ML060340004

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Examination Outline Documents for the Kewaunee Initial Examination - November 2005
ML060340004
Person / Time
Site: Kewaunee Dominion icon.png
Issue date: 11/14/2005
From: Johnson S
Dominion Energy Kewaunee
To:
Office of Nuclear Reactor Regulation
Shared Package
ML060060458 List:
References
50-305/05-301 50-305/05-301
Download: ML060340004 (40)
v • d • e


Text

EXAMINATION OUTLINE DOCUMENTS FOR THE KEWAUNEE INITIAL EXAMINATION - NOVEMBER 2005

Dominion Energy Kevauncc, Inc.

N490 Highway 42. Kewuaunct, WI 54216-951 I Dominion" AUG 1 2 2005 U. S. Nuclear Regulatory Commission Serial No. 05-484A Attention: Document Control Desk KPSILICIGR: RO Washington, DC 20555 Docket No. 50-305 License No. DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION RESPONSE TO REQUEST TO FURNISH EXAMINATION OUTLINES In response to a letter from the NRC dated July 14, 2005 regarding the administration of licensing examinations at the Kewaunee Power Station, examination outlines were hand delivered to Mr. Bruce Palagi and Mr. Hironori Peterson at Region 111 on August 8, 2005.

If you have any questions, please contact Frank Winks at (920) 388-8303.

Very truly yours,

.. ,i i $ , + .

Michael G. Gaffn Site Vice President Commitments made by this letter: NONE cc: Mr. J. L. Caldwell Administrator Region 111 US. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Mr. J. F. Stang Project Manager U S . Nuclear Regulatory Commission Mail Stop 0-8-H-4a Washington, D. C. 20555 Mr. S. C. Burton NRC Senior Resident Inspector Kewaunee Power Station

Kewaunee 2005 NRC Ooerator License Exam Examination Outline Submittal Contents

1. Form ES-201-2, Exam Outline Quality Checklist - 1 page
2. Form ES-201-3, Examination Security Agreement (copies) - 4 pages
3. Form ES-3 10-1, Administrative Topics Outline - 2 pages - 1 RO, 1 SRO
4. Form ES-301-2, Control R o o d - P l a n t Systems Outline - 3 pages - 1 RO, 1 SRO-I, 1 SRO-U 5 . Form ES-D-I, Scenario Outline - 9 pages - Scenario 1,3 pages; Scenario 2 , 2 pages; Scenario 3,2 pages; Scenario 4 (Spare), 2 pages
6. Form ES-301-5, Transient and Event Checklist - 4 pages - Group 1 (RO, SRO-I, SRO-U

- Scenarios 1 & 2), 1 page; Group 2 (2 SRO-I, SRO-U - Scenarios 1,2 & 3), 1 page; Group 3 (SRO-I, SRO-U - Scenario 1 & 2), 1 page; Scenario 4,1 page

7. Form ES-401-2, PWR Examination Outline, and Form ES-401-3, Generic Knowledge and Abilities Outline (Tier 3) - 14 pages - RO - ES-401-2,7 pages & ES-401-3, 1 page; SRO - ES-401-2,5 pages & ES-401-3,1 page
8. Form ES-401-4, Record of Rejected WAS- 2 pages
9. Kewaunee Power Station (KPS) Random Generation Technique and WA Suppression Report - 1 page
10. Additional Information: Last two NRC Exam JPM topics
a. 2002 NRC Examination record of Administrative Topics, and Control Room Systems and Facility Walk-Through subjects
b. 2004 NRC Examination record of Administrative Topics, and Control Room Systems and Facility Walk-Through subjects

ES-201 Examination Outline Quality Checklist Form ES-201-2 Date of Examination: 1/07/05

-aciii@: Kewaunee Power Station tern 1.

Task Description W

R I

T T

E

- N 2.

s I I M b. Assess whether mere are enough scenario sets (and spares) to test the projected number and mix of applicants in acwrdanca with the expedad UBW composition and rotation schedule qzkf U

L without compromisingexam integnty. and ensure that each applicant can be tested using A at least one new oi significantly modified Sc8nario. that no scenarios are duplicated T from me appllcants audit test(s). and that scanarloswill not be rewated on subsequent days.

0 c. To me extent possible, assess whether me outline(s)wnform(s) with the qualitative R

- 3.

W I

T (1) the tasks am distributed among the topics as specified on the form

- 4.

G E

N E

R A

L I. Author

). Facjlity Rewiewer (3

. NRC Chief Examiner (#)

I. NRC SupeNisor ES-201, Page 25 of 27

ES-301 Administrative Topics Outline F o ES-301-1

~

~~ ~ ~~ ~

~ ~ ~ i l i Kewaunee ty: Power Station Date of Examination: 14/07/05 Examination Level: RO SRO 0 Operating Test Number: 1 Administrative Topic Describe activity to be performed see Note Code' RO-036-JP03A.Reactor Coolant System Leak Rate Check -

Conduct of Operations SP-36-082 JPM for determining RCS leak rate using PPCS Computer and response.

WA2.1.19 RO value 3.0 I -

RO-033-JP05A, Verify SI Lineup N-SI-33-CL Alt A Conduct of Operations S. M JPM to check SI At-Power Lineup (> 100 psig RCS). Two errors

- one Initial missing from 1st Operator and one valve mispositioned.

WA 2.1.29 RO value 3.4 Record Individual Rod Positions with Control Rod Supervision Program (PPCS) Out of Service - A-CP46 Data Sheet 1 Record rod positions with one exceeding Tech Spec allowed limits.

IUA 2.2.12 RO value 3.0 RO-OlEJP03A,Perform Actions Prior To Discharge for a Containment Purge Discharge Permit SP-328.116 Attachment D.

Record rad monitor values, perform check source, align rad monllor to vent, and get met data.

WA 2.3.9 RO value 2.5 N/A NOTE: All items (5total) 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 p -3 for ROs; 4 for SROs & RO retakes)

(N)ew or (M)odified from bank p '1)

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

ES-301, Page 22 of 27

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Kewaunee Power Station Date of Examination: 11/07/05 Examination Level: RO 0 SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code' RO-036-JP03A.Reactor Coolant System Leak Rate Check -

SP-36-062 JPM for determining RCS leak rate using PPCS Computer and response.

WA2.1.19SROvalue3.0 Evaluate Crew Stafflng for Shin Compliment with Severe Weather Conditions - Technical Specification 6.2.b.

Conduct of Operations R. N JPM for determining if plant operations can continue based on Tech Spec requirements during severe snowstorm.

WA 2.1.5 SRO value 3.4 Record Individual Rod Positions with Control Rod Supervision Equipment Control SIN Program (PPCS) Out of Service - A-CP-46 Data Sheet 1.

Record rod positions with one exceeding Tech Spec allowed limits.

WA 2.2.12 SRO value 3.4

~~~~~

SO-I1SJPOIA, Review and Authorize a Gas Decay Tank Discharge Permit - SP-32B-116 Attachment E.

Total Gas activity is >I .OE-2 .Cl/cc but only ONE Aux Bidg Fan will be running (marked). [Requires both fans]

WA 2.3.6 SRO value 3.1 SO-I 19-JP03K. Classify Emergency Event - Control Emergency Plan Room Evacuation - EPIP-AD-02.

Determine event Classification for given conditions.

KIA 2.4.41 SRO value4.1 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 p -3 for ROs; 4 for SROs 8 RO retakes)

(N)ew or (M)odified from bank p -1)

(P)revious 2 exams p -1; randomly selected)

ES-301, Page 22 of 27

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 I

~~~~l~~ Kewaunee Power Station Date of Examination: 11/07/05 Exam Level R O M SRO-I u SRO-U 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 / JPM Title Type Code' Safety Function

a. Rod Control / Perform a dropped rod recovery (K/A 003A1.02 3.6) D, s 1
b. NISI Loss af Inbenediale Range InsVumentation wivl Failure of RsaclarTnp (KIA 033bA207 3.91 A, L, N, S 7 C. DG / Start and Load the Diesel Generator (WA 064A4.06 3.9) A, D, S 6
d. RHW Operate the RHR System In Split Train Mode (WA E03EA1.3 3.7)

D, 4

e. SI ISlop SI Pump duing Pori-LOCA Caaldorn and OsprB(I~uNyllion(WA m9EA7.M 3.6 E03EAZ.Z 3.51 A, N, S 2
f. CVCS / Respond to High Reactor Coolant Activity 076AA2.02 2.8) N, s 9
9. PR PressureITransfer from Manual to Automatic PressureContml (WA 010A4.01 3.7) D, s 3
h. PRT I Lower PRT level (K/A 007A1.01 2.9) N. S 5
i. ATWS / Locally Isolate Dilution Flowpaths (WA 029A 2.4.35 3.3) 1 D, E, R 11 k ~ ~ ~ ~ . 1 ~ . n o r m
  • n ~ m . ~ . a u ~ ~ ~ ~ m m b . s u . m n 1 ~ n . r 1pt iM

~ ~I ~I ,11 D, E

~l 4~1 11 0 ~ 1

  • 1 ~ 8 (A)lternate path 4-6 /4-6/ 2-3 (C)ontrol room (0)irect From bank - 4 /..8 /
  • 4 (E)mergency or abnormal in-plant (L)ow-Power / Shutdown * -

-1 / * -1 / * .I

-1 / '1 / -1 (N)ew or (M)odified from bank including 1(A)

(P)revious2 exams (R)CA 3 - .2/ s . 2 / -1

. I -3 / * -2 (randomly selected)

.-1/..1/*.1 (S)imulator ES-301,Page 23 of 27

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2

~ ~ Kewaunee ~ i l Power ~ ~Station  : Date of Examination: 11/07/05 Exam Level: R O U SRO-I SRO-U 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*

I Safety Function a Rod Control I Perform a dropped rod recovery [K/A 003A1.02

b. NIs i LOPS ofin~mdiam Range lnshumsntalMnwim Failure of Reador Ttip ilvA 0 3 3 W . 0 7 4.21 A, L, N, S 7 C. DG / Start and Load the Diesel Generator (WA 064A4.06 3.9) A, D. S 6
d. RHiUOperate the RHR System In Split Train Mode (KIA E03EA1.3 4.1)

D, 4 e S I I S ~ ~ ~ S I P " ~ ~ ~ " ~ ~ ~ P ~ ~ ~ - L O CE O s

A ~C E ~A ~Z ~ ~~ ~ ~) ~ D ~ ~ ~ ~ ~ ~ ~ ~ ~ I U A W O E ~ ~ ~ Z A, N, 2

f. CVCS I Respond to High Reactor Coolant Activity (WA 076AA2.02 3.4) N, s 9 9.PR Pressure 1 Transfer from Manual to Aufmallc Pressure Control (WA 01OA4.01 3.5) D, S 3 h

In-Plant Systems" (3 for RO). (3 for SRO-I): (3 or 2 for SRO-U)

~~

i. ATWS / Locally Isolate Dilution Flowpaths (WA 029A 2.4.35 3.5) D, E, R 1
j. AFWI LacsllyConlrolFeedFla*ioMinimireRCSCmldamwNIWlhSG~Fauiled~~AE12EAl.f 3.8)

A, N, E 4 k CR srsc 1 panorm -ions N-UW n c c o n ~Room Euasu-n F b i IWA O n W I 08 4 2 m@A111 1 2 )

IC~WQW D, E 8

@ All RO and SRO-l control room (and in-plant) systems must be different and serve different safety functions, all 5 SRC-U systems must serve different safety functions, in-plant systems and functions may

  • Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 (4-6 / 2-3 (C)ontrol room (Dlirect from bank

()mergemy or abnormal inplant (L)ow-Power / Shutdown

..g/.*/.=l

  • I/. .1 /
  • 1
  • .I

/ * .I/ * .1 (N)ew or (M)odifiedfrom bank including 1(A) * -2 / - .2 / * *1 (P)revious 2 exams -3 / e3 / a 2 (randomly selected)

IRlCA

~, * .1 I

  • e1 I * -1 I

I (S)imulator ES-301.Page 23 of 27

I I I I

ES-301 Transient and Event Checklist Form ES-301-5 acil ty rtowawee Power Slat on Date of Exam 11 07r05 Operalinq Test tu0 1 iGroLD 1, A I E l Scenarios M

I N

I M

U M()

Instructions

1. Check the applicant level and enter the operating test number and Form ES-D-I 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 do one scenario, including at least two instrument or component (IC) malfunctions and one major transient. in the ATC position.
2. Reactivity manipulations may be conducted under normal or contmlled 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 I-for-I basis,
3. 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.

ES-301, Page 26 of 27

ES-301 Transient and Event Checklist Form ES-301-5 acdty 6ewaLnee Power Station Date of Exam 11107105 Operatina Test No 1 1Gro.o 21 A I E l Scenarios M

I N

I M

U M(')

1 Cnecd the appl cant eve and enter the operat,ng test n m b e r ana Form ES-D-1 event numbers for eacn event lype. TS are not applicable for RO applicants ROs mJSt Serve In OOtn the -at-the-conlro s (ATC) ana 'Dalance-of-plant(BOP( pos~t.ons.Instant SROs mJst do one scenaro inclLd ng at least two instrunen1 or component (I&) malbnctions ana one ma.or transient. n the ATC position 2 React v ty man pdat ons may oe conducted under normal or conlrolled abnormal w n d lions (refer to Section D 5 a) bJt must be s gniicant per Sen on C 2 a of Append x D ('1 React v ty ana normal evoutions may De rep aced w th aaait onal mstrument or component ma fLnnions on a 1-for-1 D a m 3 Wnenever practica , both instrunent and component malfJnctionS shodd De .nclLoed only those that reqd re ver fiab e a n ons tnat prov de insight 10 the applicant's competence coLn1 towara the m nimLm req, rements specfied for the applcant s .cerise level in tne rghl-nana co.t.mns ES-301, Page 26 of 27

ES-301 Transient and Evont Checklist F O I ES-301-5

~

nstructions:

I. 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-thecontrols (ATCY and 'balance-of-plant (BOP)' positions; Instant SROs must do one scenario. induding at least two instrument or component (IC) malfunctions and one major transient. in the ATC position.

2. Reactivity manipulations may be conducted under normal or controyed abnormal conditions (refer to Section D.5.d) but must be significant per Seclion C.2.a of Appendix D. r) Reactivity and normal evolutions may be replaced with additional inlltrumenl or component malfunctions on a 1-for-1 basis.
3. Wenever practical. both instrument and component malfunctlons should be induded; only those that require verifiable actions that provide insighl 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

ES-301 Transient and Event Checklist Form ES-301-5

acility: Kewaunee Power Station Date of Exam: 11/07/05 Operatinq Test No.: 1 IAnv)

A I E l Scenarios 1 Check tne applicant level an0 enter [he operating test nJmDer and Form ES-D-1 event n m o e r s for eact event type TS are not app.icaole for RO applcants ROs mLst serve in both me at-tne-contro s rATC)'

ano -0alance-of-plant (BOP)" posit ons. Instant SROs mdst do one scenar o mc1t.d ng at least two instrLment or component (IC)malfbnctions an0 one ma.or transient n the ATC position 2 React v.ty man pblat.ons may oe conducted under normal or conlrolled abnorma cond t ons (refer to Section D 5 d) bLt mast be s gniicant per Section C 2 a of Append x D ('I React vity an0 normal evolJt ons may be replace0 witn addlt,onal instrument or component ma fdnctions on a I-for-I basis 3 Whenever practical both nstrdment and component malfLnct on$ mod 0 be ,nc Jdea on y tnose tnat requ re verifiable actions tnat prov8de insght to tne app icants competence codnt toward the m n i m m reqJ rements spec fied for the appl cant s cense level n the rgnt-hand co Jmns ES-301,Page 26 of 27

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ES-401 PWR Examination Outline Form ES-401-2 Faciliw: Kewaunee Nuclear Power Plant Printed: 08/04/2005 Date Of Exam: 11/07/2005 RO WA Category Point:

I SRO-Only Points

1. Ensure that at least two topics from every K/A category are sampled within each tier of the RO outline (i.e., the "Tier Totals" in each K/A category shall not be less than two). Refer to Section D.1.c for additional guidance regarding the SRO sampling.
2. 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 i 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.
3. Select topics from many systems and evolutions; avoid selecting more than two KIA topics from a given system unless they relate to plantspecific priorities
4. Systerns/evolutionswithin each group are identified on the associated outline
5. The shaded areas are not applicable to the category /tier 6.' The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. The SRO WAS must also be linked to 10 CFR 55.43 OT an SRO-level learning objective.
7. On the following pages, enter the K/A numben. a brief description of each topic, the topics' importance ratings (IR) 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 columns labeled "K' and "A'.

Use duplicate pages for RO and S R O d y exams.

8. For Tier 3, enter the WA numbers, descriptions, importance ratings, and point totals on Form ES-401-3.
9. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate K/A statements.

1

ElAPE # I Name I Safety Function K1 K2 K3 A1 A2 G KATopic Imp. Points Recovery 1 I Trip - Stabilization -

000008 Pressurizer Vapor Space Accident /

X X

IIII EKI .04 - Decrease in reactor power following reactor trip (prompt drop and subseauent decav)._

AK1.02 -Change in leak rate with 3.1 I 3 change in pressure 000009 Small Break LOCA / 3 X 2.1.32 Ability to explain and apply all

~ 3.4 1 system limits and precautions.

0000 I 1 Large Break LOCA / 3 X EA2.05 - Significance of charging pump 3.3 I operation xll I I I I I

~

Malfunctions / 4 AK3.03 - Sequence of events for manually tripping reactor and RCP as a r

result of an RCP malfunction 3.7 I 000022 Loss of Rx Coolant Makeup / 2 X AKI.01 - Consequences of thermal 2.8 1 shock to RCP seals 000025 Loss of RHR System / 4 I AA2.05 Limitations on LPI flow and temperature rates of change I I I 3.'*

000027 Pressurizer Pressure Control System X 2.1.33 -Ability to recognize indications 3.4 1 Malfunction / 3 for system operating parameters which are entry-level conditions for technical specifications.

000058 Loss of DC Power / 6 X AA1.02 - Static inverterdc input 3.1' I breaker, frequency meter, ac output breaker, and ground fault detector 000065 Loss of Instrument Air / 8 X AA1.05 RPS- 3.38 I

~

WIE04 LOCA Outside Containment / 3 X EK2. I - Components, and functions of 3.5 I control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features WIEO5 Inadequate Heat Transfer - Loss of X 2.4.3 1 - Knowledge of annunciators 3.3 I Secondary Heat Sink / 4 alarms and indications, and use of the response instructions.

PWR RO Examination Outline Printed: 08/04/2005 Facility: Kewaunee Nuclear Power Plant ES 401

~~

Emergency and Abnormal Plant Evolutions - Tier 1 /Group I Form ES-401-2 E/APE # I Name / Safety Function K1 K2 K 3 Points WiE12 - Stem Line Rupture - Excessive X EK3.4 - RO or SRO function within the 3.5 I Heat Transfer 14 control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated WA Category Totals: 3 1 4 Group Point Total: 18 2

PWR RO Examination Outline Printed: 0810412005 Facility: Kewaunee Nuclear Power Plant ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 I Group 2 Form ES-401-2 E/APE # / Name / Safety Function Q AI A2 2 KATopic Imp. Pointr 000005 lnoperable1Stuck Control Rod 1 X 3.6 I 000028 Pressurizer Level Malfunction 1 2 X AK3.05 - Actions contained in EOP for 3.7 1 000037 Steam Generator Tube Leak / 3 X 3.4 I 00005 1 Loss of Condenser Vacuum / 4 X 2.4.50 -Ability to verify system alarm 3.3 I setpoints and operate controls identified in the alarm response manual.

000060 Accidental Gaseous Radwaste Rel. / AK1.02 - Biological effects on humans 2.5 1 9 of the various trpes of radiation, exposure levels that are acceptable for personnel in a nuclear reactor power plant; the units used for radiation intensity measurements and for radiatior exposure levels 000061 ARM System Alarms I I AK2.01 -Detectors at each ARM 2.5* 1 system location 000068 Control Room Evac. / 8 lAK2.02 - Reactor trip system 3.7 1

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WIE03 LOCA Cooldown Depress. 14

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X EA 1.1 - Components, and functions of 4.0 I control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features W/E16 High Containment Radiation / 9 X EA22 - Adherence to appropriate 3 .O I procedures and operation within the limitations in the facility's license and amendments KIA Category Totals: 2 2 1 I .

Groun Poi1 Total: 9

i 1 PWR RO Examination Outline Printed 08/04/2005 Facility: Kewaunee Nuclear Power Plant 3s - 401 E I I I

~4 1T

Opci ;1 003 Reactor Coolant Pump K3.03 - Feedwater and emergency feedwater 003 Reactor Coolant Pumo A3.02 - Motor current 004 Chemical and Volume Control KS.30 - Relationship between temperature and pressure in CVCS components during solid plant operation 004ChemicaI and Volume Control I l l K6.24 -Controllers and 1~ - 1 I I.

~~

positioners 005 Residual Heat Removal I I Ix K3.06 - CSS 006 Emergency Core Cooling X K2.01- ECCS pumps 006 Emergency Core Cooling A3.06 - Valve lineuos 007 Pressurizer ReliefiQuench AI .03 - Monitoring quench Tank tank temperature 007 Pressurizer RelieUQuench X K3.01 -Containment Tank 008 Component Cooling Water X K1.02 - Loads cooled by ccws 0 I O Pressurizer Pressure Control X I 1 A4.01 - PZR spray valve I 3.7 012 Reactor Protection I 1A1.O1 - Trip. setpoint. I 2.9' Iadjustment 0 13 Engineered Safety Features I I K5.01 - Definitions of safety I 2.8 train and ESF channel K4.05 - Containment 2.6' cooling after LOCA destroys ventilation ducts 026 Containment Spray A2.05 -Failure of chemical 3.7 addition tanks to inject A3.02 - Isolation of the 3.1 MRSS K1.02 - AFW System 3.4' OS9 Main Feedwater X 2.2.1 - Ability to perform 3.1 pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

KZ.02 - AFW electric driven 3.7*

Feedwater Feedwater T abnormal indications for system operating parameters which are enhy-level conditions for emergency

PWR RO Examination Outline Printed: 08/04/2005 Facility: Kewaunee Nuclear Power Plant ES - 401 Plant Systems -Tier 2 /Group 1 Form E .401-2 SyslEvol # I Name K1 C KATopic Point!

procedures. -

062 AC Electrical Distribution A4.03 - Synchroscope, I including an understanding of running and incoming voltages 063 DC Electrical Distribution I A2.0 1 - Grounds 1 2.5

- 1 I K1.O1 - AC distribution 1 4 . 1 I system -

I A4.02 - Radiation I 3.7 I monitoring system control X 2.1.32 -Ability to explain

- I and apply all system limits 076 Service Water I A4.04 -Emergency heat loads I

078 Instrument Air I K4.01 - ManuaVautomatic transfers of control K3.02 - Loss of containment 1

1 103 Containment integrity under normal operations -

Group Point Total: 28 2

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PWR RO Examination Outline Printed: 08/04/2005 Facility: Kewaunee Nuclear Power Plant Sys/Evol# / Name K l K2 K3 K4 K5 K6 A1 A2 A3 A4 G KATopic Imp. Points 001 Control Rod Drive X 2.4.50 Ability to verify

~ 3.3 I system alarm setpoints and operate controls identified in the alarmresponse manual.

002 Reactor Coolant X K6.03- Reactor vessel level 3.1 1 indication

~ _ _

Generic Knowledge and Abilities Outline (Tier 3)

PWR RO Examination Outline Printed: 08/04/2005 Fafilitv: Kewaunee Nuclear Power Plant Form ES-401-3 Conduct ofOperations 2.1.10 Knowledge of conditions and limitations in the 2.1 I facility license.

Ability to perform specific system and integrated 3.9 I plant procedures during all modes of plant operation.

Ability to locate control room switches, controls 4.2 I and indications and to determine that they are correctly reflecting the desired plant lineup.

Equipment Control 2.2.28 Knowledge of new and spent fuel movement 2.6 I procedures.

2.2.30 Knowledge of RO duties in the control room 3.5 1 during fuel handling such as alarms from fuel handling area, communication with fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

Category Total: 2 Radiation Control 2.3. I Knowledge of IO CFR: 20 and related facility 2.6 1 radiation control requirements.

2.3.9 Knowledge of the process for performing a 2.5 I containment purge.

2.3.11 Ability to control radiation releases. 2.1 I I Emergency ProceduresPlan 2.4.3 Ability to identify post-accident instrumentation, 3.5 1 2.4.45 Ability to prioritize and interpret the significance 3.3 I of each annunciator or alarm.

Generic Total: 10

ES-401 PWR Examination Outline Form ES-401-2 Facilitv: Kewaunee Nuclear Power Plant Printed: 08/04/2005 Date Of Exam 11/07/2005 SRO-Only Points 3 Genenc Knowledge And Abilities Categories Note

1. Ensure that at least two topics from evely KIA category are sampled within each tier of the RO outline (i.e., the 'Tier Totals" in each KIAcategoly shall not be less than two). Refer to Section D.1.c for additional guidance regarding the SRO sampling.
2. The point total for each group and tier in fhe proposed outline must match that specified in the table. The final point total for each group and tier may deviate by i1 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.
3. Select topics from many systems and evolutions; avoid selecting more than two WA topics from a given system unless they relate to plant-specific priorities.
4. Systems/evoluSonswithin each group are identified on the associated outline.
5. The shaded areas are not applicable to (he category /tier.

6.* The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. The SRO WAS must also be linked to 10 CFR 55.43or an SRO-level learning objective.

7. On the following pages, enter (he KIA numben, a brief description of each topic, the topics' importance ratings (IR) 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 columns labeled " K and "A'.

Use duplicate pages for RO and SRO-only exams.

8. For Tier 3, enter the WA numbers, descriptions, importance ratings, and point totals on Form ES401-3.
9. Refer to ES401,Attachment 2,for guidance regarding the elimination of inappropriate KIA statements.

I

PWR SRO Examination Outline Printed: 08/04/2005 Facility: Kewaunee Nuclear Power Plant Form ES-401-2 conditions and immediate action steps.

000054 Loss of Main Feedwater / 4 X AA2.01 - Occurrence of reactor andlor 4.4 I turbine trip 000058 Loss of DC Power / 6 X AA2.03 DC loads lost; impact on to

~ 3.9 1 operate and monitor plant systems WACategory Totals: 0 0 0 0 3 3 Group Point Total: 6 I

PWR SRO Examination Outline Printed: 08/04/2005 Facility: Kewaunee Nuclear Power Plant ES - 401 Emergency and Abnormal Plant Evolutions -Tier 1 / Group 2 Form ES-401-2 E/APE #/Name / Safety Function G 1 KATopic I Imp. 1 Points I II

~

000005 InoperableRtuck Control Rod / 1 AA2.04 - Interpretation of computer in-core TC map for dropped rod location 13.41 1 I 000068 Control Room Evac. / 8 lAA2.07 - PZR level I 4.3 I 1 000076 High Reactor Coolant Activity / 9 WiEO8 RCS Overcooling - PTS 14 criteria which require the notification of plant personnel.

WA Category Totals: 21 Group Point Total: 4 I I

I PWR SRO Examination Outline Printed: 08/04/2005 Facility: Kewaunee Nuclear Power Plant ES - 401 Plant Systems - Tier 2 I Group 2 Form ES-401-2 Rl I I I I I I I I I 1 I I I 033 Spent Fuel Pool Cooling

-K1 -KS - K6 -AI A2 A3 A4 G in technical specifications for limiting conditions for operations and safety limits.

07I Waste Gas Disposal 2.2.22 -Knowledge of limiting conditions for t

~

operations and safety limits.

075 Circulating Water X A2.02 - Loss of circulating 2.7 I water pumps KIA Category Totals: 0 0 0 0 0 0 0 1 0 0 2 Group Point Total: 3 1

Generic Knowledge and Abilities Outline (Tier 3)

PWR SRO Examination Outline Printed. 08/04/2005 Facilitv: Kewaunee Nuclear Power Plant Form ES-401-3 Generic Cateeorv -

KA KA Tonic -

'oints Conduct of Operations 2. I .5 Ability to locate and use procedures and 3.4 I directives related to shift staffing and activities.

2.1.25 Ability to obtain and interpret station reference 3.1 1 materials such as graphs, monographs, and tables which contain performance data. --

2 Equipment Control 2.2.1 Knowledge of the process for conducting tests or 3.2 I experiments not described in the safety analysis report. -

2.2.32 Knowledge of the effects of alterations on core 3.3 I configuration. -

Category Total: 2 Radiation Control 2.3.3 Knowledge of SRO responsibilities for auxiliary 2.9 I systems that are outside the control room (e.g.,

waste disposal and handling systems).

Emergency ProceduresiPlan I 2.4.1 1 1 Knowledge of abnormal condition procedures.

2.4.28 Knowledge of procedures relating to emergency 3.3 I response to sabotage.

I

Kewaunee Scenario No.: 1 Op-Test No.: I xaminers: Operators: SRO I RO I Initial Conditions: IC-12, 100% power, Middle of Cycle (MOC)

BOP Motor Driven AFW A Pump is in PULLOUT.

SI Pump A is in PULLOUT.

Turnover: The plant is at 100% power. AFW Pump A has been out of service for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> due to inboard bearing replacement. SI Pump A has been out of service for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after an inspection of the breaker cubicle indicated the lugs on one of the supply line connectors is loose.

At turnover, the Equipment Operator reported Steam Traps 23 and 24 on the lines from the individual steam header supplies to the TD AFW Pump, and Steam Trap 25 and 26 on the common steam line to the Turbine-Driven AFW Pump are cold. The Shift Manager has just declared the TD AFW Pump inoperable and directs the crew to remove it from service as soon as turnover is complete. (No action taken yet.)

2 RX211, I BOP Controlling SG A level channel (LT461) fails high over I O seconds.

100 SRO 3 swoSc, C BOP Running Service Water Pump 1BI trips on overcurrent. Standby Pump IO0 DI-46118-CLOSE SRO auto starts. SW-3A header isolation valve closes improperly.

ON 4 CC03B, c RO RXCP B Thermal Barrier leak resulting in automatic isolation of CC 2.0 SRO cooling from thermal barrier. This will eventually lead to a manual reactor trip directed due to RXCP parameters exceeding limits. (EOP entry) 5 A1-4301-02-R', C BOP SG B PORV Controller fails giving PORV open signal. Manual control 0 SRO available.

6 SGOIB, 7 M BOP SG B tube rupture of - 250 gpm ramped in over 10 minutes starting 2:OO 1o:oo RO minutes after the reactor trip.

~~

  • (N)ormal, (R)eactivity (I)nstrument, (C)omponent, (M)ajor Rev. A

SCENARIO 1-1 OVERVIEW

- The plant is at 100% power. SI Pump A and AFW Pump A are inoperable. At turnover, the Turbine-Driven AFW Pump is determined to be inoperable due to the failure of steam traps on the steam lines to the TD AFW Pump. The CRS will direct removal from service (Placing TD AFW Pump in PULLOUT, and possible closing the Steam Supply valves fro the SGs MS-IOA and MS-IOB). Technical Specification 3.4.b will be addressed and LCO 3.4.b.3 requires reactor power be reduced to less than or equal to 1673 MWt within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Operations Management will inform the crew that power reduction needs to begin as soon as possible. 1673 MWt corresponds to 94.4% power. Crew will use N-0-03 and N-TB-54 to reduce load.

- Following clearly observable plant response from the reactivity changes during the load reduction, the controlling level channel for SG A will fail high. The actual level will begin to lower due to FW-7A closing in response to the high level. The BOP will take FW-7A Controller to MANUAL and verify that SG level recovers toward program level. Procedure A-FW-SA and A-MI-87 will be used to address the failure. The failed instrument will be removed from service using A-MI-87. The SRO will address Technical Specification 3.5.b to ensure continued plant operation is allowed with the failed channel.

&& - After the operator actions are completed for removal of the failed SG Level channel from service, the running Service Water Pump IBI will trip on overcurrent. This will result in automatic start of the standby SW PumplB2 when SW header pressure drops below 72 psig. Train B Header Isolation valve SW-3A will inadvertently close. It normally closes on SI signal or on sensed low pressure in the associated header. Train A pressure will remain above the closure setpoint. The crew will respond by taking actions directed in A-SW-02, and re-open SW-3A when SW Header A Operating is checked. The SRO will address Technical Specification 3.3.e.2 for actions associated with the loss of the SW Pump. Also by direction ofN-SW-02 Precaution and Limitation 2.2, the SRO will determine in accordance with this item that with SI Pump B also inoperable, Technical Specification 3.0.c is applicable requiring plant shutdown.

- Once the requirement for plant shutdown is reviewed, a leak will develop in RXCP B Thermal Barrier. This will result in leakage of RCS water into the Component Cooling System. CC radiation levels will rise and be detected on process monitor R17, Component Cooling Surge Tank level will rise and CC-610B, RXCP B Thermal Barr Comp Cooling Return, will close on high flow. The crew will address the problem by entering A-RM-45, A-CC-3 1 and/or A-RC-36C. RXCP conditions will deteriorate for bearing temperature and will require the RXCP be stopped. The CRS should then direct a manual reactor trip and stopping of RXCP B when the immediate operator actions of E-0, Reactor Trip Or Safety Injection, are complete.

- At the time of the reactor trip (based on NI power) SG B PORV Controller will fail so that SD-2B, SG B PORV fully opens. Manual control is available and when the open SG PORV is identified by the BOP, the Controller will be placed to MAN.

Event 6 & 7 -Associated with the transient placed on SG B, a SG tube rupture will occur in SG 8. The rate will increase to a value of 250 gpm over a ten-minute period. The crew will recognize conditions requiring a Safety Injection and manually initiate SI (if required). Five minutes following SI actuation, Component Cooling Pump A trips. Component Cooling Pump B fails to automatically start and should be manually started after verification that SI sequencer completes its sequence. The scenario ends following cooldown and depressurization of the RCS, and if possible, equalization of RCS and ruptured SG pressure, OR at the discretion of the Chief Examiner.

Critical Tasks I , E-0 K - Manually start at least one CCW pump required to provide adequate component cooling for the operating safeguards trains before transition out of E-0. [Event 71

2. E-3 A - Isolate feedwater flow into and steam flow from the ruptured SG before a transition to ECA-3.1 occurs

[Event 61 2 Rev. A

SCENARIO 1-1 OVERVIEW

3. E-3 E ~ Establishimaintain an RCS temperature so that transition from E-3 does not occur because the RCS temperature is in either of the following conditions: [Event 61 Too high to maintain 50°F [85"F for adverse Containment] subcooling OR Below 270°F.
4. E-3 C - Depressurize RCS until either 1) P U R level is > 74%, 2) RCS subcooling based on CETs is < 30°F [6S"F for adverse containment], or 3) RCS pressure is < S/G pressure and PRZR level is > 5% [30% for adverse containment] before 96% level is exceeded in the ruptured S/G. [Event 61 3 Rev. A

acility: Kewaunee Scenario No.: 2 Op-Test No.: 1 xaminers: Operators: SRO RO BOP nitial Conditions: IC-10; 49% power, Beginning of cycle (BOC)

Ready to start Feedwater Pump B and continue power increase.

Motor Driven AFW A Pump is in PULLOUT.

SI Pump A is in PULLOUT.

urnover: The plant is at 49% power. Currently at step 4.1.9.d of N-0-03, Plant Operation Greater Than 35%

Power and step 8.b (4.1.4.i) of N-FW-OSA, Feedwater System Normal Operation. The EO is standing by at Feedwater Pump B, ready for pump starting. Once FW Pump B is started, the crew is directed to continue power increase to 100% power. All fuel pre-conditioning requirements for normal power increase have been met. AFW Pump A has been out of service for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> due to inboard bearing vibration. SI Pump A has

-Event No.

-Malf. No. Event I been out of service for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> due to failing to develop the minimum dP during IST surveillance.

I Event

?reload Reactor Trip Breakers Fail to Open on TRIP Signal RP03 AMSAC Actuation Override ON DI-46624-CLOSE Preload Bkr 13301 Bus 33 Supply fails to trip. (CRD MG Set A power source)

D14MY.TRIP dr PTL OFF 01.16621 -CLOSE Preload ON Bkr 1-308 Bus 33 Supply fails to trip. (CRD MG Set A power source)

DI-46621-TRIPk p n om Preload SIOSB SI Pumo B fails to automaticallv start. Manual start is available.

Preload I 1 RBV- 1SOC and RBV-I SOD fail to open on high containment pressure 1 IN BOP 1 Start FW Pump B and resume power increase.

- SRO R RO Follow turbine load increase using rods and/or dilution.

2 RX217 100 I BOP SG B controlling steam flow channel fails high.

3 NIOSD 100 II RO I Power Range Nuclear Instrument channel fails high 1

0,-(611,.TRIPON I M BOP Turbine trips without reactor trip. Operation of (Rod Drive MG Set)

R O I SUOO~V SRO

.. ,breakers to Bus 33 fails.

MSOZA75 M RO SG A steam line rupture inside containment. Delayed for 3 minutes BOP following opening of the reactor trip breakers, and ramped over 5 minutes.

SRO O,dssICIMON DI-IWldCIOsEON c BOP Train B Containment Fan Coil Units Emergency Dampers fail to auto SRO open. Manual control is available.

1 SI055 c RO SI Pump B fails to auto start. Manual start available. (Only SI Pump) r _I

  • (N)ormal, (R)eactivity (I)nstrument, (C)omponent, (M)ajor Transient 4 Rev. A

SCENARIO 1-2 OVERVIEW

- The plant is at 49% power during a startup. Feedwater Pump B is to be started using N-FW-OSA, Feedwater System Normal Operation. The Equipment Operator (EO) is standing by following turnover and as completed local checks of the pump. Once the FW Pump is started the crew should resume the normal power increase to 100% at the rate of !4 %/minute, as allowed by fuel pre-conditioning.

- Following clearly observable plant response from the reactivity changes, the controlling channel for SG B steam flow will fail high. The BOP will take manual control of FW-7B, Main Feed Control Valve, and ensure SG level restores to normal. The channel will be removed from service using A-MI-87. The CRS will address Technical Specification requirements for steam flow channels (TS 3.5.b) in Table TS 3.5-2 and Table TS 3.5-4

- Following completion of the removal of service and Technical Specification review for the failed steam flow channel, Power Range NI channel N-44 will fail high. This will cause control rods to insert in AUTO, and will generate a rod withdrawal rod stop. The RO will respond, after identifying the failed channel, by taking rod control to MANUAL. A-NI-48 will be used to identify and respond to the failure, and A-MI-87 will be entered to remove the failed NI channel from service. The CRS will address Technical Specification requirement for the failed channel (TS 3.5.b) in Table TS 3.5-2.

- Following completion of the removal of service and Technical Specification review for the failed NI channel, an inadvertent turbine trip signal will be generated. The turbine will trip; however, the reactor will not trip. The crew will enter E-0 and during the immediate operator actions take actions to manually trip the reactor and deenergize the buses supply the rod drive MG sets. Neither action will be successful and the CRS will direct transition to FR-S. 1. The immediate actions will be performed including dispatching the Auxiliary Operator ( A 0 to locally trip the reactor trip breakers. Following the completion of step 2 (Verify Turbine Trip) and no sooner than 2 minutes after being notified, the A 0 will locally trip the reactor trip breakers and stop both MG sets. The A 0 will then inform the Control Room of actions completed. The crew will continue with actions of FR-S.1, including establishing charging flow and boration flow path.

The crew should transition to E-0 at step 19 of FR-S. I .

Event 5 . 6 & 7 -Four minutes following the opening of the reactor trip breakers, a steam line break on SG A will initiate and worsen over the next 5 minutes. The crew should recognize the conditions for Safety Injection and manually actuate SI, if it has not automatically occurred. The BOP will report that RBV-15OC and RBV-ISOD, Containment Fan Coil Unit Emergency Dampers (Train B), have not opened when Containment pressure has risen above 4 psig. The SRO will direct the dampers be opened manually (Control Room switch). The RO will recognize that no SI Pumps are running, and after SI sequencing is complete, start SI Pump B. The crew will recognize the conditions indicating SG A is faulted inside containment. (NOTE: Depending on the maximum value for containment pressure and transition time from E-0, entry may be made into FR-2.1 Response to Containment High Pressure, on an ORANGE path for Containment CSF.)

Transition will be made to E-2 to isolate the faulted SG. Following isolation and check of secondary radiation indication, transition will be made to E-I. The scenario terminates following transition to E-I if SG A has completed blowdown, or after evaluation of SI termination criteria in step 12 of E-I, OR at the discretion of the Chief Examiner.

Critical Tasks

1. FR-S. I C - Insert negative reactivity into the core by at least one of the following methods before completing the immediate-action steps of FR-S. 1: [Event 41 Open the Bus 33 and Bus 43 supply breakers to de-energize the Rod Drive MG sets Manually insert Control Rods
2. E-0 I - Establish flow from at least one SI pump before transition out of E-0. [Event 71
3. E-2 A - Isolate the faulted SG before transition out of E-2. [Event SI 5 Rev. A
acility: Kewaunee Scenario No.: 3 Op-Test No.: 1

.xaminers: Operators: SRO RO BOP nitial Conditions: IC-14; 85% power, Beginning of Cycle (BOC)

Motor Driven AFW A Pump is in PULLOUT.

Turbine First Stage Pressure channel PT-486 has failed high. Steam Dump Control is in STM PRESS mode.

rurnover: The plant is at 85% power, returning to 100% power following successful completion of SP-54-086, Turbine Stop and Governor Valve Operability Test. Turbine First Stage Pressure PT-486 failed high about I-hour ago. The actions for A-TB-54, Abnormal Turbine Generator Operation Section 4.12 are complete.

Steam Dumps are in STEAM PRESSURE mode. Continue load increase to 100% at %% per minute load rate. AFW Pump A has been out of service for 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> due to inboard bearing vibration.

-Event No.

P Malf. No.

_r Event Event Preload CS03A Type* - Description Failure to Auto Start ICs Pump A (Containment Spray)

RX223.600 Preload Turbine Impulse Pressure Channel PT-486 fails high. ( 0 0 s )

1 N BOP Resume power increase.

SRO R RO Follow turbine load increase using rods and/or dilution.

2 I BOP SG A Pressure Controlling channel PT-468 fails high. Affects feed flow SRO and SG A PORV opens.

3 C RO Pressurizer PORV PR-2B fails off its seat.

SRO 4 C RO ESF 480V Bus 61 Supply Breaker 16101 trips open and fails.

BOP SRO 5 C BOP Circulating Water condenser bellows failure flooding Turbine Building SRO Basement. Circulating Water Pump fails to auto trip. (Also requires reacto trip.)

6 RD04A M RO Cold Leg A LOCA increasing to loop shear over IO minutes. (Delayed 4 100 BOP minutes following reactor trip.)

SRO I CSO3A C RO Failure to Auto Start ICs Pump A. Must be manually started.

  • (N)ormal,

- (R)eactiv 7 -- SRO 6 Rev. A

SCENARIO 1-3 OVERVIEW m-The plant is at 85% power. Load increase is in progress to 100% power. Crew will resume load increase following turnover.

- Following clearly observable plant response from the reactivity changes, the SG A controlling pressure channel will fail high. This will affect the controlling steam flow channel (density compensation) for the SG, increasing level as the Feed Control valve FW-7A stokes open. Also the SG A PORV SD-3A will open in automatic. The BOP will take manual control of FW-7A and stabilize level. The BOP will also take manual control of S D J A and observe that the valve shuts. A-MI-87 will be entered to remove the failed channel from service and place control to an operable seam flow channel. Once this is completed, the BOP may place FW-7A in AUTO. The CRS will address Technical Specification requirement for the failed channel (TS 3.5.b) in Table TS 3.5-2, Table TS 3.5-3 and Table TS 3.5-4.

- After the SG pressure and steam flow channel has been addressed, Pressurizer PORV PR-2B fails 2% open giving dual light indication. The RO will close the associated PORV Block valve PR-IB. A-RC-36D will be entered to address the RCS leakage. The CRS will address Technical Specification 3.1 .a.5 for PORV operability.

- After the SG pressure and steam flow channel has been addressed, Bus 61 Supply Breaker will fail and trip open, resulting in a loss of power to the bus. ESF equipment affected include Component Cooling Pump B (standby),

Containment Fan Coil Unit C, Containment Fan Coil Unit D and ICs (Containment Spray) Pump B. The crew will respond with actions of A-ELV-40, placing the opening the feeder breakers to the affected equipment (place their control switches in PULLOUT). The CRS will address the affected equipment in Technical Specifications 3.3.c. 1 .A.3 (ICs Pump and CFCUs), and 3 .3.d.2 (CC Pump). The most limiting LCO is 72-hours.

- Once the electrical bus issues have been addressed, a failure of one of the bellows from Circulating Water (CW) to the Condenser will occur. This will result in flooding in the Turbine Building basement and should trip any running CW Pumps. The BOP is expected to check the Alarm Response procedures, and manually trip CW Pump B to minimize flooding. The crew is directed to verify both CW Pumps tripped and to manually trip the reactor. They are to stabilize the plant per E-0 and continue with actions of the A M . These actions include isolating SW to turbine building, placing the secondary pumps in PULLOUT. Further actions are directed in A-MDS-30.

Event 6 & 7 - Four minutes after the reactor trip, a LOCA will occur in RCS Cold Leg A. The LOCA will progress to a loop shear over ten minutes. The crew should return to Step 1 of E-0 and verify Safety Injection actuation. Following completion of sequencing of the ESF equipment, and with containment pressure rising above 23 psig, the RO will report the failure of ICs Pump A to start and will manually start the pump. Transition from E-0 is made to E- I . The operators should stop the RXCPs when SI flow and RCS subcooling conditions are met. The operators should establish charging flow, evaluate conditions for SI termination, and verify recirculation capability. At step 19, the crew will be directed back to step 17 to check conditions for recirculation capability, until RWST level lowers to 37%. If RWST level lowers to 37%,

transition is made to ES-I .3, Transfer To Containment Sump Recirculation, and recirculation is established with Train B RHR. The scenario is terminated following verification of Train B RHR Recirculation flow, Step12 of ES-I .3, or at the completion of the second evaluation of the recirculation capability in E-I if RWST does not lower to 37%, OR at the discretion of the Chief Examiner.

Critical Tasks I, RCS A - Close the block MOV upstream of the stuck-open PZR PORV prior to reaching reactor trip conditions.

2. E-0 N - Establish at least one Train of Containment Spray before transition out of E-0.

If transition to ES- I .3 occurs

3. ES-I .3 A - Transfer to containment sump recirculation and establish recirculation flow with at least one train prior to reaching 4% RWST level.

7 Rev. A

Facility: Kewaunee Scenario No.: 4 (SPARE) Op-Test No.: I

xaminers: Operators: SRO RO BOP lnitial Conditions: IC-12; 100% power, middle of cycle (MOC).

Motor Driven AFW A Pump is in PULLOUT.

rurnover: The plant is at 100% power. Continue normal operation.

OFF SLllOlll CRYWOLP SRO capability)

SER12D BLOCK 3 FWl7B, 2 c BOP Feedwater Pump B bearing failure (vibrations) leading to pump trip. Rapid SRO power reduction R RO Follow rapid power reduction with auto rods and boration.

4 1 FW'7A, loo I M BOP RO SRO I Feedwater Pump A trips resulting in turbine tripkeactor trip.

5 RC 1OA, C RO PORV PR-2A fails open on reactor trip.

100 MS07 6 FW,SB, ,oo C BOP Steam supply piping to turbine driven AFW fails.

sRo AFW Pump B subsequently trips on overcurrent.

I I I ~ I

  • (N)ormal, (R)eactivity (1)nstrument. (C)omponent, (J4)ajor Transient 8 Rev. A

SCENARIO 1-4 (SPARE) OVERVIEW

=-Containment Air Particulate radiation monitor R-1 1 fails high resulting in a Containment Ventilation Isolation signal. The crew will respond using A-RM-45, Abnormal Radiation Monitoring System. The crew will verify automatic actions have occurred and determine the channel has failed. The channel will then be removed from service using N-RM-

45. The CRS will address Technical Specification 3. I .d.5 to ensure adequate containment leak detection channels remain operable.

- Following the addressing of Technical Specification requirements for RCS leak detection systems, Bus 46 will lockout. This will lock out the breakers supplying the bus and prevent the load restoration sequence for the TSC DG from actuating. The crew will address the actions of A-ELV-40,48OV AC Supply Distribution System Abnormal, and place the TSC DG control switch to PULLOUT. The CRS will address the actions of the Technical Requirements Manual (TRM) 3.7.1, and apply the Administrative LCO (ALCO) actions of3.7.1.b.

- After the ALCOs for loss of Bus 46 are addressed, Feedwater Pump B bearing will begin to fail. This will be show by increasing fluctuations in pump amps and by high vibration alarms. The crew will respond using A-FW-OSA, Abnormal Feedwater System Operation. Based on pump conditions and filed reports, the crew should initiate a rapid power reduction to approximately 56% power to remove the feedwater pump from service. The crew should use A-0-03, Rapid Power Reduction, to perform the power reduction. If the power is reduced to the above value before the pump trips (approximately 18 minutes), the crew will stop Feedwater Pump B. If the pump trips during the power reduction, the crew will determine if it can reduce power and stabilize plant conditions or if a reactor trip is required. (NOTE: If the plant trips or the crew determines a reactor trip is required when Feedwater Pump B trips, the following Feedwater Pump A trip will be initiated upon the trip).

Event 4 & 5 - After the plant is stabilized (or tripped), Feedwater Pump A will trip on overcurrent. This will generate a turbine tripireactor trip signal. When the reactor trips, PORV PR-2A will fail open lowering Pressurizer pressure, with SI likely unless the RO recognizes and closes PR-1A prior to reaching the auto SI actuation setpoint.

- Ten minutes following the trip, the steam line supplying the turbine driven AFW pump will break resulting in high Steam Exclusion Area actuation. The crew should isolate the steam supply to the TDAFW Pump (MS-100A and MS-110B). Subsequently, AFW Pump B will trip resulting in a loss of all feed to the SGs. If a Safety Injection has occurred due the failure of the turbine trip, the crew will transition to FR-H.l, Response to Loss of Secondary Heat Sink, at step 14 of E-0, or if SI has not occurred, FR-H. I will be entered upon transition from E-0 to ES-0. I , Reactor Trip Recovery, on a RED PATH condition with less than total flow of 205 gpm to the SGs and level less than 4% in both SGs. The crew will perform the action of FR-H. I to attempt to restore feed flow using a Condensate Pump. The scenario will end upon the establishment of feed flow to at least one SG, OR at the discretion of the Chief Examiner.

Critical Tasks E-0 M -Close the block MOV upstream of the stuck-open PZR PORV by completion of the first step in the ERG network that directs the crew to close the block MOV. (This is expected to occur in E-0 at step 20.)

FR-H.1 A - Establish feedwater flow into at least one SG before either S/G wide range level is < 15% and before RCS pressure increases to > 2335 psig due to Loss of Secondary Heat Sink 9 Rev. A

ES-401 Record of Rejected K/As Form ES-401-4 1 Tier / Randomly Reason for Rejection Group Selected WA 1/1 056A 2.1.33 Replaced due to duplication of KA within Tier 1 of RO Outline.

027A 2.1.33 (Prz Pressure Control System Malfunction)

I maintained. Replacedwith randomly selected KA 056A 2.1.33.11

~~ ~~~

(1 I/ 2 I 060A2.2.22 I Could not write an RO level question for Waste Gas since item is- I(

~

1I I Inot include in KPS Technical Specifications (LCO). ODCMIREMM )I 1I I Idoes contain such statement in the ODC, but items in the ODCM 1 have been the responsibility of SRO-licensed individuals. Due to limited selection in Generic 2.2 and after reviewing the overall outline, A KA was randomly selected from the K1 category I I I(previously none selected in U l ) . Replacement KA is 060AK1.02. 1 112/1 I 004K6.15 I Could not write a question with greater than level of difficulty 1 for 1 I1 I 1 the KA. Plant task does not support RO level question. Replaced 1)

/l2/1 I

I 022K4.02 I

with randomly selected KA 004K6.24.

I Facility fans do not have changeable speeds. Could not write 1 a question with greater than level of difficulty 1 for the KA. I 1

1I I IReplaced with randomly selected KA 022K4.05. II 1 2/1 1 061 2.4.50 I Replaced due to duplication of KA within Tier 2 of the RO Outline. 1 11212 1 071K1.05 001 2.4.50 (Control Rod Drive System) maintained. Replaced with randomly selected KA 061 2.4.4.

I Could not write a question with greater than level of difficulty 1 for l1 the KA. Plant task does not support RO level question. Replaced with randomly selected KA 071K1.04.

2 12 (SRO: WE08 2.1.32 Replaced due to duplication of KA within the RO Outline. In Tier 1.

009E 2.1.32 (Small Break LOCA) and in Tier 2,076 2.1.32 (Service Water System) maintained. Replaced with randomly selected KA WIE08 2.1.14.

ES-401, Page 27 of 33

Tier / Randomly Reason for Rejection Group Selected WA 3 2 1 22 Replaced due to dbplicatlon of the KA witnin the SRO Outline Tler 1 025A 2 1 22 (Loss of RHR) maintaineo Rep ace0 with 1 r-1 - I randomly selected KA 2.1 25.

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ES-401, Page 27 of 33

Kewaunee Power Station Random Generation Technique and WA Suppression Random Generation Techniaue The written examination outline was generated in conjunction with the individual initiating the Point Beach written examination outline. The techniques used was similar to that described in NUREG-1021, Rev. 9, ES-401 Attachment 1. The token used were number @oker) chips that correspond to the topics and K/A available, as appropriate. One individual selected the chip, unseen from a basket, after scrambling of the chips was performed.

K/A Sumression A review was made of applicable systems and system 025, Ice Condenser since Kewaunee Power Station does not have this system. Additionally, following system and EPEiAPE selection, WA with a value less than 2.0 were reviewed and rejected (not placed in the basket for selection), if it was determined that no plant specific priority existed for the item. The WAS that are identified in NUREG-1021, Rev 9, ES-401 Attachment 2 section 1, The remaining Section 2 WAS may be excluded from the random selection process and/or rejected without explanation or justification, were also excluded from the selection process.

There were no NRC comments on the as submitted outlines

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