ML101650280
| ML101650280 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/07/2010 |
| From: | NRC/RGN-II |
| To: | Southern Nuclear Operating Co |
| References | |
| 50-348/10-301, 50-364/10-301 | |
| Download: ML101650280 (97) | |
Text
ES-301 Administrative Topics Outline Form ES-SO1-1 OPERATING OUTLINE SUBMITTAL Facility: Farley Nuclear Plant Date of Examination: March 8, 2010 Examination Level: SRO + RO Operating Test Number:
FA2010301 Administrative Topic Type Describe activity to be performed Code (see Note)
Verification of Initial Conditions Prior to Core Alterations.
Given a set of plant conditions with fuel movement in A1.1.A Conduct of Operations N
R progress, determine if all Core Alterations initial conditions are satisfied using UOP-4.1.
RO pOrtlOfl G2.l.40 (2.8/3.9)
G2.1.36 (3.0/4.1)
G2.1.32 (3.8/4.0)
Verification of Initial Conditions Prior to Core Alterations.
Given a set of plant conditions with fuel movement in Al. I.A progress, determine if all Core Alterations initial conditions are satisfied using UOP-4.1, and then list all Conduct of Operations N / R SRO portion Tech Spec conditions, REQUIRED ACTIONS and COMPLETION TIMES for LCOs not met.
G2.l.40 (2.8/3.9)
G2.1.36 (3.0/4.1)
G2.2.1.35 (2.2/3.9)
G2.1.32 (3.8/4.0)
Al.2.S D / R Conduct A Safety Shutdown Assessment and Determine Conduct of Operations Time to Saturation.
SRO ONLY G2.1.25 (3.9/4.2)
A2.1.A Perform a Shutdown Margin Calculation in modes 1 & 2 Equipment Control D I R for a stuck rod (STP-29.5)
SRO + RO 001A4.11 (3.5/4.1) APEOO5 AKI.05 (3.3/4.1)
A3. 1.A M / R Calculate the Maximum Permissible Stay Time within Radiation Control Emergency Dose Limits.
A4. I.A M I S Monitor the Critical Safety Function Status Trees Emergency Plan
manually (CSF-0.0)
SRO + RO G2.4.14 (3.8/4.5) G2.4.21 (SRO 4.0/4.6)
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (. 3 for ROs; 4 for SROs & RO retakes) [112]
(N)ew or (M)odified from bank (> 1) [313]
(P)revious 2 exams ( I; randomly selected) [010]
Farley OCT 2010 NRC ADMIN exam outline Page 1 of 1 ES-301 Administrative Topics Outline Form ES-301-1 OPERATING OUTLINE SUBMITTAL Facility: Farle~ Nuclear Plant Date of Examination: March 8,2010 Examination Level: SRO + RO Operating Test Number: FA2010301 Administrative Topic TYJ!e Describe activity to be performed (see Note)
Co e*
Verification of Initial Conditions Prior to Core Alterations.
A1.1.A Given a set of plant conditions with fuel movement in Conduct of Operations N/R progress, determine if all Core Alterations initial conditions are satisfied using UOP-4.1.
RO portion G2.1.40 (2.8/3.9)
G2.1.36 (3.0/4.1)
G2.1.32 (3.8/4.0)
Verification of Initial Conditions Prior to Core Alterations.
Given a set of plant conditions with fuel movement in A1.1.A progress, determine if all Core Alterations initial Conduct of Operations N/R conditions are satisfied using UOP-4.1, and then list all Tech Spec conditions, REQUIRED ACTIONS and COMPLETION TIMES for LCOs not met.
G2.1.40 (2.8/3.9)
G2.1.36 (3.0/4.1)
G2.2.1.3S{2.2/3.9)
G2.1.32 (3.8/4.0)
A1.2.S D/R Conduct A Safety Shutdown Assessment and Determine Conduct of Operations Time to Saturation.
SRO ONLY G2.1.2S (3.9/4.2)
A2.1.A Perform a Shutdown Margin Calculation in modes 1 & 2 Equipment Control D/R for a stuck rod (STP-29.S)
+RO 001A4.11 (3.S/4.1) APEOOS AK1.0S (3.3/4.1)
A3.1.A M/R Calculate the Maximum Permissible Stay Time within Radiation Control Emergency Dose Limits.
SRO+RO G2.3.4 (3.2/3.7)
A4.1.A M/S Monitor the Critical Safety Function Status Trees Emergency Plan -
manually (CSF-O.O)
SRO+RO G2.4.14 (3.8/4.S) G2.4.21 (SRO 4.0/4.6)
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank <:5. 3 for ROs; ~ 4 for SROs & RO retakes) [1/2]
(N)ew or (M)odified from bank e 1) [3/3]
(P)revious 2 exams (~ 1; randomly selected) [0/0]
Farley OCT 2010 NRC ADMIN exam outline Page 1 of 1
ES-301 Administrative Topics Outline Form ES-301-1 OPERATING OUTLINE SUBMITTAL Facility: Farley Nuclear Plant Date of Examination:
Examination Level: SRO + RO Operating Test Number:
FA2010301 Administrative Topic JY$e Describe activity to be performed (see Note) 0 e Verification of Initial Conditions Prior to Core Alterations.
Given a set of plant conditions with fuel movement in progress, determine if refueling operations can continue, and if not, the reason(s) that prohibit refueling from continuing. UOP-4.1 P&Ls and TS and TRM requirements.
A1.1.A G2.140 (2.8/3.9)
G2.1.36 (3.0/4.1)
Conduct of Operations N / R G2.1.32 (3.8/4.0)
RO POI Perform UOP-4.1 appendix 6 and determine if all core alteration initial conditions are satisfied. This will be done in the classroom and we will use pictures of the Nis and radiation monitors that will show that one NI is not working and the audio count rate selector switch is selected to that NI. Also with fuel movement in the SEP there will be one radiation monitor that is inoperable due to it being in test.
Verification of Initial Conditions Prior to Core Alterations.
Given a set of plant conditions with fuel movement in progress, determine if all Core Alterations initial conditions are satisfied using UOP-4. 1, and then list all Al.1.A Tech Spec conditions, REQUIRED ACTIONS and Conduct of Operations N I R COMPLETION TIMES for LCOs not met.
SRO portion G2.1.40 (2.8/3.9)
G2.1.36 (3.014.1)
G2.2.l.35 (2.2/3.9)
G2.l.32 (3.8/4.0)
This is the same JPM as above with the requirement to evaluate Tech Specs and TRM requirements that cover the conditions given and the REQUIRED ACTIONS and COMPLETION_TIMES_of TS_3.9.2_and_TRM_13.3.4.
Al.2.S D / R Conduct A Safety Shutdown Assessment and Determine Conduct of Operations Time to Saturation.
SRO ONLY G2.l.25 (3.9/4.2)
This JPM will have the candidate evaluate plant conditions, use a table to determine time to boiling and then fill out UOP 4.0, figure la Core Cooling section only. This is only performed by the SRO job function at Farley Nuclear Plant.
A2.l.A Perform a Shutdown Margin Calculation in modes 1 & 2 Equipment Control for a stuck rod (STP-29.5)
SRO + RO D / R 001A4.l 1 (3.5/4.1) APEOO5 AKi.05 (3.3/4.1)
One Bank D rod is 30 steps below the other seven Bank D rods.
Determine the SDM and that an emergency boration is_required.
Farley OCT 2010 NRC ADMIN exam outline Page 1 of2 ES-301 Administrative Topics Outline Form ES-301-1 OPERATING OUTLINE SUBMITTAL Facility: Farley Nuclear Plant Examination Level: SRO + RO Administrative Topic (see Note)
A1.1.A Conduct of Operations RO portion A1.1.A Conduct of Operations A1.2.S Conduct of Operations A2.1.A Equipment Control
+RO Type Code
- N/R N/R D/R D/R Date of Examination:
Operating Test Number: FA2010301 Describe activity to be performed Verification of Initial Conditions Prior to Core Alterations.
Given a set of plant conditions with fuel movement in progress, determine if refueling operations can continue, and if not, the reason(s) that prohibit refueling from continuing. UOP-4.1 P&Ls and TS and TRM requirements.
G2.1.40 (2.8/3.9)
G2.1.36 (3.0/4.1)
G2.1.32 (3.8/4.0)
Perform UOP-4.1 appendix 6 and determine if all core alteration initial conditions are satisfied. This will be done in the classroom and we will use pictures of the Nt's and radiation monitors that will show that one NI is not working and the audio count rate selector switch is selected to that NI. Also with fuel movement in the SFP there will be one radiation monitor that is inoperable due to it being in test.
Verification of Initial Conditions Prior to Core Alterations.
Given a set of plant conditions with fuel movement in progress, determine if all Core Alterations initial conditions are satisfied using UOP-4.1, and then list all Tech Spec conditions, REQUIRED ACTIONS and COMPLETION TIMES for LCOs not met.
G2.1.40 (2.8/3.9)
G2.1.36 (3.0/4.1)
G2.2.1.35 (2.2/3.9)
G2.1.32 (3.8/4.0)
This is the same JPM as above with the requirement to evaluate Tech Specs and TRM requirements that cover the conditions given and the REQUIRED ACTIONS and COMPLETION TIMES of TS 3.9.2 and TRM 13.3.4.
Conduct A Safety Shutdown Assessment and Determine Time to Saturation.
G2.1.25 (3.9/4.2)
This JPM will have the candidate evaluate plant conditions, use a table to determine time to boiling and then fill out UOP-4.0, figure 1a Core Cooling section only. This is only performed by the SRO job function at Farley Nuclear Plant.
Perform a Shutdown Margin Calculation in modes 1 & 2 for a stuck rod (STP-29.5) 001A4.11 (3.5/4.1) APE005 AK1.05 (3.3/4.1)
One Bank D rod is 30 steps below the other seven Bank D rods.
Determine the SDM and that an emergency boration is required.
Farley OCT 2010 NRC ADMIN exam outline Page 1 of2
ES-301 Administrative Topics Outline Form ES-301 -1 OPERATING OUTLINE SUBMITTAL A3.1.A M I R Calculate the Maximum Permissible Stay Time within Radiation Control Emergency Dose Limits.
SRO + RO This JPM has the candidate assess a job where two workers will be assigned a task during an emergency event to save a valuable piece of equipment. There will be three stages to the task in which the dose rate is given and time required to complete the task is given. The year to date dose rates will be given and the task will be to determine if either of the workers will exceed their dose limits of EIP-14. Information required to be known is that EDLs do not take into account current dose, admin limits and NRC limits do not apply and the EIP-14 limits must be applied properly.
G2.3.4 (3.2/3.7)
A4.1.A M I S Monitor the Critical Safety Function Status Trees Emergency Plan manually (CSF-0.0)
SRO + RO G2.4.14 (3.8/4.5) G2.4.21 (SRO 4.0/4.6)
The simulator will be used requiring the candidate to determine the appropriate CSF that applies. We have a snap setup that has N-42 failed high which will cause evaluation of FRP-S, to show a yellow path on FRP-C and H, an Orange path on FRP P and Z. A manual determination of CSF-O will be required as to which FRP is applicable based on the setup.
The RO will identify all the CSFs that are challenged and the SRO will have to use the procedures to determine which procedure is required due to the plant conditions.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; < 4 for SROs & RO retakes) [112]
(N)ew or (M)odifled from bank ( 1) [3I3]
(P)revious 2 exams ( 1; randomly selected) [OlO]
Changes since the draft submittal:
A1.1.A We could not find procedural guidance for the RHR flow requirement so a radiation monitor for the procedure being used was substituted for the RHR flow. Changed the appropriate Tech Spec to TRM 13.3.4 to adjust for the radiation monitor Inoperable.
A3.1.A added the KA value to ES-301-1 G2.3.4 (3.2/3.7)
A4.1.A We could not get the submitted simulator results fo having FRP-P.1 a red path, but we developed one with an Orange path. We also put in a failed NI into the scenario that will have FRP-S evaluated since one NI is >5% power, but failed, so it is not relevant.
Farley OCT 2010 NRC ADMIN exam outline Page 2 of 2 ES-301 A3.1.A Radiation Control
+RO A4.1.A Emergency Plan -
+RO Administrative Topics Outline Form ES-301-1 OPERATING OUTLINE SUBMITTAL MIR MIS Calculate the Maximum Permissible Stay Time within Emergency Dose limits.
This JPM has the candidate assess a job where two workers will be assigned a task during an emergency event to save a valuable piece of equipment. There will be three stages to the task in which the dose rate is given and time required to complete the task is given. The year to date dose rates will be given and the task will be to determine if either of the workers will exceed their dose limits of EIP-14. Information required to be known is that EDLs do not take into account current dose, admin limits and NRC limits do not apply and the EIP-14 limits must be applied properly.
G2.3.4 (3.2/3.7)
Monitor the Critical Safety Function Status Trees manually (CSF-O.O)
G2.4.14 (3.8/4.5) G2.4.21 (SRO 4.0/4.6)
The simulator will be used requiring the candidate to determine the appropriate CSF that applies. We have a snap setup that has N-42 failed high which will cause evaluation of FRP-S, to show a yellow path on FRP-C and H, an Orange path on FRP-P and Z. A manual determination of CSF-O will be required as to which FRP is applicable based on the setup.
The RO will identify all the CSFs that are challenged and the SRO will have to use the procedures to determine which procedure is required due to the plant conditions.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ~ 3 for ROs; ~ 4 for SROs & RO retakes) [1/2]
(N)ew or (M)odified from bank e 1) [313]
(P)revious 2 exams << 1; randomly selected) [010]
Changes since the draft submittal:
A1.1.A We could not find procedural guidance for the RHR flow requirement so a radiation monitor for the procedure being used was substituted for the RHR flow. Changed the appropriate Tech Spec to TRM 13.3.4 to adjust for the radiation monitor Inoperable.
A3.1.A added the KA value to ES-301-1 G2.3.4 (3.2/3.7)
A4.1.A We could not get the submitted simulator results fo having FRP-P.1 a red path, but we developed one with an Orange path. We also put in a failed NI into the scenario that will have FRP-S evaluated since one NI is >5% power, but failed, so it is not relevant.
Farley OCT 2010 NRC ADMIN exam outline Page 2 of2
FNP HLT-33 ADMIN Page 1 of 10 A.1.1.A Conduct of Operations ADMIN G2.1.40 RO & SRO JPM DIRECTIONS:
1.
This task is to be administered in a group, classroom setting with access to a variety of procedures (UOP5, STPs, SOPs, et al) available via the Exam Reference disk.
2.
This task is designed to be performed with evaluation of pictures of the equipment from which operational information will be evaluated.
3.
Due to the utilization of pictures, upon request, the status of the lights and switches may be clarified (See standards for the applicable light/switch).
4.
Upon completion of elements 1-10, all candidates will be evaluated at the RO level.
5.
FOR SRO candidates ONLY: after evaluation of elements 1-10, they will be provided a second Cue sheet (handout #2) and returned to the classroom to evaluate TS.
TASK STANDARD: Required for successful completion of this JPM:
Evaluates all steps of UOP-4. 1, Appendix 6, Verification Of Initial Conditions Prior To Core Alterations.
Identifies all inoperable equipment or unsatisfied conditions that prevent completing the attachment, if any.
(For SRO ONLY):
o Based on the conditions identified List all TS CONDITIONS, REQUIRED ACTIONS and COMPLETION TIMES for LCOs not met.
EXAMINER:
A.1.1.A TITLE: Verification of Initial Conditions Prior to Core Alterations.
PROGRAM APPLICABLE: SOT SOCT OLT X
LOCT____
ACCEPTABLE EVALUATION METHOD:
X PERFORM SIMULATE DISCUSS EVALUATION LOCATION:
SIMULATOR CONTROL ROOM X
CLASSROOM PROJECTED TIME:
20 MIN SIMULATOR IC NUMBER:
N/A ALTERNATE PATH TIME CRITICAL PRA____
Examinee:
Overall JPM Performance:
Satisfactory Unsatisfactory D
Evaluator Comments (attach additional sheets if necessary)
Developer H. Fitzwater 10/26/09 NRC Approval SEE NUREG 1021 FORM ES-301-3 FNP HL T-33 ADMIN Page 1 of 10 A.1.1.A Conduct of Operations ADMIN G2.1.40 - RO & SRO A.1.1.A TITLE: Verification of Initial Conditions Prior to Core Alterations.
PROGRAM APPLICABLE: SOT SOCT OL T -.-lL LOCT __
ACCEPTABLE EVALUATION METHOD: ~
PERFORM SIMULATE DISCUSS EVALUATION LOCATION:
SIMULATOR CONTROL ROOM ~
CLASSROOM PROJECTED TIME:
20 MIN SIMULATOR IC NUMBER:
N/A ALTERNATE PATH TIME CRITI CAL PRA __
JPM DIRECTIONS:
- 1. This task is to be administered in a group, classroom setting with access to a variety of procedures (UOPs, STPs, SOPs, et al) available via the Exam Reference disk.
- 2. This task is designed to be performed with evaluation of pictures of the equipment from which operational information will be evaluated.
- 3. Due to the utilization of pictures, upon request, the status of the lights and switches may be clarified (See standards for the applicable light/switch).
- 4. Upon completion of elements 1-10, all candidates will be evaluated at the RO level.
- 5. FOR SRO candidates ONLY: after evaluation of elements 1-10, they will be provided a second Cue sheet (handout #2) and returned to the classroom to evaluate TS.
TASK STANDARD: Required for successful completion of this JPM:
Evaluates all steps of UOP-4.1, Appendix 6, Verification OfInitial Conditions Prior To Core Alterations.
Identifies all inoperable equipment or unsatisfied conditions that prevent completing the attachment, if any.
(For SRO ONLY):
o Based on the conditions identified List all TS CONDITIONS, REQUIRED ACTIONS and COMPLETION TIMES for LCOs not met.
Examinee:
Overall JPM Performance:
Satisfactory 0
U nsatisfactorv 0 Evaluator Comments (attach additional sheets if necessary)
EXAMINER: ________________ _
H. Fitzwater 10/26/09 SEE NUREG 1021 FORM ES-301-3
FNPHLT-33ADMIN A.1.1.A Page2oflO CONDITIONS When I tell you to begin, you are to perform or evaluate all steps FNP-I UOP-4. 1, Appendix 6, Verification of Initial Conditions Prior to Core Alterations. The conditions under which this task is to be performed are:
a.
Core Offload was suspended 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> ago.
b.
RCS temp is 100°F.
c.
Time to boil is >2 hours.
d.
A Fuel shuffle in the Spent Fuel Pool is ongoing.
e.
Communications between the MCR and a System Operator acting as the Cavity Watch have been verified available 10 minutes ago.
f.
RCS boron concentration is 2450 ppm following an over boration performed last shift.
g.
Rx Makeup water is being used to dilute the Reactor Cavity by chemistrys request.
h.
FNP-1-STP-18.4, Containment Midloop and/or Refueling Integrity Verification and Containment Closure, was completed yesterday at 1400.
i.
CTMT Main Purge system is in operation.
j.
The equipment hatch and all containment air locks are currently closed.
k.
Applicable equipment can be assessed using the provided photographs.
I.
You have been directed to perform or evaluate all steps of FNP-I -UOP-4. 1, Appendix 6.
- m. Determine whether or not all Core Alterations initial conditions are satisfied.
INITIATING CUE: You may begin.
EVALUATION CHECKLIST RESULTS:
ELEMENTS:
STANDARDS:
(CIRCLE)
START TIME
- 1. (STEP 2.1 & 2.2) Checks correct version and
- 1) Determines step 2.1 & 2.2 has S / U conducted on appropriate unit.
been completed. Utilizes WebTop and verifies unit from initial conditions
- 2. (STEP 3.1) Communications between MCR and
- 2) Determines step 3.1 has been S / U the Reactor Cavity verified within one hour.
completed. Utilizes initial conditions to verify task complete.
NOTE:
Elements 3 through 5, 7 and 8 are performed by use of pictures. The pictures are sufficient to demonstrate light status and switch position, however IF light status or position is questioned, then the cues may be provided (see the standards column for the applicable light or switch).
FNP HLT-33 ADMIN A.1.1.A Page 2 of 10 CONDITIONS When I tell you to begin, you are to perform or evaluate all steps FNP-I-UOP-4.1, Appendix 6, Verification of Initial Conditions Prior to Core Alterations. The conditions under which this task is to be performed are:
- a. Core Offload was suspended 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> ago.
- b. RCS temp is 100°F.
- c. Time to boil is >2 hours.
- d. A Fuel shuffle in the Spent Fuel Pool is ongoing.
- e. Communications between the MCR and a System Operator acting as the Cavity Watch have been verified available 10 minutes ago.
- f.
RCS boron concentration is 2450 ppm following an over boration performed last shift.
- g. Rx Makeup water is being used to dilute the Reactor Cavity by chemistry's request.
- h. FNP-I-STP-18.4, Containment Midloop and/or Refueling Integrity Verification and Containment Closure, was completed yesterday at 1400.
I.
CTMT Main Purge system is in operation.
J.
The equipment hatch and all containment air locks are currently closed.
- k. Applicable equipment can be assessed using the provided photographs.
- 1.
You have been directed to perform or evaluate all steps of FNP-I-UOP-4.1, Appendix 6.
- m. Determine whether or not all Core Alterations initial conditions are satisfied.
INITIATING CUE: "You may begin."
EVALUATION CHECKLIST ELEMENTS:
START TIME
- 1. (STEP 2.1 & 2.2) Checks correct version and conducted on appropriate unit.
- 2. (STEP 3.1) Communications between MCR and the Reactor Cavity verified within one hour.
STANDARDS:
RESULTS:
(CIRCLE)
- 1) Determines step 2.1 & 2.2 has S / U been completed. Utilizes WebTop and verifies unit from initial conditions
- 2) Determines step 3.1 has been S / U completed. Utilizes initial conditions to verify task complete.
NOTE:*
Elements 3 through 5, 7 and 8 are performed by use of pictures. The pictures are sufficient to demonstrate light status and switch position, however IF light status or position is questioned, then the cues may be provided (see the standards column for the applicable light or switch).
FNP HLT-33 ADMIN A.1.1.A Page 3 of 10 EVALUATION CHECKLIST ELEMENTS:
- 3. (STEP 3.2.1) Verify N-3 1, Source Range NI, detector operable.
(reference SOP-39.O section 4.1)
- a. Check N-3 I Instrument power on AND control power on lights illuminated.
- b. Check switches aligned:
- i. Level Trip ii.Operation Selector iii.High Flux at shutdown STANDARDS:
- 3) Determines step 3.2.1 has been satisfied; Utilizes picture ofN-31 and determines that the instrument is operable.
a.
lights are illuminated b.
Switches are in position:
- i. Normal ii.
Normal iii.
Normal RESULTS:
(CIRCLE)
S/U c.
Perform a channel check per STP-l.0 (SR 3.9.2.1).
- 4. (STEP 3.2.2)Verify N-32, Source Range NI, detector operable.
(reference SOP-39.0 section 4.1)
- a. Check N-32 Instrument power on AND control power on lights illuminated.
- b. Check switches aligned:
- i. Level Trip ii.Operation Selector iii.High Flux at shutdown c.
Performs a channel check per STP-l.0 (SR 3.9.2.1).
- c. Acceptable per STP-I.0 allowances
- 4) Determines step 3.2.2 can not be satisfied; Utilizes picture of N-32 and determines that the instrument is NOT operable.
- a. Lights are checked Control power light is NOT illuminated.
Instrument power is illuminated.
b.
Switches are determined;
- i. Normal ii. Normal iii.
Normal
- c. Acceptable per STP-1.0 allowances.
S/U FNP HL T-33 ADMIN EVALUATION CHECKLIST ELEMENTS:
A.1.1.A
- 3. (STEP 3.2.1) VerifY N-31, Source Range NI, detector operable.
(reference SOP-39.0 section 4.1)
- a. Check N-31 Instrument power on AND control power on lights illuminated.
- b. Check switches aligned:
- i. Level Trip ii.Operation Selector iii.High Flux at shutdown
- c. Perform a channel check per STP-1.0 (SR 3.9.2.1).
- 4. (STEP 3.2.2)VerifY N-32, Source Range NI, detector operable.
(reference SOP-39.0 section 4.1)
- a. Check N-32 Instrument power on AND control power on lights illuminated.
- b. Check switches aligned:
- i. Level Trip ii.Operation Selector iii.High Flux at shutdown
- c. Performs a channel check per STP-1.0 (SR 3.9.2.1).
Page 3 of 10 RESULTS:
STANDARDS:
(CIRCLE)
- 3) Determines step 3.2.1 has been S / U satisfied; Utilizes picture ofN-31 and determines that the instrument is operable.
- a. lights are illuminated
- b. Switches are in position:
I. Normal ii. Normal iii. Normal
- c. Acceptable per STP-1.0 allowances
- 4) Determines step 3.2.2 can not S / U be satisfied; Utilizes picture of N-32 and determines that the instrument is NOT operable.
- a. Lights are checked Control power light is NOT illuminated.
Instrument power is illuminated.
- b. Switches are determined; I. Normal II. Normal Ill. Normal
- c. Acceptable per STP-1.0 allowances.
FNP HLT-33 ADMIN A.1.1.A Page 4 of 10 EVALUATION CHECKLIST RESULTS:
ELEMENTS:
STANDARDS:
(CIRCLE)
- 5. (STEP 3.2.3)Verify N-48, Gamma.metrics
- 5) Determines step 3.2.3 has been S / U Source Range NI, detector operable.
satisfied. (May identify that the (reference SOP-39.O section 4.3)
Refueling Coordinator must first be consulted.)
a.
Checks N-48 in service, a.
Aligned and in service as indicated by picture and bulleted items listed on picture.
b.
Performs a channel check per STP-1.0 (SR b.
Acceptable per STP-1.0 3.9.2.1).
Table I allowances.
- 6. (STEP 3.3) Reactor Coolant system 2000 ppm.
- 6) Determines step 3.3 has been S / U satisfied; Utilizes initial conditions to verify boron concentration.
NOTE:
Element 7 would require either R-5 being in service OR required action being taken. The candidate might question whether or not surveys are being performed. Additionally, due to rules of usage may stop the task and attempt to notify the SS.
IF required CUE: The SS acknowledges the issue you have identified and directs you to continue. No compensatory actions have yet been completed.
7.
(STEP 3.4.1) Verifies R-5 operable/operating.
- 7) Identifies R-5 is NOT S / U operable/operating; Determines step 3.4.1 has NOT been satisfied. (May annotate failure
(
Reference:
SOP-45.0 section 4.1.3; TRM 13.3.4) or need to initiate survey).
a.
Verify Operation Selector switch position.
- a. Level Cal.
b.
Verify Range Selector switch position.
b.
WIDE c.
Verify Drawer indication lights:
c.
Lights are:
i.
Power i.
ON ii.
Channel Test light ii.
ON iii.
High Alarm iii.
OFF iv.
Low Alarm iv.
OFF FNP HLT-33 ADMIN EVALUATION CHECKLIST ELEMENTS:
- 5. (STEP 3.2.3)Verify N-48, Gamma-metrics Source Range NI, detector operable.
(reference SOP-39.0 section 4.3)
- a. Checks N-48 in service.
A.1.1.A
- b. Performs a channel check per STP-I.O (SR 3.9.2.1).
- 6. (STEP 3.3) Reactor Coolant system 2000 ppm.
Page 4 of 10 RESULTS:
STANDARDS:
(CIRCLE)
- 5) Determines step 3.2.3 has been S / U satisfied. (May identify that the Refueling Coordinator must first be consulted.)
- a. Aligned and in service as indicated by picture and bulleted items listed on picture.
- b. Acceptable per STP-I.O Table 1 allowances.
- 6) Determines step 3.3 has been satisfied; Utilizes initial conditions to verify boron concentration.
S / U NOTE:
Element 7 would require either R-5 being in service OR required action being taken. The candidate might question whether or not surveys are being performed. Additionally, due to rules of usage may stop the task and attempt to notify the SS.
IF required CUE: "The SS acknowledges the issue you have identified and directs you to continue. No compensatory actions have yet been completed."
- 7. (STEP 3.4.1) Verifies R-5 operable/operating.
(
Reference:
SOP-45.0 section 4.1.3; TRM 13.3.4)
- a. Verify Operation Selector switch position.
- b. Verify Range Selector switch position.
- c. Verify Drawer indication lights:
I.
Power
- 11.
Channel Test light iii.
High Alarm IV.
Low Alarm
- 7) Identifies R-5 is NOT S / U operable/operating; Determines step 3.4.1 has NOT been satisfied. (May annotate failure or need to initiate survey).
- a. Level Cal.
- b. WIDE
- c. Lights are:
- l. ON ii. ON Ill. OFF IV. OFF
FNP HLT-33 ADMIN A.1.1.A Page 5 of 10 EVALUATION CHECKLIST RESULTS:
ELEMENTS:
STANDARDS:
(CIRCLE)
- 8. (STEP 3.4.2 through 3.4.4) Verifies R-24AorB,
- 8) Identifies radiation monitors are S / U R-25AorB, and R-35AorB operable using SOP-operable; Initials step 3.4.2 45.0 section 4.1.5.
through 3.4.4.
(uses pictures of LOCAL radiation monitor The following indications are for indications to determine the following conditions) all steps 3.4.2 through 3.4.4:
a.
Verify local switch for vacuum pump.
- a. AUTO b.
Verify LOCAL switch position.
- b. ON (and Fans are running)
(uses pictures of radiation monitor drawers to determine the following conditions) c.
Verify MCR switch:
- c. MCR switches are:
i.
Operation selector i.
Oper ii.
Pump Power ii.
ON iii.
Pump Start iii.
Start/Neutral d.
Verify MCR lights:
- d. MCR lights are:
i.
Power light i.
ON ii.
Pump ON light ii.
ON iii.
Flow fault light iii.
OFF iv.
Alert light iv.
OFF v.
High light v.
OFF vi.
Fail/Reset light vi.
ON vii.
Power on light vii.
ON
- 9. (STEP 3.5 through 3.6) Evaluates Refueling
- 9) Utilizes initial conditions and S / U integrity has been verified within 100 hrs, and evaluates that step 3.5 is that step 3.6 is not applicable, satisfied and 3.6 in N/A.
NOTE:
REVIEW of completed Appendix 6 to evaluate the above elements.
CUE:
Are core alterations initial conditions satisfied?
10.Reports/annotates that step 3.2.2 and 3.4.1 are not
- 10) Identifies N32 inoperable, R-S / U completed or satisfied and that core Alts should 5 inoperable and that Core not be started.
alterations should not be permitted since the initial conditions have not been satisfied.
FNP HL T-33 ADMIN A.1.1.A EVALUATION CHECKLIST ELEMENTS:
- 8. (STEP 3.4.2 through 3.4.4) Verifies R-24AorB, R-25AorB, and R-35AorB operable using SOP-45.0 section 4.1.5.
(uses pictures of LOCAL radiation monitor indications to determine the following conditions)
- a. Verify local switch for vacuum pump.
- b. Verify LOCAL switch position.
(uses pictures of radiation monitor drawers to determine the following conditions)
- c. Verify MCR switch:
I.
Operation selector Il.
Pump Power iii.
Pump Start
- d. Verify MCR lights:
I.
Power light Il.
Pump ON light iii.
Flow fault light IV.
Alert light
- v.
High light VI.
Fail/Reset light vii.
Power on light
- 9. (STEP 3.5 through 3.6) Evaluates Refueling integrity has been verified within 100 hrs, and that step 3.6 is not applicable.
Page 5 of 10 RESULTS:
STANDARDS:
(CIRCLE)
- 8) Identifies radiation monitors are S / U operable; Initials step 3.4.2 through 3.4.4.
The following indications are for all steps 3.4.2 through 3.4.4:
- a. AUTO
- b. ON (and Fans are running)
- c. MCR switches are:
I.
Oper
- 11.
ON Ill.
StartlN eutral
- d. MCR lights are:
I.
ON Il. ON Ill. OFF IV. OFF
- v. OFF VI. ON vii. ON
- 9) Utilizes initial conditions and evaluates that step 3.5 is satisfied and 3.6 in N/A.
S / U NOTE:
REVIEW of completed Appendix 6 to evaluate the above elements.
CUE:
Are core alterations initial conditions satisfied?
- 10. Reports/annotates that step 3.2.2 and 3.4.1 are not completed or satisfied and that core Alts should not be started.
- 10) Identifies N32 inoperable, R-5 inoperable and that Core alterations should not be permitted since the initial conditions have not been satisfied.
S / U
FNPHLT-33ADMIN A.1.1.A Page6oflo EVALUATION CHECKLIST RESULTS:
ELEMENTS:
STANDARDS:
(CIRCLE)
RO ONLY:
Terminate when elements 1-10 of the task have been completed.
SRO ONLY Cue: PROVIDE Handout #2, read and escort back to classroom to allow TS review.
Based on the conditions you identified perform the following:
- 1) List all TS CONDITIONS, REQUIRED ACTIONS and COMPLETION TIMES for LCOs not met, if any.
- 11. Evaluates TS 3.9.2 Condition C: Inoperable II) Identifies the need to S / U Audio Counts.
IMMEDIATELY verify un borated water sources are isolated.
- 12. Evaluates TRM 13.3.4 condition B.
- 12) Identifies the need to S I U IMMEDIATELY notify HP to perform survey the SFP area within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
NOTE:
TS 3.9.2 Condition A is not required to be entered due to Gamma Metrics (as allowed by BASIS) and N-31 are OPERABLE. CONDITION A may be implemented and does not challenge personnel safety, equipment or license requirements.
However, IF CONDITION A is implemented, THEN element 13 is critical.
SEE
- 13. IF TS 3.9.2 implemented, THEN evaluates
- 13) Identifies the need to S / U NOTE TS 3.9.2 CONDITION A: One source range IMMEDIATELY Suspend neutron flux monitor inoperable.
CORE ALTERATIONS AND positive reactivity additions (secure Reactor Cavity Makeup).
SRO ONLY:
Terminate when elements 1-13 of the task are complete.
STOP TIME CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.
GENERAL
REFERENCES:
- 1. UOP-4.1, VER 51 2.KA:#G2.1.40 2.8 3.9 G2.1.36 3.0 4.1 G2.1.32 3.8 4.0 FNP HL T-33 ADMIN A.1.1.A EVALUATION CHECKLIST ELEMENTS:
STANDARDS:
RO ONLY:
Terminate when elements 1-10 of the task have been completed.
Page 6 of 10 RESULTS:
(CIRCLE)
SRO ONLY Cue: PROVIDE Handout #2, read and escort back to classroom to allow TS review.
Based on the conditions you identified perform the following:
- 1) List all TS CONDITIONS, REQUIRED ACTIONS and COMPLETION TIMES for LCOs not met, if any.
- 11. Evaluates TS 3.9.2 Condition C: Inoperable Audio Counts.
- 12. Evaluates TRM 13.3.4 condition B.
- 11) Identifies the need to S I U IMMEDIATELY verify un-borated water sources are isolated.
- 12) Identifies the need to S I U IMMEDIATELY notify HP to perform survey the SFP area within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
NOTE:
TS 3.9.2 Condition A is not required to be entered due to Gamma Metrics (as allowed by BASIS) and N-31 are OPERABLE. CONDITION A may be implemented and does not challenge personnel safety, equipment or license requirements.
However, IF CONDITION A is implemented, THEN element 13 is critical.
SEE NOTE
- 13. IF TS 3.9.2 implemented, THEN evaluates TS 3.9.2 CONDITION A: One source range neutron flux monitor inoperable.
- 13) Identifies the need to IMMEDIATELY Suspend CORE ALTERATIONS AND positive reactivity additions (secure Reactor Cavity Makeup).
I SRO ONLY:
Terminate when elements 1-13 of the task are complete.
STOP TIME S I U CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.
GENERAL
REFERENCES:
- 1. UOP-4.I, VER 51
- 2. KA: # G2.1.40 G2.1.36 G2.I.32 2.8 3.9 3.0 4.1 3.8 4.0
FNP HLT-33 ADMIN A.1.1.A Page 7 of 10 disk/files only.
Provide UOP-4. 1, Appendix 6 Pen/pencil Pictures of the following components: (II pages total)
- a. R-24A Local indications g.
- b. R-24B Local indications h.
- c. R-25A Local indications i.
- d. R-25B Local indications j.
- e. R-35A Local indications k.
- f. R-35B Local indications I.
R24A & R25A MCR indications R-24B & R25B MCR indications R-35A & R35B MCR indications N3 drawer indications N32 drawer indications N3 1, N32 & N48 Board indications
- m. Audio Count Rate Channel Drawer Critical ELEMENT justification:
STEP Evaluation NOTE:
The listed sub-steps to each element (a.b.c...) describe the anticipated actions taken by the candidate. Completion of all substeps is not required to make the determination of operability of the components, nor is it intended that these actions be observed. These sub-steps are provided for the examiner as an evaluation tool.
1.
NOT critical--this step is for satisfying site procedural protocol and expectations.
2.
NOT critical this step is for satisfying site level expectations or requirements.
3.
NOT criticalAlthough Task objective-component is operable.
4.
Critical Element Task objective-evaluate operability of required equipment.
5.
NOT criticalAlthough Task objective-component is operable.
6.
NOT critical this step does not require any evaluation from the candidate.
7.
Critical Element Task objective-evaluate operability of required equipment.
8.
NOT criticalAlthough Task objective-components are operable.
9.
Critical ElementFINAL Task objective-identifies that initial conditions are not satisfied for Core Alts.
10.
NOT critical this step does not require any evaluation from the candidate.
SRO Critical ElementIDENTIFIES and IMPLEMENTS the Actions for License 11 requirement.
SRO Critical ElementIDENTIFIES and IMPLEMENTS the Actions for License requirement.
Conditional Critical ElementIDENTIFIES and IMPLEMENTS the Actions for License requirement.
IF the candidate implements this TS condition, then the actions must also be initiated to comply with License requirements.
GENERAL TOOLS AND EQUIPMENT:
I.
References to be accessed via classroom computer (using exam ID logon) and Exam Reference 2.
3.
4.
12 SRO 13 FNP HL T-33 ADMIN A.1.1.A Page 7 of 10 GENERAL TOOLS AND EQUIPMENT:
I. References to be accessed via classroom computer (using exam ID logon) and Exam Reference disk/files only.
- 2. Provide UOP-4.1, Appendix 6
- 3. Pen/pencil
- 4. Pictures of the following components: (11 pages total)
- a. R-24A Local indications
- g. R24A & R25A MCR indications
- b. R-24B Local indications
- h. R-24B & R25B MCR indications
- c. R-25A Local indications
- i.
R-35A & R35B MCR indications
- d. R-25B Local indications
- e. R-35A Local indications
- f. R-35B Local indications J.
N31 drawer indications
- k. N32 drawer indications
- l.
N3I, N32 & N48 Board indications
- m. Audio Count Rate Channel Drawer Critical ELEMENT justification:
NOTE:
- 1.
- 2.
- 3.
- 4.
- 5.
- 6.
- 7.
- 8.
- 9.
- 10.
SRO II SRO 12 SRO 13 Evaluation The listed sub-steps to each element (a.b.c... ) describe the anticipated actions taken by the candidate. Completion of all sub steps is not required to make the determination of operability of the components, nor is it intended that these actions be observed. These sub-steps are provided for the examiner as an evaluation tool.
NOT critical-this step is for satisfying site procedural protocol and expectations.
NOT critical-this step is for satisfying site level expectations or requirements.
NOT critical-Although Task objective-component is operable.
Critical Element -Task objective-evaluate operability of required equipment.
NOT critical-Although Task objective-component is operable.
NOT critical -
this step does not require any evaluation from the candidate.
Critical Element -Task objective-evaluate operability of required equipment.
NOT critical-Although Task objective-components are operable.
Critical Element-FINAL Task objective-identifies that initial conditions are not satisfied for Core Alts.
NOT critical-this step does not require any evaluation from the candidate.
Critical Element-IDENTIFIES and IMPLEMENTS the Actions for License requirement.
Critical Element-IDENTIFIES and IMPLEMENTS the Actions for License requirement.
Conditional Critical Element-IDENTIFIES and IMPLEMENTS the Actions for License requirement.
IF the candidate implements this TS condition, then the actions must also be initiated to comply with License requirements.
FNP HLT-33 ADMIN A.1.1.A Page 2 of 10 COMMENTS:
References expected to be utilized:
STP-1.0 Table 1 and Appendix 3, Ver. 97.0; channel check information M-50 Master list of Surveillance Tests, Ver. 25.0; determine channel check procedure SOP-3 9.0, Nuclear Instrumentation System, Ver. 9.0; evaluate switch alignment/operability SOP-45.0, Radiation Monitoring System, Ver. 35.0; evaluate switch alignment/operability TRM 13.3.4 and Basis, Ver. 3.0 TS 3.9.2 and Basis, Amendment 146 (U-I), Amendment 137 (U-2)
UOP-4. 1, CONTROLLING PROCEDURE FOR REFUELING, Ver. 51.0 TO [)ISCUSS with the examiner:
Due to procedure rules of usage, a candidate may stop performing the steps upon encountering the first malfunction; therefore the candidate may not have evaluated all components/steps expected.
Possible options:
Brief all students verbally prior to providing the task or provide the following as part of the initial conditions:
For the purposes of this task, complete the Appendix in its entirety. Any actions, permissions or notifications required by any identified discrepancies, or procedure step, will be conducted upon completion of all other steps of the Appendix. Annotate any condition identified or the action, the permission, or the notification required for discussion upon completion of the appendix for any, if any, step that can not be completed.
Facilitate with Cues at the evaluation phase to resolve the expected actions while monitoring the completion of the remaining elements.
Facilitate with Cues at the evaluation phase then return the candidate to complete the remaining elements.
FNP HL T-33 ADMIN A.1.1.A Page 2 of 10 COMMENTS:
References expected to be utilized:
STP-1.0 Table 1 and Appendix 3, Ver. 97.0; channel check information M-50 Master list of Surveillance Tests, Ver. 25.0; determine channel check procedure SOP-39.0, Nuclear Instrumentation System, Ver. 9.0; evaluate switch alignment/operability SOP-45.0, Radiation Monitoring System, Ver. 35.0; evaluate switch alignment/operability TRM 13.3.4 and Basis, Ver. 3.0 TS 3.9.2 and Basis, Amendment 146 (U-l), Amendment 137 (U-2)
UOP-4.1, CONTROLLING PROCEDURE FOR REFUELING, Ver. 51.0
'1"0 DISCUSS with the examiner:
Due to procedure rules of usage, a candidate may stop performing the steps upon encountering the first malfunction; therefore the candidate may not have evaluated all components/steps expected.
Possible options:
Brief all students verbally prior to providing the task or provide the following as part of the initial conditions:
For the purposes of this task, complete the Appendix in its entirety. Any actions, permissions or notifications required by any identified discrepancies, or procedure step, will be conducted upon completion of all other steps of the Appendix. Annotate any condition identified or the action, the permission, or the notification required for discussion upon completion of the appendix for any, if any, step that can not be completed.
Facilitate with Cues at the evaluation phase to resolve the expected actions while monitoring the completion of the remaining elements.
Facilitate with Cues at the evaluation phase then return the candidate to complete the remaining elements.
FNP HLT-33ADMIN A.1.1.SRO HANDOUT #2 Pg 1 of 1 CONDITIONS Based on the conditions you identified evaluate TS and TRM requirements and perform the following:
a.
List all TS CONDITIONS, REQUIRED ACTIONS and COMPLETION TIMES for LCOs not met, if any.
FNP HLT-33 ADMIN A.1.1.SRO HANDOUT #2 Pg 1 of 1 CONDITIONS Based on the conditions you identified evaluate TS and TRM requirements and perform the following:
- a.
List all TS CONDITIONS, REQUIRED ACTIONS and COMPLETION TIMES for LCOs not met, ifany.
FNPHLT-33ADMIN A.1.1.A HANDOUT Page 1 of 1 CONDITIONS When I tell you to begin, you are to perform or evaluate all steps FNP-l-UOP-4.1, Appendix 6, Verification of Initial Conditions Prior to Core Alterations. The conditions under which this task is to be performed are:
a.
Core Offload was suspended 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> ago.
b.
RCStempis 100°F.
c.
Time to boil is >2 hours.
d.
A Fuel shuffle in the Spent Fuel Pool is ongoing.
e.
Communications between the MCR and a System Operator acting as the Cavity Watch have been verified available 10 minutes ago.
f.
RCS boron concentration is 2450 ppm following an over boration performed last shift.
g.
Rx Makeup water is being used to dilute the Reactor Cavity by chemistrys request.
h.
FNP-1-STP-18.4, Containment Midloop and/or Refueling Integrity Verification and Containment Closure, was completed yesterday at 1400.
i.
CTMT Main Purge system is in operation.
j.
The equipment hatch and all containment air locks are currently closed.
k.
Applicable equipment can be assessed using the provided photographs.
I.
You have been directed to perform or evaluate all steps of FNP-1-UOP-4.1, Appendix 6.
- m. Determine whether or not all Core Alterations initial conditions are satisfied.
FNP HLT-33 ADMIN A.1.1.A HANDOUT Page 1 of 1 CONDITIONS When I tell you to begin, you are to perform or evaluate all steps FNP-I-UOP-4.1, Appendix 6, Verification of Initial Conditions Prior to Core Alterations. The conditions under which this task is to be performed are:
- a. Core Offload was suspended 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> ago.
- b. RCS temp is 100°F.
- c. Time to boil is >2 hours.
- d. A Fuel shuffle in the Spent Fuel Pool is ongoing.
- e. Communications between the MCR and a System Operator acting as the Cavity Watch have been verified available 10 minutes ago.
- f.
RCS boron concentration is 2450 ppm following an over boration performed last shift.
- g. Rx Makeup water is being used to dilute the Reactor Cavity by chemistry's request.
- h. FNP-I-STP-18.4, Containment Midloop and/or Refueling Integrity Verification and Containment Closure, was completed yesterday at 1400.
I.
CTMT Main Purge system is in operation.
J.
The equipment hatch and all containment air locks are currently closed.
- k. Applicable equipment can be assessed using the provided photographs.
- 1.
You have been directed to perform or evaluate all steps ofFNP-I-UOP-4.1, Appendix 6.
- m. Determine whether or not all Core Alterations initial conditions are satisfied.
07/02/09 06:39:16 FNP-1-UOP-4.1 APPENDIX 6 VERIFICATION OF INITIAL CONDITION S PRIOR TO CORE ALTERATIONS Performed by:
Date Reviewed by:
Date This appendix consists of 2 pages Version 51.0 07/02/0906:39: 16 APPENDIX 6 VERIFICA TION OF INITIAL CONDITIONS PRIOR TO CORE ALTERATIONS FNP-I-UOP-4.1 Perfurmedby: ______________________ __ Date -----------------------
Reviewed by:
Date -----------------------
This appendix consists of 2 pages Version 51.0
07/02/0906 39 16 FNP-1-UOP-4 I APPENDIX 6 APPENDIX 6 VERIFICATION OF INITIAL CONDITIONS PRIOR TO CORE ALTERATIONS
1.0 Purpose
Provide separate guidance for specific initial conditions which will be needed to be verified more than once due to core alterations occurring at various time during an outage period.
2.0 Initial Conditions:
2.1 The version of this appendix is the current version. (OR 1-98-498) 2.2 This appendix is correct for the unit for which the task applies. (OR 1-98-498)
3.0 Instructions
NOTE:
This appendix should be completed each time core alterations are commenced, including suspensions of core alterations exceeding 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.1 The direct communications system between the control room and the reactor
/
cavity is verified available for use within one hour prior to core alterations.
CAUTION:
The Gamma-Metrics source range channel may only be used as a back-up to N-31 or N-32 during certain core configurations. The Refueling Coordinator should be consulted if N-31 or N-32 becomes inoperable when the core is not fully loaded.
3.2 Verify at least two of the required source range neutron flux monitors are operable with continuous visual indication in the Control Room and a channel check performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to any core alterations. Trip functions and associated TSLBs are not required for Mode 6 or core alterations.
3.2.1 Source Range Nuclear Instrument, Channel N-31 3.2.2 Source Range Nuclear Instrument, Channel N-32 3.2.3 Gamma-Metrics Neutron Flux Monitor, Channel N-48 Page 1 of 2 Version 51.0 07/02109 06:39: 16 FNP-1-UOP-4.1 APPENDIX 6 APPENDIX 6 VERIFICA nON OF INITIAL CONDITIONS PRIOR TO CORE ALTERA nONS
1.0 Purpose
Provide separate guidance for specific initial conditions which will be needed to be verified more than once due to core alterations occurring at various time during an outage period.
2.0 Initial Conditions:
2.1 The version of this appendix is the current version. (OR 1-98-498) 2.2 This appendix is correct for the unit for which the task applies. (OR 1-98-498)
3.0 Instructions
NOTE:
This appendix should be completed each time core alterations are commenced, including suspensions of core alterations exceeding 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.1
_1-The direct communications system between the control room and the reactor cavity is verified available for use within one hour prior to core alterations.
CAUTION:
The Gamma~Metrics source range channel may only be used as a back~up to N~31 or N~32 during certain core configurations. The Refueling Coordinator should be consulted ifN-31 or N~32 becomes inoperable when the core is not fully loaded.
3.2 Verify at least two of the required source range neutron flux monitors are operable with continuous visual indication in the Control Room and a channel check performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to any core alterations. Trip functions and associated TSLBs are not required for Mode 6 or core alterations.
3.2.1 Source Range Nuclear Instrument, Channel N-31 3.2.2 Source Range Nuclear Instrument, Channel N-32 3.2.3 Gamma-Metrics Neutron Flux Monitor, Channel N-48 Page 1 of2 Version 51.0
07/02/09 06 39 16 FNP-1-UOP-4 1 APPENDIX 6 3.3 The reactor coolant system boron concentration is> 2000 ppm.
3.4 The following radiation monitors are in operation per FNP-1-SOP-45.0, RADIATION MONITORING SYSTEM, or required action is being taken per Tech. Specs 3.3.6, 3.3.7, and 3.3.8, and TR 13.3.4.
3.4.1 R-5 Spent fuel storage or portable monitoring instrumentation used to monitor SFP area at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (for fuel movement in SFP).
3.4.2.
R-25A or B Spent fuel storage Gaseous (for fuel movement in SFP).
3.4.3 R-24A or B
- Containment purge system with main or mini purge in operation.
3.4.4 R-35A or B
- Control Room HVAC 3.5 Refueling integrity has been verified per FNP-1-STP-1 8.4, CONTAINMENT
/
MIDLOOP AND/OR REFUELING INTEGRITY VERIFICATION AND CONTAINMENT CLOSURE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of core alterations.
3.6 jf the containment equipment hatch, personnel air lock, or auxiliary air lock is open during core alterations and/or during movement of irradiated fuel assemblies within containment, THEN ensure the following items are complete (TS 3.9.3):
/
3.6.1 A designated, trained Maintenance Closure Response Team (MCRT)
MM is available to shut the containment equipment hatch, a door in the personnel air lock, and a door in the auxiliary air lock within two hours after notification and direction from the control room.
/
3.6.2 For the equipment hatch, the surveillance is current for SR 3.9.3.3, i.e.,
FNP-0-STP-610.0 has been completed by Maintenance within the last seven days Page 2 of 2 Version 51.0 07/02/0906:39:16 FNP-I-UOP-4.1 APPENDIX 6
_/-
_/-
MM 3.3 The reactor coolant system boron concentration is:::: 2000 ppm.
3.4 The following radiation monitors are in operation per FNP-l-S0P-45.0, RADIA nON MONITORING SYSTEM, or required action is being taken per Tech. Specs 3.3.6, 3.3.7, and 3.3.8, and TR 13.3.4.
3.5 3.4.1 R-5 Spent fuel storage or portable monitoring instrumentation used to monitor SFP area at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (for fuel movement in SFP).
3.4.2. R-25A or B Spent fuel storage - Gaseous (for fuel movement in SFP).
3.4.3 R-24A or B - Containment purge system with main or mini purge in operation.
3.4.4 R-35A or B - Control Room HVAC Refueling integrity has been verified per FNP-l-STP-18.4, CONTAINMENT MIDLOOP AND/OR REFUELING INTEGRITY VERIFICA nON AND CONTAINMENT CLOSURE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of core alterati ons.
3.6 IF the containment equipment hatch, personnel air lock, or auxiliary air lock is open during core alterations and/or during movement of irradiated fuel assemblies within containment, THEN ensure the following items are complete (TS 3.9.3):
3.6.1 A designated, trained Maintenance Closure Response Team (MCRT) is available to shut the containment equipment hatch, a door in the personnel air lock, and a door in the auxiliary air lock within two hours after notification and direction from the control room.
3.6.2 For the equipment hatch, the surveillance is current for SR 3.9.3.3, i.e.,
FNP-0-STP-61 0.0 has been completed by Maintenance within the last seven days Page 2 of2 Version 51.0
CD NUCLEAR INSTRUMENTATION PROTECTION CHANNEL I
SOURCE RANGE V
d
-.I (104
/
7J N1C55N10031 CHANNtL Off TEST LOSS OF O!TgCTOfl VOLT.
TRIP I
I 118V, 5A, sm POll/ER I
Pung Coro1 Pc Fuses wrtout prior Deacivaon w4 Scn Ct EvacuaDon Maim 118V 3A.AC COf4UOL POWER I
o 11 V "J' }'\\(
III.; III POI/,TI\\
o NUCLEAR INSTRUMENTATION PROTECTION CHANNEL I 11~V. ~;A, AC CO'lmOL POWER
C NUCLEAR INSTRUMENTATION PROTECTION CHANNEL II SOURCE RANGE 0
L Ni C55NIO32 INSTRUMCWr CHANNEL ON POWER ON TEST CONTROL LGSS or POWER ON OCIECTOR VOLT.
F IISV, 5A,AC INSTR POWER LEVEL TRIP OPERATION SELECTOR HIGH FLUX AOJ AT SHUTDOWN fJC:rA I
F r:rr-!L\\
6LOC$
I PfALi S
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Ft.se pror Dea-iaor w!
SOUndCT Evacuaon Jarm 118V, 5A,AC CONTROL POWER o
11 ~l V 5fl." l'.. C III ~ r I~
PO\\NER o
NUCLEAR INSTRUMENTATION PROTECTION CHANNEL II 118V 51\\ Ae CCHIll
- 2 SGs Avail with loops filled (Ref.
- 2) Zero (0) written on the line for S /
- 3. LINE 2:Cavity level> 152 9
- 3) Zero (0) written on the line for S / U item 2.
- 4. LINE 3:RHR subsystems Available (0,1, or 2)
- 4) Two (2) written on the line for S /
- 5. LINE 4:RCS level> 126 6
- 5) Zero (0) written on the line for S /
- 6) Zero (0) written on the line for S /
- 7) Two (2) written on line for Core S /
- 8. Evaluates Core Cooling condition
- 8) Circle around ORANGE S /
- 2. LINE 1:22 SGs Avail with loops filled (Ref.
- 3. LINE 2:Cavity level:::: 152' 9"
- 4. LINE 3:RHR subsystems Available (0,1, or 2)
- 5. LINE 4:RCS level:::: 126' 6"
- 6. LINE 5:Time to saturation >30 minutes OR RCS press >325 psig with at least one RCP available for operation and at least one SG available
- 7. Core Cooling Subtotal Page 3 of 5 STANDARDS:
- 2) Zero (0) written on the line for item 1.
- 3) Zero (0) written on the line for S / U item 2.
- 4) Two (2) written on the line for S / U item 3.
- 5) Zero (0) written on the line for S / U item 4.
- 6) Zero (0) written on the line for S / U item 5.
- 7) Two (2) written on line for Core S / U Cooling Subtotal.
- 8. Evaluates Core Cooling condition
- 8) Circle around ORANGE condition for the Core Cooling Function OR., if cued per NOTE State the condition is ORANGE Terminate when all elements of the task have been completed.
- 1. FNP-0-UOP-4.0, VER 36.0
- 1. FNP-0-UOP-4.0, VER 36.0
- 1. References to be accessed via classroom computer (using exam 10 logon) and Exam Reference disk/files only.
- 2. Provide copy ofFNP-0-UOP-4.0 Table A, Table B, Appendix 1
- 3. Pen/Pencil
- 4. Calculator (optional-not required to complete this task)
- 1.
- 2.
- 3.
- 4.
- 5.
- 6.
- 7.
- 8.
- 1. FNP-0-UOP-4.0, ver 36.0
- 2.
- 3.
- 4.
- 2: 2 SGs A vail with loops filled (Ref step 2.7) 0
- Q,-L _
- _
- _~IiP. _
- _
- _
- _4f _
- _ * "':'
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- 5.
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- 4.
- 2: 2 SFP Makeup Sources (RWST, DW, RMW to Blender, Boric Acid to 7
- a. Unit 1 is in a Refueling Outage, and has been shutdown for 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />.
- c. RCS temperature is 120°F.
- d. The RCS is at mid-loop with a FULL core.
- e. Both trains of RHR are in operation in the cooldown mode.
- f.
- g. The IPC is not available.
- h. Another operator is evaluating the other SHUTDOWN SAFETY FUNCTION/CRITERIAS.
- 1.
- The Shutdown Safety Assessment form may be a two-side copy.
- The Shift Supervisor will ensure that the appropriate Shutdown Safety Assessment Form is completed at about 0200 and 1400 when in Mode 4, Mode 5, Mode 6 or defueled. (CR 2004102447) 4.1 Obtain a copy of the appropriate Shutdown Safety Assessment form. For modes 5, 6, and defueled use Figure IA and for mode 4 use Figure lB.
- The Shutdown Safety Assessment form may be a two-side copy.
- The Shift Supervisor will ensure that the appropriate Shutdown Safety Assessment Form is completed at about 0200 and 1400 when in Mode 4, Mode 5, Mode 6 or defueled. (CR 2004102447) 4.1 Obtain a copy of the appropriate Shutdown Safety Assessment form. For modes 5, 6, and defueled use Figure 1A and for mode 4 use Figure 1 B.
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- 4. z 2 SFP Makeup Sources (RWST, OW, RMW to Blender, Boric Acid to 7
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- An offsite power feed through its associated startup transformer.
- An operable charging pump in the boration flow path
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- 2. Provide the examinee with the required materials to perform this JPM.
- 2) Value entered: 90 % entered S / U FNP HL T-33 ADMIN A.2.1.A Page 2 of 7 CONDITIONS When I tell you to begin, you are to Determine if Shutdown Margin is adequate using STP-29.5, SHUTDOWN MARGIN CALCULATION IN MODES 1 AND 2 (TAVG 2: 547°F), for Unit 1. The conditions under which this task is to be perfonned are:
- a. Unit 1 is stable at 90% with the ramp on hold
- b. Bank D indicates 192 by Group Demand.
- c. Seven of the Bank D rods (H2, B8, H 14, F6, FlO, K 10, K6) are at 192 steps by D RPI.
- d. Rod P8, in the D bank, has been detennined to be stuck.
- e. Rod P8 is at 162 steps by DRPI.
- f.
- g. Core bumup is 9,800 MWD/MTU bumup.
- h. FNP-I-STP-29.5, SHUTDOWN MARGIN CALCULATION IN MODES 1 AND 2 (TA VG ~
- 1.
- 1. Step A.I: Document Core Burnup
- 1) Value entered: 9,800 MWD S I U
- 2. Step A.2: Document Power Level
- 2) Value entered: 90 % entered S I U
- 3) Value entered: ZERO (0)
- 4) Value entered: 5 (RNG: 3-8)
- 5) Value entered: 6268 pcm.
- 6) Value entered of 0 (zero) pcm
- 7) Value entered of 50 pcm (range: 30-80)
- 3. Step A.3; Determine penalty steps for Banks below RIL.
- 4. Step A.4: Determine number of penalty steps for RODS below RIL
- 5.
- 6.
- 7.
- 3) Value entered; ZERO (0)
- 4) Value entered: 5 S f U (RNG; 3-8)
- i. Rod # -7 P8 @ 162
- 5) Value entered: 6268 pcm.
- 6) Value entered of 0 (zero) pcm
- 7) Value entered of 50 pcm (range; 30-80)
- 8) Value entered of 1406 pcm S /
- of stuck: 1 rod (b).
- 10. Step B.5; Calculate penalized rod worth
- 10) Value entered of:
- 11. Step B.6; Determine power defect for given
- 11) Value entered of:
- 8. Step B.4.a; Determine most reactive Rod worth
- 9. Step B.4.b; Calculate rod worth of stuckluntrippable rod
- 10. Step B.5; Calculate penalized rod worth
- 11. Step B.6; Determine power defect for given conditions Page 4 of 7 STANDARDS:
- 8) Value entered of 1406 pcm (range: NONE)
- 9)
- 10) Value entered of:
- 11) Value entered of:
- 12. Step B.8; Calculate available Shutdown 12)
- 1. B.5: (-) 2432.7 ii. B.6: 1634 iii. Calculate
- 13. Step B.9; Calculate Excess Shutdown Margin 13)
- i. B.8: (-) 748.7 ii. Calculate NOTE: Element 15 is not required to be completed but is likely to be performed if initiated as a group/classroom setting, therefore it may be necessary to provide the following cue during review/evaluation of work sheet:
- 14. Step B.1O: Identifies Need to Emergency
- 14) Notifies Shift Supervisor of S /
- 15. completes task:
- 15) Completes sign off.
- 12. Step B.8; Calculate available Shutdown Reactivity STANDARDS:
- 12) Value entered of:
- i. B.5: (-) 2432.7 ii. B.6: 1634 iii. Calculate Page 5 of 7 RESULTS:
- 13. Step B.9; Calculate Excess Shutdown Margin
- 13) Value entered of:
- i. B.8: (-) 748.7 ii. Calculate NOTE: Element 15 is not required to be completed but is likely to be performed if initiated as a group/classroom setting, therefore it may be necessary to provide the following cue during review/evaluation of work sheet:
- 14. Step B.I0: Identifies Need to Emergency Borate OR Notifies SS that SDM is NOT adequate (does not meet acceptance criteria).
- 15. completes task:
- 14) Notifies Shift Supervisor of S / U need to Emergency Borate.
- 15) Completes sign off.
- 1. FNP-I-STP-29.5, VER4.0
- 1. FNP-I-STP-29.5, ver 4.0
- 2. PCB-VOLl-CRV77, Cycle 23 rev 8
- 3. PCB-VOLl-CRV78, Cycle 23 rev 8
- 5. Pen/Pencil
- 6. calculator Critical ELEMENT justification:
- I 0
- ~_~"'1-6-2-6-8-""lp:~-:
- rrnt, emergency boration completed Verified by:
- .;X[~5~,~9It35161 Penalized rcfrT'~"""~~mm~.,.;MIMT!"""""""'~~I"IT'I"i~---I misaligned rods, rods below the insertion limit, and uncertainty:
- .,) *. l..
- 1.
- 2.
- 3.
- 150 658
- 1000 676
- 2000 732
- 3000 781
- 4000 889
- 5000 1024
- 6000 1160
- 7000 1232
- 9000 1355
- 10000 1406
- 11000 1450
- 12000 1494
- 13000 1528
- 14000 1561
- 15000 1592
- 16000 1623
- 17000 1653
- 18000 1684
- 19000 1713
- 20000 1741
- 20885 1767
- 1.
- 2.
- 3.
- 4.
- 1.
- i 100 0
- t 125 I'
- See special test exceptions of above specifications TEST RESULTS (TO BE COMPLETED BY TEST PERFORMER)
- See special test exceptions of above specifications TEST RESULTS (TO BE COMPLETED BY TEST PERFORMER)
- For calculating shutdown margin at power, the amount of reactivity by which the reactor would be subcritical from its present condition must be determined.
- Verification of shutdown margin calculation by a licensed individual other than the test performer, should take place as soon as practical, and any differences resolved immediately.
- A Shift Foreman that serves as verifier of the calculation may also serve as reviewer on the Surveillance Test Review Sheet.
- For calculating shutdown margin at power, the amount of reactivity by which the reactor would be subcritical from its present condition must be determined.
- Verification of shutdown margin calculation by a licensed individual other than the test performer, should take place as soon as practical, and any differences resolved immediately.
- A Shift Foreman that serves as verifier of the calculation may also serve as reviewer on the Surveillance Test Review Sheet.
- ()
- 150
- 1000
- 2000
- 3000
- 4000
- 5000
- 6000
- 7000
- 8000
- 9000
- 10000
- 11000
- 12000
- 13000
- 14000
- 15000
- 16000
- 17000
- 18000
- 5:19000
- 5:20000
- 5:20885 Notes:
- 1.
- 2.
- 3.
- /#
- k. f-o/
- 150 658
- 1000 676
- 2000 732
- 3000 781
- 4000 889
- 5000 1024
- 6000 1160
- 7000 1232
- 8000 1304
- 9000 1355
- 10000 1406
- 11000 1450
- 12000 1494
- 14000 1561
- 15000 1592
- 17000 1653
- 18000 1684
- 19000 1713
- 20000 1741
- 20885 1767
- 1.
- 2.
- 3.
- 4.
- 1.
- iii 100 0
- a. Unit I is stable at 90% with the ramp on hold
- b. Bank 0 indicates 192 by Group Demand.
- c. Seven of the Bank 0 rods (H2, B8, H14, F6, FlO, KlO, K6) are at 192 steps by DRPI.
- d. Rod P8, in the 0 bank, has been detennined to be Stuck.
- f. All other rods are at 229 steps.
- g. Core bumup is 9,800 MWD/MTU bumup.
- h. FNP-1-STP-29.5, SHUTDOWN MARGIN CALCULATION IN MODES I AND 2 (TAVG 2::
- 1.
- 1. Initiation of task may be in group setting, evaluation performed individually upon completion.
- 2. The references for this task will be provided as listed or the student may be provided a computer with a generic exam login and access to the EXAM reference disk.
- 3. Elements 1 through 7 may be evaluated by reviewing the responses on the Handout.
- If any member of the team will NOT exceed limits prior to performing task 4, THEN calculate the maximum allowable stay time to complete task 4, and remain within the dose limits.
- IF any member of the team WILL exceed limits prior to performing task 4, THEN identify the last tasks (if any) that the team can complete without either member exceeding dose limits.
- a. A General Emergency has been declared on Unit 1.
- c. The TSC has requested that 1 B RHR pump motor bearing oil levels be checked and filled as required after restoring lA RHR and suggests using the same Repair team. The time required to perform this task is unknown, but is estimated to range between 1 to 20 minutes.
- d. Ted and Joel have been selected to perform the task, their exposure information is stated below.
- e. The Tasks are provided in the table below, and estimated times and doses have been provided.
- f.
- g. Both operators are expected to receive equal dose for each job.
- h. The Emergency Director (ED) directs you to perform the following with the information provided:
- If any member of the team will NOT exceed limits prior to performing task 4, THEN calculate the maximum allowable stay time to complete task 4, and remain within the dose limits.
- IF any member of the team WILL exceed limits prior to performing task 4, THEN identify the last tasks (if any) that the team can complete without either member exceeding dose limits.
- c. Task #3 (20min)x 5.65R / hr x1 lhr
- 1) Calculates (in REM) a.
- 2) Circles YES on handout OR S / U limits of EIP-14.0 for completion of tasks I states that team can perform through 3.
- 3. Calculates stay time for task #4.
- 3) Documents or States: STAY S / U 4
- 1. Determine the dose received for tasks 1 through 3.
- a. Task#1 (30min)x 5.31R/hr(
- b. Task #2 (I5min)x19.75RIhrx 1h
- 1. Determine the dose received for tasks 1 through 3. 1) Calculates (in REM)
- a. Task #1
- a. 2.66
- b. Task#2
- b. 4.938 {4.9 to 5.0}
- c. Task #3 (20min)X5.65Rlhrx( Ihr. )=1.88REM
- c. 1.88
- 2. Evaluates Team exposure within emergency limits ofEIP-14.0 for completion of tasks 1 through 3.
- a.
- b.
- c.
- 3. Calculates stay time for task #4.
- 2) Circles YES on handout -OR-states that team can perform tasks 1 through 3.
- a. lOR is limit
- b. 0.52 R remains available
- c. Circles the YES choice
- 3) Documents or States: STAY time is 4.5 mins
- 1. FNP-O-EIP-I4.0, ver 22
- 2. FNP-O-M-l.O, ver 18.0
- 3. KA: 02.3.4 RO 3.2 GENERAL TOOLS AND EQUIPMENT:
- 1. Computer with access to Exam Reference Disk, or EIP 14.0, ver 22.0 and M-I.O, version 18.0.
- 2. Calculator
- 3. pens/pencils
- 4. Scrap paper Critical ELEMENT justification:
- If any member of the team will NOT exceed limits prior to performing task 4, THEN calculate the maximum allowable stay time to complete task 4, and remain within the dose limits.
- IF any member of the team WILL exceed limits prior to performing task 4, THEN identify the last tasks (if any) that the team can complete without either member exceeding dose limits.
- a. A General Emergency has been declared on Unit 1.
- c. The TSC has requested that 1 B RHR pump motor bearing oil levels be checked and filled as required after restoring 1 A RHR and suggests using the same Repair team. The time required to perform this task is unknown, but is estimated to range between 1 to 20 minutes.
- d. Ted and Joel have been selected to perform the task, their exposure information is stated below.
- e. The Tasks are provided in the table below, and estimated times and doses have been provided.
- f.
- g. Both operators are expected to receive equal dose for each job.
- h. The Emergency Director (ED) directs you to perform the following with the information provided:
- If any member of the team will NOT exceed limits prior to performing task 4, THEN calculate the maximum allowable stay time to complete task 4, and remain within the dose limits.
- IF any member of the team WILL exceed limits prior to performing task 4, THEN identify the last tasks (if any) that the team can complete without either member exceeding dose limits.
- 1. This task will be conducted on the Simulator.
- 2. The simulator will remain frozen for the duration to the task, if the candidate attempts to operate any component, no plant response will occur.
- 3. The plant computer screens will be turned off.
- 4. ALL will perform elements 1 through 7.
- 5. ONLY SRO will perform elements 8-10.
- 1) Determines Subcriticality SAT /
- a. YES (N42 is failed high) b.
- b. YES c.
- d. YES 2.
- 2) Determines Core Cooling OFF-S / U NORMAL / YELLOW (C.3).
- a. YES b.
- b. NO c.
- c. YES 3.
- 3) Determines Heat Sink is OFF-S / U NORMAL / YELLOW (H.5).
- a. NO b.
- b. YES c.
- c. YES d.
- d. YES e.
- e. YES f.
- f. NO FNP HLT-33 ADMIN AA.l.A Page 2 of6 CONDITIONS When I tell you to begin, you are to monitor the Critical Safety Function Status Trees (CSFST).
- a. A Large Break LOCA and loss of Off-site power has occurred on Unit 1 from 100% 10 minutes ago.
- b. The team has transitioned to EEP-1.0, Primary or Secondary Loss of Coolant.
- c. The Integrated Plant Computer has failed.
- d. The STA is en route to the Control Room.
- e. You have been directed to manually monitor Critical Safety Functions using CSp -0.0, Critical Safety Function Status Trees and:
- 1. Identify all applicable Critical Safety functions which are challenged, if any.
- 2. If applicable, identify the highest level challenge to the CSFSTs and the required procedure entry for that condition.
- 1.
- a. Power Rng < 5%
- b. Both Int RNG SUR zero or neg
- c. Both SR detectors ARE energized
- d. IR Range < -0.2 DPM
- 2.
- a. CET <1200 of
- b. Subcooling> 16°F {45°}
- c. CET <700 of
- 3.
- a. NR levels> 31 {48% }
- b. Total AFW flow >395 gpm.
- c. Press in all SG <1129 psig
- d. NR lvl <82%
- e. Press <1075 psig
- f.
- 1) Determines Subcriticality SAT /
- a. YES (N42 is failed high)
- b. YES
- c. NO
- d. YES
- 2) Determines Core Cooling OFF-S / U NORMAL / YELLOW (C.3).
- a. YES
- b. NO
- c. YES
- 3) Determines Heat Sink is OFF-S / U NORMAL / YELLOW (H.5).
- a. NO
- b. YES
- c. YES
- d. YES
- e. YES
- f. NO
- 4) Determines Integrity is S / U SEVERELY Challanged /
- a. NO b.
- b. YES (LOSP will result in instrument spike < 25 OF-indication onlyshould be determined as not valid).
- c. NO NOTE:
- The candidate may inform the SS of the FRP-Z. 1/ORANGE path condition upon discovery but should continue to evaluate the remaining CSFSTs. IF CUE required, then provide: SS Acknowledges.
- 5) Determines containment is in S / U SEVERELY Challenged!
- a. YES
- b. NO (35 psig)
- c. NO; recognizes I B CS pump has no flow path aligned (MOV-8820B closed) and IA CS pump is not running.
- 6) Determines containment is in S / U OFF-NORMAL / YELLOW (1.2).
- a. YES b.NO 5.
- 4.
- a. Temp decr <100 OF last 60 min
- b. All Press and CL temps to right of LIMIT A
- c. Al I CL temps> 250°F
- 4) Determines Integrity is SEVEREL Y Challanged /
- a. NO
- b. YES (LOSP will result in instrument spike < 250F-indication only-should be determined as not valid).
- c. NO S / U NOTE:
- The candidate may inform the SS of the FRP-Z.1I0RANGE path condition upon discovery but should continue to evaluate the remaining CSFSTs. IF CUE required, then provide: "SS Acknowledges."
- 5.
- a. CTMT press <54 psig
- b. CTMT press <27 psig
- c. AT least ONE CTMT Spray pump running with flow> 1000 gpm.
- 6.
- a. PRZR level < 92%
- b. PRZR level> 15%
- 5) Determines containment is in SEVEREL Y Challenged/
- a. YES
- b. NO (35 psig)
- c. NO; recognizes 1 B CS pump has no flow path aligned (MOV -8820B closed) and lA CS pump is not running.
- 6) Determines containment is in OFF-NORMAL / YELLOW (1.2).
- a. YES
- b. NO S / U S / U
- 8) RCS pressure checked on S / U than 435 psig:
- 9) Checks flow on FI-605A and S / U greater than 1500 gpm.
- 10. Transitions to procedure step in effect.
- 10) Properly assesses Procedure S / U step in effect is FRP-Z.I.
- 7.
- a. FRP-P.l, ORANGE due to Cooldown
- 8.
- 9.
- 10. Transitions to procedure step in effect.
- 8) RCS pressure checked on PI-402B and 403B (50 psig) and determined less than.
- 9) Checks flow on FI-605A and FI-605B. (3250 gpm)
- 10) Properly assesses Procedure step in effect is FRP-Z.l.
- 1. CSF -0.0, ver 17.0
- 2. FNP-1-FRP-P.1, ver 19
- a. A Large Break LOCA and loss of Off-site power has occurred on Unit 1 from 100% 10 minutes ago.
- b. The team has transitioned to EEP-I.O, Primary or Secondary Loss of Coolant.
- c. The Integrated Plant Computer has failed.
- d. The ST A is en route to the Control Room.
- e. You have been directed to manually monitor Critical Safety Functions using CSF -0.0, Critical Safety Function Status Trees and:
- 2. If applicable, identify the highest level challenge to the CSFSTs and the required procedure entry for that condition.
REFERENCES:
I.
References to be accessed via classroom computer (using exam ID logon) and Exam Reference disk/files only.
2.
Provide copy of FNP-0-UOP-4.0 Table A, Table B, Appendix 1 3.
Pen/Pencil 4.
Calculator (optional-not required to complete this task)
Critical ELEMENT justification:
STEP Evaluation I.
Critical
- this is first of the two major objectives of the task. Utilization of the plant parameters and selection of the appropriate table are required to properly assess Shutdown Core Cooling function.
2.
Critical
- Each line item is critical to evaluate proper assessment of each parameter.
The aggregate effect of errors made within each line could result in obtaining the correct end point; therefore each sub-step must be evaluated to ensure examinee is 6
correctly assessing the parameters.
7.
This is NOT a critical element since documentation of this summation is not procedurally driven nor will failure to complete this element necessarily impact the outcome.
8.
Critical
- This is the final objective of the task; proper assessment of this function is required to identify the need of mitigating actions necessary to ensure the health and safety of the public is not jeopardized.
COMMENTS:
References expected to be utilized:
1.
FNP-0-UOP-4.0, ver 36.0 FNP HLT-33 ADMIN A.1.2.S Page 4 of 5 GENERAL
REFERENCES:
Critical ELEMENT justification:
STEP
Evaluation Critical - this is first of the two major objectives of the task. Utilization of the plant parameters and selection of the appropriate table are required to properly assess Shutdown Core Cooling function.
Critical - Each line item is critical to evaluate proper assessment of each parameter.
The aggregate effect of errors made within each line could result in obtaining the correct end point; therefore each sub-step must be evaluated to ensure examinee is correctly assessing the parameters.
This is NOT a critical element since documentation of this summation is not procedurally driven nor will failure to complete this element necessarily impact the outcome.
Critical - This is the final objective of the task; proper assessment of this function is required to identify the need of mitigating actions necessary to ensure the health and safety of the public is not jeopardized.
COMMENTS:
References expected to be utilized:
1.
2 SGs Avail with loops filled (Ref step 2.7) 2.
Cavity level 1529 3.
RHR Subsystems Available (0, 1 or 2) 4.
RCS level 126 6 5.
Time to saturation > 30 minutes OR RCS press > 325 psig with at least one RCP available for operation and at least one SO available Core Cooling Subtotal I.
I A Train DG Available 2.
1 B Train DG Available 3.
F 4160 V BUS normal (Aligned to A SOT) 4.
G 4160 V BUS normal (Aligned to B SUT) 5.
2 Feeds available to the HV Switchyard (0, 1 or 2)
Power Availability Subtotal 0-I RED 2-3 ORANGE GREEN (GREEN if Defueled)
SHUTDOWN SAFETY FUNCTION! CRITERIA CONDITION (No/False=0, Yes/True=l, Use number within range when required)
(Circle Condition)
REACTIVITY Subtotal Condition AOP I.
No Core Alterations in Progress 0-I RED 41 2.
Number of Boration Flow Paths (0, 1, 2) (Ref step 2.13) 2 ORANGE 41 3.
RCS Boron: CSD/Refueling Concentration 3-4 YELLOW 41 4.
Source Range Instrumentation Available 5
GREEN (GREEN if Defueled)
Reactivity Subtotal CORE COOLING Subtotal Condition AOP 0
0 2
0 0
2 42 42
POWER AVAILABILITY Subtotal Condition Mm required is to CIRCLE the condition ORANGE including the subtotal or the AOP is not required.
0-2 RED 3
ORANGE 4-5 YELLOW 6
GREEN CONTAINMENT Subtotal Condition AOP I.
Refueling Integrity Set 0l RED 44 2.
CTMT Closure Set 2-4 ORANGE 44 3.
No Core Alterations in Progress (2 pts) 5-6 YELLOW 44 4.
Equipment Hatch & Air Locks Closed or Capable of Being Closed on 7
GREEN Short Notice 5.
RCS level 126 6 (3 pts)
(GREEN if Defueled)
Containment Subtotal INVENTORY Subtotal Condition AOP 1.
Refueling Cavity 23 Feet (142 1 ) Above Fuel 0
RED 45 2.
LHSI Pump/Flowpath Available I
ORANGE 45 3.
I-IHSI PumpfFlowpath Available 2
YELLOW 45 4.
RCS is Intact below the Reactor Vessel Flange GREEN (GREEN if Defueled)
Inventory Subtotal RCS INTEGRITY Subtotal Condition AOP I.
All S/G Manways or Nozzle Dams Installed 0-I ORANGE 46 2.
RCS is Intact below the Reactor Vessel Flange 2
YELLOW 46 3.
Pressurizer level < 100%
3 GREEN RCS Integrity Subtotal (GREEN if Defueled)
SPENT FUEL COOLING Subtotal Condition AOP I.
SFP level 23 feet (151 6) above fuel (4 pts) 0-4 RED 47 2.
A Tm SFP Cooling available 5
ORANGE 47 3.
B Tm SFP Cooling available 6
YELLOW 47 4.
2 SFP Makeup Sources (RWST, DW, RMW to Blender, Boric Acid to 7
GREEN Blender. RHT to transfer canal with weir gate removed)
SFP Subtotal Time to saturation If core cooling were lost:
0 flours 9.2 minutes Page 6 of 18 Current Version Header information is not required for this Task; but if requested use current date and time.
Prepared By:-=====~ __
ULv *.J....._--'-Time:_IH_H_M_MI App-l, Fig 2 Evaluated 0 SHUTDOWN SAFETY FUNCTIONI CRITERIA CONDITION (NolFalse=O, Yes/True=l, Use number within range when required)
(Circle Condition)
REACTIVITY Subtotal Condition AOP J.
No Core Alterations in Progress 0-1 RED 41
Number of Boration Flow Paths (0, 1,2) (Ref step 2.13) 2 ORANGE 41
RCS Boron: CSDfRefueling Concentration 3-4 YELLOW 41
Source Range Instrumentation Available 5
GREEN Reactivity Subtotal (GREEN ifDefueled)
CORE COOLING Subtotal Condition AOP I.
Cavity level 2152'9" 0
2-3 ORANGE 42
RHR Subsystems Available (0, lor 2)
_2_ F- '4"....... )-"ttI':<M".... 4'2"......
RCS level:2: 126' 6"
_0_
25 GREEN
Time to saturation> 30 minutes OR RCS press> 325 psig with at least 0
(GREEN ifDefueled)
Min required is to one RCP available for operation and at least one SG available "CIRCLE" the Core Cooling Subtotal 2
condition "ORANGE" POWER A V AILABILITY Subtotal Condition J.
I "A" Train DG Available 0-2 RED including the
I "B" Train DG Available 3
ORANGE subtotal or the AOP
F 4160 V BUS normal (Aligned to A SUT) 4-5 YELLOW is not required.
G 4160 V BUS normal (Aligned to B SUT) 6 GREEN
2 Feeds available to the HV Switchyard (0, I or 2)
Power Availabilitv Subtotal CONTAINMENT Subtotal Condition AOP I.
Refueling Integrity Set 0-1 RED 44
CTMT Closure Set 2-4 ORANGE 44
No Core Alterations in Progress (2 pts) 5-6 YELLOW 44
Equipment Hatch & Air Locks Closed or Capable of Being Closed on 7
GREEN Short Notice (GREEN ifDefueled)
RCS level:2: 126' 6" (3 pts)
Containment Subtotal INVENTORY Subtotal Condition AOP J.
Refueling Cavity :2: 23 Feet (142' 1") Above Fuel 0
RED 45
LHSI PumpfFlowpath Available 1
ORANGE 45
HHSI PumpfFlowpath Available 2
YELLOW 45
RCS is Intact below the Reactor Vessel Flange 3-4 GREEN (GREEN ifDefueled)
Inventory Subtotal RCS INTEGRITY Subtotal Condition AOP I.
All SfG Manways or Nozzle Dams Installed 0-1 ORANGE 46
RCS is Intact below the Reactor Vessel Flange 2
YELLOW 46
Pressurizer level < 100%
3 GREEN RCS Intee:ritv Subtotal (GREEN if Defueled)
SPENT FUEL COOLING Subtotal Condition AOP I.
SFP level :2: 23 feet (lSI '6") above fuel (4 pts) 0-4 RED 47
A Trn SFP Cooling available 5
ORANGE 47
B Trn SFP Cooling available 6
YELLOW 47
GREEN Blender, RHT to transfer canal with weir gate removed)
SFP Subtotal Time to saturation IE core cooling were lost:
0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 9.2 minutes Page 6 of 18 Current Version
FNPHLT-33ADMIN A.1.2S HANDOUT Pg 1 of I CONDITIONS When I tell you to begin, you are to determine the time to saturation and prepare the CORE COOLING SECTION ONLY of a shutdown safety assessment on Unit 1 per FNP-O-UOP-4.O, Appendix 1, step 4.3 and step 4.4. The conditions under which this task is to be performed are:
a.
Unit us in a Refueling Outage, and has been shutdown for 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />.
b.
1A SG wide range level is 74%; 113 SG wide range level is 75%; IC SG wide range level is 10%.
c.
RCS temperature is 120°F.
d.
The RCS is at mid-loop with a FULL core.
e.
Both trains of RHR are in operation in the cooldown mode.
f.
Mid-loop integrity per STP-1 8.4 has been verified current.
g.
The IPC is not available.
h.
Another operator is evaluating the other SHUTDOWN SAFETY FUNCTION/CRITERIAS.
i.
You are directed to determine the time to saturation and prepare the CORE COOLING SECTION ONLY of a shutdown safety assessment on Unit 1 per FNP-0-UOP-4.0, Appendix 1, step 4.3 and step 4.4.
FNP HLT-33 ADMIN A.1.2S HANDOUT Pg 1 of 1 CONDITIONS When I tell you to begin, you are to determine the time to saturation and prepare the CORE COOLING SECTION ONLY of a shutdown safety assessment on Unit 1 per FNP-O-UOP-4.0, Appendix 1, step 4.3 and step 4.4. The conditions under which this task is to be performed are:
Mid-loop integrity per STP-18.4 has been verified current.
You are directed to determine the time to saturation and prepare the CORE COOLING SECTION ONLY of a shutdown safety assessment on Unit 1 per FNP-0-UOP-4.0, Appendix 1, step 4.3 and step 4.4.
11/13/09 12:03:44 FNP-0-UOP-4.0 TABLE A TABLE A Power-Uprated Unit Time To Saturation: Full Core Assumed Initial Temperature = 100°F Time Time To Time To Time To Time To After Saturation Saturation Saturation Saturation RCS Shutdown at Midloop 3 ft Below Flange Full Reactor Cavity Full (hours)
(mm)
(mm)
(hours)
(hours) 40 7.7 10.5 5.6 1.3 60 8.7 11.9 6.3 1.4 80 9.5 13.0 6.9 1.6 100 10.4 14.2 7.5 1.7 120 11.3 15.4 8.2 1.9 140 11.9 16.3 8.6 2.0 160 12.7 17.4 9.2 2.1 180 13.3 18.2 9.6 2.2 200 13.9 19.0 10.1 2.3 336 17.1 23.4 12.4 2.9 504 20.8 28.5 15.1 3.5 VOLUME REFERENCE TABLE Midloop Volume (ft3):
945 Volume: 3 ft Below Flange (ft3) 348 Total =
1293 Volume: Full Reactor Cavity (ft3) 39750 Total =
41043 Volume: RCS Full (including Pzr) (ft3) 9591 Page 1 of 3 Version 36.0 11113/09 12:03:44 FNP-0-UOP-4.0 TABLE A TABLE A Power-Up rated Unit Time To Saturation: Full Core Assumed Initial Temperature = lOO°F Time Time To Time To Time To Time To After Saturation Saturation Saturation Saturation RCS Shutdown at Midloop 3 ft Below Flange Full Reactor Cavity Full (hours)
(min)
(min)
(hours)
(hours) 40 7.7 10.5 5.6 1.3 60 8.7 11.9 6.3 1.4 80 9.5 l3.0 6.9 1.6 100 10.4 14.2 7.5 1.7 120 11.3 15.4 8.2 1.9 140 11.9 16.3 8.6 2.0 160 12.7 17.4 9.2 2.1 180 13.3 18.2 9.6 2.2 200 l3.9 19.0 10.1 2.3 336 17.l 23.4 12.4 2.9 504 20.8 28.5 15.1 3.5 VOLUME REFERENCE TABLE Midloop Volume (ft3):
945 Volume: 3 ft Below Flange (ft3) 348 Total =
1293 Volume: Full Reactor Cavity (ft3) 39750 Total =
41043 Volume: RCS Full (including pzr) (ft3) 9591 Page 1 of3 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 TABLE A TABLE A Power-Uprated Unit Time To Saturation: Full Core Assumed Initial Temperature = 120°F Time Time To Time To Time To Time To After Saturation Saturation Saturation Saturation RCS Shutdown at Midloop 3 ft Below Flange Full Reactor Cavity Full (hours)
(mm)
(mm)
(hours)
(hours) 40 6.3 8.6 4.5 1.1 60 7.1 9.8 5.2 1.2 80 7.8 10.6 5.6 1.3 100 8.5 11.7 6.2 1.4 120 9.2 12.6 6.7 1.6 140 9.8 13.4 7
1 1.7 160 10.4 14.2 7.5 1.8 180 10.9 14.9 7.9 1.8 200 11.4 15.6 8.2 1.9 336 14.0 19.1 10.1 2.4 504 17.0 23.3 12.3 2.9 VOLUME REFERENCE TABLE Midloop Volume (ft3):
945 Volume: 3 ft Below Flange (ft3) 348 Total =
1293 Volume: Full Reactor Cavity (ft3) 39750 Total 41043 Volume: Full RCS (including Pzr) (ft3) 9591 Page 2 of 3 Version 36.0 11/13/09 12:03:44 TABLE A Power-Up rated Unit Time To Saturation: Full Core FNP-0-UOP-4.0 TABLE A Assumed Initial Temperature = 120°F Time Time To Time To Time To Time To After Saturation Saturation Saturation Saturation RCS Shutdown at Midloop 3 ft Below Flange Full Reactor Cavity Full (hours)
(min)
(min)
(hours)
(hours) 40 6.3 8.6 4.5 1.1 60 7.1 9.8 5.2 1.2 80 7.8 10.6 5.6 1.3 100 8.5 11.7 6.2 1.4 120 9.2 12.6 6.7 1.6 140 9.8 13.4 7.1 1.7 160 10.4 14.2 7.5 1.8 180 10.9 14.9 7.9 1.8 200 11.4 15.6 8.2 1.9 336 14.0 19.1 10.1 2.4 504 17.0 23.3 12.3 2.9 VOLUME REFERENCE TABLE Midloop Volume (ft3):
945 Volume: 3 ft Below Flange (ft3) 348 Total =
1293 Volume: Full Reactor Cavity (ft3) 39750 Total =
41043 Volume: Full RCS (including pzr) (ft3) 9591 Page 2 of3 Version 36.0
I 1/13/09 12:03:44 FNP-0-UOP-4.0 TABLE A TABLE A Power-Uprated Unit Time To Saturation: Full Core Assumed Initial Temperature 140°F Time Time To Time To Time To After Saturation Saturation Saturation Time To Saturation Shutdown at Midloop 3 ft Below Flange Full Reactor RCS Full (hours)
(mm)
(mm)
Cavity (hours)
(hours) 0.8 40 4.9 6.7 3.5 0.9 60 5.6 7.6 4.0 1.0 80 6.1 8.3 4.4 1.1 100 6.6 9.1 4.8 1.2 120 7.2 9.8 5.2 1.3 140 7.6 10.4 5.5 1.4 160 8.1 11.1 5.9 1.4 180 8.5 11.6 6.1 1.5 200 8.9 12.1 6.4 1.8 336 10.9 14.9 7.9 2.2 504 13.3 18.2 9.6 VOLUME REFERENCE TABLE Midloop Volume (ft3):
945 Volume: 3 ft Below Flange (ft3) 348 Total =
1293 Volume: Full Reactor Cavity (ft3) 39750 Total =
41043 Volume: Full RCS (including Pzr) (ft3) 9591 Page 3 of 3 Version 36.0 11113/0912:03:44 TABLE A Power-Up rated Unit Time To Saturation: Full Core Assumed Initial Temperature = 140°F Time Time To Time To Time To After Saturation Saturation Saturation Shutdown at Midloop 3 ft Below Flange Full Reactor (hours)
(min)
(min)
Cavity (hours) 40 4.9 6.7 3.5 60 5.6 7.6 4.0 80 6.1 8.3 4.4 100 6.6 9.1 4.8 120 7.2 9.8 5.2 140 7.6 10.4 5.5 160 8.1 11.1 5.9 180 8.5 11.6 6.1 200 8.9 12.1 6.4 336 10.9 14.9 7.9 504 13.3 18.2 9.6 VOLUME REFERENCE TABLE Midloop Volume (ft3):
945 Volume: 3 ft Below Flange (ft3) 348 Total =
Volume: Full Reactor Cavity (ft3) 39750 Total =
Volume: Full RCS (including pzr) (ft3) 9591 Page 3 of3 FNP-0-UOP-4.0 TABLE A Time To Saturation RCS Full (hours) 0.8 0.9 1.0 1.1 1.2 1.3 1.4 1.4 1.5 1.8 2.2 1293 41043 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 TABLE B TABLE B Power-Uprated Unit Time To Saturation: One-Third New Fuel Assumed Initial Temperature = 100°F Time Time To Time To Time To Time To After Saturation Saturation Saturation Saturation RCS Shutdown at Midloop 3 ft Below Flange Full Reactor Full (hours)
(mm)
(mm)
Cavity (hours)
(hours) 100 15.6 21.4 11.3 2.6 200 20.9 28.5 15.1 3.5 300 24.7 33.7 17.8 4.2 400 27.5 37.6 19.9 4.7 500 31.1 42.5 22.5 5.3 600 34.5 47.3 25.0 5.8 700 37.2 51.0 27.0 6.3 800 40.4 55.3 29.2 6.8 VOLUME REFERENCE TABLE Midloop Volume (ft3):
945 Volume: 3 ft Below Flange (ft3) 348 Total =
1293 Volume: Full Reactor Cavity (ft3) 39750 Total =
41043 Volume: Full RCS (including Pzr) (ft3) 9591 Page 1 of 3 Version 36.0 11113/0912:03:44 Time TABLEB Power-U prated Unit Time To Saturation: One-Third New Fuel Assumed Initial Temperature = lOO°F Time To Time To Time To FNP-0-UOP-4.0 TABLEB Time To After Saturation Saturation Saturation Saturation RCS Shutdown at Midloop 3 ft Below Flange Full Reactor Full (hours)
(min)
(min)
Cavity (hours)
(hours) 100 15.6 21.4 11.3 2.6 200 20.9 28.5 15.1 3.5 300 24.7 33.7 17.8 4.2 400 27.5 37.6 19.9 4.7 500 31.1 42.5 22.5 5.3 600 34.5 47.3 25.0 5.8 700 37.2 51.0 27.0 6.3 800 40.4 55.3 29.2 6.8 VOLUME REFERENCE TABLE Midloop Volume (ft3):
945 Volume: 3 ft Below Flange (ft3) 348 Total =
1293 Volume: Full Reactor Cavity (ft3) 39750 Total =
41043 Volume: Full RCS (including pzr) (ft3) 9591 Page 1 of3 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 TABLE B TABLE B Power-Uprated Unit Time To Saturation: One-Third New Fuel Assumed Initial Temperature = 120°F Time Time To Time To Time To Time To After Saturation Saturation Saturation Saturation RCS Shutdown at Midloop 3 ft Below Flange Full Reactor Full (hours)
(mm)
(mm)
Cavity (hours)
(hours) 100 12.8 17.5 9.2 2.2 200 17.1 23.4 12.4 2.9 300 20.2 27.6 14.6 3.4 400 22.5 30.8 16.3 3.8 500 25.4 34.8 18.4 4.3 600 28.3 38.7 20.5 4.8 700 30.5 41.7 22.1 5.2 800 33.0 45.2 23.9 5.6 VOLUME REFERENCE TABLE Midloop Volume (ft3):
945 Volume: 3 ft Below Flange (ft3) 348 Total =
1293 Volume: Full Reactor Cavity (ft3) 39750 Total 41043 Volume: Full RCS (including Pzr) (ft3) 9591 Page 2 of 3 Version 36.0 11113/09 12:03:44 Time TABLEB Power-Up rated Unit Time To Saturation: One-Third New Fuel Assumed Initial Temperature = 120°F Time To Time To Time To FNP-0-UOP-4.0 TABLE B Time To After Saturation Saturation Saturation Saturation RCS Shutdown at Midloop 3 ft Below Flange Full Reactor Full (hours)
(min)
(min)
Cavity (hours)
(hours) 100 12.8 17.5 9.2 2.2 200 17.1 23.4 12.4 2.9 300 20.2 27.6 14.6 3.4 400 22.5 30.8 16.3 3.8 500 25.4 34.8 18.4 4.3 600 28.3 38.7 20.5 4.8 700 30.5 41.7 22.1 5.2 800 33.0 45.2 23.9 5.6 VOLUME REFERENCE TABLE MidloopVolume (ft3):
945 Volume: 3 ft Below Flange (ft3) 348 Total =
1293 Volume: Full Reactor Cavity (ft3) 39750 Total =
41043 Volume: Full RCS (including pzr) (ft3) 9591 Page 2 of3 Version 36.0
11/13/09 12:03:44 TABLE B Power-Uprated Unit Time To Saturation: One-Third New Fuel Assumed Initial Temperature = 140°F FNP-0-UOP-4.0 TABLE B Time Time To Time To Time To Time To After Saturation Saturation Saturation Saturation RCS Shutdown at Midloop 3 ft Below Flange Full Reactor Full (hours)
(mm)
(mm)
Cavity (hours)
(hours) 100 10.0 13.6 7.2 1.7 200 13.3 18.2 9.6 2.2 300 15.7 21.5 11.4 2.7 400 17.5 24.0 12.7 3.0 500 19.8 27.1 14.3 3.3 600 22.0 30.1 15.9 3.7 700 23.7 32.5 17.2 4.0 800 25.7 35.2 18.6 4.3 VOLUME REFERENCE TABLE 945 Midloop Volume (ft3):
Volume: 3 ft Below Flange (ft3) 348 Total =
1293 Volume: Full Reactor Cavity (ft3) 39750 Total =
41043 Volume: Full RCS (including Pzr) (ft3) 9591 Page 3 of 3 Version 36.0 11113/09 12:03 :44 Time TABLEB Power-Uprated Unit Time To Saturation: One-Third New Fuel Assumed Initial Temperature = 140°F Time To Time To Time To FNP-0-UOP-4.0 TABLE B Time To After Saturation Saturation Saturation Saturation RCS Shutdown at Midloop 3 ft Below Flange Full Reactor Full (hours)
(min)
(min)
Cavity (hours)
(hours) 100 10.0 13.6 7.2 1.7 200 13.3 18.2 9.6 2.2 300 15.7 21.5 11.4 2.7 400 17.5 24.0 12.7 3.0 500 19.8 27.1 14.3 3.3 600 22.0 30.1 15.9 3.7 700 23.7 32.5 17.2 4.0 800 25.7 35.2 18.6 4.3 VOLUME REFERENCE TABLE Midloop Volume (ft3):
945 Volume: 3 ft Below Flange (ft3) 348 Total =
1293 Volume: Full Reactor Cavity (ft3) 39750 Total =
41043 Volume: Full RCS (including pzr) (ft3) 9591 Page 3 of3 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I FARLEY NUCLEAR PLANT SHARED UNIT OPERATING PROCEDURE UOP-4.0 APPENDIX 1
1.0 Purpose The purpose of the Shutdown Safety Assessment is to provide a means for evaluating the safety condition of the plant when in Modes 4, 5, 6 or de-fueled and to provide appropriate contingency actions. (CR 2004102447) 2.0 Definitions 2.1 Green Condition
- The plant is fully capable of performing the associated safety function.
2.2 Yellow Condition
- The plants ability to perform the associated safety function is reduced but is at an acceptable level.
2.3 Orange Condition
- The plants ability to perform the associated safety function has been severely reduced and steps should be taken to minimize the amount of time in this condition.
2.4 Red Condition
- The plants ability to perform the associated safety function is in jeopardy and steps must be taken immediately to correct the cause of the condition.
2.5 Source Range Instrumentation Available
- The audible count rate and at least one indication of source range counts (preferably NR-45) are available to indicate a dilution of the RCS.
2.6 RCS Intact Below the Vessel Flange
- No opening exist which would result in spillage if the RCS level was raised to the vessel flange.
2.7 One or More SGs Available
- A S/G is available to remove heat if the tubes are filled and vented, the wide range level is> 75%, a steam flow path to the atmosphere can be made available with minor valve manipulations and a means of adding water to the S/G via the auxiliary feed water system exists. The S/G tubes are considered filled and vented if the RCS has been maintained greater than or equal to 100 psig since the last RCS fill and vent.
2.8 RHR Subsystems Available
- The RHR system is capable of removing heat from the RCS.
Page 1 of 18 Version 36.0 11113/09 12:03:44 1.0 Purpose FARLEY NUCLEAR PLANT SHARED UNIT OPERATING PROCEDURE UOP-4.0 APPENDIX 1 FNP-0-UOP-4.0 APPENDIX 1 The purpose of the Shutdown Safety Assessment is to provide a means for evaluating the safety condition of the plant when in Modes 4,5,6 or de-fueled and to provide appropriate contingency actions. (CR 2004102447) 2.0 Definitions 2.1 Green Condition - The plant is fully capable of performing the associated safety function.
2.2 Yellow Condition - The plant's ability to perform the associated safety function is reduced but is at an acceptable level.
2.3 Orange Condition - The plant's ability to perform the associated safety function has been severely reduced and steps should be taken to minimize the amount of time in this condition.
2.4 Red Condition - The plant's ability to perform the associated safety function is in jeopardy and steps must be taken immediately to correct the cause of the condition.
2.5 Source Range Instrumentation Available - The audible count rate and at least one indication of source range counts (preferably NR-45) are available to indicate a dilution of the RCS.
2.6 RCS Intact Below the Vessel Flange - No opening exist which would result in spillage if the RCS level was raised to the vessel flange.
2.7 One or More SGs Available - A S/G is available to remove heat if the tubes are filled and vented, the wide range level is ~ 75%, a steam flow path to the atmosphere can be made available with minor valve manipulations and a means of adding water to the S/G via the auxiliary feed water system exists. The S/G tubes are considered filled and vented if the RCS has been maintained greater than or equal to 100 psig since the last RCS fill and vent.
2.8 RHR Subsystems Available - The RHR system is capable of removing heat from the RCS.
Page I of 18 Version 36.0
11/13/09 12:03 :44 FNP-0-UOP-4.0 APPENDIX I 2.9 One D/G Available
- The selected DIG is capable of being started (either manually or automatically) and supplying 4160 V power to its respective bus.
2.10 Equipment Hatch & Airlocks Capable of Being Closed on Short Notice The equipment hatch meets this definition when the MCRT is established. The personnel and aux airlocks meet this definition when a routine check of the hoses and cables going through the airlocks reveals that at least one door in each airlock can be closed in less than the current time to boil. (Al 2008207932) 2.11 HHSI Pump / Flow Path Available
- A charging pump is capable of injecting water into the RCS from the RWST which contains> 50,000 gallons of water.
Only minor valve manipulations are required.
2.12 LHSI Pump/Flow Path Available
- A RHR pump is capable of injecting water into the RCS from the RWST which contains> 50,000 gallons of water. Only minor valve manipulations are required.
2.13 Based on the unit outage, boration flow paths via charging pumps and ECCS injection lines are determined using the guidance of Table 4 of FNP-1-STP-2.1 OR FNP-2-STP-2.l
, BORON INJECTION FLOW PATH VERIFICATION AND BORIC ACID TRANSFER PUMP OPERABILITY TEST, MODES 5 & 6 AND by a current FNP-1-STP-3.2 OR FNP-2-STP-3.2, BORATED WATER SOURCE OPERABILITY TEST MODE 5,6.
Additionally, Normal charging can be considered an available boration flow path under the following conditions:
2.13.1 QIE2IMOV8IO7 and Q1E2IMOV8IO8 are open, OR Q2E21MOV8IO7 and Q2E21MOV8IO8 are open.
2.13.2 With one of the following charging pump requirements met.
IA or 2A Charging Pump available I B Charging Pump Available with Ql E2 1 MOV8 I 32A and B open, OR 2B Charging Pump Available with Q2E2 I MOV8 I 32A and B open 1C Charging pump available with QIE2IMOV8I32A and B open, OR 2C Charging pump available with Q2E2IMOV8I32A and B open Page 2 of 1 8 Version 36.0 11113/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I 2.9 One DIG Available - The selected DIG is capable of being started (either manually or automatically) and supplying 4160 V power to its respective bus.
2.10 Equipment Hatch & Airlocks Capable of Being Closed on Short Notice - The equipment hatch meets this definition when the MCRT is established. The personnel and aux airlocks meet this definition when a routine check of the hoses and cables going through the airlocks reveals that at least one door in each airlock can be closed in less than the current time to boil. (AI 2008207932) 2.11 HHSI Pump I Flow Path Available - A charging pump is capable of injecting water into the RCS from the RWST which contains> 50,000 gallons of water.
Only minor valve manipulations are required.
2.12 LHSI Pump I Flow Path Available - A RHR pump is capable of injecting water into the RCS from the RWST which contains> 50,000 gallons of water. Only minor valve manipulations are required.
2.13 Based on the unit outage, boration flow paths via charging pumps and ECCS injection lines are determined using the guidance of Table 4 of FNP-I-STP-2.1 OR FNP-2-STP-2.l,BORON INJECTION FLOW PATH VERIFICATION AND BORIC ACID TRANSFER PUMP OPERABILITY TEST, MODES 5 & 6 AND by a current FNP-I-STP-3.2 OR FNP-2-STP-3.2, BORATED WATER SOURCE OPERABILITY TEST MODE 5, 6.
Additionally, Normal charging can be considered an available boration flow path under the following conditions:
2.13.1 QIE21MOV8107 and QIE21MOV8108 are open, OR Q2E21MOV8107 and Q2E21MOV8108 are open.
2.13.2 With one of the following charging pump requirements met.
IA or 2A Charging Pump available IB Charging Pump Available with QIE21MOV8132A and B open, OR 2B Charging Pump Available with Q2E2IMOV8132A and B open IC Charging pump available with QIE21MOV8132A and B open, OR 2C Charging pump available with Q2E21 MOV8132A and B open Page 2 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I 3.0 Precautions and Limitations 3.1 The Shift Supervisor will evaluate the affect of removing shutdown safety equipment from service. If proposed changes to the schedule appear to reflect a reduction in the key safety functions, refer the proposed changes to the Outage Manager for assessment per FNP-0-AP-94, OUTAGE NUCLEAR SAFETY, prior to implementing the changes.
3.2 The appropriate AOP-40 series procedure will be implemented upon unexpected entry into or discovery of the existence of a Red, Orange or Yellow condition.
3.3 Intentional entry into a red condition must be approved by the Plant Manager and documented per FNP-0-AP-94, OUTAGE NUCLEAR SAFETY. FNPs current position is that the plant will not be intentionally placed in this condition.
Review FNPs commitment to NUMARC 9 1-06.
3.4 Intentional entry into an orange condition must be approved by the Outage Manager or his designee prior to entry into that condition and documented per FNP-0-AP-94.
3.5 Intentional entry into a yellow condition must be approved by the Shift Manager or Operations Superintendent prior to entry into that condition.
3.6 Notify the Outage Manager immediately if a red condition or an unexpected orange condition exists.
4.0 Instructions NOTE:
4.2 Fill in the header information on the form.
Unit Prepared By Date Time Page 3 of 18 Version 36.0 11113/09 12:03 :44 3.0 Precautions and Limitations FNP-0-UOP-4.0 APPENDIX 1 3.1 The Shift Supervisor will evaluate the affect of removing shutdown safety equipment from service. If proposed changes to the schedule appear to reflect a reduction in the key safety functions, refer the proposed changes to the Outage Manager for assessment per FNP-0-AP-94, OUTAGE NUCLEAR SAFETY, prior to implementing the changes.
3.2 The appropriate AOP-40 series procedure will be implemented upon unexpected entry into or discovery of the existence of a Red, Orange or Yellow condition.
3.3 Intentional entry into a red condition must be approved by the Plant Manager and documented per FNP-0-AP-94, OUTAGE NUCLEAR SAFETY. FNP's current position is that the plant will not be intentionally placed in this condition.
Review FNP's commitment to NUMARC 91-06.
3.4 Intentional entry into an orange condition must be approved by the Outage Manager or his designee prior to entry into that condition and documented per FNP-0-AP-94.
3.5 Intentional entry into a yellow condition must be approved by the Shift Manager or Operations Superintendent prior to entry into that condition.
3.6 Notify the Outage Manager immediately if a red condition or an unexpected orange condition exists.
4.0 Instructions NOTE:
4.2 Fill in the header information on the form.
Unit Prepared By Date Time Page 3 of 18 Version 36.0
11/13/09 12:03 :44 FNP-0-UOP-4.0 APPENDIX I 4.3 IF in mode 5 or 6, THEN determine the time to saturation for existing plant conditions using the IPC (preferred) or FNP-0-UOP-4.0 Table A or B. Enter time on Figure 1A of Appendix I.
4.4 Evaluate the criteria listed below each Shutdown Safety Function as follows.
NOTE:
Refer to Definitions to assist in the evaluation as appropriate.
I 4.4.1 IF the reactor is defueled, THEN N/A the blanks for all Safety Functions except power availability and spent fuel cooling.
4.4.2 IF the criteria is met, THEN place a I in its blank or use the number/range of numbers listed in ().
4.4.3 IF the criteria is NOT met, THEN place a 0 in its blank.
4.4.4 For each individual Safety Function add up the numbers in the blanks and circle the condition corresponding to the subtotal.
4.5 IF an unexpected Red, Orange or Yellow condition is determined to exist, THEN implement the appropriate AOP-40 series procedure referenced on the Shutdown Safety Assessment form.
4.6 a yellow or orange condition exists that result in a single train available, THEN place a caution sign concerning equipment required for a safe shutdown condition at the applicable locations per the following:
4.6.1 Use Figure 2 to determine posting requirements.
4.6.2 Check the box on Figure IA or lB to indicate the review of the Figure 2 posting requirements {A1 2001203573).
4.7 If a red, orange, or yellow condition currently exists for any Safety Function due to an intentional entry, verify 4.7.1 Appropriate approval per FNP-0-AP-94 has been obtained, and 4.7.2 Contingency Actions per Table I have been performed for the affected Shutdown Safety Functions.
Page 4 of 1 8 Version 36.0 11113/09 12:03:44 FNP-0-UOP-4.0 APPENDIX 1 4.3 IF in mode 5 or 6, THEN determine the time to saturation for existing plant conditions using the IPC (preferred) or FNP-0-UOP-4.0 Table A or B. Enter time on Figure 1 A of Appendix 1.
4.4 Evaluate the criteria listed below each Shutdown Safety Function as follows.
NOTE:
Refer to Definitions to assist in the evaluation as appropriate.
4.4.1 IF the reactor is defueled, THEN N/ A the blanks for all Safety Functions except power availability and spent fuel cooling.
4.4.2 IF the criteria is met, THEN place a 1 in its blank or use the number/range of numbers listed in ( ).
4.4.3 IF the criteria is NOT met, THEN place a 0 in its blank.
4.4.4 For each individual Safety Function add up the numbers in the blanks and circle the condition corresponding to the subtotal.
4.5 IF an unexpected Red, Orange or Yellow condition is determined to exist, THEN implement the appropriate AOP-40 series procedure referenced on the Shutdown Safety Assessment form.
4.6 IF a yellow or orange condition exists that result in a single train available, THEN place a caution sign concerning equipment required for a safe shutdown condition at the applicable locations per the following:
4.6.1 Use Figure 2 to determine posting requirements.
4.6.2 Check the box on Figure 1 A or 1 B to indicate the review of the Figure 2 posting requirements {AI 2001203573}.
4.7 If a red, orange, or yellow condition currently exists for any Safety Function due to an intentional entry, verify 4.7.1 Appropriate approval per FNP-0-AP-94 has been obtained, and 4.7.2 Contingency Actions per Table 1 have been performed for the affected Shutdown Safety Functions.
Page 4 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I 4.8 Evaluate the activities planned for the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to determine if the plant may potentially enter a red, orange, or yellow condition.
4.8.1 jf intentional entry into a red condition is expected THEN entry must be approved by the Nuclear Plant General Manager and documented per FNP-0-AP-94, OUTAGE NUCLEAR SAFETY. FNPs current position is that the plant will not be intentionally placed in this condition. Review FNPs commitment to NUMARC 91-06.
4.8.2 fl intentional entry into an orange condition is expected THEN A. Entry must be approved by the Outage Manager or his designee prior to entry into that condition and documented per FNP-0-AP-94.
B. Perform the Contingency Actions in Table 2 for the affected Shutdown Safety Function(s).
4.8.3 j! intentional entry into a yellow condition is expected THEN A. Entry must be approved by the Shift Manager or Operations Superintendent prior to entry into that condition and documented per FNP-0-AP-94.
B. Perform the Contingency Actions in Table 2 for the affected Shutdown Safety Function(s).
4.9 Post the Shutdown Safety Assessment at the following locations.
Control Room Entrance to the PAP Entrance to the SAP (only required if SAP opened for protected area access)
Outage Control Center 5.0 References 4.1 NUMARC 9 1-06, Guidelines For Industry Actions To Assess Shutdown Management.
4.2 IOCFR5O.65 (a)(4) 4.3 NUMARC 93-01, Section 11, Assessment of Risk Resulting from Performance of Maintenance Activities, February 2, 2000.
4.4 FNP-0-AP-94, OUTAGE NUCLEAR SAFETY.
Page 5 of 18 Version 36.0 11113/09 12:03:44 FNP-0-UOP-4.0 APPENDIX 1 4.8 Evaluate the activities planned for the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to determine if the plant may potentially enter a red, orange, or yellow condition.
4.8.1 IF intentional entry into a red condition is expected THEN entry must be approved by the Nuclear Plant General Manager and documented per FNP-0-AP-94, OUTAGE NUCLEAR SAFETY. FNP's current position is that the plant will not be intentionally placed in this condition. Review FNP's commitment to NUMARC 91-06.
4.8.2 IF intentional entry into an orange condition is expected THEN A. Entry must be approved by the Outage Manager or his designee prior to entry into that condition and documented per FNP-0-AP-94.
B. Perform the Contingency Actions in Table 2 for the affected Shutdown Safety Function(s).
4.8.3 IF intentional entry into a yellow condition is expected THEN A. Entry must be approved by the Shift Manager or Operations Superintendent prior to entry into that condition and documented per FNP-0-AP-94.
B. Perform the Contingency Actions in Table 2 for the affected Shutdown Safety Function(s).
4.9 Post the Shutdown Safety Assessment at the following locations.
Control Room Entrance to the PAP Entrance to the SAP (only required if SAP opened for protected area access)
Outage Control Center 5.0 References 4.1 NUMARC 91-06, Guidelines For Industry Actions To Assess Shutdown Management.
4.2 10CFR50.65 (a)(4) 4.3 NUMARC 93-01, Section 11, Assessment of Risk Resulting from Performance of Maintenance Activities, February 2, 2000.
4.4 FNP-0-AP-94, OUTAGE NUCLEAR SAFETY.
Page 5 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I
Figure IA SHUTDOWN SAFETY ASSESSMENT (Modes 5, 6 and Defueled)
Unit:
Prepared By:_________________ Date:
Time:______
App-I, Fig 2 Evaluated El SHUTDOWN SAFETY FUNCTION! CRITERIA CONDITION (No/False=O, Yes/TrueI, Use number within range when required)
(Circle Condition)
REACTIVITY Subtotal Condition AOP I.
No Core Alterations in Progress 0-1 RED 41 2.
Number of Boration Flow Paths (0, 1, 2) (Ref step 2.13) 2 ORANGE 41 3.
RCS Boron: CSDfRefueling Concentration 34 YELLOW 41 4.
Source Range Instrumentation Available 5
GREEN Reactivity Subtotal (GREEN if Defueled)
CORE COOLING Subtotal Condition AOP 1.
2 SGs Avail with loops filled (Ref step 2.7) 0-1 RED 42 2.
Cavity level 1529 2-3 ORANGE 42 3.
RHR Subsystems Available (0, 1 or 2)
YELLOW 42 4.
RCS level 126 6 GREEN (GREEN if Defueled) 5.
Time to saturation > 30 minutes OR RCS press> 325 psig with at least one RCP available for operation and at least one SG available Core Cooling Subtotal POWER AVAILABILITY Subtotal Condition AOP I.
I A Train DG Available 0-2 RED 43 2.
1 B Train DG Available 3
ORANGE 43 3.
F 4160 V BUS normal (Aligned to A SUT) 45 YELLOW 43 4.
G 4160 V BUS normal (Aligned to B SUT) 6 GREEN 5.
2 Feeds available to the HV Switchyard (0. 1 or 2)
Power Availability Subtotal CONTAINMENT Subtotal Condition AOP 1.
Refueling Integrity Set 0-1 RED 44 2.
CTMT Closure Set 2-4 ORANGE 44 3.
No Core Alterations in Progress (2 pts) 5-6 YELLOW 44 4.
Equipment Hatch & Air Locks Closed or Capable of Being Closed on 7
GREEN 5.
RCSlevel 126 6 (3 pts)
(GREEN if Defucled)
Containment Subtotal INVENTORY Subtotal Condition AOP I.
Refueling Cavity 23 Feet (142l) Above Fuel 0
RED 45 2.
LHSI Pump/Flowpath Available I
ORANGE 45 3.
HHSI Pump/Flowpath Available 2
YELLOW 45 4.
RCS is Intact below the Reactor Vessel Flange 3-4 GREEN (GREEN if Defueled)
Inventory Subtotal RCS INTEGRITY Subtotal Condition AOP I.
All S/G Manways or Nozzle Dams Installed 0-I ORANGE 46 2.
RCS is Intact below the Reactor Vessel Flange 2
YELLOW 46 3.
Pressurizer level < 100%
3 GREEN RCS Integrity Subtotal (GREEN if Defueled)
SPENT FUEL COOLING Subtotal Condition AOP 1.
SFP level 23 feet (1516) above fuel (4 pts) 0-4 RED 47 2.
A Tm SFP Cooling available 5
ORANGE 47 3.
B Tm SFP Cooling available 6
YELLOW 47 4.
2 SFP Makeup Sources (RWST, DW, RMW to Blender, Boric Acid to 7
GREEN Blender, RHT to transfer canal with weir gate removed)
SFP Subtotal Time to saturation Jf core cooling were lost:
nours minutes Page 6 of 18 Version 36.0 11113109 12:03:44 Figure lA FNP-0-UOP-4.0 APPENDIX 1 SHUTDOWN SAFETY ASSESSMENT (Modes 5, 6 and Defueled)
Unit:_ Prepared 8y: ________ Date: ____ Time:___
App-l, Fig 2 Evaluated 0 SHUTDOWN SAFETY FUNCTION/ CRITERIA CONDITION (No/False 0, Yes/True 1, Use number
~,
(Circle Condition)
REACTIVITY Subtotal Condition AOP
No Core Alterations in Progress 0-1 RED 41
Number of Boration Flow Paths (0, I, 2) (Ref step 2.13) 2 ORANGE 41
RCS Boron: CSO/Refueling Concentration 3-4 YELLOW 41
Source Range Instrumentation Available 5
GREEN Reactivity Subtotal (GREEN if Defueled)
CORE COOLING Subtotal Condition AOP
z 2 SGs A vail with loops filled (Ref step 2.7) 0-1 RED 42
Cavity levelz 152'9" 2-3 ORANGE 42
RHR Subsystems Available (0, I or 2) 4 YELLOW 42
RCS levelz 126' 6" 2':5 GREEN
Time to saturation> 30 minutes OR RCS press> 325 psig with at least (GREEN if Defueled) one RCP available for operation and at least one SG available Core Cooling Subtotal POWER AVAILABILITY Subtotal Condition AOP
I "A" Train OG Available 0-2 RED 43
I "B" Train DG Available 3
ORANGE 43
F 4160 V BUS normal (Aligned to A SUT) 4-5 YELLOW 43
G 4160 V BUS normal (Aligned to B SUT) 6 GREEN
2 Feeds available to the HV Switchyard (0, I or 2)
Power Availability Subtotal CO NT AINMENT Subtotal Condition AOP
Refueling Integrity Set 0-1 RED 44
CTMT Closure Set 2-4 ORANGE 44
No Core Alterations in Progress (2 pts) 5-6 YELLOW 44
Equipment Hatch & Air Locks Closed or Capable of Being Closed on 7
GREEN Short Notice (GREEN if Defueled)
RCS levelz 126' 6" (3 pts)
Containment Subtotal INVENTORY Subtotal Condition AOP
Refueling Cavity z 23 Feet (142'1") Above Fuel 0
RED 45
LHSI PumplFlowpath Available 1
ORANGE 45
HHSI Pump/Flowpath Available 2
YELLOW 45
RCS is Intact below the Reactor Vessel Flange 3-4 GREEN (GREEN ifDefueled)
Inventory Subtotal RCS INTEGRITY Subtotal Condition AOP
All S/G Manways or Nozzle Dams Installed 0-1 ORANGE 46
RCS is Intact below the Reactor Vessel Flange 2
YELLOW 46
Pressurizer level < 100%
3 GREEN RCS Integrity Subtotal (GREEN ifDefue1ed)
SPENT FUEL COOLING Subtotal Condition AOP
SFP level z 23 feet (151 '6") above fuel (4 pts) 0-4 RED 47
A Tm SFP Cooling available 5
ORANGE 47
B Tm SFP Cooling available 6
YELLOW 47
GREEN Blender, RHT to transfer canal with weir gate removed)
SFP Subtotal Time to saturation IF core cooling were lost:
hours minutes Page 6 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I FIGURE lB SHUTDOWN SAFETY ASSESSMENT (Mode 4)
Unit:
Prepared By:________________ Date:
Time:______
App-I, Fig 2 Evaluated SHUTDOWN SAFETY FUNCTION! CRITERIA CONDITION (Circle Condition)
(No/FalseO, YeslTrue=1, Use number within range when required)
REACTIVITY Subtotal Condition AOP I.
No positive reactivity changes in Progress 0-1 RED 41 2.
Number of Boration Flow Paths (0, 1. 2) (Ref step 2.13) 2 ORANGE 41 3.
RCS Boron: CSD boron concentration 3-4 YELLOW 41 4.
Number of Source Range Instrumentation Available 5
GREEN Reactivity Subtotal CORE COOLING Subtotal Condition AOP I.
2 SGs Available (2 pts if 2 S/Gs Available (Ref step 2.7) 0-1 RED 42 2.
RHR Subsystems Available (0, I or 2) 2-3 ORANGE 42 3.
PRZ level within normal operating band 4
YELLOW 42 4.
RCS Subcooling Margin > 16° F GREEN Core Cooling Subtotal POWER AVAILABILITY Subtotal Condition AOP 1.
1 A Train DG Available 0-2 RED 43 2.
1 B Train DG Available 3
ORANGE 43 3.
F 4160 V BUS normal (Aligned to A SUT) 4-5 YELLOW 43 4.
G 4160 V BUS normal (Aligned to B SUT) 6 GREEN 5.
2 Feeds available to the HV Switchyard (0. 1 or 2)
Power Availability Subtotal CONTAINMENT Subtotal Condition AOP 1.
CTMT Integrity meets tech spec requirements 0-1 RED 44 2.
RCS borated to cold shutdown boron concentration (2 pts) 2-4 ORANGE 44 3.
CTMT Cooling tech spec are met (3 pts) 5-6 YELLOW 44 Containment Subtotal 6
GREEN INVENTORY Subtotal Condition AOP 1.
PRZ level within normal operating band 0
RED 45 2.
LHSI Pump/Flowpath Available I
ORANGE 45 3.
HHSI Pump/Flowpath Available 2
YELLOW 45 4.
Normal make-up capability available 3-4 GREEN Inventoiy Subtotal RCS INTEGRITY Subtotal Condition AOP I.
LTOP TECH SPEC MET 0-1 ORANGE 46 2.
RCS/PRZ cooldown rate < tech spec allowable 2
YELLOW 46 3.
Pressurizer operable per tech specs 3
GREEN RCS Integrity Subtotal SPENT FUEL COOLING Subtotal Condition I.
SFP level 23 feet (151 6) above fuel (4 pts) 0-4 RED 47 2.
A Tm SFP Cooling 5
ORANGE 47 3.
B Tm SFP Cooling 6
YELLOW 47 4
2 SFP Makeup Sources (RWST, DW, RMW to Blender, Boric 7
GREEN Acid to Blender, RHT to transfer canal with weir gate removed)
SFP Subtotal Page 7 of 18 Version 36.0 1111310912:03:44 FIGURE IB FNP-0-UOP-4.0 APPENDIX 1 SHUTDOWN SAFETY ASSESSMENT (Mode 4)
Unit __ Prepared By: _______ Date: ____ Time:___
App-l, Fig 2 Evaluated 0 SHUTDOWN SAFETY FUNCTION/ CRITERIA CONDITION (Circle Condition)
(No/False=O, Yes/True= 1, Use number within range when required)
REACTIVITY Subtotal Condition AOP
No positive reactivity changes in Progress 0-1 RED 41
Number of Boration Flow Paths (0, 1, 2) (Ref step 2.13) 2 ORANGE 41
RCS Boron: CSO boron concentration 3-4 YELLOW 41
Number of Source Range Instrumentation Available 5
GREEN Reactivity Subtotal CORE COOLING Subtotal Condition AOP
2 SGs Available (2 pts if2 2 S/G's Available (Ref step 2.7) 0-1 RED 42
RHR Subsystems Available (0, 1 or 2) 2-3 ORANGE 42
PRZ level within normal operating band 4
YELLOW 42
RCS Subcooling Margin> 16° F 25 GREEN Core Cooling Subtotal POWER A V AILABILITY Subtotal Condition AOP
1 "A" Train OG Available 0-2 RED 43
1 "B" Train OG Available 3
ORANGE 43
F 4160 V BUS normal (Aligned to A SUT) 4-5 YELLOW 43
G 4160 V BUS normal (Aligned to B SUT) 6 GREEN
2 Feeds available to the HV Switchyard (0, 1 or 2)
Power Availability Subtotal CONTAINMENT Subtotal Condition AOP
CTMT Integrity meets tech spec requirements 0-1 RED 44
RCS borated to cold shutdown boron concentration (2 pts) 2-4 ORANGE 44
CTMT Cooling tech spec are met (3 pts) 5-6 YELLOW 44 Containment Subtotal 6
GREEN INVENTORY Subtotal Condition AOP
PRZ level within normal operating band 0
RED 45
LHSI Pump/Flowpath Available 1
ORANGE 45
HHSI Pump/Flowpath Available 2
YELLOW 45
Normal make-up capability available 3-4 GREEN Inventory Subtotal RCS INTEGRITY Subtotal Condition AOP
LTOP TECH SPEC MET 0-1 ORANGE 46
RCS/PRZ cooldown rate < tech spec allowable 2
YELLOW 46
Pressurizer operable per tech specs 3
GREEN RCS Integrity Subtotal SPENT FUEL COOLING Subtotal Condition
SFP level 2 23 feet (151 '6") above fuel (4 pts) 0-4 RED 47
A Trn SFP Cooling 5
ORANGE 47
B Trn SFP Cooling 6
YELLOW 47 4
22 SFP Makeup Sources (RWST, OW, RMW to Blender, Boric 7
GREEN Acid to Blender, RHT to transfer canal with weir gate removed)
SFP Subtotal Page 7 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I Figure 2 Yellow/Orange Condition Signs Required for Single Train Availability f required by step 4.6, THEN place caution signs at applicable location specified in Table 2 for Unit 1 or Table 3 for Unit 2.
2.
Signs should be worded similar to the following:
TRAIN OUTAGE IN PROGRESS NO WORK IS TO BE DONE ON THE CALL EXT.
3.
Complete Table 2 or 3 to document the review of posting requirements.
3.1 Mark as N/A any components which are not required to be posted for the given yellow or orange condition.
3.2 Post signs at the applicable locations and initial the appropriate block of Table 2 or 3.
3.3 When the yellow/orange condition no longer exists, ensure the signs have been removed and initial the appropriate block of Table 2 or 3.
Page 8 of 18 Version 36.0 11113/0912:03:44 Figure 2 FNP-0-UOP-4.0 APPENDIX 1 Yellow/Orange Condition Signs Required for Single Train Availability
IF required by step 4.6, THEN place caution signs at applicable location specified in Table 2 for Unit 1 or Table 3 for Unit 2.
Signs should be worded similar to the following:
TRAIN OUTAGE IN PROGRESS NO WORK IS TO BE DONE ON THE TRAIN C&LEXT.
Complete Table 2 or 3 to document the review of posting requirements.
3.J Mark as N/ A any components which are not required to be posted for the given yellow or orange condition.
3.2 Post signs at the applicable locations and initial the appropriate block of Table 2 or 3.
3.3 When the yellow/orange condition no longer exists, ensure the signs have been removed and initial the appropriate block of Table 2 or 3.
Page 8 of 18 Version 36.0
11/13/09 12:03 :44 FNP-0-UOP-4.0 APPENDIX I TABLE I Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Reactivity Orange Condition Contingency Actions for Entry 1.
Voluntary entry into a Reactivity Orange condition is not anticipated.
Reactivity Yellow Condition Contingency Actions for Entry 1.
Verify Technical Specification and Technical Requirements Manual requirements are met and will continue to be met for boron concentration, boration flow path, and source range instrumentation.
Page 9 of 18 Version 36.0 11113/09 12:03:44 TABLE 1 FNP-0-UOP-4.0 APPENDIX I Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Reactivity Orange Condition Contingency Actions for Entry
Voluntary entry into a Reactivity Orange condition is not anticipated.
Reactivity Yellow Condition Contingency Actions for Entry
Verify Technical Specification and Technical Requirements Manual requirements are met and will continue to be met for boron concentration, boration flow path, and source range instrumentation.
Page 9 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I TABLE I Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Core Cooling Orange Condition Contingency Actions for Entry NOTE:
Voluntary entry into a core cooling Orange condition is not normally foreseen for evolutions other than Mid Loop Operations or other reduced inventory evolutions (e.g. drain down for reactor vessel head lift). Under certain plant conditions with time to saturation 30 mm., an orange condition may be encountered.
1.
IF Orange condition entry is being made as part of Mid Loop Operations, THEN ensure the requirements of FNP-1-UOP-4.3 QE FNP-2-UOP-4.3, MID LOOP OPERATIONS, have been met, including the Mid-Loop Compensatory measures in Appendix 3 of FNP-l-UOP-4.l OR FNP-2-UOP-4.1, CONTROLLING PROCEDURE FOR REFUELING.
2.
IF orange condition entry is due to certain plant conditions, with time to saturation 30 minutes, THEN ensure the contingency actions specified in the memo for compensatory measures for RCS level at the reactor vessel flange as required by Appendix 3 of FNP-1-UOP-4.1 OR FNP-2-UOP-4.1, CONTROLLING PROCEDURE FOR REFUELING, have been met. These same contingency actions are applicable for non-refueling outages when RCS level is at the reactor vessel flange in these conditions.
Core Cooling Yellow Condition Contingency Actions for Entry 1.
Ensure both trains of RHR are operable if in Mode 6 prior to lowering refueling cavity level to <23 feet (142l) above the fuel.
Page 10 of 18 Version 36.0 11113/09 12:03:44 TABLE 1 FNP-0-UOP-4.0 APPENDIX 1 Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Core Cooling Orange Condition Contingency Actions for Entry NOTE:
Voluntary entry into a core cooling Orange condition is not normally foreseen for evolutions other than Mid Loop Operations or other reduced inventory evolutions (e.g. drain down for reactor vessel head lift). Under certain plant conditions with time to saturation ~ 30 min., an orange condition may be encountered.
IF Orange condition entry is being made as part of Mid Loop Operations, THEN ensure the requirements ofFNP-I-UOP-4.3 OR FNP-2-UOP-4.3, MID LOOP OPERATIONS, have been met, including the Mid-Loop Compensatory measures in Appendix 3 of FNP-I-UOP-4.1 OR FNP-2-UOP-4.1, CONTROLLING PROCEDURE FOR REFUELING.
IF orange condition entry is due to certain plant conditions, with time to saturation
~ 30 minutes, THEN ensure the contingency actions specified in the memo for compensatory measures for RCS level at the reactor vessel flange as required by Appendix 3 of FNP-I-UOP-4. 1 OR FNP-2-UOP-4.1, CONTROLLING PROCEDURE FOR REFUELING, have been met. These same contingency actions are applicable for non-refueling outages when RCS level is at the reactor vessel flange in these conditions.
Core Cooling Yellow Condition Contingency Actions for Entry
Ensure both trains of RHR are operable if in Mode 6 prior to lowering refueling cavity level to < 23 feet (142' 1") above the fuel.
Page 1 0 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I TABLE I Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Power Availability Orange Condition Contingency Actions For Entry 1.
Verify RCS level 126 6 prior to voluntary Orange condition entry.
2.
Refer to Technical Specification 3.8.2.
3.
Based on the unit outage, as a minimum, ensure the Unit 1 or Unit 2 F ORG 4160V Bus is maintained with:
An offsite power feed through its associated startup transformer.
Its associated DG(s) operable An operable charging pump in the boration flow path An operable BATP if a BATP is required as part of the boration flow path An operable RWST TO CHG PUMP SUCT MOV if the RWST is part of the boration flow path An operable RHR, CCW, and SW Pump 4.
iF the head is on the reactor vessel, THEN at least one PRZR PORV is operable with backup nitrogen available for use in feed and spill cooling in the event that RHR is lost.
Notify the Shift Manager, Alabama Control Center (ACC), or outside SO as appropriate to ensure that no work is in progress that could adversely affect the operable train of AC power and to ensure that no such work is started while this Orange condition exists.
5.
Ensure that no work is in progress in the low voltage switchyard that could adversely affect the operable train of AC power.
6.
Notify EM supervisory personnel of AC power conditions and requirements and direct their personnel (including appropriate Williams and EFS personnel) be briefed on these conditions and requirements.
7.
Install Caution Tags or place barricades as appropriate to identify components such as ESF bus feeder breaker, DGs, DG output breakers and bus tie breakers, and startup transformer associated with the ESF bus to be maintained operable.
Page 11 of 1 8 Version 36.0 11113/0912:03:44 TABLE 1 FNP-0-UOP-4.0 APPENDIX 1 Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Power Availability Orange Condition Contingency Actions For Entry
Verify RCS level ~ 126' 6" prior to voluntary Orange condition entry.
Refer to Technical Specification 3.8.2.
Based on the unit outage, as a minimum, ensure the Unit I or Unit 2 F OR G 4160V Bus is maintained with:
IF the head is on the reactor vessel, THEN at least one PRZR PORV is operable with backup nitrogen available for use in feed and spill cooling in the event that RHR is lost.
Notify the Shift Manager, Alabama Control Center (ACC), or outside SO as appropriate to ensure that no work is in progress that could adversely affect the operable train of AC power and to ensure that no such work is started while this Orange condition exists.
Ensure that no work is in progress in the low voltage switchyard that could adversely affect the operable train of AC power.
Notify EM supervisory personnel of AC power conditions and requirements and direct their personnel (including appropriate Williams and EFS personnel) be briefed on these conditions and requirements.
Install Caution Tags or place barricades as appropriate to identify components such as ESF bus feeder breaker, DGs, DG output breakers and bus tie breakers, and startup transformer associated with the ESF bus to be maintained operable.
Page 11 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I TABLE I Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Power Availability Yellow Condition Contingency Actions For Entry I.
Verify RCS level 126 6 prior to voluntary Yellow condition entry.
2.
Refer to Technical Specification 3.8.2.
3.
Based on the unit outage, as a minimum, ensure the Unit I or Unit 2 F OR G 41 60V Bus is maintained with:
An offsite power feed through its associated startup transformer.
Its associated DG(s) operable An operable charging pump in the boration flow path An operable BATP if a BATP is required as part of the boration flow path An operable RWST TO CHG PUMP SUCT MOV if the RWST is part of the boration flow path An operable RHR, CCW, and SW Pump 4.
IF the head is on the reactor vessel, THEN at least one PRZR PORV is operable with backup nitrogen available for use in feed and spill cooling in the event that RHR is lost.
5.
Ensure that no work is in progress in the low voltage switchyard that could adversely affect the operable train of AC power.
6.
Notify EM supervisory personnel of AC power conditions and requirements and direct their personnel (including appropriate Williams and EFS personnel) be briefed on these conditions and requirements.
Page 12 of 1 8 Version 36.0 11113/09 12:03 :44 TABLE 1 FNP-0-UOP-4.0 APPENDIX 1 Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Power Availability Yellow Condition Contingency Actions For Entry
Verify RCS level;::: 126' 6" prior to voluntary Yellow condition entry.
Refer to Technical Specification 3.8.2.
Based on the unit outage, as a minimum, ensure the Unit 1 or Unit 2 F OR G 4160V Bus is maintained with:
An offsite power feed through its associated startup transformer.
Its associated DG(s) operable An operable charging pump in the boration flow path An operable BA TP if a BA TP is required as part of the boration flow path An operable RWST TO CHG PUMP SUCT MOV if the RWST is part of the boration flow path An operable RHR, CCW, and SW Pump
IF the head is on the reactor vessel, THEN at least one PRZR PORV is operable with backup nitrogen available for use in feed and spill cooling in the event that RHR is lost.
Ensure that no work is in progress in the low voltage switchyard that could adversely affect the operable train of AC power.
Notify EM supervisory personnel of AC power conditions and requirements and direct their personnel (including appropriate Williams and EFS personnel) be briefed on these conditions and requirements.
Page 12 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I TABLE I Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Containment Orange Condition Contingency Actions For Entry 1.
Verify no core alterations in progress or planned.
2.
IF Orange condition entry is being made as part of Mid Loop Operations, THEN base on the unit outage ensure the requirements of FNP-1-UOP-4.3 Q FNP-2-UOP-4.3, MID LOOP OPERATIONS, have been met.
Containment Yellow Condition Contingency Actions For Entry I.
Not applicable.
Page 13 of 18 Version 36.0 11113/0912:03:44 TABLE 1 FNP-0-UOP-4.0 APPENDIX 1 Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Containment Orange Condition Contingency Actions For Entry
Verify no core alterations in progress or planned.
IF Orange condition entry is being made as part of Mid Loop Operations, THEN base on the unit outage ensure the requirements ofFNP-I-UOP-4.3 OR FNP-2-UOP-4.3, MID LOOP OPERATIONS, have been met.
Containment Yellow Condition Contingency Actions For Entry
Not applicable.
Page 13 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX 1
TABLE 1 Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Inventory Orange Condition Contingency Actions For Entry 1.
Voluntary entry into an Inventory Orange condition is not anticipated.
Inventory Yellow Condition Contingency Actions For Entry 1.
Ensure both trains of RHR are operable before reducing refueling cavity level to <23 feet above fuel.
2.
If entry is due to Mid Loop Operations, THEN based on the unit outage, ensure the requirements of FNP-1-UOP-4.3 OR FNP-2-UOP-4.3, have been met.
3.
Refer to applicable Technical Specifications before removing RHR or HHSI systems from service.
Page 14 of 18 Version 36.0 11113/09 12:03:44 TABLE 1 FNP-0-UOP-4.0 APPENDIX 1 Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Inventory Orange Condition Contingency Actions For Entry
Voluntary entry into an Inventory Orange condition is not anticipated.
Inventory Yellow Condition Contingency Actions For Entry I.
Ensure both trains ofRHR are operable before reducing refueling cavity level to < 23 feet above fuel.
IF entry is due to Mid Loop Operations, THEN based on the unit outage, ensure the requirements of FNP-I-UOP-4.3 OR FNP-2-UOP-4.3, have been met.
Refer to applicable Technical Specifications before removing RHR or HHSI systems from service.
Page 14 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX 1
TABLE 1 Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
RCS integrity Orange Condition Contingency Actions For Entry 1.
IF entry is due to Mid Loop Operations, THEN based on the unit outage ensure the requirements of FNP-1-UOP-4.3 OR FNP-2UOP-4.3, MID LOOP OPERATIONS, have been met.
2.
WHEN SG cold leg manways are to be opened OR cold leg nozzle dams are to be removed, THEN verify that at least one hot leg manway, diaphragm, and nozzle dam has been previously removed.
RCS integrity Yellow Condition Contingency Actions For Entry I.
11 entry is due to Mid Loop Operations, THEN based on the unit outage ensure the requirements of FNP-1-UOP-4.3 OR FNP-2-UOP-4.3, MID LOOP OPERATIONS, have been met.
2.
WHEN SG cold leg manways are to be opened Q cold leg nozzle dams are to be removed, THEN verify that at least one hot leg manway, diaphragm, and nozzle dam has been previously removed.
Page 15 of 18 Version 36.0 11/13/09 12:03 :44 TABLE 1 FNP-0-UOP-4.0 APPENDIX 1 Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
RCS Integrity Orange Condition Contingency Actions For Entry
IF entry is due to Mid Loop Operations, THEN based on the unit outage ensure the requirements of FNP-I-UOP-4.3 OR FNP-2-UOP-4.3, MID LOOP OPERATIONS, have been met.
WHEN SG cold leg manways are to be opened OR cold leg nozzle dams are to be removed, THEN verify that at least one hot leg manway, diaphragm, and nozzle dam has been previously removed.
RCS Integrity Yellow Condition Contingency Actions For Entry
IF entry is due to Mid Loop Operations, THEN based on the unit outage ensure the requirements ofFNP-I-UOP-4.3 OR FNP-2-UOP-4.3, MID LOOP OPERATIONS, have been met.
WHEN SG cold leg manways are to be opened OR cold leg nozzle dams are to be removed, THEN verify that at least one hot leg manway, diaphragm, and nozzle dam has been previously removed.
Page 15 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I TABLE I Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Spent Fuel Cooling Orange Condition Contingency Actions for Entry 1.
Voluntary entry into an Orange condition is not anticipated.
Spent Fuel Cooling Yellow Condition Contingency Actions for Entry I.
Verify that at least one train of SFP cooling and at least one makeup source are maintained operable.
2.
Ensure that operating train SFP HX outlet temperature is being monitored every four hours on Rad Side SO logs.
Page 16 of 18 Version 36.0 11113/0912:03:44 TABLE 1 FNP-0-UOP-4.0 APPENDIX 1 Contingency Actions Required for Voluntary Entry to a Shutdown Safety Assessment Orange or Yellow condition.
Spent Fuel Cooling Orange Condition Contingency Actions for Entry
Voluntary entry into an Orange condition is not anticipated.
Spent Fuel Cooling Yellow Condition Contingency Actions for Entry
Verify that at least one train of SFP cooling and at least one makeup source are maintained operable.
Ensure that operating train SFP HX outlet temperature is being monitored every four hours on Rad Side SO logs.
Page 16 of 18 Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I TABLE 2 Protected Train Postings Unit I Components Sign Posted Sign Removed Initials Initials Charging Pump I A (B,C)
BATP 1A (B)
Diesel Generator 1-2A (IC, 2C, IB, 2B)
SFP Cooling IA TRN (B TRN) 4160 V Buses F and K (Buses G and L)
Aux Bldg DC Swgr, A TRN (B TRN)
Aux Bldg Battery, A TRN (B TRN)
SW Valve Boxes Start Up Transformers I A, or I B Other Page 17 of 18 Version 36.0 11113/0912:03:44 TABLE 2 Protected Train Postings Unit 1 Components Sign Posted Initials Charging Pump lA (B,C)
RHR Pump lA (B)
CCW Pump 1A (B,C)
BATP 1A(B}
Diesel Generator 1-2A (1 C, 2C, 1 B, 2B)
SFP Cooling lA TRN (B TRN) 4160 V Buses F and K (Buses G and L)
Aux Bldg DC Swgr, A TRN (B TRN)
Aux Bldg Battery, A TRN (B TRN)
SW Pumps 1A (B, C, D, E)
SW Valve Boxes Start Up Transformers 1 A, or 1 B Other Page 17 of 18 FNP-0-UOP-4.0 APPENDIX 1 Sign Removed Initials Version 36.0
11/13/09 12:03:44 FNP-0-UOP-4.0 APPENDIX I TABLE 3 Protected Train Postings Unit 2 Components Sign Posted Sign Removed Initials Initials Charging Pump 2 (B,C)
RFIR Pump 2 (B)
CCW Pump 2 (B,C)
BATP 2 (B)
Diesel Generator 1-2A (IC, 2C, or 2B)
SFP Cooling A TRN (B TRN) 4160 V Buses F and K (Buses G and L)
Aux Bldg DC Swgr, A TRN (B TRN)
Aux Bldg Battery, A TRN (B TRN)
SW Pumps 2A (B, C, D, E)
SW Valve Boxes Start Up Transformers 2A or 2B Other Page 18of18 Version 36.0 11113/09 12:03:44 TABLE 3 Protected Train Postings Unit 2 Components Sign Posted Initials Charging Pump 2 (B,C)
RHR Pump 2 (B)
CCW Pump 2 (B,C)
BATP 2 (B)
Diesel Generator 1-2A (1 C, 2C, or 2B )
SFP Cooling A TRN (B TRN) 4160 V Buses F and K (Buses G and L)
Aux Bldg DC Swgr, A TRN (B TRN)
Aux Bldg Battery, A TRN (B TRN)
SW Pumps 2A (B, C, D, E)
SW Valve Boxes Start Up Transformers 2A or 2B Other Page 18 of 18 FNP-0-UOP-4.0 APPENDIX 1 Sign Removed Initials Version 36.0
FNP HLT-33 ADMIN Page 1 of 7 A.2.1.A Equipment Control ADMIN 001A4.11-RO & SRO JPM DIRECTIONS:
1.
Initiation of task may be in group setting; evaluation performed individually upon completion of the task by reviewing the completed form.
2.
Provide the examinee with the required materials to perform this JPM.
TASK STANDARD: Required for successful completion of this JPM:
Determine if Adequate Shutdown Margin exists while in Mode 1, using STP-29.5.
EXAMINER:
Developer H. Fitzwater I 10/28/09 I
NRC Approval HLT33-A.2. 1.A TITLE: Perform A Shutdown Margin Calculation in modes I & 2.
PROGRAM APPLICABLE: SOT SOCT OLT X
LOCT___
ACCEPTABLE EVALUATION METHOD:
X PERFORM SIMULATE DISCUSS EVALUATION LOCATION:
SIMULATOR CONTROL ROOM X
CLASSROOM PROJECTED TIME:
30 MIN SIMULATOR IC NUMBER:
N/A ALTERNATE PATH TIME CRITICAL PRA____
Examinee:
Overall JPM Performance:
Satisfactory Unsatisfactory D
Evaluator Comments (attach additional sheets if necessary)
I SEE NUREG 1021 FORM ES-301-3 FNP HL T-33 ADMIN Page 1 of 7 A.2.1.A Equipment Control ADMIN -OOlA4.11-RO & SRO HL T33-A.2.1.A TITLE: Perfonn A Shutdown Margin Calculation in modes 1 & 2.
PROGRAM APPLICABLE: SOT SOCT OLT~ LOCT __
ACCEPTABLE EVALUATION METHOD: ~
PERFORM SIMULATE DISCUSS EV ALUA TION LOCATION:
SIMULA TOR CONTROL ROOM --.lL CLASSROOM PROJECTED TIME:
30 MIN SIMULATOR IC NUMBER:
N/A AL TERNA TE PATH TIME CRITICAL PRA __
JPM DIRECTIONS:
I. Initiation of task may be in group setting; evaluation perfonned individually upon completion of the task by reviewing the completed fonn.
TASK STANDARD: Required for successful completion of this JPM:
Detennine if Adequate Shutdown Margin exists while in Mode 1, using STP-29.5.
Examinee:
Overall JPM Performance:
Satisfactory 0
Unsatisfactory 0 Evaluator Comments (attach additional sheets if necessary)
EXAMINER: _________ _
H. Fitzwater 10/28/09 SEE NUREG 1021 FORM ES-301-3
FNP HLT-33 ADMIN A.2.1.A Page 2 of 7 CONDITIONS When I tell you to begin, you are to Determine if Shutdown Margin is adequate using STP-29.5, SHUTDOWN MARGIN CALCULATION IN MODES 1 AND 2 (TAVG 547°F), for Unit 1. The conditions under which this task is to be performed are:
a.
Unit I is stable at 90% with the ramp on hold b.
Bank D indicates 192 by Group Demand.
c.
Seven of the Bank D rods (H2, B8, H14, F6, FlO, K10, K6) are at 192 steps by DRPI.
d.
Rod P8, in the D bank, has been determined to be stuck.
e.
Rod P8 is at 162 steps by DRPI.
f.
All other rods are at 229 steps.
g.
Core burnup is 9,800 MWD/MTU burnup.
h.
FNP-1-STP-29.5, SHUTDOWN MARGIN CALCULATION IN MODES I AND 2 (TAVG 547°F), initial conditions are satisfied.
i.
The Shift Supervisor has directed you to complete FNP-1-STP-29.5 starting at step 5.1.
INITIATING CUE:
You may begin.
EVALUATION CHECKLIST RESULTS:
ELEMENTS:
STANDARDS:
(CIRCLE)
START TIME NOTE:
Elements I through 13 are evaluated by comparing the completed Calculation page to the values and ranges of each element. The acceptable responses are ALSO listed on the calculation KEY.
I.
Step A.1: Document Core Burnup I) Value entered: 9,800 MWD S /
U 2.
Step A.2: Document Power Level
All other rods are at 229 steps.
547°F), initial conditions are satisfied.
The Shift Supervisor has directed you to complete FNP-I-STP-29.5 starting at step 5.1.
INITIATING CUE:
"You may begin."
EVALUATION CHECKLIST ELEMENTS:
STANDARDS:
START TIME RESULTS:
(CIRCLE)
NOTE: Elements 1 through 13 are evaluated by comparing the completed Calculation page to the values and ran es of each element. The acce table res onses are ALSO listed on the calculation KEY.
FNP HLT-33 ADMIN EVALUATION CHECKLIST ELEMENTS:
A.2.1.A STANDARDS:
Page 3 of 7 RESULTS:
(CIRCLE) 3.
Step A.3: Determine penalty steps for Banks below RIL.
a)
Using COLR figure 1 determines RIL for 90% power level.
b)
Gathers data from Conditions page c)
Determines penalty steps per bank.
4.
Step A.4: Determine number of penalty steps for RODS below RIL a)
Determines stuck rod in CBD, Determines CBD R1L.
b) Documents given Data for rod P-8 c)
Calculates difference 5.
Step B.1: Determine Rod Worth for all control and shutdown bands at Zero steps using given power and burnup:
a)
Uses Curve 77 pg 1 for 90% power and 1 0000MWD 6.
Step B.2: Calculate penalty value of rod banks 7.
Step B.3: Calculate penalty value of individual
RNG: NONE)
(a). Determines RIL from COLR CBD/RIL=167(165-170) from COLR.
CBC/RIL=FULLOUT (or 229).
(b). Documents given Demand positions CB D/ Demand 192.
All others: Full out (or 229)
(c). Calculates difference All values: 0 (RNG: NONE)
(a). Determines RIL from COLR P8-CBD/RlL=167 (Range: 165-170) from COLR.
(b). Documents given data I. Rod#-P8@162 (c). Calculates difference (RIL DRPI) 5 (RNG: 3-8)
(range: NONE)
S/U S/U s/U S/U SI U
rods FNP HL T-33 ADMIN EVALUATION CHECKLIST A.2.1.A ELEMENTS:
a) Using COLR figure 1 determines RIL for 90% power level.
b) Gathers data from Conditions page c)
Determines penalty steps per bank.
a) Determines stuck rod in CBD, Determines CBD RIL.
b) Documents given Data for rod P-8 c) Calculates difference Step B.1; Determine Rod Worth for all control and shutdown bands at Zero steps using given power and bumup:
a) Uses Curve 77 pg 1 for 90% power and 10000MWD Step B.2: Calculate penalty value of rod banks Step B.3: Calculate penalty value of individual rods Page 3 of 7 STANDARDS:
(RNG: NONE)
(a). Determines RIL from COLR CB Df RIL = 167 (165-170) from COLR.
CBC f RIL = FULL OUT (or 229).
(b). Documents given Demand positions CB Df Demand 192.
All others; Full out (or 229)
(c). Calculates difference All values: 0 (RNG; NONE)
RESULTS:
(CIRCLE)
S f U
(a). Determines RIL from COLR P8 -7 CB Df RIL = 167 (Range; 165-170) from COLR.
(b). Documents given data
( c). Calculates difference (RIL - DRPI) = 5 (RNG; 3-8)
(range; NONE)
S f U S f U S f U
FNP HLT-33 ADMIN A.2.1.A Page 4 of 7 EVALUATION CHECKLIST RESULTS:
ELEMENTS:
STANDARDS:
(CIRCLE) 8.
Step B.4.a; Determine most reactive Rod
U worth (range: NONE)
(a).
From Curve 77 pg 2 for I 0000MWD.
9.
Step B.4.b; Calculate rod worth of 9)
Value entered of 3515 S I U
stuck/untrippable rod (RNG: 3515-3516)
(a).
B.4.a: 1406.
(+1 / -0) value for various rounding technique effects)
S /
U
(-) 2432.7 NOTE: NEG (-) SIGN IS IMPORTANT FOR THIS CALCULA TION/Evaluation (RNG: (-)2403 to (-) 2451)
Transpose data from other steps:
(a).
B.I:6268 (b).
B.2: 0 (c).
B.3: 50 (d).
B.4.b 3515.
Calculate using equation.
S I U
conditions 1634 pcm (RNG: NONE)
(a).
From Curve 78 for <
1 0000MWD.
FNP HLT-33 ADMIN EVALUATION CHECKLIST ELEMENTS:
A.2.1.A
(a). From Curve 77 pg 2 for ::::
10000MWD.
Value entered of3515 (RNG: 3515-3516)
(a). # of stuck: 1 rod (b). B.4.a: 1406.
(+ 1 I -0) value for various rounding technique effects)
(-) 2432.7 NOTE: NEG (-) SIGN IS IMPORTANT FOR THIS CALCULA TIONlEvaluation (RNG: (-)2403 to (-) 2451 )
Transpose data from other steps:
(a). B.l: 6268 (b). B.2: 0 (c). B.3: 50 (d). B.4.b 3515.
Calculate using equation.
1634 pcm (RNG: NONE)
(a). From Curve 78 for::::
10000MWD.
RESULTS:
(CIRCLE)
S I U S I U S I U S I U
FNP HLT-33 ADMIN A.2.1.A Page 5 of 7 EVALUATION CHECKLIST RESULTS:
ELEMENTS:
STANDARDS:
(CIRCLE)
Value entered of:
S /
U Reactivity
(-) 748.7 (RNG: -718 to -767)
NOTE: NEG (-) SIGNIS IMPORTANT FOR THIS CALCULA TION/Evaluation (a).
Transpose data:
Value entered of:
S /
U
(+) 1021.3 (RNG: 1003 to 1051)
NOTE: POS (+) SIGN IS IMPOR TANT FOR THIS CALCULA TION/Evaluation (a).
Transpose data:
CUE: Is the SDM adequate?
U Borate need to Emergency Borate.
OR OR Notifies SS that SDM is NOT adequate (does Reports SDM is NOT not meet acceptance criteria),
adequate.
S /
U a)
Signs and dates Calculation Sheet (a). Signs and dates form b) Initials step 5.1 (b).lnitials step 5.1 of procedure.
c)
Asks for verification of calculation LIMINATE After calculation of Excess Shutdown Margin AND element 14 completed.
STOP TIME FNP HL T-33 ADMIN EVALUATION CHECKLIST ELEMENTS:
A.2.1.A
(-) 748.7 (RNG: -718 to -767)
NOTE: NEG (-) SIGN IS IMPORTANT FOR THIS CALCULATIONlEvaluation (a). Transpose data:
(CIRCLE)
S / U
S / U
(+) 1021.3
( RNG: 1003 to 1051)
NOTE: POS (+) SIGN IS IMPORTANT FOR THIS CALCULA TIONlEvaluation (a). Transpose data:
CUE: "Is the SDM adequate?"
a) Signs and dates Calculation Sheet b) Initials step 5.1 c) Asks for verification of calculation
OR Reports SDM is NOT adequate.
S / U (a). Signs and dates form (b ).Initials step 5.1 of procedure.
I TERMINATE After calculation of Excess Shutdown Margin AND element 14 completed.
STOP TIME
FNP HLT-33 ADMIN A.2.1.A Page 6 of 7 CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.
GENERAL
REFERENCES:
I. FNP-I-STP-29.5, VER4.O
1.
FNP-I-STP-29.5, ver4.O 2.
PCB-VOLI-CRV77, Cycle 23 rev 8 3.
PCB-VOLI-CRV78, Cycle 23 rev 8 4.
Unit 1, COLR for FNP Unit I Cycle 23, Rev 0 Figure 1.
5.
Pen/Pencil 6.
calculator Critical ELEMENT justification:
STEP Evaluation 4-13 CRITICAL - Calculation completion; each step is critical in accurately determining available SDM and Excess SDM.
15 CRITICAL
- Task Objective; Identification that Acceptance Criteria NOT met and/or Emergency boration IS required to ensure corrective action is initiated and compliance with T.S. is restored.
COMMENTS:
FNP HLT-33 ADMIN A.2.1.A Page 6 of 7 CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.
GENERAL
REFERENCES:
STEP 4-13 15 Evaluation CRITICAL - Calculation completion; each step is critical in accurately determining available SDM and Excess SDM.
CRITICAL - Task Objective; Identification that Acceptance Criteria NOT met and/or Emergency boration IS required to ensure corrective action is initiated and compliance with T.S. is restored.
COMMENTS:
KEY A.
PRESENT CONDITIONS A.1 Core Burnup 19800 IMWD/MTU A.2 Power Level.
190 I
°° A.3 Determine number of penalty steps for ROD BANKS below RIL.
(Use COLR or convert the RIL computer reading from % to steps)
A.3.a Record data in table ROD BANK RIL BANK STEPS BELOW BANKS HEIGHT DEMAND RIL
CBA FULL OUT Full Out (229) 0 CBB 11l1 I flTTT Full Out (229) 0 CBC jFull Out(229)r Full Out (229) 0 CBD j 167 (166-170) 192 0
SDA FULL OUT Full Out (229) 0 SDB FULL OUT Full Out (229) 0 TOTAL STEPS BELOW RIL LI I.
RANGE for CBD RIL:
LO: Readability of COLR RIL curve may allow allow as low as 165 steps; although this is <than the curve limit of 166.9, if calculated. The assumptions within the curve development provides for this readability error.
HI: conservatism allowed by P&L 4.1 may be instituted and a value of up to 170 steps may be selected since the specific value between the Y-axis increments may not be calculated.
Page 1 of 3 Revision 4 A.
PRESENT CONDITIONS A.I Core Bumup....
19_80_0_.... IMWD/MTU A.2 Power Level. 190 I %
A.3 Determine number of penalty steps for ROD BANKS below RIL.
(Use COLR or convert the RIL computer reading from % to steps)
A.3.a Record data in table ROD BANKRIL BANK STEPS BELOW BANKS HEIGHT DEMAND RIL,....-
CBA FULL OUT Full Out (229) 0 CBB RT IT T nT TT Full Out (229) 0 CBC Full Out(229) I-- Full Out (229) 0 CBD 167 (166-170) I--
192 0
SDA FULL UUT Full Out (229) 0 SDB FULL OUT Full Out (229) 0 TOTAL STEPS BELOW RIL
I :
I... ___ _
---..I RANGE for CBD RIL:
LO: Readability of COLR RIL curve may allow allow as low as 165 steps; although this is < than the curve limit of 166.9, if calculated. The assumptions within the curve development provides for this readability error.
HI: conservatism allowed byP&L4.1 maybe instituted and a value of up to 170 steps may be selected since the specific value between the Y-axis increments may not be calculated.
Page I of3 Revision 4
A.4 KEY uetermine riuniber 01 penaity steps ror 1ULTh oeiow iciL wnicn were not countea in A.i (Rods below RIL on DRPI when the BANK demand is above RIL)
(Use COLR or convert the RIL computer reading from % to steps)
A.4.a Record data in table ROD NUMBER I RIL POSITION I DRPI POSITION I DIFFERENCE P8 167 (+3,-2) 162 5 (3-8)
TOTAL STEPS BELOW RIL 15 L_1ANGE3-8) I.
B.
SHUTDOWN MARGIN NOTE: Write the values from Curves 77 & 78 directly into the surveillance test.
The correct sign convention has been entered in the STP.
B. I Rod worth for all control and shutdown banks at zero steps at Present Burnup (A. 1) and Power Level (A.2).
B.2 Penalty value of rod banks below insertion limit:
[o steps x 75 pcm/step
=
(A.3.a) 16268 I
I ()
pcm i
(Curve 77 Page 1 L
B.3 Penalty value of individual rods below RIL which were not counted in A.3 (1?rdc li1-vij1 RIL on DRPI when the BANK demand is above RIL) 5 I
r (RANGE 3-8)__
1 s x 10 pcm/step
=
(+)j 50 cm (A.4.a)
138) i Page 2 of 3 Revision 4 A.4.a Record data in table ROD NUMBER RIL POSITION DRPI POSITION DIFFERENCE
~
P8 167 (+3,-2) 162 5 (3-8)
TOTAL STEPS BELOW RIL I
5 I
(RANGE 3-8)
L.._
B.
SHUTDOWN MARGIN NOTE: Write the values from Curves 77 & 78 directly into the surveillance test.
The correct sign convention has been entered in the STP.
B.l Rod worth for all control and shutdown banks at zero steps at Present Burnup (A.l) and Power Level (A.2).
(Curve 77 Page 1) I B.2 Penalty value of rod banks below insertion limit:
o steps x 75 pcm/step
=
(A.3.a)
L __________ _
B.3 Penalty value of individual rods below RIL which were not counted in A.3 r--~;w.£...QIi:,~ RIL on DRPI when the BANK demand is above RIL) 5 (RANGE 3-8) s x 10 pcm/step =
(A.4.a) r------------
I (+) 50 em:
I _ _
(RNG 30-80) ___ J
~
Page 2 of3 Revision 4
KEY I
B.4 Stuck / Untrippable Rod penalty:
(B.4.a is N/A and B.4.b is zero if there are no stuck/untrippable rods)
B.4.a Worth of most reactive rod worth at present burnup (A.l)
(_)j1406 ipcm from Curve 77 pg. 2 B.4.b Calculate worth of Stuckluntrippable rods 1
1406 I
1406 3515
[
x()
k 1.75] + [()
xO.75] =
ran e 3515 to 3516
( Stuck!
(B.4.a)
(B.4.a) untripPj 4
oosto 246 ii
+
[ -1054.5 to -1055]
I
[35150R3516J B.5 Penalized rdu WUI LII UH.IU11H LUIJUHL1LppauI IUU, misaligned rods, rods below the insertion limit, and uncertainty:
50 3515 I
[(_)
I + El
+
(30-80)
(+1,..Oj x 0.9 l6268+80+3515]*0.9 = 2405.7 (B.4.b)
[-6268+80+3517] *09 = 2403.9 (Round down to 2403)*LO
[_6268+30+3515]*0.9 = 2450.7 (Round up to 2451) (hi) l-6268+30+35 171 *0.9= 2448.9 nd Burnup (A. 1)
B.8 Available Shutdown Reactivity:
( )[24327 [_ (_)
11634 I
+
(B.5)
(B.6) 50 (B.7)
(-) 748.7 Range:
-719 to -7670 pcnl I B.9 Shutdown Margin in excess of required Shutdown Margin (B.8 Required SDM from the COLR):
1-748.7 pcm ()
1770 pcm (B.8)
Required SDM from the COLR i(+/-)102131:
Range: 1003 m
tolOSi I
L...________
B. 10 If B.9 is positive, THEN emergency borate per FNP-l-AOP-27.0, EMERGENCY Emergency Boration BORATION, to establish the required shutdown margin otherwise Shutdown IS required Margin is adequate for the Present plant condition.
May also state that a DATE-- EIenent 15-if f
d Signed-- element 15-if assumes/states Rx trip may he assumes/states Per orme by. ____Emerg Boration completed
(Curve 78).
I
(-) 2432.7 range: I
( ) -2403 to-2451 L_
1634
()
pcm B.7 Void Collapse Defect.
(+)
50 pcm Page 3 of 3 Revision 4 B.4 Stuck I Untrippable Rod penalty:
B.5 (B.4.a is N/A and B.4.b is zero if there are no stuck/untrippable rods)
B.4.a Worth of most reactive rod worth at present burnup (A.1) (-)J 1406 Ipcm from Curve 77 pg. 2 B.4.b Calculate worth of Stuckluntrippable rods
[ EJ x (-)~
1.75] + [(_)1 1406 x 0 75] = (_) 3S1S range: 3S1S to 3S16
(# Stucki (B.4.a)
(B.4.a) untripp r,:,24QO~5)tq~~~1]..*..........
[ -1054.5to -1055] =:
r--"'~~-.......Lrl,;;;(--;;J,._).!:16=2:::::68:::::::::::::!.I..:.+~lo:;:::!.1 ~+..L,~~~0~-~80.;...) ~(_) ~~~~O) x 0.9 1
(-) 2432.7 range:
H268+$(}+3~15)"'O.97'~10~;~:I;)\\<}.>>)ri>r~(;;).(;
(B.4. b)
£-(j26.8+8.0+35)7J.~().~.*724~3.~*tJt~W!~qo~p~~40~1*:LO H268+3(}+3515]"'(}19.;,;,* *. 2450~!@()uti(f:UP)tO~~451)*(hi)
L.[_-6_26_8_+_30_+_.35.;..1_7_l*_O...
,9.;..b_*.~
..... *4
...... 8.... 9_*"................................
H.;..i~i...
)........... --Jlnd Burnup (A. 1 )
(Curve 78).
B.7 Void Collapse Defect.
B.8 Available Shutdown Reactivity:
()1-2432.7 L - (_)---1.1_
163_4 _..1- +
50 (B.5)
(B.6)
(B.7)
B.9 Shutdown Margin in excess of required Shutdown Margin (B.8 - Required SDM from the COLR):
1-748.7 a~,==~L...... pcm - (-)
1770 pcm (B.8)
Required SDM from the CO LR 1 ( ) -2403 to -2451 n:
L _ "'__ ____ (-) 1 1634 lpcm
(+)
50 pcm
(-) 748.7 1
Range:
1
-719 to -767.0 pcml r - - - - - - - - - - -I 1
(+) 1021.3 1
1 (
Range: 1003
~m 1
to IOS1 1..._...... ___
B.1O IfB.9 is positive, THEN emergency borate per FNP-I-AOP-27.0, EMERGENCY
...--~~-~-
EmergencyB~ration BORATION, to establish the required shutdown margin otherwise Shutdown ISrequired.
. Margin is adequate for the Present plant condition.
May.also state that a Rx tripmay be warranted (P&L43).
DA TE-- Element IS-if assumes/states emergency boration completed Signed-- element I S-if assumes/states Performed by: _
Emerg Boration completed Verified by: _______________ _
Page 3 of3 Revision 4
Westinghouse Proprietary Class 2 Notes:
UNIT 1 CYCLE 23 CURVE 77 4831 4935 5017 5063 5131 5072 5013 4965 4916 4884 4852 4828 4803 4795 4787 4787 4787 4800 5071 5180 5267 5320 5396 5346 5296 5254 5211 5182 5154 5129 5105 5096 5087 5086 5085 5096 5311 5425 5517 5578 5660 5619 5579 5542 5506 5480 5455 5431 5408 5398 5387 5385 5382 5391 5483 5594 5680 5740 5814 5778 5743 5714 5685 5669 5652 5641 5629 5631 5633 5642 5652 5669 PCB-1-VOL1-CRV77 1.
Rod worth data assumes the starting point is rods at the Rod Insertion Limit. If a bank is below the RIL, reduce the table value by 75 pcm for each step below the RIL.
2.
The rod worth data represents negative reactivity. However, STP-29.1, STP-29.2 and STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the STP.
3.
ARI4 is defined as All Rods In -- less the most reactive rod. The SDM calculation requires that one rod be assumed stuck. Therefore, use the data directly from the table above when there are not any known stuck rods.
150
> 150 1000 4587 4667 4710 4801
> 1000 2000
>2000 3000
>3000 S 4000
> 4000 5 5000 4717 4755 4818 4767 5655 4867 4909 4975 4920 5792
> 5000 6000 5762 5927 5905 4717 j
KEY I
REV.
APPROVED:
)
ENGINEERING SUPPORT MANAGER DATE ARI-1 Rod Worth (pcm) for Shutdown Margin Calculations Burnup Range Power Level (%)
(MWD/MTU) 0 10 20 30 40 50 60 70 80 90 100
>17000 518000 4682 4748 4813 5107 5400 5687 5973 6028 6083 6304 6526
>18000 519000 4709 4771 4833 5125 5418 5710 6002 6056 6109 6328 6547
>19000 20000 4735 4794 4852 5144 5435 5734 6032 6083 6135 6351 6568
>20000 520885 4760 4818 4876 5165 5454 5759 6063 6112 6161 6377 6592 6162
> 6000 7000
> 7000 5 8000 6048 4865 6397 6285 4689 4660 5844 5902 5969 5938 6523 4827 4788 5995 6052 6113 6073 6145 6201 6257 6208 6384 6442 6483 6439 5906
>8000 59000
> 9000 10000
>10000 511000
>11000 512000
>12000 513000
>13000 514000 6623 6683 6708 6670 6033 4646 4632 4622 4612 4617 4621 6159 5885 5864 4765 4742 4725 4707 4706 4704 6395 6631
>14000 515000
>15000 16000
>16000 517000 5857 5850 5850 5850 5865 5879 6001 6118 finent #6 6028 6016 6004 6010 6015 4631 4641 4662 5j 5939 5933 5927 5937 5947 4709 4714 4731 6355 6316 6292 6268 6253 6239 6242 6245 6593 6555 6531 6507 6490 6473 6475 6476 5900 5921 5947 5964 5981 6004 6028 6041 6062 6256 6267 6286 6485 6493 6510 Page 1 of 2 Westinghouse Proprietary Class 2 PCB-l-VOU-CRV77 UNIT 1 CYCLE 23 CURVE 77 I
REV.~
~~
APPROVED: <?(~...e &
Iv,.-
pry"..,
C'7r'n~'"
ENGINEERING SUPPORT MANAGER DATE ARI-l Rod Worth (pcm) for Shutdown Margin Calculations BumupRange Power Level (%)
(MWDIMTU) 0 10 20 30 40 50 60 70 80 90 100
=:;150 4587 4710 4831 5071 5311 5483 5655 5792 5927 6162 6397
> 150
=:;1000 4667 4801 4935 5180 5425 5594 5762 5905 6048 6285 6523
> 1000
=:;2000 4717 4867 5017 5267 5517 5680 5844 5995 6145 6384 6623
> 2000
=:;3000 4755 4909 5063 5320 5578 5740 5902 6052 6201 6442 6683
>3000
=:;4000 4818 4975 5131 5396 5660 5814 5969 6113 6257 6483 6708
>4000
=:;5000 4767 4920 5072 5346 5619 5778 5938 6073 6208 6439 6670
> 5000
=:;6000 4717 4865 5013 5296 5579 5743 5906 6033 6159 6395 6631
> 6000
=:;7000 4689 4827 4965 5254 5542 5714 5885 6001 6118 6355 6593
> 7000
=:;8000 4660 4788 4916 5211 5506 5685 5864 51 Element #6 6316 6555
.> 800.0
=:;c;)OQO..
.464.6.
4:7'6.5
.4~84 5182 5480 5669 5857 54 6292 6531
>9000
=:;10000 4632 4742 4852 5154 5455 5652 5850 5939 6028 +~~}~ 6507
>10000
=:;11000 4622 4725 4828 5129 5431 5641' 5850 5933 6016 h 1----6490
>11000
=:;12000 4612 4707 4803 5105 5408 5629 5850 5927 6004 6239 6473
>12000
=:;13000 4617 4706 4795 5096 5398 5631 5865 5937 6010 6242 6475
>13000
=:;14000 4621 4704 4787 5087 5387 5633 5879 5947 6015 6245 6476
>14000
=:;15000 4631 4709 4787 5086 5385 5642 5900 5964 6028 6256 6485
>15000
=:;16000 4641 4714 4787 5085 5382 5652 5921 5981 6041 6267 6493
>16000
=:;17000 4662 4731 4800 5096 5391 5669 5947 6004 6062 6286 6510
>17000
=:;18000 4682 4748 4813 5107 5400 5687 5973 6028 6083 6304 6526
>18000
=:;19000 4709 4771 4833 5125 5418 5710 6002 6056 6109 6328 6547
>19000
=:;20000 4735 4794 4852 5144 5435 5734 6032 6083 6135 6351 6568
>20000
=:;20885 4760 4818 4876 5165 5454 5759 6063 6112 6161 6377 6592 Notes:
Rod worth data assumes the starting point is rods at the Rod Insertion Limit. If a bank is below the RIL, reduce the table value by 75 pcm for each step below the RIL.
The rod worth data represents negative reactivity. However, STP-29.1, STP-29.2 and STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the STP.
ARI-l is defined as All Rods In --less the most reactive rod. The SDM calculation requires that one rod be assumed stuck. Therefore, use the data directly from the table above when there are not any known stuck rods.
Page 1 of 2
Westinghouse Proprietary Class 2 cr ENGINEERING SUPPORT MANAGER PCB-1-VOL1-CRv77 Worth of Most Reactive Rod (pcm) for Shutdown Margin Calculations Notes:
Burnup Range Power Level (MWD/MTU) 0% to 100%
150 658
> 150 1000 676
> 1000 2000 732
> 2000 3000 781
>3000 4000 889
> 4000 5000 1024
> 5000 6000 1160
> 6000 7000 1232
>10000 11000 1450
>11000 12000 1494
>12000 13000 1528
>13000 14000 1561
>14000 15000 1592
>15000 16000 1623
>16000 17000 1653
>17000 18000 1684
>18000 19000 1713
>19000 20000 1741
>20000 20885 1767 1.
Rod worth data assumes the starting point is rods at the Rod Insertion Limit.
2.
The rod worth data represents negative reactivity. However, STP-29.1, STP-29.2 and STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the STP.
3.
ARI-1 is defined as All Rods In
-- less the most reactive rod.
4.
For multiple stuck rods, the ARI-1 value on Page 1 of this curve should be reduced by the following calculation:
[KUR x WSR x 1.75] + [WSR x 0.75]
where KUR Number of known untrippable rods (does not include rod assumed stuck in ARI-1 table) and WSR = Worst stuck rod (i.e. most reactive rod)
UNIT 1 CYCLE 23 CURVE 77 Control Rod Worth for SDM Calculations KEY REV.
APPROVED:
4 1 LI_/c c DATE
> 7000 8000 element #9 :i-
> 8000 9000 1304
> 9000 10000 1355 1406 Page 2 of 2 Westinghouse Proprietary Class 2 PCB-I-VOLl-CRV77 Notes:
UNIT 1 CYCLE 23 CURVE 77 Control Rod Worth for SDM Calculations APPROVED: 0/ ~
A': f-bf lor ptrtfYl f1'-;,jW ~
ENGINEERING SUPPORT MANAGER Worth of Most Reactive Rod (pern) for Shutdown Margin Calculations Burnup Range Power Level (MWDIMTU) 0% to 100%
> 150
>1000
>2000
>3000
>4000
> 5000
> 6000
> 7000
<8000 1304 I element #9 I
> 8000
> 9000
>10000
>11000
>12000
>13000
>14000
>15000
>16000
>17000
>18000
>19000
>20000
Rod worth data assumes the starting point is rods at the Rod Insertion Limit.
The rod worth data represents negative reactivity. However, STP-29.1, STP-29.2 and STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the STP.
ARI-1 is defined as All Rods In -- less the most reactive rod.
For multiple stuck rods, the ARl-l value on Page 1 of this curve should be reduced by the following calculation:
[KUR x WSR x 1.75] + [WSR x 0.75]
where KUR== Number of known untrippable rods (does not include rod assumed stuck in ARl-l table) and WSR = Worst stuck rod (i.e. most reactive rod)
Page 2of2
Westinghouse Proprietary Class 2 UNIT 1 CYCLE 23 CURVE 78 PCB-1-VOL1-CRV78 Power Defect (pcm) for Shutdown Margin Calculations Burnup Range Power Level (%)
(MWDiMTU) 0 10 20 30 40 50 60 70 80 90 100 150 0
361 537 717 898 958 1019 1119 1219 1326 1432 150 1000 0
351 522 701 879 943 1006 1106 1206 1307 1409
> 1000 2000 0
330 495 667 839 903 967 1068 1169 1265 1361
> 2000 3000 0
317 476 647 818 879 941 1041 1142 1237 1332
> 3000 4000 0
326 489 664 839 901 964 1062 1161 1261 1360
> 4000 5000 0
344 514 697 879 942 1004 1103 1202 1306 1410
>5000 6000 0
362 539 729 919 982 1045 1144 1242 1351 1459
>6000 7000 0
388 576 774 972 1037 1103 1202 1301 1416 1531
> 7000 8000 0
414 613 818 1024 1092 1160 1260 1360 1481 1603
> 8000 9000 0
440 651 863 1074 1147 1220 entii]
1558 1684
>9000 10000 0
466 690 907 1124 1202 1280 1391 1503 1634 1765
>10000 11000 0
492 728 950 1173 1258 1344 1461 1579 1717 1854
>11000 12000 0
518 766 994 1222 1315 1408 1531 1655 1800 1944
>12000 13000 0
543 804 1035 1267 1369 1471 1603 1734 1884 2034
>13000 14000 0
567 841 1077 1313 1424 1535 1674 1813 1968 2123
>14000 15000 0
589 875 1115 1356 1477 1597 1744 1890 2054 2217
>15000 16000 0
611 908 1153 1399 1529 1660 1813 1967 2139 2310
>16000 17000 0
633 940 1189 1438 1578 1718 1882 2046 2223 2400
>17000 18000 0
654 971 1225 1478 1627 1777 1951 2125 2307 2489
>18000 19000 0
673 1001 1259 1516 1677 1837 2020 2203 2392 2582
>19000 20000 0
692 1031 1293 1554 1726 1897 2089 2280 2477 2674
>20000 20885 0
709 1057 1323 1590 1771 1953 2152 2351 2554 2757 Notes:
1.
STP-29.1, STP-29.2 and STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the Si]?.
REV.
APPROVED:
ENGINEERING SUPPORT MANAGER DATE Burnup Range (MWDIMTU)
S 150
>150 s1000
>1000 s2000
>2000 s3000
>3000 s4000
>4000 s 5000
> 5000 s6000
> 6000 s7000
>7000 s8000
... > 8(100.. s 90(10.....
>9000 s10000
>10000 s11000
>11000 s12000
>12000 s13000
>13000 s14000
>14000 s15000
>15000 s16000
>16000 s17000
>17000 s18000
>18000 s19000
>19000 s20000
>20000 s20885 Notes:
Westinghouse Proprietary Class 2 UNIT 1 CYCLE 23 CURVE 78 APPROVED: ~<t" (fL""" ~
ENGINEERING SUPPORT MANAGER PCB-l-VOLI-CRV78 Power Defect (pcm) for Shutdown Margin Calculations Power Level (%)
0 10 20 30 40 50 60 70 80 90 100 0
361 537 717 898 958 1019 1119 1219 1326 1432 0
351 522 701 879 943 1006 1106 1206 1307 1409 0
330 495 667 839 903 967 1068 1169 1265 1361 0
317 476 647 818 879 941 1041 1142 1237 1332 0
326 489 664 839 901 964 1062 1161 1261 1360 0
344 514 697 879 942 1004 1103 1202 1306 1410 0
362 539 729 919 982 1045 1144 1242 1351 1459 0
388 576 774 972 1037 1103 1202 1301 1416 1531 0
414 613 818 1024 1092 1160 1260 1360 1481 1603
.0.................. ~Q 651 1~~I: 1074 1147 1220 11 Element #12 I 1558 1684 0
466 690 907 1124 1202 1280 1391 1503 1634 1765 0
492 728 950 1173 1258 1344 1461 1579 1717 1854 0
518 766 994 1222 1315 1408 1531 1655 1800 1944 0
543 804 1035 1267 1369 1471 1603 1734 1884 2034 0
567 841 1077 1313 1424 1535 1674 1813 1968 2123 0
589 875 1115 1356 1477 1597 1744 1890 2054 2217 0
611 908 1153 1399 1529 1660 1813 1967 2139 2310 0
633 940 1189 1438 1578 1718 1882 2046 2223 2400 0
654 971 1225 1478 1627 1777 1951 2125 2307 2489 0
673 1001 1259 1516 1677 1837 2020 2203 2392 2582 0
692 1031 1293 1554 1726 1897 2089 2280 2477 2674 0
709 1057 1323 1590 1771 1953 2152 2351 2554 2757
STP-29.1, STP-29.2 and STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the STP.
COLR for FNP Unit 1 Cycle 23 Revision 0 Pacie 10 of 13 225 200 175 C
-c Q. 125 U)
C0
C 0
150 75 50 25 0
Figure 1 Rod Bank Insertion Limits versus Rated Thermal Power Fully Withdrawn 225 to 231 steps, inclusive Elements 4 & 5 0.0 0.1 0.2 0.3 04 0.5 0.6 0.7 0.8 0.9 1.0 Fraction of RATED THERMAL POWER Fully Withdrawn shall be the condition where control rods are at a position within the interval 225 and 231 steps withdrawn.
Note: The Rod Bank Insertion Limits are based on the control bank withdrawal sequence A, B, C, D and a control bank tip-to-tip distance of 128 steps.
COLR for FNP Unit 1 Cycle 23 Revision 0 Figure 1 Rod Bank Insertion Limits versus Rated Thermal Power Fully Withdrawn - 225 to 231 steps, inclusive I
Elements 4 & 5 I
225 I I I I I I I....
(.552,225)
J;!'
I' 200 J;!'
175 I;'
I' Banke
'2
== 150
~
,/
I I I I I II II-f.
"t:I
.J::. -
§:
.1' = (_1Jl.7 - 0.000 )(0.900 - 0.(70)+ 0.000 1.00 - 0.070 i;"
Y = 166.9 (rounded using 4 signticJnt digits)
=> 167sfeps
.l!!
!a-s:::
0 if not calculated and only read from graph I~
(0, 114) then> 165 And < 170; Therefore Half the accuracy of the scale II'
=> 167.5 I'
0Cij 100 0 a.
olo:
I....
I.;'
BankO s:::
C\\I IX!
"t:I 75 0
I....
IJ a::
II' 50 II I I
,/
25
(.070,0) o I I I I r 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 Fraction of RATED THERMAL POWER Page 10 of 13 m
o-(1,187) _
,/
,/
I' i;"
0.9 1.0 Fully Withdrawn shall be the condition where control rods are at a position within the interval ~ 225 and ~ 231 steps withdrawn.
Note: The Rod Bank Insertion Limits are based on the control bank withdrawal sequence A, B, C, 0 and a control bank tip-to-tip distance of 128 steps.
06/18/01 10 31 04 FNP-1-STP-29 5 April 21,2000 Revision 4 FARLEY NUCLEAR PLANT SURVEILLANCE TEST PROCEDURE FNP-I -STP-29.5 S
A F
E T
Y SHUTDOWN MARGIN CALCULATION IN MODES I AND 2 (TAVG 547°F)
R E
L A
T E
D PROCEDURE USAGE REQUIREMENTS PER FNP-0-AP-6 SECTIONS Continuous Use ALL Reference Use Information Use Approved:
C. D. COLLINS Operations Manager Date Issued 4-27-00 06/18/01 10:31 :04 FARLEY NUCLEAR PLANT SURVEILLANCE TEST PROCEDURE FNP-I-STP-29.5 FNP-I-STP-29.5 April 2], 2000 Revision 4 SHUTDOWN MARGIN CALCULATION IN MODES] AND 2 (TAVG ~ 547°F)
PROCEDURE USAGE REQUIREMENTS PER FNP-0-AP-6 SECTIONS Continuous Use ALL Reference Use Infonnation Use Approved:
C. D. COLLINS Operations Manager S
A F
E T
Y R
E L
A T
E D
Date Issued
_4..:....-=27'---"'-00~ _____ _
06/18/01 10 31 04 FNP-I-STP-29 5 FARLEY NUCLEAR PLANT SURVEILLANCE TEST REVIEW SHEET SURVEILLANCE TEST NO.
TECHNICAL SPECIFICATION REFERENCE FNP-I -STP-29.5 TR 13.1.1; LCO 3.1.4; LCO 3.1.5; LCO 3.1.6 TITLE MODE(S) REQUIRING TEST:
SHUTDOWN MARGIN CALCULATION IN MODES I AND 2 1 *,2*
(TAVG 547°F)
PERFORMED BY DATE/TIME_________________________
COMPONENT OR TRAIN TESTED (if applicable)
[j ENTIRE STP PERFORMED
[] FOR SURVEILLANCE CREDIT
{] PARTIAL STP PERFORMED:
[1 NOT FOR SURVEILLANCE CREDIT REASON FOR PARTIAL:
TEST COMPLETED:
[] Satisfactory
[ j Unsatisfactory
[] The following deficiencies occurred:
[] Corrective action taken or initiated:
SHIFT FOREMAN REVIEW REVIEWED BY DATE
[j Procedure properly completed and satisfactory
[] Comments:
TECHNICAL GROUP REACTOR ENG REVIEW REVIEWED BY
[ J Satisfactory and Approved
[ j Comments:
Revision 4 06/18/01 10:31:04 FNP-l-STP-29.5 FARLEY NUCLEAR PLANT SURVEILLANCE TEST REVIEW SHEET SURVEILLANCE TEST NO.
TECHNICAL SPECIFICA nON REFERENCE FNP-1-STP-29.5 TR 13.1.1; LCO 3.1.4; LCO 3.1.5; LCO 3.1.6 TITLE MODE(S) REQUIRING TEST:
SHUTDOWN MARGIN CALCULATION IN MODES 1 AND 2 1 *,2*
(TAVG ~ 547°F)
PERFORMED BY DATE/TIME COMPONENT OR TRAIN TESTED (if applicable)
[ ] ENTIRE STP PERFORMED
[ ] FOR SURVEILLANCE CREDIT
[] PARTIAL STP PERFORMED:
[ ] NOT FOR SURVEILLANCE CREDIT REASON FOR PARTIAL:
TEST COMPLETED:
[ ] Satisfactory
[ ] Unsatisfactory
[ ] The following deficiencies occurred:
[ ] Corrective action taken or initiated:
SHIFT FOREMAN REVIEW REVIEWED BY DATE
[ ] Procedure properly completed and satisfactory
[ ] Comments:
TECHNICAL GROUP-REACTOR ENG REVIEW REVIEWED BY DATE
[ ] Satisfactory and Approved
[ ] Comments:
Revision 4
06/18/01 10:31:04 FNP-1STP-29.5 TABLE OF CONTENTS Procedure Contains Number of Pages Body 2
STRS I
Page 1 of I Revision 4 06118/01 10:31 :04 FNP-1-STP-29.5 TABLE OF CONTENTS Procedure Contains Number of Pages Body......................................................... 2 Shutdown Margin..................................... 3 STRS........................................................ 1 Page 1 of 1 Revision 4
06/18/01 10 31 04 FNP-1-STP-29 5 FARLEY NUCLEAR PLANT UNIT I SURVEILLANCE TEST PROCEDURE STP-29.5 SHUTDOWN MARGIN CALCULATION IN MODES 1 AND 2 (TAVG 547°F) 1.0 Purpose 1.1 The purpose of this procedure is to verify that the SHUTDOWN MARGIN is greater than the limits of the Technical Requirements Manual TR 13.1.1 as directed by Technical Specifications in Modes I or 2:
.1.1 One or more rod(s) untrippable (LCO 3.1.4),
1.1.2 One or more rod(s) not aligned within 12 steps of their group step counter demand position (LCO 3.1.4),
1.1.3 One or more shutdown banks not within insertion limits specified in the COLR(LCO 3.1.5),
1.1.4 Control bank insertion, sequence, or overlap not within limits specified in the COLR (LCO 3.1.6). (This condition does not affect shutdown margin unless control banks are below insertion limits) 2.0 Acceptance Criteria 2.1 The Shutdown Margin shall be greater than or equal to the value specified in the COLR.
3.0 Initial Conditions HBF 3.1 The revision of this procedure has been verified to be the current revision and correct unit for the task. (OR I 498)
HBF 3.2 TAVG 547° F.
HBF 33 TAVG +/- I OF of programmed value. (Needed for Power Defect to be accurate)
HBF 3.4 The plant is in Mode I or 2 with Keff 1.
HBF 3.5 Technical Specifications required action entered to verify SDM to be within the limits provided in the COLR Qj Shift Supervisor desires to perform.
-I-Revision 4 06118/01 10:31:04 FNP-I-STP-29.5 FARLEY NUCLEAR PLANT UNIT 1 SURVEILLANCE TEST PROCEDURE STP-29.5 SHUTDOWN MARGIN CALCULATION IN MODES 1 AND 2 (TAVG ~ 547°F) 1.0 Purpose 1.1 The purpose of this procedure is to verify that the SHUTDOWN MARGIN is greater than the limits of the Technical Requirements Manual TR 13.1.1 as directed by Technical Specifications in Modes 1 or 2:
1.1.1 One or more rode s) untrippable (LCO 3.1.4),
1.1.2 One or more rode s) not aligned within 12 steps of their group step counter demand position (LCO 3.1.4),
1.1.3 One or more shutdown banks not within insertion limits specified in the COLR (LCO 3.1.5),
1.1.4 Control bank insertion, sequence, or overlap not within limits specified in the COLR (LCO 3.1.6). (This condition does not affect shutdown margin unless control banks are below insertion limits) 2.0 Acceptance Criteria 2.1 The Shutdown Margin shall be greater than or equal to the value specified in the COLR.
3.0 Initial Conditions HBF 3.1 HBF 3.2 HBF 3.3 HBF 3.4 HBF 3.5 The revision of this procedure has been verified to be the current revision and correct unit for the task. (OR 1-98-498)
T AVG +/- 1 ° F of programmed value. (Needed for Power Defect to be accurate)
The plant is in Mode 1 or 2 with Keff ~ 1.
Technical Specifications required action entered to verify SDM to be within the limits provided in the COLR OR Shift Supervisor desires to perform. Revision 4
06/18/01 10:31:04 FNP-1-STP-29.5 4.0 Precautions and Limitations 4.1 Read all curves as accurately as possible OR most conservatively.
4.2 Observe proper algebraic sign notation throughout the calculation.
4.3 If necessary to emergency borate due to inadequate shutdown margin. THEN consideration should be given to tripping the reactor and entering FNP-I -EEP-0, REACTOR TRIP OR SAFETY INJECTION concurrent with the emergency boration.
5.0 Instructions NOTE:
If the reactor were to shutdown instantaneously, the changes in reactivity would be from control rod position and power defect. In the instant that the reactor shut down, reactivity changes from xenon, samarium and boron concentration would be zero.
5.1 Enter the information required on the attached calculation sheets and calculate the shutdown margin.
NOTE:
5.2 Verify shutdown margin calculated in step 5.1.
6.0 References 6.1 Technical Specifications: LCOs 3.1.4, 3.1.5, & 3.1.6.
6.2 Technical Requirement Manual TR 13.1.1 6.3 Core Operating Limits Report (COLR). Revision 4 06118/0110:31:04 FNP-1-STP-29.5 4.0 Precautions and Limitations 4.1 Read all curves as accurately as possible OR most conservatively.
4.2 Observe proper algebraic sign notation throughout the calculation.
4.3 IF necessary to emergency borate due to inadequate shutdown margin, THEN consideration should be given to tripping the reactor and entering FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION concurrent with the emergency boration.
5.0 Instructions NOTE:
If the reactor were to shutdown instantaneously, the changes in reactivity would be from control rod position and power defect. In the instant that the reactor shut down, reactivity changes from xenon, samarium and boron concentration would be zero.
5.1 Enter the information required on the attached calculation sheets and calculate the shutdown margin.
NOTE:
5.2 Verify shutdown margin calculated in step 5.1.
6.0 References 6.1 Technical Specifications: LCOs 3.1.4, 3.1.5, & 3.1.6.
6.2 Technical Requirement Manual TR 13.1.1 6.3 Core Operating Limits Report (COLR). Revision 4
06/18/01 10:31:04 FNP-1-STP-29.5 SHUTDOWN MARGIN IN MODES 1 AND 2 A.
PRESENT CONDITIONS A.1 Core Bumup MWD/MTU A.2 Power Level.
A.3 Determine number of penalty steps for ROD BANKS below RIL.
(Use COLR or convert the RIL computer reading from % to steps)
A.3.a Record data in table ROD BANK RIL BANK STEPS BELOW BANKS HEIGHT DEMAND RIL CBA FULL OUT CBB FULL OUT CBC CBD SDA FULL OUT SDB FULL OUT TOTAL STEPS BELOW RIL i
Page 1 of 3 Revision 4 06118101 10:31 :04 FNP-I-STP-29.5 SHUTDOWN MARGIN IN MODES 1 AND 2 A.
PRESENT CONDITIONS A.I Core Bumup ___ MWDIMTU A.2 Power Level.
A.3 Determine number of penalty steps for ROD BANKS below RIL.
(Use COLR or convert the RIL computer reading from % to steps)
A.3.a Record data in table ROD BANKRIL BANK STEPS BELOW BANKS HEIGHT DEMAND RIL CBA FULL OUT CBB FULL OUT CBC CBD SDA FULL OUT SOB FULL OUT TOTAL STEPS BELOW RIL
"- ________..1 Page 1 of 3 Revision 4
06/18/01 10 31 04 FNP-1-STP-29 5 A.4 Determine number of penalty steps for RODS below RIL which were not counted in A.3 (Rods below RIL on DRPI when the BANK demand is above RIL)
(Use COLR or convert the RIL computer reading from % to steps)
A.4.a Record data in table ROD NUMBER RIL POSITION DRPI POSITION DIFFERENCE TOTAL STEPS BELOW RIL
.1 B.
SHUTDOWN MARGIN NOTE: Write the values from Curves 77 & 78 directly into the surveillance test.
The correct sign convention has been entered in the STP.
B. I Rod worth for all control and shutdown banks at zero steps at Present Burnup (A.l) and Power Level (A.2).
()
i (Curve 77 Page 1)
L B.2 Penalty value of rod banks below insertion limit:
x 75 pcm/step
=
pcm (A.3.a)
L B.3 Penalty value of individual rods below RIL which were not counted in A.3 (Rods below RIL on DRPI when the BANK demand is above RIL)
I (+)
pcm x 10 pcm/step
=
(A.4.a)
Page 2 of 3 Revision 4 06118101 10:31 :04 FNP-I-STP-29.5 AA Determine number of penalty steps for RODS below RIL which were not counted in A.3 (Rods below RIL on DRPI when the BANK demand is above RIL)
(Use COLR or convert the RIL computer reading from % to steps)
AA.a Record data in table ROD NUMBER RIL POSITION DRPI POSITION DIFFERENCE TOTAL STEPS BELOW RIL B.
SHUTDOWN MARGIN NOTE: Write the values from Curves 77 & 78 directly into the surveillance test.
The correct sign convention has been entered in the STP.
B.l Rod worth for all control and shutdown banks at zero steps at r - - - - - - - - - - -I 1 (_)
pcm 1
1 1
(Curve 77 Page 1) 1 Present Bumup (A. 1 ) and Power Level (A.2).
L __________ _
B.2 Penalty value of rod banks below insertion limit:
B.3
___ steps x 75 pcm/step (A.3.a) 1 1 (+) ____ pcm 1
1 L __________ _
Penalty value of individual rods below RIL which were not counted in A.3 (Rods below RIL on DRPI when the BANK demand is above RIL)
___ steps x 10 pcm/step (AA.a)
Page 2 of3 r------------
1 (+)
pcm:
!... ___________ J Revision 4
06/18/01 10:31:04 FNP-1-STP-29.5 B.4 Stuck / Untrippable Rod penalty:
(B.4.a is N/A and B.4.b is zero if there are no stuckluntrippable rods)
B.4.a Worth of most reactive rod worth at present burnup (A.1)
()
pcm from Curve 77 pg. 2 B.4.b Calculate worth of Stuck/untrippable rods
[
x ()
x 1.75] + [()
x 0.75] = ()
(# Stuck!
(B.4.a)
(B.4.a) untrippable rods)
(Most reactive (Most reactive rod worth) rod worth)
B.5 Penalized rod worth considering stuck!untrippable rods, misaligned rods, rods below the insertion limit, and uncertainty:
[(-)
+
+
- (-)
1 x 0.9 to pcm (B.1)
(B.2)
(B.3)
(B.4.b)
L B.6 Power Defect at Power Level (A.2) and Burnup (A.1)
(Curve 78).
()
B.7 Void Collapse Defect.
(+)
50 pcm B.8 Available Shutdown Reactivity:
.7
()______ ()________ +
50 pcmi (B.5)
(B.6)
(B.7)
B.9 Shutdown Margin in excess of required Shutdown Margin (B.8 Required SDM from the COLR):
()___________ pcm ()
1770 pcm 0)
(B.8)
Required SDM t
(SDMExcess) fromtheCOLR L..
B.10 If B.9 is positive, THEN emergency borate per FNP-l-AOP-27.0, EMERGENCY BORATION, to establish the required shutdown margin otherwise Shutdown Margin is adequate for the Present plant condition.
Date:___________________
Performed by:
Date:____________________
Verified by:
Page 3 of 3 Revision 4 06118/01 10:31 :04 FNP-I-STP-29.5 B.4 Stuck 1 Untrippable Rod penalty:
(B.4.a is N/A and B.4.b is zero if there are no stuck/untrippabJe rods)
B.4.a Worth of most reactive rod worth at present burn up (A. I ) (-) ____ pcm from Curve 77 pg. 2 B.4.b Calculate worth of Stuckluntrippable rods
[
x (-)
x 1.75] + [(-)
x 0.75] = (-) ___ pcm B.5 B.6 B.7
(# Stucki (B.4.a)
(B.4.a) untrippable rods)
(Most reactive (Most reactive rod worth) rod worth)
Penalized rod worth considering stuckluntrippable rods, misaligned rods, rods below the insertion limit, and uncertainty:
[(-.L-) ___ +
+
- (-)
] x 0.9 (B.I)
(B.2)
(B.3)
(B.4.b)
Power Defect at Power Level (A.2) and Burnup (A. I)
(Curve 78).
Void Collapse Defect.
B.8 A vailable Shutdown Reactivity:
( ),----(-),---- +
50 (B.5)
(B.6)
(B.7)
B.9 Shutdown Margin in excess of required Shutdown Margin (B.8 - Required SDM from the COLR):
0 ____ pcm - (-)
1770 pcm (B.8)
Required SDM from the CO LR 1
1 1 ( )
pcm 1 L __________ _
(-) ____ pcm
(+) _---'5o....=O'---_pcm 1- -
I 1
!... ______ __ ~c~
~
r -
- -I
pcm 1
(SDMExcess)
I 1... __________ _
B.IO If B.9 is positive, THEN emergency borate per FNP-I-AOP-27.0, EMERGENCY BORA TION, to establish the required shutdown margin otherwise Shutdown Margin is adequate for the Present plant condition.
Date:
Performed by: _______________ _
Date:
Verified by: _______________ _
Page 3 of3 Revision 4
Westinghouse Proprietary Class 2 PCB-1-VOL1-CRV77 UNIT 1 CYCLE 23 CURVE 77 Control Rod Worth for SDM Calculations REV.
APPROVED:
fd, ENGINEERING SUPPORT MANAGER DATE ARI-1 Rod Worth (pcm) for Shutdown Margin Calculations Bumup Range Power Level (%)
(MWDIMTU) 0 10 20 30 40 50 60 70 80 90 100 150 4587 4710 4831 5071 5311 5483 5655 5792 5927 6162 6397
> 150 1000 4667 4801 4935 5180 5425 5594 5762 5905 6048 6285 6523
> 1000 2000 4717 4867 5017 5267 5517 5680 5844 5995 6145 6384 6623
>2000 3000 4755 4909 5063 5320 5578 5740 5902 6052 6201 6442 6683
> 3000 4000 4818 4975 5131 5396 5660 5814 5969 6113 6257 6483 6708
> 4000 5000 4767 4920 5072 5346 5619 5778 5938 6073 6208 6439 6670
> 5000 6000 4717 4865 5013 5296 5579 5743 5906 6033 6159 6395 6631
> 6000 7000 4689 4827 4965 5254 5542 5714 5885 6001 6118 6355 6593
> 7000 8000 4660 4788 4916 5211 5506 5685 5864 5970 6076 6316 6555
> 8000 9000 4646 4765 4884 5182 5480 5669 5857 5955 6052 6292 6531
> 9000 10000 4632 4742 4852 5154 5455 5652 5850 5939 6028 6268 6507
>10000 11000 4622 4725 4828 5129 5431 5641 5850 5933 6016 6253 6490
>11000 12000 4612 4707 4803 5105 5408 5629 5850 5927 6004 6239 6473
>12000 13000 4617 4706 4795 5096 5398 5631 5865 5937 6010 6242 6475
>13000 14000 4621 4704 4787 5087 5387 5633 5879 5947 6015 6245 6476
>14000 15000 4631 4709 4787 5086 5385 5642 5900 5964 6028 6256 6485
>15000 16000 4641 4714 4787 5085 5382 5652 5921 5981 6041 6267 6493
>16000 17000 4662 4731 4800 5096 5391 5669 5947 6004 6062 6286 6510
>17000 18000 4682 4748 4813 5107 5400 5687 5973 6028 6083 6304 6526
>18000 19000 4709 4771 4833 5125 5418 5710 6002 6056 6109 6328 6547
>19000 20000 4735 4794 4852 5144 5435 5734 6032 6083 6135 6351 6568
>20000 20885 4760 4818 4876 5165 5454 5759 6063 6112 6161 6377 6592 Notes:
1.
Rod worth data assumes the starting point is rods at the Rod Insertion Limit. If a bank is below the RIL, reduce the table value by 75 pcm for each step below the RIL.
2.
The rod worth data represents negative reactivity. However, STP-29.1, STP-29.2 and STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the STP.
3.
ARI-1 is defined as All Rods In -- less the most reactive rod. The SDM calculation requires that one rod be assumed stuck. Therefore, use the data directly from the table above when there are not any known stuck rods.
Page 1 of 2 Burnup Range (MWDIMTU)
>150
> 1000
>2000
> 3000
>4000
> 5000
> 6000
> 7000
> 8000
> 9000
>10000
>11000
>12000
>13000
>14000
>15000
>16000
>17000
>18000
>19000
>20000
REv.l Westinghouse Proprietary Class 2 UNIT 1 CYCLE 23 CURVE 77 Control Rod Worth for SDM Calculations APPROVED: t')(~..e (<4; Iv,-
P'Pl-n c'?r'no-.,#
ENGINEERING SUPPORT MANAGER PCB-l-VOLl-CRV77 DATE ARI-l Rod Worth (pcm) for Shutdown Margin Calculations Power Level (%)
0 10 20 30 40 50 60 70 80 90 100 4587 4710 4831 5071 5311 5483 5655 5792 5927 6162 6397 4667 4801 4935 5180 5425 5594 5762 5905 6048 6285 6523 4717 4867 5017 5267 5517 5680 5844 5995 6145 6384 6623 4755 4909 5063 5320 5578 5740 5902 6052 6201 6442 6683 4818 4975 5131 5396 5660 5814 5969 6113 6257 6483 6708 4767 4920 5072 5346 5619 5778 5938 6073 6208 6439 6670 4717 4865 5013 5296 5579 5743 5906 6033 6159 6395 6631 4689 4827 4965 5254 5542 5714 5885 6001 6118 6355 6593 4660 4788 4916 5211 5506 5685 5864 5970 6076 6316 6555 4646 4765 4884 5182 5480 5669 5857 5955 6052 6292 6531 4632 4742 4852 5154 5455 5652 5850 5939 6028 6268 6507 4622 4725 4828 5129 5431 5641 5850 5933 6016 6253 6490 4612 4707 4803 5105 5408 5629 5850 5927 6004 6239 6473 4617 4706 4795 5096 5398 5631 5865 5937 6010 6242 6475 4621 4704 4787 5087 5387 5633 5879 5947 6015 6245 6476 4631 4709 4787 5086 5385 5642 5900 5964 6028 6256 6485 4641 4714 4787 5085 5382 5652 5921 5981 6041 6267 6493 4662 4731 4800 5096 5391 5669 5947 6004 6062 6286 6510 4682 4748 4813 5107 5400 5687 5973 6028 6083 6304 6526 4709 4771 4833 5125 5418 5710 6002 6056 6109 6328 6547 4735 4794 4852 5144 5435 5734 6032 6083 6135 6351 6568 4760 4818 4876 5165 5454 5759 6063 6112 6161 6377 6592
Rod worth data assumes the starting point is rods at the Rod Insertion Limit. If a bank is below the RIL, reduce the table value by 75 pcm for each step below the RIL.
The rod worth data represents negative reactivity. However, STP-29.1, STP-29.2 and STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the STP.
ARI-l is defined as All Rods In --less the most reactive rod. The SDM calculation requires that one rod be assumed stuck. Therefore, use the data directly from the table above when there are not any known stuck rods.
Page 1 of 2
Westinghouse Proprietary Class 2 UNIT 1 CYCLE 23 CURVE 77 Control Rod Worth for SDM Calculations PCB-1-VOL1-CRV77 Notes:
Burnup Range Power Level (MWD/MTU) 0% to 100%
150 658
> 150 1000 676
> 1000 2000 732
> 2000 3000 781
> 3000 4000 889
> 4000 5000 1024
> 5000 6000 1160
> 6000 7000 1232
> 7000 8000 1304
> 8000 9000 1355
> 9000 10000 1406
>10000 11000 1450
>11000 12000 1494
>12000 13000 1528
>13000 14000 1561
>14000 15000 1592
>15000 16000 1623
>16000 17000 1653
>17000 18000 1684
>18000 19000 1713
>19000 20000 1741
>20000 20885 1767 1.
Rod worth data assumes the starting point is rods at the Rod Insertion Limit.
2.
The rod worth data represents negative reactivity. However, STP-29.1, STP-29.2 arid STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the STP.
3.
ARI-1 is defined as All Rods In -- less the most reactive rod.
4.
For multiple stuck rods, the ARt-i value on Page 1 of this curve should be reduced by the following calculation:
[KUR x WSR x 1.751 + [WSR x 0.751 where KUR = Number of known untrippable rods (does not include rod assumed stuck in ARI-1 table) and WSR = Worst stuck rod (i.e. most reactive rod)
REv.g APPROVED:
/i by (rr ENGINEERING SUPPORT MANAGER DATE Worth of Most Reactive Rod (pcm) for Shutdown Margin Calculations Page 2 of 2 Westinghouse Proprietary Class 2 PCB-I-VOLl-CRV77 Notes:
REV.~
UNIT 1 CYCLE 23 CURVE 77 Control Rod Worth for SDM Calculations
APPROVED: y ~
lor grrtf~ (1'I-:,ltr ENGINEERING SUPPORT MANAGER Worth of Most Reactive Rod (pcm) for Shutdown Margin Calculations Bumup Range Power Level (MWDIMTU) 0% to 100%
> 150
>1000
>2000
>3000
>4000
> 5000
> 6000
> 7000
> 8000
> 9000
>10000
>11000
>12000
$;13000 1528
>13000
>14000
>15000
$;16000 1623
>16000
>17000
>18000
>19000
>20000
Rod worth data assumes the starting point is rods at the Rod Insertion Limit.
The rod worth data represents negative reactivity. However, STP-29.1, STP-29.2 and STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the STP.
ARI-1 is defined as All Rods In --less the most reactive rod.
For multiple stuck rods, the ARl-l value on Page 1 of this curve should be reduced by the following calculation:
[KUR x WSR x 1.75} + [WSR x 0.75]
where KUR = Number of known untrippable rods (does not include rod assumed stuck in ARl-l table) and WSR = Worst stuck rod (i.e. most reactive rod)
Page 2 of2
Westinghouse Proprietary Class 2 PCB-1-VOL1-CRV78 UNIT 1 CYCLE 23 CURVE 78 Total Power Defect for 5DM Calculations REV.
APPROVED:
ENGINEERiNG SUPPORT MANAGER DATE Power Defect (pcm) for Shutdown Margin Calculations Burnup Range Power Level (%)
(MWD/MTU) 0 10 20 30 40 50 60 70 80 90 100 150 0
361 537 717 898 958 1019 1119 1219 1326 1432
> 150 1000 0
351 522 701 879 943 1006 1106 1206 1307 1409
>1000 2000 0
330 495 667 839 903 967 1068 1169 1265 1361
> 2000 3000 0
317 476 647 818 879 941 1041 1142 1237 1332
> 3000 4000 0
326 489 664 839 901 964 1062 1161 1261 1360
> 4000 5000 0
344 514 697 879 942 1004 1103 1202 1306 1410
> 5000 6000 0
362 539 729 919 982 1045 1144 1242 1351 1459
> 6000 7000 0
388 576 774 972 1037 1103 1202 1301 1416 1531
>7000 8000 0
414 613 818 1024 1092 1160 1260 1360 1481 1603
> 8000 9000 0
440 651 863 1074 1147 1220 1326 1431 1558 1684
>9000 10000 0
466 690 907 1124 1202 1280 1391 1503 1634 1765
>10000 11000 0
492 728 950 1173 1258 1344 1461 1579 1717 1854
>11000 12000 0
518 766 994 1222 1315 1408 1531 1655 1800 1944
>12000 13000 0
543 804 1035 1267 1369 1471 1603 1734 1884 2034
>13000 14000 0
567 841 1077 1313 1424 1535 1674 1813 1968 2123
>14000 15000 0
589 875 1115 1356 1477 1597 1744 1890 2054 2217
>15000 16000 0
611 908 1153 1399 1529 1660 1813 1967 2139 2310
>16000 17000 0
633 940 1189 1438 1578 1718 1882 2046 2223 2400
>17000 18000 0
654 971 1225 1478 1627 1777 1951 2125 2307 2489
>18000 19000 0
673 1001 1259 1516 1677 1837 2020 2203 2392 2582
>19000 20000 0
692 1031 1293 1554 1726 1897 2089 2280 2477 2674
>20000 20885 0
709 1057 1323 1590 1771 1953 2152 2351 2554 2757 Notes:
1.
STP-29.1, STP-29.2 and STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the STP.
Westinghouse Proprietary Class 2 PCB-l-VOLI-CRV78 UNIT 1 CYCLE 23 CURVE 78 Total Power Defect for SDM Calculations REV.~
APPROVED: ~'1;1 (<L<f j"r fjrya",
41',-",,, ~
ENGINEERING SUPPORT MANAGER Power Defect (pcm) for Shutdown Margin Calculations Burnup Range Power Level (%)
(MWD/MTU) 0 10 20 30 40 50 60 70 80 90 100
~150 0
361 537 717 898 958 1019 1119 1219 1326 1432
>150
~1000 0
351 522 701 879 943 1006 1106 1206 1307 1409
>1000
~2000 0
330 495 667 839 903 967 1068 1169 1265 1361
>2000
~3000 0
317 476 647 818 879 941 1041 1142 1237 1332
>3000
~4000 0
326 489 664 839 901 964 1062 1161 1261 1360
>4000
~ 5000 0
344 514 697 879 942 1004 1103 1202 1306 1410
> 5000
~6000 0
362 539 729 919 982 1045 1144 1242 1351 1459
> 6000
~7000 0
388 576 774 972 1037 1103 1202 1301 1416 1531
>7000
~8000 0
414 613 818 1024 1092 1160 1260 1360 1481 1603
> 8000
~9000 0
440 651 863 1074 1147 1220 1326 1431 1558 1684
>9000
~10000 0
466 690 907 1124 1202 1280 1391 1503 1634 1765
>10000
~11000 0
492 728 950 1173 1258 1344 1461 1579 1717 1854
>11000
~12000 0
518 766 994 1222 1315 1408 1531 1655 1800 1944
>12000
~13000 0
543 804 1035 1267 1369 1471 1603 1734 1884 2034
>13000
~14000 0
567 841 1077 1313 1424 1535 1674 1813 1968 2123
>14000
~15000 0
589 875 1115 1356 1477 1597 1744 1890 2054 2217
>15000
~16000 0
611 908 1153 1399 1529 1660 1813 1967 2139 2310
>16000
~17000 0
633 940 1189 1438 1578 1718 1882 2046 2223 2400
>17000
~18000 0
654 971 1225 1478 1627 1777 1951 2125 2307 2489
>18000
~19000 0
673 1001 1259 1516 1677 1837 2020 2203 2392 2582
>19000
~20000 0
692 1031 1293 1554 1726 1897 2089 2280 2477 2674
>20000
~20885 0
709 1057 1323 1590 1771 1953 2152 2351 2554 2757 Notes:
STP-29.1, STP-29.2 and STP-29.5 have been written to handle the correct sign convention when the above data is entered as a positive number. Enter the above numbers directly into the STP.
C 0.
U)
C0 000 C
0 0
0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 Fraction of RATED THERMAL POWER Fully Withdrawn shall be the condition where control rods are at a position within the interval 225 and 231 steps withdrawn.
Note:
The Rod Bank Insertion Limits are based on the control bank withdrawal sequence A, B, C, D and a control bank tip-to-tip distance of 128 steps.
COLR for FNP Unit I Cycle 23 Page 10 of 13 Revision 0 Figure 1 Rod Bank Insertion Limits versus Rated Thermal Power Fully Withdrawn 225 to 231 steps, inclusive 225 200 175 150 125 100 75 50 25 COLR for FNP Unit 1 Cycle 23 Revision 0 Figure 1 Rod Bank Insertion Limits versus Rated Thermal Power Fully Withdrawn - 225 to 231 steps, inclusive 225 II n. I I..-
(.552,225)
IJ 200 IJ 175 Banke "2
~ 150 IV
~
~
"C
.s:::: -
§
~
~
li 125
~
.s
~
~
c::
0 (0, 114)
~
c...
~
1..-
Bank D c::
IV
/Xl "C
75 0
0:::
~
50
~
25
~ (.070,0) o I I I I I 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 Fraction of RATED THERMAL POWER Page 10 of 13 I
(1,187)_
~
~
0.9 1.0 Fully Withdrawn shall be the condition where control rods are at a position within the interval ~ 225 and S 231 steps withdrawn.
Note: The Rod Bank Insertion Limits are based on the control bank withdrawal sequence A, B, C, D and a control bank tip-to-tip distance of 128 steps.
FNPHLT-33ADMIN A.2.IA HANDOUT Pg 1 of I CONDITIONS When I tell you to begin, you are to Determine if Shutdown Margin is adequate using STP-29.5, SHUTDOWN MARGIN CALCULATION IN MODES 1 AND 2 (TAVG 547°F), for Unit 1. The conditions under which this task is to be performed are:
a.
Unit I is stable at 90% with the ramp on hold b.
Bank D indicates 192 by Group Demand.
c.
Seven of the Bank D rods (H2, B8, H14, F6, FlO, Kb, K6) are at 192 steps by DRPI.
d.
Rod P8, in the D bank, has been determined to be Stuck.
e.
Rod P8 is at 162 steps by DRPI.
f.
All other rods are at 229 steps.
g.
Core burnup is 9,800 MWD/MTU bumup.
h.
FNP-1-STP-29.5, SHUTDOWN MARGIN CALCULATION IN MODES I AND 2 (TAVG 547°F), initial conditions are satisfied.
i.
The Shift Supervisor has directed you to complete FNP-I -STP-29.5 starting at step 5.1.
FNP HL T-33 ADMIN A.2.1.A HANDOUT Pg I of I CONDITIONS When I tell you to begin, you are to Determine if Shutdown Margin is adequate using STP-29.5, SHUTDOWN MARGIN CALCULATION IN MODES 1 AND 2 (TAVG 2:: 547°F), for Unit I. The conditions under which this task is to be perfonned are:
547°F), initial conditions are satisfied.
The Shift Supervisor has directed you to complete FNP-1-STP-29.5 starting at step 5.1.
FNP HLT-33 ADMIN Page 1 of 5 A.3.1.A Radiation Control ADMIN G2.3.9 RO & SRO CRO-A.3.1.A TITLE: Calculate the Maximum Permissible Stay Time within Emergency Dose Limits.
PROGRAM APPLICABLE: SOT SOCT____ OLT X
LOCT____
ACCEPTABLE EVALUATION METHOD:
X PERFORM SIMULATE DISCUSS EVALUATION LOCATION:
SIMULATOR CONTROL ROOM X
CLASSROOM PROJECTED TIME:
20 MIN SIMULATOR IC NUMBER:
N/A ALTERNATE PATH TIME CRITICAL PRA____
JPM DIRECTIONS:
1.
Initiation of task may be in group setting, evaluation performed individually upon completion.
2.
The references for this task will be provided as listed or the student may be provided a computer with a generic exam login and access to the EXAM reference disk.
3.
Elements I through 7 may be evaluated by reviewing the responses on the Handout.
TASK STANDARD: Required for successful completion of this JPM:
Calculate dose expected for Tasks 1 through 3 Determine if a Repair team would exceed equipment protection emergency dose limits of EIP 14.0, Personnel Movement, Relocation, Re-Entry and Site Evacuation.
Calculate the maximum allowable stay time for a task; or determine what tasks can be performed, if any, without exceeding limits.
Examinee:
Overall JPM Performance:
Satisfactory Unsatisfactory D
Evaluator Comments (attach additional sheets if necessary)
EXAMINER:
Developer Howard Fitzwater 1 1/06/09 NRC Approval SEE NUREG 1021 FORM ES-301-3 FNP HL T-33 ADMIN Page 1 of5 A.3.1.A Radiation Control ADMIN G2.3.9 - RO & SRO CRO-A.3.1.A TITLE: Calculate the Maximum Permissible Stay Time within Emergency Dose Limits.
PROGRAM APPLICABLE: SOT SOCT OLT~ LOCT __
ACCEPTABLE EV ALUA TION METHOD: ~
PERFORM SIMULA TE DISCUSS EVALUATION LOCATION:
SIMULATOR CONTROL ROOM ----.lL CLASSROOM PROJECTED TIME:
20 MIN SIMULATOR IC NUMBER:
NIA ALTERNATE PATH TIME CRITICAL PRA __
JPM DIRECTIONS:
TASK STANDARD: Required for successful completion of this jPM:
Calculate dose expected for Tasks 1 through 3 Determine if a Repair team would exceed equipment protection emergency dose limits of EIP-14.0, Personnel Movement, Relocation, Re-Entry and Site Evacuation.
Calculate the maximum allowable stay time for a task; or determine what tasks can be performed, if any, without exceeding limits.
Examinee:
Overall JPM Performance:
Satisfactory 0
Unsatisfactory 0 Evaluator Comments (attach additional sheets if necessary)
EXAMINER: _________ __
Howard Fitzwater 11/06/09 SEE NUREG 1021 FORM ES-301-3
FNP HLT-33 ADMIN A.3.1.A Page 2 of 5 CONDITIONS When I tell you to begin, you are to determine if exposure is within emergency dose limits of EIP 14.0, Personnel Movement, Relocation, Re-Entry and Site Evacuation. The conditions under which this task is to be performed are:
a.
A General Emergency has been declared on Unit 1.
b.
IA RHR pump is air bound and must be vented per AOP-12, Attachment 1, RHR Pump Venting.
c.
The TSC has requested that lB RHR pump motor bearing oil levels be checked and filled as required after restoring IA RHR and suggests using the same Repair team. The time required to perform this task is unknown, but is estimated to range between 1 to 20 minutes.
d.
Ted and Joel have been selected to perform the task, their exposure information is stated below.
e.
The Tasks are provided in the table below, and estimated times and doses have been provided.
f.
Each task must be completed by both operators, and must be performed in the order listed.
g.
Both operators are expected to receive equal dose for each job.
h.
The Emergency Director (ED) directs you to perform the following with the information provided:
o Calculate the expected dose for tasks I through 3 and document in the table.
o Determine the tasks, if any, for which the members of the team could be permitted to perform without exceeding the equipment protection emergency dose limits of EIP-14.0.
Task Time allowed!
Dose Location/Task description reqd Rate Dose ft (minutes)
(RIhr) 83 1A RHR Pump and HX rooms! Vent rig installation and 30 5 31 venting 2
100 piping penetration room! Vent rig installation and 15 19 75 venting 121 piping penetration room! Vent rig installation and 20 5 65 venting Inspect 1 B RHR pump motor bearing oil levels and fill as 7
required Year-to-date DOSE Records (REM TEDE)
TED 1.26 JOEL 0.4 Can this team perform all of the tasks?
(Circle one)
YES NO IF yes, then state the maximum permitted stay time for task #4.
IF no, then state the highest sequential task # that can be performed, if any.
INITIATING CUE: You may begin.
FNP HL T-33 ADMIN A.3.1.A CONDITIONS Page 2 of5 When I tell you to begin, you are to determine if exposure is within emergency dose limits of EIP-14.0, Personnel Movement, Relocation, Re-Entry and Site Evacuation. The conditions under which this task is to be performed are:
Each task must be completed by both operators, and must be performed in the order listed.
o Calculate the expected dose for tasks 1 through 3 and document in the table.
o Determine the tasks, if any, for which the members of the team could be permitted to perform without exceeding the equipment protection emergency dose limits of EIP-14.0.
Task Time allowed/
Dose Location/Task description req'd Rate Dose (minutes)
(R/hr) 1 83' 1 A RHR Pump and HX rooms/ Vent rig installation and 30 5.31 venting 2
100' piping penetration room/ Vent rig installation and 15 19.75 venting 3
121' piping penetration room! Vent rig installation and 20 5.65 venting 4
Inspect IB RHR pump motor bearing oil levels and fill as 7
required Year-to-date DOSE Records (REM TEDE)
TED 1.26 JOEL 0.4 Can this team perform all of the tasks? (Circle one)
YES I
NO IF yes, then state the maximum permitted stay time for task #4.
IF no, then state the highest sequential task # that can be performed, if any.
INITIATING CUE: "You may begin."
FNP HLT-33 ADMIN A.3.1.A Page 3 of 5 EVALUATION CHECKLIST ELEMENTS:
STANDARDS:
RESULTS:
(CIRCLE)
START TIME
=1.88REM 60min,)
2.66
{Range 2.6 to 2.7}
S / U b.
4.938 {4.9 to 5.0}
S I U c.
1.88
{l.8to 1.9}
S/U NOTE:
A Cue may be required to obtain responses for each element if the candidate does not clearly document results. Provide the Cue as stated on the Handout.
2.
Evaluates Team exposure within emergency
tasks I through 3.
a.
Emergency Dose limit for equipment a.
10 R is limit protection is 10 R and DOES NOT include current exposure.
b.
Summation of Task 1 through 3 1 OREM (2.66 + 4.94 + 1.88)REM 0
b.
0.52 R remains available 0.52REM0 c.
Determines BOTH team members can c.
Circles the YES choice complete all tasks.
hr V6OminN time is 4.5 mins (0.52Rem II 1=4.493mm 7REM,A hr
,)
{Range: 3 mm to 6 mm)
{RANGE : 3mmnto6mmn}
STOP TIME Terminate when all elements of the task have been completed.
CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.
lhr
= 2.66REM 60 mm
}
=4.938REM 60mmn}
FNP HLT-33 ADMIN A.3.1.A Page 3 of5 EVALUATION CHECKLIST RESULTS:
ELEMENTS:
STANDARDS:
(CIRCLE)
START TIME
{Range 2.6 to 2.7}
S I U (30min)X5.31Rlhrx( lhr. )=2.66REM 60mm
S I U (15min)XI9.75Rlhrx( lhr. )=4.938REM 60mm
{1.8 to 1.9}
S I U 60mm NOTE:
A Cue may be required to obtain responses for each element if the candidate does not clearl document results. Provide the Cue as stated on the Handout.
Emergency Dose limit for equipment protection is lOR and DOES NOT include current exposure.
Summation of Task 1 through 3 1 OREM - (2.66 + 4.94 + 1.88)REM ~ 0 0.52REM~0
Determines BOTH team members can complete all tasks.
(0.52Rem{
hr )(60min) = 4.493 min
\\7 REM hr
{RANGE: 3 min_to _ 6min}
STOP TIME
{Range: 3 min to 6 min)
Terminate when all elements of the task have been completed.
S I U S I U CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.
FNP HLT-33 ADMIN A.3. 1.A Page 4 of 5 GENERAL
REFERENCES:
I.
FNP-0-EIP-14.0, ver22 2.
FNP-0-M-1.0, ver 18.0 3.
KA: G2.3.4 RO 3.2 SRO 3.7 GENERAL TOOLS AN]) EQUIPMENT:
Provide/Acquire:
1.
Computer with access to Exam Reference Disk, or EIP 14.0, ver 22.0 and M-1.0, version 1 8.0.
2.
Calculator 3.
pens/pencils 4.
Scrap paper Critical ELEMENT justification:
STEP Evaluation l.a CRITICAL
- Task Objective I.b CRITICAL
- Task Objective I.c CRITICAL
- Task Objective 2
CRITICAL
- Task Objective 3
CRITICAL
- Task Objective COMMENTS:
IF Teds year to date dose is used, the emergency dose limit will be exceeded and the expected response for element 2 is NO and highest task that can be completed is Task #3.
IF Joels year to date dose is used and the Emergency dose is reduced by this value then the stay time will range from 0 mins to 1.1 mins.
IF emergency dose limit of 25 Rem Then Stay time will be excessive.
IF ADMIN dose limit of 2 REM used then NONE of the tasks can be performed, or Legal limit of 5 REM then Task #2 will be the highest.
FNP HLT-33 ADMIN GENERAL
REFERENCES:
Provide/Acquire:
A.3.I.A Page 4 of5 SRO 3.7
STEP Evaluation I.a CRITICAL - Task Objective I.b CRITICAL - Task Objective I.c CRITICAL - Task Objective 2
CRITICAL - Task Objective 3
CRITICAL - Task Objective COMMENTS:
IF Ted's year to date dose is used, the emergency dose limit will be exceeded and the expected response for element 2 is "NO" and highest task that can be completed is Task #3.
IF Joel's year to date dose is used and the Emergency dose is reduced by this value then the stay time will range from ° mins to 1.1 mins.
IF emergency dose limit of 25 Rem Then Stay time will be excessive.
IF ADMIN dose limit of2 REM used then NONE of the tasks can be performed, or Legal limit of 5 REM then Task #2 will be the highest.
FNP HLT-33 ADMIN A.3.1.A HANDOUT Page 1 of I CONDITIONS When I tell you to begin, you are to determine if exposure is within emergency dose limits of EIP 14.0, Personnel Movement, Relocation, Re-Entry and Site Evacuation. The conditions under which this task is to be performed are:
a.
A General Emergency has been declared on Unit 1.
b.
IA RHR pump is air bound and must be vented per AOP-12, Attachment I, RHR Pump Venting.
c.
The TSC has requested that I B RHR pump motor bearing oil levels be checked and filled as required after restoring IA RHR and suggests using the same Repair team. The time required to perform this task is unknown, but is estimated to range between I to 20 minutes.
d.
Ted and Joel have been selected to perform the task, their exposure information is stated below.
e.
The Tasks are provided in the table below, and estimated times and doses have been provided.
f.
Each task must be completed by both operators, and must be performed in the order listed.
g.
Both operators are expected to receive equal dose for each job.
h.
The Emergency Director (ED) directs you to perform the following with the information provided:
o Calculate the expected dose for tasks 1 through 3 and document in the table.
o Determine the tasks, if any, for which the members of the team could be permitted to perform without exceeding the equipment protection emergency dose limits of El P-I 4.0.
Task Time allowed!
Dose Location/Task description reqd Rate Dose (minutes)
(R!hr) 1 83 IA RHR Pump and HX rooms! Vent rig installation and 30 5 31 venting 2
100 piping penetration room! Vent rig installation and 15 19 75 venting 121 piping penetration room! Vent rig installation and 20 5 65 venting Inspect I B RHR pump motor bearing oil levels and fill as 7
required Year-to-date DOSE Records (REM TEDE)
TED 1.26 JOEL 0.4 Can this team perform all of the tasks?
(Circle one)
YES NO IF yes, then state the maximum permitted stay time for task #4.
IF no, then state the highest sequential task # that can be performed, if any.
FNP HL T -33 ADMIN A.3.1.A HANDOUT Page 1 of 1 CONDITIONS When I tell you to begin, you are to determine if exposure is within emergency dose limits of EIP-14.0, Personnel Movement, Relocation, Re-Entry and Site Evacuation. The conditions under which this task is to be performed are:
Each task must be completed by both operators, and must be performed in the order listed.
o Calculate the expected dose for tasks 1 through 3 and document in the table.
o Determine the tasks, if any, for which the members of the team could be permitted to perform without exceeding the equipment protection emergency dose limits of EIP-14.0.
Task Time allowed/
Dose Location/Task description req'd Rate Dose (minutes)
(Rlhr) 1 83' 1 A RHR Pump and HX rooms/ Vent rig installation and 30 5.31 venting 2
100' piping penetration room/ Vent rig installation and 15 19.75 venting 3
121 ' piping penetration room/ Vent rig installation and 20 5.65 venting 4
Inspect 1 B RHR pump motor bearing oil levels and fill as 7
required Year-to-date DOSE Records (REM TEDE)
TED 1.26 JOEL 0.4 Can this team perform all of the tasks? (Circle one)
YES
/
NO IF yes, then state the maximum permitted stay time for task #4.
IF no, then state the highest sequential task # that can be performed, if any.
FNP HLT-33 ADMIN Page 1 of 6 A.4.1.A Emergency Plan G2.4. 14 RO & SRO CRO-A.4. 1.A TITLE: Monitor the Critical Safety Function Status Trees.
PROGRAM APPLICABLE: SOT SOCT OLT X
LOCT____
ACCEPTABLE EVALUATION METHOD:
X PERFORM SIMULATE DISCUSS EVALUATION LOCATION:
X SIMULATOR PROJECTED TIME:
20 MIN SIMULATOR IC NUMBER:
IC-216 ALTERNATE PATH TIME CRITICAL PRA____
JPM DIRECTIONS:
1.
This task will be conducted on the Simulator.
2.
The simulator will remain frozen for the duration to the task, if the candidate attempts to operate any component, no plant response will occur.
3.
The plant computer screens will be turned off.
4.
ALL will perform elements 1 through 7.
5.
ONLY SRO will perform elements 8-10.
TASK STANDARD: Required for successful completion of this JPM:
Identi1v all applicable Critical Safety functions which are challenged.
Identify the highest level challenge to the CSFSTs and the required procedure entry, if any.
SRO ONLY: Perform a proper assessment and procedure transition for the given conditions.
Examinee:
Overall JPM Performance:
Satisfactory 1
Unsatisfactory D
Evaluator Comments (attach additional sheets if necessary)
EXAMINER:
Developer Howard Fitzwater 1 1/09/09 LNRC Approval SEE NUREG 1021 FORM ES-301-3 FNP HLT-33 ADMIN A.4.1.A Emergency Plan G2.4.14 - RO & SRO CRO-A.4.1.A TITLE: Monitor the Critical Safety Function Status Trees.
PROGRAM APPLICABLE: SOT SOCT OLT X LOCT __
ACCEPTABLE EVALUATION METHOD: ~
PERFORM EVALUATION LOCATION: ~
SIMULATOR SIMULATE PROJECTED TIME:
20 MIN SIMULATOR IC NUMBER:
IC-216 ALTERNATE PATH TIME CRITICAL PRA __
JPM DIRECTIONS:
Page 1 of6 DISCUSS
TASK STANDARD: Required for successful completion of this JPM:
Identify all applicable Critical Safety functions which are challenged.
Identify the highest level challenge to the CSFSTs and the required procedure entry, if any.
SRO ONLY: Perform a proper assessment and procedure transition for the given conditions.
Examinee:
Overall JPM Performance:
Satisfactory 0
Unsatisfactory 0 Evaluator Comments (attach additional sheets if necessary)
EXAMINER: _________ __
Howard Fitzwater 11109/09 SEE NUREG 1021 FORM ES-301-3
FNP HLT-33 ADMIN A.4.1.A Page 2 of 6 CONDITIONS When I tell you to begin, you are to monitor the Critical Safety Function Status Trees (CSFST).
The conditions under which this task is to be performed are:
a.
A Large Break LOCA and loss of Off-site power has occurred on Unit I from 100% 10 minutes ago.
b.
The team has transitioned to EEP-l.0, Primary or Secondary Loss of Coolant.
c.
The Integrated Plant Computer has failed.
d.
The STA is en route to the Control Room.
e.
You have been directed to manually monitor Critical Safety Functions using CSF-0.0, Critical Safety Function Status Trees and:
1.
identify all applicable Critical Safety functions which are challenged, if any.
2.
If applicable, identify the highest level challenge to the CSFSTs and the required procedure entry for that condition.
INITIATING CUE: You may begin.
EVALUATION CHECKLIST RESULTS:
ELEMENTS:
STANDARDS:
(CIRCLE)
START TIME I.
Evaluates CSFST at CSF-0.i Subcriticality.
S / U GREEN.
a.
Power Rng < 5%
Both mt RNG SUR zero or neg
Both SR detectors ARE energized c.
NO d.
IR Range <-0.2 DPM
Evaluates CSFST at CSF-0.2 Core Cooling.
a.
CET<1200°F
Subcooling> 16°F {45°}
CET <700 °F
Evaluates CSFST at CSF-0.3 Heat Sink.
a.
NR levels> 31 {48%}
Total AFW flow >395 gpm.
Press in all SG <1129 psig
NR lvl <82%
Press <1075 psig
NR lvl >31 % {48%}
The conditions under which this task is to be performed are:
INITIATING CUE: "You may begin."
EVALUATION CHECKLIST ELEMENTS:
START TIME
Evaluates CSFST at CSF-O.l Subcriticality.
Evaluates CSFST at CSF-O.2 Core Cooling.
Evaluates CSFST at CSF-0.3 Heat Sink.
NRlvl>31%{48%}
STANDARDS:
RESULTS:
(CIRCLE)
S / U GREEN.
FNP HLT-33 ADMIN A.4.1.A Page 3 of 6 EVALUATION CHECKLIST ELEMENTS:
STANDARDS:
RESULTS:
(CIRCLE)
NOTE:
Candidate may inform the SS of the FRP-P.1/ ORANGE path condition upon discovery but should continue to evaluate the remaining CSFSTs. IF CUE required, then provide: SS Acknowledges.
4.
Evaluates CSFST at CSF-0.4 INTEGRITY.
ORANGE. (P.1) a.
Temp decr <100 °F last 60 mm
All Press and CL temps to right of LIMIT A
c.
Al 1 CL temps> 250°F
a.
CTMT press <54 psig b.
CTMT press <27 psig c.
AT least ONE CTMT Spray pump running with flow >1000 gpm.
a.
PRZR level <92%
b.
PRZR level> 15%
ORANGE (Z.1)
Evaluates CSFST at CSF-0.5 Containment.
6.
Evaluates CSFST at CSF-0.6 INVENTORY.
FNP HLT-33 ADMIN AA.l.A EVALUATION CHECKLIST ELEMENTS:
STANDARDS:
Page 3 of6 RESULTS:
(CIRCLE)
NOTE:
Candidate may inform the SS of the FRP-P.lI ORANGE path condition upon discovery but should continue to evaluate the remaining CSFSTs. IF CUE required, then provide: "SS Acknowledges."
Evaluates CSFST at CSF-OA INTEGRITY.
ORANGE. (P.I)
Evaluates CSFST at CSF-0.5 Containment.
Evaluates CSFST at CSF-0.6 INVENTORY.
ORANGE (Z.l)
FNP HLT-33 ADMIN A.4.1.A Page 4 of 6 EVALUATION CHECKLIST RESULTS:
(CIRCLE)
ELEMENTS:
STANDARDS:
7.
Reports to the Shift Supervisor identified
ORANGE paths exist and FRP P.1 entry required.
a.
FRP-P. 1, ORANGE due to Cooldown
>100°F in 60 minutes and < 250°F.
b.
FRP-Z.1 ORANGE due to failed IA CTMT Spray pump and I B CTMT Spray pump discharge valve.
RO ONLY: Terminate when Elements 1-7 have been completed.
IF element 7 correctly assessed then:
SRO CUE: SS acknowledges and directs you to perform FRP-P. I.
8.
(FRP-P.l Step I) Checks RCS pressure Greater
P1-402B and 403B (50 psig) and determined less than.
9.
(FRP-P.l step I RNO) Checks LHSI flow
F1-605B. (3250 gpm)
Transitions to FRP-Z.l or states the FRP-Z.1 is required to be entered.
STOP TIME I SRO ONLY: Terminate when all Critical Elements have been completed.
CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.
FNP HLT-33 ADMIN A.4.1.A EVALUATION CHECKLIST ELEMENTS:
STANDARDS:
Page 4 of6 RESULTS:
(CIRCLE)
Reports to the Shift Supervisor identified conditions and FRP-P.I is highest priority:
> 100°F in 60 minutes and < 250°F.
P.I entry required.
RO ONLY: Terminate when Elements 1-7 have been completed.
IF element 7 correctly assessed then:
SRO CUE: "SS acknowledges and directs you to perform FRP-P.l."
(FRP-P.l Step 1) Checks RCS pressure Greater than 435 psig:
(FRP-P.l step 1 RNO) Checks LHSI flow greater than 1500 gpm.
STOP TIME
Transitions to FRP-Z.l or states the FRP-Z.l is required to be entered.
I SRO ONLY: Terminate when all Critical Elements have been completed.
S / U S / U S / U CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.
FNP HLT-33 ADMIN A.4.l.A Page 5 of 6 GENERAL
REFERENCES:
1.
CSF-0.0, ver 17.0 2.
FNP-1-FRP-P.1, ver 19 3.
KA: G2.4.14 RO 3.8 SRO 4.5 GENERAL TOOLS AND EQUIPMENT:
NONE Critical ELEMENT justification:
STEP Evaluation I
NOT criticalprocedure protocol
, condition not challenged (GREEN) 2 NOT criticalprocedure protocol, conditions Off-Normal (Yellow) 3 NOT criticalprocedure protocol, condition Off-Normal (Yellow) 4 CRITICAL
- Task Objective, condition CHALLENGED (ORANGE) 5 CRITICAL
- Task Objective, condition CHALLENGED (ORANGE) 6 NOT criticalprocedure protocol, condition Off-Normal (Yellow) 7 CRITICAL
- Task Objective, reports both challenged conditions to SS and identifies highest priority procedure.
(END RO ONLY) 8 NOT criticalprocedure protocol 9
NOT criticalprocedure protocol 10 CRITICAL Task Objective; properly assess next procedure transition for conditions. (END SRO ONLY)
COMMENTS:
If the candidate attempts to operate the CTMT Spray pump or discharge valve, and notes that the simulator is frozen, a Cue may be required: The response is as you see it.
FNP HL T-33 ADMIN A.4.1.A GENERAL
REFERENCES:
NONE Critical ELEMENT justification:
STEP 1
2 3
4 5
6 7
8 9
10 Evaluation NOT critical-procedure protocol, condition not challenged (GREEN)
NOT critical-procedure protocol, conditions Off-Normal (Yellow)
NOT critical-procedure protocol, condition Off-Normal (¥ellow)
CRITICAL - Task Objective, condition CHALLENGED (ORANGE)
CRITICAL - Task Objective, condition CHALLENGED (ORANGE)
NOT critical-procedure protocol, condition Off-Normal (Yellow)
CRITICAL - Task Objective, reports both challenged conditions to SS and identifies highest priority procedure. (END RO ONLY)
NOT critical-procedure protocol NOT critical-procedure protocol CRITICAL -Task Objective; properly assess next procedure transition for conditions. (END SRO ONLY)
COMMENTS:
Page 5 of6 If the candidate attempts to operate the CTMT Spray pump or discharge valve, and notes that the simulator is frozen, a Cue may be required: "The response is as you see it."
FNP l-ILT-33 ADMIN A.4.1.A HANDOUT Page 1 of I CONDITIONS When I tell you to begin, you are to monitor the Critical Safety Function Status Trees (CSFST).
The conditions under which this task is to be performed are:
a.
A Large Break LOCA and loss of Off-site power has occurred on Unit 1 from 100% 10 minutes ago.
b.
The team has transitioned to EEP-I.0, Primary or Secondary Loss of Coolant.
c.
The Integrated Plant Computer has failed.
d.
The STA is en route to the Control Room.
e.
You have been directed to manually monitor Critical Safety Functions using CSF-0.0, Critical Safety Function Status Trees and:
I.
Identify all applicable Critical Safety functions which are challenged, if any.
2.
If applicable, identify the highest level challenge to the CSFSTs and the required procedure entry for that condition.
FNP HLT-33 ADMIN A.4.I.A CONDITIONS HANDOUT Page 1 of 1 When I tell you to begin, you are to monitor the Critical Safety Function Status Trees (CSFST).
The conditions under which this task is to be performed are:
I. Identify all applicable Critical Safety functions which are challenged, if any.