Browns Ferry Initial Exam 2014-3014 Draft Administrative ... · RS) 1 f\2.O4- 295003 Partial or ... RO SPO 295001AK1.03 Partial or Complete Loss of Forced 3.6 4.1 El El [1 C]

$Page 1: Browns Ferry Initial Exam 2014-3014 Draft Administrative ... · RS) 1 f\2.O4- 295003 Partial or ... RO SPO 295001AK1.03 Partial or Complete Loss of Forced 3.6 4.1 El El [1 C]$
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* Total1 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 7Categories 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 ROand 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 donot apply at the facility should be deleted and justified; operationally important, site-specific systems that arenot included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regardingthe elimination of inappropriate K/A statements.

)O/14... 4. Select topics from as many systems and evolutions as possible; sample every system or evolutionin 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 beselected. 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 topicsmust 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’ importanceratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enterthe group and tier totals for each category in the table above; if fuel handling equipment is sampled in otherthan 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 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 #

ta295001 Partial or Complete Loss of Forced

— ( A .I .03 —

Core Flow Circulation / 1 & 4 — —RS)1f\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 / 7p (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. 0Temp/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 (.5Disturbances / 6

K1A Category Totals: Z) 4 .3I 4II3 Group PointTotal: = 20/7

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ES-401 3 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1Emergency 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! 2R () A Z .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. D2Area Radiation Levels! 9 (5) A2.. P3295034 Secondary Containment

Ventilation High Radiation! 9 — —

295035 Secondary Containment High ( ‘. .A .01Differential Pressure! 5

295036 Secondary Containment HighSump/Area Water Level I 5 — —

500000 High CTMT Hydrogen Conc. ! 5 — —

K/A Category Point Totals: JjJ\]I)Q) j Group Point Total:

21

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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 $) 3SMode

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)A’2.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 . olWater

K/A Category Point Totals: !ZO i2 3 13 1 2 12 I i 3 i3 l Group Point Total: 26/5

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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 DriveMechanism

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 CoolingMode

223001 Primary CTMT and Aux. — — — — —

226001 RHR/LPCI: CTMT Spray Mode

230000 RHR/LPCI: Torus/Pool Spray R (f.) A4, 08Mode

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 A2.09272000 Radiation Monitoring R () K.03

286000 Fire Protection

288000 Plant Ventilation R 03

290001 Secondary CTMT

290003 Control Room HVAC ft (R 14 3v3290002 Reactor Vessel Internals () 1<5, OS

K/A Category Point Totals: Ro L LILI iJT I I Group Point Total: ]cR.c?

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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

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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 performedCode

*

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 correctprocedure 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 theyare 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


Reactor Operator

1. For restoring an HCU to service determines valve position, sequencing, verificationrequirements, 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 thendetermines that unidentified leakage is outside the acceptance criteria.

• 2.1.7 Ability to evaluate plant performance and make operational judgments based onoperating 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 Systemcomponents are in the required position on Control Room Panels for Condensate Systemstartup.

• 2.2.15 Ability to determine the expected plant configuration using design and configurationcontrol 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


• EPIP-5, General Emergency

• Completion of ERO Activation and State Notification in accordance with Appendix B andAppendix C of EPIP-5.

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


Senior Reactor Operator

1. Determine Electrical Safety requirements and the correct procedure to Rack Out 480V SDBU 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 theElectrical Safety PPE and Protective Boundary requirement.

• 2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, hightemperature, 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 theFatigue 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 andMaintenance Rule Availability of RIIRSW and EECW Pumps


• 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 inthat room Inoperable and Unavailable. Determines Technical Specification actions conditionrequired 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 emergencyevolution to save equipment.


• EPIP 15, Emergency Exposure

• Determine amount of time an operator has to perform an emergency evolution due toradiation levels and authorize on the correct form.

• 2.3.4 Knowledge of radiation exposure limits under normal and emergency conditions.Importance SRO 3.7


5. Emergency Action Level Classification


• 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 SRO4.6

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* SafetyFunction

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 safetyfunctions; all 5 SRO-IJ systems must serve different safety functions; in-plant systems and functionsmay 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 thecontrol room: Recirculation pumps IMPORTANCE: RO 3.7 SRO 3.7

• Operator directed to perform dual Recirculation Pump quick restart lAW 2/3-01-68, theoperator will start both Recirculation Pumps.

b. RFPT Trip Recovery (Unit 3)

• New / Simulator / Low Power

• 3-01-3, Reactor Feedwater System

• 259001 Reactor Feedwater System A201. Ability to (a) predict the impacts of the followingon the Reactor Feedwater System; and (b) based on those predictions, use procedures tocorrect, 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 ReactorLevel.

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 onthe 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: Stuckopen 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 thecontrol room proceeds to backup control panel in the Unit 2 simulator and performs actions toclose 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

• 2/3-AOI-74-1 Loss of Shutdown Cooling, 2/3-ARP-9-3D

• 295021 Loss of Shutdown Cooling AA1.02 Ability to operate andlor monitor thefollowing as they apply to Loss of Shutdown Cooling: RHRlshutdown coolingIMPORTANCE: 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 isstarted, the RHR System II Pump B Seal Leakage High alarm will be received. Inaccordance with the ARP the operator will secure RHR B pump and establish cooldownwith 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

• 2/3-EOI Appendix-13 Emergency Venting Primary Containment

• 295024 High Drywell Pressure EA1.19 Ability to operate and/or monitor the following asthey apply to High Drywell Pressure: Containment atmosphere control: Plant-SpecificIMPORTANCE: RO 3.3 SRO 3.4

• The operator vents the Drywell and Suppression Chamber through the hardened suppressionchamber 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 followingon the AC Electrical Distribution; and (b) based on those predictions, use procedures tocorrect, 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 operatortakes actions to open the Generator PCB lAW 2/3 -AOl- 100-1.


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 / ormonitor the following as they apply to Partial or Complete Loss of Instrument Air: BackupAir Supply IMPORTANCE: RO 3.5 SRO 3.3

• Operator crossties CAD to Drywell Control Air JAW 2/3-EOI Appendix-8G. When CADSystem A shows indications of being 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 systemhigh radiation IMPORTANCE: RO 3.7 SRO 4.1

• Operator is directed to respond to Offgas Post-Treatment Radiation HI-HI-HI alarm inaccordance 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 inserta 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 SuppressionChamber Head Tank Pumps

• 295031 Reactor Low Water Level EA1 .08 Ability to operate and / or monitor the followingas 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 thePSC Pumps fail to start, the operator drains the PSC Tank Level switch.


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 thefollowing as they apply to Control Room Abandonment: Control Room / local controltransfer mechanisms IMPORTANCE: RO 4.2 SRO 4.3

• Simulates placing Shutdown Cooling in service from outside the control roomJAW 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 theyapply to Plant Fire on Site: Vital equipment and control systems to be maintained andoperated 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 forinjection, JAW 0-SSI-16. When the 3A RHR Pump fails to start the operator will utilize thenote 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

1N-BOP and 30B, Refuel damper 64-9 fails in mid position when

TS-SRO Refuel Fans are in Off and is Open when Refuel Fansoperating

R-ATC2 Commence power increase with flow to 85%

R-SRO

C-ATC

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

TS-SRO

C-BOP4 Batch File Stator Water Cooling Pump trip

C-SRO

5 fw30a RFPT 3A Governor fails low

I-ATC6 tcl0b

1SR0EHC 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.


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.

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

ANDControl Rod insertion commenced in accordance EOI Appendixes.

4. Feedback:Reactor Power trend.

Control Rod indications.

SLC tank level.

\ \\


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.


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.


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

ANDObservation - 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.


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


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. Refueldamper 64-9 fails in mid position when Refuel Fans are in off and is open when Refuel Fansoperating. 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 torestore power to 480V SD BD 3B, reset RPS, reset PCIS and restore systems. The SRO will alsohave to enter the following AOIs; 3-AOI-l-3, 3-AOI-70-l, and 3-AOJ-99-1. SRO will refer to theTRM and determine Technical Surveillance Requirement 3.4.1.1 to monitor Reactor CoolantConductivity continuously cannot be met and samples must be drawn every 4 hours. SRO willrefer to Tech Spec 3.6.1.3 for failed closed MS1V and enter condition A. SRO will refer to TechSpec 3.4.5 and determine Condition B is required for inoperable containment atmosphericmonitoring 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 restoresystem 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 theATC or Crew notices. The controller will fail to respond until the ATC takes manual control withhandswitch. The Operator will be able to restore RFPT 3A speed in manual. SRO should directentry into 3 -AOI-3 -1.

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

7. With RCIC, CRD and SLC as the only source of high pressure makeup as the LOCA degradesRPV Level will continue to lower. The SRO will determine Emergency Depressurization isrequired 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 EOI’s 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 isaligned for operation.

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

Event Maif. No. Event Type* Event DescriptionNo.

N-BOP1 Return LPRM 8-49B to Operate lAW 2-OI-92B

N-SRO

R-ATC2 Commence power decrease with flow to 90%

R-SRO

C-BOP3 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 standby5 thl8a

C-SRO pump to auto start

C-ATCLOCA - Recirculation Pump 2A Inboard and Outboard seal

6 thlO/1 laR-ATC

failureTS-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 Auto9 Batch I open

10 rcO8 C RCIC Steam Valve fails to Auto open


1

NRC Scenario 5

Critical Tasks - Three

With RPV pressure below the Shutoff Head of the available Low Pressure system(s), operate availableLow 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 injectsinto 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 leveldrops 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 SRV’s 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 maintainedabove -162 inches, inhibit ADS.



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 isolateand numerous alarms will come in. The BOP operator will shift SJAE’s to B or reset SJAE Aand return to service lAW 2-01-66 or 2-AOI-47-3. Reactor Zone Differential pressure low willalarm and the operator will have to reset Refuel and Reactor Zone fans. When one of the SJAE’sare restored high H2 will result in Off Gas, the SRO will evaluate IRM 3.7.2 and enterCondition A. The H202 analyzer will isolate requiring the SRO to evaluate TRM 3.3.11 and3.6.2. 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 toI&C Bus A the alarm will clear but ADS SRV 1-22 will be leaking by and the acoustic monitorwill indicate the leak by. SRO should enter 2-AOl-i-i, the ATC will lower power to less than90%. When power is below 90% the BOP operator will perform 2-AOl-i-i actions to attempt toclose 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 standbypump will fail to start. The ATC will start the standby VFD Cooling Water Pump torestore cooling water preventing a VFD and Reactor Recirc Pump trip.

6. #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 hours 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 steamsupply valve to RCIC will fail to auto open. The Crew will maintain reactor level until theLOCA is beyond the ability of RCIC to control. The SRO will determine ED is required in orderto restore level with available low pressure systems.

8. After the scram 480V Shutdown Board 2A will fail due to a lockout, this will prevent operationof Core Spray Division 1 System for injection. RHR Loop 1 may be used for injection but nothrottle capability with exist. RHR Loop 1 will not be available for Containment coolingoperation.

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

4

NRC Scenario 5


Emergency Depressurization complete

Reactor Level is restored


SCENARIO NUMBER: 5










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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-03SRO 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 4.2SRO S-57C-AB-03

ADS SRV leaking

RO U-001-AB-01 239002 A2.03 4.1 4.2SRO S-001-AB-01

VFD Cooling Water Pump Failure

RO U-068-AL-19 202001 A2.22 3.1 3.2SRO S-068-AB-01

RR Pump Seal Failure

RO U-068-AL-09 202001 A2.10 3.5 3.9SRO S-068-AB-01

Loss of 480V SD BD 2A

RO U-57B-AL-06 262001 A4.05 3.3 3.3SRO S-57B-NO-07

LOCA/Low Level ED

RO U-003-AL-24 295031 EA2.04 4.6 4.8RO U-000-EM-01RO U-000-EM-13SRO S-000-EM-14SRO S-000-EM- 15SRO 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 ModeChange.

Event Maif. No. Event Event DescriptionNo. Type*

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

R-ATC2 R-SRO

Raise power with Control Rods

C-ATC Control Rod will difficult to withdraw, control rod at position3 rd05r3847 C-SRO otherthanOO

4 dOEC-ATC Control Rod will triple notch after drive water header pressure

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

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

5 e 07aC-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



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 - 2-FIC-84-20 in manual and set at 0 SCFM.Observation - 2-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 hourcooldown 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.


Within 10 minutes of recorded time in SSI an Operator has placed the following switches inTest/Inhibit, at Panel 2-9-3: ECCS SYS I HI DW PRESS Test/Inhibit, 2-HS-75-59 AND ECCSSYS 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 ofsubsequent 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 - 2-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 respondlAW 3-01-85. Once Drive water pressure is at 350 psig or greater the control rod will triplenotch 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 attemptsto insert the control rod to the correct position will fail. The control rod will be declared stuckand 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 andthe standby pump will not auto start, BOP operator will start EHC Pump 2B to prevent a lossof 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 RCICinadvertent initiation. The BOP Operator will respond lAW ARPs. BOP Operator will verifythat level is in normal band and secure RCIC. The SRO will evaluate Technical Specification3.5.3 Condition A, 3.3.4.2 Condition A, 3.3.5.1 Condition A, B and F, 3.3.5.2 Condition Aand 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 StationPump 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 SSIguidance.

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

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

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

Reactor Level is maintained

Controlled Cooldown in progress


SCENARIO NUMBER: 6










\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-N0-07SRO 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 3.1SRO S-000-AD-3 1

Control Rod Mispositioned

RO U-085-AB-07 201002 A2.02 3.2 3.3SRO S-085-AB-07

Loss of 480V Unit BD 2A

RO U-57B-AL-06 226001 A2.04 3.8 4.2SRO S-57B-N0-07

RCIC Inadvertent Start

RO U-071-N0-5 217000 A2.01 3.8 3.7SRO S-000-AD-27

SSI FIRE

RO U-000-EM-85 600000 AA2.16 3.0 3.5RO U-000-SS-30RO U-000-N0-32SRO S-000-EM-30SRO S-000-SS-30SRO S-000-SS-3 1

\\,\\


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 andSteam Packing Exhauster 3A.

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

Event Maif. No. Event Type* Event DescriptionNo.

N-BOP1 Inert the Primary Containment in accordance with 3-01-76

N-SRO

R-ATC2 Power increase with Control Rods 30%

R-SRO

I-ATC3 rd25r5035 Failed RPIS position indication on rod 50-3 5 at position 36

I-SRO

C-ATC4 cuo 1 RWCU Leak with failure to Auto isolate

TS-SRO

dg03b C-BOPLoss of EECW C3 Pump through loss of 4KV SID Board 3EB

TS-SRO

C-BOP6 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.


Appendix D Scenario Outline Form ES-D-1


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 radiationrelease 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, operatortakes 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 trendHPCI injection valve open



With RPV pressure below the Shutoff Head of the available Low Pressure system(s), operate availableLow 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 theRPV 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 maintainedabove -162 inches, inhibit ADS.



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.

\



When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe regionof the Drywell Spray Initiation Limit (DSIL) curve and prior to exceeding the PSP limit.


2. Cues:Procedural complianceHigh Drywell Pressure and Suppression Chamber Pressure

3. Measured by:Observation - US directs Drywell Sprays lAW with EOI Appendix 1 7B


4. Feedback:Drywell and Suppression Pressure loweringRHR flow to containment

OR

Before Drywell temperature rises to 280°F, initiate Drywell Sprays while in the safe region of theDrywell Spray Initiation Limit (DSIL) curve.


2. Cues:Procedural complianceHigh Drywell Pressure and Suppression Chamber Pressure

3. Measured by:Observation - US directs Drywell Sprays JAW with EOI Appendix 17B


4. Feedback:Drywell and Suppression Pressure loweringRHR flow to containment


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-69-1will fail to automatically isolate. The ATC will isolate RWCU and take actions lAW3-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, TRMTechnical Surveillance Requirement 3.4.1.1 to monitor Reactor Coolant Conductivitycontinuously cannot be met and samples must be drawn every 4 hours.

5. The crew will respond to a loss of 4KV Shutdown Board 3EB. This will result in a loss of therunning EECW pump C3. The operator will take action to start EECW pump Cl. The SRO willrefer to Tech Specs and initially determine TS 3.7.2 Condition A. Once the Cl EECW pump hasbeen 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 whenthe main turbine is tripped the SRO will enter 3-AOI-47-1. Condenser Vacuum will continue todegrade.

7. Prior to the SRO directing a Reactor Scram on vacuum a Loss of Offsite Power will occur. Thecrew 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 maintainReactor Level and Containment parameters.

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


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



4 Abnormal Events: List (1 -3)






0 YES Technical Specifications Exercised (Yes/No)


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-07SRO 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-03SRO S-047-AB-03

LOOP

295002 AA1.05

RO U.-57A-AB-01RO U-082-AL-07SRO S-57A-AB-01

LOCA

295003 AA1.03

RO U-000-EM-01RO U-000-EM-02RO U-000-EM-05RO U-000-EM-3 1RO U-000-EM-32RO U-000-EM-80SRO S-000-EM-0lSRO S-000-EM-02

295024 EA1.11295031 EA2.04

SRO S-000-EM-05

3.4 3.4

2.2.2

RO U-069-AL-10SRO S-000-EM-12

4.6 4.1

RO U-085-AL-14SRO S-085-AB-04

3.2 3.2

4.4 4.4

4.2 4.24.6 4.8

Documents

Browns Ferry Initial Exam 2014-3014 Draft Administrative ... · RS) 1 f\2.O4- 295003 Partial or ... RO SPO 295001AK1.03 Partial or Complete Loss of Forced 3.6 4.1 El El [1 C]