ML112000654

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2011 Kewaunee Power Station, Initial Examination Outline Submittal
ML112000654
Person / Time
Site: Kewaunee Dominion icon.png
Issue date: 02/07/2011
From: Walton R
NRC/RGN-III/DRS/OLB
To:
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ML110460236 List:
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Download: ML112000654 (49)
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2011 KEWAUNEE POWER STATION INITIAL EXAMINATION ourLINE SUBMITTAL

Dominion Energy Kewaunee, Inc.

N490 .Hwy 42, Kewaunee, WI 54216 Dominiow Web Address: .www.dom.com WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 SEP 21 2010 Regional Administrator, Region III Serial No.10-534 U. S. Nuclear Regulatory Commission LlC/NW/RO 2443 Warrenville Road, Suite 210 Docket No.: 50-305 Lisle, IL 60532-4352 License No.: DPR-43 Attention: Mr. Keith Walton DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION RESPONSE TO FURNISH INTEGRATED EXAMINATION OUTLINES In response to an NRC letter dated August 31, 2010 (reference 1) regarding the administration of licensing examinations at Kewaunee Power Station, enclosed are the integrated examination outlines and the following documents as required by NUREG 1021 Revision 9, Supplement 1:

ES-201-2 Exam Outline Quality Checklist ES-201-3 Examination Security Agreement ES-301-1 Administrative Topics Outline (RO and SRO)

ES-301-2 Control Room/ln-Plant Systems Outline (RO and SRO)

ES-D-1 Scenario Outlines (4)

ES-301-5 Transient and Event Checklist ES-301-6 Competencies Checklist ES-401-2 PWR Exam Outline (RO and SRO)

ES-401-3 Generic KIA Outline ES-401-4 Record of Rejected KlAs Explanation of Random Generation Techniques and KIA Suppression Report Proposed Scenario Schedule NUREG-1021 physical security requirements state that the enclosed examination materials must be withheld from public disclosure until after the examinations are complete.

If you have questions or require additional information, please feel free to contact Mr. Dan Laing at 920-388-8691.

Very truly yours,

~~

Site Vice President, Kewaunee Power Station NOTICE: Enclosures to this letter contain confidential information. Upon separation from the enclosures. this letter is DECONTROLLED.

OCT 1 2 2010

Serial No.10-534 Page 2 of 2

Reference:

1. Letter from Hironori Peterson (NRC) to David A. Heacock (Dominion Energy Kewaunee),

"Confirmation of Initial License Examination," dated August 31,2010.

Enclosures Commitments made by this letter: NONE cc without enclosures:

NRC Senior Resident Inspector Kewaunee Power Station U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

ES-201 ExaminatiQI1 Qutline Quality Checklist Facility: Kewaunee Date of Examination: 0217-14/2010 Initials Item Task Description a b* c#

1.

W

a. Verify that the outline(s) fit(s) the appropriate model, in accordance with ES-401.

(tI// JQl,JJ) ~

b. Assess whether the outline was systematically and randomly prepared in accordance with R

I Section D.1 of ES-401 and whether all KIA categories are appropriately sampled. ~ ~ Rw.J 1II\JJ ~~

T T c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics.

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N d. Assess whether the justifications for deselected or rejected KIA statements are appropriate.

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a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number

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2. of normal evolutions, instrument and component failures, technical specifications, Jl),JI)

S and major transients.

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I b. Assess whether there are enough scenario sets (and spares) to test the projected number M and mix of applicants in accordance with the expected crew composition and rotation schedule U

L without compromising exam integrity, and ensure that each applicant can be tested using ~~

c:t'A 1rrJe at least one new or significantly modified scenariO, that no scenarios are duplicated A from the applicants' audit testes), and scenarios will not be repeated on subsequent days.

T o c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative

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3. a.

and quantitative criteria specified on form ES-301-4 and described in AppendiX D.

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

"" ..11 ~~

W distributed among the safety functions as specified on the form 1 (2) task repetition from the last two NRC examinations is within the limits specified on the form T (3) no tasks are duplicated from the applicants' audit testes) _

(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 I r-on the form ~

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 r~eated from the last two NRC licensing examinations

c. Determine if there are enough different outlines to test the projected number and mix of lilPplicants and ensure that no items are duplicated on subsequent days.
4. a. Assess whether plant-specific priorities (including PRA and IPE insights) are covered in the appropriate exam section.

G E b. Assess whether the 10 CFR 55.41/43 and 55.45 sampling is appropriate.

N c. Ensure that KIA importance ratings (except for plant-specific priorities) are at least 2.5.

E R d. Check for duplication and overlap among exam sections.

A e. Check the entire exam for balance of coverage.

L

f. Assess whether the exam fits the appropriate job level (RO or SRO).

P~nted NamE I"'" :Ife.......-7 .AI"" Date

a. Author Andrew P. Fahrenkru 1 ~. 1'.. .. ~~ -£/~

r: 09/30/2010

b. Facility Reviewer (0) Mark Goolsbev I 1TWL fJ./J.JI ,.., .., 09/3012010

~~W7~WA\.:\aJ t" c.

d.

NRC Chief Examiner (#)

NRC Supervisor 1t1....

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¢;2t o NOTE: D

  1. Independent NRC Reviewer initial items in Column *c chief examiner concurrence required.

KPS-ES-201-2-A-L-09302010-036

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Kewaunee Power Station Date of Examination: 02107-14/2011 Examination Level: RO 181 SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

RO-119*JP16A - Calculate Reactivity Plaque Conduct of Operations Information per SP-87-151 S,N RO-119-JP17A - Determine Allowable work hours Conduct of Operations S/R,N RO-119-JP121- Equipment Tagging Faulted Equipment Control S/R,D Not Selected Radiation Control N/A AO-119-JP231 - Make Initial Event Notifications Emergency Procedures/Plan S,D 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 (S 3 for ROs; S 4 for SROs & RO retakes)

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

(P)revious 2 exams (S 1; randomly selected)

KPS-ES-30 l-1-A-L-093020 10-029

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ES-301 Administrative Topics Outline Form ES-301-1 Facility: Kewaunee Power Station Date of Examination: 02/07-14/2011 Examination Level: RO D SRO [g] Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

SO-119-JP131 - Review Fuel Assembly Handling Conduct of Operations- Deviation Report S/R,D SO-119-JP011 - Review SP-87-151, Weekly Instrument Conduct of Operations Channel Checks SIR,N RO-119-JP121 - Equipment Tagging Faulted Equipment Control S/R,D SO-119-JP03B - Approve Liquid Discharge Permit prior Radiation Control to release THEN approve and disposition Liquid S/R,D,P Discharge permit after discharge SO-119-JP05E - Determine PAR Change: Wind Shift Emergency Procedures/Plan and Change in Release Status S,N 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 (S 3 for ROs; S 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (0: 1)

(P}revious 2 exams (S 1; randomly selected)

KPS-ES-30 I-I-A-L-0930201 0-029

Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Kewaunee Power Station Date of Examination: 02/07-14/2011 Exam Level: RO ~ SRO-I 0 SRO-U 0 Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / .1 PM Title Type Code* Safety Function

a. ESFAS I RO-EOO-JP012 - Failure of ICS to Automatically Initiate A,D,E,EN,S 2
b. Rod Control/Perform Control Rod Exercise RO-049-JP01 0 D,S 1
c. Pressurizer Pressure Control I RO-EOO-JP011 - Pressurizer A,D,E,S 3 Pressure Control Malfunction
d. Main Turbine Generator I RO-054-.IP061 - Rapid Power A,D,E,P,S 4 Reduction to Approximately 570 MWe
e. Containment Cooling I RO-018-JP011 - Respond to Containment A, E,EN,N,S 5 Fan Coil Unit Emergency Discharge Damper Open
f. AC Electrical Distribution I RO-039-JP02A - Transfer 4160 V AC N,S 6 Bus 1 From RAT To MAT
g. RHR I RO-034-JP04A - Respond to a Shutdown Loss of Coolant E,L,N,S 4 Accident
h. Process Radiation Monitoring I RO-045-.IP01A - Operate the D,S 7 Process Radiation Monitors (Startup R-11)

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. eves I AO-036-JP09A - Locally Isolate Dilution Flow Paths D,E,P,R 1
j. Fire (RCS) 1 RO-E07-JP01H - Remove Pressurizer PORV Fuses D,E 8
k. Diesel Generator I AO-010-JP021 - Emergency Shutdown of A,D,E 6 Diesel Generator B KPS-ES-301-2-A-L-09302010-031

@ All RO anaSRO-1 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-II SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank  :$9/:$8/S4 (E)mergency or abnormal in-plant 2:1/2:1/2:1 (EN)gineered safety feature - I - I ;::1 (control room system)

(L)ow-Power I Shutdown 2:1/2:1/2:1 (N)ew or (M)odified from bank including 1(A) 2:2/2:2/2:1 (P)revious 2 exams s 31 s 3/ s 2 (randomly selected)

(R)CA 2:1/;::1/;::1 (S)imulator KPS-ES-301-2-A-L-093020 10-031

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: Kewaunee Power Station Date of Examination: 02/07-14/2011 Exam Level: RO D SRO-I ~ SRO-U 0 Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code* Safety Function

a. ESFAS I RO-EOO-JP012 - Failure of ICS to Automatically Initiate A,D,E,EN,S 2
b. Rod Control/Perform Control Rod Exercise RO-049-JP01 D D,S 1
c. Pressurizer Pressure Control I RO-EOO-JP011 - Pressurizer A,D,E,S 3 Pressure Control Malfunction
d. Main Turbine Generator I RO-054-JP061 - Rapid Power A,D,E,P,S 4 Reduction to Approximately 570 MWe
e. Containment Cooling I RO-018-JP011 - Respond to Containment A, E,EN,N,S 5 Fan Coil Unit Emergency Discharge Damper Open
f. AC Electrical Distribution I RO-039-JP02A - Transfer 4160 V AC N,S 6 Bus 1 From RAT To MAT
g. RHR I RO-034-.IP04 - Respond to a Shutdown Loss of Coolant E,L,N,S 4 Accident I h.

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. CVCS 1AO-036-.IP09A - Locally Isolate Dilution Flow Paths D,E,P,R 1
j. Fire (RCS) I RO-E07-JP01H - Remove Pressurizer PORV Fuses D,E 8
k. Diesel Generator I AO-010-JP021 - Emergency Shutdown of A,D,E 6 Diesel Generator B KPS-ES-301-2-A-L-09302010-031

@ 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 1 SRO-II SRO-U (A)lternate path 4-61 4-612-3 (C)ontrol room (D)irect from bank s9/:$8/:$4 (E)mergency or abnormal in-plant  ;?:1/~1/~1 (EN)gineered safety feature - I - 1 ;?:1 (control room system)

(L)ow-Power 1 Shutdown ~1/~1/~1 (N)ew or (M)odified from bank including 1(A) ~2/~2/<?1 (P)revious 2 exams  :$ 3 I :$ 3/ :$ 2 (randomly selected)

(R)CA i:!:1/i:!:11;?:1 (S)imulator KPS-ES-30 1-2-A-L-093020 10-031

~ --

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Kewaunee Power Station Date of Examination: 02/07-14/2011 Exam Level: RO D SRO-I D SRO-U ~ Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code* Safety Function

a. ESFAS I RO-EOO-JP012 - Failure of les to Automatically Initiate A,D,E,EN,S 2 b.

c.

d.

e.

f.

g. RHR I RO-034-JP04 - Respond to a Shutdown Loss of Coolant E,L,N,S 4 Accident h.

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. CVCS 1 AO-036-JP09A - Locally Isolate Dilution Flow Paths D,E,P,R 1
j. Fire (RCS) I RO-E07-JP01H - Remove Pressurizer PORV Fuses D,E 8
k. Diesel Generator I AO-010-JP021 - Emergency Shutdown of A,D,E 6 Diesel Generator B

@ 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/ SRO-U (A)lternate path 4-6 /4-61 2-3 (C)ontrol room (D)irect from bank S9/s8/S4 (E)mergency or abnormal in-plant =:::1/=:::1/=:::1 (EN)gineered safety feature 1 =:::1 (control room system)

(L)ow-Power 1 Shutdown =:::1/=:::1/2:1 (N)ew or (M)odified from bank including 1(A) =:::2/=:::2/2:1 (P)revious 2 exams s 3 1 s 3 1s 2 (randomly selected)

(R)CA  ;::1/;::1/2:1 (S)imulator KPS-ES-301-2-A-L-09302010-031

ES..301'tiansient and Event Checklist Form ES-301-5 Facility: Kewaunee Power Station Date of Exam: 2/7-1412010 Operating Test Number: Crews A, 8, & C A E Scenarios P V T 1 2 3 4 M P E L N 0 I T CREW CREW CREW CREW T N I

POSITION POSITION POSITION POSITION A I C

A T S A 8 S A 8 S A 8 S A 8 L M N Y R T 0 R T 0 R T 0 R T 0 U T P 0 C P 0 C P 0 C P 0 C P M(*)

E R I U RO RX - 1 1 1 1 0

[8J NOR SRO-I 1 - 1 1 1 1 0 I/C 3,6 3,5 I4 I4 4 2 SRO-U MAJ 4 2 2 2 1 0 TS - - 0 0 2 2 RO RX 1 1 2 1 1 0 0 NOR - 1 1 1 1 1 SRO-I

[8J IIC 2,4 2,3,5 5 4 4 2 SRO-U MAJ 5 4 2 2 2 1 0 TS - 2,3 2 0 2 2 RO RX 1 - 1 1 1 0 0 NOR 1 1 2 1 1 1 SRO-I IIC 0 2,3,4,6 2 5 4 4 2 SRO-U MAJ 5 4 2 2 2 1

[8J TS 2,4 - 2 0 2 2 RO RX 1 1 0 0 NOR 1 1 1 SRO-I 0 IIC 4 4 2 SRO-U MAJ 2 2 1 0 TS 0 2 2 Instructions:

1. Check the afElicant level and enter the operating test number and Form ES-D-1 event numbers for each event tYfae; are not a~licable for RO applicants. ROs must serve in both the "at-the-controls (ATC)"

and "ba ance-of-plant (8 P)" positions' Instant SROs must do one scenario, including at least two instrument or component (I/C) malfunctions and one major tranSient, in the ATC pOSition.

2. Reactivi~ manipulations may be conducted under normal or controlled abnormal conditions (refer to Section .5.d) but must be Significant per Section C,2.a of Appendix D. P Reactivity and normal evolutions may be replaced With additional instrument or component mal unctions on a 1-for-1 basis.
3. Whenever practical both instrument and component malfunctions should be included; only those that require verifiable actions that provide inSight to the applicant's com~etence count toward the mimmum requirements specified for the applicant's license level in t e right-hand columns.

ES-301. Page 26 of 27 KPS-ES-30 1-5-CREWABC-A-L-0930201 0-033

ES-301 Transleni and Event ~tiecKlistForm ES-~Q14.

Facility: Kewaunee Power Station Date of Exam: 2/7-1412010 Operating Test Number: Crew D A E Scenarios P V P E T M 1 2 3 4 I L N 0 N I T CREW CREW CREW CREW T I C POSITION POSITION POSITION POSITION A M A T S A B S A B S A B S A B L U N Y R T 0 R T 0 R T 0 R T 0 M(*)

T P 0 C P 0 C P 0 C P 0 C P R I U E

RO RX

- 1 - 1 1 1 0 t8J NOR 1 - 1 2 1 1 1 SRO-I I/C 0 3,6 3,5 3 5 4 4 2 SRO-U MAJ 5 4 3 3 2 2 1 0 TS - 0 0 2 2 RO RX 1 1 3 1 1 0 0 NOR 1 1 2 1 1 1 SRO-I -

I/C t8J 2,4 2,3,5 2,4,5 8 4 4 2 SRO-U MAJ 5 4 3 3 2 2 1 0 TS - 2,3 1 3 0 2 2 RO RX 1 - 1 2 1 1 0 0 NOR 1 1 2 1 1 1 SRO-I IIC t8J 2,3.4,6 2 2,5 7 4 4 2 SRO-U MAJ 5 4 3 3 2 2 1 D TS 2,4 - 2 0 2 2 RO RX 1 1 0 0 NOR 1 1 1 SRO-I I/C 0 4 4 2 SRO-U MAJ 2 2 1 0 TS 0 2 2 Instructions:

1. Check the afElicant level and enter the operating test number and Form ES-D-1 event numbers for each event tyli:,; are not a~licable for RO applicants. ROs must serve in both the "at-the-controls (ATC)"

and "ba ance-of-plant (8 P)" positions; Instant SROs must do one scenario, including at least two instrument or component (IIC) malfunctions and one major transient, in the ATC position.

Reactivi~ manipulations may be conducted under normal or control/ed abnormal conditions (refer to 2.

Section .5.d) but must be Significant per Section C.2.a of Appendix D. P Reactivity and normal evolutions may be replaced With additional instrument or component mal unctions on a 1-for-1 basis.

3. Whenever practical both instrument and component malfunctions should be included; only those that r~uire verifiable actions that provide insight to the applicant's com~etence count toward the mimmum requirements specified for the applicant's license level in f e right-hand columns.

ES-301, Page 26 of 27 KPS-ES-301-5-A-L-CREWD-D-09302010-032

ES-301 Transient and Event Checklist Form ES-301::S Facility: Kewaunee Power Station Date of Exam: 2/7-1412010 Operating Test Number: 1 A E Scenarios P V P E 1 2 3 4 IT M L N 0 I I T CREW CREW CREW CREW T N C POSITION POSITION POSITION POSITION A I A T S A B S A B S A B S A B L M N Y R T 0 R T 0 R T 0 R T 0 U T P 0 C P 0 C P 0 C P 0 C P M(*)

E R I U RO RX - 1 1 0 0 NOR 1 1 1 1 SRO-I 0 I/C 2,5 4 4 2 SRO-U MAJ 5 2 2 1 0

TS - 0 2 2 RO RX 1 1 1 0 0

SRO-I NOR - 1 1 1 0 IIC 3,4 4 4 2 SRO-U MAJ 2 5 2 1 0

TS - 0 2 2 RO RX 1 1 1 0 0 NOR 1 1 1 1 SRO-I i

0 I/C 2,3,4 4 4 2 SRO-U MAJ 2 2 1 5

0 TS 2,3,4 0 2 2 RO RX 1 1 0 0 NOR 1 1 1 SRO-I 0 I/C 4 4 2 SRO-U MAJ 2 2 1 0

TS 0 2 2 Instructions:

1. 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-ot-plant (BOP)" positions; Instant SROs must do one scenario, including at least two instrument or component (IIC) malfunctions and one major transient, in the ATC position.

2. 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-1 basis.
3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicant's competence count toward the minimum requirements specified for the applicant's license level in the right-hand columns.

ES-301, Page 26 of 27 KPS-ES-301-5-SPARE-A-L-09302010-034

~----~--~-------------~------------------------------

ES-301 Competencies Checklist Form ES-301-6 Facility: Kewaunee Date of Examination: 2n-14/2010 Operating Test No.: 1 APPLICANTS

[&1 D D Competencies RO SRO-I SRO-U D D

RO SRO-I SRO-U D

[&1 RO SRO-I SRO-U [&1 D

RO SRO-I SRO-U D 8

i SCENARIO SCENARIO SCENARIO SCENARIO 1 2 3 4 1 2 3 4 1 2 3 4 - - -

Interpret! Diagnose 2*5 20S 1*5 1-5 2-5 20S 1*5 1-5 2-5 20S 1*5 1-5 Events and Conditions Comply With and Use 1-5 loS 1-5 15 1*5 loS 1-5 15 1-5 loS 1-5 15 Procedures (1)

Operate Control 1-5 loS 1-5 1-5 1-5 loS 1-5 1-5 1-5 loS 1-5 1-5 Boards (2)

Communicate 1-5 loS 1-5 1-5 1-5 10S 1-5 1-5 1-5 loS 1-5 1-5 and Interact Demonstrate 1-5 loS 1-5 1-5 10S 1-7 1-5 1-5 Supervisory Ability (3)

Comply With and 2,4 2,3 1 2,3,4 2,4 2,3 1 2,3,4 Use Tech Specs. (3)

Notes:

(1 ) Includes Technical Specification compliance for an RO.

(2) Optional for an SRO-U.

(3) Only applicable to SROs.

Instructions:

Check the applicant's license type and enter one or more event numbers that will allow the examiners to evaluate every applicable competency for every applicant.

ES-30 1, Page 27 of 27 KPS-ES-301-6-A-L-09302010-030

ES-401 PWR Examination Outline Form ES-401-2 ty: Kewaunee Date of Exam: 2/07-14/2011 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G 1 2 3 4 5:6 1 2 3 4 . Total A2 G* Total

1. 1 3 3 3 3 3 3 18 3 3 6 Emergency

& Abnormal 2 1 1 1 N/A 2 2 N/A :2 9 2 2 4 Plant Evolutions Tier Totals 4 4 4 5 5 27 10 1 3 3 3 3 2 2 3 2 3 2 2 28 5 2.

Plant 2 1 0 1 1 1 1 1 1 1 1 10 3 Systems Tier Totals 4 3 4 4 3 3 4 3 4 3 3 :38 8

3. Generic Knowledge and Abilities 1 2 Categories 7

3 2 Note: 1. Ensure that at least two topics from every applica and SRO outlines (i.e. except for one category in in each KIA category shall not be less than two).

2. The point total for each group and tier in the at specified in the table.

The final pOint total for each group and tier from t ifled in the table based on NRC revisions. The finalll)R~.~l9: and the SRO-only exam must total 25 points.

3. Systems/evolutions within each ed outline; systems or evolutions that do not apply at the facility should be deleted nt, site-specific systems that are not included on the outline should be added. Att,~".hmAlnt 2, for guidance regarding the elimination of inappropriate KIA
4. a!1ari.e.Vi~M~~Y~$ possible; sample every system or evolution in the group OUI.ystelm or evolution.
5. KlAs having an importance rating (IR) of 2.5 or higher shall be selected.

and SRO-only portions, respectively.

1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics applicable evolution or system.

8. ages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IRs) license level, and the point totals (#) for each system and category. Enter the group and tier totals ry in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-o/lly exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Fonm ES401-3. Limit SRO selections to KlAs that are linked to 10 CFR 55.43.

ES-401, Page 21 of 33 KPS-ES-401-2/3-A-L-09302010-028

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO I-SRG)

ElAPE # I Name 1 Safety Function KKK KIA Topic(s) IR #

123 1.07 - Ability to operate and/or monitor the 000007 (8!JINE02&E10; CElE(2) Reactor following as they apply to a reactor trip:

4.3 Trip - Stabilization - Recovery 11 MT/G trip; verification that the MT/G has been tripped 2.03 - Ability to determine and interpret the following as they appl)';tothe 000008 Pressurizer Vapor Space Pressurizer Vapor Space Accident:

Accident/3 control board, valve conlmJJ, and indicators 000009 Small Break LOCA 13 000011 Large Break lOCA I 3 x 4.1 1 000046117 RCP Malfunctions 14 x 2.8 of the reasons for the responses as they apply to the Reactor Coolant Pump Makeup: 3.5 000022 Loss of Rx Coolant Makeup I 2 contained in SOPs and EOPs for Ps, loss of makeup, loss of charging, d abnormal charging 1.22

  • Ability to operate andlor monitor the following as they apply to the Loss of 000025 Loss of RHR Residual Heat Removal System: 2.9 Obtaining of water from BWST for LPI system 2.04 - Ability to determine and interpret the following as they apply to the loss of Component Cooling Water: The normal 2.5 values and upper limits for the temperatures of the components cooled byCCW 000027 Pressurizer Pre'~~ttf';1p~tltrol G 2.2.22 - Knowledge of limiting 4.0 System Malfunction I 3 conditions for operations and safety limits.

1.01 - Knowledge of the operational implications of the following concepts as 000029 ATWS 11 x they apply to the ATWS: Reactor 2.8 nucleonics and thermo-hydraulics behavior 000038 Steam Gen. Tube Rupture I 3 G 2.4.1 - Knowledge of EOP entry 4.6 conditions and immediate action steps.

ES-401, Page 22 of 33 KPS-ES-40 1-2/3-A-L-093020 10-028

ES-401 2 Form ES-401-2 000040 (BWIE05; CE/E05; W/Et2)

Steam Line Rupture - Exce$sive Heat Transfer 14 3.05 - Knowledge of the reasons for the 000054 (eli/EOE) Loss of Main following responses as they apply to the Feedwater 14 x Loss of Main Feedwater (MFW): Actions 4.4 contained in EOPs for loss of MFW 1.06 - Ability to operate andlor monitor the following as they apply to a Station 000055 Station Blackout 16 4.1 Blackout: Restoration of power Y\lith one ED/G ...

000056 Loss of Off-site Power I 6 000057 Loss of Vital AC Inst. Bus 16 000058 Loss of DC Power I 6 000062 Loss of Nuclear Svc Water I 4 2.9 000065 Loss of Instrument Air I 8 x 2.9 of the operational of the following concepts as WIE04 LOCA Outside Containment 1 3 to the lOCA Outside 3.5 Containment: Components, capacity, and function of emergency systems WlE11 Loss of Recirc./4 2.2 - Knowledge of the interrelations between the loss of Secondary Heat Sink and the following: Facility's heat removal systems, including primary coolant, 3.9 emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility 2.04 - Knowledge of the interrelations between the Generator Voltage and 3.0 Electric Grid Disturbances and the following: Controllers, positioners KIA Category Totals: 333 Group Point Total: 181@

ES-401, Page 22 of 33 KPS-ES-401-2/3-A-L-09302010-028

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions Tier lIGroup 2 (RO I-SRG)

E/APE # I Name I Safety Function KKK KIA Topic(s) IR #

1 2 3 000001 Continuous Rod Withdrawal I 1 000003 Dropped Control Rod 11 000005 Inoperable/Stuck Control Rod f 1 1.13 - Ability to operate ancl/Qr*mo,*

000024 Emergency Boration , 1 following as they apply Soration: Soric acid 000028 Pressurizer Level Malfunction I 2 000032 Loss of Source Range NI/7 000033 Loss of Intermediate Range NI/7 4.2 000036 (BW/A08) Fuel Handling Accident 18 000037 Steam Generator Tube Leak I 3 3.B 000051 Loss of Condenser Vacuum 14 000059 Accidentel Liquid RadWaste ReI. 19 000060 Accidental Gaseous Radwaste Ret. I 9 1.01 - Knowledge of the operational implications of the following concepts as 000061 ARM System ~ 17 they apply to Area Radiation Monitoring 2.5 (ARM) System Alarms: Detector limitations 2.02 - Knowledge of the interrelations between the Control Room Evacuation 3.7 and the following: Reactor trip system (Not Selected) 3.07 - Knowledge of the reasons for the following responses as they apply to the x Inadequate Core Cooling: Starting up 4.0 emergency feedwater and RCPs 1.04 - Ability to operate and/or monitor the following as they apply to the High 000076 High Reactor Coolant Activity I 9 3.2 Reactor Coolant Activity: Failed fuel-monitoring equipment W/E01 & E02 Rediagnosis & SI Termination 13 WIE13 Steam Generator Over-pressure I 4 W/E15 Containment Flooding' 5 (Not Selected)

WIEle High Containment Radiation 19 (Not Selected)

ES-401, Page 23 of 33 KPS-ES-40 1-2/3-A-L-0930201 0-028

ES-401 3 Form ES-401*2 BW/A01 Plant Runback 11 (Not Applicable to plant)

BW/A02&A03 Loss of NNI-X/Y 17 (Not Applicable to plant)

BW/A04 Turbine Trip I 4 (Not Applicable to plant)

BW/A05 Emergency Diesel Actuation 1 6 (Not Applicable to plant)

BW/A07 Flooding I 8 (Not Selected)

BW/E03 Inadequate Subcooling Margin 14 (Not Applicable to plant) 8WIeo8r WlE03 LOCA COoldown - Depress. 14 BWIE09; CElA13; W/E09&E10 Natural Circ./4 BW/E13&E14 EOP Rules and Enclosures CEIM4; WlE08 RCS Overcooling - PTS 14 CElA16 Excess RCS Leakage 12 CElE09 Functional Recovery KIA Category Totals: 914 ES-401, Page 23 of 33 KPS-ES-40 1-2/3-A-L-093020 10-028

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 21Group 1 (RO t-SRO)

System # I Name KIA Topic(s) IR #

5.01 - Knowledge ofthe operational implications of the following concepts as they apply to the RCPS: The relationship between the RCPS flow 003 Reactor Coolant Pump rate and the nuclear reactor core 3.3 operating parameters power tilt, imbalance, DNB p~r density, difference -hot pressure) 004 Chemical and Volume Control 005 Residual Heat Removal x 2.5 3.6 006 Emergency Core Cooling - Ability to predict and/or monitor 2 in parameters (to prevent design limits) associated with operating the ECCS controls 3.3 including: Pressure, high and low 2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS and (b) based 007 Pressurizer on those predictions, use procedures to 3.9 Tank correct, control, or mitigate the consequences of those malfunctions or operations: Stuck-open PORV or code safety 2.02 - Knowledge of bus power supplies to the following: CCW pump, including emergency backup 3.0 008 Component 2 3.01 - Ability to monitor automatic operation of the CCWS, including:

Setpoints on instrument signal levels for 3.2 normal operations, warnings, and trips that are applicable to the CCWS 4.02 - Ability to manually operate and/or 010 Pressurizer Pressure monitor in the control room: PZR 3.7 Control heaters ES-401, Page 24 of 33 KPS-ES-401-2/3-A-L-09302010-028

ES-401 4 Form ES401*2 2.02

  • Ability to (a) predict the impacts of the following malfunctions or operations on the RPS and (b) based 3.6 on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of Instrument Power 2 012 Reactor Protection G 2.4.49 - Ability to perform without reference to procedures those actions 4.6 that require immediate of system components 013 Engineered Safety Features Actuation x 022 Containment Cooling 2 that a CCSwili have Containment 3.0 readings 025 Ice Condenser Knowledge of the effect that a 026 Containment Spray or malfunction of the CSS will have 3.9 on the following: CCS 3.05 - Knowledge of the effect that a loss or malfunction of the MRSS will 3.6 have on the following: RCS 039 Main and Reheat 2 4.07 - Knowledge of MRSS design feature(s) and/or interlock(s) which provide for the following: Reactor 3.4 building isolation 4.19* Knowledge of MFW System design feature(s) and/or interlock(s) x 3.2 which provide for the following:

Automatic feedwater isolation of MFW 6.01

  • Knowledge of the effect of a loss 061 Auxiliary/Emergency or malfunction of the following will have 2.5 Feedwater on the AFW System components:

Controllers and positioners 1.01

  • Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated 3.4 062 AC Electrical Distribution with operating the A.C. Distribution System controls including: Significance of DIG load limits ES-401, Page 24 of 33 KPS-ES-40 1-2/3-A-L-093020 10-028

ES-401 4 Form ES-401-2 4.01 - Knowledge of D.C. Electrical System design feature(s) and/or interlock(s) which provide for the following: ManuaVautomatic transfers of 2.7 control 063 DC Electrical Distribution x 2 1.01 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the D.C. Electrical 2.5 System controls capacity as it is rate 064 Emergency Diesel Generator 073 Process Radiation Monitoring 3.7 recognize abnormal operating 076 Service Water are entry-level 4.5 for emergency and abnormal procedures.

1.05 - Knowledge of the physical connections and/or cause-effect 3.4 relationships between the lAS and the following systems: MSIVair 078 Instrument Air 2 3.01 - Ability to monitor automatic operation of the lAS, including: Air 3.1 pressure 1.05 - Knowledge of the physical connections and/or cause-effect relationships between the Containment 2.8 System and the following systems:

Personnel access hatch and emergency access hatch KIA Category Totals: roup Point Total: 2815 ES-401, Page 24 of 33 KPS-ES-401-2/3-A-L-09302010-028

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 21Group 2 (RO t-SRO)

System # I Name KIA Topic(s) IR #

001 Control Rod Drive 4.07 - Knowledge of RCS design feature(s) and/or interlock(s) which 002 Reactor Coolant provide for the Contraction 3.1 and expansion during cooldown 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor x 2.7 027 Containment Iodine Removal 1.01 - Knowledge of the physical connections and/or cause-effect 028 Hydrogen Recombiner and relationships between the HRPS and 2.5 Purge Control the following systems: Containment annulus ventilation system (including pressure limits) 2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System and (b) based on 2.7 those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of SFPCS 1.02 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated 034 Fuel Handling Equipment 2.9 with operating the Fuel Handling System controls including: Water level in the refueling canal 035 Steam Generator (Not Selected) 041 Steam DumplTurbine Bypass (Not Selected)

Control ES-401, Page 25 of 33 KPS-ES-401-2/3-A-L-09302010-028

ES-401 5 Form ES-401-2 3.11

  • Ability to monitor automatic 045 Main Turbine Generator operation of the MT/G System, 2.6 including: Generator trip 055 Condenser Air Removal (Not Selected) 056 Condensate 066 Liquid Radwaste 071 Waste Gas Disposal 3.2 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 066 Fire Protection 2.7 KIA Category Totals: Group Point Total: 1013 ES-401, Page 25 of 33 KPS-ES-401-2/3-A-L-09302010-028

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 tRG-I SRO)

ElAPE # 1 Name I Safety Function KKK KIA Topic(s) IR #

123 000007 (BW/E02&E10; CElE02) Reactor (Not Selected)

Trip - Stabilization - Recovery 11 000008 Pressurizer Vapor Space Accident 1 3 (Not Selected) 000009 Small Break LOCA 13 000011 Large Break LOCA I 3 000015/17 RCP Malfunctions 14 000022 Loss of Rx Coolant Makeup 1 2 000025 Loss of RHR System 14 Pressurizer Pressure Control Malfunction 1 3 000029 ATWS 11 000038 Steam Gen. Tube Rupture 1 3 4.8 000040 (B'JIJlIi06; CIiI~06; V1Jl1i12)

Steam Line Rupture - Excessive Heat 4.7 Transfer 14 000054 (CElE06) Loss of Main Feedwater 1 4 000055 Station Black(l~1 G 2.4.6 - Knowledge of EOP mitigation 4.7 strategies.

2.07 - Ability to determine and interpret the following as they apply to the Loss of 4.1 Vital AC Instrument Bus: That a loss of ac has occurred G 2.1.20 Ability to interpret and execute 4.6 procedure steps.

(Not Selected)

WlE04 LOCA Outside Containment 13 (Not Selected) 2.2 - Ability to determine and interpret the following as they apply to the Loss of W/E11 Loss of Emergency Coolant Emergency Coolant Recirculation:

Recirc./4 4.2 Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments BW/E04; WlE05 .....".,,,,,..,.,

Transfer - Loss Sink/4 ES-401, Page 22 of 33 KPS-ES-401-2/3-A-L-09302010-028

ES-401 2 Form ES-401-2 000077 Generator Voltage and Electric Grid (Not Selected)

Disturbances I 6 KIA Category Totals: Group Point Total:

ES-401, Page 22 of 33 KPS-ES-40 1-2/3-A-L-0930201 0-028

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RQ.,I SRO)

ElAPE # I Name I Safety Function KKK KIA Topic(s) IR #

1 2 3 000001 Continuous Rod Withdrawal I 1 000003 Dropped Control Rod f 1 000005 lnoperablefStuck Control Rod f 1 000024 Emergency Boration f 1 000028 Pressurizer Level Malfunction f 2 000032 Loss of Source Range NI f 7 000033 Loss of Intermediate Range NI/7 000036 (BWfA08) Fuel Handling Accident f 8 000037 Steam Generator Tube Leak 13 000051 Loss of Condenser Vacuum f 4 000059 Accidental Liquid RadWaste ReI. f 9 000060 Accidental Gaseous Radwaste ReI. f 9 000061 ARM System Alarms f 7 000067 Plant Fire On-site f 8 000068 ~ Control Room Evac. f 8 000069 (WfE14) Loss ofCTMT Integrity f 5 000076 High React 2.02 - Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity: Corrective 3.4 actions required for high fISsion product activity in RCS (Not Selected)

G 2.4.21 - Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor 4.6 coolant system integrity, containment conditions, radioactivity release control, etc.

2.02 - Ability to determine and interpret the following as they apply to the WfE15 Containment Flooding /5 Containment Flooding: Adherence to 3.3 appropriate procedures and operation within the limitations in the facility's license and amendments WfE16 High Containment Radiation f 9 BW/A01 Plant Runback 11 (Not Applicable to plant)

BW/A02&A03 Loss of NNI-XN 17 (Not Applicable to plant)

BW/A04 Turbine Trip f 4 (Not Applicable to plant)

ES-401, Page 23 of 33 KPS-ES-40 1*2/3-A-L-093020 10*028

ES-401 3 Form ES-401-2 BW/A05 Emergency Diesel Actuation I 6 (Not Applicable to plant)

BW/A07 Flooding I 8 (Not Selected)

BW/E03 Inadequate Subcooling Margin 14 (Not Applicable to plant)

~ W/E03 LOCA Cooldown

  • Depress. 14 (Not Selected)

BW/E09; CElA13; W/E09&E10 Natural Circ./4 (Not Selected)

BW/E13&E14 EOP Rules and Enclosures (Not Applicable to plant)

Ge/M4f W/E08 RCS Overcooling

  • PTS 14 (Not Selected)

CElA16 Excess RCS Leakage 12 CElE09 Functional Recovery KIA Category Totals:

ES-401, Page 23 of 33 KPS-ES-401-2/3-A-L-09302010-028

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 21Group 1 {RO+-SRO)

System # / Name KIA Topic(s) IR #

003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection 3.2

- Ability to (a) predict the impacts follOwing malfunctions or on the ESFAS and (b) based 013 Engineered Safety Features on those predictions, use procedures to 4.8 Actuation correct, control, or mitigate the consequences of those malfunctions or operations: LOCA 022 Containment 025 Ice Condenser (Not Applicable to plant) 2.08 - Ability to (a) predict the impacts of the following malfunctions or operations on the CSS and (b) based on those predictions, use procedures to 026 correct, control, or mitigate the 3.7 consequences of those malfunctions or operations: Safe securing of containment spray (when it can be done) 059 Main Feeclwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution (Not Selected) 063 DC Electrical Distribution (Not Selected)

ES-401, Page 24 of 33 KPS-ES-401-2/3-A-L-09302010-028

ES-401 4 Form ES-401-2 G1.1.7 - Ability to evaluate plant performance and make operational 064 Emergency Diesel judgments based on operating 4.7 Generator characteristics, reactor behavior, and instrument interpretation.

Radiation 076 Service Water 078 Instrument Air 103 Containment KIA Category Totals:

~5 ES-401, Page 24 of 33 KPS-ES-401-2/3-A-L-093020 10-028

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 21Group 2 (RG I SRO)

System # I Name KIA Topic(s) IR #

001 Control Rod Drive (Not Selected) 002 Reactor Coolant (Not Selected) 011 Pressurizer Level Control (Not Selected) 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment (Not Selected) 2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the SDS and (b) based 041 Steam DumplTurbine on those predictions, use procedures 3.9 Control to correct, control, or mitigate the consequences of those malfunctions or operations: Steam valve stuck open 056 Condensate 2.1.32 - Ability to explain and apply system limits and precautions. 4.0 068 Liquid Radwaste (Not Selected) 071 Waste Gas Disposal (Not Selected) 072 Area Radiation MonitOring (Not Selected) 075 Circulating Water (Not Selected) 079 Station Air 086 Fire Protection ES-401, Page 25 of 33 KPS-ES-401-2/3-A-L-09302010-028

ES-401 5 Form ES-401-2 KIA Category Totals: roup Point Total:

ES-401, Page 25 of 33 I KPS-ES-401-2/3-A-L-09302010-028

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Category KIA # Topic RO IR # IR #

2.1.23 4.3 1 1.

Conduct 2.1.39 1 of Operations 2.1.30 Ability to locate and operate components. including local 1 controls.

2.1.40 2.1.20 2.1.

SUbtotal 2.2.40 2.2.17 2.

Equipment Control 2.2.36 4.2 1 2.2.43 3.3 1 2

3.

Radiation 3.2 1 Control 3.2 1 1

ES-401, Page 25 of 33 KPS-ES-401-2/3-A-L-09302010-028

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Knowledge of RO responsibilities in emergency plan 3.9 4.

Emergency Ability to prioritize and interpret the significance of each 4.1 1 Procedures I Plan 2.4.11 4.2 1 2.4.27 of "fire in the 3.9 2.4.

2.4.

Subtotal 2 Tier 3 Point Total 7 ES-401, Page 25 of 33 KPS-ES-40 1-2/3 -A-L-093020 10-028

Appendix D Scenario Outline Form ES-D-1 Facility: Kewaunee Power Station Scenario No.: 1. Op-Test No.:-.1 Examiners: Operators: SRO:

ATC:

BOP:

Initial Conditions: 100% 120wer EOL Eguilibrium Xenon. RCS boron concentration is 77 I2l2m. RCS Tave is 572°F. Generator load is 602 MWe gross, AS-31/AS-35 R-11 & R-12 Sam(2le Return Aligned to Containment, PR-1A PRZR PORV Block Valve is closed with (2ower maintained due to PR-2A seat leakage, TO AFW (2um(2 is OOS for corrective maintenance on its Aux Lube Oil Pum(2, 2 inch containment vent is in I2rogress (2er NOP-RBV-002 section 5.6 Turnover: Notified b~ DEMI that MISO has escalated a (2revious Minimum Generation Alert to a Minimum Generation Warning with an actual event, KPS has agreed to Lower Power to 95%.

TS 3.7.5 (AFW s:istem} Condition B with one AFW Train ino(2erable. Reguired Action B.1 is to restore train(s} to OPERABLE status with a Coml2letion Time of 72 Hours. Start Time 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before scenario start time.

TS 3.4.11 (Reactor Coolant S~stem Pressurizer Power O(2erated Relief Valves} Condition A One or more PORVs ino(2erable and ca(2able of being manuall:i c:icled. Reguired Action A.1 is to close and maintain (2ower to the associated PORV with a com(2letion time of one hour (Coml2leted}.

Event Malf. Event Event No. No. Type* Description Pre- PR-2A CAUTION tagged CLOSED due to seat leakage. PR-1A Load closed TS LC03.4.11, ACTION A.1.

D046113-G Pre- OFF TID AFW Pump OOS. MS-102 in PULLOUT.

DO-46114-G N/A Load OFF MS-100AlB in CLOSE and lights OFF Pre- FW16A BOP-C AFW Pump A fails to auto start. Manual start remains available.

Load ATC-R 1 N/A Power reduction required due to Minimum Generation Warning.

BOP-N After approximately 5% load decrease, controlling PRZR pressure RX201 2 25000:30 ATC-I blue channel (III), PT-431 fails high.

Heaters de-energize and sprays open to lower PRZR pressure.

FW-7B controller output signal fluctuates resulting in unstable RX02B 3 7515:00 BOP-C operation of FW-7B, SG B Feed Reg Valve. Value for fluctuation increases to 75% over 15 minutes.

N105A 4 1.2 ATC-I NI Red Channel, N41, fails low.

SG01B ATC, SGTR occurs in SG B ramping to maximum value over a 5-minute 5.6 ramp BOP 5 over 5 period. Crew responds to radiation alarms and rising SG level.

M minutes When the reactor trips, the SGTR goes to its maximum value, AFW Pump A fails to auto start on low SG level or SI signal. BOP 6 FW16A BOP-C establishes flow to SG A.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 5 KPS-ES-O-1-Scenario1-1-A-L-0930201 0-040

SCENARIO 1.:1 OVERVIEW

~-~EVENT DI;SCR1PTlON Power reduction. Crew is directed during turnover to perform a power reduction to reduce thermal power to 1673 MW (- 95% RTP) due to the Minimum Generation Warning with an actual event. Crew should review 1 Standard Reactivity Plan and set up for 5% power reduction at % %/min.

KW-GOP-307, Hold-At Power Greater Than 35%, will be used to direct the power reduction. The ATC operator will borate in accordance with the Reactivity Plan using NOP-CVC-001, and the BOP operator will reduce turbine load using NOP-TB-001.

After the power decrease the controlling PRZR Pressure blue channel, PT-431, fails high to 2500 psig. This will result in PRZR heater output going to zero, and PRZR Spray valves opening. Actual PRZR pressure will lower. The crew will perform actions of AOP-GEN-001, Immediate Operator Actions, Attachment H for Pressurizer Spray Valves Open and/or the ARP(s) associated with High Pressure alarms. The failed instrument will be addressed using AOP-MISC-001, Response to Instrument Failure, Attachment G Pressurizer Pressure.

Technical Specifications:

Condition A One or more Functions with one or more required channels or trains inoperable. Required Action A.1 enter the condition referenced in Table 3.3.1-1 for the channel(s) or train(s) with a completion time of immediately.

Condition E one channel inoperable with Required Action E.1 to 2 place channel in trip with a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Table 3.3.1-'1 Item 6. Overtemperature LlT (Loop B Chan 3 OTLlT) & item B.b Pressurizer Pressure High.

Condition K one channel inoperable with Required Action K.1 to place channel in trip with a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Table 3.3.1-1 Item B.a Pressurizer Pressure Low

Instrumentation Condition A One or more Functions with one or more required channels or trains inoperable. Required Action A.1 enter the condition referenced in Table 3.3.2-1 for the channel(s) or train(s) with a completion time of immediately.

Instrumentation Condition D one channel inoperable with Required Action D.1 to place channel in trip with a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Table 3.3.2-1 Item 1.d Pressurizer Pressure - Low Note: The option in TS for shutting down the plant is not expected to be exercised for this failure The Unit supervisor will direct tripping of bistables. It is expected to trip bistables for the failed channel(s) during the scenario.

Page 2 of 5 KPS-ES-D-1-Scenario 1-1-A-L-0930201 0-040

Following the addressing of the failed PRZR pressure channel, FW-7B, SG B Feed Reg Valve, will experience oscillation of the output signal from

_ _+-wi'+",

ii-"'Gc;""'m""tro1Ier in AUTO. This wiU-result in *fluctyation in SG B level,and feed

{ft)w. The BOP operator is expected to respond to the changes in SG B 1evel and/or associated alarms. The crew will respond in accordance with AOP-GEN-001, Immediate Operator Actions, Attachment B, Abnormal 3 Steam Generator Level, or the ARPs for SG level or steam/feed flow. The operator will transfer FW-7B controller to MAN and establish "normal" level in SG B (33% to 50%). The crew will also enter AOP-FW-001, Abnormal Feedwater System Operation, which also contains direction for maintaining SG level with the FW-7B controller in manual.

If the operator fails to control SG level and a reactor rip signal is generated, then Event 5, SGTR, will be initiated at its final value.

When FW-7B failure has been addressed, Power Range Nuclear Channel N41 (Red) will fail low. This will result in a rapid change in NI power rate and result in control rods in AUTO stepping OUT. The A TC operator will identify rod movement and determine rod movement is NOT required by the plant condition (stable). Actions of AOP-GEN-001, Immediate Operator Actions, Attachment C Uncontrolled Rod Motion will be performed. Once it is determined that a turbine runback or rapid power reduction is NOT in progress, the ATC operator will place the Rod Bank Selector to MAN and verify rod motion stops. The crew will check for instrument failures and determine N41 channel has failed low. The failed instrument will be addressed using AOP-MISC-001, Response to Instrument Failure, Attachment J Nuclear Power Range. The crew will remove N41 from service. Tave-Tref should be restored to within +/- 1°F using rod control in manual.

  • TS 3.3.1 {Reactor Protection System (RPS) Instrumentation}

Condition A One or more Functions with one or more required channels or trains inoperable. Required Action A.1 enter the condition referenced in Table 3.3.1-1 for the channel{s) or train(s) with a completion time of immediately.

4 Either of the following for Table 3.3.1-1 Item 2.a Neutron Flux High

  • TS 3.3.1 {Reactor Protection System (RPS) Instrumentation}

Condition 0 one channel inoperable with Required Action 0.1.1 to place channel in trip with a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND Required Action 0.1.2 reduce Thermal Power to .::: 75% RTP with a completion time of 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />. Table 3.3.1-1 Item 2.a PR Neutron Flux High OR

Condition 0 one channel inoperable with Required Action D.2.1 to place channel in trip with a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND Required Action 0.2.2 perform SR 3.2.4.2 with a completion time of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Table 3.3.1-1 Item 2.a PR Neutron Flux High OR

Condition D one channel inoperable with Required Action 0.3 to be in MODE 3 with a completion time of 78 Table 3.3.1-1 Item 2.a PR Neutron Flux High Page 3 of 5 KPS-ES-D-1-Scenario1-1-A-L-09302010-040

Either of the following for Table 3.3.1-1 Item 2.a Neutron Flux High

Condition E one channel inoperable with Required Action E.1 to place channel rnJrip with a completion time of 72-hours. --=Tc--,ab"l-e---+----

3.3.1-1 Item 2..t:t~R Neutron Flux Low & Item 6. Overtemperature 8T.

OR

Condition E one channel inoperable with Required Action E.2 to be in MODE 3 with a completion time of 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />. Table 3.3.1-1 Item 2.b PR Neutron Flux Low & Item 6. Overtemperature 8T.

The crew should note that tripping of the Loop B Chan 1 OT8 T will result in TWO channels of OT8 T being tripped and result in a reactor trip signal being generated. [Tripping of OT8 T Loop B Chan 1 should NOT be directed.] Management should be contacted to resolve problem and prioritize work.

If the crew elects to trip bistables the unit will trip and will move to the next event SGTR.

Page 4 of 5 KPS-ES-O-1-Scenario1-1-A-L-09302010-040

After the failed NI channel has been addressed, a tube rupture will occur in SG B. The rupture will build in to a value of approximately over 5

",,~,~- -+----~ ~nutes. The crew shoukff.eGQgrnze secondary system-radiation indications for monitors R-43, R-15 and R-19 (SG B steam line N-14, Condenser Air Ejector and SG blowdown), and respond to the lowering RCS pressure, lowering PRZR level and increased charging. When letdown is isolated and maximum charging flow is established with two Charging Pumps, AND PRZR level is still decreasing, the Unit Supervisor (US) will direct a reactor trip as directed by AOP-RC-004 (or AOP-RC 001). When the reactor is tripped the SGTR will increase to its maximum input value (the ramp is stopped).

5 The crew will perform immediate actions of E-O:

1. CHECK reactor trip and reactor subcritical
2. ENSURE turbine trip
3. CHECK Bus 5 OR Bus 6 (ESF Bus) energized
4. CHECK SI actuated.

It is expected that SI will be manually actuated, or will be required, by this time as PRZR level continues to lower.

The US will provide and brief. The US should address the FOLD OUT Page Criteria 3 for isolating feed flow to a ruptured SG when narrow range level goes above 5%. The A TC operator will perform Attachment A steps while the US directs the BOP operator performing E-O actions.

The BOP operator is expected to report the failure of AFW Pump A to start. [It should have started on low-low SG level and/or SI signaL] The BOP operator will manually start AFW Pump B, after closing AFW-2A, (and SI sequencer complete) by taking its control switch to STOP and then START positions. [fHIS IS A CRITICAL TASK and must be accomplished by the initiation of the RCS cooldown directed in E-3, step 11]

The crew will continue the actions of E-O to ensure Safeguards equipment is operating as required. Diagnosis will be made of a SGTR and transition will be made to E-3.

The crew will identify SG B as ruptured and isolate steam flow from SG B by closing the MSIV. [fHIS IS A CRITICAL TASK and must be 6 accomplished by the isolation of major flow paths from the SG by completion of E-3, step 4]. Feed flow to SG B will also be stopped. A RCS target temperature based on SG B pressure will be determined and a cooldown initiated using the condenser [MS-1A open] or SD-3A [MS-1A closed]. The cooldown will be stopped and stabilized after the target temperature is reached. The RCS will then be depressurized using PRZR Sprays or a PRZR PORV. Conditions will be checked for SI termination and SI flow stopped. [fHIS IS A CRITICAL TASK and must be accomplished before SG B level is at 100% narrow range AND SG B pressure rises above 1050 pSig, indicative of SG overfill]

Conditions will be established to allow balance between RCS pressure and SG B pressure.

The scenario may be terminated at the point SI flow is stopped in E-3.

Page 5 of 5 KPS-ES-D-1-Scenario1-1-A-L-09302010-040

Appendix 0 Scenario Outline Form ES-D-1 Facility: Kewaunee Power Station Scenario No.:.£ Op-Test No.: 1 Examiners: Operators: ~S~R~O:.:...:_ _ _ _ _ _ _ _....:.

ATC:

BOP:

Initial Conditions: IC-17: 51% power EOl Equilibrium Xenon. RCS boron concentration is 256 ppm.

RCS Tave is 558°F. Generator load is 282 MWe gross. AS-31/AS-35 R-11 & R-12 Sample Retum Aligned to Containment. PR-1A PRZR PORV Block Valve is closed with power maintained due to PR-2A seat leakage, 2 inch containment vent is in progress per NOP-RBV-002 section 5.6, N41 is OOS (failed low previous shift)Power increase is to be initiated to 56% power. Main Feedwater Pump A start is in progress.

Turnover: Power initially lowered due to problems associated with Main Feed Pump 1A. Problems have been corrected and the pump is ready for retesLPlant Management has directed starting Main Feed Pump 1A and raising power to 56% at 1/2% per minute. Then hold at 56% for testing of Main Feed Pump 1A.

  • OP-KW-GOP-106 has been completed up-to and including 5.1.12.
  • NOP-FW-001 has been completed up-to and including step 5.1.5
  • Reactor Engineering has provided a reactivity plan and the plan has been reviewed and approved.

TS 3.4.11 (Reactor Coolant System Pressurizer Power Operated Relief Valves) Condition A One or more PORVs inoperable and capable of being manually cycled. Required Action A.1 is to close and maintain power to the associated PORV with a completion time of one hour (Completed).

TS 3.3.1 (Reactor Protection System (RPS) Instrumentation) Condition A One or more Functions with one or more required channels or trains inoperable. Required Action A.1 enter the condition referenced in Table 3.3.1-1 for the channel(s) or train(s) with a completion time of immediately.

TS 3.3.1 Reactor Protection System (RPS) Instrumentation) Condition D one channel inoperable with Required Action D.2.1 to place channel in trip with a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND Required Action D.2.2 perform SR 3.2.4.2 with a completion time of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Table 3.3.1-1 Item 2.a PR Neutron Flux High - Channel in Trip and SR 3.2.4.2 performed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> before start of the scenario TS 3.3.1 Reactor Protection System (RPS) Instrumentation) Condition E one channel inoperable with Required Action E.1 to place channel in trip with a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. - Channel in Trip Page 1 of6 KPS*ES*Q*1*Scenario1*2*A*L*09302010-038

Event Malf. Event Event No. No. Type'" Description Pre PR-2A CAUTION tagged CLOSED due to seat leakage. PR-1A Load closed TS LC03.4.11, ACTION A.1.

N105A 1.2 RF:

Pre- N41 Power Range NI Red Channel is OOS failed low. OT.AT Trip RP133 N/A Load Trip bistable and Rod Stop bistable are tripped.

RP134 Trip 01-46355*

CLOSE ON 01-46355*

Pre- OPEN OFF Containment Sump 8 Supply to RHR Pumps, SI-350A and SI 01-46356* ATC-C Load 3508, fail to open CLOSE ON 01-46356*

1 OPEN OFF N/A ATC-R 80P-N Turbine load is increased and Main Feedwater Pump B is started prior to exceeding 285 psig impulse pressure (56% turbine power)

=

SG A red pressure channel (PT-468) fails high to 1400 psig over 30 RX213 2 1400 0:30 80P-C seconds resulting in increase in indicated steam flow to SG A, increase in feed flow to SG A and opening of SG A PORV, SD-3A RC08 An unisolable RCS leak to containment atmosphere of I 3 1.4 5:00 ATC-C approximately 20 gpm develops over a 5-minute period.

ATC, RC03A 80P 4 10%,0:30 RCS leak propagates into a large-break LOCA.

M 01-46355 CLOSE ON 01-46355 SI-350A and 51-3508, CNTMT Sump 8 Supply To RHR Pump Al8, OPEN OFF fail to open in ES-1.3, Transfer To Containment Sump 5 DI-46356 ATC-C Recirculation, requiring transition to ECA-1.1, Loss of Emergency CLOSE ON 01-46356 Coolant Recirculation OPEN OFF

.. (N}ormal, (R)eactivity, (I}nstrument. {C}omponent, (M)ajor Page 2 of6 KPS-ES-D-1-Scenario1-2-A-L-09302010-038

SCENARIO 1-2 OVERVIEW

..... _.. .~ .. -- .

EVENT DESCRIPTION Feedwater Pump A start and Power increase.

Crew is directed during turnover to start MFP 1A and perform a power increase to approximately 56% RTP at 1/2 %/min. The crew should start 1 MFP 1A per NOP-FW-001. After starting MFP 1A the crew should commence the power increase using OP-KW-GOP-106.

The ATC operator will dilute in accordance with the Reactivity Plan using

. NOP-CVC-001, and the BOP operator will raise turbine load using NOP-i TB-001.

Page 3 of6 KPS-ES-D-1-Scenario1-2-A-L-09302010-038

After the power increase, SG A red channel pressure, PT-468, fails high to 1400 psig over 30 seconds. This will result in indicated SG A steam flow reading higher than actuaUtow, and-cause FW-7 A to throttle open to increase feed flow in response to the increased steam flow. Additionally, SO-3A, SG A PORV will open in response to the overpressure condition sensed by its controller.

The BOP operator will respond according to AOP-GEN-001, Immediate Operator Actions, Attachment D for a SG PORV Failure, and/or the appropriate Alarm Response Procedures. Once SO-3A is noted to be open with SG A pressure less than 1005 psig, the BOP operator will manually close SD-3A by taking its controller to the MANUAL position and ensuring the control potentiometer is rotated fully clockwise to 0%

demand.

The operator should also take action to place FW-7A, Feed Reg Valve, in MAN, to control SG level at 44% (AOP-GEN-001, Attachment A).

The failed pressure instrument will be identified and removed from service using AOP-MISC-001, Response to Instrument Failure, Attachment D Steam Generator Pressure. The alternate steam flow channel (FT-465 White) will be selected. When SG A level is stabilized FW-7A controller will be restored to Automatic.

Technical Specifications:

2

Condition A One or more Functions with one or more required channels or trains inoperable. Required Action A.1 enter the condition referenced in Table 3.3.1-1 for the channel{s) or train{s) with a completion time of immediately.

Condition E one channel inoperable with Required Action E.1 to place channel in trip with a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Table 3.3.1-1 Item 15. SG Water Level- Low coincident with Steam Flow/Feedwater Flow Mismatch.

Instrumentation Condition A One or more Functions with one or more required channels or trains inoperable. Required Action A.1 enter the condition referenced in Table 3.3.2-1 for the channel(s) or train(s) with a completion time of immediately.

Instrumentation Condition D one channel inoperable with Required Action D.1 to place channel in trip with a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Table 3.3.2-1 Item 1.e Safety Injection - Steam Line Pressure - Low

  • TS 3.3.3 Post Accident Monitoring Instrumentation all functions are met The Unit supervisor will direct tripping of bistables. It is expected to trip bistables for the failed channel(s) during the scenario Page 4 of6 KPS-ES-O-1-Scenario1-2-A-L-09302010-038

After the failed SG pressure channel is addressed. a small non-isolable leak will develop in containment. The leak will reach a maximum value of approximately 20 gpm overa 2;.;minUte-pe~ontainment radiation monitors will indicate a marked increase in radiation levels inside containment. containment pressure and humidity level will rise, Fireworks RXCP vault detection will alarm with riSing temperatures in containment.

Charging Pump A in auto will increase speed to attempt to maintain PRZR level on program The crew will respond by entering AOP-RC-001, Reactor Coolant Leak.

With PRZR level lowering the crew is expected to isolate letdown and establish charging flow to maintain PRZR level on program. The crew will 3 then attempt to diagnose and isolate the leak. The crew should determine the leak is in containment based on containment conditions. The Unit Supervisor should inform the Shift Manager (or crew member) to contact RP and set up for a containment entry to determine leak location. The Unit Supervisor will direct determining the leak rate. and based upon the value.

determine that if the condition continues a unit shutdown would be required.

Technical Specifications:

Page 5 of6 KPS-ES-O-1-Scenario1-2-A-L-09302010-038

After Technical Specifications for the RCS leak have been addressed the RCS leak will develop into a large break LOCA. RCS pressure and PRZR level will rapidly decrease. The crew is-expected to recognize the need for a reactor trip and initiate Safety Injection on PRZR pressure and/or PRZR level (cannot be maintain> 3%). The reactor will trip and Safety Injection will actuate.

The crew will perform immediate actions of E-O:

1. Verify reactor trip
2. Verify turbine rip
3. Verify Bus 5 OR Bus 6 (ESF Bus) energized
4. Check SI actuated.

It is expected that SI will have been actuated, or will be required, by this time.

4 The US will provide and brief. The US should address the FOLD OUT Page Criteria stopping RXCPs. The ATC operator will perform Attachment A steps while the US directs the BOP operator performing E-O actions.

The crew will continue the actions of E-O to ensure Safeguards equipment is operating as required. Diagnosis will be made of a LOCA and transition will be made to E-1, Loss of Reactor Or Secondary Coolant. A RED Path is expected to exist in the Integrity CSF Status Tree. If so, entry is made into FR-P .1. When performing step 1, RCS pressure is less than 300 psig and RHR injection flow (FI-626 or FI-928) is above 700 gpm, so transition will then be made to E-1.

The crew will continue action of E-1 and monitor RWST level. When RWST level is < 37%, as indicated by annunciator 47023-B, RWST i

LEVEL lOW, the crew will transition to ES-1.3, Transfer To Containment Sump Recirculation.

The crew will identify Train A ECCS flow, and stop Train Band unnecessary equipment.

The crew will attempt to open both Containment Sump B to RHR Pump Suction valves SI-350A and SI-350B at step 7 of ES-1.3. Neither valve will operate resulting in transition to ECA-1.1, Loss of Emergency Coolant Recirculation.

In ECA-1.1, the BOP operator will check Containment Cooling with four Fan Coil Units in operation. The crew will determine the required number of Containment Spray Pumps required operating, stop the ICS Pumps to 5

conserve RWST inventory. The ATC operator will be directed to initiate refill of the RWST using NOP-CVC-001. The BOP operator will continue with actions of ECA-1.1 by maintaining SG levels, initiating cooldown, if necessary, and monitoring EECS Pump operation.

Actions will continue until it is determined that SI flow CANNOT be terminated (Step 17), the minimum required injection flow is determined, RHR Pumps stopped, and direction provided to adjust SI-7B locally to achieve the determined minimum flow. [THIS IS A CRITICAL TASK]

The scenario may be terminated minimum injection flow has been determined.

Page 6 of6 KPS-ES-D*1*Scenario 1-2-A-L-0930201 0-038

Appendi~D Scenario Outline Form ES-D-1 Facility: Kewaunee Power Station Scenario No.: ~ Op-Test No.: 1 Examiners: Operators: SRO:

ATC:

BOP:

Initial Conditions: 73% 120wer EOl Eguilibrium Xenon. RCS boron concentration is 162 I2l2m. RCS Tave is 557°F. Generator load is 423 MWe gross. AS-31/AS-35 R-11 & R-12 Saml2le Return Aligned to Containment, PR-1A PRZR PORV Block Valve is closed with 120wer maintained due to PR-2A seat leakage, 2 inch containment vent is in progress per NOP-RBV-002 section 5.6, Turnover: Maintain 73% Power, TS 3.4.11 (Reactor Coolant S~stem Pressurizer Power Ooerated Relief Valves} Condition A One or more PORVs inol2erable and cal2able of being manuall~ c~cled. Reguired Action A.1 is to close and maintain 120wer to the associated PORV with a coml2letion time of one hour {Coml2leted}.

Event Malf. Event Event No. No. Type* Description Pre PR-2A CAUTION tagged CLOSED due to seat leakage. PR-1A Load closed TS LC03.4.11 ACTION A.1.

DI-4623Q.

Pre- CLOSE ON Load ATC-C CVC-440, Emergency Boration to Charging Pumps, fails closed DI-4623Q.

OPEN OFF P~!d I S~~~8 Lo 05A& ATC-C Failure of SI pumps to Auto Start ATC-R SG tube leak of approximately 12 gpm in SG B entered over 5 SG038 1 minutes 30 5:00 BOP-N Crew initiates a Dower reduction to 45% power 01-46230 CLOSE ON 2 01-46230 ATC-C CVC-440, Emergency Boration to Charging Pumps, fails closed OPEN OFF MS03B 1.0 ATC.

B steam line leak outside Containment. EO calls in. - ATC the 3 BOP-3:00 M Controls Trip the Reactor and BOP Initiate Main Steam Isolation Steam Generator 'B' Safety Opens when the crew closes the MS 4 MS048 BOP-C 1B - Operator Isolate Feed Flow to Steam Generator B SI05A, 5 SI05B I ATC-C Failure of Safety Injection Pumps to Auto Start

  • (N)ormal, (R)eactivlty, (l)nstrument, (C)omoonent, (M)aior Page 1 of 3 KPS-ES-0-1-Scenario1-3-A-l-09302010-037

~~. ~~~~_~~-=-SC~ENARIO 1-3 OVERVIEW EVENT DESCRIP,.ION A SG tube leak of approximately 15 gpm occurs in SG B ramped in over 5 minutes. SG B recorder will show an increase in N16 count rate as leak develops, and a PPCS alarm actuating TLA-15, RMS ABOVE NORMAL.

The crew will respond by entering AOP-RC-004, Steam Generator Tube Leak, check that the leakage is within charging system capability to maintain PRZR level> 3%. Charging flow will be adjusted to maintain PRZR level on program value. The crew will compare R-15 reading to the Chemistry estimated 100 GPD from CY-KW-059-003. With leakage greater than 100 GPD as indicated by R-15 count rate, the crew will check R19 (SG Blow Down monitor) has increased> 5% form background, and 1 then initiate ACTION LEVEL 3 actions. A load decrease of 3%/min to <

45% power to ensure the unit is less than 50% within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of leak initiation.

The leaking SG will be identified and SG blow down isolated. The crew will direct actions to minimize secondary system contamination. At 45%

power, the {un)loading rate may be adjusted to 1%/min provided that MODE 3 can be achieved within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of leak initiation.

Technical Specification:

  • TS 3.4.13 RCS Operational LEAKAGE Condition B Primary to secondary LEAKAGE not within limit Required Action B.1 be in MODE 3 with a completion time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Required Action B.2 be in MODE 5 with a required completion time in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> During the load back down while performing actions of AOP-GEN-002, Rapid Power Reduction, the ATC operator will initiate a boration of 50 gallons of boric acid using the emergency boration flow path (Attachment A). When the operator attempts to open CV-440, Emergency Boration to Charging Pumps, the valve fails to open. The operator will then initiate a boration path using the normal boration flow path (Step E5) by setting the 2

Boration Totalizer to 50.0, setting CVC-403 controller to the desired flow rate, placing the Reactor Makeup Mode Selector to BORATE, and then placing the Reactor Makeup Control switch to START. When the 50 gallons have been added, the ATC operator will restore the respective Makeup control switches to AUTO and START. Subsequent borations will be accomplished using Attachment F.

When power has been reduced by at least 5% and not greater than 10%,

a steam line leak outside containment will occur on the B steam line header. The leak is NOT expected to actuate any protective or safeguards 3

function. The crew should recognize condition and determine a reactor trip is required as direct by AOP-GEN-001, followed by clOSing the MSIVs .

The crew will perform immediate actions of E-O:

The BOP will initiate a main steam isolation and the crew will complete the immediate actions of E-O, "Reactor Trip or Safety Injection. When MS-1 B, 4 Main Steam Isolation closes, Steam Generator 'B' safety will fail open.

The BOP will then isolate feed flow to Steam Generator 'B' per the foldout page criteria.

Page 2 of3 KPS-ES-D-1-Scenario1-3-A-L-0930201 0-037

Safety Injection will automatically initiate. The safety injection pU!!!p=s-=:,w:::::iII=--t--_

fail to automatically start. The ATe operator will start both safety injection pumps per Attachment A of E-O, "Reactor Trip or Safety Injection.

5 The crew will continue in E-O and then transition to E-2, "Faulted Steam Generator Isolation" the scenario will end after the crew has completed the steps in E-2 to isolate all steam flow from the faulted Steam Generator Page 3 of 3 KPS-ES-O-1-Scenario1-3-A-L-09302010-037

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