ML13098A252

From kanterella
Jump to navigation Jump to search

Peach Bottom - Final Outlines (Folder 3)
ML13098A252
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 02/25/2013
From:
Exelon Generation Co
To: Todd Fish
Operations Branch I
Jackson D
Shared Package
ML12328A009 List:
References
TAC U01867
Download: ML13098A252 (24)
v • d • e


Text

ES-401 BWR Examination Outline FORM ES-401*1 Facility Name: Peach Bottom Date of Exam: 04/04/2013 RO KlA Category Points SRO-Only Points Tier Group K K K K K A A A A G Total A2 G* Total 1 2 3 6 1 2 3 4 *

1. 1 3 3 4 3 4 3 20 3 4 7 Emergency &

Abnormal 2 1 1 1 N/A 1 1 N/A 2 7 1 2 3 Plant 4 6 Evolutions Tier Totals 1

4 3

2 5

2 2 2 2 l!J!

3 3 2 3 5

2 27 26 3 2 10 5

2.

Plant 2 1 1 1 1 1 2 1 1 1 1 1 12 1 1 1 3 Systems Tier Totals 4 3 3 3 3 4 4 4 3 4 3 38 5 3 8

3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 10 7 Categories 3 2 2 3 2 2 2 1 Note: 1 Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each 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.

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.

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 Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.
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.
5. Absent a plant-specific priority, only those KlAs 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 1 and 2 from the shaded systems and KIA categories.

7.* The generic (G) KlAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topiCS must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KlAs.

8. On the following pages, enter the KIA 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 (Nole #1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs.

and point totals (#) on Form ES-401-3. Limit SRO selections to KlAs that are linked to 10 CFR 55.43.

ES-401, PagE~ 16 of 33

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

K K K A A Q# E/APE # / Name f Safety Function G KIA Topic(s) IR #

1 2 3 1 2 52 295001 Partial or Complete Loss of Forced Core Flow Circulation f 1 & 4 ..

(\

Individual jet pump flows' Not-BWR-1 &2 3.

(\

57 295003 Partial or Complete Loss of AC /6 System lineups 3.5 1

"' 61..

56 295004 Partial or Total Loss of DC Pwr /6 4.4 1

,SO' AbIlity to locate and operate components, indudlng local controls 49 295005 Main Turbine Generator Trip f 3 Main generator Gontmls 2.7 1 4

0 50 295006 SCRAM f 1 3

Reactor~urbine pres"ure regulating system 3.7 I 1

'.',,,' r 0 "',

45 295016 Control Room Abandonment I 7 i,C"i}, Reactor SCRAM 4.

1 43 295018 Partial or Total Loss of CCW f 8 0 I i Plant operanons 1 2 I 53 295019 Partial or Total Loss of Inst. Air { 8 Instrument air system pressure 3.5 1 55 295021 Loss of Shutdown Cooling /4  ;; 94* Ability to identify posl*accldent instrumentation 3.7 1 I",,', 'OS 0

47 295023 Refueling Acc f 8 Ventilation isolation 3.3 1 3 I,'"'

0 I""",..

40 295024 High Drywell Pressure / 5 Drywell integrity Pieri-Specific 4.1 1

r~1'"

0 48 295025 High Reactor Pressure I 3 Reactorlturbine pressure regulatIng system 3.8 2

295026 Suppression Pool High Water .

44 Suppression pool level 3.5 1 Temp. f 5 .. ,. ;,;

I; fie*""

295027 High Containment Temperature 15 0 0 I 42 295028 High Drywell Temperature 15 .' Drywell spray Mark-I&II 3.7 1 1 ..

51 295030 Low Suppression Pool Wtr Lvl/5 if Suppression pool lemperature 3.9 1

~ ..**....,.,.. ,.

58 295031 Reactor Low Water Level/2 4

0 Steam COOling

~

295037 SCRAM Condition Present and  ;

0 46 Reactor Power Above APRM Downscale or .,'", Hal shutdown boron weight Plant*Specific 3.2 1 4

Unknown 11 " 7/'>"

54 ~95038 High Off-site Release Rate 19

64. Knowledge of EOP "'illgation strategies 3.7 1 00 0

41 ~OOOOO Plant Fire On Site I 8 *. .*.** Fire Fighting 2.9 1 2

700000 Generator Voltage and Electric Grid 0

~I 39 Over-excllation Disturbances I 6 2 KIA Category Totals: 3 3 4 r;= 4 3. Group Point Total:

E5-401, Page 17 of 33

ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO)

K K K A A Q# E/APE # / Name / Safety Function G KiA Topic(s) IR #

1 2 3 1 2 0

62 295002 Loss of Main Condenser Vac I 3 RPS 3.4 1 3

295007 High Reactor Pressure / 3 0 295008 High Reactor Water Levell 2 0 63 295009 Low Reactor Water Levell 2 0 Reactor water cleanup blowdown rate 29 1 3

AbIlity to diagnose and recognIze trends tn an accurate and 64 295010 High Drywell Pressure I 5

04. timely manner utilizing the appropriate control room reference 4.2 1 47 material.

295011 High Containment Temp I 5 0 295012 High Drywell Temperature 15 0 295013 High Suppression Pool Temp. I 5 0 295014 Inadvertent Reactivity Addition 11 0 65 295015 Incomplete SCRAM 11

04. Ability to verify system alarm setpoints and operate controls 4.2 1 50 identified in the alarm response manual 295017 High Off-site Release Rate I 9 0 0

59 295020 Inadvertent Cont. Isolation / 5 & 7 Loss of normal heat sink 3.7 1 1

0 61 295022 Loss of CRD Pumps I 1 Reactor SCRAM 3.7 1 1

295029 High Suppression Pool Wtr LviI 5 0 295032 High Secondary Containment Area 0 60 PCIS/NSSSS 3.6 1 Temperature / 5 4 295033 High Secondary Containment Area 0

Radiation Levels I 9 295034 Secondary Containment Ventilation 0

High Radiation I 9 295035 Secondary Containment High 0

Differential Pressure / 5 295036 Secondary Containment High 0

Sump/Area Water Levell 5 500000 High CTMT Hydrogen Conc. / 5 0 KiA Category Totals: 1 1 1 1 1 2 Group Point Total: 7 ES-401, Page 18 of 33

ES-401 4 Form ES-401-1 ES-401 BWI~ Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO)

Q# System # / Name K K K K K K A A A A G KIA Topic(s) IR #

1 2 3 4 5 6 1 2: 3 4 1

11 203000 RHR/LPCI: Injection Mode Component cooling water systems 3.0 1 0

0 13 205000 Shutdown Cooling Reactor temperatures (moderator, vessel, flange) 3.7 1 6

0 1 Turbine shaft sealing. BWR-2, 3, 4, System pumps BWR 2.8; 9, 19 206000 HPCI 2 2 0 2,3,4 3.7 207000 Isolation (Emergency) 0 Condenser 10 0 209001 LPCS System venting 2.5 1 5

209002 HPCS 0 0

4 211000 SLC SBLC pumps 2.9 1 1

- Nuclear boiler instrumentation; Knowledge of local auxiliary 1,22 212000 RPS 0 04. 3.7; operator tasks during an emergency and the resultant 2 2 3$: 3.8

- operational effects.

16,2 2150031RM 0 0 Reactor power Indication response to rod position 3.7; changes; Up scale or down scale trips 2 2 4 3.7 215004 Source Range Monitor 0 0

7 215005 APRM / LPRM Rod withdrawal blocks 3.7 1 1

0 1 Suppression pool water supply; Condensate storage tank 3.5; 12,2 217000 RCIC .. 2 3 1 level 3.5 0

17 218000 ADS . Reactor pressure 4.2 1 8

223002 PCIS/Nuclear Steam Supply 1 2 Traversing in-core probe system 2.7 1 Shutoff 3 0

3 239002 SRVs SRV solenoids 2.8 1 1

20,2( 259002 Reactor Water Level Control 0 0 Loss of any nurnber of main steam flow inputs, All 3.3; individual component controllers in the automatic mode 2

1 2 3.7 0

5 261000 SGTS Primary containment pressure Mark-I&II 3.2 1 3

15 262001 AC Electrical Distribution 0 Exceeding volt~lge limitations 3.1 1 9

0 8 262002 UPS (AC/DC) Transfer from preferred power to alternate power supplies 3.1 1 1

0 14 263000 DC Electrical Distribution Battery charginq/discharging rate 2.5 1 1

0 01_ Automatic starting of compressor and emergency 18,2 264000 EDGs 3; 3.9 2 1 27 generator; Knowledge of system purpose andlor function 0

25 300000 Instrument Air Main Steam Isolation Valve air 3.1 1 5

0 6 400000 Component Cooling Water Loads cooled by CCWS 2.9 1 1

0 IKIA Category Totals: 131212121212131312 3 2 Group Point Total: 26 ES-401, Page 19 of 33

ES-401*1 5 Form ES-401*1 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (ROj K K K K K Q# KIA TopiC(s) IR #

2 3 4 5 0

28 3 solenoids 3.5 0

201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 0 201006 RWM 0 202001 Recirculation 0 0

ntrol Minimum and maximum pump speed setpoints 2.9 0

0 0

0 3.0 216000 Nuclear Boiler Inst. 0 219000 RHR/LPCI: Torus/Pool Cooling Mode 0 37 223001 Primary CTMT and Aux. ration of system 4.6 226001 RHR/LPCI CTMT Spray Mode 0 36 230000 RHR/LPCI: Torus/Pool Spray Mode 3.6 233000 Fuel Pool Cooling/Cleanup 0 0

3 2.9 0

0 0

0 ux.

8 Reactor/turbine pressure control system Plant~Specmc 3A 0

a 3.7 0

2.7 4

33 271000 Offgas 3.1 272000 Radiation Monitoring 0 286000 Fire Protection 0 35 288000 Plant Ventilation 3.8 290001 Secondary CTMT 0 290003 Control Room HVAC 2

32 290002 Reactor Vessellntemals a 29 2

ES-401, Page 20 of 33

ES-401 2 Form ES-401-1 ES 401 BWR Examination Outline a*

Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (SRO)

A I**

IT K K K A Q# EIAPE # I Name I Safety Function KIA Topic(s) 1 2 3 1 2 I****

295001 Partial or Complete Loss of Forced 0 78 Neutron monitoring Core Flow Circulation / 1 & 4 2 295003 Partial or Complete Loss of AC I 6 0 295004 Partial or Total Loss of DC Pwr 16 I 0 295005 Main Turbine Generator Trip 13 I~< t r:l r

295006 SCRAM I 1 0 I .....

295016 Control Room Abandonment 17 0 82 02.

295018 Partial or Total Loss of CCW I 8 . Ability to apply Technical Specifications for a system. 4.7 1 40

04. Ability to perform without reference to procedures those 80 295019 Partial or Total Loss of Inst. Air 18 ~.,."' '""' req"'"

components and controls.

"'mod'.' op","" ~ .,",m 4.4 1 Knowledge of annunciator .

79 295021 Loss of Shutdown Cooling 14 I 4.1 1 i* ... response procedures.

h*'

()

76 295023 Refueling Acc I 8 Fuel pool level I 3.7 1 2

04~ Ability to verify that the alarms ar 81 295024 High Drywell Pressure I 5 1 46 plant conditions.

I--

295025 High Reactor Pressure I 3 I--

295026 Suppression Pool High Water Temp. . ...

5 I--

295027 High Containment Temperature I 5 0 295028 High Drywell Temperature 15 .0 0

()

L=Jt95030 Low Suppression Pool Wlr Lvii 5 Reactor pressure 3.9 1 3

.? .*.*)

1295031 Reactor Low Water Level 12 ..... 0 i ..

1295037 SCRAM Condition Present land Reactor Power Above APRM I I ., ..

. 0 Downscale or Unknown 11 295038 High Off-site Release Rate I 9 0 600000 Plant Fire On Site I 8 0 700000 Generator Voltage and Electric Grid 0

Disturbances 16 KIA Category Totals. 0  :!m:Group Point Tolal: I 7 ES-401, Page 17 of 33

ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (SRO)

~

IAPE # I Name I Safety Function K K A A #

2 3 1 2 G KIA Topic(s} IR I

Loss of Main Condenser Vac I 3 ..... 0

  • :c High Reactor Pressure I 3
> 0 295008 High Reactor Water Levell 2 0 295009 Low Reactor Water Levell 2

...... - ~

295010 High Drywell Pressure 15 0 1 High Containment Temp 15 0

'.~i.

83 !295012 High Drywell Temperature 15 Drywell temperature 3.9 1 295013 High Suppression Pool ,~ ... t'. '-' 0 295014 Inadvertent Reactivity Addition 11 0 1:< '.:,**:1 85 295015 Incomplete SCRAM 11 ***.f~i Knowledge

.~

of RO tasks performed outside the main control room during an emergency and the resultant operational effects 1---:

295017 High Off-site Release Rate 19 0 r- i 295020 Inadvertent Cont. Isolation I 5 & 7 0 I--

1295022 Loss of CRD Pumps 11 0 295029 High Suppression Pool Wtr Lvl/5 0 295032 High Secondary Containment Area '" !*..

0 emperature I 5 I--

295033 High Secondary Containment Area 0

Radiation Levels I 9

- 295034 Secondary Containment Ventilation 0

High Radiation I 9 35 Secondary C l1il~!; 0 ntial Pressure I 5 6 Secondary 01: 1 AbIlity to interpret and execute procedure steps 4.6 p/Area Water Levell 5 @

i:';

500000 High CTMT H '-' 0 KIA Category Totals: 0 0 0 0 Total: 3 ES-40 1, Page 18 of 33

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems Tier 2/Group 1 (SRO)

--~--~-~----------------~--~--~i K

Q# System # I Name KIA Topic(s) IR #

03000 RHR/LPCI: Injection o 205000 Shutdown Cooling Mode o 206000 HPCI o 207000 Isolation (Emergency)

Condenser 209001 LPCS 209002 HPCS o Inadequate system fiow 4

04, Ability to verify system alarm setpoIOts and operate controls

'50' identified In the alarm response manual 4.0 Ability to prioritize and interpret the signjficance of each 2150031RM 4.3 annunciator or alarm 215004 Source Range Monitor o 215005 APRM I LPRM o 217000 RCIC o 18000 ADS Large break LGCA 3.6 223002 PCIS/Nuclear Steam Supply Shutoff o

239002 SRVs o 25!}OQ!2 F~eaictor Water Level Control o 261000 SGTS Q 262001 AC Electrical Distribution o 262002 UPS (AC/DC) o 87 263000 DC Electrical Distribution 264000 EDGs 113000()0 li1strum,entAi*r o II4C)OO,OOCol'l1pl:meint ICooling Water o o

Group Point Total: 5 ES-401, Page 19 of 33

ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (SRO)

~ System # / Name

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ KIA Topic(s)

~

- 201001 CRD Hydraulic 201002RMCS

!  ;::c:

0 0

Control Rod and Drive Mechantsm 0 m

RSCS 0 f--

RCIS '<, a f-

)1006 RWM  !;;< 0 r---

I I rr=

202001 Recirculation 0 f-202002 Recirculation Flow Control f-

~04000 RWCU ," 0 f-

~14000 RPIS 0 Traversing In-core Probe 5002 RBM

~ ";;

0 0

~16000 Nuclear Boiler Inst 0

~19000 RHRlLPCI: I oruswool Cooling 0

IMode

~23001 Primary CTMT and Aux, 0

~26001 RHR/LPCI: CTMT Spray Mode a boooo RHR/LPCI: Torus/Pool Spray Mode 0

~33000 Fuel Pool Cooling/Cleanup 0 92 ,1234000 Fuel Handling Equipment

~ Fuel orientation 137 1

~39001 Main and Reheat Steam 0

- 1239003 MSIV Leakage Control 0

241000 ReactorfTurbine Pressu 0

245000 Main Turbine Gen, / Aux, 0 1256000 Reactor Condensate 0

~59001 Reactor Feedwater 0

- 268000 Radwaste 0

- 1°,1 IA~jfjtY to interpret reference matenais, such as 93 2710000ffgas 4,2 1 Itables, etc, Radiation Monitoring 0 Fire Protection 0 91 288000 Plan! Ventilation ri Low reactor water leveL Plant-Specific 3.6 1 290001 Secondary CTMT 0 290003 Control Room HVAC 0 1290002 Reactor Vessel Internals 0 I~'

0 0 0 0 0 1 0 0 1 0 0 IT:

ES-401, Page 20 of 33

ES-401 Generic Knowle(tge and Abilities Outline (Tier 3) Form ES-401-3 2.1, to explain and apply system limits and precautions, 4.0 1 rd,nate personnel activities outside the control room 3.4 pret and execute procedure steps 4.6 1 75 2.1 Knowledge of facility requirements for controlling vital/controlled access. 2.5 1 95 Knowledge of the process for making design or operating changes to the facility. 3.2 1 Knowledge of the process for controlling equipm.ent configuration or status 4.3 1 Ability to Interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions 4.2 1 IKrlowiled\je of the process for managing maintenance activities during shutdown IODoaretiorl*., such as fisk assessments, work priOritization, etc 2.6 1 radiation Or contamination hazards that may arise during normal. abnormal, conditions or activities 3,8 1 r= liKnOWle<:llle of radlation monitoring systems, such as fixed radiation monitors and alarms, instruments, personnel monltonng equipment, etc 3,1 radiation or contamination hazards that may anse during nonm,lI, abnormal, 3.4 1 1

abnormal indications for sys1em operating parameters that are entry for emergency and abnormal operating procedures.

1 1

Emergency Procedures 1 Plan ~_--!JOlQ.!lli!IiI:lmll~!l!!i!il;~il!!!!!!l!£lli:!!:J:..!l!!li~;om!rJ<w!!!"--_ _ _ __

2.4, ES-40 1. Page 26 of 33

ES-401 Record of Rejected KlAs Form ES-401-4 Tier I Randomly Reason for Rejection Group Selected KIA R02/1 300000 Unable to write discriminating question for this KIA..

0#19 A4.01 (Replaced with 206000 A4.10)

R02/1 212000 PBAPS does not have RPS PAM instrumentation.

0#22 2.4.3 (Replaced with 212000 2.4.35)

Unable to construct an SRO question for this KIA that meets the SRO 1/2 295036 requirements of NUREG-1021 0#84 2.1.30 (Replaced with 295036 2.1.20)

Unable to construct an SRO question for this KIA that meets the SRO 2/1 215003 requirements of NUREG-1 021 0#88 2.4.35 (Replaced with 215003 2.4.45)

Unable to construct an SRO question for this KIA that meets the SRO 2/1 205000 requirements of NUREG-1 021 0#90 A2.11 (Replaced with 211000 A2.04)

Unable to construct an SRO question for this KIA that meets the SRO 2/1 234000 requirements of NUREG-1021 0#92 2.4.4 (Replaced with 234000 K5.05)

Unable to construct an SRO question for this KIA that meets the SR03 2.3.4 requirements of NUREG*1021 0#96 (Replaced with 2.3.14)

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Peach Bottom Date of Examination: 03/25/2013 Examination Level: RO 0 SRO [gI Operating Test Number: 2013 NRC Administrative Topic Type Describe activity to be performed (See Note) Code*

Conduct of Operations D,R G2.1.34 (3.5) - Review And Evaluate Reactor Coolant System Chemistry Limits - Condenser Tube Leak at Power (PLOR-259C)

Conduct of Operations D, R G2.1.32 (4.0) - Evaluation Of High CRD Temperature On Control Rod Scram Time (PLOR 347CA)

Equipment Control N,R G2.2.40 (4.7) - Compensatory Actions for an Inoperable Fire Door (NEW)

Radiation Control D,R G2.3.13 (3.8) - Review And Approve Primary Containment PurgeNent Isolation Valve Cumulative Hour Log (PLOR 256C)

Emergency Plan D,R G2.4.40 (4.5) - Make EAL Classification And State/Local Notifications For SITE AREA EMERGENCY - Loss of Two Fission l:lroduct Barriers (PLOR-230C)

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from banI< C:: 3 for ROs; .:: 4 for SROs & RO retakes)

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

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

ES 301, Page 22 of 27

ES-301 Administrative Outline Form ES-301-1 Facility: Peach Bottom Date of Examination: 03/25/2013 Examination Level: RO [g] SRO 0 Operating Test Number: 2013 NRC Administrative Topic Type Describe activity to be performed (See Note) Code*

G2.1.29( 4.1) Lineup Standby Gas Treatment System Conduct of Operations D,S For Automatic Operation Alternate Path, Control Switches Are Out of Position (PLOR 337CA)

G2.1.5 (2.9) Evaluate Overtime Work Request (PLOR Conduct of Operations N,R 279C)

G2.2.41 (3.5) - Isolating the 3B RBCCW Heat Exchanger Equipment Control N,R Due to;:1 Leak (PLOR 274C)

D,R,P Radiation Control (2011 G2.3.14 (3.4) Perform Reactor Coolant Leakage NRC) Surveillance (PLOR 244C)

Emergency Plan N/A Not Required NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class{R)oom

{D)irect from bank (.::: 3 for ROs; < 4 for SROs & RO retakes)

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

{P)revious 2 exams {.::: 1; randomly selected}

ES 301, Pa~le 22 of 27

ES-301 Control Room/In-Plant "'U'I!,..aoW\Q Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 03/25/2013

  • Exam Level: RO I:8J SRO-I D SRO-U D Operating Test Number: 2013 NRC Control Room Systems@ (S for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function

a. 233000 A2.02 (3.1/3.3) - Fuel Pool Cooling and Cleanup / HPSW A, L, N, S 9 Injection into the Fuel Pool {Alternate Path - HPSW Pump Overcurrent, Use Other Pum~} (NEW)
b. 206000 A2.09 (3.5/3.7) - High Pressure Coolant Injection 1 Raise HPCI A. D, EN, P, 2 Flow (Alternate Path - Suction Valves Fail to Swap on Low Condensate S Storage Tank Level) (PLOR-333CA)
c. 239001 A4.01 {4.2/4.0} - Main Steam System 1 Open Main Steam D, L, S 3
  • Isolation Valves After a Group-1 Isolation (PLOR-OS3C)
d. 209001 A4.04 (2.9/2.9) - Core Spray System / Perform Pump Capacity A,D,EN,S 4 Test For 1ST (Alternate Path - Min Flow Valve Fails To Open) (PLOR 335CA)
e. 223002 A4.03 {3.6/3.S} - Primary Containment Isolation System / D, EN, L, S 5 Perform a Grou~ 1 PCIS Isolation Reset (GP-SA) (PLOFt-024C)
f. 262001 A4.04 (3.6/3.7) - AC Distribution / Excite the Main Generator D,S 6 (PLOR-031C)
g. 212000 A4.14 (3.S/3.S) - Reactor Protection System / Reset a Full D, EN, L, S 7 Scram (PLOR-004C)
h. 400000 A4.01 (3.113.0) - Component Cooling Water I Verify Isolation Of A,D,S S Drywell Chilled Water And RBCCW (Alternate Path - RBCCW Is
0"',....

Su~~I~ing Drywell Chilled Water Loads)- (PLOR-310CA)

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

i. 217000 A4.07 (3.9/3.S) - Reactor Core Isolation Cooling / Defeat RCIC D,E,R 2 Interlocks lAW T-251-2 (PLOR156P)
j. 218000 K4.04 (3.5/3.6) - Bypass of SV-9130A and B lAW T-331-3 (NEW) N,E, R 3
k. 2S6000 A4.06 (3.4/3.4) - Fire Protection System I Diesel Driven Fire D 8

_pum~ Manual Start(PLOR-327PA}

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO I SRO-l / SRO-U (A)ltemate path 4-6 1 4-6 1 2-3 (C)ontrol room (D)irect from bank  :::9/:::8/:,:4 (E)mergency or abnonmal in-plant =.1 1=.1/=.1 (EN)gineered safety feature I - I G 1 (control room system)

(L)ow-Power I Shutdown =.1 I =.11 =.1 (N)ew or (M)odified from bank including 1(A) =-2/=-2/=.1 (P)revious 2 exams  :': 3 1 :': 3 I ::: 2 (randomly selected)

(R)CA

=-1 I=-1 I=.1 (S)imulator ES-301, Pa9e 23 of 27

ES-301 Control Room/In-Plant <IO:u,,,,b,.,rn .. Outline Form ES-301-2 Facility: Peach '3~ JJ Date of Examination: 03/25/2013 Exam Level: RO D SRO-I ~ SRO-U D Operating Test Number: 2013 NRC tR.

r SRO-I); (2 or 3 for SRO-U, including 1 ESF)

~

System I JPM Title Type Code*

a. 233000 A2.02 (3.1/3.3) Fuel Pool Cooling and Cleanup I HPSW A, L, N, S 9 Injection into the Fuel Pool (Alternate Path - HPSW Pump Overcurrent, Use Other Pump) (NEW)
b. 206000 A2.09 (3.5/3.7) - High Pressure Coolant Injection / Raise HPCI A, D, EN, P, 2 Flow (Alternate Path - Suction Valves Fail to Swap on Low Condensate S Storage Tank Level} (PLOR-333CA}
c. 239001 A4.01 (4.2/4.0) - Main Steam System I Open Main Steam D,L,S 3 Isolation Valves After a Group-1 Isolation (PLOR-083C)
d. 209001 A4.04 (2.9/2.9) - Core Spray System 1 Perform Pump Capacity A,D,EN,S 4 Test For 1ST (Alternate Path - Min Flow Valve Fails To Open) (PLOR 335CA)
e. 223002 A4.03 (3.6/3.5) - Primary Containment Isolation System 1 D, EN, L,S Perform a Group 1 PCIS Isolation Reset (GP-8A) (PLOR-024C) f.
g. 212000 A4.14 (3.8/3.8) - Reactor Protection System / Reset a Full D, EN, L, S 7 Scram (PLOR-004C)
h. 400000 A4.01 (3.113.0) Component Cooling Water I Verify Isolation Of A, D,S 8 Drywell Chilled Water And RBCCW (Alternate Path - RBCCW Is Supplying Dryvvell Chilled Water Loads)- (PLOR-310CA)

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

i. 217000 A4.07 (3.9/3.8) Reactor Core Isolation Cooling I Defeat RCIC D,E, R 2 Interloc~l:;If\YY T-251-2 {PLOR156P)
j. 218000 K4.04 (3.5/3.6) - Bypass of SV-9130A and B lAW T-331-3 N, E, R 3 (NEW)
k. 286000 A4.06 (3.4/3.4) Fire Protection System / Diesel Driven Fire D 8 Pump Manual Start (PLOR-327PA)

@ All RO and SRO-I control room (and in-plant) systems functions; all 5 SRO-U systems must serve different overla those tested in the control room.

  • Type Codes Criteria for RO / SRO-II SRO-U (A)lternate path 4-6 1 4-6 1 2-3 (C)ontrol room (D)irect from bank ~9/~8/~4 (E)mergency or abnormal in-plant .::-.1/.::-.1/.::-.1 (EN)gineered safety feature
  • 1 - I ~ 1 (control room system)

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

(R)CA .::-.1/.::-.1/.::-.1 (S)imulator ES-301, Page 23 of 27

ES-301 Control Roomlln-Plant ~U'l:!.rArnC! Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 03/25/2013 Exam Level: RO D SRO-I D SRO-U [8J Operating Test Number: 2013 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

S'y<><c.111 JPM Title Type Cod F

a. 233000 A2.02 (3.1/3.3) - Fuel Pool Cooling and Cleanup I HPSW A,L, N,S 9 Injection into the Fuel Pool (Alternate Path - HPSW Pump Overcurrent, Use Other Pump) (NEW) b.

c.

d. 209001 A4.04 (2.9/2.9) - Core Spray System / Perform Pump Capacity A,D,EN,S 4 Test For 1ST (Alternate Path - Min Flow Valve Fails To Open) (PLOR 335CA) f.

g.

h. 400000 A4.01 (3.113.0) - Component Cooling Water / Verify Isolation Of

. A, D, S 8 Drywell Chilled Water And RBCCW (Alternate Path - RBCCW Is

~I lIy Drywell Chilled Water Loads)- (PLOR-310CA) t Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. 217000 A4.07 (3.9/3.8) - Reactor Core Isolation Cooling / Defeat RCIC D,E, R 2 Interlocks lAW T-251-2 (PLOR156P)
j. 218000 K4.04 (3.5/3.6) - Bypass of SV-9130A and B lAW T-331-3 (NEW) I'J,E,R 3 k.

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank ~9/~8/~4 (E)mergency or abnormal in-plant .::: 1 I .::: 1 I .::: 1 (EN)gineered safety feature - I - / ;:: 1 (control room system)

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

(R)CA .:::1/~1/~1 (S)imulator E~ -301, Pa9 e 23 of 27

Scenario Outline ES-D-l Simulation Facility Peach Bottom Scenario No. #1 OpTest No. 2013 NRC Examiners Operators CRS (SRO) lIRO (ATC)

PRO (BOP)

Scenario The scenario begins with the reactor at 100% power.

Summary Following shift turnover, the crew will stroke Main Steam Sample Valves AO-2-02-316 and 317 as part of a surveillance test for primary containment isolation valves. Shortly after stroking the valves, Reactor Building to Torus vacuum breaker isolation valve AO-2S02A will fail partially open requiring the crew to declare the valve inoperable per Technical Specifications.

Next, the running Service Water pump will trip on overcurrent, requiring the crew to place the standby pump in service using the system operating procedure. Following this, a drywell pressure instrument will fail upscale without causing the expected half scram. The crew will apply Tech Specs and (with time compression) insert a half scram lAW GP-25 "Ins,tallation of Tripsllsolations to SatiSfy Tech SpeclTRM Requirements" .

Next the 'A' Condensate pump will trip without the expected Recirc System runback. Power must be manually reduced using recirc flow to prevent a low-level scram.

When conditions have stabilized, #2 Auxiliary Bus will trip on overcurrent, causing a loss of the remaining Condensate pumps. HPCI and RCIC will initiate on low RPV level. The HPCI system flow controller will fail in automatic and must be adjusted in manual to allow the system to inject. The HPCI system will trip shortly after it injects and will not be recoverable. An RPS failure will prevent the automatic and manual scrams, requiring entry into T-101 "RPV Control" and the use of Alternate Rod Insertion (ARI) to shutdown the reactor. A small Reactor coolant leak will occur in the drywell and require the use of containment sprays. The crew should enter T-102 "Primary Containment Control". A containment spray logic failure will complicate the crew's efforts to spray containment. The crew will not be able to spray containment with the initial loop of RHR selected. The other loop of RHR will be available and should be used to spray containment.

The reactor coolant leak inside the drywell will be greater than the capacity of RCIC (the only remaining high-pressure feed source). The crew should enter T-111 "Level Restoration". As level deteriorates, the crew should start available low pressure ECCS pumps and when it is determined that level cannot be restored and maintained above -172 inches, the reactor should be depressurized in accordance with T-112 "Emergency Blowdown". Low pressure ECCS will be available to recover reactor level. The scenario will be terminated when the reactor has been depressurized and reactor level has been recovered and controlled .

. Initial IC-118, 100% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Event No. No. Type* Description 1 See Scenario Guide N PRO Stroke time primary containment isolation valves for surveillance CRS testing 2 See Scenario Guide TS CRS Reactor Bldg to Torus vacuum breaker isolation valve fails open (Tech Spec) 3 See Scenario Guide C URO Service Water pump trip / manual start of the standby pump CRS 2013 NRC Scenario # 1 - T -111 Low Level, Rev 1

Event Malfunction Event Event No. No. Type* Descri~tion 4 See Scenario Guide I PRO Drywell pressum instrument fails upscale without the expected half TS CRS scram (Tech Spec) I insert half scram lAW GP-25 5 See Scenario Guide R URO Condensate pump trip with recirc run back failure I power reduction CRS 6 See Scenario Guide M ALL Loss of #2 auxiliary bus 1 loss of condensate & feedwater 1 reactor coolant leak inside the drywell 7 See Scenario Guide C URO RPS failure requires ARI to scram the reactor CRS 8 See Scenario Guide C PRO HPCI controller fails in automatic CRS 9 See Scenario Guide C ALL HPCI turbine trip, requiring an emergency blowdown to restore level with low-pressure ECCS 10 See Scenario Guide I PRO Containment spray logic failure hampers effort to spray the CRS containment, requiring crew to use alternate RHR loop

  • (N)ormal, (R)eactlvity, (I)nstrument, (C)omponent, (M)aJor, (TS) Tech Spec 2013 NRC Scenario #1 T-lil Low Level, Rev I

Scenario Outline ES-D-l Simulation Facility Peach Bottom Scenario No. ~#2::::.-_ _ Op Test No. 2013 NRC Examiners Operators _ _ _ _ _ _ _ _ _ CRS (SRO)

URO (ATC)

PRO (BOP)

Scenario The scenario begins with the reactor at 100% power. After taking the shift, the crew will perform the Summary Master Trip Solenoid Valve Routine Test.

Next, a turbine stop valve will fail closed, requiring the crew to execute OT-102 "Reactor High Pressure",

which will require reducing reactor power to less than or equal to 95% in accordance with GP-5 "Power Operations" .

Next, a failure in the controller for the 'A' Recirc M-G set will cause the Recirc pump speed to oscillate. The crew should recognize the changes in core and jet pump flows and "lock up" the 'A' Recirc pump. The crew should verify compliance with Technical Specifications for recire loop flow differentials.

Next, a spurious HPCI initiation will occur due to a logic system failure. The crew should enter OT-104 "Positive Reactivity Insertion" and shutdown HPCI. This event will cause a steam leak from the HPCI system piping in the HPCI pump room, requiring the crew to enter and execute T-1 03 "Secondary Containment Control". Initial attempts to isolate HPCI using the Isolation Pushbutton results in "split" indication for the MO-2-23-15 and -16 steam supply valves. Operator should attempt manual closure of these valves. All attempts to isolate HPCI will be unsuccessful due to logic system and control switch failures. The leak will gradually worsen, requiring a reactor scram and entry into T-101 "RPV Control",

While performing scram actions, the PRO should recognize the generator lockout failure following the main turbine trip and manually open the generator output breakers and exciter field breaker. The URO should respond to the 'C' reactor feedpump discharge bypass valve failure by batch feeding through the 'c' reactor feedpump discharge valve. When depressurization using Bypass Valves is performed, Bypass valves will initially function normally but then fail closed, requiring operator to complete depressurization using SRVs.

Conditions will continue to deteriorate in the Reactor Building due to the HPCI steam leak. When the second Reactor Building area (Torus Room) exceeds its T-103 Action Level, the crew should perform a T-112 "Emergency Blowdown". The scenario will end when the RPV is depressurized and RPV level is being maintained with Condensate .

. Initial IC-119, 100% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Event No. No. Type* Description 1 See Scenario Guide N PRO Perform the master trip solenoid valve routine test CRS 2 See Scenario Guide R URO Turbine stop valve fails closed I power reduction CRS 3 See Scenario Guide C URO 'A' Recirc pump speed oscillations (Tech Spec) I Lock up the 'A' TS CRS Recirc pump 4 See Scenario Guide C PRO Inadvertent HPCI initiation I shutdown HPCI (Tech Spec)

TS CRS 2013 NRC Scenario #2 - T-103 HPCI Steam Leak, Rev I

Event Malfunction Event Event No. No. Type* Description 5 See Scenario Guide M ALL HPCI steam leak into secondary containment I

6 See Scenario Guide C PRO HPCI Isolation System pushbutton and control switch failure I CRS 7 See Scenario Guide I PRO Generator lockout fails to occur following main turbine trip CRS 8 See Scenario Guide C URO 'C' reactor feedpump discharge bypass valve fails to open, CRS complicating post-scram and post-blowdown reactor level control 9 See Scenario Guide C URO Bypass Valves fail closed, depressurize using SRVs CRS 10 See Scenario Guide ALL Emergency blowdown due to exceeding Reactor Building temperature limits in more than one area

  • (N)ormal, (R)eactlvlty, (I)nstrument, (C)omponent, (M)aJor. (TS) Tech Spec 2013 NRC Scenario #2 - T-103 BPCI Steam Leak, Rev I

Scenario Outline ES-D-l Simulation Facility Peach Bottom Scenario No. Op Test No. 2013 NRC Examiners Operators _ _ _ _ _ _ _ _ _ CRS (SRO)

URO (ATC)

PRO (BOP)

Scenario The scenario begins with the reactor at 100% power. After taking the shift the crew is required to Summary swap operating TBCCW pumps for inspection of a noisy bearing on the 'A' TBCCW pump.

Next, an individual control rod drive scram accumulator will experience low pressure and alarm in the main control room. The crew will initiate corrective action but the accumulator pressure will remain low requiring the crew to declare the control rod slow or inoperable per Technical Specifications.

Shortly after this, the E-4 diesel generator will inadvertently start, requiring the crew to shutdown the E-4 diesel generator and apply Technical SpeCifications for an inoperable diesel generator.

The crew should then recognize and respond to lowering main condenser vacuum caused by a failure of the in service steam jet air ejector steam supply valve. The crew must enter OT-106 "Condenser low Vacuum" and reduce reactor power in accordance with GP-9-2 "Fast Power Reduction".

Following the power reduction, a turbine lube oil malfunction will result in a high bearing temperature and vibration condition for the main turbine, requiring the crew to scram the reactor and trip the main turbine. A CRD hydraulic malfunction will result in a low-power ATWS, requiring the crew to execute T-101 "RPV Control" and T-117 "level/Power Contro!." In addition, the scram discharge volume (SDV) will fail to completely isolate, requiring the crew to manually isolate the SDV.

When SBlC is initiated the SBlC pump will trip, requiring thE~ URO to place the alternate SBlC pump in service. The second SBlC pump will trip shortly after being placed in service. A failure of the only available EHC pump will cause the turbine bypass valves to close, requiring the crew to utilize HPCI and/or SRVs for reactor pressure control. The crew should perform T-220 "Driving Control Rods During Failure to Scram" to insert control rods. The crew will need to adjust control rod drive water pressure in order to successfully insert the control rods. The scenario may be terminated when the crew has control of RPV power and level using T-240 "Termination and Prevention of Injection into the RPV" and the crew is inserting control rods.

Initial IC-120, 100% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Event No. No. Type* Description 1 See Scenario Guide N PRO Swap operating TBCCW Pumps CRS 2 See Scenario Guide TS CRS i Individual control rod drive scram accumulator low pressure (Tech Spec) 3 See Scenario Guide I PRO E4 diesel generator spurious start / diesel generator shutdown TS CRS (Tech Spec) 4 See Scenario Guide C PRO Failure of Steam Jet Air Ejector steam supply valve I re-open by placing additional valve air supply in service 2013 NRC Scenario #3 - T-117 Hydraulic ATWS, Rev I

Event Malfunction Event Event No. No. Type* Description 5 See Scenario Guide R URO Fast reactor power reduction (wI recirc)

CRS 6 See Scenario Guide C URO Main turbine high temperature and vibration I reactor scram CRS 7 See Scenario Guide M ALL ATWS (hydraulic) / turbine bypass valves fail closed 8 See Scenario Guide C URO Standby liquid control (SBLC) pump trips I start second SBLC pump CRS /second pump trips 9 See Scenario Guide C PRO Two in-series scram discharge volume (SDV) vent valves fail to CRS automatically isolate 10 See Scenario Guide C URO Low CRD drive water pressure I adjust to drive control rods

  • (N)ormal, (R)eactlvlty, (I)nstrument, (C)omponent, (M)aJof, (TS) Tech Spec 2013 NRC Scenario #3 - T-117 Hydraulic A TWS, Rev I
Retrieved from "https://kanterella.com/w/index.php?title=ML13098A252&oldid=2520782"