ML103130188
| ML103130188 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/15/2010 |
| From: | Chuck Zoia NRC/RGN-III/DRS/OLB |
| To: | Nuclear Management Co |
| Shared Package | |
| ML093500371 | List: |
| References | |
| Download: ML103130188 (42) | |
Text
2010 PRAIRIE ISLAND NUCLEAR GENERATING PLANT INITIAL EXAMINATION OUTLINE SUBMITTAL
ENCLOSURE WITHHELD FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH NUREG-I021 UNTIL AFTER THE EXAMINATION IS COMPLETE.
Xcel Energy@
DEC 0 7 2009 L-PI-09-126 NUREG-I 021 Regional Administrator, Region I l l U S Nuclear Regulatory Commission 2443 Warrenville Road, Suite 21 0 Lisle, Illinois 60532-4352 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 Prairie Island Nuclear Generatinq Plant (PINGP) Initial Operator Licensinq Examination Outlines
Reference:
- 1. Nuclear Regulatory Commission (NRC) letter to Mr. Mark A. Schimmel, Prairie Island Nuclear Generating Plant, Units 1 and 2, Confirmation of Initial License Examination, dated November 5, 2009, Accession Number ML093170483.
In response to Reference 1, Northern States Power Company, a Minnesota Corporation (NSPM), submits the integrated examination outlines for the initial operator licensing examinations to be administered at PINGP the weeks of March 15 and 22, 2010. This information is provided in accordance with guideline ES-201 of NUREG-I 021, Operator Licensing Examination Standards for Power Reactors, Revision 9, Supplement 1.
NUREG-I 021 physical security requirements state that the enclosed examination materials shall be withheld from public disclosure until after the examination is complete.
1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1 121
Regional Administrator, Region Ill Page 2 Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
Mark A. Schimmel Site Vice President, Prairie Island Nuclear Generating Plant Units 1 and 2 Northern States Power Company - Minnesota Enclosure cc:
Charles Zoia, US NRC Region Ill, with enclosure Hironori Peterson, US NRC Region Ill, without enclosure
ENCLOSURE WITHHELD FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH NUREG-I021 UNTIL AFTER THE EXAMINATION IS COMPLETE.
ENCLOSURE 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Initial Operator Licensing Examination Outlines Form Number Title or Description Number of Pages ES-201-2 Examination Outline Quality Checklist Examination Security Agreement Administrative Topics Outline - Reactor Operator (RO)
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Administrative Topics Outline - Senior Reactor Operator Control Room/ln-Plant Systems Outline - RO
( S W 2
Control Room/ln-Plant Systems Outline - SRO-U Control Room/ln-Plant Systems Outline - SRO-I ES-301-2 2
1 ES-301-5 3
1 Transient and Event Checklists ES-D-1 l o I
Scenario Outline PWR Examination Outline (RO) 7 ES-401-2 ES-401-2 PWR Examination Outline (SRO)
Generic Knowledge and Abilities Outline (Tier 3) (RO)
ES-401-3 Generic Knowledge and Abilities Outline (Tier 3) (SRO)
ES-401-3 ES-401-4 Record of Rejected WAS 2010 ILT Exam Random Selection Methodology N/A 45 pages follow
ES-201 Examination Outline Quality Checklist Form ES-201-2 Facility: Prairie Island Date of Examination:
311 512010 Task Description
)f.*.;
- 1.
- a. Verify that the outline(s) fit(s) the appropriate model, in accordance with ES-401.
H w I I
I V I
- 2.
S I
- b. Assess whether the outline was systematically and randomly prepared in accordance with Section D.l of ES-401 and whether all WA categories are appropriately sampled.
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- c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics. J
- d. Assess whether the justifications for deselected or rejected WA statements are appropriate.
- a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions, instrument and component failures, technical specifications, and major transients.
M U
L A
T
- b. Assess whether there are enough scenario sets (and spares) to test the projected number and mix of applicants in accordance with the expected crew composition and rotation schedule without compromising exam integrity, and ensure that each applicant can be tested using at least one new or significantly modified scenario, that no scenarios are duplicated from the applicants audit test(s), and that scenarios will not be repeated on subsequent days.
- c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative and quantitative criteria specified on Form ES-301-4 and described in Appendix D.
3.L ocl
- 3.
w I T
- a. Verify that the systems walk-through outline meets the criteria specified on Form ES-301-2:
(1) the outline(s) contain(s) the required number of control room and in-plant tasks distributed among the safety functions as specified on the form (2) task repetition from the last two NRC examinations is within the limits specified on the form (3) no tasks are duplicated from the applicants audit test(s)
(4) the number of new or modified tasks meets or exceeds the minimums specified on the form (5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria on the form.
7
- b. Verify that the administrative outline meets the criteria specified on Form ES-301-1:
(1) the tasks are distributed among the topics as specified on the form (2) at least one task is new or significantly modified (3) no more than one task is repeated from the last two NRC licensing examinations
- c. Determine if there are enough different outlines to test the projected number and mix of applicants and ensure that no items are duplicated on subsequent days.
- 4.
G E
N E
I R
A L
- a. Author
- b. Facility Reviewer (*)
d NRC Supervisor 12\\la/07
- c. IdRC Chief Examiner
/3//7/07 I
- Independent NRC reviewer initial items in Column c; chief examiner concurrence required.
- Not applicable for NRC-prepared examination outlines NUREG 1021, Revision 9 Supplement 1
ES-301 Administrative Topics Outline Form ES-301-1 P.0 Facility: -Prairie Island Examination Level: RO x SRO Operating Test Number:
Date of Examination: -March 201 0-Administrative Topic (see Note)
Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan Type Code*
Describe activity to be performed Admin 43, Determine the time to boil during reduced P, D, R inventory (2.1.25 3.9/4.2)
Admin 48, RCS/Steam Generator temperature verification. (2.1.20 4.6/4.6)
N, R Admin 61, Approve isolation for A CC Pump (2.2.13 4.1/4.3)
N. R Admin 62, Verification of RWP limits (2.3.7 3.5/3.6)
N/A NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (I 3 for ROs; 5 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (2 1)
(P)revious 2 exams (I 1; randomly selected)
NUREG 1021, Revision 9 Supplement 1
- Bank JPM - This JPM is to determine if the personnel airlock can be open with the given plant conditions. This JPM was used on the 2007 NRC Exam.
A I b
- Bank JPM - This JPM is to determine if plant conditions support starting the first RCP by verifying RCS and S/G temperature difference.
A2 A3 A4
- New JPM - This is a new JPM to approve the isolation of "A" CCW pump for ma i n tena nce.
- New JPM - This is a new JPM to verify if work can be performed with a given RWP.
- Not required NUREG 1021, Revision 9 Supplement 1
L o Form ES-301-1 Facility: -Prairie Island Date of Examination: -March 2010-Examination Level: RO SRO x Administrative Topic (see Note)
Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan Type Code*
N, R N, R Operating Test Number:
Describe activity to be performed Admin 43, Determine the time to boil during reduced inventory (2.1.25 3.9/4.2)
Admin 48, RCS/Steam Generator temperature verification. (2.1.20 4.6/4.6)
Admin 61, Approve isolation for A CC Pump (2.2.13 4.1/4.3)
Admin 62, Verification of RWP limits (2.3.7 3.5/3.6)
Admin 47, Perform Interim Emergency Director Actions (2.4.41 4.6)
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (I 3 for ROs; I 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (2 1)
(P)revious 2 exams (5 1; randomly selected)
NUREG 1021, Revision 9 Supplement 1
- Bank JPM - This JPM is to determine if the personnel airlock can be open with the given plant conditions. This JPM was used on the 2007 NRC Exam.
A I b
- Bank JPM - This JPM is to determine if plant conditions support starting the first RCP by verifying RCS and S/G temperature difference.
A2
- New JPM - This is a new JPM to approve the isolation of A CCW pump for maintenance.
A3 A4
- New JPM - This is a new JPM to verify if work can be performed with a given RWP.
- New JPM - This is a new JPM to perform the Interim Emergency Director actions during a plant event NUREG 1021, Revision 9 Supplement 1
ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Facility:
Prairie Island Date of Examination:
March 2010 Exam Level: RO x SRO-I 0 SRO-U Operating Test No.:
Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System / JPM Title Type Code*
Safety Function
A, D, S 2
- b. EO-31 SF-1, Perform Attachment L: Containment Isolation Actuation Failure
- c. PS-3, Respond to a Pressurizer Level Channel Failing Low
- e. RC-22SF-1, Lower PRT Level A, P, D, S 5
D, S D, S 3
4 s
- f. EA-ISSF, Restore Power to Bus 15 from Unit 2 A, N, EN, S 6
7
- h. CC-6S, Loss of Component Cooling Water N, s 8
Implant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i. VC-16, Borate Unit 1 RCS from Outside the Control Room D, E, R 1
- g. NI-4SF-1, N35 Failure High With Failure Of Reactor to Trip A, D, S
- j. IP-3, Respond to Bypassed Instrument Inverter D, E, P 6
- k. AF-18, Control S/G Water Levels N, E, L 4 s All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO I SRO-I I SRO-U (A)lternate path 4-6 14-6 12-3 (C)ontrol room (D)irect from bank
~ 9 1 s a 1 1 4 (E)mergency or abnormal in-plant (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A)
(R)CA 2 l l 2 l l 2 1 2 1 I 2 1 I 2 1 2 1 121 I21 2 2 1 2 2 1 2 1 I
3 IS 3 IS 2 (randomly selected)
(EN)gineered safety feature
- I - I 21 (control room system)
(P)revious 2 exams (S)imulator NUREG 1021, Revision 9 Supplement 1
b d
e f
cr h
i j
k
- Bank JPM - This JPM aligns Charging flow from the RWST during an ATWS condition
- Bank JPM - This JPM performs Attachment L to complete the isolation of containment by manually aligning components that did not automatically isolate.
- Bank JPM - This JPM has the operator respond to a PZR level channel failing and taking actions IAW the procedure.
- Bank JPM - This JPM has the operator restore the 12 MDAFW pump after a trip on low discharge pressure.
- Bank JPM - This JPM has the operator restore PRT conditions to normal using the RCDT pump.
- New JPM -This is a new JPM to supply Bus 15 from Unit 2.
- Bank JPM - This JPM has the operator respond to conditions as they occur. N35 will fail during a reactor startup and the reactor will not automatically trip, but will trip manually.
- New JPM - This is a new JPM to isolate a leak in the CCW system using IC14 AOPI.
- Bank JPM - This JPM requires the operator borate the RCS using MV-32086 locally.
- Bank JPM - This JPM has the operator to respond to a bypassed instrument inverter and restore the inverter to service.
- New JPM - This is a new JPM that has the operator control SG water levels during a plant shutdown from outside the control room.
NUREG 1021, Revision 9 Supplement 1
- c.
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System I JPM Title
- a. NIA
- b. EO-31 SF-1, Perform Attachment L: Containment Isolation Actuation Failure
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Facility: -Prairie Island Exam Level: RO 0 SRO-I 0 SRO-U x Operating Test No.:
Date of Examination: -March 201 0-
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Type Code*
Safety Function A, D, S 2
Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
- f. EA-I 9SF, Restore Power to Bus 15 from Unit 2
- g. NI-4SF-1, N35 Failure High With Failure Of Reactor to Trip
- h. NIA A, N, EN, S 6
A, D, S 7
- c. NIA
- i. VC-16, Borate Unit 1 RCS from Outside the Control Room
- d. NIA
- e. NIA D, E, R 1
- j. NIA
- k. AF-18, Control SIG Water Levels N, E, L 4 s
- Type Codes (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator Criteria for RO I SRO-I / SRO-U 4-6 14-6 I 2-3 5 915 8 / 5 4 2 1 121 I 2 1 2 1 / 2 1 1 2 1 2 2 1 2 2 1 2 1 5 3 1 5 3 / 5 2 (randomly selected) 2 1 I 2 1 I 2 1
- I - I 21 (control room system)
NUREG 1021, Revision 9 Supplement 1
a b
h I
j k
SRO-U JPM Summah
- NIA
- Bank JPM - This JPM performs Attachment L to complete the isolation of containment by manually aligning components that did not automatically isolate.
- NIA
- NIA
- NIA
- New JPM - This is a new JPM to supply Bus 15 from Unit 2.
- Bank JPM - This JPM has the operator respond to conditions as they occur. N35 will fail during a reactor startup and the reactor will not automatically trip, but will trip manually.
- Bank JPM - This JPM requires the operator borate the RCS using MV-32086 locally.
- New JPM - This is a new JPM that has the operator control SG water levels during a plant shutdown from outside the control room.
- NIA
- NIA NUREG 1021, Revision 9 Supplement 1
ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Facility:
Prairie Island 1 Exam Level: RO 0 SRO-I x SRO-U 0 Date of Examination:
March 2010 Operating Test No.:
Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System / JPM Title Type Code*
Safety Function
- b. NIA
- c. PS-3, Respond to a Pressurizer Level Channel Failing Low 3
- e. RC-22SF-1, Lower PRT Level A, P, D, S 5
D, S D, S
- f. EA-ISSF, Restore Power to Bus 15 from Unit 2 A, N, EN, S 6
7
- h. CC-6S, Loss of Component Cooling Water N, S 8
Implant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i. VC-16, Borate Unit 1 RCS from Outside the Control Room D, E, R 1
- g. NI-4SF-1, N35 Failure High With Failure Of Reactor to Trip A, D, S
- j. IP-3, Respond to Bypassed Instrument Inverter D, E, P 6
- k. AF-18, Control SIG Water Levels N, E, L 4 s All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I I SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (S)imulator
( R W 4-6 14-6 12-3 s 9 1 s a i s 4 2 1 I 2 1 I 2 1 2 1 I 2 1 121 2 2 / 2 2 / 2 1 5 3 I I 3 I I 2 (randomly selected) 2 1 I 2 1 I 2 1
- I - I 21 (control room system)
NUREG 1021, Revision 9 Supplement 1
h i
j k
SRO-I JPM Summary a
C d
e f
g
- Bank JPM - This JPM aligns Charging flow from the RWST during an ATWS condition
- Bank JPM - This JPM has the operator respond to a PZR level channel failing and taking actions IAW the procedure.
- Bank JPM - This JPM has the operator restore the 12 MDAFW pump after a trip on low discharge pressure.
- Bank JPM - This JPM has the operator restore PRT conditions to normal using the RCDT pump.
- New JPM - This is a new JPM to supply Bus 15 from Unit 2.
- Bank JPM - This JPM has the operator respond to conditions as they occur. N35 will fail during a reactor startup and the reactor will not automatically trip, but will trip manually.
- New JPM - This is a new JPM to isolate a leak in the CCW system using IC14 AOPI.
- Bank JPM - This JPM requires the operator borate the RCS using MV-32086 locally.
- Bank JPM - This JPM has the operator to respond to a bypassed instrument inverter and restore the inverter to service.
- New JPM - This is a new JPM that has the operator control SG water levels during a plant shutdown from outside the control room.
b
- NIA NUREG 1021, Revision 9 Supplement 1
E 0 ES-301 Transient and Event Checklist Form ES-301-5 13 14 (Instructions:
A P
P L
I C
A N
T 11 Island Date of Exam:
March 201 0 Operating Test No.: 1 12 4
CREW POSITION Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance.
of-plant (BOP) positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an Instant SRO additionally serves in the BOP position, one I/C malfunction can be credited toward the two IlC malfunctions required for the ATC position.
Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. () Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1 -for-I basis.
Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.
NUREG 1021, Revision 9 Supplement 1
P 6 - 2 3 ES-301 Transient and Event Checklist Form ES-301-5
-acility: Prairie Island Date of Exam:
March 201 0 ODeratina Test No.: 1 15 16 17 J1 RX NOR IIC MAJ TS RX NOR IIC MAJ TS RX NOR IIC MAJ TS RX NOR IIC MAJ TS Scenarios 1
2 3
4 CREW CREW CREW CREW POSITION 1 POSITION 1
POSITION I POSITION M
I N
I M
U M U nstructions:
Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an Instant SRO addifionally serves in the BOP position, one IIC malfunction can be credited toward the two IIC malfunctions required for the ATC position.
Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1 -for-I basis.
Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.
NUREG 1021, Revision 9 Supplement 1
R 0 ES-301 Transient and Event Checklist Form ES-301-5 Facility: Prairie Island Date of Exam:
March 2010 Operating Test No.: 1 R1
?3 34 RX NOR IIC
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MAJ TS RX NOR I/C MAJ TS RX NOR IIC MAJ TS RX NOR I/C MAJ TS Scenarios nstructions:
Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type: TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position.
If an Instant SRO additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1 -for-I basis.
Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.
NUREG 1021, Revision 9 Supplement 1
ES-401 RWIS)OU -0
( f f w o )
PWR Examination Outline Form ES-401-2 Facilitv:
Prairie Island Date Of Exam:
March 201 0 Printed:
Note:
- 1.
- 2.
- 3.
- 4.
- 5.
- 6.
7.*
- 8.
- 9.
Ensure that at least two topics from every applicable WA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each WA category shall not be less than two).
The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by f l from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
Systemslevolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.l.b of ES-401 for guidance regarding the elimination of inappropriate WA statements.
Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
Absent a plant-specific priority, only those WAS having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories.
The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.l.b of ES-401 for the applicable WAS.
On the following pages, enter the WA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #I does not apply). Use duplicate pages for RO and SRO-only exams.
For Tier 3, select topics from Section 2 of the WA catalog, and enter the WA numbers, descriptions, IRs. ana point totals (#) on Form ES-401-3.
I Limit SRO selections to WAS that are linked to
K2 X
X X
K:
X PWR RO Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 I Group 1 Facility:
Prairie Island ES - 401 Printed e *Q Form ES-401-2 Imp.
EIAPE # I Name I Safety Function Point 2.6 000007 Reactor Trip - Stabilization -
Recovery / 1 1
000008 Pressurizer Vapor Space Accident I 3 4.2 1
AA1.01 - PZR spray block valve and PORV block valve 3.9 00001 1 Large Break LOCA I 3 1
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4.2 000022 Loss of Rx Coolant Makeup I 2 3.9 000025 Loss of RHR System / 4 1
2.9
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000026 Loss of Component Cooling Water 18 1
2.6 000027 Pressurizer Pressure Control System Malfunction 1 3 1
000029 ATWS / 1 4.2 1
000038 Steam Gen. Tube Rupture / 3 3.4*
1 000054 Loss of Main Feedwater / 4 X
X -
4.1 1
4.1 3.7 -
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000055 Station Blackout / 6
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000058 Loss of DC Power 16 00062 Loss of Nuclear Svc Water / 4 3.8 1
3.7 1
AK3.08 - Actions contained in EOP for loss of instrument air 000065 Loss of Instrument Air / 8 WlE04 LOCA Outside Containment / 3
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EK2.2 - Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility 3.8 1
4.4 1
WNIE05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 2.1.7 - Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
J ElAPE # I Name I Safety Function K1 PWR RO Examination Outline Printed R.
Facility:
Prairie Island ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 I Group 1 Form ES-401-2 K2 KIA Category Totals:
WIE11 Loss of Emergency Coolant Recirc. 14 3
3 W/E12 Uncontrolled all Steam Generators / 4 K3 X
X A
temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics 2
PWR RO Examination Outline Facility:
Prairie Island ES - 401 EIAPE # I Name I Safety Function Emergency and Abnormal Plant Evolutions Tier I I Group 2 I
000001 Continuous Rod Withdrawal / 1 r 1000005 Inoperable/Stuck Control Rod I 1 000028 Pressurizer Level Malfunction I 2 000059 Accidental Liquid RadWaste Rel. I 9
000076 High Reactor Coolant Activity / 9 I W/E06 Degraded Core Cooling / 4 W/E10 Natural Circ.-With steam void in vessel withlwithout RVLIS / 4 W/E13 Steam Generator Over-pressure
/ 4 W/E16 High Containment Radiation / 9 1
KIA Category Totals: -
1 K2 K3 X
X X
X 2
2 Printed Form ES-401-2 of intensityand the location of the sources of radiation in a nuclear procedures and operation within the limitations in the facility's license and amendments EK2.2 - Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility I EK2.1 - Components, and functions 0' control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features EA2.2 - Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments Imp. -
3.2 3.6 4.2 2.7 2.9 3.5 3.6 3.0 3.0 Point 1
1 1
1 1
1 1
1
PWR RO Examination Outline Printed Facility:
Prairie Island R. 0 ES - 401 Plant Systems - Tier 2 / Group 1 Form ES-401-2 039 Main and Reheat Steam 063 DC Electrical Distribution
PWR RO Examination Outline Printed
'f-e Facility:
Prairie Island ES - 401 Plant Systems -Tier 2 I Group 1 Form ES-401-2 slEvol # I Name
, temperature, and 2
ES-401 PWR Examination Outline Form ES-401-2 Facilitv:
Prairie Island Printed:
Date Of Exam:
March 2010
- 3. Generic Knowledge And Abilities Categories Note:
I.
- 2.
- 3.
- 4.
- 5.
- 6.
7.*
- 8.
- 9.
Ensure that at least two topics from every applicable WA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each WA category shall not be less than two).
The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +_1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systemslevolutions that are not included on the outline should be added. Refer to Section D.l.b of ES-401 for guidance regarding the elimination of inappropriate WA statements.
Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
Absent a plant-specific priority, only those WAS having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories.
The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.l.b of ES-401 for the applicable KIAs.
On the following pages, enter the WA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #I does not apply). Use duplicate pages for RO and SRO-only exams.
For Tier 3, select topics from Section 2 of the WA catalog, and enter the WA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3.
Limit SRO selections to WAS that are linked to 1
ElAPE # I Name I Safety Function 000009 Small Break LOCA / 3 000015/000017 RCP Malfunctions / 4 W/E12 Uncontrolled Depressurization of all Steam Generators / 4
~~~~
~
~
~
000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 K1 000077 Generator Voltage and Electric Grid Disturbances / 6 KIA Category Totals:
0 K2 I I on RCS temperature and pressure, I saturated and superheated AA2.02 - Abnormalities in RCP air vent flow paths and/or oil cooling
.41 - Knowledge of the emergency ion level thresholds and 2.4.9 - Knowledge of low Dower/shutdown implications in Printed P
4 O
PWR SRO Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 I Group 1 Facility:
Prairie Island ES - 401 Form ES-401-2 Imp. -
4.8 3.0 4.6 4.5 4.2 accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.
Point 1
1 1
1 1
1 1
Facility:
Prairie Island ES - 401 K2 PWR SRO Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 I Group 2 KI 0
0 ElAPE # I Name I Safety Function 000024 Emergency Boration I 1
~
000069 Loss of CTMT Integrity / 5 000051 Loss of Condenser Vacuum / 4 i 1000067 Plant Fire On-site I 8 I
WA Category Totals:
Printed Form ES-401-2 KA Topic 2.4.4 - Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
W2.02 - Conditions requiring reactor and/or turbine trip 2.2.40 -Ability to apply Technical Specifications for a system.
AA2.02 - Verification of automatic and manual means of restoring integrity Imp.
4.7 4.1 4.7 4.4 Group Point Total:
Point 1
1 1
1 4
PWR SRO Examination Outline Printed Facility:
Prairie Island rl.0 Point 1
1 1
1 1
5
Facilitv:
Prairi Generic Cateaory Conduct of Operations Is1 2.1.29 Knowledge of how to conduct system lineups, 4.1 1
2.1.41 Knowledge of the refueling process.
2.8 1
Category Total:
2 such as valves, breakers, switches, etc.
Generic Knowledge and Abilities Outline (Tier 3)
Printed:
2.3.14 i d Knowledge of radiation or contamination abnormal, or emergency conditions or activities.
hazards that may arise during normal, 3.4 1
Form ES-401-3 Equipment Control Radiation Control Emergency ProcedureslPlan 2.2.6 Knowledge of the process for making changes to procedures.
2.2.20 Knowledge of the process for managing troubleshooting activities.
~~
2.6 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
3.1 Cateaorv Total:
2.3.1 1 Ability to control radiation releases.
1 3.8 I 1
I 2.4.2 2.4.6 2.4.42
~
~~~
I Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.
Knowledge of EOP mitigation strategies.
Knowledge of emergency response facilities.
3.7 1
2.6 1
1 Category Total:
3 Generic Total:
10 1
Generic Knowledge and Abilities Outline (Tier 3)
PWR SRO Examination Outline Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.
Facilitv:
Prairie Island
3.9 Printed
g - 6 Form ES-401-3 Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc.
Conduct of Operations 3.1 KA Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.
KA Totic Imp. Points 3.9 2.1.5 2.2.35 2.1.I4 Ability to determine Technical Specification 4.5 Mode of Operation.
Radiation Control Emergency ProcedureslPlan 2.3.4 Knowledge of radiation exposure limits under 3.7 1
Category Total:
1 2.4.1 1 Knowledge of abnormal condition procedures.
4.2 1
2.4.50 Ability to verify system alarm setpoints and 4.0 1
normal or emergency conditions.
operate controls identified in the alarm response manual.
Category Total:
2 1
1 Category Total:
2 Equipment Control 2.2.18 1
Generic Total:
7
R,-
Appendix D Scenario Outline Form ES-D-1 Facility: -Prairie Island Scenario No.:
1 Op-Test No.:
1 Examiners:
Operators:
Initial Conditions:
50% Power with xenon increasing Turnover:
No EauiPment 00s 6
T C l l A C(B0P)
Failure of Turbine to Auto Trip.
7
- S105A, C (BOP)
SI Pumps fail to auto start.
SI056 (N)ormal, (Rleactivity, (I Instrument, (C)omponent, (M)ajor NUREG 1021, Revision 9 Supplement 1
Scenario 1 Summary The crew assumes the duty at 50% power with xenon increasing, No equipment 00s. The crew is directed to increase power to 60%. During the power rise a second main feedwater pump will be started.
After the crew has completed the power increase, 11 Charging Pump Trips. The RO will stabilize seal injection flows and/or start 13 charging pump. The crew responds IAW C47 and 1c12.1.
After the plant is stabilized, Turbine First Stage Pressure channel (PT-485) fails high. The RO will take rod control to manual and stabilize charging. The BOP will address the instrument failure and swap steam dumps to Pressure Mode. The SRO will determine T.S. requirements After the plant is stabilized, a 20 gpm RCS leak develops. The SRO will determine T.S.
requirements. The crew responds IAW 1 C4 AOPI.
After the crew identifies the T.S requirements for the RCS leak, a Small Break LOCA occurs.
The crew will trip the reactor and enter 1 E-0. Upon the trip, BOP will recognize the turbine fails to auto trip and both SI pumps fail to auto start and perform manual actions to remedy these failures. The crew will transition to 1 E-I and the scenario will continue until safeguards pumps are stopped in 1 FR-P. 1 or the first SI pump is stopped in 1 ES-1.I.
Critical Tasks:
E-0 -- H: Manually start at least one safety Injection pump before transition out of E-0.
E TCOA4: Control AFW flow within 38 minutes following a Reactor Trip.
(SWI 0-35 identified Time Critical Operator Action).
E-0 -- Q: Trip turbine before leaving E-0.
E-I-- C: Trip all Reactor Coolant Pumps so that a severe challenge to Core Cooling does not occur when forced circulation in the RCS stops (Small Break LOCA).
NUREG 1021, Revision 9 Supplement 1
Appendix D Scenario Out1 i ne Form ES-D-1 Event Type*
R (RO)
Facility: -Prairie Island Scenario No.:
2 Op-Test No.:
1 Examiners:
Operators:
Initial Conditions:
71 % Power and stable Event Description 11 Main Feedwater Pump Trip.
Turnover:
11 CC Pump 00s I
TS (SRO)
Event t N44 Power Range NI Fails High z
M (All) 3 Steam Break I/S Containment on Trip 4
C (BOP) 5 11/12 MSlVs fail to automatically close Malf.
No.
FW 13A N105D ED08C MSOlA RP06 C (SRO, BOP) I I
C (All)
I Loss of Power to Instrument Bus 113 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG 1021, Revision 9 Supplement 1
Scenario 2 Summary The crew assumes the duty at 71% power and stable. 11 CC pump is 00s.
After the crew takes the duty, 11 MFW pump will trip. The RO will reduce power to -50% using IAW 1 C1.4 AOPl. BOP will address the ARP for the failure and manipulate control room switches as necessary to facilitate trouble shooting and ensure proper system response.
After the crew has completed the power reduction, N44 Power Range NI Fails High. The BOP will address the failure and remove the affected channel from service IAW C47 and 1 C51. The RO will place rod control to manual and restore Tave to Tref. The SRO will determine T.S.
requirements After the NI failure is addressed, Instrument Bus 11 3 loses power. The loss of Bus 11 3 compounded with N44 being 00s causes an automatic reactor trip. The crew will respond IAW 1 E-0 and transition to 1 ES-0.1. The SS will direct performance of 1 C20.8 AOPI to restore power to Instrument Bus 11 3 in conjunction with 1 ES-0.1.
After power is restored to Instrument Bus 113, 11 SIG Main Steam line break in containment.
The crew will transition back to 1 E-0. The BOP will recognize the failure of 1111 2 MSIVs to auto close and manually perform the action using 1 E-0 Att. L. The crew will then transition to 1 E-2 to isolate the faulted SG and transition to 1 E-I. The scenario will be terminated when the crew transitions to 1 FR-P.l or upon termination of SI pumps in 1 ES-0.2.
Critical Tasks:
E-0 -- P: Manually close MSlVs before transition out of E-0.
E TCOA4: Control AFW flow within 38 minutes following a Reactor Trip.
(SW I 0-35 identified Time Critical Operator Action).
E-2 -- A: Isolate the faulted STEAM GENERATOR before transition out of E-2.
NUREG 1021, Revision 9 Supplement 1
Appendix D Scenario Outline Form ES-D-I Event No.
1 2
3 Facility: -Prairie Island Scenario No.:
3 Op-Test No.:
1 Malf. No.
Event Type*
IA03B C (BOP)
DI-462420FF C (RO)
TS (SRO)
Examiners:
Operators:
Initial Conditions:
Power is I x ~ O - ~
amps and stable 5
Turnover:
13 Charqina Pump is 00s SG02B M (All)
Event Description Increase Power to POAH.
12 Condensate pump will be started to replace 11 for Filter Demin Support.
122 Air Compressor trips, 123 fails to auto start.
PZR Heater B/U Group 1 B Breaker Trip.
PZR pressure channel (PT-431) fails High.
12 S/G Tube Rupture (on trip).
SI signal CL System Fails.
k (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG 1021, Revision 9 Supplement 1
Scenario 3 Summary I?,
0 The crew assumes the duty with power at I x I O - ~ amps and stable. 13 Charging Pump is 00s.
The crew is directed to increase power to the POAH. During the power rise the BOP will start 12 Condensate pump and stop 11 Condensate pump.
After the crew has completed the power increase, 122 Air Compressor trips with the 123 A/C failed to auto start. The BOP will recognize that 123 Air Compressor does not automatically start and will manually start 123 Air Compressor. The crew responds IAW C47 and C34 AOPl.
After the plant is stabilized, PZR Heater B/U Group 1 B Breaker will trip. The RO will co-ordinate with the BOP to align the heater group to the alternate power supply and restore the heaters to operation IAW 1 C20.6. The SRO will determine T.S. requirements After the plant is stabilized, the controlling pressurizer pressure channel 1 P-431 fails High. The RO will respond to stabilize RCS pressure. The RO and BOP will coordinate to swap the controlling pressurizer channel to another channel IAW C47 and 1 C51.3. The SRO will determine T.S. requirements.
After the plant is stabilized, 12 S/G will suffer a Tube Rupture. The crew will trip the reactor and enter 1 E-0. The BOP will recognize the CL systems fails to respond to the SI signal and will manually align the system IAW 1 E-0 Att L. The crew will transition to 1 E-3. The scenario will be terminated upon securing SI pumps in 1 E-3.
Critical Tasks:
E TCOA4: Control AFW flow within 38 minutes following a Reactor Trip. (SWI 0-35 identified Time Critical Operator Action).
E-3 -- A: Isolate feedwater flow into and steam flow from the ruptured Steam Generator before a transition to ECA-3.1 occurs.
E 6: Establish/maintain an RCS temperature so that transition from E-3 does not occur because of the inability to maintain required subcooling or such that an extreme or severe challenge to the Subcriticality and/or the Integrity CSF occurs.
E-3 -- C: Depressurize RCS to meet SI termination criteria prior to overfilling the ruptured Steam Generator.
E-3 -- D: Terminate SI prior to overfilling the ruptured Steam Generator.
NUREG 1021, Revision 9 Supplement 1
Appendix D Scenario Outline Form ES-D-1 Event Type*
R (RO)
N (BOP/SRO)
C (All)
TS (SRO)
Facility: -Prairie Island Scenario No.:
4 Op-Test No.:
1 Event Description Reduce Power to 90% for 12 Heater Drain Pump brush checks.
At 95% the BOP remove 12 Heater Drain Pump from service.
Loss of Safeguards Bus 16.
Examiners:
Operators:
TS (SRO)
M (All)
Initial Conditions: -1 00% Power and stable Turbine Trip with ATWS.
Turnover: 12 MDAFW DumD 00s. 13 Heater Drain Pumr,
~
C (BOP)
I 11 TDAFW Fails to Auto start.
ED1 78, DI-46924T I
4-5 I TC12, RP07A.
D 1-464478, FW34A
-+
C (BOP)
I 12 S/G Pressure Instrument PT-478 Fails High, causing 12 S/G PORV to open.
TS (SRO)
I (ALL)
I Tavg Instrument Fails High.
I I
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG 1021, Revision 9 Supplement 1
R.b Scenario 4 Summary The crew assumes the duty at 100% power and stable. 11 MDAFW pump and 13 Heater Drain pump are 00s. The crew will Reduce Power to 90% to remove 12 Heater Drain pump from service for brush inspections.
After the power reduction, Loss of Safeguards Bus 16 occurs. D2 D/G starts and does not auto load. The RO will stabilize RCP seal injection flows. The BOP will restore power to Bus 16 by restoring power from CT-11 IAW 1 C20.5 AOP2. The SRO will determine T.S. requirements.
After Bus 16 is restored, 12 S/G Pressure instrument PT-478 Fails high causing 12 S/G PORV to open. The BOP will recognize the failure and manually close 12 S/G PORV. The SRO will determine T.S. requirements After the plant is stabilized, RCS Loop Thot RTD fails high causing Tavg and Delta T channels to indicate high. The RO will take manual control of charging and rod control. The BOP will coordinate with the RO to take the appropriate channel to defeat. The SRO will determine T.S.
req u i re me n ts After the plant is stabilized, a spurious turbine trip signal occurs. The turbine will trip but the reactor does not. The Crew will enter 1 E-0 and transition immediately to 1 FR-S.l. The RO will recognize the ATWS and attempt to manually trip the Reactor, but will be unsuccessful. Rods will be inserted and a boration commenced IAW 1FR-S.1. The BOP will recognize the 11 TDAFW pump failed to automatically start and will manually start the pump. The scenario will be terminated upon successful tripping of the Reactor per 1 FR-S.l Critical Tasks:
E TCOA4: Control AFW flow within 38 minutes following a Reactor Trip. (SWI 0-35 identified Time Critical Operator Action).
FR-S. 1 - C: Insert negative reactivity into the core by inserting rods or establishing emergency boration flow to the RCS during the performance of FR-S.l.
NUREG 1021, Revision 9 Supplement 1
Appendix D Scenario Out1 i ne Form ES-D-1 1
Facility: -Prairie Island Scenario No.: -SB ll Event Malf. No.
Event No.
Type*
0 p-Tes t N 0. :
RC22B C (RO)
I Examiners:
Operators:
Initial Conditions: -71 % Power and stable.
Turnover:
12 MDAFW PumD is 00s.
Breaker 16-10, Bus 26 to 16 Bus Tie Breaker. is 00s RD0909 C (RO) 3 4
5 6
- TC07B, R (RO)
TC07C C (SRO)
DI-46133C C (RO)
RD03F C (RO)
RC07A M (All) 7 RP05 C (RO)
Event Description PZR PORV 431 C leak.
Rod F-6 RPI Fails.
Two Turbine Intercept Valves fail closed.
Letdown Line Isolation Valve fails closed.
Continuous Uncontrolled Rod Withdrawal.
Containment Isolation Failure.
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG 1021, Revision 9 Supplement 1
Scenario SB Summary The crew assumes the duty at 71% power and stable. 12 MDAFW pump and Breaker 16-10 are 00s.
After the crew assumes the duty, PZR PORV PCV-431 C starts leaking by. The RO will recognize the failure and coordinate with the BOP to close the associated block valve IAW C47 and 1 C4 AOPl. The SRO will determine T.S. requirements.
After the plant is stabilized, Rod F-6 RPI fails to zero. The RO will recognize the failure and coordinate with the BOP to respond IAW C47 and 1 C5 AOP4 & AOP5. The SRO will determine T.S. requirements.
After the plant is stabilized, two Turbine Intercept valves fail closed. The BOP will recognize the failure and respond per C47 and 1 C23 AOP2 to determine a rapid downpower is required. The RO will reduce power below 50% IAW 1 C1.4 AOPl.
After the plant is stabilized, a letdown line isolation valve fails closed. The RO will recognize the failure and coordinate with the BOP to place excess letdown in service IAW 1C12.1 AOP3.
After the plant is stabilized, a continuous uncontrolled rod withdrawal occurs. The RO will recognize the failure and place Rod control in manual. Rods will continue to withdraw and the crew will respond IAW 1 C5 AOPI and trips reactor and enters 1 E-0.
After the crew trips the reactor, a large break LOCA occurs. The crew will continue in 1 E-0 and transition to E-I. The BOP will recognize automatic containment isolation signal fails to actuate and the Lead operator will ensure all containment isolation valves are closed IAW E-0 Att. L.
The crew will transition to 1 ES-1.2. The scenario will be terminated when one safeguards train has been restarted on recirc sump B.
Critical Tasks:
E-0 --0: Manually actuate Containment Isolation or close containment isolation valves such that at least one valve is closed on each penetration before the end of the scenario.
E TCOA4: Control AFW flow within 38 minutes following a Reactor Trip. (SWI 0-35 identified Time Critical Operator Action).
NUREG 1021, Revision 9 Supplement 1