ML13154A297
| ML13154A297 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 02/25/2013 |
| From: | Exelon Generation Co |
| To: | Operations Branch I |
| Jackson D | |
| Shared Package | |
| ML12328A009 | List: |
| References | |
| ES-401, TAC U01867 | |
| Download: ML13154A297 (26) | |
Text
ES-401 BWR Examination Outline FORM ES-401 M1 Facility Name: Peach Bottom Tier 1.
Emergency &
Group 1
K 1
3 K
2 3
K 3
4 Date of Exam: 03/25/2013 RO KIA Category Points K K K A A A A G 4 5 6 1 2 3 4
- 3 4 3 Total 20 A2 3
SRO-*
Abnormal 2 1 1 1 N/A 1 1 N/A 2 7 1 2 3 Plant Evolutions 4 5 4 5 5 27 4 6 10 3 2 2 2 2 2 3 3 2 3 2 26 3 2 5 2.
Plant 2 1 1 1 1 1 1 1 1 12 1 1 1 3 Systems Tier Totals 4
- 3. Generic Knowledge and Abilities Note: 1.
Categories 3 3 3 1
3 3
- 2 2
3 2
3 4 4
3 3 38 10 1
2 5
2 2
Ensure thai at least two topics from every applicable KIA category are sampled within each tier of the RO lli 3 8 7
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 (IR) 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. Enler the group and tier Iota Is 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 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. Umit 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 BWR Examination Outline Form ES-401 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO)
K K K A E/APE # I Name 1 Safety Function KIA Topic(s) 2 3 1 295001 Partial or Complete Loss of Forced pumpfiows: Not-BWR-1&2 3.0 Core Flow Circulation' 1 & 4 Partial or Complete Loss of AC 16 3.5 Partial or Total Loss of DC Pwr 16 and operate components, Including local controls 4.4 0
Main Turbine Generator Trip I 3 controls 2.7 4
0 pressure regulating system 3
0 16 Control Room Abandonment 17 0
8 Partial or Total Loss of CCW I 8 2
9 Partial or Total Loss of Inst. Air 1 8 air system pressure 1 Loss of Shutdown Cooli post-accident instrumentation 0
Refueling Acc! 8 n 3.3 3
0 High Drywell Pressure I 5 4.1 0
High Reactor Pressure I 3 pressure regulating system 3.8 2
0 pool level 3.5 6
0 0
High Drywell Temperature 15 Mark-I&II 3.7 Low Suppression Pool Wlr Lvl!5 01 temperature 3.9 0
1 Reactor Low Water Levell 2 4.0 4
SCRAM Condition Present and 0
Hot shutdown boron weIght: Plant*Specific 3.2 4
EOP mitIgatIon strategies 3.7 0 2.9 Plant Fire On Site I 8 2
Generator Voltage and Electric Grid 0 .3
/6 2 Totals: 3 3 4 3 20 ES-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 Q# E/APE # / Name / Safety Function KIA Topic(s) IR #
1 2 3 1 0
62 295002 Loss of Main Condenser Vac / 3 IRPs 3.4 1 3
c, 295007 High Reactor Pressure I 3 I 0 295008 High Reactor Water Level 12 , 0 c
63 295009 Low Reactor Water Levell 2 f;;:, Reactor water cleanup blowdown rate 2,9 1 Ability to diagnose and recognize trends in an accurate and 64 ;?95010 High Drywell Pressure / 5 04; ' timely manner utilizing the appropriate control room reference 4,2 1 material,
~95011 High Containment Temp / 5 0
~95012 High Drywell Temperature / 5 0
~95013 High Suppression Pool Temp, 15 0
" 1>~11f 295014 Inadvertent Reactivity Addition /1 0 m
Ability to verify system alarm setpOints and operate controls 65 295015 Incomplete SCRAM 11 4,2 1 identified in the alarm response manual.
,:c~
295017 High Off-site Release Rate 19 0 0 ~, ' i"""
59 1295020 Inadvertent ConI, Isolation I 5 & 7 m loss of normal heat Sink 3,7 1 1 1,,<[/
0 ",
61 1295022 Loss of CRD Pumps I 1 , Reactor SCRAM 3,7 1 1
- ""'1:
1295029 High Suppression Pool Wtr Lvii 5 ," 0
"'7'"
ff:?;i eM?':
[~;~
1295032 High Secondary Containment Area 0 60 PCISINSSSS 3,6 1 ITemperature 15 4 295033 High Secondary Containment Area 0
Radiation Levels I 9 H~"~mve~~ 0 tainment High 0
Differential Pressure I 5 295036 Secondary Containment High 0
SumplArea Water Levell 5 r--
500000 High CTMT Hydrogen Cone. I 5 KIA Category Totals: 1 1 1 1 :{
ES-401, Page 18 of 33 2 Group Point Total:
f11
ES-401 4 Form ES-401-1 ES-401 BWR 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 9 206000 HPCI Turbine shaft sealing: BWR-2, 3, 4 2.8 1 2
207000 Isolation (Emergency) 0 Condenser 0
10 209001 LPCS System venting 2.5 1 5
209002 HPCS 0 0
4 211000 SLC SBLC pumps 2.9 1 1
1,22 212000 RPS 0 M. Nuclear boiler instrumentation, Knowledge of local auxlhary operator tasks during an emergency and the resultant 3.7; 2
2 $ operational effects 3.8 16,2 2150031RM 0 0 .' Reactor power indication response to rod position 3.7; 2
2 4 changes; Up scale or down scale trips 3.7 215004 Source Range Monitor 0
0 7 215005 APRM / LPRM Rod withdrawal blocks 3.7 1 1
- C'~
0 1 " Suppression pool water supply; Condensate storage tank 3.5; 12,2 217000 RCIC level 2
3 1 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
0 0 Loss of any number of main steam flow inputs; All 3.3;
~0,2 259002 Reactor Water Level Control 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
0 3.1 1 15 262001 AC Electrical Distribution Exceeding voltage limitations 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 charging/discharging rate 2.5 1 1
0 01. Automatic starting of compressor and emergency 3; 3.9 2 18,2 264000 EDGs generator; Knowledge of system purpose and/or function 1
"flf.
0 01,,+" 3.1; 19,2 300000 Instrument Air " Main Steam Isolation Valve air: Pressure gauges 2 5 1 2.6 0
6 400000 Component Cooling Water Loads cooled by CCWS 2.9 1 1 ;,;
0 KIA Category Totals: 3 2 2 2 2 2 3 ...
3' 2 3 2 Group Point Total: 26 ES-40 1, Page 19 of 33
ES-401-1 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO)
- K K K K K K A A A A <:,
Q# System # I Name 1 2 3 4 5 6 1 3 4 ',:(;" KIA Topic(s) IR #
2
- 0 28 201001 CRD Hydraulic Backup SCRAM valve solenoids 3,5 1 3
201002 RMCS 0 201003 Control Rod and Drive Mechanism 0 201004 RSCS 0 201005 RCIS 0 201006 RWM ,
0 202001 Recirculation 0 0
30 202002 Recirculation Flow Control Minimum and maximum pump speed setpoints 2,9 1 7
204000 RWCU 0 214000 RPIS 0 215001 Traversing In-core Probe 0 0
38 215002 RBM RPS BWR-3, 4, 5 3,0 1 1
216000 Nuclear Boiler Inst 0 219000 RHR/LPCI Torus/Pool Cooling Mode 0 37 223001 Primary CTMT and Aux,
- 04. ""''Y'UP"U' w""uu, '"'"'"""" 'UPLU""U",", ,"U,"
actions that require Immediate operation of system 1 4,6 49 226001 RHR/LPCI CTMT Spray Mode 0 0
36 230000 RHR/LPCI: Torus/Pool Spray Mode Indicating lights and alarms 3,6 1 9
233000 Fuel Pool Cooling/Cleanup 0 0
31 234000 Fuel Handling Equipment Water as a shield against radiation 2,9 1 3
239001 Main and Reheat Steam 0
~39003 MSIV Leakage Control 0 241000 ReactorfTurbine Pressure Regulator 0 0
27 1245000 Main Turbine Gen, / Aux, 8
Reactorlturbine pressure control system. Plant-Specific 3A 1 256000 Reactor Condensate 0 0 Pump trip 3,7 1 34 259001 Reactor Feedwater 1
0 29 268000 Radwaste Drain sumps 2,7 1 4
0 33 271000 Offgas System flow 3,1 1 8
272000 Radiation Monitoring 0 286000 Fire Protection 0 0
35 288000 Plant Ventilation Isolation/initiation signals 3,8 1 1
290001 Secondary CTMT 0 290003 Control Room HVAC 0 2 - Main steam system 1 32 290002 Reactor Vessel Internals 0 I,; 29 0
KIA Category Totals: 1 1 1 1 1 2 1 1 1 1 1 Group Point Total: 12 ES-40 1, Page 20 of 33
ES-401 2 Form ES-401 ~1 BWR Examination Outline Form ES-401-1 KKK A IR #
2 3 1 Partial or Complete Loss of Forced 3.2 Flow Circulation 11 & 4 Partial or Complete Loss of AC 16 0 Partial or Total Loss of DC Pwr 16 0 Main Turbine Generator Trip 13 0 0
16 Control Room Abandonment I 7 0 8 Partial or Total Loss of CCW I 8 ificalions for a system. 4.7 t reference to procedures those 9 Partial or Total Loss of Ins!. Air I 8 ediate operation of system 4.4 Is.
Loss of Shutdown Cooling I 4 4.1 Refueling Acc I 8 3,7 10 verify Ihat the alarms are consistent with the High Drywell Pressure I 5 42
'i o
001 High Water Temp, o
29~i027 t-lighl C(mt,linrnerltTemperalure I 5 o High Drywell Temperature 15 o Low Suppression Pool Wtr LviI 5 39 o
o o
o 000 0 ES-40 1, Page 17 of 33
ES-401 3 Form ES-401-1 BWR Examination Outline Emergency and Abnormal Plant Evolutions Tier 1/Group 2 (SRO)
KIA Topic(s)
Loss of Main Condenser Vac /3 High Reactor Pressure / 3 High Reactor Water Levell 2 Low Reactor Water Levell 2 o High Drywell Pressure I 5 o 1 High Containment Temp / 5 o 2 High Drywell Temperature / 5 3.9 3 High Suppression Pool Temp. 15 o 4lnadvertentReacti~rityAddition/1 I Knowledge of RO tasks penonned outsIde the malO control 5 Incomplete SCRAM { 1 room dunng an emergency and the resultant operatio Inadvertent Cont Isolation I 5 & 7 o Loss of CRD Pumps I 1 o High Suppression F'ooll Wtr LvII 5 Ability to interpret and execute procedure steps Category Totals:
ES-401, Page 18 of 33
ES-401 4 Form ES-401-1 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2fGroup 1 (SRO)
- f Name KIA Topic(s) IR #
RHRfLPCI: Injection o I 111205Q()0 Shutdown Cooling Mode o HPCI o Isolation (Emergency) o o
HPCS o 3.4 system alarm setpomts and operate controls 4,0 4,3 5004 Source Range Monitor PCISfNuclear Steam Supply SRVs 259002 Reactor Water Level Control 261000 SGTS AC Electrical Distribution UPS (ACIDC)
DC Electrical Distribution 3.2 o
o o
o 5
ES-401. Page 19 of 33
~rS%--re--m#-I-N-am-e-----------'~~~~~~~~~~~~~~A1~T~~~~~--~--~~-A-T-O-PiC-(S-)--------~-IR~-#~1 201001 CRD Hydraulic
- o r-r--------------+~~~ *~+-~I~+------------------+~--~I 201002 RMCS :i~q 0 r-r_--------------------+-+-+_+_+_~
201003 Control Rod and Drive Mechanism o
~r-------------------------~
201004 RSCS o r-r_--------------------+_
~1~~ 0
--r~O-1-0-06-R-W-M-----------------+-+-4~--t__I_ *. ~----------------------------+---r--O~I t202001 Recirculation IT ". 0 F02002 Recirculation Flow Control I I *. I:::; o
~~_W_I:_U ______________ ~~~~~~~~0_.:.*+-+-~~~*~:+-*________________________~--+--:~I
~15001 Traversing In-core Probe o
~15002 RBM o
~16000 Nuclear Boiler Inst. It: o
--j~~9OcmFR~H~RUt~jP~CII:1T~oruru;SI~Po.oo;ICCo.oo~li~ng~--r-t-II-r-t-1--r-1:-1~1r-t--i H Primary CTMT and Aux.
',;i,'
i.f Ii,.:!
° o
Y !!;; I,;g 1-__!f-22--60-0'I-R-H-R-'-LP-C-I:-C-T-M-T-S-pr-a-M-o-d-e---+-I_ill 230000 RHRILPCI: ToruslPool Spray Mode JJI~! I k. o U~Fuel Pool Cooling/Cleanup ~ cn:".,..,-t-+::;""I-----------------+--+-o--..,
192 I~~~~~~ Fuel Handling Equipment i Fuel orientation 3.7 1 f----!lr~--390--01-M-a-in-a-nd--R-eh-e-a-tS-t-ea-m--------!f--r-4-~-+~-=:r:J:-T__~I~~~------------------------------r---r--oO~1 239003 MSIV Leakage Control --Lli- .' I;*;
24' ReactorlTurbine Pressure Regulator o F~Y
~45000 Main Turbine Gen. / Aux. o
~ n~
r __ ~59001
_56_000_R_e_a_c_lo_r_c_on_d_e_n_sa_te
_ Reactor F e _e _ _d
_ _w f__+-~-+_4--I ~~~ I_~.~>.::,,~..--------- -----------f----r--O~
+__ a t e r ! ' ' ' . ..*. __ 0
~68000 Radwaste 0 1--1-----------------------+-+-+- '" i (to interpret reference matenals. such as graphs. curves. 42 93 ~71000 Offgas iF I ,etc. .
---f------------------------_+_+~-- Ie
~72000 Radiation Monitoring t!::"~
---r_--------------------~-+-+_+_+_+_+_ii-i'-1-1~r--------------------------i---r--~
o
~~~ 0
---f------------------------r_+_+-+-~~
~~Planl Ventilation . J Low reactor water levet Plant-Specific 3.6 I~ Secondary CTMT o
-f------------------------r_+_
~90003 Control Room HVAC o
-r_--------------------+_+_
~90002 Reactor Vessel Internals o
.i;::i~~~================~=*==091 KJA Category Totals: o 0 o~!! Group Point Total 3 ES-401, Page 20 of 33
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401*3 IKnIowledcle of refueling administrative requirements.
to explain and apply system limits and precautions. 4.0 coordinate personnel activities oulside the control room 3.4 to interpret and execute procedure steps 4,6 75 Knowledge of facility requirements for controlling vital/controlled access, 95 Knowledge of tha process for making design or operating changes to the facility. 3.2 100 2.2. ge of the process for controlling equipment configuration or status 4.3 i ity to interpret control room indications to verify the status and operation of a system, understand how operator actions and directives affect plant and system conditions 4.2 1 Knowledge of the process for managing maintenance activities during shutdown 2.2. 18 operations, such as risk assessments, work prioritization, etc 2.6 1 2.2.
2,2, 96 2.3. 04 Knowledge of radiation exposure timits under normal or emergency conditions Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, 98 2.3. 15 portable survey instruments, personnel monitoring equipment, etc, 3,1 1 Knowledge of radiation or contamination hazards that may arise dUring normal, abnormal, 70 2.3. 14 1 71 3.4 1 Ability to recognize abnonnal indications for system operating parameters that are entry 97 level conditions for emergency and abnormal operating procedures, 72 Knowledge of the specific bases for EOPs IKnlowledlJB of the bases for pnoritizing safety functions during abnormal/emergency ES-401, Page 26 of 33
ES-401 Record of Rejected KJAs Form ES-4014 Tier / Randomly Reason for Rejection Group Selected KIA RO 2/1 212000 PBAPS does not have RPS PAM instrumentation.
Q#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 Q#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-1021 Q#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-1021 Q#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 Q#92 2.4.4 (Replaced with 234000 K5.05)
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Peach Bottom Date of Examination: 03/25/2013 Examination Level: RO r2l 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)
G2.1.4 (3.3) - Recognize and Report License Medical Conduct of Operations N.R Condition Challenge (NEW)
G2.2.41 (3.5) -Isolating the 38 RBCCW Heat Exchanger Equipment Control N,R Due to a Leak (NEW)
D,S,P G2.3.11(3.8) - Perform PRO Duties For A Liquid Radiation Control (2011 Radwaste Discharge (PLOR 258C)
NRC)
Emergency Plan N/A Not Required NOTE: All items (5 total) are required forSROs. RO applicants require only 4 items unless they are retaking only the administrative~s, when 5 are required.
" Type Codes & Criteria: (C)onIRJI room, (S)imulator, or Class(R)oom (D)iredfrom bank (=s. 3 for ROs; =s. 4 for SROs & RO retakes)
(N)ewar (M)odified from bank (~ 1)
{P)reWJus 2 exams (=s. 1; randomly selected)
ES 301, Page 22 of 27
ES*301 Administrative Topics Outline Form ES-301*1 Facility: Peach Bottgm Date of Examination: 03125/2013 Examination level: RO 0 SRO ~ 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 Product 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 bank ~ 3 for ROs; ~ 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (.:::. 1)
(P)revious 2 exams (:5, 1; randomly selected)
ES 301, Page 22 of 27
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 03/25/2013 Exam Level: RO ~ SRO-I 0 SRO-U 0 Operating Test Number: 2013 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System 1 JPM Title Type Code* Safety
. Function
- a. 233000 A2.02 (3.1/3.3) -- Fuel Pool Cooling and Cleanup 1 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 I Raise HPCI A, 0, 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-083C)
- d. 209001 A4.04 (2.9/2.9) - Core Spray System I 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 0, EN,L, S 5 Perform a Group 1 PCIS Isolation Reset (GP-8A) (PLOR-024C)
- f. 262001 A4.04 (3.9/3.7) -- AC Distribution 1 Excite the Main Generator D,S 6 WLOR-031C)
- g. 212000 A4.14 (3.8/3.8) - Reactor Protection System 1 Reset a Full 0, EN, L, S 7 Scram (PLOR-004C)
- h. 400000 A4.01 (3.1/3.0) - Component Cooling Water 1Verify Isolation Of A,D,S 8 Drywell Chilled Water And RBCCW (Alternate Path - RBCCW Is Supplying 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 (3.9/3.8) -- Reactor Core Isolation Cooling I Defeat RCIC D,E, R 2 Interlocks lAW T-251-3 (PLOR157P)
- j. 218000 K4.04 (3.5/3.6) -- Bypass of SV-9130A and B lAW T-331-3 (NEW) N,E,R 3
- k. 286000 A4.06 (3.4/3.4) - Fire Protection System I Diesel Driven Fire A,D 8 Pump Manual Start (Alternate Path - Battery Status Lights Not Lit (PLOR 327PA)
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; atl-5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO I SRO-II SRO-U (A)ltemate 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/;:::1/;:::1 (EN)gineered safety feature - I - 1 :2: 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 1 ~ 2 (randomly selected)
(R)CA ;:::1/;:::1 1 ;::: 1 (S)imulator ES-301, Page 23 of 27
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 03/25/2013 Exam Level: RO D SRO-I 181 SRO-U D Operating Test Number: 2013 NRC Control Room SystemsO (B for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System 1 JPM Title Type Code* Safety Function
- a. 233000 A2.02 (~.1J3.3) - Fuel Pool Cooling and Cleanup 1 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 / Open Main Steam D, L,S 3 Isolation Valves After a Group-1 Isolation (PLOR-OB3C)
- 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 (3.6/3.5) - Primary Containment Isolation System / D, EN,L,S 5 Perform a Group 1 PCIS Isolation Reset (GP-BA) (PLOR-024C) f.
- g. 212000 A4.14 (3.B/3.B) - Reactor Protection System I Reset a Full D, EN, L, S 7 Scram {pLOR-004C) h'. 400000 A4.01 (3.1/3.0) - Component Cooling Water I Verify Isolation Of A,D,S 8 Drywell Chilled Water And RBCCW (Alternate Path - RBCCW Is Supplying 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 (3.9/3.8) - Reactor Core Isolation Cooling / Defeat RCIC D,E, R 2 Interlocks IAWT-251-3 (PLOR157P)
- j. 21BOOO 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 I Diesel Driven Fire A,D 8 Pump Manual Start (Alternate Path - Battery Status Lights Not Lit (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-II SRO-U (A) Item ate path 4-6 I 4-6 I 2*3 (C)ontrol room (D)irect from bank ~g/~8/~4 (E)mergency or abnormal in-plant ~1/~1/~1 (EN)gineered safety feature - I - 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 I S. 2 (randomly selecled)
(R)CA ~1 f:::1/~1 (S)imulator ES-301 Page 23 of 27 f
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 0312512013 Exam Level: RO 0 SRO-I 0 SRO-U t8J Operating Test Number: 2013 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System I JPM Title I T"",,,,
1 Code" Safety Function
- 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 I Perform Pump Capacity A,D,EN,S 4 Test For 1ST (Alternate Path - Min Flow Valve Fails To Open) (PLOR 335CA) e.
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 Supplying 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 (3.9/3.8) -- Reactor Core Isolation Cooling / Defeat RCIC D,E, R 2
, Interlocks IAWT-251-3 (PLOR157P)
- j. 218000 K4.04 (3.5/3.6) - Bypass of SV-9130A and B lAW T-331-3 (NEW) N,E,R 3 I
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 5.9/,;:8l~4 (E)mergency or abnormal in-plant ?1/~1/~1 (EN)gfneered 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 5. 3 I ~ 2 (randomly selected)
(R)CA ~1/~1/~1 (S)imulator ES-301, Page 23 of 27
Scenario Outline I':S-D-1 Simulation Facility Peach Bottom Scenario No. #1 OpTest No. 2013 NRC Examiners Operators _ _ _ _ _ _ _ _ CRS (SRO)
_ _ _ _ _ _ _ _ _ _ _ URO (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-2502A 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 "Installation of Tripsllsolations to Satisfy Tech SpecfTRM Requirements" .
Next the 'A' Condensate pump will trip without the expected Recire 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 drywall 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 RCle (the only remaining high-rressure 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 I~~~~
No. ____~~ No. ____~__~~~I Type* __ -+____________________~D~e~s~c~r~ip~t~io~n~_________________~_~__
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 # I - T-ill Low Level, Rev O.doc
Event Malfunction Event Event No. No. Type* Description 4 See Scenario Guide I PRO Drywell pressure 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 runback failure I power reduction CRS 6 See Scenario Guide M ALL Loss of #2 auxiliary bus 1 loss of condensate & feedwater I reactor coolant leak inside the drywell 7 See Scenario Guide C PRO HPCI controller fails in automatic CRS 8 See Scenario Guide C URO RPS failure requires ARI to scram the reactor 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)eacttvlty, (I)nstrument, (C)omponent. (M)aJor, (TS) Tech Spec 2013 NRC Scenario #1 - T-111 Low Level, Rev O.doc
Scenario Outline ES-f)..1 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 Recire 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", 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, 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 'N TS CRS Recirc pump 4 See Scenario Guide C PRO Inadvertent HPCI initiation I shutdown HPCI (Tech Spec)
TS CRS
~ Scenario Guide M All HPCI steam leak into secondary containment 2013 NRC Scenario #2 - T-I 03 HPC I Steam Leak, Rev O,doc
Event Malfunction Event Event No. No. Type* Description 6 See Scenario Guide I PRO Generator lockout fails to occur following main turbine trip CRS 7 See Scenario Guide C URO 'C' reactor feed pump discharge bypass valve fails to open, CRS complicating post-scram and post-blowdown reactor level control 8 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 HPCI Steam Leak, Rev O.doc
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 Control." 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 Individual control rod drive scram accumulator low pressure (Tech Spec) 3 See Scenario Guide I PRO E4 diesel generator spurious start I 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 -I 17 Hydraulic ATWS, Rev O.doc
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 Isecond 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)aJor, (TS) Tech Spec 2013 NRC Scenario #3 - T-117 Hydraulic ATWS, Rev O.doc
Scenario Outline ES-D-)
Simulation Facility Peach Bottom Scenario No. #4 Op Test No. 2013 NRC Examiners Operators _ _ _ _ _ _ _ _ CRS (SRO)
_ _ _ _ _ _ _ _ URO (ATC)
PRO (BOP)
Scenario The scenario begins with the reactor at approximately 6% power during a reactor startup.
Summary Following shift turnover, the crew is directed to secure drywell purge in preparation for inerting the drywell. Once drywell purge is secured, the 'B' drywell chiller will trip. The crew should place a standby drywell chiller in service in accordance with the system operating procedure. Next, a blown fuse will cause an ARI power supply failure, requiring the crew to initiate repairs and evaluate ARI RPT operability per Tech Specs.
Following the ARI failure, the crew should continue with the reactor startup by pulling control rods in accordance with the approved startup sequence. During this evolution a control rod will drift out, requiring the crew to execute ON-121 "Drifting Control Rod" and declare the affected control rod inoperable in accordance with Tech Specs. After the Tech Spec determination is made, while still executing ON-121. a second control rod will drift in, requiring the crew to perform an immediate reactor scram and enter T-100 "Scram". A subsequent trip of the 'c' reactor feed pump will complicate RPV level control post-scram.
While T-100 actions are in progress, a leak will develop in the torus, requiring the crew to enter T-103 "Secondary Containment Control" and T-102 "Primary Containment Control". When torus level reaches 12.5 feet, the crew will be directed to enter T-101 "RPVControl" and perform a depressurization.
A failure of the turbine bypass jack will require the crew to use alternate methods to depressurize the reactor in accordance with T-101 "RPV Control". Torus level will continue to lower to the point where the crew will be required to perform T-112 "Emergency Blowdown". The scenario may be terminated when the RPV is depressurized and HPSW is injecting into the torus.
Inilial IC-121, 6% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Event No. No. Type* Description 1 See Scenario Guide N PRO Secure drywell purge CRS 2 See Scenario Guide C PRO Drywell chiller trip I place standby chiller in service CRS 3 See Scenario Guide TS CRS ARt power supply failure (Tech Spec) 4 See Scenario Guide R URO Power ascension with control rods CRS
'5 See Scenario Guide C URO Drifting control rod (Tech Spec)
TS PRO CRS 2013 NRC Scenario #4 - T-102 Torus Leak, Rev O.doc
Event Malfunction Event Event No. No. Type* Description See Scenario Guide nd 6 C ALL 2 Drifting control rod, Manual Scram, T -100
.7 See Scenario Guide I URO 'C' reactor feed pump trip CRS 8 See Scenario Guide M ALL Torus leak into secondary containment I emergency blowdown 9 See Scenario Guide C PRO Turbine bypass jack fails, preventing rapid depressurization to the CRS main condenser
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)aJor, (TS) Tech Spec 2013 NRC Scenario #4 - T -102 Torus Leak, Rev O.doc