ML100620150
| ML100620150 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 02/09/2010 |
| From: | NRC/RGN-II |
| To: | |
| References | |
| 50-335/09-301, 50-389/09-301 | |
| Download: ML100620150 (100) | |
Text
{{#Wiki_filter:FACILITY FACI LlTY NAME: NAM E: _ _.;:;.St;::., S.;:;.t;;.*;. ..=L.=..uc=.:,i.=..e
-=L~uc;:;.;;.ie~ __ _ _ _ _ _ _ Section 6 REPORTNUMBER: __
REPORTNUMBER:_~20~O~9~-3~O~1
~2~OO~9~-3~O~1~ ________________________
DRAFT ADMINISTRATIVE JPMS CONTENTS: o D Draft ADMIN JPMs ( Location of Electronic Files: h A' A" '~
~::=-:::,
Submitted By: /~~
/~d',6 Verified By: _
ES-301 Administrative Topics Outline Form ES-301-1 Facility: St. Lucie Plant Date of Examination: 10/19/09 Examination Level: RO ~ [gJ SRO ~ lM Operating Test Number: HLC-19 NRC Administrative Topic Type Describe activity to be performed (see Note) Code* M Perform an Estimated Critical Condition Calculation - Unit 2 Conduct of Operations M Perform a Manual Calorimetric at 45% power. Conduct of Operations N Develop ECO for Unit 1 Containment Spray Pump. Equipment Control ( N Perform Immediate Operator Actions For a Dropped New Fuel Element (SRO) Radiation Control D Operate Portable Radiation Monitoring Equipment (RO) M,P Determine PARs During Core Melt Scenario (SRO) Emergency Procedures/Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom
(:::; 3 for ROs; :s (D)irect from bank (~ :::; 4 for SROs & RO retakes) (N)ew or (M)odified from bank (~ (?: 1) (P)revious 2 exams (:s (:::; 1; randomly selected) (
ES-301 Administrative Topics Outline Form ES-301-1 ADMINISTRATIVE JPM
SUMMARY
DESCRIPTION ( CONDUCT OF OPERATION Perform a Unit 2 Estimated Critical Condition Calculation 2-GOP-302, 'Reactor Plant Startup - Mode 3 to Mode 2' Appendix C 'Estimated Critical Conditions'. This may be performed on the Simulator or in the classroom. CONDUCT OF OPERATION Perform a Unit 2 Manual Calorimetric at 45% power. 2-3200020, 'Primary System Manual Calorimetric' will be used. This may be performed on the simulator or in the classroom. EQUIPMENT CONTROL Develop a clearance for the 1B Containment Spray Pump. The pump cooler is to be removed from service due to a leak. RADIATION CONTROL (SRO) Perform Operator Actions for a Dropped New Fuel Element. During offloading of new fuel in the new fuel area a dropped fuel element occurs. Without reference to procedures, state the Immediate Operator actions that must be carried out. Using 2-NOP-26.01 'Radiation Monitors', obtain Main Steam Line radiation readings using (RO) USing the portable RM-23 in the field from the RM-80 enclosure. EMERGENCY PROCEDURES/PLAN Determine PAR's During Core Melt Scenario (Modified from 2008 NRC exam)
ST LUCIE INITIAL LICENSE EXAMINATION JOB PERFORMANCE MEASURE NRC JPM A-1 ADM Calculate Estimated Critical Conditions Unit 2 ( CANDIDATE EXAMINER NRC ADM JPM COO ECC REV. a REV, 0 Page 1 of 11
TASK: Perform an Estimated Critical Conditions Calculation ( K/A 2.1.25 (3.9 I 4.2) Ability to interpret reference materials, such as graphs, curves, tables, etc. POSITION: RO I SRO Location: Simulator I Classroom Method: Perform Estimated JPM Completion Time: 25 Minutes Actual JPM Completion Time: Minutes The JPM Operator's performance was evaluated against the standards of this JPM and is determined to be: SATISFACTORY/UNSATISFACTORY (circle one) ( Evaluator's Signature _ _ ___ __ ____ __ __ __ __ __ ___ Date / /
References:
- 2-GOP-302 Reactor Plant Startup - Mode 3 to Mode 2, Appendix C.
- Unit 2 Plant Physics Curves Cycle 18 NRC ADM JPM COO ECC REV. 0 Page 2 of 11
INITIAL CONDITIONS Unit 2 has been operating for 118 days at 100% power. The unit was shut down at 1800 on Sept. 29, 2009, due to Main Feedwater problems. We anticipate reactor startup #09-2 to reach criticality at 1500 on Sept. 30, 2009. The following conditions exist:
- Reactor core exposure is 2800 EFPH.
- RCS Boron concentration is 1200 ppm.
- A JPN Engineering Evaluation is unavailable for this startup.
The Shift Manager instructs you to perform a manual determination of Estimated Critical Conditions by completing Appendix C (Estimated Critical Conditions) of 2-GOP-302 (Reactor Plant Startup - Mode 3 to Mode 2) in preparation for the unit startup_ startup. JPM OVERALL STANDARD: Calculate required Estimated Critical Conditions using Appendix C (Estimated Critical Conditions) of 2-GOP-302 (Reactor Plant Startup - Mode 3 to Mode 2). NOTES: This JPM may be performed in the Simulator or the classroom. Physics Curves for Unit 2 cycle 18 are required as references. Last last Stable RCS Boron and Criticality Data LJ RC System Boron was 948 ppm at 1300 on Sept 29, 2009. o Reactor Criticality Data from RCO Log dated 1300 Sept 29, 2009.
- CEA Group/Position (inches) 5 / 136
- T-avg T -avg (OF) 574
- Reactor Power (%) 100 o Reactivity Deviation (PCM) o NRC ADM JPM COO ECC REV.
REV, 0 Page 3 of 11
START TIME _ _ __ Appendix C Estimated Critical Conditions and Inverse Count Rate Ratio STEPS ELEMENTS STANDARD StU S/U COMMENTS REQUIRED FOR UNSAT Record required data 1.A Unit - Operator records Unit 2 Operator records Startup 1.B S/U StU -- 09-2. 1.C Date of ECC Operator records: Sept. 30,2009 1.0 Record the Burnup: Operator determines this EFPH ( based on initial conditions and records 2800 EFPH Operator records date: Record shutdown and 9/29/09 at 1800 2.A.1.a startup times: Date/time of shutdown
/
2.A.2.a Anticipated date/time of Operator records date: criticality 9/30/09 at 1500.
/
Operator notes that NC Obtain last stable Boron System Boron sample sample critical conditions was within 24 hours of the before shutdown. shutdown. NRC ADM JPM COO ECC REV. 0 Page 4 of 11
STEPS ELEMENTS STANDARD S/U Stu COMMENTS REQUIRED FOR UNSAT COLUMN 1 CONTINUED Record the power defect Operator determines this from PPCS PPCB Figure A.1A1 based on initial conditions and PPCS: PPCB: + 1335 pcm pem (1300 to 1370) Record the xenon worth Operator determines this PPCB Figure A.4 from PPCS AA based on initial conditions PPCB: and PPCS: + 2550
- pcm pem (2525 to 2600)
Record the samarium and neptunium worth from Operator determines this PPCS A5 PPCB Figure A.5 based on initial conditions
- PPCB:
and PPCS: + 806 pcm pem ( (805 to 808) Record the diff. boron Operator determines this worth from PPCS, PPCB, Figure based on data provided A8 A.8 and PPCS: PPCB:
+ 8.18 pem/ppm pcm/ppm
* (7.95 to 8.2)
Record prior to shutdown Operator determines this RCS Soron Boron concentration based on initial conditions
- ppm + 948 ppm
- Calculate the boron worth: Operator calculates boron
_ _ ppm Soron Boron x __ __ worth: pcm/ppm Soron Boron = 948 ppm Soron Boron x 8.18 pcm pcm/ppm = + 7755 pem. (7749 to 7773)
- DENOTES CRITICAL NRC ADM JPM COO ECC REV. o Page 5 of 11
STEPS ELEMENTS STANDARD S/U COMMENTS REQUIRED FOR UNSAT
- Record the CEA reactivity Operator determines this worth, Figure A.6 or A. 7 based on data, Figure A.6: + 8580 pcm pcm (NO band since pcm given on curve) 2a2 FILL IN COLUMN 2:
Record the xenon worth Operator determines this from PPCB Figure AA based on initial conditions and PPCB: + 2460 pcm pcm (2500 to 2550) ( Operator determines this Record the samarium and based on initial conditions neptunium worth from and PPCB: + 826
- PPCB Figure A.5 pcm pcm (825 to 827)
Record the diff. boron worth from Column 1. Operator records: pcm/ppm + 8.18 pcm/ppm (8.1 to 8.2) Record current RCS Boron Concentration Operator determines this based on initial
- ppm conditions:
+ 1200 ppm Calculate the boron worth:
_ _ ppm Boron x _ __ _ Operator calculates boron ( pcm/ppm Boron = worth: pcm 1200 ppm Boron x 8.18 pcm/ppm = + 9816 pcm.
- DENOTES CRITICAL NRC ADM JPM COO ECC REV. 0 Page 6 of 11
STEPS ELEMENTS STANDARD S/U COMMENTS REQUIRED FOR UNSAT
* (9720 to 9840)
Determine whether to go critical at normal point of Operator determines that 60 inches on group 5. the normal critical rod position is desired. Record the CEA reactivity PPCS, Figure worth from PPCB, Operator determines this A.6 or A.7 A.60rA.7 based on data provided PPCS, Figure A.6: and PPCB, pcm Group 5 @ 60"
+ 8289 pcm Enters Reactivity (8275 to 8300)
- Deviation Enters 0 given in cue sheet.
2a2b Verify that the critical rod Operator verifies that 60 position is above the inches on Group 5 is
- insertion limits and initial above 55 inches on group Note 2 of the worksheet. 3 and initials Note 2 on the work sheet.
2a2c Calculate the CEA Operator calculates CEA position at +500 pcm and positions at +500, -500 & -
-500 pcm from the 1000 pcm from the esti-estimated critical position: mated critical position.
+500 pcm _ in on Grp
=
8789pcm = ARO 7789pcm = = 111 in on Group 3
-500 pcm _ in on Grp (104to (104 to 118)
-1 OOOpcm _
-1000pcm in on Grp 7289 pcm = 45 in on Group 3
( 37 to 52 ) 2a2d Calculate line-by-line differences and Net Reactivity:
- DENOTES CRITICAL NRC ADM JPM COO ECC REV. a 0 Page 7 of 11
STEPS ELEMENTS STANDARD S/U COMMENTS REQUIRED FOR UNSAT Operator subtracts Column 1 from Column 2 Calculate the power line-by-line. defect difference: Column 1 - Column 2 = 1335 - 0 = +1335 pcm pcm (1300 to 1370) Calculate the xenon worth difference: Column 1 - Column 2 = 2550 - 2460 = + 90 pcm pcm (25 to 100) Calculate the samarium and neptunium worth ( difference: Column 1 - Column 2 = 806 - 826 = -20 pcm pcm (-17 to -22) Calculate the boron worth:
- difference: 7755 - 9816 = - 2061 Column 1 - Column 2 = pcm pcm (-1947 to -2091)
Calculate the CEA reactivity worth difference. Column 1 - Column 2 = 8580 - 8289 = -291 pcm pcm (-280 to -305) ( Reactivity Deviation o pcm Net Reactivity: 2a2e
- DENOTES CRITICAL NRC ADM JPM COO ECC REV. 0 Page 8 of 11
( STEPS ELEMENTS STANDARD S/U COMMENTS REQUIRED FOR UNSAT
** Total up all the -947 pcm differences.
(- 749 to - 1068) 2a2f Total change in Boron:
** Net Reactivity) diff. boron -947/8.18 =
-947 I 8.18 = -116 ppm worth. (-91 to -114.2)
~ pcm ) _
_ ~ pcm/ppm = _ _ ppm. Dilute 2a2f Estimated Critical Boron Concentration. Present RCS Boron + ( Total Change in Boron _ppm+_ppm _ppm +~ppm = = 1200 ppm - 116 = 1084 _ppm (91 to 144.2) (1055 to 1109) 3.0 Per Note 3 the operator Cue: Tell the operator must compare the ECC that the Desk RCO with the shutdown margin completed Data Sheet 2 requirements of 2-NOP- for 2-NOP-1 00.04 and 100-04. determined that the Required Boron The operator locates 2-Concentration is 841 NOP-1 00.04 to determine shutdown margin for these ppm. conditions.
- DENOTES CRITICAL NRC ADM JPM COO ECC REV. 0 Page 9 of 11
STEPS ELEMENTS STANDARD S/U COMMENTS REQUIRED FOR UNSAT Per Note 2 the operator 4.0 INITIALS Note 2 of ECC ensures that the critical Calculation Worksheet
- CEA position is above the insertion limits of Technical Specification 3.1.3.6 and initial Note 2 of ECC Calculation Worksheet.
Calculates CEA position at 5.0 ENTERS the CEA
-1000 PCM, +500 PCM, positions on the
- peM from the and -500 PCM appropriate line. This was estimated critical condition calculated on step 2a2c.
and enter on the appropriate line. ( The licensed operator 6.0 SIGNS the Prepared by preparing the worksheet line. shall sign the Prepared by line 7.0 Reactor Engineering shall Cue: Tell the operator independently verify the that this verification will ECC and sign the be completed by RE Ll}degendently: LfJdegendently: Verified by: later. END of JPM. line. STOP TIME _ _ _ __
- DENOTES CRITICAL NRC ADM JPM COO ECC REV. 0 Page 10 of 11
CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK) INITIAL CONDITIONS Unit 2 has been operating for 118 days at 100% power with all rods out. The unit was shut down at 1800 on Sept. 29, 2009, due to Main Feedwater problems. We anticipate reactor startup #09-2 to reach criticality at 1500 on Sept. 30, 2009. The following conditions exist:
- Reactor core exposure is 2800 EFPH.
- RCS Boron concentration is 1200 ppm.
- A JPN Engineering Evaluation is unavailable for this startup.
The Shift Manager instructs you to perform a manual Estimated Critical Condition ( Calculation by completing Appendix C (Estimated Critical Conditions) of 2-GOP-302 (Reactor Plant Startup - Mode 3 to Mode 2) in preparation for the unit startup. Last Stable RCS Boron and Criticality Data o [j RC System Boron was 948 ppm at 1300 on Sept 29, 2009. o Reactor Criticality Data from RCO Log dated 1300 Sept 29, 2009.
- CEA Group/Position (inches) 5 / 136
- T-avg (OF)
T-avg(OF) 574
- Reactor Power (%) 100 o
[] Reactivity Deviation (PCM) 0 ( Page 11 of 11
~+
EVISION NO.: FEVISION PROCEDURE TITLE: PAGE: 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 52 of 61 ~ROCEDURE ROCEDURE NO.: 2-GOP-302 ST. LUCIE UNIT 2 APPENDIX C APPENDIXC ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 6 of8) of 8) ECC CALCULATION WORKSHEET (Page 1 of 2) Unit No.: 2 Startup No.: 01 -~ ::2 07' L Date of ECC: q / .Jo/ 9;3b/ Or;. oQ, Exposure: :z tfOO
- 2tfOO EFPH COLUMN 1 COLUMN 2 Z Conditions Prior to Shutdown Estimated Critical Conditions Difference Sign PARAMETER Date~/$!1.!l.-
!L~Q£ Time 1: Date' I I. IjO/)
Time: ItO!) . Time 2: Date i 1_3...12.-, ()9 L..1!L, (}9 Time:L5 () i) Time:1.5 () (PCM) Determinatio n Determination f! 3()1J 3()()- ... J370 1370 1300-1370 11300-1370 Power Defect At LOO % Power FigUr~A.)3s) FigUr~A.1 ) NA NA (+) PCM (+)Always (+) Always
... 1 3>3:;
13>,./ PCM ... .' (113~:n 13o§J
;<!:"~~ -.<,00
,;)..5.:l,'t" ~..<'OO From Figure AA or as provided by R.E. 2-5 (DO ZS )0 ..- 25'""{PtM
'25'5fl'CM t<. S':)[) D
- I:.S":.-I() 0 Xenon Worth From DCS or Figure AA PCM (2560 ) ~
(2%D PCM ( '2..~(.,
'2.~("O) 0) i-f(
Lf( '10)q 0) (+) If Column 1 is Samarium and From Figure A.S. A.5. foS
?O~ -gor From Figure A.S.
A.5. or as provided by R.E. 5-"';.~7pCM ft? $--:'f).. f~ 7PCM
..... 17-2.v
)
17-Z""l/ PCM PCM greater than Column 2 Neptunium Worth (DoC. \'I ('80' PCM ( ?2.c.) PZc.) (- 2.0) (-) If Column 1 is Boron Worth PI.L
~ zZt{C;-- tf:::GO qs-:- tf::GO -'77 '17 -777
'7"1 V7'-777 ~ lZ IZD Present Boron conlntration conz.ntration - 9[1/0
~720-9jlJo e(1720 /'117- ZOCJt
/t;t/l- 209/ less than Column 2 PPM X Boron Worth oDPPM X £1- f;f-,2. ,2- -'-') PCM (Note 1)
A('lr*I'r) From Figure A.("6.1't) (( 77~ PCM 77 S:;-") Boron~)orth Boron Worth A.f.!?, If)" From Figure A'¥?I/ fi-
~
I'
.q?/~ }
*Cfi'/b}
PCM ""') (-2.06 l) 1(-2.oG nPC 2jO-30~
-0/
/
(-) If Column 1 is
//'
fo<. 7??Jo6 75- ?306 2!O-30£' CEA Reactivity GrouP4~ Group4e:v- withdraw to 10 ~ inches ?5f'O P5fO £:;
. Group _ _ withdraw to ft
.b..12- Inche:;l.,
inches PCM (-) PCM greater than Column 2 Worth Use Figure A.6. A.S. or A.7. ~ Use Figure A.6. A.S. or A.7. (Note 2) ( ?2r9)
?zrCj) if/ -2 q I)
(+) If Column 1 is
,) less than Column 2 Reagivity Reactivity Deviation N/A N/A From Figure 8.3 B.3 or as provided by R.E. ~PCM ()~PCM ( )~PCM NeYReactivity Net Reactivity Total up the reactivities in the Difference column and enter the value here. Observe signs. ( ) peM PCM
-_..- _J
(~ EVISION NO.: P.EVISION PROCEDURE TITLE: PAGE: 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 53 of 61 53 of 61 ROCEDURE NO.: PROCEDURE 2-GOP-302 ST. LUCIE UNIT 2 APPENDIXC APPENDIX C ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 7 of 8) ECC CALCULATION WORKSHEET (Page 2 of 2)
. Sign PARAMETER PARAMETER Difference Difference Determination Determination
({ 1tf9-I06Y 1L/9- l q6Y
,..,. q /.;. / .;. 14!f/z,.
It/4t Z, . If sign is +, Total Change in ( Net Reactivity ) = _ ( ( (-) (-) (Cj + 7) PCM) ~) (<=j 17) ~) II II (; (g PPM Borate PPM Borate Boron Boron Column 2 Boron Worth £/ f;/ _ z,. PCM/PPM f, Z,
-' l(, PCMIPPM If sign If sign is is -- ,,
" ", (<;.1 Y)
\¥.IY) Dilute Estimated Critical Estimated Critical (( I ;{J ;Z 00O!) ) + ( (-)(-) (ell C~ fZ {'
( --, /<f,(.?/)
. I(L£G2'
" ~
~ r-- 1P 55 r-..I()
( )! 5503'
) IOJ"f
--1°9)
/ 10 9)
+ Total Change i~
Boron PPM PPM (Note 3) Concentration Present RCS Boron Concentration in Boron Concentration GEA CEA Position at -1000 PGM PCM from EGG =31; 5 ECC = 3.j - <j Y
'Y Inches withdrawn on Group:1 Group 3 S1artup (Mode 2) ent",
Startup entry pOln( pOin,( 7.;:( 7.:?!f ffj7c"" c",) IOtj -/I (' CEA Position at +500 PCM from ECC loti -/I ( CEA Position at +500 PCM from ECC = Inches withdrawn on Group Inches withdrawn on Group 3 3 77J-9 pc. (( 77J-y pCJn) ~) .
= /lIZ
/If( 0D j>J ;-7 r7 le)y, CEA Position at -500 PCM from ECC Inches withdrawn on Group (t-J
( I jJc)y,
))
." .. _ ........".... * .... *. ..~ " _..... -" . -........
......................--~." - . . ." ....,.... , . "....... .......l ..
.--l
REVISION NO.: PROCEDURE TITLE: PAGE: 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 1---------------; I------~------~ 54 of 61 PROCEDURE NO.: 2-GOP-302 ST. LUCIE UNIT 2 APPENDIX C ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 8 of 8) ; ECC CALCULATION WORKSHEET (continued) NOTES: 1. If the difference in time between when the boron sample was drawn and the time of shutdown exceeds 24 hours, contact Reactor Engineering.
- 2. PJ(J\~ has been verified to be above PDIL The critical CEA pXO\~ POlL (Technical Specification COLR Fig. 3.1-2) (initials).
- 3. The ECC Boron Concentration must meet or exceed the boron concentration requirements for shutdown margin with CEAs at Column 2 position in accordance with 2-NOP-100.04, 2-NOP-1 00.04, ;
Surveill~;e Requirements for Shutdown Margin, Modes 2, 3, 4 and 5, Subcritical. surveill"N;e Prepared Prepared By: By, lAv t \./ ~
'1/\ ~
RCO l----' L--. ACTUAL CRITICAL CONDITIONS ACTUAL CRITICAL CONDITIONS Date of Criticality _ _,/_ _/ _ Independently Time (24 hour clock) Verified By: _ _ _ _:=--_--::=--:-----:--_ _ _ _ _ _ _ __ __ _____ CEA Group/Position (inches) ____ _ _....:'
-.0/_ _
Reactor Engineering RCS Boron Concentration (PPM) _ _ __ ( Xenon Worth (DCS or Fig A.4) T-avg (OF) I have reviewed the requirements of this Reactor Power (%) _ _ _ __ procedure including other surveillances performed during this procedure, if any (i.e., datasheet(s), PMT sheet(s), etc.). Any deviations, abnormal results, equipment problems, failures, or human performance issues must be documented via a Condition Report for each individual item. C.R.# _ _ _ _ _ _ _ __ Reviewed By: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ Reactivity Manager ApprovedBy: _ _ _ __ __ __ __ __ __ __ __ _~_ ___ ~------------
-~-------------
SM or US
REVISION NO.: PROCEDURE TITLE: PAGE: 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 ( I--------------~ 54 of 61 PROCEDURE NO.: 2-GOP-302 ST. LUCIE UNIT 2 APPENDIX C ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 8 of 8) ECC CALCULATION WORKSHEET ((continued) continued) NOTES: 1. If the difference in time between when the boron sample was drawn and the time of shutdown exceeds 24 hours, contact Reactor Engineering.
- 2. The critical CEA position has been verified to be above PDIL (Technical Specification COLR Fig. 3.1-2) (initials).
- 3. The ECC Boron Concentration must meet or exceed the boron concentration requirements for shutdown margin with CEAs at Column 2 position in accordance with 2-NOP-100.04, :'
Surveillance Requirements for Shutdown Margin, Modes 2, 3, 4 and 5, Subcritical. Prepared By: _ _ _ __ __ __ __ __ __ __ __ __ __ ___ __ ACTUAL CRITICAL CONDITIONS RCO Date of Criticality _ _1 __
/_ _1_/_
Independently Time (24 hour clock) , Verified By: _ ___ __ __ __ __ _ __ ____ __ __ __ ____ CEA GrouplPosition (inches) Group/Position (inches)::---c----'/ 1 Reactor Engineering (PP=-M-:-)=~::::::::==_ RCS Boron Concentration (PPM) ______ ( Xenon Worth (DCS or Fig AA) A.4) T-avg (OF) I have reviewed the requirements of this Reactor Power (%) _______ procedure including other surveillances performed during this procedure, if any (i.e., datasheet(s), PMT sheet(s), sheet(s}, etc.). etc.}. Any deviations, abnormal results, equipment problems, failures, or human performance issues must be documented via a Condition Report for each individual item. C.R.# _ _ _ _ _ _ _ _ __ Reviewed By: ____________________________ Reactivity Manager Approved By: ____________________________________________________ SM or US (
REV[S[ON NO.: REVISION T[TLE: PROCEDURE TITLE: PAGE: 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 1---.....:....::...----1 [---------1 47 of 61 PROCEDURE NO.: 2-GOP~302 2-GOP-302 ST. LUCIE UNIT 2 APPENDIX C ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 1 of 8) NOTE
- The ECC worksheet should be considered valid for a period not to exceed 1/2 112 hour before or 1/2 112 hour after the time entered in Column 2 of the ECC Calculation Worksheet. The decision to change ECCs during the approach to criticality shall be made by the Reactivity Manager with concurrence of the US or SM. !f If the intended time of criticality exceeds 4 days from the time of shutdown, Then Xenon and Samarium effects will be negligible and the 1/2112 hour guideline is NOT applicable.
- Data being used is required to be consistent with current core burnup.
- 1. Enter the following data on the ECC Calculation Worksheet:
A. Unit No: Enter the unit number for which the ECC applies. B. Startup No: Enter the Startup number as determined from the Unit 2 Control Room working copy of OP-0030129, Reactor Trip ( Records. C. Date of ECC: Enter the date that the ECC calculation is prepared. D. Exposure: Enter the current Exposure of the core from DCS. NOTE At the discretion of the Reactor Engineering Supervisor, Estimated Critical Conditions provided by the Nuclear Fuel Department, BEACON modeling, or equivalent, may be used instead of the ECC Worksheet Manual calculation. !fIf one of these alternate methods is used, Then skip Step 2A and perform Step 2B.
- 2. Determine Estimated Critical Conditions as follows:
NOTE
- The values entered in column 1 are for the last known equilibrium conditions which the reactor operated at a stable power level for a minimum of 40 hours.
- The values entered in column 2 correspond to the reactor conditions in column 1.
A. Manual determination of Estimated Critical Conditions:
REVISION NO.: PROCEDURE TITLE: PAGE: 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 { 1 I ---- --- -- --
-i ; 48 of 61
\ PROCEDURE NO.: 2-GOP-302 ST. LUCIE UNIT 2 APPENDIX C ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 2 of 8)
- 2. A. (continued)
- 1. Complete Conditions Prior to Shutdown (Column 1) as follows:
- a. Time 1:
Enter the date and time (24 hour clock) of Reactor shutdown.
- b. Conditions Prior to Shutdown:
Using the last stable boron sample (chemical analysis) and recorded critical conditions from the RCO log, calculate the reactivity for each parameter. (Note that if the time difference between the last stable boron sample and the time of the Reactor shutdown is greater than 24 hours, Reactor Engineering should be contacted for a more precise value of the reactivity to be used for Column 1 boron worth.)
- c. 11 at the time of shutdown equilibrium conditions were not present, Then ensure the data in column 1 reflects the last known equilibrium operating conditions at the operating power level prior to the time of shutdown.
- 2. Complete Estimated Critical Conditions (Column 2) as follows:
- a. Time 2:
Enter estimated date and time (24 hour clock) that criticality is expected to be achieved. NOTE Reactivity defect associated with temperature changes are accounted for in power defect values which assume criticality is attained at 532°F.
- b. Estimated critical conditions:
Using the plant physics curves or data provided by RE, calculate the reactivity for each parameter at Time 2. When estimating CEA withdrawal height (if possible, apprOXimately 60 inches on Group 5), verify that the critical position is above the insertion limits and initial Note 2 of ECC Calculation Worksheet that this has been checked. (
- c. Calculate CEA position at +500 PCM and -500 PCM from the estimated critical condition and enter on the appropriate line.
REVISION NO.: NO .. PROCEDURE TITLE: PAGE: 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 I-I ---
- ------t f 49 of 61 PROCEDURE NO.:
2-GOP-302 ST. LUCIE UNIT 2 APPENDIX C ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 3 of 8)
- 2. A. 2. (continued)
- d. Subtract the absolute value of the parameter in Column 2 from the absolute value of the parameter in Column 1 and enter the difference in the difference column. Refer to the determination of sign column for each parameter and assign the sign accordingly.
- e. Add all values in the difference column, applying signs as appropriate and enter the total. Divide this Net Reactivity by Boron Worth. The sign of the total change in boron determines if boron concentration needs to be increased or decreased.
- f. Enter the total change in Boron required from the present concentration and calculate the Estimated Critical Boron Concentration.
( g. If equilibrium conditions were not present at the time of shutdown, Then ensure that the data used in column 2 corresponds to the reaction conditions in column 1. B. Determine the Estimated Critical Conditions using an Alternate Method as follows:
- 1. The Reactor Engineer shall obtain an estimate of critical conditions.
This may be provided by the Nuclear Fuel department, completion of a BEACON model, or an equivalent output document, which determines critical boron concentration for a specific Lead Group position (normally 60 inches withdrawn on Group 5).
- 2. The estimated critical conditions, determined by one of the preceding methods, shall include a minimum of 24 critical boron concentrations spaced at one hour intervals to allow for possible delays.
- 3. Complete the ECC Worksheet by marking all Column 1 and Column 2 variables as "N/A", with the exception of Column 2 CEA position.
Attach the document used to determine the boron concentrations to the back of the ECC Worksheet. Engineering Document Type and Number: _ _ _ _ _ _ Date _ __ _1___ _1__ Reactor Engineer Providing the Estimate: _ _ _ _ _ _ _ Date __1_ __ _1__
REVISION NO. PROCEDURE TITLE: PAGE: 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 ( 1 1- ----- ------ -- --- 1 / 50 of 61 PROCEDURE NO.: 2-GOP-302 ST. LUCIE UNIT 2 APPENDIX C ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 4 of 8)
- 3. §1 Ensure the Estimated Critical Boron Concentration is greater than or equal to the boron concentration required for shutdown margin needed prior to Startup (Mode 2), in accordance with 2-NOP-100.04, 2-NOP-1 00.04, Surveillance Requirements for Shutdown Margin, Modes 2, 3, 4, and 5 (Subcritical).
- 4. §1 Ensure that the critical CEA position is above the insertion limits of Technical Specification 3.1.3.6 and initial Note 2 of ECC Calculation Worksheet.
- 5. Calculate CEA position at -1000 PCM, +500 PCM, and -500 PCM from the estimated critical condition and enter on the appropriate line.
- 6. The licensed operator preparing the worksheet shall sign the Prepared by line.
- 7. Reactor Engineering shall Independently Verify the ECC and sign the Independently Verified by line.
- 8. The Reactivity Manager shall review ECC in accordance with Appendix E and sign
( the Reviewed by line.
- 9. The SM or US shall review the worksheet and sign the ~pproved Approved by line.
- 10. Following the boration / dilution to achieve the Estimated Critical Boron Concentration, the present boron concentration should be determined by chemical analysis. There should be approximately a 20 minute wait between the boron concentration change and RCS sampling.
NOTE An ICRR shall be plotted for all approaches to criticality. The initial approach to criticality following refueling shall be performed in accordance with Preop 2-3200088, Initial Criticality Following Refueling. After a Reactor trip from low power or during start-up testing, the critical conditions at the time of trip may be used for recovery or restart.
- 11. Complete the ICRR Data Sheet and Plot Sheet as follows:
A. Verify that the ECC Calculation Worksheet has been completed and enter the estimated critical CEA position on a copy of the ICRR Data Sheet (Figure A.9 Data Sheet in the Physics Curve Book). B. Select two Startup or Wide Range Log Safety channels and record the stable CPS or percent power with CEAs at the initial position and record initial CEA position.
REVISION NO.: PROCEDURE TITLE: PAGE: 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 1 --------1 I-------~ 51 of 61 PROCEDURE NO.: 2-GOP-302 ST. LUCIE UNIT 2 APPENDIX C ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 5 of 8)
- 11. (continued)
C. Commence a Reactor Startup in accordance with this procedure and at approximate positions indicated on the ICRR Data Sheet In Plant Physics Curves, stop CEA withdrawal and allow the count rate to stabilize. D. Record the CPS or percent power indication on the ICRR Data Sheet and compute the ICRR for both selected channels using the formula:
%0 Powen. or=~i cps*
cps' ICRRn=
% Powern cpsn Where:
% Power ij = Initial % Power CPS i = Initial counts per second
% Power n = % Power at present CPS n = Counts per second at present
( ICRR n = Present Inverse Count Rate Ratio E. Plot the ICRR on the Figure A.9 Inverse Count Rate Ratio Plot Sheet. F. If the ICRR plot indicates that criticality will occur at a CEA height which is less than the POI L limits of Technical Specification 3.1.3.6, Then: POlL
- 1. Stop CEA withdrawal.
- 2. Recompute the ECC using the current verified RCS boron concentration by chemistry analysis, if necessary.
- 3. Determine a boron concentration that will establish criticality above the POlL.
- 4. Borate to the calculated concentration.
- 5. Start the ICRR plot over again (re-normalize) by substituting the count rate after the boron change for CPS i.j.
- 6. Continue CEA withdrawal as before.
- 12. Between 3 x 10-44 % and 7 x 10- 10-44 % of rated power, following criticality, stabilize conditions then record actual critical data on the ECC Calculation Worksheet.
(
EVISION NO.: PEVISION PROCEDURE TITLE: PAGE: 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 PROCEDURE ROCEDURE NO.: 52 of 61 2-GOP-302 ST. LUCIE UNIT 2 APPENDIX C ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 6 of 8) ECC CALCULATION WORKSHEET (Page 1 of 2) Unit No.: 2 Startup No.: Date of ECC: I/ I/ Exposure: EFPH COLUMN 1 COLUMN 2 Conditions Prior to Shutdown Estimated Critical Conditions Difference Sign PARAMETER (PCM) Determination Time 1: Date I I Time: Time 2: Date
--'--'-- I I Time:
Power Defect At % Power Figure A.1 -PCM
-- NA NA (+) PCM (+) Always From Figure AA or as provided by R.E. RE.
Xenon Worth From DCS or Figure A.4 AA PCM ( ) PCM PCiV1 PCrV1 (+) If Column 1 is From Figure A.S. or as provided by R.E. greater than Samarium and PCM Column 2 From Figure A.S. -peM
-- ( ) PCM Neptunium Worth PCiV1
(-) If Column 1 is Present Boron Concentration less than Column 2 Boron Worth PPM X Boron Worth ( ) peM PCM
--- PPM X Boron Worth PCM (Note 1) peM PCM From Figure A.B. From Figure A.B.
(-) If Column 1 is greater than Group _ _ _ withdraw to '. Group _ _ withdraw to CEA Reactivity - - - inches -PCM
-- - - - inches PCM ( ) peM PCM Column 2 Worth Use Figure A.S.
A.6. or A.7.' A.7. peM Use Figure A.S. A,6. or A.7. (Note 2) I (+) If Column 1 is less than Column 2 Reactivity N/A N'A N/A N'A B.,3 oras From Figure B.3, Of,as provided by R.E. PCM ( ) peM PCM Deviation Net Reactivity Total up the reactivities in the Difference column and enter the value here. Observe signs. ( ) PCM
'~--~-.-.....,---~--,
---~--.,---~~-- -
~
REVISION NO.: PROCEDURE TITLE: PAGE: 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 53 of 61 PROCEDURE NO.: 2-GOP-302 ST. LUCIE UNIT 2 APPENDIXC APPENDIX C ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 7 of 8) ECC CALCULATION WORKSHEET (Page 2 of 2) Sign PARAMETER Difference Determination If sign is +, Borate Total Change in ( Net Reactivity ) _ ( ( ) PCM ) ( ) PPM Boron Column 2 Boron Worth PCM/PPM If sign is -* , Dilute T~tai Change in Boron concentration) Estimated Critical Boron ( ) PPM (Note 3) Co ncentratio n Concentration ( Present RCS Boron concentration) + [( (present Change in Boron concentration) CEA Position at *1000
-1000 PCM from ECC =
= Inches withdrawn on Group Startup (Mode 2) entry point CEA Position at +500 PCM from ECC = Inches withdrawn on Group
*500 PCM from ECC =
CEA Position at -500 Inches withdrawn on Group
REVISION NO.: PROCEDURE TITLE: PAGE: 38 38 REACTOR PLANT STARTUP - MODE 3 TO MODE 2 r--------------;
~------------~ 54 of 61 PROCEDURE NO.: ~ 2-GOP-302 ST. LUCIE UNIT 2
_ _ _ _ _ _ _ _ _ _ _ _ _ _L -_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~_ _ _ _ _ _ _ _ _ _~'*t APPENDIX C ESTIMATED CRITICAL CONDITIONS AND INVERSE COUNT RATE RATIO (Page 8 of 8) ECC CALCULATION WORKSHEET (continued) NOTES: 1. If the difference in time between when the the. boron sample was drawn and the time of shutdown exceeds 24 hours, contact Reactor Engineering.
- 2. The critical CEA position has been verified to be above POlL PDIL (Technical Specification COLR Fig. 3.1-2) (initials).
- 3. The ECC Boron Concentration must meet or exceed the boron concentration requirements for shutdown margin with CEAs at Column 2 position in accordance with 2-NOP-100.04, '
Surveillance Requirements for Shutdown Margin, Modes 2, 3, 4 and 5, Subcritical. Prepared Prepared By:By: -________
- - - - - - = -:::--:-___
= c- -= --- -- - ---- --- - ACTUAL CRITICAL CONDITIONS RCO Date of Criticality _ _, _ _, /_
Independently Time (24 hour clock) Verified By: _ _ _ __ __ __ _:--:::-_
,--_ _ __ __ __ _ CEA Group/Position Group'Position (inches) ___ _~,_ _
Reactor Engineering RCS Boron Concentration (PPM) _ _ __ ( Xenon Worth (DCS or Fig A.4) T-avg (OF) I have reviewed the requirements of this Reactor Power (%) ________ procedure including other surveillances performed during this procedure, if any (i.e., (Le., datasheet(s), PMT sheet(s), etc.). Any deviations, abnormal results, equipment problems, failures, or human performance issues must be documented via a Condition Report for each individual item. C.R.# _ __ __ __ __ _ _ _ __ __ ___ Reviewed By: 8y: ______________________________________________ Reactivity Manager Approved By: 8y: ____________________________ SM or US
REGION II ST. LUCIE NUCLEAR PLANT INITIAL LICENSE EXAMINATION ADMINISTRATIVE JPM Perform a Manual Calorimetric NRC A-2 CANDIDATE EXAMINER NRC ADM JPM A-2 fRey 0 Classroom ( PAGE 1
REGION II ST LUCIE NUCLEAR PLANT INITIAL LICENSE EXAMINATION ADMINISTRATIVE JPM Perform Manual Calorimetric KA Statement: Ability to evaluate plant performance and make operational judgments based on I operating characteristics, reactor behavior, and instrument interpretation. KA#: 2.1.7(4.4/4.7) 2.1.7(4.414.7)
References:
2-3200020, Primary System Manual Calorimetric, Steam Tables Candidate: Time Start --- Name Time Finish --- Performance Rating: Sat Unsat Validation Time 30 minutes Examiner: Signature: Comments NRC ADM JPM A-2 IRev fRey 0 PAGE 2
REGION II ST LUCIE NUCLEAR PLANT INITIAL LICENSE EXAMINATION ADMINISTRATIVE JPM Perform Manual Calorimetric Directions to the candidate for Administrative JPMS: will explain the initial conditions and state the task to be performed. You will be I wi" allowed to use any reference normally available in the Control Room to complete the task. Ensure you indicate to me when you finish your assigned task by returning the material needed for the task that I provided to you. Initial Conditions Today is Thursday 10-01-09, 11 :30am. Unit 2 is at approximately 45% power steady state. Initiating Cue ( The Unit Supervisor has instructed you to perform a primary system manual calorimetric lAW 2-3200020', Primary System Manual Calorimetric' Data Sheet 1 and determine core power. NO computers are available to perform this calculation. Classroom Data Sheet
- Steam Generator Pressures are as follows:
PI-8013A: 885 psia PI-8023A: 886 psia PI-8013B: 887 psia PI-8023B: 886 psia PI-8013C: 888 psia PI-8023C: 887 psia PI-8013D: 889 psi psiaa PI-8023D: 886 psia psi a
- Feedwater Temperatures are 362.rF.
- Total blowdown = = 80 gpm
- Feed flow 9011-2 (ERDADS) = =2.50 x1 0666 Ibm/hr
- Feed flow 9021-2 (ERDADS) = 2.50 x1 0 6 Ibm/hr
- DCS pt KFA = = 29101
- DCS pt KFB = = 29345
- Charging and Letdown are in operation.
- DCS 10 min avg: 46.1%
{( NRC ADM JPM A-2 (Rev 0 PAGE 3
JOB PERFORMANCE MEASURE 2-3200020, Primary System Manual Calorimetric Section 6.0 Data Sheet 1 STEP 1 (6.2): Date and current time. _ _ SAT STANDARD: ENTERS date and current time on Data Sheet 1 per classroom data. UNSAT EXAMINER'S CUE: None COMMENTS:
.--------------------------------------+--------1 r--.-----------------------------------------------r--------~
STEP 2 (6.2): MAIN STEAM PRESSURE CRITICAL STEP PI-8013A: 885 psia ( PI-8023A: 886 psia PI-8013B: 887 psia PI-8023B: 886 pSia psia _ _ SAT PI-8013C: 888 psia PI-8023C: 887 psia PI-8013D: 889 psi a psia PI-8023D: 886 psia psi a UNSAT STANDARD: ENTER MAIN STEAM PRESSURE for all pressures per classroom data. EXAMINER'S CUE: None COMMENTS: ( L- _-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- - 1_ _ _ _ _1
- - - - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - -_ _ _ _ _ _ _ _ _ _ _----"-_ _ _ _ _ _ _----l NRC JPMI Rev 0 Classroom Page 4 of 14
JOB PERFORMANCE MEASURE r*-----*- IS-TEP STEP 3 (6.3): COMPUTE TOTAL FEEDWATER FLOW CRITICAL STEP A A. A Loop x =-
-;- 29178 =
.;- ~~_x 66
- - x 10 1b/hr Ib/hr (FT 9011-2 (ERDADS), or F-9011) (DCS PT KFA)
_ _ SAT STANDARD: ENTERS and CALCULATES FEEDWATER FLOW A Loop. UNSAT A A. A Loop 2.50 x 29101 .;- =2.493 x 1066 Ib/hr
-;- 29178 =
FEEDWATER FLOW A Loop = =(2.490 - 2.493 X 1066 1b/hr) Ib/hr) EXAMINER'S CUE: None COMMENTS: ( ----.-~.-.--:-__=_----__:_c_:::______c---------------__+_---
~-.--~.-.~~-~~~~~~~~~~~~~~~~~~~~~+-~----------.-- ..--....
STEP 4 (6.3) COMPUTE TOTAL FEEDWATER FLOW CRITICAL STEP B. B Loop x .;- =___
-;- 29050 = ~~_xx 10 6 1b/hr 6
Ib/hr (FT 9021-2 (ERDADS), or F-9021) (DCS PT KFB) _ _ SAT STANDARD: ENTERS and CALCULATES FEEDWATER FLOW BLoop. 6 UNSAT B. BLoop 2.50 x 29345.;-29345 -;- 29050 = 2.525 x 106 Ib/hr 6 FEEDWATER FLOW BLoop = := (2.52 - 2.53 X10 6 Ib/hr) EXAMINER'S CUE: None COMMENTS: IIL......... __ _ __ _____ ~ __ _ _ _ _-"-_ _ _ _ _ _ _ _ _..._.. _________L _ _ _ . _____ . ___.1 .J NRC JPMI Rev 0 Classroom Page 5 or or 14
JOB PERFORMANCE MEASURE STEP 5 (6.3) COMPUTE TOTAL FEEDWATER FLOW CRITICAL 6 STEP MFW =
= - - - - x 106 1b/hr Jb/hr A+8 Loop)
(Total A+B _ _ SAT STANDARD: ENTERS and CALCULATES TOTAL FEEDWATER FLOW. UNSAT 6 MFW = (2.493 + 2.525 = 5.018 X 10 6 1b/hr A+8 Loop) (Total A+B
=
( 2.49 + 2.52 = 5.01)
=
( 2.493 + 2.53 =5.023) MFW Flow total = ( 5.01 - 5.023 X10X1 066 Ib/hr) Ih/hr) EXAMINER'S CUE: None COMMENTS: ( I I--------------------------------+-----~-- I STEP 6 (6.3) COMPUTE Average Steam Pressure. CRITICAL I STEP Average Steam Pressure = =_ __ __ __ _ divided by 8 = _ _ _ _ _ TOTAL AVERAGE _ _ SAT STANDARD: ENTERS and CALCULATES Average Steam Pressure per classroom data. UNSAT 885 +886 +887 +886 +888 +887 +889 +886 = =7094/8 = 886.75
=
Average Steam Pressure = 886.8 psia. EXAMINER'S CUE: None COMMENTS: NRC JPMI Rev 0 Classroom Page 6 of 14
JOB PERFORMANCE MEASURE STEP 7 (6.3): ENTER FEEDWATER TEMPERATURE o _ _ SAT TR-09-5B Channel ChanneI11_ _ _oF Channel 2_ _ _ oF ChanneI2 F (behind RTGB 202) Average Feedwater Temperature _ _ _ _ divided by 2 = UNSAT __ __ ___ ___ TOTALAVERAGE STANDARD: ENTERS FEEDWATER TEMPERATURE and CALCULATES Average Feedwater Temperature. Both channels were given as 362. rF per classroom data so total Average Feedwater Temperature is 362.7°F. No calculation is required. I I EXAMINER'S CUE: None I COMMENTS: I ( I 1------------------------------------1-------1 STEP 8 (6.4.1): Determine the enthalpy for the average main steam CRITICAL pressure (hsTEAM). STEP STANDARD: DETERMINES and ENTERS the enthalpy for the average main steam pressure (hsTEAM). _ _ SAT 886.8 psia average main steam pressure psia 860 psi a : hSTEAM= 1197.5 BTU/Ibm UNSAT 900 psia : hSTEAM= 1196.2 BTU/Ibm hSTEAM= 1196.6 BTU/Ibm
=
hSTEAM: = (1196.4 - 1197.3 BTUllbm) EXAMINER'S CUE: None ( - -... - - - - - - - - - - . - - - - - - - - - - -_ _ _ _ _ _ _ _ _ _ _ _- - ' - -_ _
~------------------------------------------'----------
~ _ _ _ _ _ .. J NRC JPMI Rev 0 Classroom Page 7 of 14
JOB PERFORMANCE MEASURE COMMENTS: rOMMENTS I STEP 9 (6.4.2): Determine the mass flow of blowdown by using Table 1, CRITICAL Data Sheet 1, to convert GPM to Ibm/hr. STEP _ _ SAT STANDARD: DETERMINES And ENTERS and the mass flow of blowdown. UNSAT
.039599 x 1066 1bm/hr Ibm/hr EXAMINER'S CUE: None COMMENTS:
I STEP 9 (6.4.3): Subtract blowdown mass flow (MBD) from feedwater mass CRITICAL ( ! flow (MFw) to obtain steam flow. STEP I _ _ SAT i STANDARD: DETERMINES And ENTERS steam flow. I I j 6 039599 X 106 )
- .039599 (5.01 - 5.023 .. = 6 (4.97- 4.983 X 106 Ibm/hr)
[ MFW - MBO] = Steam flow UNSAT 6 Steam Flow: 4.97- 4.983 X 10 6 Ibm/hr EXAMINER'S CUE: None I COMMENTS: I~-~~-----------------~---------------+--------~-~ I STEP 9 (6.4.4): Multiply the quantity from step 6.4.3 times the enthalpy from CRITICAL step 6.4.1. This resultant is QSTEAM. STEP 6 6 66 6 10 - - -10 106 BTU/hr
----:-- - _ _ SAT
( hSTEAM X [MFW - MBO] =
= QSTEAM NRC JPM! Rev 0 Classroom Page 8 of 14
JOB PERFORMANCE MEASURE STANDARD: DETERMINES QSTEAM. UNSAT (1196.4 - 1197.3) X (1196.4-1197.3) (5.01 - 5.023 .. X(5.01-5.023 - .039599 039599X10 6
=
X 106 ) =(5946.1-(5946.1 - 5962.6 X X10 6 106 BTU/hr) hSTEAM X [ MFW - MBD] = QSTEAM (4.97 - 4.983) QSTEAM = (( 5946.1 - 5966.1 X 1066 Ib/hr )
=
EXAMINER'S CUE: None COMMENTS: STEP 10 (6.5.1) : From the steam tables, enter the enthalpy of the average CRITICAL feedwater temperature: STEP hFw =____ _____ BTU/Ibm _ _ SAT ( STANDARD: ENTER the enthalpy of the average feedwater temperature using steam tables: UNSAT TFW 362.rF 360°F = 332.34 BTU/Ibm BTU/Ibm 364°F = 336.57 BTUllbm hFW == 334.45 BTU Ibm hFw = hFw = (334.0- 335.0 BTU/Ibm) BTUllbm) EXAMINER'S CUE: None COMMENTS: ( NRC JPMI Rev 0 Classroom Page 9 of 14
JOB PERFORMANCE MEASURE I STEP 11 (6.5.2) : Calculate the heat feedwater heat input (QFW): CRITICAL STEP I ( x 10 6
) X ( ) = 6 x 10 BTU I M FW h FW OFW hr
_ _ SAT I I
\ STANDARD: ENTER and CALCULATE the heat feedwater heat input.
(QFW). (OFW). UNSAT (5.01 - 5.023 x 106 ) x (334.0 - 335.0) = 6 6
= 1673.34 - 1691.75 x 10 6 BTU MFW hFW QFW OFW hr 6
QFW OFW = (1673.0 - 1683.0 X 10 6 BTU/hr) BTu/hr) EXAMINER'S CUE: None COMMENTS: ( ( NRC JPMI Rev 0 Classroom or 14 Page 10 of
JOB PERFORMANCE MEASURE STEP 12 (6.6): Circle the total blowdown flow from the S/Gs and the CRITICAL corresponding heat output: (Interpolation is not required, STEP circle the matching blowdown flow below or the lower flow if between flow rates, to be conservative.) _ _ SAT Total blowdown flow Mass flow Heat output from (both steam generators) of blowdown MsoMBD blowdown Qso QBD UNSAT 40 GPM .019799 x 10e6 1bm/hr lbm/hr 9.728 x 1 10Oe6 BTU/hr 80 GPM .039599 x 10e6 lbm/hr 1bm/hr 19.456 x 10 66 BTU/hr 120 GPM .059398 x 10 1bm/hr 6 e lbm/hr 29.183 x 1OeOB BTU/hr 160 GPM 1bm/hr
.079198 x 10e6 lbm/hr 38.912 x 1066 BTU/hr 200 GPM .098997 x 10e6 1bm/hr lbm/hr 48.639 x 110Oe6 BTU/hr 240 GPM .118797 x 10 e lbm/hr 10"lbm/hr 58.367 x 1066 BTU/hr STANDARD: CIRCLES the total blowdown flow from the S/Gs and the corresponding heat output per classroom data. Determines QBD to be 19.456 X 1066 BTU/hr and enters on Data Sheet 1 QBO EXAMINER'S CUE: None COMMENTS:
---~-------.-------------------------t---------~ STEP 13 (6.7) If charging and letdown flow is secured for any reason, use CRITICAL a value of 48.952 for QOTHER. This was provided by step 6.7 STEP If charging and letdown flow is operating normally, use a value of 46.905 for QOTHER. STANDARD: CHOOSES 46.905 for QOTHER. Based on classroom data sheet that Charging and Letdown are in operation an enters ~ ~~ ~ on Data Sheet 1 _________o_n_D_a_t_a_S_h_e_e_t_1______________________________ ________ NRC JPMI Rev 0 Classroom Page II of 14
JOB PERFORMANCE MEASURE I STEP 13 (6.8) Calculate the heat output from the core: CRITICAL STEP II
- -+- - ~---,l
~_--,l x 1066 BTU = =___
_ _ _.xx 1066 BTU _ _ SAT
~~SAT
~~~~
QSTEAM Q QBO BD QOTHER hr Q CORE hr ii i I STANDARD: ENTER and CALCULATE the heat output from the core. I I UNSAT [(5946.1 - 5966.1) - (1673.0-1683.0)]+ [19.456 - 46.905]x1 46.905]x10"0 6 BTU = 4245.6- 4279.7x 10"BTU 6 10 BTU hr OCORE QCORE =(( 4235.0 - 4265.65 BTU/hr)
=
I EXAMINER'S CUE: None I Ii i COMMENTS: i I ( I1-____
.______ . ~~~~~~--~~--'---------
( NRC JPMI Rev 0 Classroom Page 12 of 14
JOB PERFORMANCE MEASURE NOTE CRITICAL CRITICALl Calculated Power in percent must be within 2% of STEP STEP I DCS Calorimetric power. If not, notify the US I SM. Initiate W/R for ICM to adjust !J. l\ T power. _ _ SAT STEP 14 (6.9): Calculate percent core power. UNSAT Core Power =
=__
_ _x 1066 divided by (92.128 x 1066 ) == __ _ _ % Manual QCORE Calorimetric Power STANDARD: ENTER and CALCULATE percent core power.
=4235.0 - 4265.65 Core Power =
6 6
=45.9 426565 x 106 I; (92.128 xX 106 ) = 459 - 46.3%
463% Manual QCORE Calorimetric Power Core Power = =(( 45.9 - 46.3%) ( EXAMINER'S CUE: Upon completion of Step 14 after manual calorimetric power is entered on Data Sheet 1 JPM is complete. COMMENTS: END OF TASK STOP TIME: NRC JPMI
.rPM/ Rev 0 Classroom 11 Page 13 of 14
(TO BE RETURNED TO THE EXAMINER UPON COMPLETION OF ANSWER) (TO BE RETURNED TO THE EXAMINER UPON COMPLETION OF ANSWER) Initial Conditions It is Thursday 10-01-09, 11 :30am. Unit 2 is at approximately 45% power steady state. Initiating Cue The Unit Supervisor has instructed you to perform a primary system manual calorimetric lAW 2-3200020', Primary System Manual Calorimetric' Data Sheet 1 and determine core power. NO computers are available to perform this calculation. Classroom Data Sheet
- Steam Generator Pressures are as follows:
PI-8013A: 887 psia PI-8023A: 886 psia PI-8013B: 887 psia ( PI-8023B: 886 psia PI-8013C: 888 psia PI-8023C: 887 psia PI-8013D: 887 psia PI-8023D: 886 psia
- Feedwater Temperatures are 362.
362.TF. rF.
- Total blowdown = 80 gpm 6
- Feed flow 9011-2 (ERDADS) = 2.50 x1 OE 6 1bm/hr 6
- Feed flow 9021-2 (ERDADS) = 2.50 x1 OE 6 Ibm/hr
- DCS pt KFA = = 29101
- DCS pt KFB = = 29345
- Charging and Letdown are in operation.
- 46.1 %
DCS 10 min avg: 46.1% JPMI Rev 0() NRC JPM! Classroom Page 14 or of 14 J4
REVISION NO.: PROCEDURE TITLE: 31 PRIMARY SYSTEM MANUAL CALORIMETRIC IPRIDCEDUF~E PROCEDURE NO.: 2-3200020 ST. LUCIE UNIT 2 c: c o .
..2......;
Eai Nffi DATA SHEET 1 (Page 1 of 3)
~
c: § Q) :J Eu E E 0 oate:L 0 ILl~ OC; Time:
>- ::l 0
- J
.....J
...J g~ 1. ENTER MAIN STEAM PRESSURE Z-oO-l 9£5' Z"'OO...J o O Q)x~ ~ ~ <{
C> Q) I-C -0 E PI-8013A . 9f~ fj/C. PYLe: Z C:"'O-Z I 9t~ rJIC~ ca co cc: Z o O..c _ ..c:'-0"'0 u-o .- PI-8023A P,-8023A'_--I:YL-I-ft-:-",0"'---f-F-.)L..!!'-=C.:!<-o-_ _ 1---02 1-"'02
<< oCt -==
c: C (5 co.):; 0:::: ca.b P'-8013B_""",=?~r~7---'fpr.....::...)~C(.:---_ PI-8013B ? 7 'vr)C{. rfr~ 4x{
,::;cc c: c:
0::: .Q 8
?f'(; I aoq oo.~.~ caco LL Q)..c:
U. zZ-~~W al..c
....... :=:o
- 0 PI-8023B r-f f'ff' f' 1/iJJ)c.*/
6ft PI-8013C fl)J)C .
-~3:w ffi~ _ __
PI-8023C_.J.£.......I./--~----,'fl£-LL.::>,.. 0::: *c ~ oO~2!+/-: ~2 ~ df LL --g
- u. -{ij a:0::: PI-8023C
(:)G, Q) u al 0 W
~ '5.. > PI-8013D_.v.,..~-L;-_.p:t!.....:..!o<:'--
PI-8013D _ __ alc.. W Q)a. o !;;:
~
..... co 0 ....
Ci>:=" 0 Qi=-o co OJ ca I-
<{
PI-8023D f1 PI-8023D_...lL--w--=----,f4J-='---_ f/tA __ f[1 COMPUTE TOTAL FEEDWATER FLOW
.2~t/l!f - 2.1/93
.2,t/q 2. t/93
( A. A Loop :2"
;)/5D SD ~~~~_
xx~~~:"",-_729178=.:l4~3 + 29178 =.;J. 493 6 x 106 Ib/hr x10 1D; 5 Zr; B L(OFTop901~ x.....,'.i:::::DD~~'!=f-.~_ (FT 9011-2 (E I}OA..QS) , or F-9011) PT K,:(:;A
;El3QAQ.S), or F-9011)
B. BLoop ~"LV e2~LO ~ xc>L. 3 . 729050
+ = 2.. 5Zr; 1066 Ib/hr 29050 =2..
(FT 9W1-2 (ERDADS), or F-9021) (DCS PT FB) S tZ.5<Z
/J
,2~S.,]
- tX-~S0 MFw=~OI M 5.0,?,,] x 1066 1b/hr Fw =c5. . 01 -- 5".023 p... 2-(Total A+B Loop)
Average e Steam Pressure = 709'1() 9f 7TOTAL divided by 8 = tK tJ-(; ?75-' AVERAGE I L ( ,;,Ci'V 5...;../ (J;Cj,V ENTER FEEDWATER TEMPERATURE TR-09-5B Channel 1 J~2.fFJt,2.. .?oF (behind.RTGB 202) (behindRTGB 21b2-.loF Channel Jb{/.I°F Average Feedwater Temperature 7U !</ divided by 2 = '3Cb. J h 7 (;r,1, 181~}) (p;" IBZ~}) AVERAGE From the steam tables, entEir entfir Jh~ the enthalpy of th~ average steam pressure: hSTEAM=/Jn.~ hSTEAM= 1/7'. ( BTUllbm BTU/Ibm VI r'. f lJl "-f ,. 1/ IllJ"b)) rz.)) C~culate the heat 0y!put o~ut due to steam flow (QSTEAM): SLOPS 11 i7. 3 ~()I-f.{)lii106, I-:--'c....::....:.=-- 03'7 5'97 106 fo/-f.olfiJ06,D),592106 6 106 BTu/hr BTU/hr J) DATE,_ _ _ __ DATE. hSTEAM X X [MFW 7 --, , !V1BO] s-'1 ~'.I-~'itG. I;
~BDl == QSTEAM( f1,/c,.I-1)CJtG.!; DOCT Procedure From the steam fables, c.
eLI.tJ. 'j - f.'} tl5rJ rJ t~bles, enter the enthalpy of the average feedwater temperature: DOCN 2-3200020 SYS OPS hFw = BTU COMP COMPLETED ITM_ _ _ __ ITM. ( Ibm Calculate the heat feedwater feedW!ter.,~ ~ input (Q FW): (QFW): ( I
'-->/--_
6 xX 106 ) x (]1</-J~ =
<71'L-'b;) 6 x 106 BTU
.. V;tl
&.6! _ }:~2.-iJ hFw hFw I/ QFWQFW hr hr
~~~~---------7~-=~--~~r---------------------------~
~~~-----rJ-'~1~J~-)-~~r7Jr-----------------~
REVISION NO.: PROCEDURE TITLE: 31 PRIMARY SYSTEM MANUAL CALORIMETRIC PROCEDURE NO.: NO.: 2-3200020 ST. LUCIE UNIT 2 DATA SHEET 1 (Page 2 of 3) TABLE 1 ircle the total blowdown flow from the S/Gs and the correspond Circle correspondin output: nterpolation is not required, circle the matching blowdown flow below or the lower flow if (Interpolation between flow rates, to be conservative.) IngnA/<"gn Total blowdown flow Mass flow Heat output from (both steam generators) of blowdown M BO blowdown QBO
~OG~
~OGeM... .019799 x 1061bmL~
~
120 GPM a~9.059398 x . x ~g: :~~~~ r
~~~;:...:..:~~~~
160 GPM .079198 x 106 Ibm/hr 200 GPM .098997 x 106 Ibm/hr 240 GPM .118797 x 106 Ibm/hr Calculate the heat output from the core: ( Iff charging and letdown flow is secured for any reason, use a value cha~ging and letdown flow is If char,ging is. operating normally, use a valu of 46.905 or QOTHER. 51f/~.*1-1Qfl(,..l- 1~7]-
/073- .. ... . .... .,_...
[~"./ [~"./ - //:;J'J 11. Ii. I/:',!'J + 'If( ~'.~x tjS( - ~C. 7D.JJ x 1066 BTU = = x x 1066 IDi
~"2iS-'I2J:S -
QSTEAM QFW QBO QOTHER hr QCORE ,/Z6r;'J hr - '/2(;~ ('5"""" NOTE Calculated Power in percent must be within 2% of DCS Calorimetric power. If not, noti the US I SM. Initiate W/R for ICM to ad'ust 11 ~ Tower. Calc~te percentper£§nt core power: powe.s-) 'It: LJ.t: 5 -1),.1
-- !)C.J -,
Core Power =
-,,'l*tf:tS/
1/];Jr-lj?4I':c.?1
*'12:-Jf 106 divided by (92.128 x 1066 ) =
x 1ov-divided I* I 0 Yo I Calorimetric Manual QCORE Power Record DCS Calorimetric Power (QRP % 10 minute ave), (QRPP (instantaneous) should be used below if the reactor has not been in a stable power configuration for at least 10 minutes): DCS Calorimetric Power = 5/ L"C, . ) (j I } %
REVISION NO.: PROCEDURE TITLE: PAGE:,> 31 PRIMARY SYSTEM MANUAL CALORIMETRIC PROCEDURE NO.: 2-3200020 ST. LUCIE UNIT 2 DATA SHEET 1 (Page 3 of 3)
- 2. On the first Monday of the quarter, complete Data Sheet 2. ;lit}
jl/tJ ::;0 Initial ,W
.~
~
¢- --
- 3. On the first Monday of the quarter, complete Data Sheet 3. ;.0 w
~
I have reviewed the requirements of this procedure including other surveillances performed during this procedure, if any (i.e., datasheet(s), PMT sheet(s), etc.). Any deviation(s) n(s) found during the performance of this procedure has (have) been listed an appropriate ppropriate actions and notifications made. A A~~ ~,-.~'
'kco RCO US/SM
(
Procedure No.
*J;;;n~*;\
ST. LUCIE UNIT 2 2-320002'0 OPERATING PROCEDURE Current t-'UUISlnn No. FPL SAFETY RELATED CONTINUOUS USE Effective Date 08/26/08
Title:
PRIMARY SYSTEM MANUAL CALORIMETRIC Responsible Department: REACTOR ENGINEERING REVISION
SUMMARY
Revision 31 - Incorporated PCR 08-3524 to change the periodicity of the performance of data sheet 2 and 3. (Carlos Diaz, 08/18/08) Revision 30 - Incorporated PCR 08-3603 for PCM 08096 to change device number. (Brad Bishop, 07/24/08) Revision 29829B - Incorporated MC 08-0040 to change verbiage to clarify when determining blowdown flow values. (Clyde Price, 01/14/08) Revision 29A - Incorporated PCR 07-4414 to label Table 1 and added verbage for ( clarification in determining values when in between flow rates. (Clyde Price, 12/17/07) Revision 29 - Incorporated PCR 07-1166 for PCM 05137 to reflect new blowdown mass flow rate and heat output values in Data Sheet 1. (Tim Bolander, 09/10/07) Revision 28C - Incorporated MC 07-2322 to add FlO stamp and SOPS stamp on Data Sheet 1. (Clyde Price, 07/10/07) Revision 28828B - Incorporated PCR 07-0228 to replace FR-9011 with F-9011. (Dave Pottorff, 01/17/07) Revision 28A - Incorporated PCR 06-3831 to add comment to note box. (R. Poppell, 10/26/06) Revision FRG Review Date Approved By Approval Date S...LOPS S....LOPS 0 03/08/83 C. M. Wethy 03/08/83 DATE Plant General Man DOCT Revision FRG Review Date Approved By Approval Date DOCN 31 N/A SYS Plant General Manager COM John Harmon 08/25/08 ITM Authorized Approver N/A Authorized Approver (Minor Correction)
"f!;{~ . "
REVISION NO.: PROCEDURE TITLE: PAGE,,: 31 PRIMARY SYSTEM MANUAL CALORIMETRIC PROCEDURE NO.: 2 of 13
~.
2-3200020 ST. LUCIE UNIT 2 TABLE OF CONTENTS SECTION 1.0 PURPOSE ...................................................................................................... 3
.0
2.0 REFERENCES
............................................................................................... 3 .0 3.0 PREREQUISITES .......................................................................................... 4 .0 4.0 PRECAUTIONS I/ LIMITATIONS .................................................................... 4 5.0 RECORDS REQUIRED ................................................................................. 4 6.0 INSTRUCTIONS ............................................................................................ 5 7.0 INFREQUENT OPERATIONS ....................................................................... 8 DATA SHEETS DATA SHEET 1 ........................................................................................................ 9
( DATA SHEET 2 ...................................................................................................... 12 DATA SHEET 3 ...................................................................................................... 13
REVISION NO.: PROCEDURE TITLE: PAGE:';':1t c 31 PROCEDURE NO.: 2-3200020 PRIMARY SYSTEM MANUAL CALORIMETRIC ST. LUCIE UNIT 2 3 of 13 1.0 PURPOSE 1.1 To provide a method to accurately calculate the reactor thermal power manually under steady state conditions.
2.0 REFERENCES
NOTE One or more of the following symbols may be used in this procedure:
§ Indicates a Regulatory commitment made by Technical Specifications, Condition of License, Audit, LER, Bulletin, Operating Experience, License Renewal, etc. and shall NOT be revised without the required Focus review and appropriate approval.
11~ Indicates a management directive, vendor recommendation, plant practice or other non-regulatory commitment that should NOT be revised without consultation with the plant staff.
'+/-' Indicates a step that requires a signn off on an attachment.
( 2.1 Miscellaneous Documents
- Unit 2 Drawing 2998-G-056-Sh. 2, Heat Balance at Rated Power
- Steam Tables
- PCM 028-295 Installation of an Acoustic Flow Meter for Measuring Feedwater Flow
** 111
~1 CR 01-0344
REVISION NO.: PROCEDURE TITLE: 31 PRIMARY SYSTEM MANUAL CALORIMETRIC PROCEDURE rl"'.uvcu'U,",C NO.: 2-3200020 ST. LUCIE UNIT 2 3.0 PREREQUISITES 3.1 The reactor and related equipment should be in a steady state condition with plant power greater than or equal to 20% of Rated Thermal Power. 4.0 PRECAUTIONS I LIMITATIONS 4.1 Reactor power, RCS temperature, steam generator level, blowdown, charging and letdown rates should be maintained as constant as possible during data taking. 4.2 If core power is determined to be above licensed or administrative limits, steps should be taken to reduce it to an acceptable level. 4.3 Any parameters that are determined by redundant inputs should be compared with each other to establish each as a believable or non-believable value. 4.4 Te should be between 548°F and 548.8°F for operation at rated thermal power. Tc 5.0 RECORDS REQUIRED ( 5.1 The data sheets from this procedure shall be maintained in the plant files in accordance with QI-17 -PSL-1, Quality Assurance Records. 5.2 Forms similar to those contained in this procedure may be used provided they contain all of the essential information as the forms attached. (
REVISION NO.: PROCEOURE TITLE: PAGE: 31 PRIMARY SYSTEM MANUAL CALORIMETRIC ( PROCEOURE NO.: PROCEDURE of 13'" 2-3200020 ST. LUCIE UNIT 2 6.0 INSTRUCTIONS 6.1 Establish plant conditions per section 3.0 NOTE Use Data Sheet 1, Manual Calorimetric Calculation, in conjunction with the ste stepss below. NOTE Reactor core thermal power shall be determined using the following equation: QCORE + QFW+ QFW + QRCP + QCH + QpZR-QpZR - Qccw-Qccw - QSTEAM-QSTEAM - QBD-QBO - QLD- = Q LD - L = 0 M is mass flow rate in Ibm/hr, h is enthal enthalpy in BTU/Ibm NOTE The DCS feedwater data points are on the SG 2A (2B) FW BLOWDOWN CALCS screens accessible throuthrough h the POWER MENU. 6.2 Collect all data as required by Data Sheet 1, Manual Calorimetric Calculation. Record the time the data was taken. ( 6.3 Compute total feedwater flow, average feedwater temperature and average main steam pressure. 6.4 Compute the net heat output due to steam flow (QSTEAM). NOTE QSTEAM =(M Fw - MBD) X hSTEAM Mso)
- 1. Determine the enthalpy for the average main steam pressure (hsTEAM)'(hSTEAM).
- 2. Determine the mass flow of blowdown by using Table 1, Data Sheet 1, to convert GPM to Ibm/hr.
- 3. Subtract blowdown mass flow (M BD BO ) from feedwater mass flow (M Fw) to obtain steam flow.
- 4. Multiply the quantity from step 6.4.3 times the enthalpy from step 6.4.1.
This resultant is QSTEAM.
REVISION NO.: PROCEDURE PROCEOURE TITLE: PAGE: f, 31 PRIMARY SYSTEM MANUAL CALORIMETRIC 6 of 1 PROCEOURE NO.: PROCEDURE 2-3200020 ST. LUCIE UNIT 2 NOTE The effect of moisture carryover has been neglected because it is difficult to measure and accounts for less than 0.2% power. r. 6.5 Compute the feedwater heat input (QFW).(OFW). NOTE QFW= OFW= MFwx hFw hFw
- 1. Determine the enthalpy for the average feedwater temperature (hFw).
- 2. Multiply the total feedwater flow (MFw)
(M Fw) times hFw (from above). This result is hFw(from QFW. OFW. 6.6 . Determine the Blowdown Heat Output (QBO). (OBO). NOTE QBO M BO OBO = X h BO Bo X Bo hBO blowdown enthal hBo (blowdown enthalpy) is assumed to be constant. (
- 1. Determine total (both S/Gs) blowdown flow rate (M BD Bo), from Table 1 of Data Sheet 1.
- 2. Enter Table 1 of Data Sheet 1 1and and determine blowdown heat output (QBD). (OBO).
(Interpolation is not required.) 7 6.7 The following is the determination for the constant used to represent other net heat losses and gains (QOTHER). (OOTHER). 6 6 OOTHER = 46.905 x 106 STU/hr-(48.952 QOTHER- BTU/hr (48.952 x 106 BTU/hr if charging and letdown flow is secured for maintenance) NOTE QOTHER OOTHER = QRCP ORCP + QCH OCH + QpZR OPZR - Qccw Occw- - QLD OLD- - L
- 1. The following constant is supplied for the Reactor Coolant Pump heat flow (QRCP).
(ORCP). 6 QRCP ORCP = 55.951 x X 106 BTU/hr
REVISION NO.: PROCEDURE TITLE: 31 PRIMARY SYSTEM MANUAL CALORIMETRIC
**,7 of 13 PROCEDURE NO.:
2-3200020 ST. LUCIE UNIT 2 6.7 (continued)
- 2. Charging heat input (OCH)
(QCH) is defined by the equation OCH = QCH = MCH X X h cH , where hCH, MCH is the mass flow rate and hCH is the enthalpy. The following constant is based onon one o.ne charging pump operating at a VCT temperature of 105°F.
=
QCH = 1.174 X OCH 6 X 106 BTu/hr BTU/hr
- 3. The constant for the net Pressurizer heat input (QpZR) (OPZR) was determined as a result of Hot Ops II.
OPZR QpZR =0.592 x 1066 BTU/hr
- 4. The constant for the Component Cooling Water heat output (Qccw) (Occw) was determined as a result of Hot Ops II.
= 6 Qccw = 0.458 Xx 106 BTu/hr Occw
- 5. Letdown heat flow (OLD)
(QLD) is defined by the equation OLD = QLD MLD X Xh LD , where hLD,
.MLD is the letdown mass flow rate and hLD is the letdown enthalpy. The following constant is based on one charging pump operating with a letdown
( temperature of 250°F out of the Regenerative Heat Exchanger and 100°F out of the Letdown Heat Exchanger. QLD = 3.221 OLD X 1066 BTu/hr BTU/hr
- 6. The following constant is supplied to account for all piping, radiation and other losses (L).
=
L = 7.133 X 6 X 106 BTu/hr 6.8 Calculate the heat output from the core (OCORE). (QCORE). NOTE QCORE = OCORE =QSTEAM - QFW OFW + OBD QBD - OOTHER QOTHER 6.9 Calculate percent core power NOTE Core Power (%) =OCORE / 92.128 xX 1 QCORE 192.128 10066 (a conversion factor) 10 6.10 Complete the attached data sheets, as required. END OF SECTION 6.0
ON NO.: PROCEDURE TITLE:
- ):~'~ :
REVISION PAGE: , 31 PRIMARY SYSTEM MANUAL CALORIMETRIC ;-\, l' ( .. 8 of 1 ,<:, PROCEDURE NO.: ;-.<;: 2-3200020 ST. LUCIE UNIT 2 ,
.0 7.0 INFREQUENT OPERATIONS ,~>:c
.i*t~'~>
7.1
.1 None END OF SECTION 7.0
(
REVISION NO.: PROCEDURE TITLE: PAGE: .. 70 *
"'*;;v 31 PRIMARY SYSTEM MANUAL CALORIMETRIC 9 of 13 IPRI:>CEDUF~E PROCEDURE NO.:
2-3200020 ST. LUCIE UNIT 2 IF DATA SHEET 1 C ,fL
.Q~ (Page 1 of 3) 1955 C E Date: __1 _ _1I_ _ Time: _ _ __
Q) E u
- l Date: - -I- --- Time: - ---
>- :::l 0 ..J g~ 1. ENTER MAIN STEAM PRESSURE Z"OO....J o:g,~ ~ PI-8013A.
PI-8013A_ _ _ _ _ _ _ _ _ __ Z c"o !:: C\'l C Z O..c::*- PI-8023A. _ U"O PI-8023A_ _ _ _ _ _ _ _ _ __ I-"O~
<<CO PI-8013B_ _ _ _ _ _ _ _ _ __ -=C\'l~ PI-8013B -=cc 0::: .Q 8 o*~ C\'l PI-8023B_ _ _ _ _ _ _ _ _ __
LL. Q)..c:: z . . ;:::
-~$:UJ 0 PI-8013C._ _ _ _ _ _ _ _ _ _ __
PI-8013C 0::: *c,....... - oLL. -.g
~~ !:!:
c::: PI-8023C._ _ _ _ _ _ _ _ _ _ __ PI-8023C Q) u UJ
~ '5. > PI-8013D_ _ _ _ _ _ _ _ _ _ __
PI-8013D Q)o..UJ
.... C\'l I-0 ..... <C PI-8023D_ _ _ _ _ _ _ _ _ _ __
1D::::- 0 PI-8023D co 111 COMPUTE TOTAL FEEDWATER FLOW 6 ( A. A Loop (FT 9011-2 (ERDADS). (ERDADS), or F-9011) x
--==(D:"::C-=-S-='
(DCS PT PT=-K=-=F:-:A-:-) KFA) -
.;- 29178 =
x 106 Ib/hr 6 B. B Loop BLoop X X .;. 29050 =
.;- xX 106 Ib/hr (FT 9021-2 (ERDADS).
(ERDADS), or F-9021) (DCS PT KFB) 6 MFw = X 10 6 1b/hr (Total A+B Loop) Average Steam Pressure = _ ............_ _ _ divided by 8 = _______=-- TOTAL AVERAGE ENTER FEEDWATER TEMPERATURE TR-09-5B ChanneI11_ _ _of Channel 0F (behindRTGB 202) (bebindR'-GB202) Channel 2 of 1LJ.\""'~"'r'Q Average Feedwater Temperature divided by 2 = _-:-:-:==-:-=-::=- TOTAL AVERAGE From the steam tables, enter the enthalpy of the average steam pressure: hSTEAM= BTUllbm BTU/lbm Calculate the heat output due to steam flow (QSTEAM): luClIll..UIICnC S...LOPS S-LOPS 66 6 6 DATE. DATE_ _ _ __
- : : - : - - -10 106 106 BTUlhrBTU/hr hSTEAM X [MFw [MFW - MSDl =
Msol QSTEAM DOCT Procedure DOCN 2-3200020 From the steam tables, enter the enthalpy of the average feedwater temperature: SYS OPS hFw = h = BTU COMP COMPLETED ITM,_ _ _ _ __ Ibm ITM Calculate the heat feedwater heat input (Q (QFW): FW): 6 6 ( 1'-__ -- x 1 06 ) X ( X 10 ) = 106 BTU x 10 X MFw hFw QFW hr
SION NO.: REVISION PROCEOURE TITLE: PROCEDURE PAGE:
'\
31 PRIMARY SYSTEM MANUAL CALORIMETRIC PROCEOURE NO.: PROCEDURE o of 13 2-3200020 ST. LUCIE UNIT 2 DATA SHEET 1 (Page 2 of 3) TABLE 1 ircle the total blowdown flow from the S/Gs and the corresponding heat output: Circle Interpolation is not required, circle the matching blowdown flow below or the lower flow if (Interpolation
,n",nM"n between flow rates, to be conservative.)
Total blowdown flow Mass flow Heat output from (both steam generators) of blowdown M MBD Bo blowdown QBD QBO 6 Ibm/hr 9.728 X 106 6 BTu/hr 40GPM .019799 x 106 BTU/hr 6 Ibm/hr 19.456 x 106 6 BTu/hr 80GPM 80 GPM .039599 x 106 BTUlhr 6 120 GPM .059398 x 106 6 1bm/hr Ibm/hr 29.183 x 10 BTu/hr 6 BTUlhr 6 6 160 GPM .079198 x 6 10 1bm/hr Ibm/hr 38.912 x 106 BTu/hr BTU/hr 200 GPM .098997 x 6 Ibm/hr 106 48.639 x 106 6 BTu/hr BTU/hr 240 GPM .118797 x 106 6 Ibm/hr 58.367 x 106 6 BTu/hr BTUlhr Calculate the heat output from the core: ( Iff charging and letdown flow is secured for any reason, use a value of 48.952 for QOTHER. If charging and letdown flow is operating normally, use a value of 46.905 for QOTHER. [_ _ _ - - __ -_ -+---::-__ _ _-oil 1066 BTU = _ _~X~1~0~6
- - - --::-_-,1 x = __ -<.X,-1,...,,0,-6 BTU QSTEAM QFW QBD QBO QOTHER hr QCORE hr NOTE Calculated Power in percent must be within 2% of DCS Calorimetric power.
noti the US I SM. Initiate W/Rfor ICM to adjust LlT If not, notify LlT power. Calculate percent core power: Manual
=
Core Power = _ _----'x~1 = I xi 00=-66 divided by (92.128 xx 1066) = % L====':"'=---.l. I Calorimetric QCORE Power Record DCS Calorimetric Power (QRP % 10 minute ave), (QRPP (instantaneous) should be used below if the reactor has not been in a stable power configuration for at least 10 minutes): DCS Calorimetric Power = - = _- __ -_ -_ -_ -%
REVISION NO.: PROCEDURE TITLE: PAGE:. 31 PRIMARY SYSTEM MANUAL CALORIMETRIC 11 of 1 PROCEDURE NO.: 2-3200020 ST. LUCIE UNIT 2 DATA SHEET 1 (Page 3 of 3)
- 2. On the first Monday of the quarter, complete Data Sheet 2.
Initial
- 3. On the first Monday of the quarter, complete Data Sheet 3. ......
;:0 w
Initial I have reviewed the requirements of this procedure including other surveillances performed ed during this procedure, if any (i.e., datasheet(s), PMT sheet(s), etc.). Any eviation(s) found during the performance of this procedure has (have) been listed an deviation(s) appropriate nnrnnr*'*!:1 actions and notifications made. RCO US/SM (
REVISION NO.: PROCEDURE TITLE: 31 PRIMARY SYSTEM MANUAL CALORIMETRIC PROCEDURE NO.: 2-3200020 ST. LUCIE UNIT 2 DATA SHEET 2 (Page 1 of 1) NOTE This data sheet will compare feedwater temperature inputs into DCS calorimetric with feedwater temperature indications used in manual calorimetric.
- 1. Record feedwater temperature inputs from DCS.
PT T0910A1 _ __ _ _ PT T0910A2 _ __ _ _ PT T0910A3 _ __ _ _ Compare PTs T0910A1 through T0910A3 to be within 2°F, if greater than 2°F notify I&C. PT T091 081 _ __ T091081 ( PT T091 082 _ __ T091082 PT T091083 T091 083 ___ _ _ Compare PTs T091081 through T091083 to be within 2°F, if greater than 2°F notify I&C.
- 2. TR-09-S8 (behind RTG8 202).
Record feedwater temperatures from DCS and from TR-09-58 DCS PT TFA ----_ _ TR-09-S8 __ TR-09-58 Channel 1 ---- Ensure feedwater temperatures to be within 5°FSOF or notify I&C and US. DCS PT TF8 _____ _ TR-09-S8 TR-09-58 Channel 2 __ Ensure feedwater temperatures to be within 5°FSOF or notify I&C and US. I have reviewed the requirements of this procedure including other surveillances performed during this procedure, if any (i.e., datasheet(s), PMT sheet(s), etc.). Any deviation(s) found during the performance of this procedure has (have) been listed an appropriate actions and notifications made. RCO US/SM
PROCEDURE TITLE: 31 PRIMARY SYSTEM MANUAL CALORIMETRIC ( PROCEDURE NO.: 3 of 1
+'
2-3200020 ST. LUCIE UNIT 2 ,,",,'
;;'<;~'
DATA SHEET 3 ., '0'1 ~.; (Page 1 of 1) NOTE This data sheet will compare calorimetric and Delta T power as computed by the DCS for consistency. Delta T power is used as the input into the DCS Xenon calculation and as a backup reactor power signal supplied to BEACON. CAUTION Calorimetric power should have been stable (i.e., no power maneuvers were in progress) over the hour time period selected for the Delta T power comparison. com arison.
- 1. RECORD the following data from the DCS Hourly Reactor Poison and Power Log.
( Delta Power - - - - - - -% Avg Pwr Last Hour% - - - - - - -%
- 2. COMPARE Delta T power to AVG PWR Last Hour %.
- 3. if Delta T Power and AVG PWR Last Hour % DO NOT agree 1f if Delta T Power adjustment is desired, Then within 2 percent, or 1f NOTIFY I&C and SCE. (N/A if notification not required)
Initial I have reviewed the requirements of this procedure including other surveillances performed during this procedure, if any (i.e., datasheet(s), PMT sheet(s), etc.) Any deviation(s) found during the performance of this procedure has (have) been listed an appropriate actions and notifications made. . RCO US/SM
REGION II ST. LUCIE NUCLEAR PLANT INITIAL LICENSE EXAMINATION ADMINISTRATIVE JPM Develop ECO for 1B Containment Spray Pump NRC A-3 CANDIDATE ( EXAMINER NRC ADM JPM A-3 fRev tRev 0 PAGE 1
REGION" ST LUCIE NUCLEAR PLANT INITIAL LICENSE EXAMINATION ADMINISTRATIVE JPM Develop ECO for 1B Containment Spray Pump. KA Statement: Ability to obtain and interpret station electrical and mechanical drawings. KA#: 2.2.41 (3.5/3.9)
References:
P&ID 8770-G-088 SH 1 P&ID 8770-G-083 SH 2 CWO 8770-B-326 SH 290 CWD ADM-09.08 Operations In-Plant Equipment Clearance Orders 1-ADM-03.01 A Unit 1 Power Distribution Breaker List Operations Policy OPS-406 Clearance Database Conventions ( Candidate: Time Start -__ -- Name Time Finish - __ -_- Performance Rating: Sat Unsat Validation Time 25 minutes Examiner: Signature: Comments NRC ADM JPM A-3 IRev fRey 0 PAGE 2
REGION II REGION" ST LUCIE NUCLEAR PLANT INITIAL LICENSE EXAMINATION ADMINISTRATIVE JPM Develop ECO for 1 B CS LPSI pump Directions to the candidate for Administrative JPMS: I will explain the initial conditions and state the task to be performed. You will be allowed to use any reference normally available in the Control Room to complete the task. Ensure you indicate to me when you finish your assigned task by returning the material needed for the task that I provided to you. Initial Conditions Unit 1 is at 100% power. The 1B Containment Spray Pump has developed a pump cooler leak. The pump cooler is to be removed from service to repair the leak. No other equipment is out of service. NOMS is unavailable. ( Initiating Cue On the attached matrix, develop an ECO lAW ADM-09.08, 'Operations In-Plant Equipment Clearance Orders', for the 1B Containment Spray Pump that would facilitate repairing the pump cooler leak. Ensure tagged components are placed in the proper sequence. IRev 0 NRC ADM JPM A-3 fRey PAGE 3
Com~onent Component Name Position Tag (type) NOTE: Plus is NOT required for position to meet Critical Steps Control Room Control 1-CS-290 In Place Info Tag Switch for 1B Containment Spray Pump
*Breaker 1-20407 1B Containment Spray Racked Out Plus Danger I Pump
*1-V407124 1B Containment Spray Locked Closed Plus Danger Pump Suction Valve
*1-V07130 1B Containment Spray Locked Closed Plus Danger Pump Discharge Valve
*1-V07118 Equalizer Valve for Closed Plus Danger V07124
*1-V07135 Equalizer Valve for Closed Plus Danger V07130
*1-V07262 Isolation Valve for Sodium Locked Closed Plus Danger I Hydroxide Feed to Containment Spray Pump 1B Suction
*1-V07226 Primary Isolation Valve for Locked Closed Plus Danger Containment Spray Pump 1B Recic.
1B
*1-V14292 Isolation for Containment Locked Closed Plus Danger Spray Pump 1B CCW inlet
*1-V14295 Isolation for Containment Locked Closed Plus Danger
( Spray Pump 1B CCW outlet 120/208 VAC PP-102 4160V SWGR 1B2 Motor Open Plus Danger CKT -31 and Space Heaters NOTE: The following drains and vent may be opened and tagged but only ONE VENT and ONE DRAIN is needed to be opened to meet the critical step. #As per Ops Policy 406, 'Vents and Drains do not require a danger tag. It is preferred, but not required'. 1-V07123 Drain Valve on Open #Danger Containment Spray Pump 1 B Suction 1-V07128 Drain Valve on Open #Danger Containment Spray Discha~ge Pump 1B Discharge 1-V07125 Root Valve for PX Open #Danger 2B Upstream of SS-07-1 B (Vent) V07128 Drain Valve on Open #Danger Containment Spray Pump 1B Discharge V14294 Root Valve for PX Open #Danger 19B CSP 1B CCW Outlet (Vent) V14293 Root Valve for PX Open #Danger 19B CSP 1B CCW ( Inlet(Vent) Inlet (Vent)
*Cntlcal Step NRC ADM JPM A-3 IRev fRey 0 PAGE 4
(TO BE RETURNED TO THE EXAMINER UPON COMPLETION OF ANSWER) Initial Conditions Unit 1 is at 100% power. The 1B Containment Spray Pump has developed a pump cooler leak. The pump cooler is to be removed from service to repair the leak. No other equipment is out of service. NOMS is unavailable. Initiating Cue On the attached matrix, develop an ECO lAW ADM-09.08, 'Operations In-Plant Equipment Clearance Orders', for the 1B Containment Spray Pump that would facilitate repairing the pump cooler leak. Ensure tagged components are placed in the proper sequence. ( NRC ADM JPM A-3 IRev fRev 0
(TO BE RETURNED TO THE EXAMINER UPON COMPLETION OF ANSWER) Component Name Position Tag (type)
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eFPL St. Lucie Nuclear Plant Operations Training JOB PERFORMANCE MEASURE Operate Portable Radiation Monitoring Equipment RM-23P UNIT 2 ( NRC JPM (RO) A-4 IN-PLANT Revised By _ _ _ _ _ La_rrL-y_Ri_ch La_rr"--yR_i_ch_ _ _ __ Approval _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ Date NRC ADM JPM Rad. Control REV. 0 Page 1 of 12
Job Performance Measure A-4 (RO) Task Number: 03203020, Operate Portable Radiation Monitoring Equipment, RM-23P Task
Title:
Operate Portable Radiation Monitoring Equipment Faulted JPM: 0 Yes DYes I:8J No [gJ Facility JPM # 0321307 K/A Ratings: 2.9 K/A Statement: 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. Duty Areas: Radiation Control Task Information: N/A Task Standard: Applicant tests a RM-80 using the RM-23P per procedure. Evaluation Location Performance Level o D Simulator I:8J In Plant [gJ I:8J Perform [gJ [gJ I:8J Simulate D 0 Discuss o Lab D o Other D
References:
(
- 2-NOP-26.01, Radiation Monitors Validation Time: 30 Minutes Time Critical 0 Yes DYes I:8J No
[gJ Tools/Equipment/Procedures Needed:
- Standard Personal Protective Equipment
- Flashlight
- Radio
- 2-NOP-26.01, Radiation Monitors
- RM-23P pre-staged in TB Switchgear Specific Safety Concerns, PPE and Hazards associated with the task:
- None Radiological Protection and RWP Requirements:
- None NRC ADM JPM Rad. Control REV. 0 Page 2 of 12
Job Performance Measure A-4 (RO) Specific Directions:
- The Task you are to perform Operate Portable Rad Monitoring Equipment is:
- The performance level to be used for this JPM is PERFORM/SIMULATE.
(Circle the performance level being used for this implementation of the task.)
- This Is Not a time critical JPM.
- During the performance of the task, I will tell you which steps to simulate or discuss.
- I will provide you with the appropriate cues for steps that are simulated or discussed.
- You may use any approved reference materials normally available in the execution of this task, including logs.
- Indicate to me that you have finished the assigned task by returning the Candidate Cue Sheet that I provided to you.
Initial Conditions: Unit 2 is operating at 100% power. The PC-11 Group 15, Loop 3 is out of service. Initiating Cues: ( The US has directed you to obtain the local reading from the Main Steam Line Monitor 1 for the A Main steam Line, RIM-26-71, using the RM-23P in accordance with 2-NOP-26.01, Radiation Monitors. NRC ADM JPM Rad. Control REV. a 0 Page 3 of 12
Job Performance Measure A-4 (RO) PERFORMANCE CHECKLIST Start Time: 2-NOP-26.01, Radiation Monitors, Step 6.3, Portable RM-23 Operation STEP 1 (6.3.1) ENSURE Section 3.0, Prerequisites, completed. Standard: Operator CHECKS Section 3.0, Prerequisites. _ _ Sat Examiner's Section 3.0, Prerequisites is checked. Cue: _ _ Unsat Comments: (6.3.~ STEP 2 (6.3.2) REVIEW Section 4.0, Precautions / Limitations. Standard: Operator REVIEWS Section 4.0, Precautions / - - Sat Limitations.
- - Unsat Examiner's Section 4.0, Precautions I Limitations is reviewed.
Cue: ( Comments: Step 6.3.3, Perform a functional test of the monitor as follows: Critical Critical Step STEP 3 (6.3.3) PERFORM a functional test of the monitor as follows: Step A. ENERGIZE the monitor by plugging into a 120V AC power outlet.
- - Sat Standard: Operator ENERGIZES the RM-23P. __
- - Unsat Examiner's RM-23P is ENERGIZED.
Cue: Comments: (( Page 4 of 12 NRC ADM JPM Rad. Control REV. 0
Job Performance Measure A-4 (RO) Critical Step STEP 4 (6.3.3.8) TURN monitor power switch ON. Sat TURN monitor power switch ON. Standard: Unsat Examiner's Cue: Monitor power switch is ON. Comments: STEP 5 (6.3.3.C} {6.3.3.C} VERIFY the following indications:
- The display window reads 0.00 Sat
- Top left button (mR/hr display I control contrallight) light) is LIT.
Unsat Standard: VERIFY the display window reads 0.00. ( VERIFY the top left button (mR/hr display I control contral light) is LIT. Examiner's Cue: Window reads 0.00. Examiner's Cue: Top left button (mR/hr display I control light) is LIT. Comments: STEP 6 (6.3.3.D} ENTER the following command: 997 [ITEM]. Critical Step {6.3.3.D} Standard: ENTER command: 997 [ITEM]. Sat Examiner's Command: 997 [ITEM] entered correctly. Cue: Unsat I Comments: ( Page 5 of 12 NRC ADM JPM Rad. Control REV. 0
Job Performance Measure A-4 (RO) STEP 7 (6.3.3.E) VERIFY the following indications:
- The display window reads 8.888 + 18.
Sat Standard: VERIFY the display window reads 8.888 + 18. Unsat Examiner's The display window reads 8.888 + 18. Cue: Comments: STEP 8 (6.3.3.E) VERIFY the following indications:
- All (12) red, yellow and green indication Sat lights are LIT.
Standard: VERIFY all (12) red, yellow and green indication - Unsat lights are LIT. ( Examiner's All (12) red, yellow and green indication Cue: lights are LIT. Comments: (6,3.3.E) STEP 9 (6.3.3.E) VERIFY the following indications:
- All (8) white channel buttons on the Sat upper keyboard are LIT.
Standard: VERIFY all (8) white channel buttons on the Unsat upper keyboard are LIT. Examiner's All (8) white channel buttons on the upper Cue: keyboard are LIT. Comments: ( Page 6 of 12 NRC ADM JPM Rad. Control REV. 0
Job Performance Measure A-4 (RO) STEP 10 (6.3.3.El VERIFY the following indications:
- ITEM and MON push pushbuttons buttons on the Sat lower keypad are LIT Standard: VERIFY ITEM and MON push buttons on the Unsat lower keypad are LIT.
Examiner's Cue: ITEM and MON push buttons on the lower keypad are LIT. Comments: STEP 11 (6.3.3.F} If any of the lights fail to illuminate during the functional test, Then PERFORM the following:
- 1. Notify M& TE to investigate and repair the Sat defective unit.
Unsat Standard: DETERMINES all lights illuminate properly .. ( Examiner's Cue: All lights illuminate properly during functional test. Comments: Critical Step STEP 12 (6.3.3.G} TURN monitor power switch OFF. TURN monitor power switch OFF. Standard: Sat Examiner's Cue: Monitor power switch is OFF. Unsat Comments: ( Page 7 of 12 NRC ADM JPM Rad. Control REV. 0
Job Performance Measure A-4 (RO) c STEP 13 (6.3.3.Hl UNPLUG monitor. Critical Step UNPLUG monitor. Standard: Sat Examiner's Cue: Monitor is UNPLUGGED. Unsat Comments: (6.3.4.A} STEP 14 {6.3.4.A} PERFORM the following to obtain RM-80 readings: Standard: ENSURE the RM-23P power switch is OFF. Sat Examiner's Cue: The RM*23P RM-23P power switch is OFF. Unsat Comments: CAUTION The RM-80 enclosure contains energized ( equipment. Critical Step CAUTION STEP 15 {6.3.4.B} (6.3.4.B} The RM-80 enclosure contains energized equipment. Sat Unsat CONNECT the RM-23P power cord to the 120V AC outlet located at the lower right corner inside the RM-80 enclosure. Standard: CONNECT the RM-23P power cord to the 120V AC outlet located inside the RM-80 enclosure. Examiner's Cue: The RM*23P RM-23P power cord is connected to the 120V AC outlet. Comments: ( Page 8 of 12 NRC ADM JPM Rad. Control REV. 0
Job Performance Measure A-4 (RO) Critical Step STEP 16 (6.3.4.C) CONNECT the RM-23P signal cable to the RM-80 communications port. Standard: CONNECT the RM-23P signal cable to the RM- _ _ Sat 80 communications port. Examiner's Cue: The RM-23P signal cable is connected to the _ _ Unsat RM-80 communications port. Comments: Critical Step STEP 17 (6.3.4.D) PLACE the RM-23P power switch in ON. Standard: PLACE the RM-23P power switch in ON. Examiner's Cue: The RM-23P power switch is ON. _ _ Sat Comments: _ _ Unsat NOTE ( The top four white buttons are the channel select pushbuttons. The buttons are used to select the appropriate channel dependent upon what type monitor the RM-23P is reading. Buttons represented from left to right. MONITOR BUTTON 1 BUTTON 2 BUTTON 3 BUTTON 4 SSG/SSL Gas I Liquid PIG Gas Particulate Iodine WRGM Low Range Mid Range RanQe HiQh His:Jh Range Effluent Main Steam SLM Line Critical Step STEP 18 PRESS the appropriate channel select (6.3.4.E) pushbutton to obtain the desired reading. Standard: PRESS the appropriate channel select _ _ Sat pushbutton to obtain the desired reading. Determines 0.3 mr/hr. is normal background reading. _ _ Unsat Examiner's Button #1 for 2A SLM is depressed and the Cue: reading indicates 0.3 mr/hr. ( I ~ments: Comments: Page 9 of 12 NRC ADM JPM Rad. Control REV. 0
Job Performance Measure A-4 (RO) Critical Step STEP 19 When the desired readings have been obtained, (6.3.4.F) Then PERFORM the following to remove the RM-23P from service:
- 1. PLACE the RM-23P power switch in OFF. _ _ Sat Standard: PLACE the RM-23P power switch in OFF.
- - Unsat Examiner's The RM-23P power switch is OFF.
Cue: TELL candidate that another operator will complete the remaining steps of the procedure. Comments: END OF TASK ( Stop Time: _ _ Page 10 of 12 Page10of12 NRC ADM JPM Rad. Control REV. 0
Job Performance Measure A-4 (RO) CANDIDATE CUE SHEET Initial Conditions: Unit 2 is operating at 100% power. The PC-11 Group 15, Loop 3 is out of service. Initiating Cues: The US has directed you to obtain the local reading from the Main Steam Line Monitor 1 for the A Main steam Line, RIM-26-71, using the RM-23P in accordance with 2-NOP-26.01, Radiation Monitors. ( ( Page 11 of 12 NRC ADM JPM Rad. Control REV. 0
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eFPL St. Lucie Nuclear Plant Operations Training JOB PERFORMANCE MEASURE Perform Immediate Operator Actions For a Dropped New Fuel Element in the New Fuel Area UNIT 1 ( NRC JPM (SRO) A-4 Revised By _ __ _____ _ ----'L=a--"-rry'----'-R~ic_'__'___h L_a--->'rrY'--R_ic_h_ _ __ Approval _ _ _ _ _ _ _ _ _ __ Date ( NRC SRO ADM JPM Rad. Rad, Control REV. REV, 0 Page 1 of 8
Job Performance Measure Task Number: Task
Title:
Perform immediate operator actions for a dropped new fuel element. Faulted JPM: DYes I:8J [g] No Facility JPM # K/A Ratings: 3.8 K/A Statement: 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. Duty Areas: Radiation Control Task Information: N/A Task Standard: Perform immediate operator actions for a dropped new fuel element. Evaluation Location Performance Level D Simulator D In Plant D Perform I:8J Simulate D Discuss [g] D Lab ( I:8J LX] Other Classroom
References:
- 1-1600030 Accidents Involving New or Spent Fuel.
Validation Time: 5 Minutes Time Critical DYes I:8J No [g] Tools/Equipment/Procedures Needed:
- 1-1600030 Accidents Involving New or Spent Fuel.
Specific Safety Concerns. Concerns, PPE and Hazards associated with the task:
- None Radiological Protection and RWP Requirements:
- None NRC SRO ADM JPM Rad. Control REV. 0 Page 2 of 8
Job Performance Measure Specific Directions:
- The Task you are to perform Perform Immediate Operator Actions For a Dropped is: New Fuel Element in the New Fuel Room
- The performance level to be used for this JPM is PERFORM/SIMULATE.
(Circle the performance level being used for this implementation of the task.)
- This Is Not a time critical JPM.
- During the performance of the task, I will tell you which steps to simulate or discuss.
- I will provide you with the appropriate cues for steps that are simulated or discussed.
- As these are IMMEDIATE OPERATOR ACTIONS REFERENCES ARE NOT ALLOWED.
- Indicate to me that you have finished the assigned task by returning the Candidate Cue Sheet that I provided to you.
Initial Conditions: Unit 1 is operating at 100% power. New fuel is being off loaded in the new fuel area. Initiating Cues: ( New fuel is being off loaded in the new fuel area and EM is performing work in the spent fuel area. You are the SRO in charge of the new fuel off load on to the ground floor next to the container. As a new fuel bundle is being raised, the grapple separates and the new fuel element falls from an elevation of 20 feet and lands in a horizontal position. You are to carry out the IMMEDIATE OPERATOR ACTIONS pertaining to this event. ( NRC SRO ADM JPM Rad. Control REV. 0 Page 3 of 8
Job Performance Measure A-4 PERFORMANCE CHECKLIST Start Time: 1-1600030 Accidents Involving New or Spent Fuel. STEP 1 {7.4.} 11 the accident occurred in the New Fuel Area, Then If perform the following: A. Ensure all material is left in a safe condition. Standard: Ensures all material is left in a safe condition. Sat Examiner's All material is in a safe condition. Unsat Cue: After step 'C' immediate operator actions DO NOT Examiner's have to be in the specific order of the procedure. Note: Comments: ( STEP 2 {7.4.} 11 the accident occurred in the New Fuel Area, Then If perform the following: Critical Step B. Evacuate the New Fuel Area. Sat Standard: Announce evacuation of the New Fuel Area. Unsat Examiner's Evacuation of the New Fuel area has been Cue:
-- announced.
Comments: Page 4 of 8 NRC SRO ADM JPM Rad. Control REV. 0
Job Performance Measure A-4 {7 A.} STEP 3 (7.4.} If if the accident occurred in the New Fuel Area, Then Critical Step perform the following: C. Ensure the Spent Fuel Area is evacuated Sat Standard: Ensures all personnel have left the Spent Fuel Pool Unsat Area. Examiner's All personnel have left the Spent Fuel Pool area. Cue: Comments: Critical Step I STEP 4 (7.4.} {7.4.} If if the accident occurred in the New Fuel Area, Then perform the following: D. Ensure all FHB doors are closed and evacuate the Sat area. Unsat Ensures all FHB doors are closed and evacuates the Standard: area. Examiner's FHB doors are closed and the area is evacuated. Cue: Comments: Page 5 of 8 NRC SRO ADM JPM Rad. Control REV. 0
Job Performance Measure A-4 STEP 5 (7A.} {7.4.} 11if the accident occurred in the New Fuel Area, Then perform the following: Critical Step E. Notify HP to perform a radiological assessment. Sat Standard: Notifies HP to perform a radiological assessment. Unsat Examiner's HP acknowledges to perform a radiological Cue:
-- assessment.
Comments: STEP 6 {7A.} {7.4.} 11 the accident occurred if in the New Fuel Area, Then perform the following: Critical Step F. Notify Reactor Engineering Sat ( I Standard: Notifies Reactor Engineering Unsat Examiner's Reactor Engineering has been notified. Cue: Comments: STEP 7 (7A.} (7.4.} 11 the accident occurred in the New Fuel Area, Then if perform the following: Critical Step G. Notify Security Sat Standard: Notifies Security Unsat Examiner's Security has been notified. Cue: Comments: ( Page 6 of 8 NRC SRO ADM JPM Rad. Control REV. 0
Job Performance Measure A-4 STEP 8 (7.4.) {7.4.} If the accident occurred in the New Fuel Area, Then Critical Step perform the following: H. Notify the SM to determine reportability
- - Sat requirements.
_ _ Unsat Standard: Notifies SM Examiner's SM has been notified. Cue: Comments: END OF TASK Stop Time: _ _ (( Page 7 of 8 NRC SRO ADM JPM Rad. Control REV. 0
Job Performance Measure A-4 CANDIDATE CUE SHEET Initial Conditions: Unit 1 is operating at 100% power. New fuel is being off loaded in the new fuel area. Initiating Cues: New fuel is being off loaded in the new fuel area and EM is performing work in the spent fuel area. You are the SRO in charge of the new fuel off load on to the ground floor next to the container. As a new fuel bundle is being raised, the grapple separates and the new fuel element falls from an elevation of 20 feet and lands in a horizontal position. You are to carry out the IMMEDIATE OPERATOR ACTIONS pertaining to this event. ( ( Page 8 of 8 NRC SRO ADM JPM Rad. Control REV. 0
ST. LUCIE NUCLEAR PLANT INITIAL LICENSE EXAMINATION ADMINISTRATIVE JPM DETERMINE PROTECTIVE ACTION RECOMMENDATIONS A-6 CANDIDATE _ _ _ _ _ _ _ _ _ _ _ __ { EXAMINER
ST LUCIE NUCLEAR PLANT INITIAL LICENSE EXAMINATION ADMINISTRATIVE JPM DETERMINE PROTECTIVE ACTION RECOMMENDATIONS KA Statement: Knowledge of emergency plan protective action recommendations. KA #: 2.4.44 (4.4) Knowledge of emergency plan protective action recom mendations. recommendations.
References:
EPIP-08 Off-Site Notifications and Protective Action Recommendations Candidate: Time Start - - - Name Time Finish - - - Performance Rating: Sat Unsat ( Validation Time 15 minutes Examiner: Signature: Comments ST. LUCIE TRAINING DEPT. ( ADMINISTRATIVE JPM PAGE 2
ST LUCIE NUCLEAR PLANT INITIAL LICENSE EXAMINATION ADMINISTRATIVE JPM DETERMINE PROTECTIVE ACTION RECOMMENDATIONS Directions to the candidate for Administrative JPMS: I will explain the initial conditions and state the task to be performed. You will be allowed to use any reference needed to complete the task. Ensure you indicate to me when you finish your assigned task by returning the material needed for the task that I provided to you. Initial Conditions Unit 1 is experiencing a total loss of feedwater and a loss of offsite power. The 1A HPSI pump was out of service prior to the event. The crew just initiated once through cooling and the 1B HPSI pump tripped and is unable to be started. RCS pressure is rising and currently is 1620 psia. At time 0230, Containment High Range Monitor indicates 1.67 x 10 55 R/hr. A containment vacuum breaker has failed resulting in a release path from the ( containment. At time 0445 Initial manual off-site dose reports have come in and are as follows (only TEDE values are available at this time; CDE values are to be determined later): 1 mile: 1800 mrem 2 miles: 950 mrem 5 miles: 400 mrem Wind direction is from 78° Initiating Cue The Shift Manager has directed you to determine Protective Action Recommendations for the next State of Florida Notification Form update. THIS IS A TIME CRITICAL JPM ( ST. LUCIE TRAINING DEPT. ADMINISTRATIVE JPM PAGE 3
START TIME: _ _ __ Attachment 2 of EPIP-08; "Off-Site Notifications and Protective Action Recommendations" STEP 1: 2. Determination of PARs A. Instructions for selecting PARs Based on Plant Conditions
- 1. Refer to the PARs Based on Plant Conditions. - - SAT STANDARD: REFER to page 5 of 7 on Attachment 2 - - UNSAT COMMENTS:
STEP 2: 2. Determination of PARs (cont.) CRITICAL A. Instructions for selecting PARs Based on Plant Conditions (cont.) STEP
- 2. Begin in the upper left hand corner of the chart by answering the General Emergency (GE) question.
- - SAT STANDARD: DETERMINE that a General Emergency Exists.
- - UNSAT
( COMMENTS: STEP 3: 2. Determination of PARs (cont.) CRITICAL A. Instructions for selecting PARs Based on Plant Conditions (cont.) STEP
- 3. Correctly answer the questions until you reach one of the boxes that provide PAR information based on plant conditions.
- - SAT STANDARD: DETERMINE that PARs based on severe core damage.
- - UNSAT COMMENTS:
ST. LUCIE TRAINING DEPT. ( ADMINISTRATIVE JPM PAGE 4
STEP 4: 2. Determination of PARs (cont.) A. Instructions for selecting PARs Based on Plant Conditions (cont.)
- 4. Circle the box with the appropriate PARs.
- - SAT STANDARD: CIRCLES the PARs based on severe Core Damage.
- - UNSAT COMMENTS:
STEP 5: 2. Determination of PARs (cont.) A. Instructions for selecting PARs Based on Plant Conditions (cont.)
- 5. J£ there is no release, Then transfer the PARs to the State Notification Form. Section 11. The sectors affected can be - - SAT determined by referring to number 8, weather data in Attachment 1A. Directions for completing the Florida Nuclear Plant Emergency Notification Form. - - UNSAT STANDARD: DETERMINE that a release is occurring, step is N/A.
COMMENTS: ( STEP 6: 2. Determination of PARs (cont.) A. Instructions for selecting PARs Based on Plant Conditions (cont.)
- 6. J£ a release is involved, Then go to Section B, Instructions for PARs based on Off-site Dose, below. - - SAT STANDARD: DETERMINE that a release is occurring.
- - UNSAT COMMENTS:
ST. LUCIE TRAINING DEPT. ( ADMINISTRATIVE JPM PAGE 5
STEP 7: 2. Determination of PARs (cont.) B. Instructions for selecting PARs Based on Off-Site Dose.
- 1. PARs are based on Total Effective Dose Equivalent (TEDE or total dose) and /I or the Committed Dose Equivalent - - SAT (CDE, throid dose). Do NOT use dose rate values.
- - UNSAT STANDARD: DETERMINE TEDE to be used for selection of PARs.
COMMENTS: STEP 8: 2. Determination of PARs (cont.) B. Instructions for selecting PARs Based on Off-Site Dose (cont.)
- 2. See below:
- a. If performing dose calculations using the Class A Model, - - SAT Then in Forecast Mode, print the State Notification Form Summary for computer generated PARs.
( - - UNSAT STANDARD: DETERMINES Dose calculations were performed manually as per initiating cue. COMMENTS: 8T. ST. LUCIE TRAINING DEPT. ( ADMINISTRATIVE JPM PAGE 6
STEP 9: 2. Determination of PARs (cont.) B. Instructions for selecting PARs Based on Off-Site Dose (cont.)
- 2. See below: SAT
- b. If dose calculations were performed manually using EPIP-09, Off-site Dose Calculations, Then calculate TEDE and CDE in accordance with the procedure. UNSAT
- 1. Refer to the PARs Based on Manual Dose Calculation flowchart.
(a) Compare the TEDE dose at 1 mile with the values on the flowchart. Enter the chart at the appropriate dose level by determining if the dose is between 500 and 999 mrem or between 1000 and 4999 mrem or 5000 mrem or greater. (b) From the selected dose level, move to the right on the chart to the first column, 0-2 miles. The PAR provided corresponds to the calculated TEDE at 1 mile. STANDARD: DETERMINE from the cue sheet that the 1 mile TEDE dose is 1800 mrem and ENTER the TEDE DOSE column at the TEDE 1000-4999 mrem dose level. Under the 0-2 Miles column, DETERMINE that 'EVACUATE ALL" is the correct ( classification. DETERMINE from Attachment 1 1AA that the affected sectors are
'LMNP' (this is the critical part of this step).
DETERMINE from the cue sheet that the 2 mile TEDE dose is 950 mrem and ENTER the TEDE DOSE column at the TEDE 500-999 mrem dose level. Under the 2-5 Miles column, DETERMINE that 'SHELTER Sectors Affected" is the correct classification. COMMENTS: ST. LUCIE TRAINING DEPT. ( ADMINISTRATIVE JPM PAGE 7
STEP 10: 2. Determination of PARs (cont.) B. Instructions for selecting PARs Based on Off-Site Dose (cont.)
- 2. See below: - - SAT
- c. Enter the PAR in the 0-2 mile block on the TEDE DOSE table below the flowchart. The sectors affected can be determined by referring to number 8, Weather - - UNSAT Data, in Attachment 1A. Directions for completing the Florida Nuclear Plant Emergency Notification Form.
STANDARD: ENTER the PAR information as per the attached Answer Key sheet in the TEDE DOSE table. DETERMINE from Attachment 1A 1A that the affected sectors are 'LMNP' (this is the critical part of this step). COMMENTS: STEP 11: 2. Determination of PARs (cont.) B. Instructions for selecting PARs Based on Off-Site Dose (cont.)
- 2. See below: - - SAT
( d. Continue to determine the corresponding PAR at 2-5 miles using the calculated 2 mile TEDE, at 5-10 miles using the calculated 5 mile TEDE and the 10 - - UNSAT mile plus (To Be Determined (TBD) distance) using the calculated 10 mile TEDE, as necessary. STANDARD: DETERMINE from the cue sheet that the 2 mile TEDE dose is 950 mrem and ENTER the TEDE DOSE column at the TEDE 500-999 mrem dose level. Under the 2-5 Miles column, DETERMINE that SHELTER SECTORS AFFECTED is the correct classification DETERMINE 400 mrem at 5 miles is NONE for the 5-10 Miles PAR. COMMENTS: ST. LUCIE TRAINING DEPT. ADMINISTRATIVE JPM PAGE 8
STEP 12: 2. Determination of PARs (cant.) (cont.) CRITICAL B. Instructions for selecting PARs Based on Off-Site Dose (cant.) (cont.) STEP
- 2. See below:
- e. Enter the PAR information in the appropriate blocks of the TEDE DOSE table. - - SAT STANDARD: ENTER the following on bottom of page 6 of 7 of ATTACHMENT 2: - - UNSAT TEDE DOSE Miles EVACUATE SHELTER NO ACTION 0-2 ALL NONE NONE 2-5 NONE LMNP NONE 5-10 NONE NONE NONE
>10 COMMENTS:
( STEP 13: (cont.)
- 2. Determination of PARs (cant.)
B. Instructions for selecting PARs Based on Off-Site Dose (cant.) (cont.)
- 2. See below:
- f. Follow the same methodology for determining the
- - SAT PARs corresponding to the calculated CDE values beginning at 1 mile - - UNSAT STANDARD: DETERMINE CDE values will be later as per initiating cue.
COMMENTS: ST. LUCIE TRAINING DEPT. ADMINISTRATIVE JPM PAGE 9
STEP 14: 2. Determination of PARs (cont.) B. Instructions for selecting PARs Based on Off-Site Dose (cont.)
- 2. See below:
- g. Enter each of the determined PARs in the CDE - - SAT (Thyroid) DOSE table below the flowchart.
STANDARD: DETERMINE CDE values will be later as per initiating cue. -- UNSAT COMMENTS: STEP 15: 2. Determination of PARs (cont.) B. Instructions for selecting PARs Based on Off-Site Dose (cont.)
- 2. See below:
- h. Go to Section C, Instructions for the PAR - - SAT worksheet, below.
STANDARD: REFERS to Section C - - UNSAT COMMENTS: STEP 16: 2. Determination of PARs (cont.) C. Instructions for the PAR Worksheet
- 1. Fill in the time, date and emergency class:
- - SAT STANDARD: FILLS in the current time, date and General Emergency
- - UNSAT COMMENTS:
ST. LUCIE TRAINING DEPT. ADMINISTRATIVE JPM PAGE 10
STEP 17: 2. Determination of PARs (cont.) C. Instructions for the PAR Worksheet (cont.)
- 2. In Part a, PAR Comparison determine the basis for the most conservative PAR by comparing the PARs - - SAT based on plant conditions against those based on off-site dose. It is important to compare PARs at each (0-2, 2-5, 5-10) because the basis of the most distance (0-2,2-5,5-10) - - UNSAT conservative PAR could be different at different distances.
STANDARD: REFERS to PARs for General Emergency 'Severe Core Damage' page 5 of 7 on Attachment 2. Compares to PARs based on TEDE dose. COMMENTS: STEP 18: 2. Determination of PARs (cont.) CRITICAL C. Instructions for the PAR Worksheet (cont.) STEP ( 3. Enter the most conservative PARs into the table in Part B, Protective Actions to be Recommended by FPL. Use the word(s) NONE, ALL, ALL REMAINING or list - - SAT the individual affected sectors by letter.
- - UNSAT STANDARD: DETERMINES PARs for General Emergency (plant conditions) are more conservative than PARs based on TEDE Dose. Enter PAR's as per answer key.
COMMENTS: END OF TASK
*STOP TIME: _ _ _ __ *Critical Step: Must be S15 minutes from start time ST. LUCIE TRAINING DEPT.
( ADMINISTRATIVE JPM PAGE 11
A. PAR Comparison Compare protective actions based on plant conditions versus dose, determine the more conservative and check the applicable box in the table below. If no PAR is necessary, Then check 'none'
~
Downwind Distance Plant Conditions Dose None 0-2 miles *.J
*{ V
*.J 2-5 miles -J/
5-10 miles I/ I >10 TBD miles L-
- Both Plant and Dose are the same for 0-2 miles. Either I or are acceptable.
B. Protective Actions to be Recommended by FPL Complete the table below with the protective actions to be recommended. Use these items: NONE, ALL REMAINING or fill in the Letters of the sectors affected Downwind Distance EVACUATE SHELTER NO ACTION 1---- SECTORS SECTORS SECTORS ALL NONE NONE 0-2 Miles L,M,N,P L, M, N, P ALL REMAINING NONE 2-5 Miles NONE ALL NONE 5-10 Miles 10-TBD Miles*
- If necessary, add to State Notification Form
( ST. ST, LUCIE TRAINING DEPT. ADMINISTRATIVE JPM PAGE 12
CANDIDATE COPY (TO BE RETURNED TO THE EXAMINER UPON COMPLETION OF ANSWER) Initial Conditions Unit 1 is experiencing a total loss of feedwater and a loss of offsite power. The 1A HPSI 1A pump was out of service prior to the event. The crew just initiated once through cooling and the 1B HPSI pump tripped and is unable to be started. RCS pressure is rising and currently is 1620 psia. At time 0230, Containment High Range Monitor indicates 1.67 x 1055 R/hr. A containment vacuum breaker has failed resulting in a release path from the containment. At time 0445 Initial manual off-site dose reports have come in and are as follows (only TEDE values are available at this time; CDE values are to be determined later): 1 mile: 1800 mrem 2 miles: 950 mrem 5 miles: 400 mrem ( Wind direction is from 78° Initiating Cue The Shift Manager has directed you to determine Protective Action Recommendations for the next State of Florida Notification Form update. THIS IS A TIME CRITICAL JPM (
E: 20 OFF-SITE NOTIF!CATIONS AND PROTECTIVE 50 PROCEDURE NO.: ACTION RECOlv1rv1ENDA TIONS ST LUC!E PLANT ATTACHMENT 2 D:ETERMINATION OF PROTECTIVE ACTION RECOMMENDATIONS (PARs) DETERMINATION (Pnge (Page 5 of of7) 7) E'vACUA~E NO ;\CTJON Aji None NOilB None NJf1B
- t " . 2-5 None N)f1B C;Y11i:,UF 5-;(: rione A; N:mB N)f1B to i\ssess t.:; l;,r;S8SS Cont!ihons Comlrh:YIS
( EV,ACUAfE SHELTER SH::Ll1::K no NO ic,.CIION liON ()~2 NOt':tt ,451
?.!1 Nonl;!
;>Jon&
2-5 None ,A,II 5-10 Nor,e None None A;I
CANDIDATE COPY (TO BE RETURNED TO THE EXAMINER UPON COMPLETION OF ANSWER) Initial Conditions Unit 1 is experiencing a total loss of feedwater and a loss of offsite power. The 1A HPSI pump was out of service prior to the event. The crew just initiated once through cooling and the 1B HPSI pump tripped and is unable to be started. RCS pressure is rising and currently is 1620 psia. 5 At time 0230, Containment High Range Monitor indicates 1.67 x 10 5 R/hr. A containment vacuum breaker has failed resulting in a release path from the containment. Wind direction is from 78° At time 0445 Initial manual off-site dose reports have come in and are as follows (only TEDE values are available at this time; CDE values are to be determined later): 1 mile: 1800 mrem 2 miles: 950 mrem ( 5 miles: 400 mrem Wind direction is from 78° Initiating Cue The Shift Manager has directed you to determine Protective Action Recommendations for the next State of Florida Notification Form update. THIS IS A TIME CRITICAL JPM
FROCSD).;RE PRClCEJ)J.:RE 7]"1...E: TFLE: 20 OFF-SITE NOTIFICATiONS.AND PROTECTIVE 41 of 56 DLF-E NO.: ACTION RECO~",'MENDATIONS S1. LUCIE PLANT ATTACHMENT 1 1AA DIRECTIONS FOR COMPLETING THE FLORIDA NUCLEAR PLANT EMERGENCY NOTIFICATION FORM (Page 4 of 8) B. (continued) F. Entertlle v'Iind cHrecticm (vifind from) in degrees in item 'A': G. Enter tile downwind sectors in item '8. Sectors Sectors Affected Affected HJK PC1R 326-348 Tllere is no "1" sector A. If there are no indications
!I,til,.*~tjf'\r*,<:: of a
;) release of nldioactive rnaterl£ll, clled, box "A" and go to item 11 ,
itern11. I. 1I1~' Re!ease
- a. Evaluation guidance dwing the early stages of a declared ernergency, one of the following is true:
- 1. An Increase of (approxirnately) 10 times or one decade above pre-transIent values on of the followinG!
rnonitors: (a) AlB Stearniine (b) AlB ECCS (c) Plant Vent (d) Fuel Building OR
A. PAR Comparison Compare protective actions based on plant conditions versus dose, determine the more conservative and check the applicable box in the table below. If no PAR is necessary, Then check 'none' Downwind Distance Plant Conditions Dose None 0-2
-2 miles */"
*,[ */"
*,[
2-5 2 -5 miles ,[
/"
5 -10 miles 5-10 ,[
/"
.<,.;4 ,"'.
> 10 TSD
>10 TBD miles <~i:j.* ..
-----~-~-
- Both Plant and Dose are the thesarne same for 0-2 miles. Either lor are acceptable.
( B. Protective Actions to be Recommended Recoml'l1ended by FPL Complete the table below with theprQtective the protective actions to be recommended. Use REMAININ~;Q(Ji" these items: NONE, ALL REMAINING or fill in the Letters of the sectors affected tbeiLetters
------~---
,;'~::f;~~b~~': <"'; ';3 Downwind Distance .. EVACUATE EVACUATE ;;C!~\;:;:!~~~~L SHELTER TER NO ACTION
;> SECTORS .i**yi SECTORS SECTORS ALL NONE NONE 0-2 Miles L, M, N, 'f' P ALL REMAINING NONE 2-5 Miles NONE ALL NONE 5-10 Miles 10-TBD Miles*
---_.- ~
- If necessary, add to State Notification Form ST. LUCIE TRAINING DEPT.
ADMINISTRATIVE JPM PAGE 12}}