ML101230159
| ML101230159 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/03/2010 |
| From: | NRC/RGN-II |
| To: | Tennessee Valley Authority |
| References | |
| 50-327/10-301, 50-328/10-301 | |
| Download: ML101230159 (32) | |
Text
ES-401 1 Rev. 9 PWR Examination Outline Form ES-401-2 Facility: Sequoyah Date of Exam: 2010 RO KIA Category Points SRO-Only Points I
Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4
- Total
- 1. 1 3 3 3 3 3 3 18 3 3 6 Emergency &
2 1 2 1 2 2 1 9 2 2 4 Abnormal Plant N/A N/A Evolutions Tier Totals 4 5 4 5 5 4 27 5 5 10 1 2 3 2 3 2 2 3 3 3 3 2 28 3 2 5 2.
2 1 1 0 1 1 1 1 1 1 1 1 10 2 1 3 Plant Systems Tier Totals 3 4 2 4 3 3 4 4 4 4 3 38 5 3 8
- 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 3 2 2 2 1 2
- 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 ofthe 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 ES-401, Attachment 2, 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 KIAs 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) KIAs in Tiers 1 and 2 shall be selected from Section 2 ofthe KIA Catalog, butthe topics must be relevant to the applicable evolution or system.
- 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; iffuel 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 ofthe KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to KIAs that are linked to 10 CFR 55.43.
ES-401 Form ES-401-2 PWR Examination Outline 1~1 Form ~~
Emerqencv and Abnormal Plant Evolutions - Tier 1/Group 1((RO 1 SHe)
'-.../
E/APE # 1 Name 1 Safety Function K K K A A G KIA Topic(s) IR #
1 2 3 1 2 007 [;ALIO 3.7 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery 1 1 R
000008 Pressurizer Vapor Space Accident 1 3 000009 Small Break LOCA 1 3 R Or&( e:~I,(J 4-, ;L 000011 Large Break LOCA 13 f\ Oll~K;;{,06L (;lAc;)
000015/17 RCP Malfunctions 14 R 01 SAl< /. () I 4,4 000022 Loss of Rx Coolant Makeup 1 2 R OJ."J. AA /,07 ;;Z .g 000025 Loss of RHR System 14 R Dc;( 5 AI< 3, {);;L 33 000026 Loss of Component Cooling Water 1 8 000027 Pressurizer Pressure Control R OOl? AA/.()3 3,c" System Malfunction 1 3 000029 ATWS 1 1 R 0;)1 '{;;<;< ,() " ;2,1 000038 Steam Gen. Tube Rupture 1 3 R 038 AC(, I£:' 3,g 000040 (BW/E05; CE/E05; W/E12)
R A I< /.0 ~ .0<
Steam Line Rupture - Excessive Heat Transfer 1 4 4,fo 000054 (CE/E06) Loss of Main Feedwater 14 R D64AG'J.,J,'6J 000055 Station Blackout 1 6 --
000056 Loss of Off-site Power 1 6 R 05b A/<3,O( 3,S" 000057 Loss of Vital AC Inst. Bus 16 f< 057 AA;;Z';£:' 3, ()
000058 Loss of DC Power 1 6 --
000062 Loss of Nuclear Svc Water 1 4 ~ ob2 AGe< 14, '1 13,g 000065 Loss of Instrument Air 1 8 I< 0(05"" A 1<3,04- 3.0 W/E04 LOCA Outside Containment 1 3 R W£(?J1- £A;<. :2 13.b W/E11 Loss of Emergency Coolant Recirc. 1 4 BW/E04; W/E05 Inadequate Heat A <../ , '3/1 Transfer - Loss of Secondary Heat Sink 14 000077 Generator Voltage and Electric 0 iG ;(,~*+4 4,;1 Grid Disturbances 1 6 R
'/1 '"' T.
"j !J Group Point Total: 18/6
ES-401, REV 9 R0 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO 007EA1.10 Reactor Trip - Stabilization - Recovery
/1 3.7 3.7 0 0 o O~O o0 S/G pressure 009EK1.01 Small Break LOCA / 3 4.2 4.7 ~ 0 0 0 0
~OOOO o0
.-=----::::--=_-.---:c-:-------c--:----:---c------:--c:- ----..- - -..
Natural circulation and cooling, including reflux boiling 011EK2.02 Large Break LOCA / 3 2.6 2.7 0 0 0 0 0 00000 0 Pumps 015AK1.01 RCP Malfunctions / 4 4.4 4.6 ~
~ 000000000 Natural circulation in a nuclear reactor power plant 022AA1.07 Loss of Rx Coolant Makeup / 2 2.8 2.7 00 Excess letdown containment isolation valve switches and indicators 025AK3.02 Loss of RHR System / 4 3.3 3.7 O~ 0000000 Isolation of RHR low-pressure piping prior to pressure increase above specified level 027AA1.03 Pressurizer Pressure Control System 3.6 3.5 Pressure control when on a steam bubble Malfunction / 3 029EK2.06 ATWS / 1 2.9 3.1 ~ DO Breakers, relays, and disconnects.
=--------_.
038EA2.06 Steam Gen. Tube Rupture / 3 3.8 4.4 0 0 0DOD 0 0 ~
~OOO0 0 0 Shutdown margins and required boron concentrations 040AK1.04 Steam Line Rupture - Excessive Heat 3.2 3.6 ~ 0DODD 0 0 0 0DOD0 0 Nil ductility temperature Transfer / 4 054AG2.1.31 Loss of Main Feedwater / 4 4.6 4.3 0 0 0 0 0 00000 ~ Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup.
Page 1 of 2 6/25/2009 12:27 PM
ES-401, REV 9 R0 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:
OS6AK3.01 Loss of Off-site Power / 6 3.S 3.9 0 ~ 0 Order and time to initiation of power for the load sequencer OS7AA2.16 Loss of Vital AC Inst. Bus / 6 3 3.1 0 0 0 0 0 0 0 ~ 0 0 0 Normal and abnormal PZR level for various modes of plant operation 062AG2.4.9 Loss of Nuclear Svc Water / 4 3.8 4.2 0 0 0 0 ~ Knowledge of low power / shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.
06SAK3.04 Loss of Instrument Air / 8 3 3.2 0 0000 Cross-over to backup air supplies 077AG2.2.44 Generator Voltage and Electric Grid Disturbances / 6 4.2 4.4 o OOO~
ooo~ 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 WE04EA2.2 LOCA Outside Containment / 3 3.6 4.2 0 0 0 0 0 ~
~ Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.
WEOSEK2.2 Inadequate Heat Transfer - Loss of 3.9 4.2 0 ~ 0 0 0 0 0 0 0 Facility's heat removal systems, including primary Secondary Heat Sink / 4 coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.
Page 2 of 2 6/2S/2009 12:27 PM
ES-401 3 Form ES-401-2 r;S-401 PWR Examination Outline !~ Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 RO I ~
E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #
1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 R O;L3 AK;;( ,O;;L R,b 000032 Loss of Source Range NI / 7 P. 0"3;2 AK;;(/Ol 0{,1 000033 Loss of Intermediate Range NI /7 ~ 0'33' AA;2" J2 ;;,S 000036 (BW/A08) Fuel Handling Accident /8 It; 03& f/(~ ~,4-, 3) 3"g 000037 Steam Generator Tube Leak / 3 R 031 It A/, 04- 5,t, 000051 Loss of Condenser Vacuum /4 000059 Accidental Liquid RadWaste ReI. / 9 000060 Accidental Gaseous Radwaste ReI. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. /8 {( 068 A K3, 1r1- 4, I 000069 (W/E14) Loss of CTMT Integrity /5 000074 (W/E06&E07) Inad. Core Cooling /4 000076 High Reactor Coolant Activity / 9 W/E01 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation /9 BW/A01 Plant Runback /1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin /4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 R WE (!) 3 GA :2,1 , 3.4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 R WelD GAf3 3,4-BW/E13&E14 EOP Rules and Enclosures CE/A 11; W/E08 RCS Overcooling - PTS / 4 1< (fiE 08 t;J!.I, J 3. 5 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery f".
D. int Totals: I
ES-401, REV 9 {<.O T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO 02SAK2.02 Pressurizer Level Malfunction / 2 2.6 2.7 D ~
~O D DO DO Sensors and detectors 032AK2.01 Loss of Source Range NI /7 2.7 3.1 0D ~ DOD DO DD Power supplies, including proper switch positions 033AA2.12 Loss of Intermediate Range NI / 7 2.5 3.1 0 0 0 0 0 0000 ~ 0 o allowable channel disagreement Maximum allc)walble 036AG2.4.35 Fuel Handling Accident / S 3.S 4.0 DOD 0 ~
~ Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects 037AA1.04 Steam Generator Tube Leak / 3 3.6 3.9 DO O~OO D~DD 0 D Condensate air ejector exhaust radiation monitor and failure indicator 06SAK3.12 Control Room Evac. / S 4.1 4.1 4.5 4.5 0D ~ 0 0 0 0DDD 0 0 Required sequence of actions for emergency evacuation of control room WE03EA2.1 LOCA Cool down - Depress. / 4 3.4 4.2 0DD 0 DOD ~O Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
WEOSEK1.1 RCS Overcooling - PTS / 4 3.5 3.S ~
~ 0 0 o DDD DOD Components, capacity, and function of emergency systems.
WE1 OEA 1.3 WE10EA1.3 Natural Circ. With Seam Void/ 4 3.4 3.4 3.7 0DD 0 0DD~DD 0 ~ 0 Desired operating results during abnormal and emergency situations.
Page 1 of 1 6/25/2009 12:27 PM
ES-401 5QN r~ 2010 - 120 EXaM 4 Form ES-401-2 ES-401 PWR Examination oUi~,
oUi~. F Form ES-401-2 cC' '" ..,
Plant S stems - Tier 2/Group 1 RO .sRG)
-SRG)
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 f)() 3 j('t;, Of
'I 3/3 003 Reactor Coolant Pump R 004 Chemical and Volume R~ &04 A 3, oS' 3*1 Control 004 A4;13 3,3 005 Residual Heat Removal IRK ooS )<5,oS)<5'05 c:(,7 006 Emergency Emerqency Core Cooling Coolinq f{ 00 b AI, OOb AI,I1-11-' 3,(;,
007 Pressurizer Relief/Quench 13,/)
Tank R
008 Component Cooling Coolina Water R f{ (Jog A /, '1-/
IvogA/,Of-/ ~
fit?l ,03 f);< 3.,-
'3./
3'0 010 Pressurizer Pressure Control K Off) I< 4-, O~. 3,0 012 Reactor Protection R R f) 1;( Ii I, 01 ) 1<::<,01 1~.1:s 013 Engineered Safety Features
~p R 0/3 ~~, '
3,("
0:(.7 Actuation () 1.3 Xi ,01 0/3 01
- 022 Containment Cooling Coolina !Z 0;((;( I< / ,01 3,S 3,S 025 Ice Condenser 1\ 0;(5' 1<1-, 021 !j 026 Containment Spray IZ O;:zro K3,O~ 4-.;(.
039 Main and Reheat Steam R 031O?1 G, Q ,J.
G,Q'}*7 7 44-144 059 Main Feedwater ~ tJ'S'1' D'S'1' )(3,0:( :3,("
061 Auxiliary/Emergency
!\ oCt! I j,;z,o j,;Zo if" Feedwater 062 AC Electrical Distribution K Ob(;l J< 4, /)::L RS 063 DC Electrical Distribution R 003 AJ ,01 ~.s
- ZS 064 Emergency Emerqency Diesel Generator K K 10&4- A4
- IJ- Kb,t)S ~:~:~
073 Process Radiation Monitoring R ~73 fl flf.o~
- f. >;L 3,7 076 Service Water R 01& A3,0;t A3.0;t 13,7 078 Instrument Air A .'D1g K;;{, )2 D1g K;;{,D2 3,3 103 Containment IA fZ 103 /l;;(,
/l;;(1 of A3,ol 3'!ij 3,.
r-. n ..
Totals:
mrmr;:
l 3 2 3 l;;t ,~ :3 k3j :) 13
,3 2 Group Point Total: 28/5
ES-401, REV 9 1<-0 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:
RO SRO 003KS.01 Reactor Coolant Pump 3.3 3.9 0 0 ~ 0 0 0 0 0
~ooooo The relationship between the RCPS flow rate and the nuclear reactor core operating parameters (quadrant power tilt, imbalance, DNB rate, local power density, difference in loop T-hot pressure) 004A3.0S Chemical and Volume Control 3.9 3.9 0 0 0 0 0 0000 ~ 0 0
~oo pressure and temperature 004A4.13 Chemical and Volume Control 3.3 2.9 ooooo~
00 oo~ VCT level control and pressure control OOSKS OS Residual Heat Removal 2.7 3.1 0 0 0 ~ 000 00 Plant response during "solid plant": pressure change due to the relative incompressibility of water 006A1.14 Emergency Cooling 3.6 3.9 0 ~oo o0 Reactor vessel level 007K1.03 Pressurizer Relief/Quench Tank 3.0 3.2 ~ 0 0 0 0 0 0 0 008A1.04 Component Cooling Water 3.1 3.2 0 0 ~ 0 Surge tank level 008A2.03 Component Water 3.0 3.0 3.2 0 0 0OO~OOO 0 ~ 0 0 0 High/low CCW temperature 010K4.02 Pressurizer Pressure Control 3.0 3.4 00 ~o 00000 00 00 Prevention of uncovering PZR heaters 012A1.01 Reactor Protection 2.9 2.9 3.4 3.4 0DO 0 0OO~O 0 ~ 0 0 Trip setpoint adjustment 012K2.01 Reactor Protection 3.3 3.7 ~ DODD000 DO 00 RPS channels, components and interconnections Page 1 of 3 6/2S/2009 12:27 PM
ES-401, REV 9 RO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO 013K2.01 Engineered Safety Features Actuation 3.6 3.8 ~ 0 0000 0000 ESFAS/safeguards equipment control
~~----.~~~--- .....---------------
013K6.01 Engineered Safety Features Actuation 2.7 3.1 0 0 oo~ooo ~ 0 0 0 Sensors and detectors
--~---.-
022K1.01 Containment Cooling 3.5 3.5 3.7 3.7 ~ 0 0 0 0
~OOOO 0 0 0 DO SWS/cooling system 025K4.02 Ice Condenser 2.8 3.0 0 0 DO ~ 0 0 0 DOD System control 026K3.02 Containment Spray 4.2 4.2 4.3 4.3 0OO~0 ~ 0 0 0 0 0 DODD Recirculation spray system 039G2.1.7 Main and Reheat Steam 4.4 4.7 nn 0 ~ Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation.
059K3.02 Main Feedwater 3.6 3.7 OO~OOOOOOOO AFW system 061G2.1.20 061 G2.1.20 Auxiliary/Emergency Feedwater 4.6 4.6 0 0 0 0 0 0 0 0 0 0 ~
OOOOOOOOOO~ Ability to execute procedure steps.
062K4.02 AC Electrical Distribution 2.5 2.7 0 0 0 ~ 0 0 0 0 OOO~OOOO Circuit breaker automatic trips
~
.--~~--------------~.::------:-------- .. ------~~----
063A2.01 DC Electrical Distribution 2.5 2.5 3.2 3.2 0 0 0 0 0 0 0 ~ 0 0 0 OOOOOOO~OOO Grounds
~--.- .. ~----~~----------------------------=---:----------------
064A4.12 Emergency Diesel Generator 2.7 2.7 2.6 2.6 0 0 0 0 0 0 ODD ~ 0 OOOOOOOOO~O Synchroscope Page 2 of 3 6/25/2009 12:27 PM
ES-401, REV 9 Ro T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO 064K6.08 Emergency Diesel Generator 3.2 3.3 0D 0D 0D 0D 0D ~ 0D 0D 0D 0D 0D Fuel oil storage tanks 073A4.02 Process Radiation Monitoring 3.7 3.7 3.7 3.7 DDD DD~ D ooo~o Radiation monitoring system control panel 076A3.02 Service Water 3.7 3.7 0D 0D 0D 0D o~o D~DD Emergency heat loads 078K2.02 Instrument Air 3.3 3.3 3.5 3.5 0D~DDDDDDDDD
~ 0 0 0 0 0 000 Emergency air compressor
-.~-----".-"----------------
103A2.04 Containment 3.5 3.6 ~ Containment evacuation (including recognition of the alarm) 103A3.01 Containment 3.9 3.9 4.2 4.2 0DDDDDDD 0 0 0 0 0 0 ~o Containment isolation Page 3 of 3 6/25/2009 12:27 PM
ES-401 5 Form ES-401-2 ES-401 PWR Examination ou~~ F Plant S stems - Tier 2/Group 2 RO SRe-)
System # 1 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 4 001 Control Rod Drive R 001 /<;<,0S; 6,1 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication R f) It A4. 0/ "33 015 Nuclear Instrumentation !<, DISA 3, os- ;;;, to 016 Non-nuclear Instrumentation R O/b /<5, 'I ;;:),7 017 In-core Temperature Monitor II? DF7 A~,OI 3, I 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge t< ODl1 At, D;;L 34-033 Spent Fuel Pool Cooling R 0'33 G ;; ,4-,;( I 14,0 034 Fuel Handling Equipment R )31- t: t:"oOL Ic;(,b 035 Steam Generator K 035 1<4.01 3/~
041 Steam DumpITurbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate IZ O!S(;' 1<1,0], ;(,10 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection KIA Category Point Totals:
IITL0lIili I I I I I Group Point Total: 10/3
ES-401, REV 9 R0 T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO 001 K2.05 Control Rod Drive 3.1 3.5 0 ~ 0 0 0 0 MIG sets 014A4.01 Rod Position Indication 3.3 3.1 0 0 0 0 0 0 0 0 0 ~ 0 Rod selection control 015A3.05 Nuclear Instrumentation 2.6 2.7 0 0 0 0 0 0 O~ Recognition of audio output expected for a given plant condition 016K5.01 Non-nuclear Instrumentation 2.7 2.8 0 ~ DO 0 0 Separation of control and protection circuits 017A2.01 In-core Temperature Monitor 3.1 3.5 o 0 0 0oooo~ooo 0 0 0 ~ 0 0 0
= - = - - - - - - - - - - - - - - - - - - -----,-----:--
Thermocouple open and short circuits 029A1.02 Containment Purge 3.4 3.4 0 ~ DOD 000 Radiation levels
::::c-----:c---:-::-------:c------~-:_::____:___:=_=_==_=____==_==__:=___=-_==__==--____:_:_-__:____:_______;_:_:___--___:-______:=_:___---:-:-----
033G2.4.21 Spent Fuel Pool Cooling 4.0 4.6 0 0 0 0 0 ~
~ Knowledge of the parameters and logic used to assess the status of safety functions 034K6.02 Fuel Handling Equipment 2.6 3.3 n n ooo~ooooD ~ 0 0 0 0 Radiation monitoring systems 035K4.01 Steam Generator 3.6 3.8 0 ~ 0 SIG level control 056K1.03 Condensate 2.6 2.6 ~ 000000 000000 o0 MFW Page 1 of 1 6/25/2009 12:27 PM
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 II ..
Facilit /" ~(RO [};ar~) n~+~ _" c.
Date of Exam: F~h c<o/D
'OlD Category KIA # Topic RO SRO-Only IR # IR #
2.1. 11 !Jse pp/tl././c:t Use /t/././c:t Cz"AcpvAcA tpU0!J .1c:A -h e vaet (?lie va.,tt(?Jfe 31 3/1 ~IA iV/A- 1A-2.1.3;;2 £X/;/,;?//A. :; a?p~sdC;;M!l~,".
£Xp!t2//./iJ a?p~c;;qC!lP/'A P:L P'L '5 3,6 1 1.
O~ r1,.:<~
{;, 4j)~/'tii
/¥j ft7 ;iJdJtlft/~~
dJtlft/~~ IZ UJ 0 .'i:r Conduct of Operations 2.1.4D 2.1.
f?r, v .
J f J' ?<, ~
\
2.1. I 2.1.
Subtotal 3 2.2.0 Proce'sr;.fvJ Prot: es:;.fvJ YP;!tiu Yr\;:dUf'!P'jt'lP] {"Ch:Vi'O', *Io h:vi 'e' ( *Ill jJ foe 'r prOC'r 3J) IJ 2.2.40 P-pp4; /.fch P-ppi.tJ uv
.. rr;:;.fCs:'
rp,fCs:' 3,4-
- 2. 2.2. 1 Equipment Control 2.2.
2.2. I 2.2.
Subtotal ~.
- (,. . .
2.3. t; ~ ya~ a;;Iv~v" vadi4. a;;Iv~v\. yy0;V!Jvt'lf'L,,;Vt(;{ ;;;'U((JfJ.ofA.I~:1.
yyV1;VI.{A'/:rL.*;vt(;{ ;;;'t4)lEJ, I.§ C)/1 2.3.7 tt* 1'4 ~j tCY"1'4 ~I K! "'JP f/Jf rrv, In~./~ /A.hM) rrv; 'tJ f1.f'ft./~ /.lthM) Mil M",F S,S 3.
Radiation 2.3. 12.
2.3.12 (:'W{71A/~~"1"{
(:y, tlvJ/1~;J iCk?
jeW ,v'l '1/,{ CvJ/?v':J If iC~ 'II. 'fl. flAi'J JI£I-,,/:11]
Jl£l-u/2lt". I r IICH IIcH 3*Z Control 2.3. r 2.3.
2.3.
Subtotal 1.3 1.3. :.
2.4.;<g Pr
°ri7(J/clv~
JeRel Vt/!.<J f~,)
f",) rv,Jop rVcP}/J,;2,C
);/J,.1*C A JJf,Ar.rr;.,.&'}.t..
..f AAr.rr;J(,&'jt. 3,2
>12
- 4. 2.4.4b V8~'v~
V8'Vil v(.
., tUJfI/;'
/7 ()t-)/i ML,l
.,/j rXriRW>/?r:f "'j rXriR!4>!?r:f "'/ If (p../.f /t?r.f C1'/;;:f'r C1'/;;1 r 4,2 Emergency 2.4.
Procedures / "- Pian 2.4.
~
2.4. .11 2.4. I/JA ,1/
'V Subtotal ::l
- 2 . ,J)4.
,J~
Tier 3 Point Total If) If) . 10 7
ES-401, REV 9 gD T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO G2.1.19 Conduct of operations 3.9 3.8 ooooo~ DDDDD~ Ability to use plant computer to evaluate system or component status .
-----------------~-__: ____=c___c=-_==_:=__==__=::_==_==__==:_==_-__:_:__:_::_.
G2.1.32 Conduct of operations 3.8 4.0 0 0O~ ~ Ability to explain and apply all system limits and precautions. G2.1.40 Conduct of operations 2.8 3.9 o DDDDDDD~ OOOOOOO~ Knowledge of refueling administrative requirements G2.2.40 Equipment Control 3.4 4.7 0 0 0 0 0 0 0 0 0 0 ~ Ability to apply technical specifications for a system. G2.2.6 Equipment Control 3.0 3.6 0 0 0 0 0 ~ Knowledge of the process for making changes to procedures G2.3.12 Radiation Control 3.2 3.2 3.7 3.7 0DDDDDDDDD~ 0 0 0 0 0 0 0 0 ~ Knowledge of radiological safety principles pertaining to licensed operator duties G2.3.5 Radiation Control 2.9 2.9 oooooooo~
--------------:-:-=------.~-------------------
0 0 0 0 0 0 ~ Ability to use radiation monitoring systems G2.3.7 Radiation Control 3.5 3.6 0DOD 0 0 0 0DDDD~ 0 0 0 ~ Ability to comply with radiation work permit requirements during normal or abnormal conditions
---~~--:--:~.-=~=-~-==-:=-~-=~=-~-==:-=-=--=-=-~-=-----------
G2.4.28 Emergency Procedures/Plans 3.2 4.1 0 0ooo~ 0 0 ~ Knowledge of procedures relating to emergency response to sabotage.
-------~--
G2.4.46 Emergency Procedures/Plans 4.2 4.2 DO DOD O~ to verify that the alarms are consistent with the plant conditions. Page 1 of 1 6/25/2009 12:27 PM
ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline ~ Form ES-401-2 EmerQencv and Abnormal Plant Evolutions - Tier 1/Group 1 SRil E/APE # 1 Name 1 Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery 1 1 p O()7 13A ;{, 4.~ 000008 Pressurizer Vapor Space Accident 1 3 000009 Small Break LOCA 1 3 S 001 G~. 4.;;;, 143 000011 Large Break LOCA 13 1$ 011 GA;),01 13*4 000015/17 RCP Malfunctions 1 4 000022 Loss of Rx Coolant Makeup 1 2 000025 Loss of RHR System 14 000026 Loss of Component Cooling Water 18 ;5 l3.b 000027 Pressurizer Pressure Control System Malfunction 1 3 000029 A TWS 1 1 000038 Steam Gen. Tube Rupture 1 3 000040 (BW/E05; CE/E05; W/E12) Steam Line Rupture - Excessive Heat Transfer 14 000054 (CE/E06) Loss of Main Feedwater 1 4 000055 Station Blackout 16 000056 Loss of Off-site Power 1 6 S OSC? AG.;;2, I. I~ 3,g 000057 Loss of Vital AC Inst. Bus 1 6 000058 Loss of DC Power 1 6 000062 Loss of Nuclear Svc Water 1 4 000065 Loss of Instrument Air 18 W/E04 LOCA Outside Containment 1 3 W/E11 Loss of Emergency Coolant Recirc. 1 4 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink 14 000077 Generator Voltage and Electric Grid Disturbances 1 6 1)17 A fl. , 4.4-KIA Category Totals: .3 13 Group Point Total: 18/6
ES-401, REV 9 SRO T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 007EA2.04 Reactor Trip - Stabilization - Recovery 11 4.6 4.4 DO o 0 ~ If reactor should have tripped but has not done so, manually trip the reactor and carry out actions in ATWS EOP 009EG2.4.20 Small Break LOCA 1 3 3.8 4.3 00 DOD OOOO~ DDDD~ Knowledge of operational implications of EOP warnings, cautions and notes. 011EA2.07 Large Break LOCA / 3 3.2 3.4 ~ That equipment necessary for functioning of critical pump water seals is operable 026AA2.02 Loss of Component Cooling Water 18 2.9 3.6 0 0 0 0 0 0 0 ~ 0 0 0 The cause of possible CCW loss 056AG2.1.19 Loss of Off-site Power 1 6 3.9 3.8 OO~ DD~ Ability to use plant computer to evaluate system or component status. 077AG2.2.44 Generator Voltage and Electric Grid 4.2 4.4 0 0 0 0 0 0 0 0 ~ Ability to interpret control room indications to verify the Disturbances 1 6 status and operation of a system, and understand how operator actions and directives affect plant and system conditions Page 1 of 1 6/25/2009 12:27 PM
ES-401 3 ES-401 PWR Examination Outline /~ Form E'" fI)" ,., Emerqency and Abnormal Plant Evolutions - Tier 1/Group 2 fRO"/ SRO E/APE # / Name / Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 t DD :; A~;2, +. 3S 11. 0 000024 Emergency Boration / 1 5 of21 AAo? 0 I 4.1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI /7 ~ 033 AA,;:?, 0'3> 3,/ 000036 (BW/A08) Fuel Handling Accident /8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum /4 000059 Accidental Liquid RadWaste ReI. / 9 000060 Accidental Gaseous Radwaste ReI. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E 14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 I's f}/1 EG r:2;), 44~ 41-000076 Hiqh Reactor Coolant Activity / 9 W/E01 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation /9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip /4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin /4 BW/E08; W/E03 LOCA Cooldown - Depress. /4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery fA r'. "'.' T. t:I:E r-. D~;n+ Total: 9/
ES-401, REV 9 SRO T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 005AG2.4.35 Inoperable/Stuck Control Rod 1 1 3.8 4.0 0 0 000000 0 0 0 0 0 0 D~ Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects 024AA2.01 Emergency Boration 1 1 3.8 4.1 DOD 0 0 0 0 0 ~ Whether boron flow and/or MOVs are malfunctioning from plant conditions 033AA2.03 Loss of Intermediate Range NI 17 2.8 3.1 0 0 0 0 ~ 0 Indication of blown fuse 074EG2.2.44 Inad. Core Cooling 1 4 4.2 4.4 0 D~ ~ 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 Page 1 of 1 6/25/2009 12:27 PM
ES*401 4 ES-401 PWR Examination Outlin~~ Form ES-401-2 Plant S stems - Tier 2/Group 1 fl SRO 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 4 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal Is I:> G; 2.~V 1,7 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection ;;1 DJ;2 6ii.;<, 4. // 14,2 013 Engineered Safety Features Actuation 022 Containment Cooling 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam lS 03J A;(,OiiL ~,7 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator I¢ 064 A ~*Ib 3,1
~ ,
073 Process Radiation Monitoring 073 !1 g,2 076 Service Water 078 Instrument Air 103 Containment KIA Category Point Totals: :3 :L Group Point Total: I 28/5
ES-401, REV 9 SRO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 005G2.2.40 Residual Heat Removal 3.4 4.7 0 0 0 0DDDDD~ 0 0 0 0 ~ Ability to apply technical specifications for a system.
- : c - - - -__ ----------------------~~--~~--.-.----------
012G2.4.11 Reactor Protection 4.0 4.2 0 0 0 0 0 0 0 0 0 ~
~ Knowledge of abnormal condition procedures.
039A2.02 Main and Reheat Steam 2.4 2.7 0 0 0 0 0 Decrease in turbine load as it relates to steam escaping from relief valves 064A2.16 Emergency Diesel Generator 3.3 3.7 0DDD~DDD 0 0 ~ 0 0 0 Loss of offsite power during full-load testing of ED/G
----------~-~-------~--~-------
073A2.02 Process Radiation Monitoring 2.7 2.7 3.2 3.2 0 0 0 0 00 00 0 0 ~ 0 0 0
~DDD Detector failure Page 1 of 1 6/25/2009 12:27 PM
ES-401 5 Form ES-401-2 Cv-'"tu PWR Examination OU~I~ F Plant S stems - Tier 2/Group 2 SRO System # 1 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 4 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 1$$ - L?;2~11 I 40 029 Containment Purge 033 Spent Fuel Pool Coolinq Cooling 034 Fuel Handling Equipment 035 Steam Generator 041 Steam DumpITurbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal ,5 GQI,~ OS)' GQd,'JO 3,2 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water IrS' . 07t; Ad, D/ 3.';;{ 3.)" 079 Station Air 086 Fire Protection I Lt' UI fA . n~ ,int
.. Totals:
~ tals: l.:l II Group p Point Total: 10/3
ES-401, REV 9 SRO T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 028A2.03 Hydrogen Recombiner and Purge 3.4 4.0 000 DD~DDD o~ooo The hydrogen air concentration in excess of limit flame Control propagation or detonation with resulting equipment damage in containment
~~--=~~-:--~-:-:---=-~----:~~~~~~~~:---=-==--==--.=--==--==-==--=--==-==-=--"----
055G2.1.20 Condenser Air Removal 4.6 4.6 D 0 D 0 D 0 D 0 D 0 D 0 ~
~ Ability to execute procedure steps.
075A2.01 Circulating 3.0 3.2 D 0 D0 D0 D0 D0 D 0 ~ Loss of intake structure Page 1 of 1 6/25/2009 12:27 PM
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 II Facility: Facili~: SON SQ,,! (Sf2-D {'"raM)
£Kt{lY') Date of Exam: Feb ;(olD
;(01 I Category KIA # Topic RO SRO-Only SRO-Onlv IR # IR #
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liYlO ./J~ [ III [;5 . j-J:.rs, SW i1-c~--; \ 4,0 v I Conduct 2.1. of Operations 2.1. 2.1. 2.1. Subtotal I I*** '. ....*... ~.. 2.2. 102 ~ved ItVvic..c!.. IjiJrTUdt.,/j 5t1 ,rvedltVv1(..e. JrTU. d /.,,, v-<Lddl 4-,/ 2.2.11 v {/ Pl PIV~.l pl 1'"i'!t7/V'Lt:'-i-'7f';
,,"f'!t7A/' Cl"?f'; Y;qy~ /V49/~J .P. P. "'" "" *Ire
'ee 6,%
- 2. 2.2.
Equipment Control 2.2. 2.2. 2.2. Subtotal ........ . ..:;;L.
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- 3. 2.3.
Radiation Control 2.3. 2.3. 2.3. Subtotal I I','
.i. '.;.; . .1 2.4. b t7t/}7V. p-ttJ\"h In" r;;?}7V,rHtJ\*% /",L10/-!
,§ ji pt?f '}'Yl.
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0r:' rlYtiI#f!J/t 4,7 4. Emergency 2.4. 1;2 2.4. 0w 0-v/ Y(b.1 Y (b.IJI}y'(jil/ f u/4.1 ;"11 Ip .,1
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*,-3 Procedures I Plan 2.4.
2.4. ,II
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'V Subtotal "f{'~r\.** C. foil;;foIJ;; ";;;L I~
... . / '.......*...
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' 7 Tier 3 Point Total **** I.' 10 I ***... 7
ES-401, REV 9 SRO T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO G2.1.26 Conduct of operations 3.4 3.6 0 0 0 0 0 0 ~ Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). G2.1.29 Conduct of operations 4.1 4.1 4.0 4.0 0 0 0 0 00 00 0 0 0ooo~ 0 0 ~ Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. G2.2.12 Equipment Control 3.7 4.1 0 0 0 0 0 0 0 0 ~ Knowledge of surveillance procedures. G2.2.17 Equipment Control 2.6 3.8 0 0 ~ Knowledge of the process for managing maintenance activities during power operations.
--~----------------------------------------------------------------
G2.3.5 Radiation Control 2.9 2.9 ~ Ability to use radiation monitoring systems G2.4.12 Emergency Procedures/Plans 4.0 4.3 0 0 0 0 0 0 0 0 0 0 ~ Knowledge of general operating crew responsibilities during emergency operations. G2.4.6 Emergency Procedures/Plans 3.7 4.7 ~
~ Knowledge symptom based EOP mitigation strategies.
Page 1 of 1 6/25/2009 12:27 PM
ES*301 Administrative Topics Outline Form ES-301-1 Facility: Sequoyah 1 & 2 Date of Examination: 2/16/2010 Examination Level: RO D SRO X Operating Test Number: 2010301 Administrative Topic Type Describe activity to be performed (see Note) Code* N,R 2.1.25 Ability to interpret reference materials, such as Conduct of Operations graphs, curves, tables, etc. (3.9/4.2) Boric Acid storage tank level calculation (JPM #2.1.a SRO 09-AP) D,R 2.1.7 Ability to evaluate plant performance and make Conduct of Operations operational judgments based on operating characteristics, reactor behavior, and instrument interpretation (4.4/4.7) Determine Mn. Turb. Startup and Loading Times (JPM2.1.b) D,R 2.2.12 Knowledge of surveillance procedures Equipment Control Review a Surveillance for approval (3.7/4.1) (JPM410) D,R,P 2.3.6 Ability to approve release permits Radiation Control Approval of a Waste Gas Decay tank Release (2.0/3.8) (JPM A-3) D,R 2.4.38 Ability to take actions called for in the facility Emergency emergency plan, including supporting or acting as Procedures/Plan emergency coordinator if required. (2.4/4.4) Classify Rep: LOCA with Fuel Fail Loss of Cntmt (JPM 019 AP2 ) NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (:::; 3 for ROs; :::; 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (~ 1) (P)revious 2 exams (:::; 1; randomly selected)
SRO 2.1.a This JPM has the candidate if the "C" boric acid storage tank, which is in standby is operable such that it can be aligned to Unit 1 to replace the "A" BAST which is being removed from service. The candidate will determine that "C" BAST is not operable and makes recommendation to make tank operable before allowing "A" to be removed from service. This is a New JPM that can be performed in the simulator or in the classroom. SRO 2.1.b This JPM has the candidate determine the Main Turbine Startup and Loading times for Unit 1 turbine from cold shutdown conditions during a Unit startup. This is a Bank JPM that can be performed in the simulator or classroom. SRO 2.2 This JPM has the candidate review a surveillance test for determining whether or not the adjust RCP seal injection supply controlled leakage. This is a Bank JPM that can be performed on the simulator or in the classroom. SRO 2.3 This JPM has the candidate review a radioactive gas decay tank release to determine if release permit is accurate and can take place as written. The candidate will determine who needs to approve the permit and actions needed due to having a radiation monitor out of service. This is a Bank JPM that can be performed in the simulator or in the classroom. This JPM is a Repeat from the 2009 NRC exam and was randomly selected from a group of Radiation Control JPMs. SRO 2.4 This JPM has the candidate determine the correct Emergency Classification based on the data provided and make the initial contacts. This is a time critical, Bank JPM that can be performed in the simulator or in the classroom.
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Sequoyah 1 & 2 Date of Examination: 2/16/2010 Examination Level: RO X SRO D Operating Test Number: 2010301 Administrative Topic Type Describe activity to be performed (see Note) Code* N, R 2.1.25 Ability to interpret reference materials, such as Conduct of Operations graphs, curves, tables, etc. (3.9/4.2) Boric Acid storage tank level calculation (JPM 2.1.a RO) D,R 2.1.26 Knowledge of industrial safety procedures (such Conduct of Operations as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen. (3.4/3.6) Containment Formaldehyde Stay Time Calculation (JPM 123) D,S 2.2.12 Knowledge of surveillance procedures (3.7/4.1) Equipment Control Perform Reactor Coolant System water inventory (JPM 43-1) Radiation Control N,R 2.4.13 Knowledge of crew roles and responsibilities Emergency during EOP usage. (4.0/4.6) Procedures/Plan Calculating maximum reactor vessel vent time Per EA-0-7 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (:::; 3 for ROs; :::; 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (2: 1) (P)revious 2 exams (:::; 1; randomly selected)
RO 2.1.a This JPM that has the candidate determine if "C" Boric Acid storage tank can be aligned to Unit 1 so that "A" Boric Acid storage tank can be removed from service. This is a New JPM can be performed in the simulator or classroom. RO 2.1.b This JPM has the candidate determine the stay time in containment based on the formaldehyde concentration and determine the respiratory protection requirements. This is a Bank JPM that can be performed in the simulator or classroom. RO 2.2 This JPM has the candidate determine the Reactor Coolant System Water inventory (leak rate). This is a Bank JPM that should be performed in the simulator (using plant computer). RO 2.4 This JPM has the candidate determine the allowable Reactor Vessel head venting time to prevent CNMT Hydrogen from exceeding 3% per EA-O-7. This is a New JPM that can be performed in the classroom.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: _ _Sequoyah Nuclear Station Date of Examination: _02/16/2010-Exam Level: RO X SRo-ID SRO-U 0 Operating Test No.: 2010301 Control Room Systemsfg/ Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / JPM Title Type Code* Safety Function
- a. 005 Inoperable/Stuck Control Rod (AA2.03 3.5/4.4) ~ D, S, A, L 1 001 AP-1 Emergency 80ration (Stuck Rods)
- b. 011 Large 8reak LOCA (EA 1.11 4.2/4.2) D, S, A, EN 2 013AP1 Transfer to Hot Leg Recirc.
- c. 008 Pressurizer (PZR) Vapor Space Accident (Relief Valve M,S,A 3 Stuck Open) (AA2.19 3.4/3.6) 070 AP Respond to Failed Open PZR Spray valve
- d. 005 Residual Heat Removal System (A4.01 3.6/3.4) D,S,L 4P 152 Swap RHR pumps (A train to 8 train) with level in the PZR.
- e. WE05 Loss of Secondary Heat Sink (EA2.1 3.4/4.4) N, S 4S 034-1 Establish MFW per EA-2-2
- f. 103 Containment System (A1.01 3.7/4.1) D, S 5 065 Re-establishment of CNMT pressure
- g. 064 Emergency Diesel Generators (A4.06 3.9/3.9) D,S,A 6 077-1AP Perform DG load test on 18-8 DG (with high crankcase pressure)
- h. 015 Nuclear Instrumentation System (A2.02 3.1/3.5) M,S, 7 021 AP Respond to a failure of N-41 In-Plant Systems@
Systems(gl (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i. 078 Instrument Air System D,E 8 096 Respond to Loss of Control Air System
- j. 005 Residual Heat Removal System D,E,R 4P 044 Venting A-A RHR pump due to cavitation
- k. 062 AC Electrical Distribution System D,E,A 6 091AP Transfer Controls to Aux Mode per AOP-C.04, Att 3 All RO and SRO-I control room (and in-plant) systems must be different and seNe different safety functions; all 5 SRO-U systems must seNe different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes I Criteria for RO I SRO-II SRO-U
(A)lternate path 4-6/4-6 / 2-3 (C)ontrol room (D)irect from bank s9/s8/s4 (E)mergency or abnormal in-plant ;::1/;::1/;::1 (EN)gineered safety feature - / - / ;::1 (control room system) (L)ow-Power / Shutdown ;::1/;::1/;::1 (N)ew or (M)odified from bank including 1(A) ;::2/;::2/;::1 (P)revious 2 exams s 3 / s 3 / s 2 (randomly selected) (R)CA ;::1/;::1/;::1 (S)imulator JPM A. Candidate has to determine that emergency boration is required due to 2 stuck rods following a Rx trip, with the normal emergency boration valve failing to open, requiring the candidate to align alternate boration through the charging pump suction. This is a Bank, Low Power, Alternate path, JPM. JPM B. Candidate is directed to make alignment change for Transfer to Hot Leg Recirculation per ES-1.4. following a trip due to large break LOCA. The Hot leg recirculation valve will fail to open requiring RHR to be aligned to Cold Leg i1\iection with High head pumps aligned to Hot Legs. This is a Bank, Low Power, Alternate path, JPM. JPM C. Candidate will be required to respond to failed PZR pressure instrument which causes the PZR spray valves to o~n with one valve sticking open. With a PZR spray valve failing to close, candidate will follow actions of AOP-1.4 bnd trip the RX and trip at least two RCPs to stop mitigate the depressurization to prevent SI actuation. This is a-Modified Bank, Alternate Path JPM. Original JPM JPM D. Plant is in Mode 4, and Candidate is directed to transfer RHR pumps from B train to A train. This is a Bank, Low Power JPM. JPM E. Candidate is directed to establish a Secondary Heat Sink using Main Feed Water System following a Rx Trip. MFW will be required due to a failure of all AFW pumps. This is a New JPM. JPM F. Candidate is directed to vent excess pressure from CNMT and then establish Ventilation alignment to re-establish CNMT vacuum atmosphere. This is a Bank JPM. JPM G. Candidate is to perform a quick start ofEDG A-A and load the EDG. While the candidate is loading the EDG, a high crankcase condition will develop requiring a manual emergency trip of the EDG. This is a Bank, Alternate Path JPM. JPM H. Candidate will respond to failed Nuclear Instrument (N41) High. Control Rods will be stepping in at maximum rate, Candidate will take Rod bank selector switch to Manual and proceed to remove failed channel from service. This is a Modified Bank, JPM. Original JPM (021) had N-41 failing low from -45%, this JPM has N-41 failing High, requiring immediate manual action to stop control rod movement prior to removing channel from service. JPM 1. Candidate is to restore Control Air pressure following a Loss of Off-site power. Candidate will restore Control and Service Air compressors using EA r. This will require a Local Manual start. This a Bank JPM. JPM J. Candidate is directed to vent the lA-A RHR pump due to pump cavitating during mid-loop operation. This venting is done locally and is required to be performed to return the RHR pump to service. This is a Bank JPM and is performed in the RCA. JPM K. Candidate is directed to perform checklist 3 of AOP-C.04 Shutdown from Auxiliary Control Room, following an event which requires Control Room Abandonment. This a Bank, Alternate Path JPM.
ES*301 Control Room/ln*Plant Systems Outline Form ES*301*2 Facility: Seauovah Nuclear Plant Date of Examination: _02/16/2010 - Exam Level: RO D SRO-I X SRO-U D Operating Test No.: 2010301 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / JPM Title Type Code* Safety Function
- a. 005 Inoperable/Stuck Control Rod (AA2.03 3.5/4.4) D, S,A,L 1 001 AP-1 Emergency Boration (Stuck Rods)
- b. 011 Large Break LOCA (EA1.11 4.2/4.2) D, S,A, EN 2 013AP1 Transfer to Hot Leg Recirc.
- c. 008 Pressurizer (PZR) Vapor Space Accident (Relief Valve N,S,A 3*
Stuck Open) (AA2.19 3.4/3.6) 070 AP Respond to Failed Open PZR Spray valve
- d. 005 Residual Heat Removal System (A4.01 3.6/3.4) D, S,L 4P .
152 Swap RHR pumps (A train to B train) with level in the PZR.
- e. WE05 Loss of Secondary Heat Sink (EA2.1 3.4/4.4) N,S 4S 034-1 Establish MFW per EA-2-2 f.
- g. 064 Emergency Diesel Generators (A4.06 3.9/3.9) D,S,A 6 077 -1 AP Perform DG load test on 1B-B DG (with high crankcase pressure)
- h. 015 Nuclear Instrumentation System (A2.02 3.1/3.5) M,S, 7 021 AP Respond to a failure of N-41 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i. 078 Instrument Air System D,E 8 096 Respond to Loss of Control Air System
- j. 005 Residual Heat Removal System D,E,R 4P 044 Venting A-A RHR pump due to cavitation
- k. 062 AC Electrical Distribution System D,E,A 6 091 AP Transfer Controls to Aux Mode per AOP-C.04, AU 3 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 I Criteria for RO I SRO-II SRO-U
(A)lternate path 4-6 / 4-6 / 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 (control room system) (L)ow-Power / Shutdown ;::1/;::1/;::1 (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 JPM A. Candidate has to determine that emergency boration is required due to 2 stuck rods following ?t Rx trip, with the normal emergency boration valve failing to open, requiring the candidate to align alternate boration through the ~. charging pump suction. This is a Bank, Low Power, Alternate path, JPM. . JPM B. Candidate is directed to make alignment change for Transfer to Hot Leg Recirculation per ES-I.4. following a trip due to large break LOCA. The Hot leg recirculation valve will fail to open requiring RHR to be aligned to Cold Leg injection with High head pumps aligned to Hot Legs. This is a Bank, Low Power, Alternate path, JPM. JPM C. Candidate will be required to respond to failed PZR pressure instrument which causes the PZR spray valves to open with one valve sticking open. With a PZR spray valve failing to close, candidate will follow actions of AOP-1.4 and trip the RX and trip at least two RCPs to stop mitigate the depressurization to prevent SI actuation. This is a Modified Bank, Alternate Path JPM. Original JPM JPM D. Plant is in Mode 4, and Candidate is directed to transfer RHR pumps from B train to A train. This is a Bank, Low PowerJPM. JPM E. Candidate is directed to establish a Secondary Heat Sink using Main Feed Water System following a Rx Trip. MFW will be required due to a failure of all AFW pumps. This is a New JPM. JPM G. Candidate is to perform a quick start ofEDG A-A and load the EDG. While the candidate is loading the EDG, a high crankcase condition will develop requiring a manual emergency trip of the EDG. This is a Bank, Alternate Path JPM. JPM H. Candidate will respond to failed Nuclear Instrument (N41) High. Control Rods will be stepping in at maximum rate, Candidate will take Rod bank selector switch to Manual and proceed to remove failed channel from service. This is a Modified Bank, JPM. Original JPM (021) had N-41 failing low from --45%, this JPM has N-4 1failing High, requiring immediate manual action to stop control rod movement prior to removing channel from service. JPM I. Candidate is to restore Control Air pressure following a Loss of Off-site power. Candidate will restore Control and Service Air compressors using EA-32-1. This will require a Local Manual start. This a Bank JPM. JPM J. Candidate is directed to vent the lA-A RHR pump due to pump cavitating during mid-loop operation. This venting is done locally and is required to be performed to return the RHR pump to service. This is a Bank JPM and is performed in the RCA. JPM K. Candidate is directed to perform checklist 3 of AOP-C.04 Shutdown from Auxiliary Control Room, following an event which requires Control Room Abandonment. This a Bank, Alternate Path JPM.}}