ML110630047

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Final Outlines (Folder 3)
ML110630047
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 01/11/2011
From: Wasong A
Exelon Generation Co
To: D'Antonio J
Operations Branch I
Hansell S
Shared Package
ML102220201 List:
References
TAC U01769
Download: ML110630047 (20)
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Text

ES-401 BWR Examination Outline FORM ES-401-1 Facility Name: Peach Bottom Date of Exam: 01/31/2011 RO KIA Category Points SRO-Only Points Tier 1.

Group K 1 KtwJ=K 2 5 K A A A A G 6 1 2 3 4

  • Total A2 G* Total 1 4 4 3 3 3 20 4 3 7 Emergency &

Abnormal 2 1 1 1 NIA 2 1  !\I/A 1 7 2 1 3 Plant Evolutions Tier Totals 5 5 4 5 4 4 27 6 4 10 1 3 2 3 3 2 2 2 2 3 2 2 26 3 2 5 2.

Plant 2 1 1 1 1 2 1 1 1 1 1 1 12 0 2 1 3 Systems Tier Totals 4 3 4 4 4 3 3 3 4 3 3 38 5 3 8

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

7.* The generic (G) KJAs in Tiers 1 and 2 shall be selected from Section 2 of the KJA 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 KJAs.

8. On the following pages, enter the KJA 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 equiproent 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 KJA catalog, and enter the KJA numbers. descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to KJAs that are linked to 10 CFR 55.43.

ES-401, Page 17 of 34 2011 NRC Form ES-401-1, 3 - Written Outline Rev 1.xls

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)

E/APE # I Name I Safety Function K

1 K

2 K

3 A

1 1, ,Q>

i KIA Topic(s) IR #

295001 Partial or Complete Loss of Forced 0 <:">. ;*i.

limiting cycle OSCillation Plant-Specific 2.5 1 Core Flow Circulation f 1 & 4 4 I

0 , "

295003 Partial or Complete Loss of AC f 6 Containment isolation 3.7 1 6

?

I Knowledge of the design. procedural, and operational 295004 Partial or Total Loss of DC Pwr f 6 differences between units 3.8 1 295005 Main Turbine Generator Trip I 3 0 I*"', I . Recirculation system 3.2 1 3 '> f, .

0 295006 SCRAM 11 I *.*** . , Reactor power 4.2 1 6 .

295016 Control Room Abandonment 17 0 ..... '.'

DC. electrical distnbulion 2.8 1

.' .~.'

5 0

295018 Partial or Total Loss of CCW f 8 .', .. ' Effects on componenVsystem operations 3.5 1 1 ,

i:r Ability to venfy system alarm setpOints and operate controls 295019 Partial or Total Loss of Ins!. Air I 8 Ii Identified In the alarm response manual.

4.2 1 0 ii, 295021 Loss of Shutdown Cooling 14 Thermal stratification 3.3 1 2 ,

295023 Refueling Acc f 8 0 ;1.**.. . . RadIation momtoring equipment 3.4 1 3 "/

[: M. Knowledge of how abnormal operating procedures are used In 295024 High Drywell Pressure 1 5 3.8 1

k. ~ conjunclion with EOPs.

.~:

295025 High Reactor Pressure 1 3 Reactor pressure 4.3 1

'.\ ,

~95026 Suppression Pool High Water 0 I" Suppression pool cooling 3.9 1 Temp. 15 2

.,i 295027 High Containment Temperature 15 0 0

295028 High Drywell Temperature 15 Drywell temperature 4.0 1 1 I.:

295030 Low Suppression Pool Wtr Lvi f 5 0

8 i",:'.:

SRV discharge submergence 3.5 1 295031 Reactor Low Water Level 12 0 ',; Adequate core cooling 4.6 1 4'

~95037 SCRAM Condition Present and ...*...

0 4.5 Reactor Power Above APRM Downscale or SBlC 1 4

Unknown 11 0

295038 High Off-site Release Rate / 9 1

'.:1 Stack-gas monitOring system. Plant-Specific 3.9 1 600000 Plant Fire On Site /8 0  : Actions contained In the abnormal procedure for plant fire on 2.8 1 4 site

.", ':,\>;

700000 Generator Voltage and Electric Grid 0 'C. Over-excItation 3.3 1 Disturbances 16 2 .

KIA Category Totals: 4 4 3 3 g> 3' Group Point Total: 20 2011 NRC Form ES-401-1, 3 - Written Outline Rev 1.xls

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 E/APE # I Name I Safety Function 1 2 3 1 f1!G I', ,i KIA Topic(s) IR #

0 295002 Loss of Main Condenser Vac , 3 iF I"~, Loss of heat sink 3.6 1 3

7:

295007 High Reactor Pressure I 3 0 295008 High Reactor Water Level' 2

'" 0 I' .."':

295009 Low Reactor Water Levell 2 0 I

0 295010 High Drywell Pressure 15 Nitrogen makeup system: Plant-Specific 2.6 1 4

295011 High Containment Temp' 5 0

\ ~.

295012 High Drywell Temperature 15 f***  ! *. ' 0 i

295013 High Suppression Pool Temp.' 5 0

";:7: .. ..

295014 Inadvertent Reactivity Addition 11 I .'. bility to perform spee,f,e system and integrated plant 4,3 1

_,:,' procedures dunng all modes of plant operation.

I ..

295015 Incomplete SCRAM! 1

'm I: 0

'.' I***.*..

295017 High Off-site Release Rate 19 i 0 295020 Inadvertent Cont. Isolation 15 & 7 o~ Drywall ventilahonleoolrng system 32 1 2:IL 0  ;.: I:.**.**.,* . Reactor water cleanup system: Plant-Specific 295022 Loss of CRD Pumps I 1 2.5 1 4 ". ,

0  ;/

1 295029 High Suppression Pool Wtr Lvi' 5 i Lowering suppression pool water level 3.6 1 2 /,:

b*

1!=2 High Secondary Containment Area 0 erature I 5 I'"

295033 High Secondary Containment Area t::, 0 Radiation Levels I 9 i': ,.;,}

295034 Secondary Containment Ventilation *<f r Ventilation radiation levels 3.8 1 High Radiation! 9 ,,"

295035 Secondary Containment High 0

Differential Pressure I 5 295036 Secondary Containment High ". 0 SumplArea Water Levell 5  ; .:

I>

500000 High CTMT Hydrogen Conc, I 5 0 KIA Category Totals: 1 1 1 2 l+/- IGroup Point Total: 7 2011 NRC Form If!~r1fo"-~,~~ ~illt~'butline Rev 1.xls

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

System # I Name K

1 K

2 K

3 K

4 K

5 K

6 AA A 1 ,,'2 3 AI"~

4 KIA Topic(s) IR #

0 .., ',,'

203000 RHR/LPCI: Injection Mode Surveillance for all operable components 3,1 1 9 [,'

0 205000 Shutdown Cooling Motor operated valves 2,5 1 2 i,,'

206000 HPCI 1 0 ,i Nuclear boiler instrumentabon. BWR-2, 3, 4, Low 3,4; suppression poollsve;' BWR-2, 3, 4 2

2 7 ',/, 3,4 207000 Isolation (Emergency)

Condenser 0 i"f.

0 209001 LPCS DC, power 2,8 1 4

209002 HPCS  :

0

[".

211000 SLC 0 'i4~ Core plate differential pressure indication: Ability to verify 2,6; 2

3 I ,~, that the alarms are conSistent with the plant conditions 4.2 0

212000 RPS RPS motor-generator sets 3.2 1 1 r',{

0 2150031RM 2 .,: 24/48 vott D,C power: Plant-Specific 3.6 1 215004 Source Range Monitor 0 i,:

215005 APRM I LPRM 0 o ;,' Control rod ~ock status; Verification of proper functioning! 3,7; I'"',; 8 6 :~,""'" operability 3.6 2

0 217000 RCIC Reactor waler level 4,0 1 3 "

0 218000 ADS ADS log'c operallon 3,8 1 1 . ,,','

223002 PGIS/Nuclear Steam Supply 1 ":'" Contalnment drainage system 2,8 1 Shutoff 4 """

239002 SRVs 0 01., Tail pipe temperatures: Ability to expla,n and apply system 3,6:

limits and precautions.

2

.:".'[ 3,8 2#1 0 ,

259002 Reactor Water Level Control ". GEMAC/Foxboro/Bailey controller operation: Plant-Spec~ic 3,1 1 1

0 ...........

261000 SGTS Fan start 3,2 1 1\. 2 .

0 I .. '

262001 AG Electrical Distribution Major system loads 3.5 1 1 ....

0 262002 UPS (AG/DC) Containment jsolatlon system: Plant~Specific 2,9 1 8

0 .....

263000 DC Electrical Distribution Battery charging/dischaf1l,ng rate 2.5 1 1

0 0 . ' " Parallel operation of emergency generator: Indicating 3,5; 264000 EOGs 2 1* 3 lights, meters, and recorders 3.4

.~.

0 0 "

2,9; 300000 tnstrument Air Cross~tied vnrts; Securing of lAS upon loss of cooling wate 2 3 3 2.8 0

~OOOOO Component Gaoling Water Automatic start of standby pump 3.4 1 1

L *..

,i '"

I

, .j\:,;,

0 KIA Category Totals: 3 i 21 3 \3 2 2 2 .2 3 2 i' Group POint Total: 26

ES-401*1 5 Form ES-401*1 ES-401 BWR Examination OuUine Form ES-401-1 Plant System!; Tier 2/Group 2 (RO)

System # I Name ~ ~ ~ ~ ~ ~ ~ I~ ~ ~ ~... *.

  • KIA Topic(s) IR #

201001 CRD Hydraulic o 201002 RMCS o If~_:_:_:_:_:_:_:_:_I_R_O_d_a_n_d_D_n_.v_e_M_e_ch_an_i_sm_+--'_1-~+-f_1-+-f:-"1-+_IT"'~ ~~"

3.2 o

o i

201006RWM wer supply loss: P-Spec(Not-BWR6) 2.5 202001 Recirculation I' . nowledge of system purpose andlor function 3.9 Ir-------------------------~~+_~_r_+~~+_~_r_+--

202002 Recirculation Flow Control o y' 0 T"

?04oooRWCU  ;' Response to system isolations 3.6 3

214000 RPIS o I

~15001 Traversing In-core Probe I o 215002 RBM ,', o 216000 Nuclear Boiler Inst.

o Recorders 3.3 1

219000 RHR/LPCI: ToruS/Pool Cooling Mode o 223001 Primary CTMT and Aux.

o ,:-. Drywall cooling fans' Piant-Specific 9

2.7 226001 RHR/LPCI: CTMT Spray Mode o 230000 RHR/LPCI ToruS/Pool Spray Mode o 233000 Fuel Pool Cooling/Cleanup o 234000 Fuel Handling Equipmen1 3.0 239001 Main and Reheat Steam o 1239003 MSIV Leakage Control o 1 241000 ReactorlTurbine Pressure Regulator o 245000 Main Turbine Gen. I Aux. I o 256000 Reactor Condensate o t ....

259001 Reactor Feedwater 6

" Reorculation 3.1 268000 Radwaste o I

271000 Offgas o 1 272000 Radiation Monitoring o Automatic actions to contain the radioactive release in the 3.7 2 I event that the predetermined release rates are exceeded o

~860oo Fire Protection ,. A.C. electrical distribution: Plant-Specific 3.1 1

o 288000 Plant Ventilation 2

Differential pressure control 3.2 290001 Secondary CTMT o 290003 Con1rol Room HVAC o ,. ..;; Control mom pressure 2.5 4

290002 Reactor Vessel Internals o o

KIA Category Totals: 1 1 1 1 2 1 11 1 1 t Group PointTotal' 12

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 (SRO)

\<~ i,,~<

K K K A E/APE # I Name I Safety Function KIA Topic(s) IR #

1 2 3 1 i""

"1 295001 Partial or Complete Loss of Forced Core Flow Circulation I 1 & 4 IA. Power/flow map 3,8 1 295003 Partial or Complete Loss of AC I 6 02" Knowledge of the bases in Technical Specifications for 4.2 1 25 limiting conditions for operations and safety limits.

295004 Partial or Total Loss of DC Pwr J 6 . '. 0 295005 Main Turbine Generator Trip J 3 ' I 0

.... 0 1295016 Control Room Abandonment 17 0 1295018 Partial or Total Loss of CCW 1 8 System pressure 2.9 1 L5i .,

295019 Partial or Total Loss of Inst. Air 18 . ', 0

~L 295021 Loss of Shutdown Cooling 14 0 295023 Refueling Acc I 8 *r. 0

.c:

295024 High Drywell Pressure I 5 I+/-

I*" ,

',c'

. ' Drywell radiation levels 4.0 1

~95025 High Reactor Pressure I 3 "

0

,-,- ~

_" , ...v t'~,g ...~.m~ g"v ."",.~ u~uu .v g~~v~~

295026 Suppression Pool High Water Temp. 04; the status of safety functions. such as reactivity control, 4.6 1 5 21' core COOling and heat removal, reactor coolant system f*

295027 High Containment Temperature 15 0 Q4, Knowledge of the operational implications of EOP 295028 High Drywell Temperature I 5 '" 4.3 1

',.. ' ~O warnings, cautions. and notes.

295030 Low Suppression Pool Wtr Lvl/5 0 295031 Reactor Low Water Level/2 0 j /,>"

295037 SCRAM Condition Present

'(i and Reactor Power Above APRM Suppression pool temperature 4.1 1 Downscale or Unknown 11 4 ..*.

fe*..

295038 High Off-site Release Rate I 9 0 iJ 600000 Plant Fire On Site I 8 , 0 700000 Generator Voltage and Electric Grid .. :..*,...... 0 Disturbances I 6 KIA Category Totals: 0 0 0 0 4 oup Point Total: 7 2011 NRC Form ES-401-1, 3 - Written Outline Rev 1.xls

ES-401 3 Form ES-401-1 ES*401 BWR Examination Outline Form ES401*1 Emergency and Abnormal Plant Evolutions* Tier 1/Group 2 (SRO)

E/APE # I Name I Safety Function K K K A A ~ KJA Topic(s) IR #

1 2 3 1 .2 295002 Loss of Main Condenser Vac I 3 0

... \

295007 High Reactor Pressure I 3 0

....... . ~

295008 High Reactor Water Levell 2 '~. Knowledge of abnormal condition procedures 4.2 1 295009 Low Reactor Water Levell 2

. .2. 0 295010 High Drywell Pressure I 5 0

./

t"****

.... " [,;,J 295011 High Containment Temp I 5

  • r, 0 295012 High Drywell Temperature /5 295013 High Suppression Pool Temp. /5 t f2 Drywell pressure 4.1 1 0

I*.d*'

295014 Inadvertent Reactivity Addition /1 0

. I******

295015 Incomplete SCRAM 11 I' 0

... :I* '.<

295017 High Off*site Release Rate / 9 0

).

295020 Inadvertent ConI. Isolation / 5 & 7 t> 0

.i {

[f/

295022 Loss of CRD Pumps 11 0 295029 High Suppression Pool Wtr LviI 5 0 I

295032 High Secondary Containment Area 0 Temperature I 5 .

295033 High Secondary Containment Area 0 ", Equipment operability 3.2 1 Radiation Levels I 9 2 **

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 I';

500000 High CTMT Hydrogen Conc. I 5 1;.;" l~/ 0

["*;.,... 1';' 3 KJA Category Totals: 0 0 0 o I - " } Group Point Total:

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

K K K K K K A A A I~ ,~,

System # I Name p. KIA Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 203000 RHRlLPCI: Injection 0 205000 Shutdown Cooling Mode 0

., i'/"

206000 HPCI 0 f

207000 Isolation (Emergency) I'.

0 Condenser ','

209001 LPCS ' ,< 0 209002 HPCS 0 i*

211000 SLC  :{:~" 0 212000 RPS Jl

<4

! >/ Nuclear instrument system failure 3,7 1 2150031RM  !",;' 0

,,?iF 215004 Source Range Monitor 0

... I'"

215005 APRM I LPRM '.' I,> 0 217000 RCIC I****,**** 0

+/-

218000 ADS 0 .. *,i" .** Loss of A C. or D.C. power to ADS valves 3,6 1 5

223002 PCIS/Nuclear Steam Supply Shutoff "": 0 239002 SRVs ':' . 0 f259002 Reactor Water Le\lel Control

. 0

'"',,' must be reported to Internal organizations or external lit, _.' , ,e,o,~.v ,,><0'" v~o.ouv, ,,><a.v" u ...

4,1 261000 SGTS 1

~: agendes, such as the State, the NRC, or the transmission 02ii Knowledge of limiting conditions for operations and safety 262001 AC Electrical Distribution 262002 UPS (AC/DC) limits 4,7 1 0

i' 263000 DC Electrical Distribution ,.;:. 0 264000 EDGs

..  : 0 300000 Instrument Air 0 0 . 5 HigMow ccw temperature 3.0 1 400000 Component Cooling Water 3 .

'.". 0 KIA Category Totals: 0 0 0 0 0 0 03 0 o I_~ Group Point Total: 5

ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-40H Plant Systems Tier 2/Group 2 (SRO)

A~A ~ ~*.~

K K K K K K System # I Name KJA Topic(s) IR #

1 2 3 4 5 6 1 3

~01001 CRD Hydraulic 0

~01002 RMCS .*....... 11r 0

~01003 Control Rod and Drive Mechanism .. ' 0 1201004 RSCS 0 k

~01005 RCIS I*,]  ;~'

I 0

~01006 RWM 0 t.'*L.

~02001 Recirculation Ii 0

~02002 Recirculation Flow Control '.,

0

~04000RWCU n

0 1214000 RPIS ...*.. 0

~15001 Traversing In-cere Probe 0

~15002 RBM I,*; 0

~16000 Nuclear Boiler Insl.

f19000 RHRILPCI: ToruS/Pool Cooling I "t.

1<

.. 0 0

~ode 223001 Primary CTMT and Aux. I; 0

.c.

~26001 RHRlLPCI' CTMT Spray Mode Of; Ablllty to interpret and execute procedure steps" 4.6 1

.'~

~30000 RHRILPCI: Torus/Pool Spray Mode ";.~

0

~33000 Fuel Pool Cooling/Cleanup c" 0 I I

~ [.

Ii'.

~34000 Fuel Handling Equipment I.: *......... '. k f* 0

~39001 Main and Reheat Steam

..; ..  : . 0

,ci'

~39003 MSIV Leakage Control 0 I*:..**

1

~41 000 ReactorlTurbine Pressure Regulator Tumina top. Plant*Specific 3.8 1 7

~45000 Main Turbine Gen. I Aux.

.. . 0 t*, Feedwater heater string top: Plant-Specific

~56000 Reactor Condensate

.J~ 2.9 1

~59001 Reactor Feedwater 0

~68000 Radwaste .'.

0

~71 000 Offgas 0

~72000 Radiation Monitoring . 0 l2S6000 Fire Protection ....... 0

~88000 Plant Ventilation 0

    • f 1290001 Secendary CTMT 0

~90003 Control Room HVAC ..... 0 290002 Reactor Vessel Internals 0 L

0 otEtttt KIA Category Tolals: 0 0 0 tTotal: 3

'd 2011 NRC Form ~*'6,,-f,"'j~1JriRb~'butline Rev 1.xls

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 IFa~ty Name:feach 8VLLVIII Date of Exam:01/31/2011 Category KJA# Topic l etc. 4.2 1 lo..,c, cnivll;:)

AbIlity idenllfy and Interpret diverse indications to validate the response of another 2.1. 45 Ind,cator. 4.3 1 2.1.

1Subtotal ~ 2 2.2. 38 "~. of conditions and limitations In the facility lioonse 3.6 1

!Ability to interpret control room indications to verify the status and operation of a system, 2.2. 44 I and how operator actions and directives affecl plant and system conditions 4.2 1 lAbility pre-startup procedures for the facility, including operating those controls 2.2. 01 4.4 1

12. i 1 plant equipment that could affect reactiv ty Equipment lAbility determinethe expected plant configuration using design and configurallon 2.2. 15 Icontrol w~"'c, "ouv, suc\1 as drawmgs. line-ups, lag-outs, atc 4.3 1 IControl 2.2.

2.2.

ISubtotal 2.3. 04 Knowledge of radiation exposure limIts under normal or emergency condilions 3.2

~

1

--=r lAbility use radiation monitoring syslems, such as flX"d radiation monitors and alarms, 2.3. 05 !portable survey instruments. personnel monitoring equipmenl. etc, 2.9 1 2.3. 11 AbIlity to control radialion releases, 3.8 1 13.

, pertaIning to licenseo operator QUileS, sucn 12.3 13 ias ""O~~:AO:{} ratii"tinn ::,~;~~

Radiation i , fuel handling 3.8 1

,Control aocess to locked i careas, aliamna fIllers. elc,

.v'mu~, radiation or contamination hazards that may arise during normal, abnormal.

2.3. 14 lor, "~' conditions or activities. 3.8 1 2.3.

,SllQ!otal ~ 2 2.4. 27 Knowledge of "fire in the plant" procedures. 34 1 2.4. 31 ..~' of annuncIator alarms, indIcations, or response procedures 4.2 1

!AboIity abnormal indications for system operating parameters thaI are entry-

[4. 2.4. 04 Ilevel for emergency and abnormal operating procedures, 4.7 1 Emergency Procedures 2.4.

II Plan 2.4.

2.4.

,Subtotal £ 1 ITier 3 Point Total j~ 7 ES-401, Page 27 of 34 2011 NRC Form ES-401-1, 3 - Written Outline Rev 1.xls

ES-401 Record of WA:,,,,,,,'PAI'I KJAs Tier I Randomly Reason for Rejection ~

Group Selected KIA RO 111 295028 This KIA is essentially the same as the one for 0 #83 (295012 AA2.02). Replaced with 0#51 EA2.04 KIA 295028 EA2.01.

Unable to construct a question for this KIA that meets the requirements of NUREG RO 1/1 295019 1021 (there are no EOP warnings, cautions or notes relative to the Instrument Air 0#55 G2.4.50 System). Replaced with KIA 295019 G2.4.50.

RO 1/1 295024 This generic KIA is not a good fit for the High Drywell Pressure EPE. Replaced with 0#56 G2.2.36 KIA 295024 G2.4.8.

RO 1/1 295005 This KIA is for BWR-2 plants and does not apply to Peach Bottom, which is a BWR-4.

0#57 AK2.09 Replaced with KIA 295005 AK2.03.

Unable to construct a question for this KIA that meets the requirements of NUREG RO 1/2 295020 1021 (cannot link this KIA to Inadvertent Containment Isolation). Replaced with KIA 0#64 G2.1.7 295020 AA1.02 (Generic KIA's were over-sampled).

I RO 1/2 295012 The High Drywell Temperature APE/EPE was over-sampled. Replaced with KIA I 0#65 G2.1.23 295014 G2.1.23.

RO 2/1 400000 This KIA is essentially the same as the one for 0 #26 (400000 K4.01). Replaced with 0#15 A2.01 KIA 264000 A2.01 (400000 was over-sampled).

Unable to construct a question for this KIA that meets the requirements of NUREG R03 G2.1.38 1021 (originally submitted question was determined to be LOD-1). Replaced with KIA 0#66 G2.1.3.

SRO 1/1 295031 This KIA is very similar to the one for 0 #52 (295031 EA2.04). Replaced with KIA 0#78 EA2.01 295037 EA2.04 (295031 was over-sampled).

Unable to construct an SRO-only question for this KIA that meets the requirements of SRO 1/1 295001 NUREG-1021 (system parameters that are entry level for Tech Specs is RO 0#79 G2.2.42 knowledge). Replaced with KIA 295001 AA2.01.

Unable to construct an SRO-only question for this KIA that meets the requirements of SRO 1/2 295008 NUREG-1021 and KIA is not tied to 10CFR55.43(b). Replaced with KIA 295008 0#84 G2.1.30 G2.4.11.

There are no RO tasks performed outside the Main Control Room associated with High SRO 1/2 295033 Secondary Containment Radiation Levels. Replaced with KIA 295033 EA2.02 0#85 G2.4.34

~ (Generic KIA's were over-sampled).

Unable to construct an SRO-only question for this KIA that meets the requirements of SRO 2 / 1 300000 NUREG-1021. In addition, System 300000 was already sampled on the RO section.

0#86 A2.01 Replaced with KIA 218000 A2.05.

SRO 2 / 1 400000 Unable to construct an SRO-only question for this KIA that meets the requirements of 0#89 G2.4.45 NUREG-1021. Replaced with KIA 400000 A2.03 (Generic KIA's were over-sampled).

Unable to construct an SRO-only question for this KIA that meets the requirements of SRO 2/2 268000 NUREG-1021 and KIA is not tied to 10CFR55.43(b). Replaced with KIA 241000 A2.17 0#92 G2.4.31 (Generic KIA's were over-sampled).

Unable to construct an SRO-only question for this KIA that meets the requirements of SRO 2 f 2 226001 NUREG-1021 (SPDS is not used for "decision-making"). Replaced with KIA 226001 0#93 G2.1.19 G2.1.20.

2011 NRC Form ES-401-4 Rev. 1.doc ES-401, Page 27 of 33

ES-301 Administrative ODICS Outline Form ES-301-1

! Facility: Peach Bottom Date of Examination: 01/31/2011 Examination Level: RO r8.J SRO 0 Operating Test Number: 2011 NRC Administrative Topic I Type I Describe activity to be performed (See Note) Code*

1 201 .45. Manually Calculate Drywell Bulk Average Conduct of Operations D,R Temperature (Alternate Path - Failed Temperature POints)

[PLOR241C]

2.1.25 - Perform RCS Leakage Test lAW ST-0-020-560 Conduct of Operations D,R

[PLOR244C]

~------------------~-------+----------------------------------------

Equipment Control D,P,S 2.3.11 - PRO Duties for Liquid RadWaste Discharge Radiation Control (2008

[PLOR258C]

2.4.28 - Perform Personnel Notifications During a Security Emergency Plan M,S Threat [PLOR350C]

I 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, Page 22 of 27 Last printed 12/8/20103:02 PM

ES-301 Administrative Outline Form ES-301-1 Facility: Peach Bottom Date of Examination: 01/31/2011 Examination Level: ROO SRO ~ Operating Test Number: 2011 NRC Administrative Topic (See Note)

I Type Code*

Describe activity to be performed 1--.. . .--~"----~--~

2.1.18 - Identify and Specify Required Notifications per OP-AA-1 06-1 02 "Accidents Involving the Transportation of Conduct of Operations N,R Rad Materials" Based on Completed Attachment 1 Transportation Accident! Incident Form [PLOR351C]

---.--.--.~.

2.1.25 - Perform RCS Leakage Test lAW ST-O-020-560 Conduct of Operations N,R and Evaluate Tech Spec Requirements [PLOR354CA]

~

Equipment Control D,R 2.2.23 - TE~ch Spec Action Log Entry [PLOR221C]

~---

2.3.4 - Review and Authorize an Emergency Exposure Radiation Control D,R

[PLOR249C]

I 2.4.41 - EAL Classification with State and Local Emergency Plan D,R I

Notification [PLOR232C]

IN-OTE: 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 C::: 1; randomly selected)

II ES 301. Page 22 of 27 Last printed 1/28/2011 10: 18 AM

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 01/31/2011 Exam Level: RO!8l SRO-I 0 SRO-U 0 Operating Test Number: 2011 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 I Safety I Code* I Function

a. 201003 A2.03 - Control Rod and Drive Mechanism** Respond to a Rod A,D, I 1 Drift (Alternate Path - Second Rod Drifts) [PLOR307CA] [Set 11 EN,S I
b. 259001 A4.02 - Reactor Feedwater System - Shutd!::>wn the "A" RFP Turbine (Alternate Path - Min Flow Valve Fails Closed) [PLOR303CA]

A,D,S  ! 2

[Set 2] , I

c. 218000 A4.03 - Automatic Depressurization System - ADS Reset I 0, EN, i 3 Following Blowdown [PLOR023C] [Set 3] L, S

. d. 206000 A3.07 - High Pressure Coolant Injection - Startup HPCI in CST A, EN, 4 to-CST Mode (Alternate Path - Exhaust Diaphragm Rupture) M, S

[PLOR353CA] [Set 4]

e. 219000 A4.13 - RHR/LPCI: Torus Cooling Mode - HPSW Injection into 0, EN, 5 the Torus [PLOR081C] [Set 1] L, S
f. 262001 A4.01 - A.C. Electrical Distribution - Restoration of 4KV Buses A, D,S 6 from 2SUE (Alternate Path - 2SU-A Breaker Closes) [PLOR344CA]

[Set 4] .

g. 400000 A4.01 - Component Cooling Water System - ECW System M,S 8 I-- Makeu[> to Tower Using a HPSW Pum[> [PLOR-270C] [Set 3] i
h. 261000 A4.03 - Standby Gas Treatment System - Manually Start SBGT I D,EN,S 9 System [PLOR044C] [Set 2l I I In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i i. 202001 A4.08 - Recirculation Flow Control System - Manual Operation of A, E, M, R I 1 Scoop Tube Positioner (Alternate Path - Clutch Fails to Engage)

I [PLOR346PA] I

j. 241000 A2.01 - ReactorlTurbine Pressure Regulating System - Swapping A, 0, E, R I I 3 EHC System Pressure Regulators - Unit 3 (Alternate Path - Backup Pressure Regulator Instabilities) [PLOR334PA] I
k. 295018 M1.01 - Component Cooling Water System - Loss of RBCCW D,E,R 8

{Plant Actions for the Instrument Nitrogen System) [PLOR096P] \ I

@ 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. I

[

  • Type Codes i

Criteria for RO 1 SRO-II SRO-U Last printed 12/8/20103:02 PM

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2

~-------------------------------

Facility: Peach Bottom Date of Examination: 01/31/2011 Level: RO 0 SRO-I [gI SRO-U Operating Test Number: 2011 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

.~.-~-.

~--

I  !

~ System I JPM Title Type Safety Code* I Function

~a. 201003 A2.03 - Control Rod and Drive Mechanism - Respond to a Rod I A,D, I 1 Drift (Alternate Path - Second Rod Drifts) [PLOR307CA] [Set 1) ~_ EN,S

b. 259001 A4.02 - Reactor Feedwater System - Shutdown the "A" RFP . A, D,S I 2 Turbine (Alternate Path - Min Flow Valve Fails Closed) [PLOR303CA] I

[Set 2]

c. 218000 A4.03 - Automatic Depressurization System - ADS Reset D, EN, 3 Following Blowdown [PLOR023C] [Set 3) L, S
d. 206000 A3.07 - High Pressure Coolant Injection - Startup HPCI in CST- A, EN, 4 to-CST Mode (Alternate Path - Exhaust Diaphragm Rupture) M,S

[PLOR353CA] [Set 4]

e. 219000 A4.13 - RHR/LPCI: Torus Cooling Mode - HPSW Injection into II D, EN, 5 I---the Torus [PLOR081 C] [~et 1] L, S
f. 262001 A4.01 - A.C. Electrical Distribution - Restoration of 4KV Buses A,D,S 6 from 2SUE (Alternate Path - 2SU-A Breaker Closes) [PLOR344CA]

[Set 4] . .

g.

c- -- - - - - -

h. 261000 A4.03 - Standby Gas Treatment System - Manually Start SBGT D, EN, S J 9 S~stem IPLOR044C] [Set 2l In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
i. 202001 A4.08 - Recirculation Flow Control System - Manual Operation of A, E, M, R I 1 Scoop Tube Positioner (Alternate Path - Clutch Fails to Engage)

[PLOR346PA] ..

j. 241000 A2.01 - ReactorlTurbine Pressure Regulating System - Swapping - A, D, E,-;r 3 EHC System Pressure Regulators - Unit 3 (Alternate Path - Backup Pressure Regulator Ins1a.!>ilities) [PLOR334PA] I I I

~

k. 295018 AA1.01 - Component Cooling Water System - Loss of RBCCW D, E, R 8 I

@ 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 :

overla those tested in the control room. __ __ I

"--_________*_T_y_p_e_C_o_d_es* i __ Criteria for RO 1 SRO-II SRO-U . II Last printed 12/8120 I0 3:02 PM

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 01/31/2011 Exam Level: RO 0 SRO-I 0 SRO-U [gI Operating Test Number: 2011 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System 1 JPM Title  ; Code* Function I

a.

I

b. 259001 A4.02 - Reactor Feedwater System - Shutdown the "A" RFP 2 Turbine (Alternate Path - Min Flow Valve Fails Closed) [PLOR303CA] I

[Set 2] i I c.

i I

d. 206000 A3.07 - High Pressure Coolant Injection - Startup HPCI in CST I A, EN, 4 to-CST Mode (Alternate Path - Exhaust Diaphragm Rupture) M,S

[PLOR353CA] [Set 41 I

I e. 219000 A4.13 - RHR/LPCI: Torus Cooling Mode - HPSW Injection into I D, EN, 5 I the Torus [PLOR081CI [Set 1] L,S

, f.

g.

h.

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

i. 202001 A4.08 - Recirculation Flow Control System - Manual Operation of A, E,M, R 1 Scoop Tube Positioner (Alternate Path - Clutch Fails to Engage)

I

[PLOR346PA1.

j.

k. 295018 AA1.01 - Component Cooling Water System - Loss of RBCCW I D,E, R (Plant Actions for the Instrument Nitrogen System) [PLOR096P] I

@ 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 Criteria for RO 1 SRO-II SRO-U Last printed 12/8/20103:02 PM

(A)lternate path 4-6 14-6 1 2-3 (C)ontrol room (D)irect from bank .:::9 I <8 1 <4 (E)mergency or abnormal in-plant ~1 / ~1 1~1 (EN)gineered safety feature / - I ~ 1 (control room system)

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

(R)CA ~1 I ~1 1~1 (S)imulator ES-301, Page 23 of 27 Last printed 12/8/20103:02 PM

Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #1 (new} OpTest No. 2011 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 cross-tie 480V Summary auxiliary load center 1PS4 with 3PS4 to allow for scheduled preventative maintenance on the 1PS4 breaker. Shortly after this, the running CRD pump will trip due to a clogged pump suction filter, requiring the crew to bypass the filter and restore a CRD pump to service in accordance with ON-107 "Loss of CRD Regulating Function". After CRD has been restored, 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 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-103 "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 feedplJmp discharge bypass valve failure by batch feeding through the 'C' reactor feed pump 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 between +5 to +35 inches with Condensate.

Initial IC-81, 100% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Event No. No. Type* Description 1 See Scenario Guide N PRO Cross-tie 480V auxiliary load center 1PS4 with 3PS4 CRS 2 See ScenariO Guide C URO Loss of CRD pump due to clogged pump suction filter I bypass filter TS CRS and restore CRD pump to service (Tech Spec) 3 See Scenario Guide R URO  ! Turbine stop valve fails closed I power reduction CRS 4 See Scenario Guide C PRO Inadvertent HPCI initiation I shutdown HPCI (Tech Spec)

TS CRS 5 See Scenario Guide M ALL HPCI steam leak into secondary containment 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, I 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

>I' (N)ormal, (R)eactlvlty, (I)nstrument, (C)omponent, (M)aJor, (TS) Tech Spec

Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #2 (modified} OpTest No. 2011 NRC Examiners Operators CRS (SRO)

URO (ATC)

PRO (BOP)

Scenario The scenario begins with the reactor at 100% power. The turnover will direct the crew to perform the Summary Master Trip Solenoid Valves Routine Test. Following this test, the operating TBCCW pump will trip and the standby pump will fail to automatically start, requiring the crew to manually start the standby TBCCW pum p and restore the system in accol'dance with ON-118 "Loss of TBCCW". Next, an emergency service water pump will spuriously start, requiring the crew to remove the pump from service and apply Tech Specs for the inoperable ESW pump.

The 'A' recirc pump will then trip, requiring the crew to carry out the actions of OT-112 "Unexplained!

Unexpected Change in Core Flow", which includes inserting GP-9-2 "Fast Reactor Power Reduction" Table 1 control rods. The crew should also establish single loop operation per GP-5 "Power Operations" and consult Technical Specifications.

Next, a sustained loss of Stator Cooling will occur, requiring the crew to scram the reactor. An ATWS (electrical) will require the crew to execute T-101 "RPV Control" and T-117 "Level/Power Control".

The main turbine will trip several minutes into this event as a result of the loss of Stator Cooling, complicating the crew's efforts to respond to the ATWS and challenging Primary Containment due to SRV actuation. When SBLC is initiated, RWCU will fail to automatically isolate, requiring the crew to manually isolate RWCU. In addition, the crew will not be able to restore normal instrument nitrogen.

which will require aligning a backup source of nitrogen to the SRVs to ensure they are available for reactor pressure control. After RPV level has been lowered to control power, the ATWS will be terminated using T-214 "Venting the Scram Air Header".

Initial IC-82. 100% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Event No. No. Type* Description 1 See Scenario Guide N PRO Master trip solenoid valves routine test CRS 2 See Scenario Guide C URO TBCCW pump trip with failure of standby pump to auto-start CRS 3 See Scenario Guide I PRO ESW pump spurious start / shutdown ESW pump (Tech Spec) i TS CRS 4 See Scenario Guide R URO Recirc pump trip / single loop (Tech Spec) / insert GP-9-2 Table 1 rods TS CRS 5 See Scenario Guide M ALL Loss of stator cooling water / scram (electric ATWS) 6 See Scenario Guide I URO RWCU fails to isolate on SBLC initiation / manually isolate RWCU CRS 7 See Scenario Guide C PRO Unable to restore drywell instrument nitrogen / place alternate CRS instrument nitrogen system(s) in service

  • (N)ormal, (R)eactlvlty, (I)nstrument, (C)omponent, (M)aJor, (rS) Tech Spec

Scenario Outline Simulation Facility Peach Bottom Scenario No. #3 {new) OpTest No. 2011 NRC Examiners Operators CRS (SRO)

URO (ATC)

PRO (BOP)

Scenario The scenario begins with the reactor at -500 psig and -5% power during a reactor startup. Following turnover, the Summary crew will continue the reactor startup in accordance with GP-2 "Normal Plant Startup" by cycling the HPCI steam supply valve and raising recirc pump speeds to the 30'Yo limiter. After these evolutions are complete, a radiation monitor will fail upscale and a drywell 1S-inch vent damper will fail to isolate, requiring the crew to isolate the penetration and consult Tech Specs. Following this event, the crew will be required to swap RBCCW pumps due to a report from the field indicating excessive seal leakage from the running RBCCW pump.

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. Following this, the 'H' SRV will inadvertently open, requiring the crew to take actions to close the valve, and to maximize torus cooling, in accordance with OT-114 "Inadvertent Opening of a Relief Valve". The crew will not be successful in closing the SRV, and a rupture in the SRV downcomer will result in pressurizing the torus air space, challenging primary containment. The crew must execute OT-101 "High Drywell Pressure", T-101 "RPV Control" and T-102 "Primary Containment Control".

When the crew inserts a manual scram due to rising drywell pressure, reactor pressure will lower below 450 psig, which along with a high drywell pressure (2 psig) signal will cause low-pressure ECCS pumps to auto-start. The crew must manually secure the ECCS pumps to prevent overfilling the RPV. Attempts to spray the primary containment will fail due to multiple spray valve failures, requiring the crew to perform a T-112 "Emergency Blowdown" when the Pressure Suppression Pressure (PSP) Limit curve is exceeded. When drywell pressure exceeds Drywell Chilled Water (DWCW) System pressure, a DWCW piping break will occur that must be manually isolated per GP-S.B, "PCIS Isolation - Groups II and III" to eliminate a release pathway into the Turbine Building via DWCW piping. The scenario will be terminated after the RPV depressurization and DWCW isolation are performed.

Initial IC-83, 5% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Event No. No. Type* Description 1 See Scenario Guide N PRO Cycle HPCI steam supply per GP-2 "Normal Plant Startup" CRS 2 See Scenario Guide R URO Raise recirc pump speeds per GP-2 "Normal Plant Startup" CRS 3 See Scenario Guide I PRO Radiation monitor upscale failure with failure of drywell18-inch vent TS CRS damper to isolate (Tech Spec) 4 See Scenario Guide C URO RBCCW pump swap due to excessive seal leakage on running pump CRS I

5 See Scenario Guide C URO 'A' Recirc pump speed oscillations (Tech Spec) I Lock up the 'A' TS CRS Recirc pump 6 See Scenario Guide C PRO SRV inadvertently opens (Tech Spec) I maximize torus cooling TS CRS 7 See Scenario Guide M ALL Rupture in SRV downcomer I valve failures prevent containment spray 8 See Scenario Guide C URO Drywell Chilled Water (DWCW) piping break I manually isolate DWCW CRS 9 See Scenario Guide ALL Emergency blowdown due to exceeding the PSP curve

  • (N)ormal, (R)eacbvlty, (I)nstrument, (C)omponent, (M)aJor, (TS) Tech Spec
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