ML111661880

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Draft -Outlines (Folder 2)
ML111661880
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 09/01/2010
From: D'Antonio J
Operations Branch I
To: Wasong A
Exelon Generation Co
Hansell S
Shared Package
ML102220201 List:
References
ES-401, ES-401-1, TAC U01769
Download: ML111661880 (22)
v • d • e


Text

ES-401 BWR Examination Outline FORM ES401-1 Facility Name: Peach Bottom Date of Exam: 01/31/2011 RO KIA Category Points SRO-Only Points Tier Group K AIA~

8:=

KKK Total A2 4 5 6 1 ")

1. 1 4 3 3 20 4 Emergency &

Abnormal Plant Evolutions 2

Totals 1

8ilii 1

3 1

2 3 N/A 2 2 2

5 2

1 4

2 3 N/A 28 1

4 7

27 26 2

6 3

1 4

2 3

10 5

2 I

2.

Plant 2 1 1 1 1 2 1 1 1 1 1 1 12 0 1 3 Systems Tier Totals 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 Categories 3 2 3 2 10 2 2 2 1 7

Note: 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (I.e .. except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA category shall not be less than two).

2. The point total for each group and tier in the proposed outline must match that specified in the table.

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. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (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. Limit SRO selections to KlAs that are linked to 10 CFR 55.43.

ES-401, Page 17 of 34 2011 NRC Form ES-401-1, 3 - Written Outline Rev O.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) r-------------------------~~~--

K A E/APE # I Name I Safety Function KIA Topic(s) IR 1

295001 Partial or Complete Loss of Forced 0 limiting cycle oscillation Plant-Specific 2.5

  • Core Flow Circulation 11 & 4 4 295003 Partial or Complete Loss of AC 16 Containment isolation 3.7 02, Knowledge of the desIgn. procedural, and operatIonal 295004 Partial or Total Loss of DC Pwr 16 03 differences between units 3.8 295005 Main Turbine Generator Trip 13 Recirculation system 3.2 295006 SCRAM 11 Reactor power 4.2 6

95016 Control Room Abandonment 17 o D C electncal distrobutlon 2.8 5

0 95018 Partial or Total Loss of CCW 18 Effects on component/system operations 3.5 295019 Partial or Total Loss of Ins!. Air I 8 04.. Ability to verify system alarm selpoints and operate controls f* . ".50 identified in the alarm response manual.

4.2 295021 Loss of Shutdown Cooling 14 Thermal stratification 3.3 Radiation monitoring eqUipment 3,4

.fM. Knowledge of how abnormal operating procedures are used ,n 24 High Drywell Pressure 15 os: conjunction with EOPs. 3.8 95025 High Reactor Pressure I 3 Reactor pressure 295026 Suppression Pool High Water Suppression pool cooling 3.9 emp./5 295027 High Containment Temperature 15 o 295028 High Drywell Temperature 15 Drywall temperature 4.0 295030 Low Suppression Pool Wtr Lvl/5 o SRV discharge submergence 3.5 8

295031 Reactor Low Water Levell 2 Adequate core cooling 4.6 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or 4.5 Unknown 11 295038 High Off-site Release Rate 19 Stack-gas mOnitoring system: Plant-SpecifIc 3.9 ActIons contained ,n the abnormal procedure for plant fire on 600000 Plant Fire On Site I 8 28 00000 Generator Voltage and Electric Grid o 3.3 isturbances I 6 2 KIA Category Totals: 4 Group Point Total: 20 2011 NRC Form ES-401-1, 3 - Written Outline Rev O.xls

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

E/APE # 1 Name 1 Safety Function KKK A KIA Topic(s) 2 3 1 95002 Loss of Main Condenser Vac I 3 o Loss of heat sink 3

295007 High Reactor Pressure 13 295008 High Reactor Water Level 12 295009 Low Reactor Water Levell 2 295010 High Drywell Pressure 15 o Nitrogen makeup system Plant-Specific 2.6 4

95011 High Containment Temp I 5 o 295012 High Drywell Temperature 15 295013 High Suppression Pool Temp. 15 Ability to perform specific system and Integrated plant 295014 Inadvertent Reactivity Addition 11 procedures dunn9 all modes of plant operation.

295015 Incomplete SCRAM 11 o 295017 High Off-site Release Rate 19 o 295020 Inadvertent Cont. Isolation 1 5 & 7 Drywall ventilation/cooling system 3.2 2 Loss of CRD Pumps I 1 Reactor water cleanup system. Plant-Specific o

295029 High Suppression Pool Wtr Lvii 5 Lowering suppression pool water level 2

295032 High Secondary Containment Area Temperature I 5 295033 High Secondary Containment Area Radiation Levels I 9 o

295034 Secondary Containment Ventilation Ventilation radiation levels 3.8 High Radiation I 9 295035 Secondary Containment High Differential Pressure I 5 o

295036 Secondary Containment High o SumplArea Water Levell 5 500000 High CTMT Hydrogen Conc. I 5 KIA Category Totals: 2 . Group Point Total:

2011 NRC Form~1,-~,a.§J~~nl~~1:>utline Rev O.xls

ES-401 4 Form ES-401*1 1ES-401 8WR Examination Outline Form ES-401-1i Plant Systems - Tier 2/Group 1 (RO)

System # / Name KKK K t i l l A A A A F~) KIA Topic(s) IR #

1234 12341'~

r---------------------~~~_r-o .... .'

1-:_:_: _:_:_:_:_:_:_~_:_:_lc_lo _nJ_'~_in_:'_n_M_O_d_e_-1-1+~+- ~m° ~ :~:=~~~=~:.,~

3.1 2.5 3.4; 2 . i,.", suppression pool level BWR-2, 3, 4 2 7 3.4

"'i:i~

207000 Isolation (Emergency)

Condenser .... o 209001 LPCS o DC power 2.8 4

209002 HPCS o 211000 SLC o Core plate differential pressure indication; Abllfty to verify 2.6; 2

3 that the alarms are consistent with the plant condftions 4.2 212000 RPS o RPS motor-generator sets 3.2 1 .......

0 2150031RM 24148 volt D.C power: Plant-Specific 3.6 2

215004 Source Range Monitor

., o

-0 lock status: Verification of proper functioning/ I ;.~;

215005 APRM / LPRM 2 8 6 217000 RCIC Reacto< water level 4.0 1 3

218000 ADS o .~ DS logic operation 3.8 1 1

223002 PCIS/Nuclear Steam Supply

. Containment drainage system 2.8 1 IISh~to~

1239002 SRVs 259002 Reactor Water Level Control o

1 A(

o ~t*

2- ;~~

Tail pipe temperatures: Ability to explain and apply system limits and precautions GEMACIFoxborolBailey controller operation: Plant-Specific 3.6; 3.8 3.1 2

1 0

261000 SGTS Fan start 3.2 1 i:. 2 262001 AC Electrical Distribution o Major system loads 3.5 1 1

262002 UPS (AC/DC) o ontainnient isolabon system: Plant-Specific 2.9 1 8

o ..

263000 DC Electrical Distribution Battery charging/discharging rate 2.5 1 1

tl 0 laralle! operation of emergency generator Indicating 3.5; 1264000 EDGs

~:J ~-: 3_ lights meters, and recorders 3.4 2

~--------------------~_r_r_r~~~-+~-+-+

2.9; 300000 Instrument Air o 0 --'-::; ross-oed units, Secunng of lAS upon loss of cooling wate 2 rr-t-+

3 3 2.8

~ooooo Component Cooling Water -+---1f-+-

O

'""""1-+-1-1" . . . . Automatic start of standby pump 3.4 1 II------t-+-+-tf---t-t-1 "

I I 1- [~<:~:**i***,**~- - - - - + - - t - - - -0 i IFKI=A==c=a=m=g=o=~=T=o=m=ls=:============~3~2=*~~ 2 P2~~=G=ro=u=P=p=o=in=t=T=ot=a=I:======================r====11 26

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

I~

K A A A

~I~

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

2 3 4 5 6 1 2 3 1201001 CRD Hydraulic ,i, 0 hH ~~"

~01002 RMCS 0 i201 003 Control Rod and Drive Mechanism 3.2 1 201004 RSCS 0 F'

201005 RCIS 0 201006 RWM ..  : Power supply loss'. P-Spec(Not-BWR6) 2.5 1 202001 Recirculation Knowledge of system purpose and/or function. 3.9 1 202002 Recirculation Flow Control .... 0 0

204000 RWCU Response to system isolations 3.6 1 3

214000 RPIS 0 215001 Traversing In-core Probe 0

215002 RBM

-- 0 0 ......

1216000 Nuclear Soiler Inst 1

Recorders 3.3 1 1219000 RHRlLPCI: Torus/Pool Cooling Mode 0 0

223001 Primary CTMT and Aux.

9 . Drywell cooling fans. Plant-Specffic 2.7 1 226001 RHRlLPCI: CTMT Spray Mode 0 1230000 RHR/LPCI: Torus/Pool Spray Mode 0 Pool Cooling/Cleanup I 234000 Fuel Handling Equipment ..

I.*. .* ~

. . Fuel orientation

~

o 1

rnt 239001 Main and Reheat Steam 0 239003 MSIV Leakage Control 0 241000 ReactorfTurbine Pressure Regulator 0 245000 Main Turbine Gen. / Aux. .'. 0 ill 256000 Reactor Condensate 0 259001 Reactor Feedwater 6 i':.

Recirculation 3.1 1 268000 Radwaste V,* 0 271000 Offgas 0 0 '< IAutomatiC actions to contain the radioactive release in the 272000 Radiation Monnoring levent that the predetermined release rates are exceeded 3.7 1 2

0 ....

286000 Fire Protection AC. electrical distribution. Plant-SpecIfic 3,1 1 1

0 288000 Plant Ventilation Differential pressure control 3.2 1 2

~ ..

290001 Secondary CTMT 0 i.,

290003 Control Room HVAC

~ L< Control room pressure 25 1 290002 Reactor Vessel Internals 0 0

KIA Category Totals: 1 1 1 roup Point Total:

ES-401 Page '~2 Qf 34 2011 NRC Form ES-40r-1, 3" - ~l\Trinen Outline Rev O.xls

ES-401 2 Form ES-401-1 E8-401 BWR Examination Outline Fonn E8-401-1 Emergency and Abnonnal Plant Evolutions - Tier 1/Group 1 (SRO)

ElAPE # 1 Name 1 Safety Function K K K A A KIA Topic(s) IR #

1 2 3 1 2 G.

/295001 Partial or Complete Loss of Forced 0 Power/flow map 3.8 1

~e Flow Circulation 11 & 4 Knowledge of the bases in Technical Specifications for ml

~95003 Partial or Complete Loss of AC / 6 4.2 1

25. limiting conditions for operations and safety limits.

artial or Total Loss of DC Pwr / 6 0 If~,<

~95005 Main Turbine Generator Trip 13 0 295006 SCRAM 11 0

~95016 Control Room Abandonment 17 0

~95a18 Partial or Total Loss of CCW 18 0 ., System pressure 2.9 1 5

295019 Partial or Total Loss of Inst Air f 8 295021 Loss of Shutdown Cooling f 4 295023 Refueling Acc 1 8

()

295024 High Drywell Pressure f 5 8

.; Drywall radiation levels I 4.0 1 1<

~95025 High Reactor Pressure I 3 0 2 , ..., u,~ pg, g" ,~<~. 0 g"y *...,.,~ YO¥Y <Y aoo~o_

~95026 Suppression Pool High Water Temp_ the status of safety functions, such as reactivity control, 5 and heat removal, reactor coolant system I 4.6 1 l

295027 High Containment Temperature I 5 0 of the operational implications of EOP

~95028 High Drywell Temperature I 5 ings, cautions, and notes.

4.3 1 I*c*.

~95a30 Low Suppression Pool Wtr Lvi f 5 a I~:

I****** *.

rl..*.*k<-'

I=AM

~95031 Reactor Low Water Level f 2 ..

a c.""moo Pm",,,, 0 1 Reactor Power Above APRM nscale or Unknown 1 1 4 I'* " Suppression pool temperature 4.1 1295038 High Off-site Release Rate I 9 I:**** 0

~OOOOO Plant Fire On Site 1 8 0 1700000 Generator Voltage and Electric Grid i

a rbances 16 Category Totals: 0 0 0 0 I"" D *. t Total:

2011 NRC Form ES-401-1, 3 - Written Outline Rev O_xls

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

K K K A A E/APE # I Name I Safety Function G KIA 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

04. Knowledge of abnormal condilion procedures 4.2 1 11 295009 Low Reactor Water Levell 2 0 295010 High Drywell Pressure I 5 0 295011 High Containment Temp I 5 0 295012 High Drywell Temperature 15 0 Drywell pressure 4.1 1 2

295013 High Suppression Pool Temp. 15 0 295014 Inadvertent Reactivity Addition 11 0 295015 Incomplete SCRAM 11 0 295017 High Off-site Release Rate I 9 0 295020 Inadvertent Cont. Isolation I 5 & 7 0 295022 Loss of CRD Pumps I 1 0 295029 High Suppression Pool Wtr LviI 5 0 295032 High Secondary Containment Area 0

Temperature I 5 295033 High Secondary Containment Area 0 Equipmenl 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 I 5 295036 Secondary Containment High 0

Sump/Area Water Levell 5 500000 High CTMT Hydrogen Conc. 15 0 KIA Category Totals: 0 0 0 0 2 1 Group Point Total: 3 2011 !\IRC Form If:%.1R)t,,-~,CIJl~#hll~~'butline Rev O.xls

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline K

System # / Name KIA Topic(s) IR #

03000 RHRlLPCI: Injection o 05000 Shutdown Cooling Mode o HPCS 0 SLC 0 RPS 0 NUClear Instrument system failure 3.7 4

IRM 215004 Source Range Monitor 215005 APRM / LPRM 0 217000 RCIC 0 0 Loss of A.C or D C power to ADS valves 3.6 5

IS/Nuclear Steam Supply 0

239002 SRVs 0 259002 Reactor Water Level Control 0 261000 SGTS 4.1 262001 AC Electrical Distribution 4.7 (AC/DC) 0 63000 DC Electrical Distribution 0 264000 EDGs 0 300000 Instrument Air 0 00000 Component Cooling Water HighAow ccw temperature 3.0 0

5

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

System # I Name K K K K K K A A A A 1 2 3 4 5 6 1 2 3 4 G KIA Topic(s) IR #

1201001 CRD Hydraulic 0

~01002 RMCS 0

~01003 Control Rod and Drive Mechanism 0

~01004 RSCS 0 1201005 RCIS 0

~01006 RWM 0 1202001 Recirculation 0 202002 Recirculation Flow Control 0

~04000 RWCU 0

~14000 RPIS 0 1215001 Traversing In-core Probe 0 1215002 RBM 0

~16000 Nuclear Boiler Inst. 0

~19000 RHR/LPCI: Torus/Pool Cooling Mode 0 1223001 Primary CTMT and Aux. 0

~26001 RHR/LPCI: CTMT Spray Mode 01.

Ability to interpret and execute procedure steps 4.6 1 20 1230000 RHR/LPCI: Torus/Pool Spray Mode 0

~33000 Fuel Pool Cooling/Cleanup 0

~34000 Fuel Handling Equipment 0 1239001 Main and Reheat Steam 0 1239003 MSIV Leakage Control 0 1

241000 ReactorfTurbine Pressure Regulator Turbine tnp: Plant-Specific 3.8 1 7

245000 Main Turbine Gen. / Aux. 0 1

256000 Reactor Condensate Feedwater heater string trip: Plant~Specific 2.9 1 7

259001 Reactor Feedwater 0 268000 Radwaste 0 271000 Offgas 0 t>72000 Radiation Monitoring 0

~86000 Fire Protection 0

~88000 Plant Ventilation 0 290001 Secondary CTMT 0 1790003 Control Room HVAC 0

~90002 Reactor Vessel Internals 0 0

KIA Category Totals: 0 0 0 0 0 0 0 2 0 0 1 Group Point Total: 3 2011 NRC Form 'E.%-~ro~,- fCljl~ ~M-iH~H'butline Rev O.xls

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 4.4 of the station's requirements for verbal communications when implementing 3.7 1 3.9 1 identify and interpret diverse indIcations to validate the response of another control room indications to verify the status and operation of a system, ator actions and directives affect plant and system conditions.

cedures for the facility, hcluding operating those controls equIpment that could affect reactivity 4.4 determine the expected plant configuration uSlrg design and confIguration "m;UH1"""~"U"'. such as drawings, line-ups, tag-outs, etc 4,3 I Knlowledlge of radiation exposure limits under normal or Elmergency conditions. 3.2 1 use radiation monitoring systems, such as fixed radiation monitors and alarms.

survey instruments. personnel monitoring eqUipment, etc.

2.9 3,8 3.8 1 IKn,nwlAdrlAof "fire in the plant" procedures.

I Knowlede,e of annunciator alarms, indications, or response procedures. 4.2 1 abnormal indications for for emergency and abnormal operating procedures 4.7 ES-401, Page :27 of 34 2011 NRC Form ES-401-1, 3 - Written Outline Rev O.xls

ES-401 Record of Rejected K/,As Form ES-401-4 Randomly Tier I Selected Reason for Rejection Group KIA RO 1 11 295028 This KIA is essentially the same as the one for Q #83 (295012 AA2.02). Replaced Q #51 EA2.04 with KIA 295028 EA2.01.

Unable to construct a question for this KIA that meets the requirements of NUREG RO 1 11 295019 1021 (there are no EOP warnings, cautions or notes relative to the Instrument Air Q#55 G2.4.50 System). Replaced with KIA 295019 G2.4.50.

RO 1 11 295024 This generic KIA is not a good fit for the High Orywell Pressure EPE. Replaced with Q#56 G2.2.36 KIA 295024 G2.4.8.

RO 1 11 295005 This KIA is for BWR-2 plants and does not apply to Peach Bottom, which is a BWR Q#57 AK2.09 4. Replaced with KIA 295005 AK2.03.

Unable to construct a question for this KIA that meets the requirements of NUREG RO 1 12 295020 1021 (cannot link this KIA to Inadvertent Containment Isolation). Replaced with KIA Q#64 G2.1.7 295020 AA1.02 (Generic KIA's were over-sampled).

RO 1 12 295012 The High Orywell Temperature APE/EPE was over-sampled. Replaced with KIA Q#65 G2.1.23 295014 G2.1.23.

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

SRO 1 11 295031 This KIA is very similar to the one for Q #52 (295031 EA2.04). Replaced with KIA Q#78 EA2.01 295037 EA2.04 (295031 was over-sampled).

Unable to construct an SRO-only question for this KIA that meets the requirements SRO 1 11 295001 of NUREG-1021 (system parameters that are entry level for Tech Specs is RO Q#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 SRO 1 12 295008 of NUREG-1021 and KIA is not tied to 10CFR55.43(b). Replaced with KIA 295008 Q#84 G2.1.30 G2.4.11.

There are no RO tasks performed outside the Main Control Room associated with SRO 1 12 295033 High Secondary Containment Radiation Levels. Replaced with KIA 295033 EA2.02 Q#85 G2.4.34 (Generic KIA's were over-sampled).

Unable to construct an SRO-only question for this KIA that meets the requirements SRO 2 11 300000 of NUREG-1021. In addition, System 300000 was already sampled on the RO Q#86 A2.01 section. Replaced with KIA 218000 A2.05.

Unable to construct an SRO-only question for this KIA that meets the requirements SRO 2 11 400000 of NUREG-1 021. Replaced with KIA 400000 A2.03 (Generic KIA's were over-Q#89 G2.4.45 sampled).

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

Unable to construct an SRO-only question for this KIA that meets the requirements SRO 2 12 226001 of NUREG-1021 (SPOS is not used for "decision-making"). Replaced with KIA Q#93 G2.1.19 226001 G2.1.20.

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

ES-301 Administrative OlDies Outline Form ES-301-1 Facility: Peach Bottom Date of Examination: 01/31/2011 Examination Level: RO [8J SRO D Operating Test Number: 2011 NRC Administrative Topic Type Describe activity to be performed (See Note) Code*

2.1.45 - Manually Calculate Drywell Bulk Average Conduct of Operations D, R Temperature (Alternate Path Failed Temperature Points)

[PLOR241C]

I 2.1.25 - Perform RCS Leakage Test lAW ST-O-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]

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

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when ~) 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 12/8/2010 3:05 PM

ES-301 Administrative Outline Form ES-301-1 Facility: Peach Bottom Date of Examination: 01/31/2011 Examination Level: RO 0 SRO [SI Operating Test Number: 2011 NRC Administrative Topic Type Describe activity to be performed (See Note) Code*

2.1.18 - Identify and Specify Required Notifications per OP-M-106-102 "Accidents Involving the Transportation of Conduct of Operations N,R Rad Materials" Based on Completed Attachment 1 Transportation Accident 1 Incident Form [PLOR351C]

2.1.26 - Determine Electrical Safety Personal Protective Cond uct of Operations D, R Equipment For Racking Out A 4 KV Circuit Breaker

[PLOR263C]

Equipment Control D,R 2.2.23 - T(~ch Spec Action Log Entry [PLOR221 C]

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

[PLOR249C]

2.4.41 - EAL Classification with State and Local Emergency Plan D,R Notification [PLOR232C]

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when ~i 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 (.s 1; randomly selected)

ES 301, Page 22 of 27 Last printed 12/8/20103:05 PM

ES-301 Control Roomlln-Plant ........ "'Tg'rTlIO;:O Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 01/31/2011 Exam Level: RO [8J SRO-I D SRO-U D Operating Test Number: 2011 NRC

~ .... ~ Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Type Safety System I JPM Title Code* Function

a. 201003 A2.03 - Control Rod and Drive Mechanism - Respond to a Rod A, D, 1 Drift (Alternate Path - Second Rod Drifts) [PLOR307CA] [Set 1] EN,S
b. 259001 A4.02 - Reactor Feedwater System - Shutdown the "An RFP A,D,S 2 Turbine (Alternate Path - Min Flow Valve Fails Closid) [PLOR303CA]

[Set 2]

c. 218000 A4.03 - Automatic Depressurization System - ADS Reset D, EN, 3 Foll()wing 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

. ~OR353CA] [Selt4]

e. 219000 A4.13 - RHR/LPCI: Torus Cooling Mode - HPSW Injection into I D, EN, S the Torus [PLOR081Q] [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 Makeup to Tower U§i...g a HPSW Pump [PLOR-270C] [SElt3]

261000 A4.03 - Standby Gas Treatment System Manually Start SBGT D,EN,S 9 System [PLOR044c] [Set 2]

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 1 Scoop Tube Positioner (Alternate Path - Clutch Fails to Engage)

[PLOR346PA] I

j. 241000 A2.01 - Reactor/Turbine Pressure Regulating System - Swapping A,D,E,R 3 EHC System Pressure Regulators - Unit 3 (Alternate Path - Backup PressureBegulator Instabilities) [PLQR334PA~ *
k. 295018 M 1,01 - Component Cooling Water System - Loss of RBCCW D, R I (Plant Actions for the Instrument Nitrogen System) [PLOR096P]

@ 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.

  • Type Codes Criteria for RO I SRO-II SRO-U I

Last printed 12/8/20103:05 PM

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

Type Safety System 1 JPM Title Code* Function

a. 201003 A2.03 Control Rod and Drive Mechanism Respond to a Rod A,D, 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 2 Turbine (Alternate Path Min Flow Valve Fails CIOSE!d) [PLOR303CA]

[Set 2]

c. 218000 A4.03 - Automatic Depressurization System - ADS Reset D, EN, 3 Following BI9wdown [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 D, EN, 5 the Torus [PLOR081C] [Set 1] L, S
f. 262001 A4.01 - AC. Electrical Distribution - Restoration of 4KV Buses A, D,S I 6 from 2SUE (Alternate Path - 2SU-A Breaker Closes) [PLOR344CA]

[Set 4]

g.

h. 261000 A4.03 - Standby Gas Treatment System - Manually Start SBGT N,S 9 System [PLOR044C1 [Set 21 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 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

,----Er~ssure Regulator Instabilities) [PLOR334PA]

n

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

@ 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 I

Last printed 12/8/2010 3 :05 PM

ES-301 Control Room/In-Plant _'TD","" Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 01/31/2011 Exam Level: RO D SRO-I D SRO-U ~ Operating Test Number: 2011 NRC

, ems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Type Safety System I JPM Title Code* Function a.

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

[Set 2]

c.

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

g.

I h.

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, M,R 1 Scoop Tube Positioner (Alternate Path - Clutch Fails to Engage)

[PLOR346PA]

j.

k. 295018 AA1.01 - Component Cooling Water System - Loss of RBCCW 0, R 8 (Plant Actions for the Instrument Nitrogen System) [PLOR096P]

@ 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 I

Last printed 12/812010 3:05 PM

(A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank .:::9 I .::: 8 I .::: 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 I >1 I ~1 (N)ew or (M}odified from bank including 1(A) ~2 I ~2 I >1 (P}revious 2 exams .:::3 I .::: 3 / < 2 (randomly selected)

(R)CA ~1 I > 1 / ~ 1 (S)imulator ES-301, Page 23 of 27 Last printed 12/8/20103:05 PM

Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. 1~1 (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-1 07 "Loss of CRD Regulating Function". After CRD has been restored, a turbine stop valve will fail closed, requiring the crew to execute OT-1 02 "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-1 04 "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, n~quiring a reactor scram and entry into T-1 01 "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 feed pump 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) exclaeds its T-1 03 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 / bypass filter TS CRS and restore CRD pump to service (Tech Spec) 3 See Scenario Guide R lIRO Turbine stop valve fails closed / power reduction CRS 4 See Scenario Guide C PRO Inadvertent HPCI initiation / 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 feedpump 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)eactivity, (I)nstrument, (C)omponent, (M)aJor, (TS) Tech Spec

Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #2 (modified) OpTestNo. 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 pump and restore the system in accordance 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 "Unexplainedl 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 SRVactuation. 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 I~

nction Event Event 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 I shutdown ESW pump (Tech Spec)

TS CRS 4 See Scenario Guide R URO Recirc pump trip / single loop (Tech Spec) I insert GP-9-2 Table 1 rods TS CRS 5 See Scenario Guide S of stator cooling water I scram (electric ATWS) 6 See Scenario Guide I URO RWCU fails to isolate on SBlC initiation I manually isolate RWCU CRS 7 See Scenario Guide C PRO Unable to restore drywell instrument nitrogen I place alternate CRS instrument nitrogen system(s) in service

  • (N)ormal, (R)eactlvlty, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec

Scenario Outline ES-D-1 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 30'1'0 limiter. After these evolutions are complete, a radiation monitor will fail upscale and a drywell 18-inch vent dal'11per 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 oV':lrfiliing 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-8.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 drywell 18-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 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 I

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

Aiai J

5~!/'r:-t iter  ??7

Scenario Outline ES-D-1 Simulation Facility Peach Bottom Scenario No. #5 (modified) OpTestNo. 2011 NRC Examiners Operators CRS (SRO)

URO (ATC)

PRO (SOP)

Scenario The scenario begins with the reactor at approximately 88% power. Following shift turnover, the crew will Summary perform ST-O-001-200-2 "Turbine Stop Valve Closure and EOC-RPT Functional Test". A failure during the test will require the crew to make a Tech Spec declaration. 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 ins.trument 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 Trips/Isolations to Satisfy Tech SpeclTRM Requirements". When this is complete, the 'A' Condensate pump will trip without the expected Recirc System runback. Power must be manually reduced using recirc flow to prevent a low-level scram.

When conditions have stabilized, #2 Auxiliary Bus will trip on overcurrent, causing a loss of the remaining Condensate pumps. An RPS failure will prevent the automatic and manual scrams, requiring entry into T 101 "RPV Control" and the use of Alternate Rod Insertion (ARI) to shutdown the reactor. A small Reactor coolant leak will occur in the drywell and require the use of containment sprays. The crew should enter T 102 "Primary Containment Control". A containmt:!nt spray logic failure will complicate the crew's efforts to spray containment; the other loop of RHR will be available and should be used to spray containment.

HPCI will trip shortly after drywell sprays are in-service and will not be recoverable. The reactor coolant leak inside the drywell will be greater than the capacity of RCIC (the only remaining high-pressure feed source). The crew should enter T-111 "Level Restoration". As level deteriorates, the crew should start available low pressure ECCS pumps and when it is determined that level cannot be restored and maintained above -172 inches, the reactor should be depressurized in accordance with T-112 "Emergency Slowdown". Low pressure ECCS will be available to recover reactor level.

Initial IC-85, 88% power Conditions Turnover See Attached "Shift Turnover" Sheet Event Malfunction Event Event No. No. Type* Description I

1 See Scenario Guide N PRO Main turbine stop valve functional test I RPS and EOC-RPT logic TS CRS failure (Tech Spec) 2 See Scenario Guide C URO Service Water pump trip I manual start of the standby pump CRS 3 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 4 See Scenario Guide R URO Condensate pump trip with recirc runback failure I power reduction CRS 5 See Scenario Guide M ALL Loss of #2 auxiliary bus 1 loss of condensate & feedwater 1 reactor coolant leak inside the drywell 6 See Scenario Guide C URO RPS failure requires ARI to scram the reactor CRS 7 See Scenario Guide I PRO Containment spray logic failure hampers effort to spray the CRS containment, requiring crew to use alternate RHR loop 8 See Scenario Guide C ALL HPCI turbine trip, requiring an emergency blowdown to restore level with low-pressure ECCS

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