the Licensed Power Linirt (LPL) (3456 Mwt) is demonstrated by Ihe following process:

A.

Na actions are allowed that would intentionalty raise core themial power above 3458 Mwt for any period of time Smail, short-term fluctuations in power that are not under the direct conirol of the unit operator are not considered intentiona1.

6.

Closely monitor the thermal poier during steady-state power operation with the goal of maintaining the t>-hour average at or below 3458 M&t If the core themial power average for a 2-hour period is found to exceed 3458 Mwt, Operations take timely action to ensure that thermal power is less than or equal to 3458 MWL (This is implemented by taking action when any running average less than or equal to the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average exceeds 3458 1wt)

C.

The core thermal power for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period ( 8 hr average) is not to exceed 3456 Mwt D.

If an evoluuon is expected to cause a transient increase in reactor power that could exceed 3458 Mwt, action should be taken to lower core power prior to performing the evolutiorL E.,

IF power is.> 3463, REDUCE power.

F.

IF power s. 3458 to 3463 MWt alter allowing time ftr recent perturbations to settle, REDU CE power and EVALUATE the trend G.

IF any running 30 mm Avg. I hr average, or 2 hr average is 3458 MWt, REDUCE power.

(2)

Cqre Thermal Power is normally recorded every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when required. However, these readings may be marked N/A during TIP trace runs, control rod pattern adjustments, or tnytirne Core Monitoring System is blocked and/or < 25% power The Reactor Engineer is responsible for monitoring Core Thermal Limits. Monitoring of Core Thetmal Poer and other Core Themial Limits is recommended following completion of planned rise in power and following any unexpected power change.

If core monitoring solIare becomes unavailable, the Shift Manager and Reactor Engineer shall determine the appropriate frequency for monitorIng Core Thermal Power but should not exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, using backup core monitoring computer, and taking into consideration current core conditions and margin to thermal limits. Power changes should not normally be mde without the core rronitQring software being available, (3)

A.

Consult the Reactor Engineer when value 05& Refer to 0-11-248 for Administrative Limits.

5.

Consult the Reactor Engmeer when value O.835 Refer to 0-11-248 for Administrative Limits.

C.

Consult the Reactor Engineer when value Refer to 0-11-243 for Adrninistraiie Limits.

• • • • • • • • • • • • • • • • • • • Student Handout *********************

Admin RO Aib Unit 2 BFN instrument Checks and Observations 2-SR-2 Unit 2 Rev, 0072 Page 31 of 149 (Page 11 of 87)

Surveillance Procedure Data Package

• Modes 1, 2,

& 3 TABLE I

NEAT BALANCE RELATED ICS ALARM SETPOINTS (Note 1)

DAY SI-lIFT WEEK:

APPLlCAlLiTY:

Model when 25% RTP Record the readings as soon as possibie after the genenator breaker has been cosed Criteria Source:

BFPER95I 914 LOCATlOJ:

ICS Computer Review Initials ICS Points HI and HI HI alarm setpoirits listed in 3-48A 3-483 3-SflA 3-SOB SSOO17 MAX Table 1.E1 & tB2 are NOT exceeded. (Note 3)

Unit (F)

( F)

(CF)

E)

(F)

DEV SAT / LJNSAT I N/A tiC Supr Fiiday Saturday Sunday 2F

Monday, (Note 2)

Tueeda Wednesdiy Thurdaj Th computer paints listed in Table I B. I and 1 B2 are inputs to the CS Core Thermal Par Heat Balance calculations. The points are monitored to ensure the inputs are in agreement and to ensure the license limits for thermal power are maintained. In addition to the aioove these points should be monitored any time reactor power changes are p.erfonl1ed.

A difference beteen Feedwats, temperature points 3-48A. 3-48B. 3-50& 3-5(18. and NSSOO17 & greater than 2 degrees will require the notification of Site Engineering and suspending any rise in power until the discrepancy is resolved.

An alarm setpoint being exceeded will require notifying the Unit Supeivisom immediately and. if action cannot be taken immediately to return the value to within limits, Site Engineering will be nctitled for assistsnce.

TABLE I.B ICS POINT DESCRiPTION HI ALARM lii HI ALARM CALCO2O Rx Power 3(1 Mm Avg.

3458 3483 CALCO2I Rx Power 1 Hr. kg.

3458 3451 CALCO83 Rx Power 2 Hr. Avg.

3458 345 CALCO(18 Generator Power 1185 1 IQO CALCO26 Eiflciency 35 36 CALCO27 Load Une N/A 113.6 CALCO24 Rx Power%

100.2 100.5 TABLE l B2 105 POINT DESCRIPTION HI ALARM HI HI ALARM 3-48A FWTemp 382 386 3-48B FWTemp 382 388 3-SOA FW Temp 382 388 3-502 FW Temp 382 388 NSSDO 1?

Avg. FW Temp 382 386 CONSO400 Total RWCU Flow 0.15 N/A T1isweet Nextwee fr

• • • • • • • • • • • • • • • • • • • Student Handout *********************

Admin RO Aib Unit 2 BFN instrument Checks and Observations 2-SR2 UnIt 2 Rev, 0072 Page 48 of 149 (Page 28 of 87)

Surveillance Procedure Data Package - Modes 1, 2, & 3 APPUClLlTY:

Modes 1 & 2 Readings are requIred at ar times.

Criteria Scurce:

Tethnical Requiremerns Manual TSR 3&&3 LOCATION:

Panel 2-t4 and ICS Computer Review Initials

1. LPRMsBYPASSED Noi Total #

1. ofLPRM LPRMs readings APRM LPRM APRM LPtRM APRM LPRM APRM LPRM Bypassed 3% on lOS MAX OEJ All Data DAY TIME

1. 2

1. 2

1. 4

1. 4

1. 3

1. 3

1. 1

1. 1 (Note 2)

Note 3)

(AC)

SATAJNSAT 1X)

Unit Suprr Friday (1800 0

0 o

0 0

(1 0

o Sanirdy 0800 Sunday 0800 0

Mondai 0800 (Note 4)

TueedaV 0803 Weesca 0800 Thursday 0800 (1) acord number of LPRMs bypassed in the four APRM and LPRM cabinets as observed at Panel 2-9-14. Add these values together and record as Total # LPRMs Bypassed (2)

Less than 20 LPRMs in OPERATE or Less than 3 per level for any APRM will result in a Rod Block and a trouble alarm on the display panel. This does not yield an automatic APRM trip, but does, however, make the associated APRM INOP (3)

Fecord number of LPRMS reading less than 3% on the LPRM printout or display on ICS (4)

MAX 0EV is not required to be met when the APRMs are downscale; however, unexpected inconsistencies should be reported to the Reactor Engtn&er. The total number of LFRMs bypassed shall equal the number of LPRMS reading less than 3% on ICS TASLE 5

LPRM INSTRUMENTATION DAY SHIFT WEEK:

Thisweel Ne)tweel

• • • • • • • • • • • • • • • • • • • Student Handout *********************

Admin RO Aib Unit 2 BFN instrument Checks and Observations 2-SR-2 Unit 2 Rev. C072 Page 54 of 149 (Page 34 of 27)

Surveillance Procedure Data Package - Modes 1, 2, & 3 ThIsweeJ NeXt weeJ TASLE REACTOR VESSEL STEAM DOME PRESSURE INSTRUMENTA]1ON DAY SHIFT WEE:

APFUCAB1LITY:

Mds I & 2 Readings are reqund a all times, Sweianc4Rquirements:

13111(13), 3331i,3.41Di LQCATION ICS (Ncte 1 & 4 2--8 2-0-8 2--B4 2-a-3 Review Initials MAX D

C B

A Reference TiME OE DEV MAX Au Data Leg (Nots 4) 3-74A 3-74B (AC) 2-PIS-3-22D 2-PlS-3-22C 2-PlS-3-228B 2-PiS-3-22AA (AC)

LIMIT SATUNSAT LJO Unit SLpr Fhthy 0600 1035 1035 1035 1035 Saturday 0600 Sunday 0600 Mcnday 40 Psi 60Psi Nte 3 (Note 2)

(Note 2) f4ote5 Tuesday 0630 Wadnesda 0600 Thursday 06313 (1)

These readings ina be obtatned from ICS using the Single Vakte Display or from the ATU output voltage translated into a PRESSURE Signal for the specific instrunients For ICS, tjpe in SVD for Sinçde Value Display, enter the point desired as 3-74A°, record reading, select F4, enter 3-74B, record the second reading.

(2 3-74A and 3-74B have a Maximum allowable deviation of 413 psiçj, AND 2-PIS-3-22D, 2-PtS.322C, 2PlS-3-228B. & 2-PIS-3-22AA, have a Maximum allowable deviation of 63 ps(g. No comparison is required between the 374A(E) and 2-PlS-3-22D(C)(8)(?A)

(3) 3-74A and 3748 SHALL be 1050 psig. 2PlS-3-22D, 2.-PlS-3-22C, 2-PlS-322B6, & 2-PIS-3-22AA SHALL be i090 psig (4) 3-74A and 3-74B are to be recorded at 0S00 The Auxiliar Instrument Room readings are not required to be taken at precisely 0800.

(5)

Folkwing a change to Reactor Power or Pressure, verify the Steam Dome Limits are within the 0-Tl-248, Administrative Liniits and Design Anasis limits (Appendix 5)

TADLE 1.1 (Page 1 of 87)

CAY SHIFT Th.S we WEEK:

fr,

• • • • • • • • • • • • • • • • • • • ANS4TER KEY ***************

BFN Instrument Checks and Observations 2-SR-2 Unit 2 Rev. 0072 Page 21 of 149 CORE ICRMAL POWER AND CORE POWER DI3TR.OLTION SurveiIIanc9 Procedure Data Package

- Modes 1, 2, & 3 Admin RO Aib Unit 2 Next wel APPLCAILITY:

f1ode 1 when 25% RTP Hecord the readings as socn as p ssible after the generator breaker las been closed.

Criteria_Surce:

3.211: &2,2.i:_3.231: DEFINITIONS_SECTION_11_- ESAR_31.7 LOCATION:

ICS Computer (Case Summar -_CSUM_______

Reukew Inttial Core Percent TIME Thermal Power UMIT MTLCPR MAPRAT MrDLRX LIMIT Unit DAY Note 2 Pnr 1MWt

% RTP (AC)

(AC)

(rrtnr Unit Supvr 08)0 1t71)i) n.

(hs72 0.769 InItials 1000 Frlda 16J0 10)0 0800 10)0 12)0 Sa:urday 14)0 1600 18)0 NOtDS Notes 08)0 1&2 34,&5 10)0 S

d 12)0 un 14)0 1WIO 18)0 0&)0 1000 M

d i2itJ on 4 14)0 16)0 iRfl NOTES ARE FOLLOWINQ THE TABLE!

• • • • • • • • • • • • • • • • • • • ANSWER KEY *********************

Admin RO Aib Unit 2 BFN instrument Checks and Observations 2-SR-2 Unit 2 Rev. 0072 Pane 23 Gt 1:49 (Page 3 of 87)

Surveillance Procedure Data Package

- MOdes 1, 2, & 3 (1)

Cdnipance With the Ucensed Power Linih {LPL) (3456 Mwt)is demonstrated by the following process:

A.

No actions are allowed that would intentionally raise core thema1 power above 3458 Mwt for any period of time. SniaU, short-term fluctuations in power that are not under the direct: conirol of the unit operator are no considered intentionaL 5.

Closely monitor the thermal power during stead,-state power operation with the goal of maintaining the two-hour average at or below 3458 Mwt.

If the core thermal power average for a 2-hour period is found to exceed 3458 Mwt, Operations take timely action to ensure that thermal power is less than or equal to 3458 vW/L (This rs implemented by taking action when any running average less than or equal ta the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average exceeds 3458 Mwt.)

C.

The core thermal power for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period ( 8 hr average) is not to exceed 3458 Mwl D.,

if an evolutton is expected to cause a transient increase in reactor power that could exceed 3458 Mt, action should be taken to lower core power prior to performing the evoiuuon.

E.

IF power is.> 3463. REDUCE power.

F.

iF power is. 3458 to 3463 MWt after allowing time for recent perturbations to settle REDUCE power and EVALUATE the trend.

G IF any running 30 miii Avg. I hr average, or 2 fir average is> 3458 MWt, REDUCE poweL (2)

Care Theimal Power is normally recorded every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when required.. However, these readings may be marked N/A during TIP trace runs, control rod pattern adjustments, or anytime Core Monitoring System is blocked andior < 25% power. The Reactor Engineer is responsible for monitoring Core Thermal Limits. Monitoring of Core Thermal Pcer and other Core Themial Limits is recommended following completion of planned rise in power and fdlowing any unexpected power change.

If core monitoring software becomes unavailable, the Shift Manager and Reactor Engiieer shall determine the appropriate frequency for monitoring Core Thermal Power but should not exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, using backup core monitoring computer, and taking into consideration current core conditions and margin to thermal limits. Power dhanges should not normally be made without the core nionitoiing software being available.

(3)

A.

Consult the Reactor Engineer when value 095& Refer to Q-Tl-248 for Adniinistratie Limits.

5.

Consult the Reactor Engineer when value 0.83& Refer to 0-11-246 for Administrative Limits.

C.

Consuft the Reactor Engineer when value D.

Refer to O-Tl-24 for Adnilnistrative Limits

C a C

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C C

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0 Li)

C)

Cl)

-n rn 31 Ii C-)

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C C

C Z

(?mo 0

+, I D

o

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Ti

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Jz_

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C 3

z.ii I

C a) rn 1

Ii Cl)rn Z

I

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-Q m

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Cl)

C) m C

ci

.4 C,

to 0

U) z (b

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z U) 3mz C,

C, U) 0 cr U) 0 U)

P Jw o,.

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-I k) z rn C

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-l k) x 0-*

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Cl) a

• • • • • • • • • • • • • • • • • • • ANS\\VER KEY *********************

Admin RO Aib Unit 2 BFN Instrument Checks and Observations 2-SR-2 Unit 2 Rev. 0072 Page 48 of 149 (Page 28 of 87)

Surveillance Procedure Data Package - Modes 1, 2, & 3 TABLE 125 LPRM 1NSTRUM4TATON DAY SHIFT WEEK:

APPLlCLETY:

Mcde I & 2 Readings are requid at al brrtes.

Criteria &urce:

Tethical Requirernerns Manual TSR 33&3 L0CATkN:

Panet 2-14 arid ICS Cornput&

Reiiew Initials LPRMS BYPASSED NolV Thtal#

1. ofLPRM LPRMs readings APRM LPRM APRM LPRM APRM LPRM APRM LPRM Bypassed 3% on ICS MAX 0EV All Data DAY TIME

1. 2

1. 2

1. 4

1. 4

1. 3

1. 3

1. 1

1. 1 iNote2)

(Noe3 (AC SATiJNSAT UO Unit3upvr Fricay 0800 0

0 o

U U

0 0

U SAT Initials Sawrdy 0803 Sunday 0800 0

v1onday 0800 Note4)

Tuesda 0803 Wednesday 0803 Thursday 3803 (1)

Riecord number of LPRMs bypassed in the four APRM and LPRM cabinets as observed at Panel 2-9-I4. Add these values together and record as Total # LPRMs Bypassed.

(2)

Less than 20 LPRMS in OPERATE or Less than 3 per level tar any APRM will result in a Rod Block and a troithle alarm on the display panel. This does not yield an thtomatic APRM trip, but does, however, make the associated APRM INOP (3)

Record number of LPRMS reading less than 3% on the LPRM printout or display on IC&

(4)

MAX 0EV is not required to be met when the APRMs are dowrscale; however, unexpected inconsistencies thould be reported to the Reactor Engineer. The total number of LPRMs bypassed shall equal the number of LPRM*s reading less than 3% on lOS.

ThIsw NeKtwe&

• • • • • • • • • • • • • • • • • • • ANSWER KEY *********************

Admin RO Aib Unit 2 BFN kistrument Checks and Observations 2-SR-2 Unit2 Rev 0072 Pa9e 54 of 149 (Page 34 of 87)

Surveillance Procedure Data Package - Modes 1, 2, & 3 TkIswee Nextwee.

REACTOR VESSEL STEAM DOME PRESSURE LNSTRUMENTAT1ON DAY SHIFT WEEK APFUCABLT(C Modes

& 2 Readings are required at all times.

SursdIance Requiremenls:

331.1.1f3), 333iJ, 34..1Di LOCATION:

105 (Note 1 & 4 2-g-85 2-9-8 2-0-84 2-0-83 Review Initials MAX D

C B

A Reference TIME DEV MAX Ad D5t Leg (Note 4 3-7$A 3-748 (AC) 2-PlS-3-22D 2-PlS-3-22C 2-PIS-3-2256 2-PIS4-22AA (AC)

LIMIT SATUNSAT UO Unit Supr Friday 1035 1035 1035 1035 1035 1035 SAT Inzia/s Saurday QS0Q Sunday 0800 Monday 0800 4

P9 JOPSi0 Note 3 (Note 2)

(Nlote3)

Note 5 Tuesday 0800 Wednesday 0800 Thursday 0830 (1)

Th8se readinqs may be obtained from ICS using the Single Value Display or from the ATU output voltage translated into a PRESSURE Signal for the specific instruments.

For ICS lype in SVD for Single Value Display, enter the point desired as 3-74A record reading, select F4, enter p3-748, record the second reading.

(2)

S-74A and 3-745 have a Maximum allowable deviation sf40 pslç AND 2-PIS-3-22fl, 2-PIS-3-22C, 2-PIS-3-2285, & 2-PIS-3-2ZAA. have a Maximum allowable deviation of 50 sig. No comparison is required between the 3-74A(B) and 2-PtS-3-22D(C)(E A).

(3) 3-74A and 3-745 SHALL be 1050 psig. 2-PIS-3-22D, 2-PIS-3-22C, 2-PIS-3-2256. & 2-P1S-3-22AA SHALL be 1090 psig.

(4) 3-74A and 3-745 are to be recorded at 0800, The AuxiIiay Instrument Room readings are not required to be taken at precisely 0500.

(5)

Following a change to Reactor Power or Pressure, verIfy the Steam Dome Umits are within the 0-Tl-248, Administrative Limits and Design Analysis limits (Appendix 5)

Adrnin RO Aib PAGE 1

OF 10 OPERATOR:

RU SRO___

DATE:_________

JPM NUMBER:

Admin RU Aib TASK NUMBER:

Conduct of Operations TASK TITLE:

3-SR-2 ICS Computer points KJA NUMBER: 2.1.19 KJA RATING: RU 3.9 TASK STANDARD: Perform Operator logs using ICS screens in accordance with 3-SR-2 Instrument Checks and Observations for log tables 1.1, 1.6, 1.25, and 1.30. Verifies acceptance criteria is satisfied in accordance with notes.

LOCATION OF PERFORMANCE:

Unit 3 Simulator (ICS computer terminal)

REFERENCES/PROCEDURES NEEDED: 3-SR-2 Rev 68 VALIDATION TIME: 20 minutes MAX. TIME ALLOWED: (Completed for Time Critical JPMs only)

PERFORMANCE TIME:

COMMENTS:

Additional comment sheets attached? YES NO RESULTS:

SATISFACTORY___

UNSATISFACTORY___

SIGNATURE:__________________

DATE:

EXAMINER

INITIAL CONDITIONS: You are a Unit Operator assigned to Unit 3, and it is Friday morning at 0800. 3-SR-2, Instrument Checks and Observations, is being performed. All 3-SR-2 instrument checks and observations are complete with the exception of table 1.1, 1.6, 1.25, and 1.30.

INITIATING CUE: The Unit Supervisor directs you as the Unit Operator to complete 3-SR-2 for tables 1.1, 1.6, 1.25 and 1.30.

Adrnin RO Aib PAGE 3

OF 10 Simulator INITIAL CONDITIONS: You are a Unit Operator assigned to Unit 3, and it is Friday morning at 0800. 3-SR-2, Instrument Checks and Observations, is being performed. All 3-SR-2 instrument checks and observations are complete with the exception of table 1.1, 1.6, 1.25, and 1.30.

INITIATING CUE: The Unit Supervisor directs you as the Unit Operator to complete 3-SR-2 for tables 1.1, 1.6, 1.25 and 1.30.

START TIME____

Adrnin RO Alb PAGE 4

OF 10 Performance Step 1:

Critical X Not Critical Refers to 3-SR-2, Instrument Checks and Observations, table 1.1 Th3LS!j DAY HFT AFFUCkSUTY:

2E% R ToL O RECORD thf 3tnelr 5 ttfl LcCATC CcnrC3n,:rCQUM T

Cc Pc Pr L3T MFLCP tkPPAT WDLRX

.T Ur Urt DAY ct2 MW

% T) e 3

3 Df c

ICD 11 13 1z33 1D T3 3.r33 33 D3 CD Mc1Y j

Standard:

TA.L Completes table 1.1 Data for Friday at 0800, Records 3456 for MWt, 100% for RTP,.878 for MFLCPR,.780 for MAPRAT, and.854 for MFDLRX N/A COMMENTS:

SAT UNSAT

Adinin RO Aib PAGE 5

OF 10 Performance Step 2:

Critical Not Critical (1)

Compliance with the Licensed Power Limit (LPL) (3458 Mwt) is demonstrated by the following process:

A.

No actions are allowed that would intentionally raise core thermal power above 3458 Mwt for any period of time. Small, short-term fluctuations in power that are not under the direct control of the unit operator are not considered intentional.

B.

Closely monitor the thermal power during steady-state power operation with the goal of maintaining the two-hour average at or below 3458 Mwt. If the core thermal power average for a 2-hour period is found to exceed 3458 Mwt. Operations take timely action to ensure that thermal power is less than or equal to 3458 MWt.

(This isimplemented by taking action when any running average less than or equal to the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average exceeds 3458 Mwt.)

C.

The core thermal power for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period ( 8 hr average) is not to exceed 3458 Mwt.

D.

If an evolution is expected to cause a transient increase in reactor power that could exceed 3458 Mwt, action should be taken to lower core power prior to performing the evolution.

E.

IF power is> 3463, REDUCE power.

F.

IF power is 3458 to 3463 MWt after allowing time for recent perturbations to settle, REDUCE power and EVALUATE the trend.

G.

IF any running 30 mm

Avg, 1 hr average, or 2 hr average is> 3458 MWt. REDUCE power.

(2)

Core Thermal Power is normally recorded every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when required. However, these readings may be marked N/A during TIP trace runs, control rod pattern adjustments, or anytime Core Monitoring System is blocked and/or < 25%

power. The Reactor Engineer is responsible for monitoring Core Thermal Limits. Monitoring of Core Thermal Power and other Core Thermal Limits is recommended following completion of planned rise in power and following any unexpected power change. If core monitoring software becomes unavailable, the Unit Supervisor/SRO and Reactor Engineer shall determine the appropriate frequency for monitoring Core Thermal Power but should not exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, using backup core monitoring computer, and taking into consideration current core conditions and margin to thermal limits. Power changes should not normally be made without the core monitoring software being available.

(3)

A. Consult Reactor Engineer when value 0.965. Refer to 0-Tl-248 for Administrative Limits.

B. Consult Reactor Engineer when value : 0.835. Refer to 0-TI-248 for Administrative Limits.

C. Consult Reactor Engineer when value ? 0.985. Refer to 0-TI-248 for Administrative Limits.

(4)

If any Turbine Bypass valve(s) are inoperable or a Recirculation Loop is out of service, contact the Reactor Engineer and refer to the COLR for Turbine Bypass Out of Service (TBOOS) or Single Loop Operation (SLO) limits which must be applied.

(5)

MAPRAT within limits is used to verify that all APLHGRs are within the limits specified within the COLR, and

<0.850.

MFDLRX within limits is used to verify that all LHGRs are within the limits specified within the COLR.

MFLCPR within limits is used to verify that all MCPRs are within the limits specified within the COLR, and < 0.980 when thermal power is > 90% RTP.

Standard:

Initial for Unit Operator for Friday at 0800 when acceptance criteria is verified in accordance with above notes.

SAT UNSAT N/A COMMENTS:

Adinin RO Aib PAGE 6

OF 10 Performance Step 3:

Critical X Not Critical ALJCALTY.

ie I 3A)

RECCO tT th ec2tt!

2-ute:

rIc:

Cccutfr R1v nL,I

-I t3V Ur4 4A I

!-5AF 17 fl 2

L Oup rIj

I4r

ThV W-ai Standard:

Completes table 1.6 Data for Friday. Records 377.2 for listed ICS points SAT_ UNSAT N/A COMMENTS:___________________

Refers to 3-SR-2, Instrument Checks and TA&E I ALNC A

AL&q) ET2 Nt I Observations, table 1.6 DAY SHIFT WYEK

Adinin RO Aib PAGE 7

OF 10 Performance Step 4:

Critical Not Critical (1)

The computer points listed in Table 1.B.l and l.B.2 are inputs to the ICS Core Thermal Power Heat Balance calculations. The points are monitored to ensure the inputs are in agreement and to ensure the license limits for thermal power are maintained. In addition to the above, these points should be monitored any time reactor power changes are performed.

(2)

A difference between Feedwater temperature points 3-48A, 3-48B, 3-50A, 3-50B, and NSSOOI7 of greater than 2 degrees will require the notification of Site Engineering and suspending any rise in power until the discrepancy is resolved.

(3)

An alarm setpoint being exceeded will require notifying the Unit Supervisor immediately and, if action cannot be taken immediately to return the value to within limits, Site Engineering will be notified for assistance.

TABLE 18:

lOS POINT DESCR1PTION HI ALARM HI HI ALARM CALO 020 Rx Pwr 30 Mm Avg.

3456 3463 CALCO21 Rx Power 1 Hr. Ag, 3456 2461 CALC 083 Rx Powef 2 Hr. Avg.

3458 345 CALC08 Gener3tor Power 1185 11 0 CALCO2S Efflcenoy 26 36 CALGO27 Load L WA 11 38 CALCO24 Rx Power %

0G2 100.5 TABLE 18.2 CS POINT DESCRIPTIGN

Hl ALARM HI HI ALARM 3-48A FW Temp 332 386 3-488 FW Temp 382 386 3-80A FW Temp 382 388 3-5DB FWTemp 322 386 NSS17 Avg. FW Temp.

322 288 CONSO400 TotI RWCU Flow 0158 WA Standard:

Documents Sat and initials for Unit Operator for Friday when Maximum Deviation between Feedwater temperature computer points are within 2 degrees (Note 2) and the conditions of Note 3 are satisfied JAW with tables 1.B.1 and 1.B.2.

SAT UNSAT N/A COMMENTS:____________________________

Adinin RO Alb PAGE 8

OF 10 Performance Step 5:

Critical X Not Critical Refers to 3-SR-2, Instrument Checks and Observations, table 1.25 TABLE 1 25 LPRM 1ISTRIJME1tiTATION 3AY SHIfl Standard:

Completes table 1.25 Data for Friday, #LPRMs reading < 3% on ICS. Records ZERO SAT_ UNSAT N/A Performance Step 6:

Critical X Not Critical (1)

Record number of LPRMs bypassed in the four APRM and LPRM cabinets as observed at Panel 3-9-14.

add these values together and record as Total # LPRMs Bypassed.

(2)

Less than 20 LPRMs in OPERATE or Less than 3 per level for any APRM will result in a Rod Block and a trouble alarm on the display panel. This does not yield an automatic APRM trip, but does, however, make the associated APRM INOP.

(3)

Record number of LPRMs reading less than 3% on the LPRM printout or display on ICS.

(4)

MAX DEV is not required to be met when the APRMs are downscale; however, unexpected inconsistencies should be reported to the Reactor Engineer. The total number of LPRMs bypassed shall equal the number of LPRMs reading less than 3% on ICS.

Standard:

satisfied.

Documents Sat and initials for Unit Operator for Friday when the conditions of Note 4 are SRPL1CAB1L1TY.

t.125s 1 & 2 Rca,ps are requred at all dries.

RnferTn P&L Step c.dtsda Snururi lertininal Rcqdremits Manual TSR 33.5 3 LOCATCN Panel 3S-14 and CS Crixuer eriew ntiaI5 P LRM BYPS Dilute 1)

Tutal s P LPRMs

,RMs nnnp MAX DEY APRM IFRM APRM LFRM APRB1 PRM ARRM

.FRM Bypassed i3.crsCS AC)

AlData DAY TIME

1. 2 cC

1. 4 nd uS

1. 3

1. 1

1. 1 Nut. 2)

Nuts 3)

Note i SAiUNSAT ND Ned Priocy 05D3 Saturday 55.3 Sunday S4 l.luiray

-StiSO S

Tuesday

-35D3 Wsnresdny 5500 Thurnnay 005t SAT UNSAT N/A COMMENTS:

Adinin RO Alb PAGE 9

OF 10 Performance Step 7:

Critical Not Critical Refers to 3-SR-2, Instrument Checks and Observations, table 1.30 TABLE 1 33 REACTOR VESSEL STEAM DOME PPESSL1RE INSTRUMENTAiDN DAY SHWT WE0 APFLICAEILTY:

t.1es I & 2 (Refer To &L Step 3.EA)

ReASn are reqotred at aø Snot.

Sane-(Iaree Recutettterrts.

33 1.1 l(S(, 3 3.3

- I 3.4 1-3.1 LOCAT(CN-CS Notes 1 4 4(

3-S-3d 3-0-35 34-44 34-43 Reese r.taiS MAX 0

C 0

A MAX P.eforerote 9ME

-0EV 0EV MAX AX Data o

(f.ote4(

3-74A 3-740 AC) 3P15.i-200 3-?IS-3-1C 3-015-3-2206 3-0)5-3-223-4

AC)

JIMIT SATJIJNSAT JO UrttSapar 0r-day 0500 Sata4ay 0400 Sot-rday 0500 M

40 psg 30 tot (Note 3) orway

)Ncde 2 NoteS)

Note 0)

Taesdsy 0500 Wedtes5ay 0400 Thttrsday 0500 Standard:

Completes table 1.30 Data for Friday. Records 1034 for ICS points 3-74A and B.

SAT UNSAT N/A COMMENTS:______________________________

Admin RO Aib PAGE 10 OF 10 Performance Step 8:

Critical Not Critical (1)

These readings may be obtained from ICS using the Single Value Display or from the ATU output voltage translated into a PRESSURE Signal for the specific instruments. For ICS, type in SVD for Single Value Display, enter the point desired as 3-74A, record reading, select F4, enter 3-74B, record the second reading.

(2) 3-74A and 3-74B have a Maximum allowable deviation of 40 psig, AND 3-PIS-3-22D, 3-PIS-3-22C, 3-PIS-3-22BB, & 3-PIS-3-22AA, have a Maximum allowable deviation of 60 psig. No comparison is required between the 3-74A(B) and 3-PIS-3-22D(C)(BB)(AA).

(3) 3-74A and 3-74B SHALL be 1050 psig. 3-PIS-3-22D, 3-PIS-3-22C, 3-PIS-3-22BB, & 3-PIS-3-22AA SHALL be 1090 psig.

(4) 3-74A and 3-74B are to be recorded at 0800. The Auxiliary Instrument Room readings are not required to be taken at precisely 0800.

(5)

Following a change to Reactor Power and/or Pressure, verify the Steam Dome Limits are within the 0-TI-248, Administrative Limits and Design Analysis Limits (Appendix 5)

Standard:

Documents Sat and initials for Unit Operator for Friday when the conditions of Notes 2 and 3are satisfied.

SAT UNSAT N/A COMMENTS:_______

END OF TASK STOP TIME

STUDENT HANDOUT Admin RO Aib Unit 3 BFN Instrument Checks and Observations 3SR2 Unit 3 Rev, 0068 Page 20 of 146 (Page 1 of 86)

Surveillance Procedure Data Package - Modes 1, 2, & 3 TABLE 11 CORE THERMAL POWER AND CORE POWER DISTRIBUTION DAY SHIFT WEEK:

liis week Nt weeJ APPLICABILITY:

Mode 1 when 25% RTP (Refer To P&L Step 3.6A)

RECORD the readings as soon as possible after the generator breaker has been closed.

CØteria_Source:

32 ii;_&22. 1;_323. 1;_DEFIHITIONS_SECTION_1:1_-_FSAR_17.7 Lt)CATION:

ICS Computer (Case Summary - CSUM)

Review Initials TIME Core Thermal Percent Power LIMIT MFLCPR MAPRAT MFDLRX LIMIT Unit Unit DAY Note 2 Power (MWt)

(% RTP)

(AC)

Note 3 Note 3 NoteS (AC)

Operator Supvr 080o 1000 1200 Frtday 1400 1600 1800 0800 1000 1200 Sturda 1400 11300 1600 Notes 1 Notes 0800

&2 34, &

1000 1200 Sunday 1600 1800 0800 1000 1200 Monday 1400 1600 1800 NflTF MF F(l I flW1M( ThF TAflI ll

STUDENT HANDOUT Admin RO Aib Unit 3 BFN Instrument Checks atid Obsefvations 3-SR-2 Utit 3 Rev. 0069 Page 21 of 146 (Page 2 cf 86 Surveillance PrDcedure Data Package

- Modes 1 2 & 3 TABLE 1!

CORE THERMAL FOWER AND CORE POWER DSTRIGUTON JOTEL ARE Oi THE OLLO%VING PAGE!

DAY SHIFT WEEK:

to APPLICABILITY:

Mn 1 hn > 2 RTP (Rfr Th PI cp RECORD tie readngs as sooi a posibIe after ie generator breaker hs beer cIsed Crtena3üuroe:

3.2:.i;_3.22.1;_3.23.1;_DEFINTIONSSTION1.1_-FEAR 3.7.7 LCCATIOl:

CS Cornouler iCse Sumnar

- CSUM)

Renew hi1ias TIME Core Thermal Pereent Power LIMIT MFLCPR MAPRAT MFCL.RX LiMIT Unit Unit DAY Nate 2 Poser (MWt)

{% RiP)

{ACi Nole IA Note 3.B Note 3 C 1AC)

Operator

&pr aaoa 1000 1200 uesy 140 1 00 1800 0800 1000 q

Notes Wedresay 1

NobeE 1 4

100 1 OO 0800 1000 1200 Thrsoy 1400 1 00 1800

STUDENT HANDOUT Admin RU Aib Unit 3 BFN Instrument Checks and Observations 3-SR4 Unit3 Rev 0069 Page 22 of 146 (Page 3 of 8)

Survei1ance Procedure Data Package

- Modes 1, 2. & 3 (1)

Compliance with the Licensed Power Hrnit(LPU (34S8 Mwt)is demonstrated by the follong process:

A.

Na actions are stowed that would intentionally raise core thermal power above 3458 Mwt for any period of time. Small, short-term Ituctuations in power that are not unc)er the direct control of the unit operator are not considered intentional.

S.

Closely monitor the thermal power during steady-state power operation with the goal of ir.taming the t-tiour average at or below 3458 tt If the core thermal power average far a 2-hour period is found to exceed 3458 Mwt, Operations take thneiy action to ensure that thermal power is less than or equal to 3458 MWt. (This is implemented by taking action wiien any running average less than or equal to the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average exceeds 3458 Mjt)

C.

The core thermal power for an 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period (8 hr average) is not to exceed 345.8 Mwt.

0.

If an evolution is expected to cause a transient in crease in reactor power that could exceed 3456 Mwt, action should be taken to lower core power prior to performing the evcAution E.

IF pcr is> 3463, REDUCE power.

F.

IF power is 3458 to 3463 fM after allowing tirnelor recent perturbations to settle. REDUCE power and EVALUATE the trend.

G.

IF any running 30 mm Avg, 1 hr average, or 2 hr average is> 3458 MWt, REDUCE power.

(2)

Core Thermal Power is normally recorded every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> vden required. however, these readings may be rnaked NIA during TIP trace runs, ccntrd rod pattern adjustments, or anylirne Core Monitoring System is blocked and/er 25% power. The Reactor Engineer is responsible for rnoni toting Core Thermal Limits. Monitoring of Core Thermal Power and other Core Thermal Limits is recommended following completion of planned rise in power and foltewthg any unexpected power change.

If core monitoring software becomes unavailable, the Unit SuperviscrfSRO d Reactor Engineer shall determine the appropriate frequency for monitoring Core Thermal Power but should not exceed 24 hrrura, using backup core monitoring computer, and taking into consideration current core conditions and margin to thermal limits. Power changes should not normally be made without the core monitoring software belng available.

(3)

A.

Connuit Reactor Engineer v.tten value 0.965. Refer to (lTI-248 for Adm nistratiw Limits.

5.

Consult Reactor Engineer wiien value 0,835. Refer to 0-Tl-248 for Administrative Limits.

C.

Consult Reactor Engineer when value 0.985. Refer to 0-Tl-248 for Administrative Limits.

(4)

If any Turbine Bypass valve(s) are inoperable or a Fcecirculaiion Loop is out of service, contact the Reactor Engineer and refer to the COLR for Turbine Sypass Out of Sehace (TBCOS) or Single Loop Operation (SLO) limits which must be applied.

(5>

PAAPRAT within limits is used to verity that all APLHGRs are within the limits specilied within Inc C OLR, and 0.850.

MFDLRX within limits is used to verify that all LHGRs re within the limits specified within the COLR.

MFLCPR within limits is used to verify that all MCPRS are within the limits specilied within the COLR, and s 0.980 when core thermal power is 90% RTP.

STUDENT HANDOUT Admin RO Aib Unit 3 BFN Instrument Checks and Observations 3-SR-2 Unit 3 Rev. 0068 Page 29 of 146 (Page 10 of 86)

Surveillance Procedure Data Package

- Modes i 2 & 3 TABLE 1.6 HEAT BALANCE RELATED CS ALARM SETPOINTS (Note 1)

DAY SHIFT WEEK:

to APPLICABILITY:

Mode 1 when 25% RTP(Refer To P&L Step 6A)

RECORD the readings as soon as possible after the generator breaker has been dosed.

Criteria Source:

BFPER9S1914 LOCATION:

ICS Computer Review Initials ICS Points Verify HI and HI HI alarm setpoints listed in MAX DEV Table 1.8.1 & 1.B.2 are NOT exceeded. (Note 3)

Unit 3-48A (F) 3-488 (F) 3-50A (F) 3-508 (°F)

NSSOO17 (F)

Note 2 SAT 1 UNSAT/ N/A UO Supvr Friday Saturday Sunday Monday 2F Tuesday Wednesday Thursday The computer points listed in Table I B and 1 B 2 are inputs to the lCi Core Thermal Powi,.r Heat Balance calculations The points are monitored to ensure the inputs dre in agreement and to ensure the license limits for thermal power are maintained. In addition to the above, these points should be monitored any time reactor power changes are performed.

A difference between Feeciwater temperature points 3-48A, 3-48B, 3-5.OA, 3-SOB, and NSSOOI7 of greater than 2 degrees will require the notification of Site Engineering and suspending any rise in power until the discrepancy is resolved.

An alarm selpoint being exceeded will require notifying the Unit Supervisor immediately and, if action cannot be talien immediately to return the value to within limits. Site Engineering will be notified for assistance.

IALh lb.

TABLE 1.8.2 kIswaI Ne,xtweeJ.

(1) 12)

(3)

[CS POINT DESCRIPTION HI ALARM Hi HI ALARM CALCO2O Rx Power 30 Mm Avg.

3458 3463 CAICO21 Rx Power 1 Hr. Avg.

3458 3461 CALCOS3 Rx Power 2 Hr. Avg.

3458 3459 CALCO98 Generator Power 1185 1190 CALCO26 Efficiency 35 36 CALCO27 Load Line N/A 113.6 CALCO24 Rx Power %

1002 100.5

[CS POINT DESCRIPTION HI ALARM HI HI ALARM 3-48A FWTemp 382 386 3-488 FW Temp 382 386 3-50A FWTemp 382 385 3-508 FW Temp 382 386 NSSOOI7 Avg. FW Temp.

382 386 CONSO400 Total RWCU Flow 0.168 NJA

STUDENT HANDOUT Admin RO Aib Unit 3 BFN Instrument Checks and Observations 3-SR2 Unit 3 Rev. 0068 Page 46 of 146 (Page 27 of 86)

Surveillance Procedure Data Package - Modes 1, 2, mis wee Ne)twe&c.

ThBLE t2 IYRIV NSTRUMENTATLON DAY SHWT WEEK:

APPLICA9LF[Y:

Mes 1 2

Rdings ar rqufrd at all nmi Refer To P&L Step 35A)

Cri&ia Source:

ThchnicaI Rquirnnrn I4anual TSR 3&3 L0CATK*

Pane[ 3-9-14 and ICS Computer Redew Initials

1. LPRMs BYPASSED (Note 1)

Tcsal #

1. IPRMs LPRMs reading MkX 0EV APRM LPRM APRM LPIIM APRM LPRM APRM LPRM Bypassed 3% cn iCS (AC)

MI Data DAY TIME

1. 2

1. 2

1. 4

1. 4

1. 3

1. 3

1. 1

1. 1 (t4oe 2)

(Note 3)

(Note 4)

SAT/tJNSkT 1)0 Unit Supvr Friday 0800 0

0 0

0 0

0 0

Saturday 0800 Sunday 0800 Monday (1800 0

Tuesday.

0800 Wednesday 0800 Thursday O

(1)

Reiord number of LPRMs bypassed in the four APRt4 and LPRM cabinets as observed at Panel 3-5-14. add these values tpgether and record as Total # LPRMs Ejpassed.

(2)

Leis than 20 LPRMs in OPERATE or Less than 3 per eve1 for any APRM Il resull in a Rod Block and a trouble alarm on the display panel. This does not yield an automatic APRM ttip, but does, however, make the associated APRM INOP, (3)

Record number of LPRM s reading less than 3% on the LPRM printout or display on [CS, (4)

MAIK 0EV is not required to be met when the APRMs are downscale; however, l.rnexpected inconsistencies should be reported to the Reactor Engineer, The total number of LPRMs bypassed shall equal the number of LPRMS reading less than 3% on ICS.

STUDENT HANDOUT Admin RO Aib Unit 3 BFN Instrument Checks and Observations 3-SR-2 Uflt 3 Rev. 0068 Page 52 of 146 (Page 33 of 86)

Surveilbnce Procedure Data Package

- Modes 1, 2, & 3 ThIsweeJ.

Next weeJ TASLE 130 REACTOR VESSEL STEAM DOME PRESSURE INSTRUMENTA11ON DAY SHIFT WEEK APPUCABILITY:

Modes I & 2 (Refer To P&L Step 3)

Rdn qthl L I Su0ance umwts 32.1.1.1(13). &3&11, 14101 LOCTlON:

103 (Notes 1 & 4) 3-9-88 3-9-85 3-9-84 3-9-83 Review Initials MAX D

C B

A 14AX Rooror*oo TLM OE DV MAX AS Et E9

()Note 4) 3-74A 3745 (AC) 3-PIS-3-22D 3-PlS-3-22C 3-PIS-3-22BB 3-PlS-2-22AX AC)

LIMIT SAT1UNSAT JO Unit Supvr Fday 0800 1035 1035 1035 1035 Siurday 0800 3urday 8808 Nonday 0800 62 psig (Note 3)

(Note 21 (Note 2)

(NoteS)

Tiesday 0800 Wecnesday 0800 Thursday 0800 (1)

These readings may be obtained from CS using the Single Value Display or from the ATU output oltage translated into a PRESSURE Signal for the specitic.instrumest&

For ICS type in SVD tsr Singje Value Display, enter the point desired as 374A, record reading, select F4, enter 3-74B, record the second reading.

(2) 3-74A and 3-745 have a Maximum allowable deviation of 40 psig, AND 3-PIS-3-22D, 3-PIS-3-22C, 3-PIS-3-22B8. & 3-PIS-3-22AA, i-ave a Maximum allowable deviation of O0psi. No coniporison is reqiired between the 3-74A10) snd 3-PlS-3-220(C)(Dfl)(AA).

(3) 3-74A and 3-745 SHALL be 1050 psig, 3-PlS-3-22D, 3PIS-3-22C, 3-PlS-3.-22E5, & 3-PIS-3-22AA SHALL be 1090 p819-(4) 3-74A and 3-745 are to be recorded at 0300, The Auxiliary Instrument Room reacings are not required to be taken at precisely 0800.

(5)

Following a change to Reactor Power and/or Plessure, verify the Steam Dome Limits are within the 0-Tl-248, Administrative Limits and Design Analysis Limits (Appendix S)

• • • • • • • • • • • • • • • • • • • ANSX,XIER J(jy *********************

Admin RO Aib Unit 3 BFN instrument Checks arid Observations 3-SR-2 Unit 3 Rev. 0068 Page 20 of 146 (Page 1 of 86)

Surveillance Procedure Data Package Modes 1, 2, & 3 ThIsweeI

,__Netwee 1ABLE 1.1 CORE THERMAL POWER AND CORE POWER DISTRIBUTiON DAY SHIFT PlPPLICAB1LflY Mode 1 when 25% RTP (Refer To P&L Step 3A)

RECORD the readings as soon as possible alter the generator brealcer has been closed.

nteria_Source:

321.1;_3221_323A:DEFINITIONS_SEC11ON_11_-FSAR 33.7 LOCATION:

I CS Computer (Case Sunuliary

- CSLJM)

Review Initials TIME coremernlal Percent Power LIMIT MFLCPR MAPRAT MFDLRX LIMIT Unit Unit DAY Note 2 Power (MWt

% RTP (AC Note 3 Note 3 Note 3 (AC)

Operator Supvr 0800 1UUJ)

U878 U7()

o.8c4 Initials 1000 1200 Fnday 1400 1600 1800 0800 1000 1200 Saturday 1400 1600 1800 Notes 1 Notes 0800 1000 1200 Sunday 1400 1600 1800 0800 1000 Monday 1800 1800 NOTES ARE FOLLOWING THE TABLEI

• • • • • • • • • • • • • • • • • • • ANSWER KEY *********************

Admin RO Aib Unit 3 BFN fristrumen Checks and Observations 3SR2 Unit 3 Rev. 0068 Page 22 of 146 (Page 3 of 86)

Surveillance Procedure Data Package - Modes 1,2. & 3 Compliance with the licensed Power limit LPL (3458 Mwt) is demonstrated by the following process:

A.

No actions are allowed that would intentionally raise core thermal power aboie 3458 Mwt for any period of time, Small, short-term fluctuations m power that are not under the direct control of the unit operator are not considered intentionaL B.

Closely monitor the thermal power during steady-state power operation with the goal of maintaining the two-hour average at or below 3455 Mwt.

If the core thermal power average for a 2-hour period is found to exceed 3458 Mwt, Operations take timely action to ensure that thermal power is less than or equal to 348 MINt. (This is irnpdemertted by taking action when any running average less than or equal 10 the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average exceeds 345B Mwt)

C.

The core thermal power for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period ( 8 hr average) is not to exceed 3458 Mvet D.

If an evolution is expected to cause a transient increase in reacto:r power that could exceed 3458 Mwt, action should be tairen to lower core power prior to performing the evolution, B.

IF power is > 3453, REDUCE power.

F.

IF power is 3458 to 345.3 MINt after allowing time for recent perturbations to settle, REDUCE power and EVALUATE the trend.

0.

IF arty running 30 mm Avg. 1 hr average, or2hr average is >3458 MINt, REDUCE power.

(2 Core Thermal Power is normally recorded every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when required. However, these readings may be marked NJA during TIP trace nina, control rod pattern adjustments, or anytime Core Jifonitoring System is blocked andfor <25% power. The Reactor Engineer is responsible for monitoring Core Thermal Limits. Monitoring of Core Thermal Power arid other Core Thermal Limits is recommended following conipletioni of planned rise in power arid following any unexpected power change. If core monitoring software becomea unavailable, the Unit Superv(sorJSRO and Reactor Engineer shall deteimmne the appropdate rrequertcy for monitoring Core Thermal Power but should not exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, using backup core monitoring.com pu.ter, and taking into consideration current core conditions and margin to thermal limits. Power changes should not normally be made without the core monitoring software being available.

A.

Consult Reactor Engineer when value 8.865.. Refer to 0-Tt-248 for Administrative Limits.

B.

Consult Reactor Engineer when value 8.83B. Refer to 8-Tl-248 for Adrninisirative Limmts.

C.

Consult Reactor Engineer when value (1.855. Refer to (1-TI-248 for Administrative Limits.

l:4 If any Turbine Bypass valve (a are inoperable or a Reciroulabon Loop is out of service, contact the Raotor Engineer and refer to the CO LR for Turbine Bypass Out of Service (TEoOS or Siri8le Loop Operation (SLO) limita which must be applied.

(5)

MAPRAT within limits is used to verrfy that all APL HORs are within the limits specified within the COLR. and 0.850, MFDLRX within limits is used so verify Ihat all Li-lOPs are within the limits specified within the COLR.

MPLCPR within limits is used so verify that all MCFRs are within the hrnits specified within the COLR, and t 0.80 when core thermal power is 80% RTP.

• • • • • • • • • • • • • • • • • • • ANSWER KEY ***********

Admin RO Aib Unit 3 BFN Instrument Checks and Observations 3-SR-2 Unit 3 Rev. 0068 Page29of 146 (Page 10 of 86)

Surveillance Procedure Data Package - Modes 1, 2, & 3 APPLICABILITY:

Mode 1 when 25% RTP(Refer To P&L Step 3.6A)

RECORD the readngs as soon as possible after the generator breaker has been closed.

Cntena Source:

8FPER951914 LOCATION:

ICS computer Review Initials ICS Points Verify HI and HI HI alarm setpoints listed in MAX DEV Table I.81 & I.8.2 are NOT exceeded.

(Note 3)

Unit 3-48A (CF) 3-488 (F) 3-50A (F) 3-508 (F NSSOQ1T (F)

Note 2 SAT! UNSAT! N)A UO Stipvr Fnday 3772 377.2 377.2 377.2 377.2 SAT I

Saturday Sunday Monday 2F Tuesday Wednesday Thursday The computer points listed in Table I B.

and I.8.2 are inputs to the IC Core Thermal Power Heat Balance calculations. The poinis are monitored to ensure the inputs are in agreement and to ensure the license limits for thermal power are maintained, in addition to the above. these points should be monitored any time reactor power changes are performed.

A difference between Faedwater temperature points 3-43A 3-48B, 3-50A, 3-5DB. and NSSOO17 of greater than 2 degrees will require the notification of Site Engineering and suspending any rise in power until tOe discrepancy is resolved.

An alarm setpoint being exceeded will require notifying the Unit Supervisor immediately and. faction cannot be taken immediately to return the value to within limits, Site Engineering will be notified for assistance.

TABLE 1.81 ICS POINT DESCRIPTION HI ALARM HI HI ALARM CALCO2O Rx Power 30 Mm Avg.

3458 3463 CALCO21 Rx Power 1 Hr. Avg.

3458 3461 CALCO83 Rx Power 2 Hr. Avg.

3458 3459 CALC098 Generator Power 118S 1190 CALCO2FJ Efficiency 35 3

CALCO27 Load Line NIA 113.6 CALCO24 RxPower%

1130.2 100.5 TABLE 1.B.2 CS POINT DESCRIPTION Hl ALARM HI HI ALARM 3-48A FWTemp 382 386 3-488 FW Temp 382 386 3-50A FW Temp 382 386 3-50B FW Temp 382 386 NSS&017 Avg. FW Temp..

382 386 CONSO400 Total RWCU Flow 0.168 NIA TABLE 1.6 HEAT BALANCE RELATED ICS ALARM SETPOINTS (Note 1)

DAY SHIFT WEEK:

Thi.s w&

to Next wee).

(1)

(2)

(3)

• • • • • • • • • • • • • • • • • • • ANSWER KEY ************

Admin RO Aib Unit 3 BFN Instrument Checks and Observations 3-SR-2 Unit 3 Rev. 0063 Page 46 of 146 Page27ot8)

Surveillance Procedure Data Packa9e - Modes 1, 2. & 3 APPUCIUTY:

Mod 1 & 2

&dins &r rquWd s a% times.

(Reter To P&L 3tep 38A)

Criri Sour Ththn)eal Rufranns MrsLl TSI 333 LOCAT0N:

Pane 3-14 and tCS Computer Rew nitis

1. LPRMs YPAS5ED (tote 1)

Total #

1. LPRMs LPRMe read!n MAX DEtI APRM LPRM APM LPRM APFM LPRM APRM LPRM Bypassd 3%crCS (AC Afi Data DAY TIME

1. 2

1. 2

1. 4

1. 4

1. 3

1. 3

1. 1

1. 1 (Not&*2)

(No 3)

(NOte 4)

SATJNSAT UO Und Supw Frklay 0800 0

0 0

0 0

0 0

0 Q

S.4T

!nitk/s

.aurthy 0800 Sunday 0800 Monday 0800 Tuocthy 0800 Wednesday 0800 Thurdsy 0800 (1)

Reor number of LPRMa bypassed in the tour APRM and LPRM eabinets an observed at Panel 3-9-1& arid these values together and reerd an Total # LPRMe Bpasee&

(2)

Lees than 20 LPRMs ir OPERATE or Lesa than 2 per level for any APRM will result in a Rod loo1c and a trouble alarm on the display panel Tels does not yield an automatto APRM trip, trut does, however, make the associated APRM INOP, (3)

Record number of LPRMs readin less than 3% on the LPRM printout or display on ICS.

(4)

MAX 0EV is net required to be met wien the APRMs are dale: however, unepeted inonsistenoles should be reported to the Reaetor Erineer. The total number of LPRMs bypassed shall slush the number of LPRMs readin lean than 3% on ICS.

TABLE 1.25 LPRM INSTRUMENTAIlON DAY SHIFT ThLcwee Netwee

• • • • • • • • • • • • • • • • • • • ANSWER KEY Admin RO Aib Unit 3 BFN Instrument Checks and Observations 3-SR-2 Unit a Rev. OO8 Page 52 of 148 (Page 33 of 86)

SurveHlance Procedure Data Package Modes 1, 2, & 3 Yswe.

Next weeJ.

TAELE 130 REACTOR VESSEL STEAM DOME PRESSURE INSTRUMENTATION DAY SKIFT WEEIC:

to AI-UILI I Y:

Modes 2 (1-teter 0 I-& step.OAI Readn are required at al times.

SuneUance Requirements:

3.3.1.1.1(13), 3&31i, 3.410,1 LOCATION:

CS (Notes I & 4) 3--38 3--s5 34m-84 34-53 Revmew Initials MX 0

0 0

A MAX Reference TIME 0EV JEV MAX Au Data

-e (Note 41 3-74A 3745 AC) 3PIS-3-220 3-PIS-3-22C

.3-S4-22SB P1S-3-22AX AC)

LIMIT SATIUNSAT UO Unit Smpmr Fiday 0-300 1034 1034 1035 1035 1035 1035 S4T (nt/a/s Saarday 0800 Smmndy 0800 Monday 0830 6] psig (Note 3)

(Nate 2)

(Note 2j (NoteS)

Tm,asthy 0300 3200 Thursday 0800 (1)

These readings may be obtained from CS using the Single Value Display or from the ATU output oltae translated irlto a PRESSURE Signal for the speciiic instrumeots.

For 105, type un SVD for Single Value Display, enter the point dssired as 3-74A, record reading, select F4. enter 3-74E, record the second reacing.

(2) 3-74A jftd 3-745 hu o Ma&iumutrm i1lowilAa UtUoIksi uf 40 piy, AND 3-PIS-3-22D, 3-PIS-3-22C. 3-.PIS-3-22B5. & 3-PS-3-22AA, e o Maimrum iIk,wia1Ae viutkm ur SC psig. No comparison is required between the 3-74A(B) and 3-PIS-3-22D(C)(8B)(AA).

(3) 3-74A and 3-745 SHALL be s 1050 psig. 3-PIS-3-22D, 3-PIS-3--22C, 3-PIS-3-22E5. & 3-PIS-3-22AA SHALL be 10 psg, (4) 3-74A and 3-745 are to be recorded at 0500. The AuiIiary lrtstnrnent Room reacings are not required to be taken at precisely 0500 (5)

FolloMng a change to Reactor Power and/or Pressure, verity the Steam Dome Limits are thin ths 0-Tl-248, Admrrtistrative Limits and Design Analysis Unilts (Appendiu 5)

Adni+/-n RO A4 PAGE 1

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OPERATOR:

RO SRO DATE:

JPM NUMBER:

Admin RO A4 TASK NUMBER:

U-000-EM-87 TASK TITLE:

EPIP-3, Appendix B, Unit Operator Notification K/A NUMBER: 2.4.43 K/A RATING: RO 3.2 TASK STANDARD: Completion of Emergency Call-out for TSC Manning LOCATION OF PERFORMANCE:

Simulator REFERENCES/PROCEDURES NEEDED: EPIP 3 VALIDATION TIME: 15 minutes MAX. TIME ALLOWED: (Completed for Time Critical JPMs only)

PERFORMANCE TIME:

COMMENTS:

Additional comment sheets attached? YES NO RESULTS:

SATISFACTORY___

UNSATISFACTORY SIGNATURE:

DATE:

EXAMINER

IMTIAL CONDITIONS:

You are the Unit 1 Operator. Unit 2 was operating at 100% (BOL) when indications of a primary system leak into the Drywell developed. Conditions have continued to the point that the SED has declared an ALERT.

IMTIATING CUES:

The SHIFT MANAGER has informed you that Unit 2 is in an ALERT status. The SHIFT MANAGER/SED directs you to COMPLETE APPENDIX B, Unit Operator NOTIFICATIONS.

Admin RO A4 PAGE 3

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Class Room INITIAL CONDITIONS:

You are the Unit 1 Operator. Unit 2 was operating at 100% (BOL) when indications of a primary system leak into the Drywell developed. Conditions have continued to the point that the SED has declared an ALERT.

INITIATING CUES:

The SHIFT MANAGER has informed you that Unit 2 is in an ALERT status. The SHIFT MANAGERJSED directs you to COMPLETE APPENDIX B, Unit Operator NOTIFICATIONS.

Adniin RO A4 PAGE 4

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START TIME____

Performance Step 1:

Critical Not Critical NOTE The Emergency Paging System (EPS) consists of a dedicated touch screen CRT.

Activation of any screen feature requires the user place their fingertip within the boundary of the select button and leave it there for at least 1 second. The CRT Screen will normally display a large rectangle that indicates that the paging system is available but currently inactive.

If the EPS fails to operate, contact the SM/SED immediately. Request that the ODS be contacted to initiate the system from his location.

1.0 Activate the Emergency Paging System (EPS) 1.1 PRESS the EPS CRT screen once to activate the paging options Standard:

OPERATOR activates the Emergency Paging System using the touch screen.

SAT UNSAT N/A COMMENTS:

Performance Step 2:

Critical Not Critical 1.2 PRESS the appropriate option as instructed by the SED

• PAGER TEST

• DRILL

• EMERGENCY

• STAGING AREA

• ABORT Standard:

Operator presses either DRILL or EMERGENCY.

SAT UNSAT N/A COMMENTS:

Adrnin RO A4 PAGE 5

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Performance Step 3:

Critical X Not Critical 1.3 PRESS the START button to initiate the option OR PRESS the ABORT button to deny the option request Standard:

Presses the START button.

SAT UNSAT N/A COMMENTS:

Performance Step 4:

Critical Not Critical X 1.4 IF the EPS fails to operate locally TFIEN CONTACT the ODS at 5-751-1700 or 5-751-2495 AND DIRECT the ODS to activate the EPS for BFN.

Standard:

NA SAT UNSAT N/A COMMENTS:

Performance Step 5:

Critical Not Critical 1.5 WHEN the EPS FAILS to operate either locally or by the ODS THEN exit this step and re-enter this Appendix at Step 2.0 Otherwise continue in this procedure.

Standard:

NA SAT UNSAT N/A COMMENTS:

Admin RO A4 PAGE 6

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Performance Step 6:

Critical Not Critical 1.6. MONITOR the Paging System Terminal Display NOTE Monitor ERO positions through OSC Document Control. Positions below OSC Document Control are courtesy pages and are not subject to call-out requirements.

1.6.1 IF A NO response is observed, OR The position being paged has not responded promptly or within approximately 20 minutes, THEN UTILIZE the Weekly Duty List and attempt to contact the position representative with available information. (No Fitness for Duty question is required.)

THEN UTILIZE the Weekly Duty List and attempt to contact the position representative with available information. (No Fitness for Duty question is required.)

Standard:

OPERATOR utilizes the weekly duty list and calls the Operations Manager, NO Fitness For Duty question is required and acknowledges a YES from the Operations Manager.

SAT UNSAT N/A COMMENTS:

I CUE:

TWENTY Minutes has elapsed and the Operations Manager has NOT responded I Driver:

When called as the Operations Manager respond YES to tIlling the Operations Manager position.

END OF TASK STOP TIME

Adinin SRO Aib PAGE 1

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OPERATOR:

RO SRO__

DATE:_______

JPM NUMBER:

Admin SRO Aib TASK NUMBER:

Conduct of Operations TASK TITLE:

NRC Event Notification K/A NUMBER: 2.1.18 K/A RATING: SRO 3.8 TASK STANDARD: Determine NRC Event Notification requirements and Technical Specification actions required.

LOCATION OF PERFORMANCE:

Class Room REFERENCES/PROCEDURES NEEDED: NPG-SPP-03.5 VALIDATION TIME: 10 minutes MAX. TIME ALLOWED: 60 Minutes PERFORMANCE TIME:

COMMENTS:

Additional comment sheets attached? YES NO RESULTS:

SATISFACTORY___

UNSATISFACTORY___

SIGNATURE:

DATE:

EXAMINER

IMTIAL CONI)ITIONS: Unit 3 is performing a normal Reactor Startup in accordance with 3-GOI-100-1A. Reactor Power is 5%, Reactor pressure is 980 psig, and core flow is 9.5% of rated core flow. A feedwater transient causes a spike in Reactor Power and an automatic Scram occurs.

The Reactor Engineer informs you that Reactor Power peaked at 28% RTP.

INITIATING CUE: As the Shift Manager, evaluate these plant conditions for appropriate notifications and technical specification actions, if any, and document your determinations on the correct form(s), if necessary.

Adinin SRO Aib PAGE 3

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Class Room INITIAL CONDITIONS: Unit 3 is performing a normal Reactor Startup in accordance with 3-GOI-100-1A. Reactor Power is 5%, Reactor pressure is 980 psig, and core flow is 9.5% of rated core flow. A feedwater transient causes a spike in Reactor Power and autnatic-Sccam-occiirs The ReaetorEngineerinforms-you-that Reactor Power peaked at 28% RTP.

INITIATING CUE: As the Shift Manager, evaluate these plant conditions for appropriate notifications and technical specification actions, if any, and document your determinations on the correct form(s), if necessary.

START TIME Admin SRO Aib PAGE 4

OF 6

Performance Step 1:

Critical X Not Critical Refers to Technical Specification Section 2.0 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow

<10% rated core flow:

2.2 SL Violations THERMAL POWER shall be 25% RTP.

With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

Standard:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

Determines that a safety limit was exceeded or violated, determines 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore compliance and insert all rods. Both actions are met and complete SAT_ UNSAT_ N/A COMMENTS:

9

Adinin SRO Aib PAGE 5

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Performance Step 2:

Critical X Not Critical Evaluates NPG-SPP-03.5 Appendix A: 3.1.B 1 B.

The following criteria require 1-hour notification:

1.

(Technical Specifications)

- Safety Limits as defined by the Technical Specifications which have been violated.

Standard:

Determines a 1 -Hr Non-Emergency notification is required.

SAT_ UNSAT N/A _COMMENTS:______________________________

Performance Step 3:

Critical Not Critical Complete Attachment 1, NPG-SPP-3.5-l

- NRC Event Notification Worksheet Standard:

Under Event Classification a check in box for 50.72 Non-Emergency SAT_ UNSAT N/A COMMENTS:______________________________

Performance Step 4:

Critical Not Critical Complete Attachment 1, NPG-SPP-3.5-1

- NRC Event Notification Worksheet Standard:

Under 1-Hr Non-Emergency 10CFR5O.72(b)(1) a check in the box TS Deviation ADEV SAT_ UNSAT N/A COMMENTS:_______________________________

Admin SRO Aib PAGE 6

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Performance Step 5:

Critical Not Critical X Complete NPG-SPP-3.5-1

- NRC Event Notification Worksheet Standard:

Under 4-Hr Non-Emergency 10CFR5O.72(b)(2) a check in box (i) TS Required S/D ASHU SAT UNSAT N/A COMMENTS:_____________________________

Performance Step 6:

Critical Not Critical Complete NPG-SPP-3.5-1

- NRC Event Notification Worksheet Standard:

Power/Mode Before will be 5%/Mode 2; Power/Mode After will be Shutdown/Mode 3 and a brief description stating that a Technical Specification Safety limit was violated.

SAT_ UNSAT N/A COMMENTS:_________________________

UE: JPM complete once an entry is made in description block on first page, Additional Information page not required to be completed.

END OF TASK STOP TIME

Admin RO/SRO A2 PAGE 1

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OPERATOR:

RO SRO DATE:________

JPM NUMBER:

Admin RO/SRO A2 TASK NUMBER:

U-066-NO-02 TITLE:

Evaluate Recombiner Performance K/A NUMBER:

2.2.44 K/A RATING: RO 4.2 SRO 4.4 TASK STANDARD: Evaluate Off-Gas Recombiner Performance. Determine that it meets Acceptance Criteria.

LOCATION OF PERFORMANCE: Classroom REFERENCES/PROCEDURES NEEDED:

3-01-66 VALIDATION TIME: 10 minutes MAX. TIME ALLOWED:

PERFORMANCE TIME:

COMMENTS:

Additional comment sheets attached? YES NO RESULTS:

SATISFACTORY___

UNSATISFACTORY SIGNATURE:

DATE:

EXAMINER

INITIAL CONDITIONS: You are a Unit Operator, A startup,,iin progress on Unit 3 and reactor power has been raised to 98% rated thermal ater Chemistry System is In) service in accordance with 3-01-4, Hydrogen Water Chemistry System. The Off-Gas Preheater, Recombiner and SJAEs are in operation in accordance with 3-01-66, Off-Gas System, Section 5.0. The in-service steam jet is operating properly.

j INITIATING CUE: The Shift Manager has directed you to evaluate Off-Gas Recombiner 3A performance, in accordance with, 3-01-66, Off-Gas System, Section 6.1 Ding.theea1natl if any the procedure-anddeterminetherequired actior totake.

z / j Conditions are as follows:

/7/

3-TI-66-75A 392 °F 3-TI-66-75B 320 °F 3-TRS-66-77A Center temp 612 °F 3-TRS-66-77B Center temp 380 °F xPowerThermaL_) t4 t-t 3402 MWth

c 3-H2R-66-96 operable

- both pens reading.26% H2

)

)JI) t;t Vsv7 ((/ (-i-)

Adrain RO/SRO A2 PAGE 3

OF 7

Classroom INITIAL CONDITIONS: You are a Unit Operator, A startup is in progress on Unit 3 and reactor power has been raised to 98% rated thermal power. The Hydrogen Water Chemistry System is in service in accordance with 3-01-4, Hydrogen Water Chemistry System. The Off-Gas Preheater, Recombiner and SJAEs are in operation in accordance with 3-01-66, Off-Gas System, Section 5.0. The in-service steam jet is operating properly.

INITIATING CUE: The Shift Manager has directed you to evaluate Off-Gas Recombiner 3A performance, in accordance with, 3-01-66, Off-Gas System, Section 6.1 During the evaluation, if any Acceptance Criteria are not met, continue in the procedure and determine the required action to take.

Conditions are as follows:

3-TI-66-75B 3-TRS-66-77A Center temp 3-TRS-66-77B Center temp 3-TI-66-75A 392 °F 3-H2R-66-96 operable

- both pens reading.26% H2 Rx Power Thermal 320°F 612 °F 380°F 3402 MWth

Adxn+/-n RO/SRO A2 PAGE 4

OF 7

START TIME____

Performance Step 1:

Critical Not Critical 6.1 Recombiner Performance Evaluation NOTES 1)

The production of hydrogen and oxygen in the reactor is dependent upon reactor power level and upon the amount of hydrogen injected by the Hydrogen Water Chemistry System if in service. Since the recombination of hydrogen and oxygen is exothermic, the operating temperature of the recombiner is also dependent upon power level and the status of the HWC System.

2)

Following startup, while still at low power, recombiner performance and hydrogen concentration should be closely monitored.

[1]

Perform a Recombiner Performance Evaluation as follows:

[1.11 DETERMINE the in-service recombiner inlet temperature as indicated on RECOMBINER 3A(3B), INLET TEMP 3-TI-66-75A(B), Panel 3-9-53.

Standard:

Determines Recombiner 3A inlet temp 3-TI-66-75A, Panel 3-9-53 (from handout)

SAT_ UNSAT N/A _COMMENTS:______________________________

Performance Step 2:

Critical Not Critical X

[1.2]

DETERMINE the in-service recombiner operating (center) temperature as indicated on RECOMBINER 3A13B TEMPERATURE recorder, 3-TRS 77, Panel 3-9-53.

Standard:

I Determines the in-service recombiner operating (center) temperature as indicated on Recombiner 3A temperature recorder, 3-TRS-66-77, Panel 3-9-53 (from handout)

SAT_ UNSAT N/A _COMMENTS:______________________________

Admin RO/SRO A2 PAGE 5

OF 7

Performance Step 3:

Critical Not Critical

[1.3]

CALCULATE the temperature difference (AT) between the values obtained in Steps 6.1[1J and 6.1[2].

Standard:

Calculates Recombiner 3A inlet/center At and determines At is 220 °F SAT_ UNSAT N/A COMMENTS:_____________________________

Performance Step 4:

Critical Not Critical X

[1.4]

DETERMINE the reactor thermal power (MWt) from process computer.

Standard:

wz Determines reactor thermal power from the handout SAT_ UNSAT N/A COMMENTS:_____________________________

Performance Step 5:

Critical Not Critical

[1.5]

USING Illustration 1, PLOT the corresponding point of reactor power in MWt and AT.

I Standard Using illustration 1, Determines At corresponding to 3402 MWT is 187.11 °F.

pgA 7

L SAT UNSAT N/A COMMENTS:______________________________

V

Admin RO/SRO A2 PAGE 6

OF 7

Performance Step 6:

Critical Not Critical

[1.61 VERIFY point on illustration 1 is above or equal to the appropriate line (HWC in service or HWC out of service)

Standard:

Determines from Illustration 1 that calculated z\\t vs MWt plots ABOVE the HWC in Service line.

SAT UNSAT N/A COMMENTS:_____________________________

Performance Step 7:

Critical Not Critical

[2]

IF the in-service recombiner performance is below the minimum allowable, THEN:

[2.11 CHECK Off-Gas Preheater, Recombiner and SJAEs are in operation in accordance with Section 5.0.

Standard:

NA Recombiner performance is satisfactory SAT_ UNSAT N/A COMMENTS:______

END OF TASK STOP TIME

Admin RO/SRO A2 PAGE 7

OF 7

Illustration I (Page 1 of I)

Recombine Performance Evaluation

- AT to Reactor Power 0

500 1000 1500 3000 3500 Evaluation is satisfactory when intersection point of AT to Reactor Power is above the appropriate line.

For 3458mwt HWC in service HWC out of service AT 190°F AT 242°F CURVE FACTORS Normal Water Chemistry (NWC)

Hydrogen Water Chemistry (F-IWC)

AT = 0.070°F per MWt AT = 0.055°F per MWt 250 200 150 100 nperature 0

MWt HWC Out of Service - -

HWC In Servic

Adrnin SRO A3 PAGE 1

OF 6

OPERATOR:

SRO___

DATE:

JPM NUMBER:

Admin SRO A3 TASK NUMBER:

Radiation Control TASK TITLE:

Radiation Exposure Limits under Emergency Conditions K/A NUMBER: 2.3.4 K/A RATING: SRO 3.7 TASK STANDARD: Determine stay time for an AUO to perform an emergency evolution due to radiation levels and authorize.

LOCATION OF PERFORMANCE:

Class Room REFERENCES/PROCEDURES NEEDED: EPW 15 VALIDATION TIME: 15 minutes MAX. TIME ALLOWED:

PERFORMANCE TIME:

COMMENTS:

Additional comment sheets attached? YES NO RESULTS:

SATISFACTORY___

UNSATISFACTORY SIGNATURE:

DATE:

EXAMINER

INITIAL CONDITIONS:

Unit 2 is in a General Emergency. No facilities are currently activated and Site Emergency Director duties remain in the Control Room. Due to multiple equipment failures the Turbine is coasting down at 1200 RPMs with no Lube Oil Pressure. An AUO has volunteered to replace blown fuses in the EBOP DC control cabinet next to the MTOT on EL 586 in the Turbine Building. Radiation Protection Supervision states that the dose rate at the cabinet is 15 REM/hr and travel path dose rates are 10 REM/hr to and from the cabinet. It is estimated that it will take him 12 minutes of total travel time to and from the cabinet. The AUO has received 500 mrem TEDE, year to date, and he has been briefed of the radiological hazards associated with this evolution per appendix A of the applicable EPIP.

INITIATING CUE:

As the Shift Manager determine how much time the AUO has to replace the fuses, in minutes, without exceeding Emergency Dose Limits and authorize the emergency exposure.

Adinin SRO A3 PAGE 3

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Class Room INITIAL CONDITIONS:

Unit 2 is in a General Emergency. No facilities are currently activated and Site Emergency Director duties remain in the Control Room. Due to multiple equipment failures the Turbine is coasting down at 1200 RPMs with no Lube Oil Pressure. An AUO has volunteered to replace blown fuses in the EBOP DC control cabinet next to the MTOT on EL 586 in the Turbine Building. Radiation Protection Supervision states that the dose rate at the cabinet is 15 REM/hr and travel path dose rates are 10 REM/hr to and from the cabinet. It is estimated that it will take him 12 minutes of total travel time to and from the cabinet. The AUO has received 500 mrem TEDE, year to date, and he has been briefed of the radiological hazards associated with this evolution per appendix A of the applicable EPIP.

INITIATING CUE:

As the Shift Manager determine how much time the AUO has to replace the fuses, in minutes, without exceeding Emergency Dose Limits and authorize the emergency exposure.

Adrn+/-n SRO A3 PAGE 4

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START TIME____

Performance Step 1:

Critical Not Critical Determine the radiation dose that he may receive to protect valuable property Standard:

Determines he may receive 10 REM to protect valuable property SAT UNSAT N/A COMMENTS:

Performance Step 2:

Critical Not Critical Determine the radiation dose that AUO may receive without exceeding 10 REM, based on previous exposure Standard:

Determines that the AUO may receive 9.5 REM, based on previous exposure of 500 mrem TEDE, year to date, to protect valuable property SAT UNSAT N/A COMMENTS:

NRC Information: Reference is EPIP 15 Section 3.4.3 CUE: Provide EPIP 15 Appendix B form which is partially completed

Adin+/-n SRO A3 PAGE 5

OF 6

Performance Step 3:

Critical Not Critical Determines the amount of time the AUO has to replace the blown fuses without exceeding the 10 REM Emergency Exposure limit to protect valuable property, based on previous exposure and travel time Standard:

Determines that the AUO has 30 minutes to replace the blown fuses, without exceeding Emergency Exposure Limit of 10 REM (Including travel time and previous exposure), If calculated in hours

- 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> SAT UNSAT N/A COMMENTS:

Performance Step 4:

Critical X Not Critical Completes Acknowledgment and Authorization to Exceed Occupational Dose Limits form Appendix B of EPIP 15 Standard:

Determines that as the Shift Manager and acting Site Emergency Director he can authorize the Emergency Dose SAT UNSAT N/A COMMENTS:

NOTE: Critical Data on form is the authorized 10 Rem and Approval signature END OF TASK STOP TIME

Adrnin SRO A3 PAGE 6

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APPENDIX B Page 1 of 1 ACKNOWLEDGMENT AND AUTHORIZATION TO EXCEED OCCUPATIONAL DOSE LIMITS READ THE FOLLOWING STATEMENT BEFORE SIGNING THIS FORM:

I acknowledge by signature on this form that I am volunteering for exposures in excess of 10 CFR 20.1201 limits and that I have been made aware through training or a briefing of the risks involved. Briefing material was presented from Appendix A of this procedure.

The persons listed below have acknowledged and volunteered to receive dose limits in excess of 1 OCFR2O. 1201 limits. Authorization is required by the Site Emergency Director to administer any emergency exposure limit. Authorization is acknowledged by Site Emergency Director signature on the bottom of this form.

Name Employee ldenttfication Signatme Doje Liimt (Please print Last. First. Ml)

Number (FIN)

(Rem)

AIJO that volunteered 123-45-6789 BriefDescription ofTask:

Authorized by:

/

Site Emergency Director TimeDate LAST PAGE

Adinin SRO A4 PAGE 1

OF 5

OPERATOR:

SRO___

DATE:

JPM NUMBER:

Admin SRO A4 TASK NUMBER:

S-000-EM-21 (SRO ONLY)

TITLE:

Classify the event per REP (Fuel Damage with RCIC Steam Leak)

K/A NUMBER:

2.4.41 K/A RATING: SRO 4.6 TASK STANDARD: The event is classified as a Site Area Emergency, EAL Designator 2.3-Si and the Initial Notification appendix is completed with the correct information. Event is classified within 15 minutes and Initial Notification is completed within 15 minutes of classification with correct Protective Action Recommendation.

LOCATION OF PERFORMANCE: Simulator or Classroom REFERENCES/PROCEDURES NEEDED:

EPIP 1, EPIP 4 VALIDATION TIME: 30 minutes MAX. TIME ALLOWED: 15 minutes to classify and 15 minutes to notify PERFORMANCE TIME:

COMMENTS:

Additional comment sheets attached? YES NO RESULTS:

SATISFACTORY___

UNSATISFACTORY SIGNATURE:

DATE:

EXAMINER

INITIAL CONDITIONS: You are the SHIFT MANAGER. Unit 2 was operating at 80% power performing a Control Rod Pattern Adjustment. During the Control Rod Pattern Adjustment, Control Rod 38-23 dropped several notches and-was-bsequenldriverrirtposit4on-99 Chemistry sampling indicated 300 jiCilgm dose equivalent Iodine-131 inut ago. Unit shutdown was in progress when a RCIC steam leak developed.

A Reactor SCRAM was inserted and the following conditions exist:

The TSC, OSC and CECC are not staffed at this time.

Reactor Level 10 inches and slowly rising Reactor Pressure 950 psig and stable DW Pressure 1.35 psig and stable DW Temperature 135°F and stable DW Radiation Panel 2-9-7C, Window 15, Drywell/Suppression Chamber Radiation High, is in alarm and 2-RM-90-272A is reading 2900 R/Hr PCIS Isolation Group 5 Is NOT complete, RCIC Valves 71-2 and 71-3 failed to auto close, RCIC Valve 71-2 was closed with Control Room operator action.

RCIC Area TE-71-41A indicates 195°F and rising RCIC Room 90-26A 600 mrem!hr and rising Wind Speed 10 mph Wind Direction 200° Projected TEDE at site boundary

<1 REM Projected Thyroid CDE at site boundary

<5 REM INITIATING CUE Classify the event and complete initial notification form. /

JPM is Time Critical

Admin SRO A4 PAGE 3

OF 5

Classroom INITIAL CONDITIONS: You are the SHIFT MANAGER. Unit 2 was operating at 80% power performing a Control Rod Pattern Adjustment. During the Control Rod Pattern Adjustment, Control Rod 3 8-23 dropped several notches and was subsequently driven in to position 00.

Chemistry sampling indicated 300 iCi/gm dose equivalent Iodine-i 31. An Alert, EAL 1.3-A, was declared 30 minutes ago. Unit shutdown was in progress when a RCIC steam leak developed.

A Reactor SCRAM was inserted and the following conditions exist:

The TSC, OSC and CECC are not staffed at this time.

Reactor Level 10 inches and slowly rising Reactor Pressure 950 psig and stable DW Pressure 1.35 psig and stable DW Temperature 135°F and stable DW Radiation Panel 2-9-7C, Window 15, Drywell/Suppression Chamber Radiation High, is in alarm and 2-RM-90-272A is reading 2900 R/Hr PCIS Isolation Group 5 Is NOT complete, RCIC Valves 71-2 and 71-3 failed to auto close, RCIC Valve 71-2 was closed with Control Room operator action.

RCIC Area TE-71-41A indicates 195°F and rising RCIC Room 90-26A 600 mrem/hr and rising Wind Speed 10 mph Wind Direction 200° Projected TEDE at site boundary

<1 REM Projected Thyroid CDE at site boundary

<5 REM INITIATING CUE:

Classify the event and complete initial notification form.

1PM is Time Critical

Adrn+/-n SRO A4 PAGE 4

OF 5

START TIME____

Performance Step 1:

Critical X Not Critical Refers to EPIP 1 to classify emergency event.

Standard:

SHIFT MANAGER refers to EPIP land declares a Site Area Emergency, EAL 2.3-Si, based on Drywell Radiation reading greater than 2263 on 2-RE-90-272A SAT UNSAT N/A COMMENTS:______________________________

TIME Classified Performance Step 2:

Critical Not Critical Implements EPIP-4 Site Area Emergency.

Standard:

SHIFT MANAGER recognizes/implements Site Area Emergency JAW EPIP-4.

SAT_ UNSAT N/A COMMENTS:______________________________

Adnün SRO A4 PAGE 5

OF 5

START TIME____

Performance Step 3:

Critical Not Critical Completes Appendix A of EP1P 4 Standard:

Shift Manager completes Appendix A of EPIP 4 within 15 minutes of event classification SAT_ UNSAT N/A _COMMENTS:_____________________________

TIME Appendix A Complete Performance Step 4:

Critical Not Critical Completes Appendix A of EPIP 4 Standard:

Following are Critical portions of Appendix A: EAL Designator 2.3-Si, Unit 2 is checked, Time and Date Event declared, and PAR recommendation NONE is checked.

SAT_ UNSAT_ N/A COMMENTS:_____________________________

END OF TASK