ML14154A337

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Initial Exam 2014-3014 Draft Administrative Documents
ML14154A337
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 06/02/2014
From:
NRC/RGN-II
To:
Tennessee Valley Authority
Shared Package
ML14155A059 List:
References
50-259/OL-14, 50-260/OL-14, 50-296/OL-14
Download: ML14154A337 (58)
v • d • e


Text

ES-401, Rev. 9 BWR Examination Outline Form ES-401-1 Facility Browns Ferry Date of Exam:

-2013 A?r( 2.0r4-.

RO K/A Category Points SRO-Only Points Tier Group

KKKKKKAAAAG A2 G*

Total 1

2 3

4 5

6 1

2 3

4 Total 1.

i 4

3 3

4 3

20 4

3 7

Emergency &

2

  • T T i

2 1

i 7

2 1

3 Abnormal Plant

NIA N/A

Evolutions Tier Totals 4

6 4

27 6

4 10 1

23322222332 26 3

2 5

2.

2 1

1 1

11 2

11 1

1 1

12 0

2 1

3 Plant

Systems Tier Totals 3

3 4

3 3

4 4

3 38 5

3 8

3. GenericKnowledgeandAbilities 1

2 3

4 10 1

2 3

4 7

Categories 2

3 3

2 1

2 2

2 1.

Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each KIA category shall not be less than two).

2.

The point total for each group and tier in the proposed outline must match that specified in the table.

1 The final point total for each group and tier may deviate by +/- 1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

7,L 3.

Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate K/A statements.

)O/14...

4.

Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

?C 5.

Absent a plant-specific priority, only those K/As having an importance rating (lR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6.

Select SRO topics for Tiers I and 2 from the shaded systems and K/A categories.

7.

  • The generic (G) K/As in Tiers I and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.

8.

On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note

  1. 1 does not apply).

Use duplicate pages for RO and SRO-only exams.

t.

9.

For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

ES-401 2

Form ES-401-1 ES-401 BWR Examination Outline Emergency and Abnormal Plant Evolutions

- Tier 1/Group Form ES-401-1 1

E/APE#/Name/SafetyFunction K

K K

A A

G K/ATopic(s)

IR ta 295001 Partial or Complete Loss of Forced

( A.I.03

Core Flow Circulation / 1 & 4

RS) 1 f\\2.O4-

295003 Partial or Complete Loss of AC / 6

R S)A1.o

($)cZI.7

295004 Partial or Total Loss of DC Pwr I 6

R)AAI.oI

295005 Main Turbine Generator Trip / 3

295006SCRAM/1 R

S.) A4j.03

()Z

).2.o

295016 Control Room Abandonment / 7 p (p.) S2.4.45

295018 Partial or Total Loss of CCW / 8 R

() APt O9.-

295019 Partial or Total Loss of Inst. Air / 8

6 P. ) (.. 4-. I

(.) AIZ. D2

295021 Loss of Shutdown Cooling /4

R

()

295023 Refueling Acc /8

A 1<3.03

295024 High Drywell Pressure / 5 (g)

295025 High ReactorPressure/3

cs) EA.D3

295026 Suppression Pool High Water

) A 1. 0 Temp/S 295027 High Containment Temperature / s : : = : =

295028 High Drywell Temperature / 5 P.

(.R) E K3 t)C,

295030 Low Suppression Pool Wtr Lvi / 5

PS

.2. 0 \\

295031 Reactor Low Water Level 12 R

E 3. o2

and Reactor Power Above APRM R

(p.)

E-4. i. 04-5) 2.

295037 SCRAM Condition Present Downscale or Unknown / 1 295038 High Off-site Release Rate / 9 R

) E < 2.

600000 Plant Fire On Site /8 R

()

,A, o4

$) PAZ. 15 700000 Generator Voltage and Electric Grid R

S )

A41D3

(.5 Disturbances / 6 K1A Category Totals:

Z) 4.3I 4 II3 Group PointTotal:

=

20/7 st-o

fl ES-401, REV 9 T1G1 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME / SAFETY FUNCTION:

IR Ki K2 1(3 K4 1(5 K6 Al A2 A3 A4 G TOPIC:

RO SPO 295001AK1.03 Partial or Complete Loss of Forced 3.6 4.1 El El El [1 C] El El [] [1 []

Thermal limits Core Flow Circulation / 1 & 4 295OO3AP2.O4 Partial or Complete Loss of AC / 6 3.5 3.7 El LI C] LI C] Elfl [i El El El System lineups 295004A41.03 Partial or Total Loss of DC Pwr / 6 3.4 3.6 C] [J [JC] C] [] [] El El El El A.C. electrical distribution 295005AA1.O1 Main Turbine Generator Trip! 3 3.1 3.3 El El [] El C] El ]

C] C] El Recirculation system: Plant-Specific 295016G2.4.49 Control Room Abandonment! 7 4.6 4.4 C] El El El El El Li El El C]

Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

295018AA1.02 Partial or Total Loss of CCW /8 3.3 3.4

[] C] El El El El ll El El El C]

System loads 295019G2.4.1 Partial or Total Loss of Inst. Air / 8 4.6 4.8 El El El El El El El El El El Knowledge of EOP entry conditions and immediate action steps.

295021AA2.02 Loss of Shutdown Cooling / 4 3.4 3.4 El El C] Li C] El El El El El RHR/shutdown cooling system flow 295023AK3.03 Refueling Acc Cooling Mode / 8 3.3 3.6 El C]

El [1 El El El El El El Ventilation isolation 295024G2.4.3 High Drywell Pressure / 5 3.7 3.9 CI El Li El El El U flfl C]

Ability to identify post-accident instrumentation.

295025EA2.O5 High Reactor Pressure / 3 3.4 3.6 El LI El El LI El El El El [}

Decay heat generation Page 1 of 2 07/01/2013 2:00 PM

Generator Voltage and Electric Grid Distrurbancecs ES-401, REV 9 T1G1 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME / SAFETY FUNCTION:

IR 1(1 1(2 1(3 1(4 1(5 1(6 Al A2 A3 A4 G TOPIC:

RO SRO 295026EA1.0l Suppression Pool High Water Temp. /

4.1 4.1 H [1 LI Li C Li LI Li H Li Suppression pool cooling 295028EK3.06 High Drywell Temperature / 5 3.4 LI Li I Li LI Li Li LI Li LI [

ADS 295030EK2.O1 Low Suppression Pool Wtr Lvi / 5 3.8 3.9 Li Li Li Li Li Li Li Li Li Li 1IPCI: Plant-Specific 295031 EK3.02 Reactor Low Water Level / 2 4.4 4.7 J

Core coverage 295037EK1.04 SCRAM Condition Present and Power 3.4 3.6 Li Li Li Li LI Li C Li Li LI Hot shutdown boron weight: Plant-Specific Above APRM Downscale or Unknown

/1 295038EK2.06 High Off-site Release Rate / 9 3.4 3.7 C

LI Li C H Li Li C C Li Process liquid radiation monitoring system 600000AK2.04 Plant Fire On Site /8 2.5 2.6 LI Li Li C Li Li Li Li LI Li Breakers I relays! and disconnects 295006AK1.03 SCRAM /1 3.7 4.0 J Li [1 Li Li Li Li C Li Li H Reactivity control 700000AKtO3 3*3 s-ll Li Li Li Li Li Li LI Li Li Li Under-excitatkin Page 2 of 2 07/01/2013 2:00PM

ES-401, REV 9 T1G1 BWR EXAMINATION OUTUNE FORM ES-401-1 700000G2.4.9 Generator Voltage and Electric Grid Distrurbancecs Knowledge of low power / shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.

KA NAME! SAFETY FUNCTION:

Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 295004G2.1.7 Partial or Total Loss of DC Pwr /6 4.4 4.7 F] Li LI LI LI LI LI F], LI LI Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation.

295006G2.1.20 SCRAM / 1 4.6 4.6 F] U U] LI F] F]

F]

Ability to execute procedure steps.

a1g 29501 9A42.02 Partial or Total Loss of Inst. Air / 8 3.6 3.7 U] LI LI [1 LI LI LI LI LI LI Status of safety-related instrument air system loads (see AK2.1 -AK2.19) 295025EA2.03 High Reactor Pressure / 3 3.9 4.1 F] LI LI LI LI LI H F] H LI Suppression pool temperature 295037EA2.05 SCRAM Condition Present and Power 4.2 4.3 LI LI LI LI LI LI LI i El LI LI Control rod position Above APRM Downscale or Unknown

/1 600000AA2.15 Plant Fire On Site / 8 2.3 3.5 LI L LI LI LI U] LI LI LI LI Requirements for establishing a fire watch 3.8 4.2 LILILILILIL1LILILILI Page 1 of 1 07/01/2013 2:00PM

ES-401 3

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal °lant Evolutions - Tier 1/Group 2 (RD SRO)

= =

==

=

E/APE #1 Name! Safety Function K

K K

A A

G K/A Topic(s)

IR 1

2 3

1 2

295002 Loss of Main Condenser Vac / 3

) Afi2i 295007 High Reactor Pressure / 3 295008 High Reactor Water Level! 2 R

() AZ.02.

295009 Low Reactor Water Level / 2 295010 High Drywell Pressure!5

$)

295011 High Containment Temp /5 295012 High Drywell Temperature / 5

295013 High Suppression Pool Temp.! 5

295014 Inadvertent Reactivity Addition! 1 P)A.1\\2.fl

295015 Incomplete SCRAM / 1

295017 High Off-site Release Rate! 9 295020 Inadvertent Cont. Isolation / 5 & 7 R

c) A 1c3.

295022 Loss of CRD Pumps! 1 295029 High Suppression Pool Wtr Lvl / 5

(\\

295032 High Secondary Containment Area Temperature I 5

295033 High Secondary Containment R S (R) E Ai. D2 Area Radiation Levels! 9 (5) A2.. P3 295034 Secondary Containment Ventilation High Radiation! 9

295035 Secondary Containment High

(..A.01 Differential Pressure! 5 295036 Secondary Containment High Sump/Area Water Level I 5

500000 High CTMT Hydrogen Conc. ! 5

K/A Category Point Totals:

JjJ\\]I)Q) j Group Point Total:

21

ES-401, REV 9 T1G2 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME! SAFETY FUNCTION:

IR Ki K2 KS K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 295008AK2.02 High Reactor Water Level 12 3.6 3.8

[1 fl LI J LI []

[J []

Reactor feedwater system 29501402.4.9 Inadvertent Reactivity Addition / 1 3.8 4.2 LI LI C] LI C] H LI LI C]

Knowledge of low power I shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.

295015AA2.O1 Incomplete SCRAM/i 4.1 4.3 LI LI LI LI LI LI LI J LI C] LI Reactor power 295020AK3.03 Inadvertent Cont. Isolation / 5 & 7 3.2 3.2 j LI I LI LI LI LI LI LI LI LI DrywelL/containment temperature response cetpe. Yc) 295032EK1.03 High Secondary Containment Area 3.5

Secondary containment leakage detection: Plant-Temperature / 5 Specific 295033EA1.02 High Secondary Containment Area 3.7 3.8 LI LI LI LI LI LI II LI LI LI LI Process radiation monitoring system Radiation Levels / 9 295035EA1.01 SecondaryContainment High 3.6 3.6 H LI LI LI LI LI LI LI LI LI Secondary containmentventilation system Differential Pressure / 5 Page 1 of 1 07/01/2013 2:00 PM

ES-401, REV 9 KA NAME / SAFETY FUNCTION:

295002AA2.01 Loss of Main Condenser Vac /3 295010G2.4.20 High Drywell Pressure /5 295033EA2.03 High Secondary Containment Area Radiation Levels /9 T1G2 BWR EXAMINATION OUTUNE IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G RO SRO 2.9 3.1 E1LJL1 3.8 4.3&

2 L1L1E1DC1L1EtJflE1L1 FORM ES-401-1 TOPIC:

Condenser vacuum/absolute pressure Knowledge of operational implications of LOP warnings, cautions and notes.

Cause of high area radiation Page 1 of 1 07/01/2013 2:00PM

ES-401 4

Form ES-401-1 ES-401 BWR Examination O!ttiQeF...

Form ES-401-1

PlantSystems-Thr2/Groupl((RO){SRO)

System #/Name K

K K

K K

K A

A A

A G

K/ATopic(s)

IR 12 34561 23_

203000 RHRILPCI: Injection K S (IZ)

A4-:03

$)

3S Mode 205000 Shutdown Cooling

(2) i.s,o2

206000 HPCI

R R

207000 Isolation (Emergency)

,_) /

Condenser

2O9001LPCS R

(13)Ki.l4

209002HPCS

= = = : :

=::

211000SLC R

CR) M-.04-

212000RPS

()A2.O3

2150031RM R

p)R303

215004 Source Range Monitor

(.R) A3 oa.

215005APRM/LPRM (K) k3,O7 cS)A2.OS

217000 RCIC A40

218000ADS (RRI.o3

223002 PCIS/Nuclear Steam R () /3.0

Supply Shutoff

) &Z,4. J 239002 SRVs R ()

2 4.46

259002 Reactor Water Level

()

14.4-.0I

Control 261000SGTS

($)2.2.4-

262001 AC Electrical P.

()

Distribution 262002 UPS (AC/DC)

R

  • )

(1) Go3

263000 DC Electrical P.

R I

0

() AZ

  • 02.

Distribution (Ps) k,.0) 264000 EDGs

1.

) k,03

300000 Instrument Air

P.

(f A3.0

400000 Component Cooling B) iL2. oa (5) A2. ol Water K/A Category Point Totals:

!ZO i2 3 13 1 2 12 I i 3 i3 l Group Point Total:

26/5 5p-D 3

2.

T2G1 BWR EXAMINATION OUTLINE FORM ES-401-1 ES-401, REV 9 KA NAME / SAFETY FUNCTION:

IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 203000A4.03 RHR/LPCI: lnjection Mode 3.4 3.4 j j Keep fill system 2050001(5.02 Shutdown Cooling 2.8 2.9

[] fl

[]

[] [] [) []

Valve operation 2060001(3.02 HPCI 3.8 3.8

[1 El ] El El U Li El UI []

Reactor pressure control: BWR-23,4 2060001<5.07 HPCI 2.8 2.8 cJ El El El El El LI U U] Cl System venting: BWR-2,3,4 209001 KI.14 LPCS 3,7 3.8 Li [1 C] LI El El [J [] U] []

Reactor vess 21 1000A4.04 SLC 4.5 4.6 U] U] [] [] U] U] U] U] [] [ []

Reactor power 212000A2.03 RPS 3.3 3.5 El U] El El C] U] LI C] C] fl Surveillance testing 2150031(3.03 IRM 3.7 3.7 U] U]

U] U] U] U] C] C] [] []

Rod control and intormation system: Plant-Specific 215004A3.02 Source Range Monitor 3.4 3.3 U] j U] U] U] U] U] U]

U] U]

Annunciator and alarm signals 2150051<3.07 APRM I LPRM 3.2 3.3 U] j U] U] U] C] U] U] El LI Rod block monitor: Plant-Specific 217000A4.08 RCIC 3.7 3.6 C] El fl El El U] El U] U] j C]

System flow Page 1 of 3 07/01/2013 2:00PM

C n

n ES-401, REV 9 T2G1 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME! SAFETY FUNCTION:

IR KI K2 K3 K4 1(5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 217000K2.02 RCIC 2.8 2.9

] J RCIC initiation signals (logic) 218000K1.03 ADS 3.7 3.8

[j jJ

[]

Nuclear boiler instrument system 223002A3.O1 PCIS/Nuclear Steam Supply Shutoff 3.4 3.4 fl System indicating lights and alarms 223002G2.4.31 PCIS/Nuclear Steam Supply Shutoff 4.2 4.1

[] fl [] [] [] [] [] [ []

Knowledge of annunciators alarms, indications or response procedures 239002G24.46 SRVs 4.2 4.2

[J [] []

- fl j J

Ability to verify that the alarms are consistent with the plant conditions.

259002K4.O1 Reactor Water Level Control 3.0 3.1 D E El HE El El El Ensuring adequate NPSH for recirculation pumps: Plant-Specific 261000A1.O1 SGTS 2.9 3.1 El El El El El El El El El El System flow 2620011<4.03 AC Electrical Distribution 3.1 3.4 El El El El El El El El Interlocks between automatic bus transfer and breakers 262002A204 UPS (AC/DC) 3.2 3.4 fl El El El El El El ] El [] El Abnormal battery operation: BWR-I 2620021<603 UPS (AC/DC) 2.7 2.9 El El El 1] Li 1 El El El El El Static inverter 263000A1.O1 DC Electrical Distribution 2.5 2.8 El [1 Li LI LI El l [1 El El El Battery charging/discharging rate Page 2 of 3 07/01/2013 2:00 PM

ES-401, REV 9 T2G1 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME / SAFETY FUNCTION:

IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 263000K2O1 DC Electrical Distribution 3.1 3.4 H

E] H H LI H H [] fl fl Major D.C. loads 264000K6.03 EOGs 3.5 3.7 LI LI H LI Li I H LI LI LI j Lube oil pumps 300000A3.02 Instrument Air 2.9 2.7 H LI LI LI El [1 LI H II LI LI Air temperature 4000O0<2.02 Component Cooling Water 2.9 2.9 LI f LI H LI LI LI H LI LI Li CCW valves n

Page3of3 07/01/2013 2:00PM

fl Th T2G1 BWR EXAMINATION OUTLINE ES-401, REV 9 FORM ES4O1-1 KA NAME / SAFETY FUNCTION:

IR Xl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 203000G2.2.38 RHRILPCI: Injection Mode 3.6 4.5 Li Li Li [] [] r] fl [1 {] fl []

Knowledge of conditions arid limitations in the facility license.

215005A205 APRM I LPRM 3.5 3.6 Li Li [] [] [] [] Li

[] [] []

Loss of recirculation flow signal 261000G2.2.4 SGTS 3.6 3.6

[J Li Li Li Li Li Li [] Li Li ]

(multi-unit) Ability to explain the variations in control board layouts. systems, instrumentation and procedural actIons between units at a facility.

263000A2.02 DC Electrical Distribution 2.6 2.9 Li Li Li Li Li Li Li ] Li Li Li Loss of ventilation during charging 400000A2.01 Component Cooling Water 3.3 3.4

[] [ [] ] Li Li Li [] Li Li Li Loss of CCW pump Page 1 of 1 07/01/2013 2:00PM

ES-401 5

Form ES-401-1 ES-401 BWR Examination Outr Form ES-401-1

=

=

Plant Systems-er2iGrou20 SRO System #1 Name K

K K

K K

K A

A A

A G

K/A Topic(s)

IR 1

234 561 2

34 201001 CRD Hydraulic

R

(R)

201002 RMCS

.) /2.04

201 003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 201006 RWM

) t.I.04-202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 2I4000RPIS

g )cz4..34

215001 Traversing In-core Probe

215002 RBM

(.S)A2.D2.

216000 Nuclear Boiler Inst.

(L.) A 3.01

21 9000 RHR/LPCI: Torus/Pool Cooling Mode 223001 Primary CTMT and Aux.

226001 RHR/LPCI: CTMT Spray Mode 230000 RHR/LPCI: Torus/Pool Spray R

(f.) A4, 08 Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment (1k) A, 02.

23900 1 Main and Reheat Steam 239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure 5

(5) 2.4. 3D Regulator 245000 Main_Turbine_Gen._/ Aux.

256000 Reactor Condensate 259001 Reactor Feedwater k

113.

W.-.4-. oa 268000 Radwaste 271000 Offgas c A2.09 272000 Radiation Monitoring R

() K.03 286000 Fire Protection 288000 Plant Ventilation R 03 290001 Secondary CTMT 290003 Control Room HVAC ft (R

14 3v3 290002 Reactor Vessel Internals

() 1<5, OS K/A Category Point Totals:

Ro L LILI iJT I

I Group Point Total:

]

cR.c?

0 ES-401, REV 9 n

T2G2 BWR EXAMINATION OUTLINE HLiHHLIHLIHLI FORM ES-401-1 KA NAME / SAFETY FUNCTION:

IR Kl K2 K3 1(4 KS KB Al A2 A3 A4 G TOPIC:

RO SRO 201001 1(6.06 CR0 Hydraulic 2.8 2.8 LI LI [] [J

[ [] [] [] [J Component cooling water systems: Plant-Specific 201002A2.04 RMCS 3.2 3.1 LI LI H [] [7 LI LI [] LI Control rod block 2010061(1.04 RWM 3.1 3.2 f] H H [1 [1 LI [J LI LI L] H Steam flow/reactor power P-Spec(Not-BWR6) 214000G2.4.34 RPIS 4.2 4.1 H [] LI LI H LI

[]

Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects 216000A3.01 Nuclear Boiler Inst.

3.4 3.4 Li H H H H LI H H H

Relationship between meter/recorder readings and actual parameter values: Plant-Specific 230000A4.08 RHR/LPCI: Torus/Pool Spray Mode 3.0 2W Li H LI H LI H H LI H II Li Pump/system discharge pressure 234000A1.02 Fuel Handling Equipment 3.3 3.8 H LI El LI LI H LI LI H H Refuel floor radiation levels/ airborne levels 259001K4.02 Reactor Feedwater 2.8 2.9 LI El H iI LI LI H H H LI Li Feedwater heating 2720001(2.03 Radiation Monitoring 2.5 2.8 LI LI LI H LI LI LI [1 LI [El Stack gas Fadiation monitoring system 2880001(6.03 Plant Ventilation 2.7 2.7 H Li H LI LI LI LI LI LI H Plant air systems 2900021(5.05 Reactor Vessel Internals 3.1 3.3 Brittle fracture Page 1 of 2 07/01/2013 2:00PM

n ES-401, REV 9 T2G2 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME / SAFETY FUNCTION:

IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 2900031(3.03 Control Room HVAC 2.9 3.1 LI II I LI [] LI [] LI LI [] [I Control room temperature Page 2 of 2 07/01/2013 2:00 PM

FORM ES-401-1 TOPIC:

Loss or reduction in recirculation system flow (flow comparator): BWR-3,4,5 Knowledge of events related to system operations/status that must be reported to internal orginizations or outside agencies.

Valve closures ES-401, REV 9 T2G2 BWR EXAMINATION OUTLINE KA NAME / SAFETY FUNCTION:

Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G RO SRO 215002A2.02 RBM 3.3 3.3 Li Li Li LI LI Li (] El LI LI 241000G2.4.30 Reactor/Turbine Pressure Regulator 2.7 4.1 LI LI Li Li LI Li LI Li Li LI 271000A2.09 Offgas 2.6 2.8 Page 1 of 1 07/01/2013 2:00PM

1 ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility:

rOlA.M R4kf Date of Exam:

Aic1 Ol4 Category K/A #

Topic IR 2.1. I L,oc.A Swirc.k5, th-5, 1.

2.1.3 (c-k-Conduct 2.1.

of Operations 2123

.Ccc-Qwed-4.4-Subtotal

(*;)

2.2.15 p

kS /ci 1

c/

2.2.22 LCC 5

q-e-+, t-4o 2.

2 2

.3g hvu/it,iuWs I\\

2,12 Equipment Control 2.2.

2.242.

L-Cs I*4.

.4kij i;64 4.7 22.38 c

£la 4._c Subtotal

- C..)

2.3.7 PM) P5 -

/. b-zQ 3, 5 2.3.13 P

-JeJvrcL 3.

2.3.14

/téti Razc Radiation J

Control 2.3.

2.3.

flQL2 p441.dfc 2.3. 13 ILA I

lIst 3

Subtotal

(:)

()

2.4.1 ti1 c/pOA c-s 4C 4.

2.4.45 vz6t 4 Emergency 2.4.

Procedures!

2.4.4Q 45 Subtotal Tier3PointTotal (i5 10 7

RO SRO T3 BWR EXAMINATION OUTLINE ES-401, REV 9 KA NAME! SAFETY FUNCTION:

IR Ki 1(2 K3 1(4 K5 K6 Al A2 A3 A4 G TOPIC:

G2.1.31 Conductof operations 4.6 4.3

[] [] [] []

FORM ES4O1-1 Abfllty to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup.

G2.1.39 Conduct of operations 3.6 4.3 LI Li Li Li [1 [1 11 Fl Li Li Knowledge of conservative decision making practices 62.2.15 Equipment Control 3.9 4.3

[ [ LI []

j []

Ability to determine the expected plant configuration using design and configuration control documentaion 62.2.22 Equipment Control 4.0 4.7 Li Li Li Li Li Li LI LI Li Li Knowledge of limiting conditions for operations and safety limits.

G2238 Equipment Control 3.6 4.5 Li Li Li Li Li

[]

Knowledge of conditions and limitations in the facility license.

G2.3.13 Radiation Control 3.4 3.8 Li Li Li Li Li [] [] []

[]

Knowledge of radiological safety procedures pertaining to licensed operator duties 62.3.14 Radiation Control 3.4 3.8 Li Li Li []F] Li Li LI Li Li Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities 62.3.7 Radiation Control 3.5 3.6 Li Li Li Li Li Li Li Li Li Li Ability to comply with radiation work permit requirements during normal or abnormal conditions 62.4.1 Emergency ProcedureslPlans 4.6 4.8 Li [1 Li Li Li Li Li Li Li Li Knowledge of EOP entry conditions and immediate action steps.

62.4.45 Emergency Procedures/Plans 4.1 4.3 Li Li Li Li Li Li Li Li Li Li Ability to prioritize and interpret the significance of each annunciator or alarm.

Page 1 of 1 07/01/2013 2:00PM

0

/fl ES-40I, REV 9 T3 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME / SAFETY FUNCTION:

IR Kl K2 K3 1(4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO G2.1.23 Conduct of operations 4.3 4.4 fi [J fl rj [j [] fi Ability to perform specific system and integrated plant

procedures during all modes of plant operation.

G2.2.22 Equipment Control 4.0 4.7 fi fi fi fi fi fi j Knowledge of limiting conditions for operations and safety limits.

G2.2.38 Equipment Control 3.6 4.5

[j [] [) fi [] fi [] [] fi [J Knowledge of conditions and limitations in the facility license.

G2.3.13 Radiation Control 3.4 3.8 fi fi El ri fl fl Knowledge of radiological safety procedures pertaining to licensed operatoi duties G2.3.6 Radiation Control 2.0 3.8 El El El El El El Li fi Ability to aprove release permits G2.4.40 Emergency Procedures/Plans 2.7 4.5 El LI El LI fi [J El fl Knowledge of the SROs responsibilities in emergency plan implementation.

G2.4.42 Emergency Procedures/Plans 2.6 3.8

[] fi fi fi El fi [ El El Knowledge of emergency response facilities.

Page 1 of 1 07/01/2013 2:00 PM

ILT 1306 Administrative Topics Outline Form ES-301-1 Facility:

Browns Ferry NPP Date of Examination:

4/28/20 14 Examination Level: RO / SRO Operating Test Number:

1404 Administrative Topic Type Describe activity to be performed Code Conduct of Operations N

2.1.29 For restoring an HCU to service determines valve sequencing, SRO/RO Ala required position, verification requirements, and torque requirements.

N 2.1.26 Determine Electrical Safety requirements and the correct procedure to Rack Out 480V SD BD 2A Compartment 2B.

Conduct of Operations D

2.1.7 2-SR-2 Drywell Floor and Equipment Drain Log Calculation SRO/RO Alb N

2.1.5 NFRESOMs Shift Staffing Equipment Control N

2.2.15 Perform Condensate Panel Lineup Checklist for Control Room SRO/RO A2 D

2.2.37 Maintenance Rule Availability for EECW and RHRSW Radiation Control D

2.3.4 Determine Stay Time under Emergency Conditions and SRO A3 authorize Emergency Plan D

2.4.43 EPIP-5 Appendix B and C Notifications RO/SRO A4 D

2.4.4 1 Emergency Action Level Classification 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 (D)irect from bank (< 3 for ROs; <4 for SROs and RU retakes)

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

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

(S)imulator

ILT 1306 Administrative Topics Outline Form ES-301-1 Reactor Operator 1.

For restoring an HCU to service determines valve position, sequencing, verification requirements, and torque requirements.

New 1-01-85, Control Rod Drive System and NPG-SPP-1O.3, Verification Program For restoring an HCU to service determines valve lineup sequence, required position, verification required and torque requirements.

2.1.29 Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

Importance RO 4.1

2. Calculate the correct Drywell Floor and Equipment Sump leakage using 2-SR-2 Direct from Bank 2-SR-2 Instrument Checks and Observations) (Applicant Handout)

Calculates the correct Drywell Floor and Equipment Sump leakage using 2-SR-2 and then determines that unidentified leakage is outside the acceptance criteria.

2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. Importance RO 4.4

3. Performs Condensate Panel Lineup Checklist for Control Room New 213-01-2 Attachment 2 Performs Condensate Panel Lineup Checklist in order to ensure the Condensate System components are in the required position on Control Room Panels for Condensate System startup.

2.2.15 Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc. Importance RO 4.5

4. EPIP-5, Appendix B and C Activation of ERO and State Notffications Direct from Bank EPIP-5, General Emergency Completion of ERO Activation and State Notification in accordance with Appendix B and Appendix C of EPIP-5.

2.4.43 Knowledge of emergency communications systems and techniques. Importance RO 3.2

ILT 1306 Administrative Topics Outline Form ES-301-1 Senior Reactor Operator 1.

Determine Electrical Safety requirements and the correct procedure to Rack Out 480V SD BU 2A Compartment 2B New 0-GOI-300-2, Electrical and 11-300, Electrical Safe Work Practices Determine correct section of 0-GOI-300-2 that is required to Rack Out the breaker and the Electrical Safety PPE and Protective Boundary requirement.

2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen, and hydrogen). Importance SRO 3.6

2. NFR ESOMs Shift Staffing New Using ESOMs the SRO determines who is available and who would be in violation of the Fatigue Rule lAW NPG-SPP-3.21.

2.1.5 Ability to use procedures related to shift staffmg, such as minimum crew complement, overtime limitations, etc. Importance SRO 3.9

3. Determine the effect that a loss of sump pumps in an RIIRSW room has on Operability and Maintenance Rule Availability of RIIRSW and EECW Pumps Direct from Bank 0-11-346, Maintenance Rule Performance Indicator Monitoring, Trending and Reporting Technical Specification 3.7.1 and 3.7.2 Determines that a loss of both sump pumps in an RHRSW Room makes the three pumps in that room Inoperable and Unavailable. Determines Technical Specification actions condition required 3.7.1 Condition A, B, C, and E required actions A.2, B.1, CA, and E.1.

2.2.3 7Ability to determine operability and/or availability of safety related equipment.

Importance SRO 4.6 4.

Determination of Stay Time and Approving Authority to perform an emergency evolution to save equipment.

Direct from Bank EPIP 15, Emergency Exposure Determine amount of time an operator has to perform an emergency evolution due to radiation levels and authorize on the correct form.

2.3.4 Knowledge of radiation exposure limits under normal and emergency conditions.

Importance SRO 3.7

ILT 1306 Administrative Topics Outline Form ES-301-1

5. Emergency Action Level Classification Direct from Bank EPIP-l and 5 Emergency Classification Procedure and General Emergency Complete Notification Handouts Appendix A 2.4.41 Knowledge of emergency action level thresholds and classifications. Importance SRO 4.6

C ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility:

Browns Ferry NPP Date of Examination:

4/28/2014 Exam Level: RO I SRO-I I SRO-U Operating Test No.:

1404 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including I ESF)

System I JPM Title Type Code*

Safety Function

a. Dual Recirc Pump Quick Restart N, L, S 1
b. RFPT Trip Recovery Unit 3 only N, L, S 2
c. Respond to a Stuck Open SRV Unit 2 only A, M, 5 3
d. Loss of Shutdown Cooling A, D, L, S 4
e. Emergency Containment Venting D, EN, S 5
f. Trip the Turbine, Generator PCB Failure to Open A, L, N, S 6
g. Cross-Tie CAD to Drywell Control Air A, D, S 8
h. Off-Gas Post-Treatment Radiation HI-HI-HI ( RO Only)

P, S 9

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

i. Alternate RPV Injection PSC Pumps D, R, E 2
j. Shutdown Cooling in Service lAW CR Abandonment N, R, E 7
k. Align 480V RMOV BD 3B and Start RHR Pump 3A A, U, E 8

© All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-IJ 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-l I SRO-U (A)lternate path 4-6/2-3 (C)ontrol room (D)irect from bank

<91<81<4 (E)mergency or abnormal in-plant 1I 1I1 (EN)gineered safety feature

- / - 1>1 (control room system)

(L)ow-Power I Shutdown

?1I?1I1 (N)ew or (M)odified from bank including 1 (A)

> 2/> 2/>1 (P)revious 2 exams 3/ 3/ 2 (randomly selected)

(R)CA

>1/>1/>1 (S)imulator

ILT 1306 Control Roomiln-Plant Systems JPM Narrative Control Room Systems:

a.

Dual Recirc Pump Quick Restart (Unit 2 or 3)

  • New / Simulator / Low Power

.2/3-01-68, Reactor Recirculation System

  • 202001 Recirculation System A4.01 Ability to manually operate and/or monitor in the control room: Recirculation pumps IMPORTANCE: RO 3.7 SRO 3.7
  • Operator directed to perform dual Recirculation Pump quick restart lAW 2/3-01-68, the operator will start both Recirculation Pumps.

b.

RFPT Trip Recovery (Unit 3)

  • New / Simulator / Low Power
  • 259001 Reactor Feedwater System A201. Ability to (a) predict the impacts of the following on the Reactor Feedwater System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Pump trip. IMPORTANCE: RO 3.7 SRO 3.7

  • The operator places an Alternate RFPT in service lAW 3-01-3 section 8.1 and restores Reactor Level.

c.

Responds to a stuck Open SRV (Unit 2 Only)

  • Alternate path / Modified from Bank / Simulator
  • 2-AOl-I-i, Relief Valve Stuck Open 239002 Relief I Safety Valves A2.03 Ability to (a) predict the impacts of the following on the Relief I Safety Valves; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Stuck open SRV IMPORTANCE: RO 4.1 SRO 4.2
  • Responds to a stuck open SRV JAW 2-AOl-i-i, when the relief valve fails to close from the control room proceeds to backup control panel in the Unit 2 simulator and performs actions to close SRV.

ILT 1306 Control Roomlln-Plant Systems JPM Narrative d.

Restore Shutdown Cooling (Unit 2 or 3)

  • Direct from Bank / Alternate Path I Low Power / Simulator
  • 295021 Loss of Shutdown Cooling AA1.02 Ability to operate andlor monitor the following as they apply to Loss of Shutdown Cooling: RHRlshutdown cooling IMPORTANCE: R03.5 SRO3.5
  • Operator is directed to restore shutdown cooling following an inadvertent RPS actuation, will commence restoration of shutdown cooling with RHR Pump B. After RHR Pump B is started, the RHR System II Pump B Seal Leakage High alarm will be received. In accordance with the ARP the operator will secure RHR B pump and establish cooldown with RHR D pump JAW with the AOl for loss of Shutdown Cooling.

e.

Emergency Containment Venting (Unit 2 or 3)

  • Direct from Bank / ENgineered Safety Feature / Simulator
  • 295024 High Drywell Pressure EA1.19 Ability to operate and/or monitor the following as they apply to High Drywell Pressure: Containment atmosphere control: Plant-Specific IMPORTANCE: RO 3.3 SRO 3.4
  • The operator vents the Drywell and Suppression Chamber through the hardened suppression chamber vents lAW 2/3-EOI Appendix-13.

f.

Trip the Turbine, Generator PCB Fails to Open (Unit 2 or 3)

  • New / Simulator I Low Power / Alternate path
  • 2/3-AOI-100-1, Reactor Scram
  • 262001 AC Electrical Distribution A2.01 Ability to (a) predict the impacts of the following on the AC Electrical Distribution; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Turbine I generator trip IMPORTANCE: RO 3.4 SRO 3.6

  • Operator trips the Main Turbine and when the Generator PCB fails to open the operator takes actions to open the Generator PCB lAW 2/3 -AOl-100-1.

ILT 1306 Control Roomlln-Plant Systems JPM Narrative g.

Cross-Tie CAD to Drywell Control Air (Unit 2 and 3)

  • Direct from Bank / Simulator / Alternate Path
  • 2/3-EOI Appendix-8G Crosstie CAD to Drywell Control Air
  • 295019 Partial or Complete Loss of Instrument Air AA1.01 Ability to operate and / or monitor the following as they apply to Partial or Complete Loss of Instrument Air: Backup Air Supply IMPORTANCE: RO 3.5 SRO 3.3
  • Operator crossties CAD to Drywell Control Air JAW 2/3-EOI Appendix-8G. When CAD System A shows indications ofbeing depressurized the operator isolates CAD System A.
h. Off-Gas Post-Treatment Radiation HI-HI-HI (RO Only) (Unit 2 or 3)
  • Previous / Simulator
  • 2/3-ARP-9-4C, Window 35
  • 2/3-AOI-66-2, Offgas Post-Treatment Radiation HI-HI-HI
  • 271000 Offgas System A2.04 Ability to (a) predict the impacts of the following on the OFFGAS SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Offgas system high radiation IMPORTANCE: RO 3.7 SRO 4.1
  • Operator is directed to respond to Offgas Post-Treatment Radiation HI-HI-HI alarm in accordance with 2/3-ARP-9-4C Window 35. Then the operator refers to 2/3-AOI-66-2, Offgas Post-Treatment Radiation HI-HI-HI, and perform the actions of 2/3-AOI-66-2 insert a core flow runback and reactor scram. Operator will then shut the MSIVs.

In-Plant Systems:

i.

Alternate RPV Injection PSC Pumps

  • Direct from Bank / Emergency or Abnormal In-Plant / RCA Entry
  • 1-EOI Appendix-7G, Alternate RPV Injection System Lineup Pressure Suppression Chamber Head Tank Pumps
  • 295031 Reactor Low Water Level EA1.08 Ability to operate and / or monitor the following as they apply to REACTOR LOW WATER LEVEL: Alternate injection systems: Plant-Specific IMPORTANCE: RO 3.8 SRO 3.9
  • Simulates aligning PSC Pumps for alternate injection LAW 1-EOI Appendix-7G. When the PSC Pumps fail to start, the operator drains the PSC Tank Level switch.

ILT 1306 Control Roomlln-Plant Systems JPM Narrative j.

Shutdown Cooling in Service lAW CR Abandonment

  • New / Emergency in Plant / RCA Entry
  • 1-AOI-100-2, Control Room Abandonment
  • 295016 Control Room Abandonment AA1.07 Ability to operate and/or monitor the following as they apply to Control Room Abandonment: Control Room / local control transfer mechanisms IMPORTANCE: RO 4.2 SRO 4.3
  • Simulates placing Shutdown Cooling in service from outside the control room JAW l-AOI-l00-2.
k. Align 480V RMOV Board 3B and Start RUR Pump 3A
  • Direct from Bank / Emergency in Plant / Alternate Path
  • 0-SSI-16, Control Building Fire EL 593 through EL 617
  • 600000 Plant Fire on Site AA2.16 Ability to determine and interpret the following as they apply to Plant Fire on Site: Vital equipment and control systems to be maintained and operated during a fire IMPORTANCE: RO 3.0 SRO 3.5
  • Simulates field actions to align 480V RMOV Board 3B and to start RHR Pump 3A for injection, JAW 0-SSI-16. When the 3A RHR Pump fails to start the operator will utilize the note to start RHR Pump 3A with the pushbutton on the breaker.

Facility:

Browns Ferry NPP Scenario No.: NRC 4 Op-Test No.: 1404 Examiners:___________________

Operators: SRO:___________________

ATC:______________

BOP:______________

Initial Conditions: 80% power, RFPT 3B and A3 RHRSW Pumps are tagged out.

Turnover: Alternate Refuel and Reactor Zone Fans lAW 3-OI-30A and 30B. Raise power to 85%

with flow and hold for RFPT 3B repairs.

Event Maif. No.

Event Type*

Event Description No.

Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A 1

N-BOP and 30B, Refuel damper 64-9 fails in mid position when TS-SRO Refuel Fans are in Off and is Open when Refuel Fans operating R-ATC 2

Commence power increase with flow to 85%

R-SRO C-ATC 3

edlOb C-BOP Loss of 480V SD BD 3B TS-SRO C-BOP 4

Batch File Stator Water Cooling Pump trip C-SRO 5

fw30a RFPT 3A Governor fails low I-ATC 6

tcl0b 1SR0 EHC Pressure Transducer failure 7

Batch File M-ALL ATWS 8

Batch File M-ALL LOCA Loss of RPV Water Level 9

hpO7 C

Loss of HPCI 120 VAC Power Supply (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Critical Tasks - Five With reactor scram required and the reactor not shutdown, to prevent an uncontrolled RPV depressurization and subsequent power excursion, inhibit ADS.

1. Safety Significance:

Precludes core damage due to an uncontrolled reactivity addition.

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation unless all required injection systems are Terminated and Prevented.

4. Feedback:

RPV pressure trend.

RPV level trend.

ADS ADS LOGIC BUS A/B 1NHLBITED annunciator status.

With a reactor scram required and the reactor not shutdown, initiate action to reduce power by injecting boron (If still critical with challenge to BuT) and inserting control rods.

1. Safety Significance:

Shutting down reactor can preclude failure of containment or equipment necessary for the safe shutdown of the plant.

2. Cues:

Procedural compliance.

Suppression Pool temperature.

3. Measured by:

Observation

- If operating lAW EOI-I and C-5, US determines that SLC is required (indicated by verbal direction or EOI placekeeping action) before exceeding 110 degrees in the Suppression Pool.

AND RO places SLC A / B Pump control switch in ON, when directed by US.

AND Control Rod insertion commenced in accordance EOI Appendixes.

4. Feedback:

Reactor Power trend.

Control Rod indications.

SLC tank level.

\\\\

\\

Critical Tasks - Five During an ATWS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection (except for CRD, SLC and RCIC) from ECCS and Feedwater until reactor pressure is below the MARFP as directed by US.

1. Safety Significance:

Prevention of fuel damage due to uncontrolled feeding.

2. Cues:

Procedural compliance.

3. Measured by:

Observation

- No ECCS injection prior to being less than the MARFP.

AND Observation

- Feedwater terminated and prevented until less than the MARFP.

4. Feedback:

Reactor power trend, power spikes, reactor short period alanns.

Injection system flow rates into RPV.

With RPV pressure <MARFP, slowly increase and control injection into RPV to restore and maintain RPV level above TAF as directed by US.

1. Safety Significance:

Maintaining adequate core cooling and preclude possibility of large power excursions.

2. Cues:

Procedural compliance.

RPV pressure indication.

3. Measured by:

Observation

- Injection not commenced until less than MARFP, and injection controlled such that power spikes are minimized, level restored and maintained greater than TAF.

4. Feedback:

RPV level trend.

RPV pressure trend.

Injection system flow rate into RPV.

Critical Tasks Five When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSTL) curve and prior to exceeding the PSP limit.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation

- US directs Drywell Sprays JAW with EOI Appendix 1 7B AND Observation

- RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment OR Before Drywell temperature rises to 280°F, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation

- US directs Drywell Sprays lAW with EOI Appendix 1 7B AND Observation

- RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RI{R flow to containment

\\\\)\\\\c

Events 1.

BOP operator will alternate Refuel and Reactor Zone Fans JAW 3-O1-30A and 30B. Refuel damper 64-9 fails in mid position when Refuel Fans are in off and is open when Refuel Fans operating. Tech Spec 3.6.4.2 Condition A, required action A.1 and A.2.

2.

ATC commences power increase 85% using recirculation flow.

3.

The Crew will respond to a loss of 480V SD BD 3B, this will cause a loss of RPS B, loss of 480V RMOV BDs 3B and 3C. The Inboard MSIV A will have inadvertently closed. The crew will need to lower power to meet the main steam line flow guidance JAW 3 -AOI-3 -1. The crew will need to restore power to 480V SD BD 3B, reset RPS, reset PCIS and restore systems. The SRO will also have to enter the following AOIs; 3-AOI-l-3, 3-AOI-70-l, and 3-AOJ-99-1. SRO will refer to the TRM and determine Technical Surveillance Requirement 3.4.1.1 to monitor Reactor Coolant Conductivity continuously cannot be met and samples must be drawn every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. SRO will refer to Tech Spec 3.6.1.3 for failed closed MS1V and enter condition A. SRO will refer to Tech Spec 3.4.5 and determine Condition B is required for inoperable containment atmospheric monitoring equipment.

4.

The running Stator Water Cooling Pump will trip and the standby pump will fail to AUTO start.

The BOP operator will be required to start the standby Stator Water Cooling pump to restore system flow and prevent an automatic Turbine Trip/Reactor scram.

5.

RFPT 3A flow controller will slowly fail low, RFPT 3A speed will continue to decrease until the ATC or Crew notices. The controller will fail to respond until the ATC takes manual control with handswitch. The Operator will be able to restore RFPT 3A speed in manual. SRO should direct entry into 3 -AOI-3-1.

6.

An ATWS will occur on the scram and the power supply to HPCI will fail, leaving RCIC as the only source of high pressure makeup besides SLC and CRD. The crew will insert control rods manually, and maintain reactor level.

7.

With RCIC, CRD and SLC as the only source of high pressure makeup as the LOCA degrades RPV Level will continue to lower. The SRO will determine Emergency Depressurization is required to restore RPV Level. The crew will ED and restore RPV Level with available systems.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:

Control Rods are being inserted Emergency Depressurization complete Reactor Level is restored

SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 4 9

Total Malfunctions Inserted:

List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events:

List (1-3) 2 Major Transients:

List (1-2) 2 EOIs used:

List (1-3) 2 EOI Contingencies used:

List (0-3) 75 Validation Time (minutes) 5 Crew Critical Tasks:

(2-5)

YES Technical Specifications Exercised (Yes/No)

Scenario Tasks TASK NUMBER K/A RO SRO Alternate Reactor and Refuel Zone Fans RO U-30A-NO-02 288000 A4.01 3.1 2.9 Raise Power with Recirc Flow RO U-000-NO-06 202002 A4.07 3.3 3.3 SRO S-000-AD-3 1

Scenario Tasks TASK NUMBER

&Q S+/-Q Loss of 480V SD BD 3B RO U-57B-AL-06 262001 A2.04 3.8 4.2 SRO S-57B-AL-09 Reactor Feed Pump Turbine Governor Failure RO U-003-AL-09 259002 A4.01 3.8 3.6 SRO S-003-AB-01 Stator Water Cooling Pump Trip RO U-35A-AL-02 245000 A4.03 2.7 2.8 SRO 5-070-AB-Ol EHC Pressure Transducer Failure RO U-047-AB-02 241000 A2.03 4.1 4.2 SRO S-047-AB-02 LOCA/Low Level ED RO U-003-AL-24 295031 EA2.04 4.6 4.8 RO U-000-EM-0 1 SRO S-000-EM-14 SRO S-000-EM-15 SRO 5-000-EM-Ol ATWS RO U-000-EM-03 295015 AA2.01 4.1 4.3 RO U-000-EM-22 RO U-000-EM-28 SRO S-000-EM-03 SRO S-000-EM-18

NRC Scenario 5

(

ci1ity:

Browns Ferry NPP PvorniflflrL Scenario No.: NRC 5 Operators:

SRO:_

ATC:

BOP:

Op-Test No.:

Initial Conditions: 100% power, DG D and RBCCW 2B Pump are tagged out. Spare RBCCW Pump is aligned for operation.

Turnover: Return LPRM 8-49B to Operate from a Bypassed Condition lAW 2-OI-92B. Lower Power with flow to 90% for Main Turbine Valve Testing.

Event Maif. No.

Event Type*

Event Description No.

N-BOP 1

Return LPRM 8-49B to Operate lAW 2-OI-92B N-SRO R-ATC 2

Commence power decrease with flow to 90%

R-SRO C-BOP 3

edl8a Loss of I&C Bus A TS-SRO R-ATC 4

adOic TS-SRO ADS SRV 1-22 leaking C-BOP C-ATC VFD Cooling Water Pump 2A trips with failure of the standby 5

thl8a C-SRO pump to auto start C-ATC LOCA

- Recirculation Pump 2A Inboard and Outboard seal 6

thlO/1 la R-ATC failure TS-SRO Two Level instruments fail high tripping Feedwater and HPCI /

7 Batch File M-ALL LOCA / ED on Reactor Level 8

edl0a C

Loss of 480V SD Board 2A RHR and Core Spray Division 2 Injection Valves will not Auto 9

Batch I

open 10 rcO8 C

RCIC Steam Valve fails to Auto open (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 1

NRC Scenario 5 Critical Tasks - Three With RPV pressure below the Shutoff Head of the available Low Pressure system(s), operate available Low Pressure system(s) to restore RPV water level above T.A.F. (-162 inches).

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance.

Pressure below low pressure ECCS system(s) shutoff head.

3. Measured by:

Operator manually starts initiates at least one low pressure ECCS system and injects into the RPV to restore water level above -162 inches.

4. Feedback:

Reactor water level trend.

Reactor pressure trend.

With an injection system(s) operating and the reactor shutdown and at pressure, after RPV water level drops to -162 inches, transition to Emergency Depressurization before RPV level lowers to

-180 inches.

1. Safety Significance:

Maintain adequate core cooling, prevent degradation of fission product barrier.

2. Cues:

Procedural compliance.

Water level trend.

3. Measured by:

Observation - At least 6 SRVs opened

4. Feedback:

RPV pressure trend.

SRV status indications.

2

NRC Scenario 5 Critical Tasks Three To prevent an uncontrolled RPV depressurization when Reactor level cannot be restored and maintained above -162 inches, inhibit ADS.

1. Safety Significance:

Maintain adequate core cooling, prevent degradation of fission product barrier.

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation.

4. Feedback:

RPV pressure trend.

RPV level trend.

ADS ADS LOGIC BUS A/B iNHIBITED annunciator status.

3

NRC Scenario 5 Events 1.

BOP operator will return LPRM 8-49B to Operate lAW 2-OI-92B.

2.

ATC lowers power to 90% using recirculation flow.

3.

The crew will respond to a momentary loss of I&C Bus A. The in-service SJAE (A) will isolate and numerous alarms will come in. The BOP operator will shift SJAEs to B or reset SJAE A and return to service lAW 2-01-66 or 2-AOI-47-3. Reactor Zone Differential pressure low will alarm and the operator will have to reset Refuel and Reactor Zone fans. When one of the SJAEs are restored high H2 will result in Off Gas, the SRO will evaluate IRM 3.7.2 and enter Condition A. The H202 analyzer will isolate requiring the SRO to evaluate TRM 3.3.11 and 3.6.2. The Drywell CAM will isolate requiring the SRO to evaluate Tech Spec 3.4.5.

4.

During I&C Bus A loss, Main Steam Relief Valve open will alarm. When power is restored to I&C Bus A the alarm will clear but ADS SRV 1-22 will be leaking by and the acoustic monitor will indicate the leak by. SRO should enter 2-AOl-i-i, the ATC will lower power to less than 90%. When power is below 90% the BOP operator will perform 2-AOl-i-i actions to attempt to close the SRV. SRO will refer to Tech Specs and determine TS 3.5.1 condition F is applicable.

5.

The VFD Cooling Water Pump for the A Reactor Recirc VFD will trip and the standby pump will fail to start. The ATC will start the standby VFD Cooling Water Pump to restore cooling water preventing a VFD and Reactor Recirc Pump trip.

6.

  1. 1 and #2 recirc pump seal failure ATC will note alarm and report #2 seal carrying full pressure. A short time later seal #2 will fail ATC will note that a small LOCA exists. ATC will trip and isolate A RR Pump lAW with 2-AOI-68-1A. ATC will insert control rods to exit Region 2 of the power to flow map. SRO will determine Technical Specification 3.4.1 Condition A, is applicable again with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to establish single loop conditions.

7.

Level instruments 208A and 208D will fail high, causing a high level trip of Main Turbine, RFPTs and HPCI. RCIC will be the only major source of high pressure injection and the steam supply valve to RCIC will fail to auto open. The Crew will maintain reactor level until the LOCA is beyond the ability of RCIC to control. The SRO will determine ED is required in order to restore level with available low pressure systems.

8.

After the scram 480V Shutdown Board 2A will fail due to a lockout, this will prevent operation of Core Spray Division 1 System for injection. RHR Loop 1 may be used for injection but no throttle capability with exist. RHR Loop 1 will not be available for Containment cooling operation.

9.

With Division 2 Accident logic bypassed RHR and Core Spray will not auto start on any accident signals. The crew will have to manually start pumps and open injection valves. RI-JR Loop 2 will be available for Containment Cooling functions until required for injection.

4

NRC Scenario 5 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:

Emergency Depressurization complete Reactor Level is restored SCENARIO REVIEW CHECKLIST SCENARIO NUMBER:

5 10 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOI entry:

List (1-4) 4 Abnormal Events:

List (1-3) 1 Major Transients:

List (1-2) 2 EOIs used:

List (1-3) 2 EOI Contingencies used:

List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks:

(2-5)

YES Technical Specifications Exercised (Yes/No)

\\

\\

\\\\1 5

NRC Scenario 5 Scenario Tasks TASK NUMBER EQ SEQ Restore an LPRM from Bypass RO U-92B-NO-05 215005 A4.04 3.2 3.2 Lower Power with Recirc Flow RO U-068-NO-03 SRO S-000-AD-31 2.1.23 4.3 4.4 Loss of I&C Bus A RO U-57C-AB-03 262001 A2.04 3.8 4.2 SRO S-57C-AB-03 ADS SRV leaking RO U-001-AB-01 239002 A2.03 4.1 4.2 SRO S-001-AB-01 VFD Cooling Water Pump Failure RO U-068-AL-19 202001 A2.22 3.1 3.2 SRO S-068-AB-01 RR Pump Seal Failure RO U-068-AL-09 202001 A2.10 3.5 3.9 SRO S-068-AB-01 Loss of 480V SD BD 2A RO U-57B-AL-06 262001 A4.05 3.3 3.3 SRO S-57B-NO-07 LOCA/Low Level ED RO U-003-AL-24 295031 EA2.04 4.6 4.8 RO U-000-EM-01 RO U-000-EM-13 SRO S-000-EM-14 SRO S-000-EM-15 SRO S-000-EM-01 6

D

Facility: Browns Ferry NPP Scenario No.: NRC - 6 Op-Test No.:

Examiners:__________________

Operators:

SRO:_________

ATC:

BOP:______

Initial Conditions: 1.3% power, operating in 2-GOI-100-1A Section 5.4 steps 63.3 and 65.

Turnover: Warm RFPT B lAW 2-01-3, section 5.6 and then Continue to pull rods for Mode Change.

Event Maif. No.

Event Event Description No.

Type*

1 Warm RFPT B lAW 2-01-3, section 5.6 R-ATC 2

R-SRO Raise power with Control Rods C-ATC Control Rod will difficult to withdraw, control rod at position 3

rd05r3847 C-SRO otherthanOO 4

dOE C-ATC Control Rod will triple notch after drive water header pressure r

TS-SRO is raised and then will remain stuck, mispositioned control rod d

C-BOP Loss of 480V Unit Board 2A, failure of EHC Pump 2B to auto 5

e 07a C-SRO start 6

rcO 1 One level instrument fails and RCIC inadvertently starts 7

batch M-ALL SSI Fire 25-1 8

hpO3 I

HPCI Flow controller will not operate in Auto 9

batch C

Numerous instrument failures due to SSI Fire (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Critical Tasks Three Within 10 minutes of recorded time in SSI an Operator has placed Path A Vent Flow Controller, 2-FIC-84-20, in MANUAL and 0 SCFM, at Panel 2-9-55.

1. Safety Significance:

Maintaining adequate RHR Pump NPSH.

2. Cues:

Procedural compliance.

Containment Pressure indication.

3. Measured by:

Observation FIC-84-20 in manual and set at 0 SCFM.

Observation FCV-84-20 closed.

4. Feedback:

Containment Pressure trend.

No flow through A vent path.

Within 10 minutes of recorded time in SSI an Operator has initiated a controlled 100°F per hour cooldown rate using HPCI and relief valves as required.

1. Safety Significance:

Prevent Drywell Temperature from exceeding design basis temperature.

2. Cues:

Procedural compliance.

Reactor Pressure indication.

3. Measured by:

Observation

- HPCI in Pressure Control Mode.

Observation

- SRVs opened to lower pressure.

4. Feedback:

Reactor Pressure trend.

Critical Tasks Three Within 10 minutes of recorded time in SSI an Operator has placed the following switches in Test/Inhibit, at Panel 2-9-3: ECCS SYS I HI DW PRESS Test/Inhibit, 2-HS-75-59 AND ECCS SYS II HI DW PRESS Test/Inhibit, 2-HS-75-60.

1. Safety Significance:

Prevent CAS initiation due to actual high Drywell Pressure, and minimize the number of subsequent additional actions (to secure/realign both credited and non-credited pumps).

2. Cues:

Procedural compliance.

No AUTO initiation of ECCS when Drywell Pressure exceeds 2.45 psig.

3. Measured by:

Observation HS-75-59 and 60 in Test/Inhibit.

Observation

- No AUTO initiation on high drywell pressure.

4. Feedback:

ECCS Pumps green lights ON and Red Lights Off.

EVENTS 1.

BOP Operator warms RFPT B lAW 2-01-3 Feedwater System, section 5.6 step 2.2.

2.

ATC Continues Power ascension with control rods.

3.

During power ascension Control Rod 34-3 5 will fail to withdraw. The crew will respond lAW 3-01-85. Once Drive water pressure is at 350 psig or greater the control rod will triple notch to position 14 which is one notch beyond the banked position of 12.

4.

The Unit Supervisor should enter 2-AOI-85-7 for a mispositioned control rod. All attempts to insert the control rod to the correct position will fail. The control rod will be declared stuck and the SRO will enter Tech Specs and determine TS 3.1.3 condition A.

5.

A loss of 480V Unit Board 2A will occur. EHC Pump 2A will trip due to loss of power and the standby pump will not auto start, BOP operator will start EHC Pump 2B to prevent a loss of EHC pressure and closure of Turbine Bypass Valves.

6.

Level transmitter 58D will fail to less than -45 inches. This failure will result in a RCIC inadvertent initiation. The BOP Operator will respond lAW ARPs. BOP Operator will verify that level is in normal band and secure RCIC. The SRO will evaluate Technical Specification 3.5.3 Condition A, 3.3.4.2 Condition A, 3.3.5.1 Condition A, B and F, 3.3.5.2 Condition A and B, and 3.8.1 Condition D.

7.

The crew will respond to a fire and enter 0-AOI-26-1 and SSI 25-1, Intake Pumping Station Pump El. 550, Cable Tunnel to Fire Door 440, RHRSW Room B, RURSW Pump Room D.

The SRO will also enter EOI-1 and 2 and perform actions that do not conflict with the SSI guidance.

8.

Shortly after entering the SSI the crew will commence a controlled cooldown lAW the SSI utilizing HPCI and SRVs, the HPCI flow controller will fail in Auto but will operate in manual.

9.

Numerous instruments fail due to the SSI Fire and spurious equipment operation occurs which the crew will respond to lAW SSI 25-1.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:

Reactor Level is maintained Controlled Cooldown in progress SCENARIO REVIEW CHECKLIST SCENARIO NUMBER:

6 10 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOI entry:

List (1-4) 4 Abnormal Events:

List (1-3) 1 Major Transients:

List (1-2) 2 EOIs used:

List (1-3) 0 EOI Contingencies used:

List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks:

(2-5)

YES Technical Specifications Exercised (Yes/No)

\\j} \\\\

\\\\\\Th

Scenario Tasks TASK NUMBER EQ SEQ Warm RFPT 2B lAW 2-01-3 RO U-003-N0-23 259001 A4.02 3.9 3.7 Raise Power with Control Rods RO U-085-N0-07 SRO S-000-AD-31 2.2.2 4.6 4.1 Control Rod difficult to withdraw from a position other than 00 ROU-085-N0-19 201001A4.04 3.1 3.1 SRO S-000-AD-3 1 Control Rod Mispositioned RO U-085-AB-07 201002 A2.02 3.2 3.3 SRO S-085-AB-07 Loss of 480V Unit BD 2A RO U-57B-AL-06 226001 A2.04 3.8 4.2 SRO S-57B-N0-07 RCIC Inadvertent Start RO U-071-N0-5 217000 A2.01 3.8 3.7 SRO S-000-AD-27 SSI FIRE RO U-000-EM-85 600000 AA2.16 3.0 3.5 RO U-000-SS-30 RO U-000-N0-32 SRO S-000-EM-30 SRO S-000-SS-30 SRO S-000-SS-3 1

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Appendix D Scenario Outline Form ES-fl-i

(

acility:

Browns Ferry NPP Scenario No.: NRC 7 Op-Test No.:

Examiners:_____________________

Operators:

SRO:______________________

ATC:_______________

BOP:________________

Initial Conditions: Reactor Power is 23%, Unit startup is in progress lAW 3-G0I-lO0-1A. EECW A3 and Steam Packing Exhauster 3A.

Turnover: Inert the Primary Containment in accordance with 3-01-76 starting at step 47. Commence a power increase with Control Rods to 30% in accordance with Reactivity Control Plan.

Event Maif. No.

Event Type*

Event Description No.

N-BOP 1

Inert the Primary Containment in accordance with 3-01-76 N-SRO R-ATC 2

Power increase with Control Rods 30%

R-SRO I-ATC 3

rd25r5035 Failed RPIS position indication on rod 50-3 5 at position 36 I-SRO C-ATC 4

cuo 1 RWCU Leak with failure to Auto isolate TS-SRO dg03b C-BOP Loss of EECW C3 Pump through loss of 4KV SID Board 3EB TS-SRO C-BOP 6

ms0 1 Loss of Condenser Vacuum C-SRO 7

edO 1 M-ALL Loss of Offsite Power 8

dgola C

DG 3EA Fails to Auto start 9

th2l M-All LOCA 10 hpO4 C

HPCI Steam Supply Valve fails to auto open.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Critical Tasks

- Five With a primary system discharging into the secondary containment, take action to manually isolate the break.

1. Safety Significance:

Isolating high energy sources can preclude failure of secondary containment and subsequent radiation release to the public.

2. Cues:

Procedural compliance.

Area temperature indication.

3. Measured by:

With the reactor at pressure and a primary system discharging into the secondary containment, operator takes action to manually isolate the break.

4. Feedback:

Valve position indication RPV Level maintained above -162 inches, HPCI has been manually initiated.

1. Safety Significance:

Maintaining adequate core cooling

2. Cues:

RPV level indication

3. Measured by:

HPCI injecting at required flow rate

4. Feedback:

RPV level trend HPCI injection valve open

Appendix D Scenario Outline Form ES-D-1 Critical Tasks Five With RPV pressure below the Shutoff Head of the available Low Pressure system(s), operate available Low Pressure system(s) to maintain or restore RPV water level above T.A.F. (-162 inches).

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance.

Pressure below low pressure ECCS system(s) shutoff head.

3. Measured by:

Operator manually starts initiates available low pressure ECCS systems and injects into the RPV to maintain or restore water level above -162 inches.

4. Feedback:

Reactor water level trend.

Reactor pressure trend.

To prevent an uncontrolled RPV depressurization when Reactor level cannot be restored and maintained above -162 inches, inhibit ADS.

1. Safety Significance:

Maintain adequate core cooling, prevent degradation of fission product barrier.

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation.

4. Feedback:

RPV pressure trend.

RPV level trend.

ADS ADS LOGIC BUS A/B iNHIBITED annunciator status.

\\

Appendix D Scenario Outline Form ES-D-1 Critical Tasks Five When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve and prior to exceeding the PSP limit.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation

- US directs Drywell Sprays lAW with EOI Appendix 1 7B AND Observation - RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment OR Before Drywell temperature rises to 280°F, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation - US directs Drywell Sprays JAW with EOI Appendix 17B AND Observation - RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment

Appendix D Scenario Outline Form ES-D-1 Events 1.

BOP will continue with Inerting Primary Containment lAW 3-01-76 2.

ATC will commence to raise power with control rods to 30%.

3.

Failed RPIS position indication on rod 50-3 5 at position 36, crew will refer to 3-AOI-85-4.

Inserting the rod one notch will restore position indication.

4.

The crew will respond to RWCU alarms indicating a leak and RWCU valve 3-FCV-69-1 will fail to automatically isolate. The ATC will isolate RWCU and take actions lAW 3-A0I-64-2A. The SRO will enter E0I-3 on High Secondary Containment Temperatures, evaluate Tech Spec 3.6.1.3, and determine Condition A must be entered. Also, TRM Technical Surveillance Requirement 3.4.1.1 to monitor Reactor Coolant Conductivity continuously cannot be met and samples must be drawn every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

5.

The crew will respond to a loss of 4KV Shutdown Board 3EB. This will result in a loss of the running EECW pump C3. The operator will take action to start EECW pump Cl. The SRO will refer to Tech Specs and initially determine TS 3.7.2 Condition A. Once the Cl EECW pump has been aligned the SRO will determine TS 3.7.1 Condition A now applies.

6.

Condenser Vacuum will begin to degrade the SRO will initially enter 3-AOI-47-3 and then when the main turbine is tripped the SRO will enter 3-AOI-47-1. Condenser Vacuum will continue to degrade.

7.

Prior to the SRO directing a Reactor Scram on vacuum a Loss of Offsite Power will occur. The crew will respond to the Reactor Scram lAW 3-AOI-100-1 and 0-AOI-57-1A.

8.

During the LOOP DG 3EA will fail to automatically start and will have to be manually started.

9.

Sometime after the LOOP a LOCA will develop requiring the crew to utilize systems to maintain Reactor Level and Containment parameters.

10. The HPCI Steam Supply Valve, 3-FCV-73-16, will fail to OPEN on an automatic HPCI initiation signal.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:

Drywell Sprays initiated Reactor Level is maintained above TAF

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Appendix D Scenario Outline Form ES-B-i C

SCENARIO NUMBER:

7 10 Total Malfunctions Inserted:

List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events:

List (1 -3) 2 Major Transients:

List (1-2) 3 EOIs used:

List (1-3) 1 EOI Contingencies used:

List (0-3) 75 Validation Time (minutes) 5 Crew Critical Tasks:

(2-5) 0 YES Technical Specifications Exercised (Yes/No)

Appendix D Scenario Outline Form ES-D-1 Scenario Tasks TASK NUMBER K/A RO SRO Inert Primary Containment RO U-076-NO-0l Raise Power with Control Rods 223001 A4.03 RO U-085-NO-07 SRO S-000-NO-3 1 RWCU Leak with Failure to Auto Isolate 223002 A3.02 3.5 3.5 Control Rod 50-35 Position failure 214000 A2.01 3.1 3.3 Loss of Condenser Vacuum RO U-047-AB-03 SRO S-047-AB-03 LOOP 295002 AA1.05 RO U.-57A-AB-01 RO U-082-AL-07 SRO S-57A-AB-01 LOCA 295003 AA1.03 RO U-000-EM-01 RO U-000-EM-02 RO U-000-EM-05 RO U-000-EM-3 1 RO U-000-EM-32 RO U-000-EM-80 SRO S-000-EM-0l SRO S-000-EM-02 295024 EA1.11 295031 EA2.04 SRO S-000-EM-05 3.4 3.4 2.2.2 RO U-069-AL-10 SRO S-000-EM-12 4.6 4.1 RO U-085-AL-14 SRO S-085-AB-04 3.2 3.2 4.4 4.4 4.2 4.2 4.6 4.8

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