ML15079A074

From kanterella
Jump to navigation Jump to search
Initial Exam 2015-301 Draft Administrative Documents
ML15079A074
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 03/20/2015
From:
NRC/RGN-II
To:
Tennessee Valley Authority
Shared Package
ML15079A140 List:
References
50-259/15-301, 50-260/15-301, 50-296/15-301
Download: ML15079A074 (52)
v • d • e


Text

{{#Wiki_filter:ES-201 Examination Outline Quality Checklist Form ES-201-2 Facility: 'B~owN.S 'F~'{ Date of Examination: 

                                                                                 'JANv..~'i ZD\5 Initials Item                                                  Task Description a      b*       c#
1. a. Verify that the outline(s) fit(s) the appropriate model, in accordance with ES-401. dl 1lJll w

R b. Assess whether the outline was systematically and randomly prepared in accordance with I Section 0.1 of ES-401 and whether all KJA categories are appropriately sampled. ettl ?JM_ T T c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics. al1. IJ;l E N d. Assess whether the justifications for deselected or rejected KJA statements are appropriate. E;- v). ~ N-l

2. a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number s

I of normal evolutions, instrument and component failures, technical specifications, and major transients. \ 

                                                                                                                   \

M b. Assess whether there are enough scenario sets (and spares) to test the projected number u and mix of applicants in accordance with the expected crew composition and rotation schedule L without compromising exam integrity, and ensure that each applicant can be tested using A at least one new or significantly modified scenario, that no scenarios are duplicated T from the applicants' audit test(s). and that scenarios will not be repeated on subsequent days. 0

c. To the extent possible, assess whether the outline(s) conforrn(s) with the qualitative R

and quantitative criteria specified on Form ES-301-4 and described in Appendix 0. \

3. a. Verify that the systems walk-through outline meets the criteria specified on Form ES-301-2: \

vA (1) the outline(s) contain(s) the required number of control room and in-plant tasks w distributed among the safety functions as specified on the form N-\ I (2) task repetition from the last two NRC examinations is within the limits specified on the form T (3) no tasks are duplicated from the applicants' audit test(s) (4) the number of new or modified tasks meets or exceeds the minimums specified on the form (5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria on the form. \ 

                                                                                                                                 \
b. Verify that the administrative outline meets the criteria specified on Form ES-301-1:

(1) the tasks are distributed among the topics as specified on the form (2) at least one task is new or significantly modified (3) no more than one task is repeated from the last two NRC licensing examinations C. Determine if there are enough different outlines to test the projected number and mix \ of applicants and ensure that no items are duplicated on subsequent days

4. a. Assess whether plant-specific priorities (including PRA and IPE insights) are covered in the appropriate exam sections. ~ ~

G E b. Assess whether the 10 CFR 55.41/43 and 55.45 sampling is appropriate. clJ( "JW.- N E

c. Ensure that KJA importance ratings (except for plant-specific priorities) are at least 2.5. !q,I( ftU-R A
d. Check for duplication and overlap among exam sections. "'- .//,

171 ......__ N-J L e. Check the entire exam for balance of coverage. 1'L

f. Assess whether the exam fits the appropriate job level (RO or SRO).

lfo.A-t_ Printed ~ignature 

                                                                                                                 ""'  f'Date
a. Author Cl'cde , ~ /J ""-/y
b. Facility Reviewer(*) AHA ~ k
c. NRC Chief Examiner(#) ~!ti.A.NO e.A~LLaD I ~:...:. / lf',,,,,//JItJJUJ ~--,:;~ 11-
d. NRC Supervisor l\\AP<' A 'Y..Ar/6/ /rJ..Jf;. 1 ~..b.. ~-~~l'-{

Note: # Independent NRC reviewer initial items in Column "c": chief examiner concurrence required.

  • Not applicable for NRG-prepared examination outlines N-1: ~ feit"m Es-201-z ~ ~ ~ w'{"n-\'~ ~
                    ~~~~ ~ "SFN J~ ~01~ 1:;,.A-h'.J. '£..ica..w.

ES-401 BWR Examination Outline Form ES-401 *1 I - ~ 

                         ~   *ry                                    Date of Exam:       Ta.A~               2.015 RO KIA Category Points                                    SRO-Only Points 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
  • Tot al
1. 1 4 3 4 3 3 3 20 4 3 7 Emergency &

Abnormal 2 1 1 1 N/A 2 0 N/A 2 7 2 1 3 Plant Evolutions Tier Totals 5 4 5 5 3 5 27 6 4 10 1 3 2 3 2 2 2 3 3 1 3 2 26 3 2 5 2. 2 1 1 2 1 1 1 1 1 1 1 1 12 0 2 1 3 Plant Systems 4 3 5 3 3 4 2 3 5 Tier Totals 3 4 4 38 3 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 3 3 2 2 1 2 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" 1JU_
2. f in each KJA category shall not be less than two).

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/evolutions that are not 1)1(.            ncluded on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.

1f<<...4. ~elect topics from as many systems and evolutions as possible; sample every system or evolution in the group before electing a second topic for any system or evolution. 1/tU... e. [$,..AfJsent a plant-specific priority, only those KIAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively. f/I(_ 6. ~8""ct SRO top;ra foe T~ 1 ""' 2 from tho " " " " ' " " " " '"" t<JA ~1'9<><1~ 7.* The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics 1$(_ 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 (I Rs) 1'J{.. <;/..-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.

1JM_ 9. ~r Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRS, nd point totals(#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43. ES-401, Page 16 of 33

ES-401 2 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emeraencv and Abnormal Plant Evolutions - Tier 1/Grouo 1 ~ ~ E/APE #I Name I Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 A.. Cl-'> A-f>ll.01.. 295001 Partial or Complete Loss of Forced Core Flow Circulation J 1 & 4 II.. (i::'> AAl*O'l. 295003 Partial or Comolete Loss of AC 16 

                                              ~                                     ~)A-4..\.o'>

295004 Partial or Total Loss of DC Pwr / 6 

                                                                              ...    ~)    c..r£.<1.z..1 295005 Main Turbine Generator Trio / 3
                                                  ~
                                                                                ~

ll.1 All- ?..O\o c;,) (o i, .~. ?.O 295006 SCRAM / 1 fl 6t.~."\'5 295016 Control Room Abandonment t 7 ll. (jl.'S A\L-1-Q \ 295018 Partial or Total Loss of CCW / 8 

                                                                                       "-'A-4..'\.01-295019 Partial or Total Loss of Inst. Air t 8 295021 Loss of Shutdown Coolina / 4
                                              ~        --                 s         l'.J\.) A~\,O \        O::> AAV:>'L I\.                                ~ l.f..) f.\t.-hO~         (..5.) (.., '2.. .~. l..\

295023 Refuelina Ace I 8 

                                                  ...                               (!\.) £: \L- 't..'ll 3 295024 Hiah Drvwell Pressure/ 5
                                                                              ,.. ey._) '1..t..'\'\

295025 Hiah Reactor Pressure I 3 

                                                      ,,.                           lit-)   ei.-~.o'i        ~)~A?..-03 295026 Suppression Pool High Water Temo. /5                                                                  !

295027 Hiah Containment Temoerature J 5 

                                                                                   ,.p..)  tr'-" ~ .o-s      (SJ e:A- t..-oS 295028 Hiah Drvwell Temoerature / 5
                                                                       ,...        ~) E>A*z.. .O'-\

295030 Low Suonression Pool V\llr Lvl J 5 (te-) £.,\ l.~"l 295031 Reactor Low Water Level/ 2 295037 SCRAM Condition Present 

                                                              -..."'"'             ~f\.)~l.C8               ~)~ft           *v::>!

and Reactor Power Above APRM Downscale or Unknown / 1 ~ 

                                                                                                             ~J '6l.-'-/      .'4,\

295038 Hiah Off-site Release Rate / 9 

                                                  ~
                                                                               .s ~)et.1--01..
                                                                       ......      t-.">AA?..O'-

600000 Plant Fire On Site I 8 

                                                      ,...                         ll-=> M4.o\

7~0000 Generator Voltage and Electric Grid Disturbances I 6 KIA Cateaorv Totals: 4 3 4 3 3 3 Groun Point Total: 2017 ES-401, Page 17 of 33

ES-401, REV 9 T1G1 BWR EXAMINATION OUTLINE FORM ES-401*1 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC: RO SRO 295001AA2.02 Partial or Complete Loss of Forced 3.1 3.2 DD DD DD D ~DD D Neutron monitoring ................................... . Core Flow Circulation / 1 & 4 295003AA1.02 Partial or Complete Loss of AC/ 6 4.2 4.3 DDDD~DDDD Emergency generators................................. . 295005G2.4.21 Main Turbine Ge1nercitor Trip 3 4.6 DD DD DD D DD~ Knowledge parameters and logic used to assess the status of safety functions 295006AK2.06 SCRAM / 1 4.2 4.3 o ~Do DDDDDD power. ....................................... . 29501BAK1.o1 Partial or Total Loss of CCW / 8 3.s 3.6 ~ D D D D D D D D D D Effects on componenVsystem operations ............... . 295019AK3.02 Partial or Total Loss of Inst. Air/ 8 3.5 3.4 D D ~ D D D D D D D D Standby air compressor operation ..................... . Loss of Shutdown Cooling I 4 heat. .......................................... . 295024EK2.03 High Drywall Pressure I 5 295028EK3.04 Suppression Pool High Water Temp./ 3.7 4.1 D D ~ D D D D 0 0 D D SBLC injection ...................................... .. 5

~---*-------

295028EK3.05 High Drywall Temperature I 5 3.6 3.7 D D ~ D D D D D D D D Reactor SCRAM ....................................... .. 295030EA2.04 Low Suppression Pool Wtr Lvl / 5 3.s 3.7 D D D D D D D ~ D D D Drywall/ suppression chamber differential pressure; Mark-1&11 ............................................ . Page 1of2 05/06/2014 12:44 PM

ES-401, REV 9 T1G1 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 295031EA1.04 Reactor Low Water Level/ 2 4.3 4.2 D D D D D D ~ D 0 0 0 High pressure core spray: Plant-Specific............ . 295037EA1.08 SCRAM Condition Present and Power 3.6 3.6 O O O 0 0 D ~ 0 0 0 0 Rod control and information system: Plant-Specific... Above APRM Downscale or Unknown 

             /1 600000AA2.02  Plant Fire On Site 18                2.8   2.9  D O O 0 0 D 0           ~ 0 0 D  Damper position 700000AK3.01  Generator Voltage and Electric Grid  3.9   4.2  D D    ~  0 0 0 D D            0 Reactor and Turbine trip criteria Distrurbancecs 295004AK1 .os Partial or Total Loss of DC Pwr I 6  3.3   3.4  ~   D D 0 0 D D D D 0 0          Loss of breaker protection ........................... .

295016G2.4.45 Control Room Abandonment/ 7 4.1 4.3 O D D D 0 0 0 0 D 0 ~ Ability to prioritize and interpret the significance of each annunciator or alarm. 295023AK1 .03 Refueling Ace Cooling Mode I 8 3.7 4.o ~ D D D 0 0 0 D 0 D 0 Inadvertent criticality..............................* 295025G2.2.44 High Reactor Pressure/ 3 4.2 4.4 D 0 D D 0 0 0 D D 0 ~ 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 295038EK2.02 High Off-site Release Rate I 9 3.6 3.8 0 ~ O 0 D D D D D 0 D Offgas system ..................... '"****************** Page 2 of 2 05/06/2014 12:44 PM

Es-401, REV 9 SRO T1G1 BWR EXAMINATION OUTLINE FORM ES-401*1 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 295006G2.4.30 SCRAM / 1 2.1 4.1 D D D D D D D D D D ~ Knowledge of events related to system operations/status that must be reported to internal orginizations or outside agencies. 2eso21AA2.02 Loss of Shutdown Cooling/ 4 3.4 3.4 D D D D D D D ~ D D D AHR/shutdown cooling system flow .................*.* 

                                                                        =-=**-=--= *---**----**          - - * **-***

295023G2.4.21 Refueling Ace Cooling Mode I 8 4.0 O O ~ Knowledge of the parameters and logic used to assess the status of safety functions 295026EA2.03 Suppression Pool High Water Temp./ 3.9 DDDDDDD~DDD Reactor pressure...................................... 5 295028EA2.05 High Drywall Temperature I 5 3.6 3.a D D D D D D D ~ D D D Torus/suppression chamber pressure: Plant-Specific ... 295037EA2.03 SCRAM Condition Present and Power DDDDDDD~DDD SBLC tank level.. .................................... . Above APRM Downscale or Unknown 

              /1 295038G2.4.41  High Off-site Release Rate/ 9       2.9   4.6 D D D D D D D D D D           ~      Knowledge of the emergency action level thresholds and classifications.

Page 1of1 05106/2014 12:44 PM

ES-401 3 Form ES-401-1 I ES-401 BWR Examination Outline ~~ Form Es-401-1 EmernenC' and Abnormal Plant Evolutions - Tier 1/Grouo 2 RO RO E/APE #I Name I Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 If.. {WI.'\ (o'l,."\.\\ 295002 Loss of Main Condenser Vac I 3 295007 Hi!lh Reactor Pressure I 3 295008 High Reactor Water Level I 2 

                                                           ~            !(.A.') .i+A-\. v"'

295009 Low Reactor Water Level I 2 29501 OHigh Drywell Pressure I 5 295011 Hi!lh Containment Temp/ 5 295012 Hiah Drvwell Temperature/ 5 295013 Hioh Suppression Pool Temo. / 5 295014 Inadvertent Reactivity Addition/ 1 

                                                                         .._fP")C\ftl*-C\

295015 Incomplete SCRAM / 1 s {I.. .J..) PtV-l,O"L 295017 Hiah Off-site Release Rate/ 9 It CJ-) P.'-~*' t 295020 Inadvertent Cont Isolation I 5 & 7 295022 Loss of CRD Pumos I 1 

                                                                  ,...   (!'-") C..'l..l."=f-295029 Hiah Suooression Pool I/I/tr Lvl / 5

(.)') eJI\ 7-(n 295032 High Secondary Containment Area Temperature/ 5 s I'.. (_fl...) £\LZ.o\ 295033 High Secondary Containment Area Radiation Levels I 9 295034 Secondary Containment 

                                                                         ~S) ~~            'L'i5 Ventilation Hiah Radiation I 9                                     s IL             1$.)ic:A-l.o\

295035 Secondary Containment High Differential Pressure I 5 295036 Secondary Containment High Sumo/Area Water Level / 5 500000 Hiah CTMT Hydroaen Cone. / 5 KIA Cateaorv Point Totals: I I

l. l int Total: 7/3 ES-401, Page 18 of 33

Es-401, REV 9 

                                                            ~

T1G2 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 295002G2.4.11 Loss of Main Condenser Vac I 3 4.o 4.2 DDDDDDDDDD~ Knowledge of abnonnal condition procedures. 295009AA 1.04 Low Reactor Water Level I 2 2.1 2.1 DDDDDD~ DDDD Reactor water cleanup ................................. 295017AK1.02 High Off-site Release Rate I 9 3.8 4.3 ~ D 295020AK3.01 Inadvertent Cont. Isolation I 5 & 7 3.8 3.8 DDDDD Reactor SCRAM .........*.....*........................... 295029G2.1. 7 High Suppression Pool Wtr Lvl I 5 4.4 4.7 DDDDDDDD~ Ability to evaluate plant pertormance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation. 295033EK2.01 High Secondary Containment Area 3.8 4.0 D~DDDDDDDDD Area radiation monitoring system .............*........ Radiation Levels I 9 295035EA1.01 Secondary Containment Differential Pressure I 5 3.6 3.6 DDDDDD~D DDD Secondary containment ventilation system ............. . Page 1of1 05106/2014 12:44 PM

ES-401, REV 9 SRO T1G2 BWR EXAMINATION OUTLINE FORM ES-401*1 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 295015AA2.01 Incomplete SCRAM/ 1 4.1 4.3 0 0 0 0 0 0 0 ~ 0 0 0 Reactor power......................................... 295032EA2.02 High Secondary Containment Area 3.3 3.s O 0 D 0 D 0 0 ~ 0 0 0 Equipment operability............................... .. Temperature 15 295034G2.4.45 Secondary Containment Ventilation 4.1 4.3 D O 0 0 0 0 0 Ability to prioritize and interpret the significance of each High Radiation / 9 annunciator or alarm. Page 1of1 05106/2014 12:44 PM

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outl~~ Plant S stems

  • Tier 2/Group 1 RO RO System # I 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 203000 RHR/LPCI: Injection 

                                                                                 ,. l-<t-)        &-z..i.. ~~        (_;) <Oi..z..;<\

Mode s 

                                                       !~                              oc-.) \lJ., .1)'\

205000 Shutdown Coolina k ~) \(...2..o"\ 206000 HPCI 207000 Isolation (Emergency) Condenser i.. 

                                                                                     ~~~'t~~

fl. 209001 LPCS 209002 HPCS {L (jl) At...l)"i" 211000SLC ill. ~JAt..z.t 212000 RPS 

                                           ,._                                        ~ > \1.\.01-215003 IRM 215004 Source RanQe Monitor It..
                                                  ,.       ,_                     ,S t tt.) "'Lo\ .o'\

(J."l "l*O'i" fCl')u C", 0 '?.,. (.<;>) (p z.. '-4. '-\:t-215005 APRM I LPRM tc.. s jt.) l'-tl,O \ (_~') f\'2..()'-\ IL ~-1,.) \(.\.\:>'{ 217000 RCIC lt.. LI.') y...<; .o \ (.<;.) f\7....0L 218000ADS  !:lo f.. UI..) '-Z..\.";f tS) A 2...o"\ 223002 PCIS/Nuclear Steam Supplv Shutoff 5 if. ~ U'" ~-s.ol. 239002 SRVs I ti,.") \C.'?>>.O'!. 259002 Reactor Water Level 

                                     ,....                        ...                 l~)      A1..C"\

Control (JO t'--!..~~ Al.oL 261000 SGTS 262001 AC Electrical 

                                  ~                        "                         (.!I..'>

(J\.) \£-l.. .() \ Distribution 

                                                                           'I..       fJl1At.t.ol 262002 UPS IAC/DC)
                             ~                                                       (.JL) ~..l.O\

263000 DC Electrical Distribution 

                                                           -....                        ~"\   A\.o\

264000EDGs 

                                                                           ,...      lJL")AA-~1..

300000 Instrument Air 400000 Component Cooling 

                                     ,,_                                             ';j-) \L.1. :Q \

Water KIA Cateaorv Point Totals: 3 2. ~ z.. z. 1.. !1 ~ \ '!> 'L Group Point Total: 26/5 ES-401, Page 19 of 33 

                                                  <.iQ)

Es-401, REV 9 T2G1 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 203000G2.2.38 RHR/LPCI: Injection Mode 3.6 4.s D D D D D D D D D D ~ Knowledge of conditions and limitations in the facility license. 205000K6.04 Shutdown Cooling 3.6 3.6 D D D D D ~ D D D D D Reactor water level 2.5 Turbine control circuits: BWR-2,3,4 209001M.03 LPCS 3.7 3.6 DD DD Injection valves 00000~0 power 211000A2.os SLC 3.1 3.4 oo DDDD~DD Loss of SBLC tank heaters 212000A2.21 RPS 3.6 3.9 D D D D D D D ~ D D D Failure of individual relays to reposition: Plant- Specific 215003K1.02 IRM 3.6 3.6 ~ D D D D D D D D D D Reactor manual control 215003K4.04 IRM 2.9 2.9 D D D ~ D D D D D D D Varying system sensitivity levels using range switches 215004A1.05 Source Range Monitor 3.6 3.8 D O D D D D ~ D D D D SCRAM, rod block, and period alarm trip setpoints 215004K5.03 Source Range Monitor 2.8 2.8 D D D D ~ D D D D D D Changing detector position Page 1 of3 05/06/2014 12:44 PM 

                                                         @)

ES-401, REV 9 T2G1 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 215005K4.01 APRM I LPRM 3.7 3.7 D D D ~ D D D D D D D Rod withdrawal blocks 217000K1.04 2.s ~ D D D D D D D D D D Main condenser 

                                                                                                                 -~*----       *-------

218000K5.01 ADS 3.8 3.8 D D D D ~ D D D D D D ADS logic operation 223002G2.1.7 PCIS/Nuclear Steam Supply Shutoff 4.4 4.7 D D D D D D 0 D D 0 ~ Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation. 239002A3.02 SRVs 4.3 4.3 D D O D D D 0 0 ~ 0 0 SAV operation on high reactor pressure 239002K3.03 SRVs 4.3 4.4 D D ~ 0 D D 0 0 0 D D Ability to rapidly depressurize the reactor 259002A2.04 Reactor Water Level Control 3.o 3.1 D D D D D D D ~ D D D RFP runout condition: Plant-Specific 259002K3.03 Reactor Water Level Control 2.1 2.9 D D ~ D D D D D D D D Rod worth minimizer: Plant-Specific 261000A1.02 SGTS 3.1 3.2 D D D D D D ~ D D D D Primary containment pressure 262001K2.01 AC Electrical Distribution 3.3 3.6 D ~ D D D D D D D D D Off-site sources of power 262002A4.01 UPS (AC/DC) 2.8 3.1 D D D D D D D D D ~ D Transfer from alternative source to preferred source Page2of3 05/06/2014 12:44 PM

Es-401, REV 9 T2G1 ~EXAMINATION OUTLINE FORM ES-401-1 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC: RO SRO 263000K1.01 DC Electrical Distribution 3.3 3.s ~ D D D D D D 0 D 0 D A.C. electrical distribution 264000A1.01 3.o ODD D 0 D ~ ODD temperature 300000A4.01 Instrument Air 2. 6 2 *7 D 0 0 0 D 0 00~0 Pressure gauges 400000K3.01 Component Cooling Water 2.9 3.3 O O ~ O O 0 D 0 0 0 0 Loads cooled by CCWS Page 3 of 3 05/06/2014 12:44 PM

Es-401, REV 9 SRO T2G1 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 20SOOOG2.2.39 RHR/LPCI: Injection Mode 3.9 4.5 0 0 0 0 0 0 0 0 0 0 ~ Knowledge of less than one hour technical specification action statements for systems. 215004G2.4.47 Source Range Monitor 4.2 4.2 0 0 0 0 0 0 0 0 0 0 ~ Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. 215005A2.04 APRM I LPRM 3.8 3.9 0 3.5 0~000 Large 223002A2.04 PCIS/Nuclear Steam Supply Shutoff 2.9 3.2 O O 0000~000 Process radiation monitoring system failures Page 1of1 05/0612014 12:44 PM

ES-401 5 Form ES-401-1 ES-401 BWR Examinabon Ou~ Form ES-401-1 Plant Svstems - Tier 2/Grouo 2 0 SRO System #I Name K K K K K K A A A A G KIA Topic(s) JR # 1 2 3 4 5 6 1 2 3 4 201001 CRD Hvdrauhc  :. L~) A-t.og 201002 RMCS 201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS II. (It.) Ai...05 201006 RWM 202001 Recirculation rl b'-'\ A \.CT.i-202002 Recirculation Flow Control 204000RWCU 214000 RPIS 215001 Traversina In-core Probe I'- '~)\(..\.01.. ()>.JG>?.-* L'O 215002 RBM 216000 Nuclear Bailer Inst. 219000 RHR/LPCI: Torus/Pool Cooling Mode 223001 Primarv CTMT and Aux 

                                                             ...                             lfL'\t4>.\0
                                               ~                                             (ft.') V..\ .01..

226001 RHR/LPCI: CTMT Soray Mode j\.. ~) l(...t..01... 230000 RHR/LPCJ: Torus/Pool Spray Mode q<.>A"l*~"> 233000 Fuel Pool Coolina/Cleanuo 234000 Fuel Handling EQuipment 239001 Main and Reheat Steam 

                                                   ,_                                        UC..'\ \<-L\     .o 'S'"

239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure Reaulator 245000 Main Turbine Gen. I Aux 

                                                                                      *fl',. U"'\ "t..1.."\

256000 Reactor Condensate 259001 Reactor Feedwater 268000 Radwaste 271000 Offaas 14'- l]L'> \(..~.'11-272000 Radiation Monitoring s l ~1 A:?AYL. 286000 Fire Protection 288000 Plant Ventilation 290001 Secondarv CTMT L !i(.:1 lo\'\ .o-z.. 290003 Control Room HVAC I- [I.) I(.<".O~ 290002 Reactor Vessel Internals

l. l I Group Point Total: 12/3 KIA Cateaorv Point Totals: l l l \ I 2...

ES-401, Page 20 of 33

ES-401, REV 9 T2G2 ~EXAMINATION OUTLINE FORM ES-401*1 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC: RO SRO 20100BA2.05 RWM 3.1 3.s O 0 O 0 0 0 D ~ 0 0 0 Out of sequence rod movement; P-Spec(Not-BWR6) 202002A1.07 Recirculation Flow Control 3.1 3.1 0 0 0 0 0 0 ~ 0 0 0 0 Recirculation loop flow: Plant-Specific 215001K1.02 Traversing In-core Probe 2.5 3.1 ~ 0 DODOOOOO Process computer: (Not-BWR1) 223001K6.10 Primary CTMT and Aux. 3.0 3.2 0 0 0 0 0 ~ DODD Containment vacuum relief system: Mark-Ill 226001K3.02 RHR/LPCI: CTMT Spray Mode 3.5 3.5 O O ~ O O D 0 ODO ContainmenVdrywell/suppression chamber temperature 230000K2.02 RHR/LPCI: Torus/Pool Spray Mode 2.s 2.9 O ~ O 0 0 0 0 0 0 0 D Pumps 233000A4.05 Fuel Pool Cooling/Cleanup 2.7 3.1 0 0 0 0 0 0 0 0 0 ~ 0 Pool temperature 239001K4.05 Main and Reheat Steam 3.1 3.2 0 0 0 ~ 0 0 0 0 0 0 0 Steam flow measurement 25600002.2.4 Reactor Condensate 3.6 3.6 O O O 0 D 0 0 0 0 0 ~ (multi-unit) Ability to explain the variations in control board layouts, systems, instrumentation and procedural actions between units at a facility. 271000K3.02 Offgas 3.3 3.9 0 0 ~ 0 DODD Off-site radioactive release rate 290002K5.03 Reactor Vessel Internals 2.1 3.o O O O 0 ~ 0 0 0 0 0 D Burnable poisons Page 1of2 05/0612014 12:44 PM

Es-401, REV 9 T2G2 ~EXAMINATION OUTLINE FORM Es-401-1 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 290003A3.02 Control Room HVAC 3.o 3.4 D D D D D D D D ~ D D Initiation/failure of fire protection system Page2 of 2 05/06/2014 12:44 PM

Es-401, REV 9 SRO T2G2 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 K5 K6 A 1 A2 A3 A4 G TOPIC: RO SRO 201001A2.08 CAD Hydraulic 2.8 2.8 D D O 0 0 0 0 ~ 0 0 0 Inadequate system flow 215001 G2.1.20 Traversing In-core Probe 4.6 4.6 DDDDD DD DDD ~ Ability to execute procedure steps. 272000A2.02 Radiation Monitoring 3.3 3.6 DDDDD DD ~ D DD Reactor protection system power failure Page 1of1 05/06/2014 12:44 PM

I ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 II .. ""rowns Ferry Date of Exam: Category KIA# Topic RO SRO-Onlv IR # IR # 2.1. 19 Ability to use plant computers to evaluate sys or comp status 3.9 1 2.1. 6 Ability to manage the control crew during plant transients 3.8 1 1. Conduct 2.1. of Operations 3.5 1 2.1. 34 Knowledge of primary and secondary plant chemistry limits 2.1. 7 Ability to evaluate plant perf and make operational judgments 4.7 1 2.1. Subtotal 2 2 2.2. 17 Knowledge of the process for managing maintenance activities 2.6 1 2.2.2 Ability to manipulate the console controls as required 4.6 1

2. 2.2.40 Ability to apply Technical Specifications for a system 3.4 1 Equipment Control 2.2.

2.2. 19 Knowledge of maintenance work order requirements 3.4 1 2.2. 21 Knowledge of pre- and post-maintenance operability requirements 4.1 1 Subtotal 3 2 2.3. ll Ability to control radiation releases 3.8 1 2.3. 15 Knowledge of radiation monitoring systems 2.9 1

3. 2.3. 7 Ability to comply with RWP requirements 3.5 1 Radiation Control 2.3.

2.3. 14 Knowledge of radiation or contamination hazards 3.8 1 2.3. Subtotal 3 1 2.4. 29 Knowledge of the emergency plan 3.1 1 2.4. 5 Knowledge of the organization of the operating proc network 3.7 1 4. Emergency 2.4. Procedures I Knowledge of abnormal condition procedures 4.2 2.4. 11 1 Plan 2.4.29 Knowledge of the emergency plan 44 1 2.4. Subtotal 2 2 Tier 3 Point Total ((a} 7) ES-401, Page 26 of 33 

                                                                -~

Es-401, REV 9 T3 BWif'EXAMINATION OUTLINE FORM ES-401-1 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO G2.1.19 Conduct of operations 3.9 3.8 0 0 0 0 0 0 0 0 0 0 ~ Ability to use plant computer to evaluate system or component status. G2.1.6 Conduct of operations 3.8 4.8 0 Ability to manage the control room crew during plant transients. G2.2.17 Equipment Control 2.6 3.8 O DODOO~ Knowledge of the process for managing maintenance activities during power operations. G2.2.2 Equipment Control 00000000~ Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. - * - - - - - - ---------** - - - * - - - -c----*--::=:-:::=- ---- --------- - - - - - - -

  • G2.2.40 Equipment Control 3.4 4.7 O O O O O O O O O ~ Ability to apply technical specifications for a system.

G2.3.11 Radiation Control Ability to control radiation releases. G2.3.15 Radiation Control Knowledge of radiation monitoring systems G2.3.7 Radiation Control 3.5 3.6 Ability to comply with radiation work permit requirements during normal or abnormal conditions G2.4.29 Emergency Procedures/Plans 3.1 4.4 0 0 0 0 0 0 0 0 0 0 ~ Knowledge of the emergency plan. 

                         ~-*~~-- ---*-----..-         - - - - - - - ---*---- -----*      *--*--- ,, _________ ------*        *---- *---~---- --~-

G2.4.5 Emergency Procedures/Plans 3.7 4.3 O O O O O O O O O ~ Knowledge of the organization of the operating procedures network for normal, abnormal and emergency evolutions. Page 1of1 05/06/2014 12:44 PM

ES-401, REV 9 SRO T3 BWR EXAMINATION OUTLINE FORM ES-401-1 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO G2.1.34 Conduct of operations 2.1 3.s DDDDDDDDDD ~ Knowledge of primary and secondary chemistry limits Conduct 4.4 4.7 oo Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation. G2.2.19 Equipment 2.3 DO~ Knowledge of maintenance work order requirements. DDDDDDDD~ post-maintenance operability G2.3.14 Radiation Control 3.4 3.s OOOD0 DD0 DD ~ Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities G2.4.11 Emergency Procedures/Plans 4.0 4.2 0 ODDDODOD~ Knowledge of abnormal condition procedures. G2.4.29 Emergency Procedures/Plans 3*1 4.4 0 0 0 0 0 Knowledge of the emergency plan. Page 1of1 05/06/2014 12:44 PM

ILT 1501 Administrative Topics Outline Form ES-301-1 Facility: Browns Ferry NPP Date of Examination: 01/19/2015 Examination Level: RO Operating Test Number: 1501 Type Administrative Topic Code Describe activity to be performed (see Note) Conduct of N 2.1.25 2/3-EOI Appendix-9 Primary Containment Water Level Operations Monitoring C00-1 Conduct of N 2.1.20 Reactor Recirc Pump Start Limitations Operations C00-2 Equipment Control N 2.2.12: 1-Sl-4.7.A.2.A Complete Primary Containment Nitrogen Consumption and Leakage Surveillance, evaluate Acceptance EC-1 Criteria 2.3.7: Review of radiological survey map to determine if a task Radiation Control D can be completed without exceeding exposure limits. RC-1 

 .,.'-""I.._. All items (5 total) are required for SROs. RO applicants require only 4 items unless they
              '-'nn "'"'- .. .__ -dministrative topics, when all 5 are required.
  • Type Codes & Criteria: (C)ontrol Room (D)irect from bank (~ 3 for ROs; ~ 4 for SROs and RO retakes)

(N)ew or (M)odified from bank ~ 1) (P)revious 2 exams (~ 1; randomly selected) (S)imulator

ILT 1501 Administrative Topics Outline Form ES-301-1 Reactor Operator

1. 2/3-EOI Appendix-9 Primary Containment Water Level Monitoring (Unit 2 or 3)
  • New
  • 2/3-EOI Appendix-9 Primary Containment Water Level Monitoring and Equipment Control
  • With Primary Containment Flooding in progress, determine Primary Containment water level in accordance with 2/3-EOI Appendix-9 attachment 2
  • 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

Importance RO 3.9

2. Determine if a Reactor Recirc Pump can be started (Unit 2 or 3)
  • New
  • 2/3-SR-3.4.9.3&4, Reactor Recirculation Pump Start Limitations
  • Perform 2/3-SR-3.4.9.3&4 and determine that one Recirc Pump can be started
  • 2.1.20 Ability to interpret and execute procedure steps RO 4.6
3. Complete Primary Containment Nitrogen Consumption and Leakage Surveillance and evaluate Acceptance Criteria (Unit 1)
  • New
  • 1-Sl-4.7.A.2.A, Primary Containment Nitrogen Consumption and Leakage
  • Completes Surveillance and determines that it does meet acceptance criteria.
  • 2.2.12 Knowledge of surveillance procedures. RO 3.7
4. Review of radiological survey map to determine if a task can be completed without exceeding exposure limits
  • Direct from Bank
  • Calculates whole body dose to complete assigned tasks and determines that both the TVA annual limit and the dose margin will be exceeded.
  • 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions. Importance RO 3.5

ILT 1501 Administrative Topics Outline Form ES-301-1 Facility: BFN Date of Examination: 01/19/2015 Examination Level: SRO Operating Test Number: 1501 Type Administrative Topic Code Describe activity to be performed (see Note) Conduct of 2.1.25 2/3-EOI Appendix-9 Primary Containment Water Level Operations N Monitoring C00-1 Conduct of N Operations 2.1.20 Reactor Recirc Pump Start Limitations C00-2 Equipment Control p 2.2.23: LCO Tracking Log entry for RWCU PCIS Valves failed EC-2 Radiation Control 2.3.7: Review of radiological survey map to determine if a task D RC-1 can be completed without exceeding exposure limits. Emergency Plan 2.4.41 Knowledge of the emergency action level thresholds and D EP-1 classifications. NOTE: All items (5 total) are required for SROs. RO applican ......... 

                                                                          - -***,1             *-,1 are retaking only the administrative topics, when all 5 are required.
  • Type Codes & Criteria: (C)ontrol Room (D)irect from bank(.:: 3 for ROs; ~ 4 for SROs and RO retakes)

(N)ew or (M)odified from bank (:: 1) (P)revious 2 exams (~ 1; randomly selected) (S)imulator

ILT 1501 Administrative Topics Outline Form ES-301-1 Senior Reactor Operator

1. 2/3-EOI Appendix-9 Primary Containment Water Level Monitoring (Unit 2 or 3)
  • New
  • 2/3-EOI Appendix-9 Primary Containment Water Level Monitoring and Equipment Control
  • With Primary Containment Flooding in progress, determine Primary Containment water level in accordance with 2/3-EOI Appendix-9 attachment 2
  • 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

Importance RO 3.9

2. Determine if a Reactor Recirc Pump can be started (Unit 2 or 3)
  • New
  • 2/3-SR-3.4.9.3&4, Reactor Recirculation Pump Start Limitations
  • Perform 2/3-SR-3.4.9.3&4 and determine that the A Recirc Pump can be started
  • 2.1.20 Ability to interpret and execute procedure steps RO 4.6
3. LCO Tracking Log entry for failed PCIS Valve
  • Previous
  • OPDP-8
  • Complete an LCO Tracking Log Entry from the results of 1/2/3-SR-3.6.1.3.5(RWCU)

RWCU Primary Containment Isolation Valve Operability

  • 2.2.23 Ability to track Technical Specifications limiting conditions for Operations.

Importance SRO 4.6

4. Review of radiological survey map to determine if a task can be completed without exceeding exposure limits
  • Direct from Bank
  • Review of radiological survey map to determine if a task can be completed without exceeding exposure limits.
  • 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions. SRO 3.6
5. Classify the event per REP (Uncontrolled water level decrease in SFSP)
  • Direct From Bank
  • The event is classified as an ALERT based on uncontrolled water level decrease in spent fuel pool with irradiated fuel assemblies expected to result in irradiated fuel assemblies being uncovered.
  • 2.4.41 Knowledge of the emergency action level thresholds and classifications.

Importance SRO 4.6

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Browns Ferry NPP Date of Examination: 1/19/2015 Exam Level: RO Operating Test No.: 1501 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) Safety System I JPM Title Type Code* Function

a. Control Room Transfer of Recirc Pump Bd 2A; Respond to A, N,S 1 Recirculation Loop A Speed Control Failure 2-AOl-68-3
b. HPCI started in Level Control IAW 2/3-EOI Appendix-SD/

A, N,S 2 Respond to HPCI Turbine Exh Rupture Disc Pressure High

c. Place SOC in service IAW 2/3-EOI Appendix 170 N, L, S 4
d. RPS MSIV 2/3-SR-3.3.1.1.8(5) w/partial closure of MSIV at A, N,S 7 power (2/3-AOl-1-3)
e. 2/3-EOI Appendix-13, Emergency Vent A,D,EN,S 5
f. Restore Offsite Power to 4KV Shutdown Bd 0/3-01-82 D,S 6
g. Restore Fuel Pool Level IAW Fuel Pool Failure 2/3-AOl-78-1 D,S 9
h. 2/3-EOI Appendix-BG Crosstie CAD to Drywell Control Air D,S 3 (RO Only)

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

i. Vent Scram Air Header D,R,E 1
j. Stuck open SRV A,D,R,E 3
k. Reset the Unit 1 Excess Flow Check Valve N,R 8
@           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-I / SRO-U (A)lternate path                                                                        4-6/2-3 (C)ontrol room (O)irect from bank                                                                    ~9/~8/S4 (E)mergency or abnormal in-plant                                                      ~ 1/~ 1/~1 (EN)gineered safety feature                                               - / - /~1 (control room system)

(L)ow-Power I Shutdown ~1/~1/~1 (N)ew or (M)odified from bank including 1(A) ~ 21~ 2/~1 (P)revious 2 exams S 3/S 3/S 2 (randomly selected) (R)CA ~1/~1/~1 (S)imulator

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Browns Ferry NPP Date of Examination: 1/19/2015 Exam Level: SRO-I Operating Test No.: 1501 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) Safety System I JPM Title Type Code* Function

a. Control Room Transfer of Recirc Pump Bd 2A; Respond to A, N,S 1 Recirculation Loop A Speed Control Failure 2-AOl-68-3
b. HPCI started in Level Control IAW 2/3-EOI Appendix-50/

A, N,S 2 Respond to HPCI Turbine Exh Rupture Disc Pressure High

c. Place SOC in service IAW 2/3-EOI Appendix 170 N, L, S 4
d. RPS MSIV 2/3-SR-3.3.1.1.8(5) w/partial closure of MSIV at A,N,S 7 power (2/3-AOl-1-3)
e. 2/3-EOI Appendix-13, Emergency Vent A,D,EN,S 5
f. Restore Offsite Power to 4KV Shutdown Bd 0/3-01-82 D,S 6
g. Restore Fuel Pool Level IAW Fuel Pool Failure 2/3-AOl-78-1 D,S 9 In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
i. Vent Scram Air Header D,R,E 1
j. Stuck open SRV A,D,R,E 3
k. Reset the Unit 1 Excess Flow Check Valve N,R 8
@            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 mav overlap those tested in the control room.
                         *Type Codes                                      Criteria for RO I SRO-I / SRO-U (A)lternate path                                                                         4-6/2-3 (C)ontrol room (D)irect from bank                                                                    5..._9/5..._8/~4 (E)mergency or abnormal in-plant                                                      ~ 1/~ 1/~1 (EN)gineered safety feature                                               - / - /~1 (control room system)

(L)ow-Power I Shutdown ~1/~1/~1 (N)ew or (M)odified from bank including 1(A) ~ 21~ 2/~1 (P)revious 2 exams ~ 31~ 31~ 2 (randomly selected) (R)CA ~1/~1/~1 (S)imulator

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Browns Ferry NPP Date of Examination: 1/19/2015 Exam Level: SRO-U Operating Test No.: 1501 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) Safety System I JPM Title Type Code* Function a.

b. HPCI started in Level Control IAW 2/3-EOI Appendix-50/

A, N,S 2 Respond to HPCI Turbine Exh Rupture Disc Pressure High

c. Place SOC in service IAW 2/3-EOI Appendix 170 N, L,S 4 d.
e. 2/3-EOI Appendix-13, Emergency Vent A,D,EN,S 5 f.

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

j. Stuck open SRV A,D,R,E 3
k. Reset the Unit 1 Excess Flow Check Valve N,R 8
@            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 mav overlao those tested in the control room.
                         *Type Codes                                      Criteria for RO I SRO-I / SRO-U (A)lternate path                                                                                 4-6/2-3 (C)ontrol room (D)irect from bank                                                                            5_.9fs_8/::=_4 (E)mergency or abnormal in-plant                                                              .:: 1/_:: 1/_::1 (EN)gineered safety feature                                                  - / - /_::1 (control room system)

(L)ow-Power I Shutdown _::1/_::1/_::1 (N)ew or (M)odified from bank including 1 (A) .:: 2/_:: 2/_::1 (P)revious 2 exams  : =_ 31::=, 31::=, 2 (randomly selected) (R)CA _::1/_::1/_::1 (S)imulator

ILT 1306 Control Room/In-Plant Systems JPM Narrative Control Room Systems:

a. Control Room Transfer of Recirculation Pump Board 2A; Respond to Recirculation Loop A Speed Control Failure (Unit 2)
  • New I Simulator /Alternate Path
  • O-Ol-57A,Switchyard and 4160V AC Electrical System; 2-AOl-68-3 Recirculation Loop A or B Speed Control Failure
  • 202001 Recirculation System A2.06 Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadvertent recirculation flow decrease IMPORTANCE:

RO 3.6 SRO 3.8 ------

  • Operator directed to perform a control room transfer of Recirc Pump Board 2A from normal to alternate IAW O-Ol-57A. Subsequently, as RR Pump 2A speed recovers from the transfer it will experience a speed control failure, and the operator will respond IAW 2- AOl-68-3.
b. HPCI started in Level Control IAW 2/3-EOI Appendix-50/Respond to HPCI Turbine Exh Rupture Disc Pressure High (Unit 2 or 3)
  • Alternate Path I New I Simulator I Low Power
  • 2/3-EOI Appendix-50 Injection System Lineup HPCI / 2/3-ARP-9-3F
  • 206000 High Pressure Coolant Injection System A2.10 Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: System Isolation. IMPORTANCE: RO 4.0 SRO 4.1
        *The operator starts HPCI for level control IAW 2/3-EOI Appendix-50.

Subsequently, HPCI experiences a failure to isolate on HPCI Turbine Exhaust Rupture Disc High Pressure and must respond IAW 2/3-ARP-9-3F (window 17) and Trip HPCI and manually close the HPCI Stm Line lnbd and outboard lsol Valves.

c. Place RHR in Shutdown Cooling in accordance with EOI Appendix 170 (Unit 2 or 3)
        *New/ Simulator I Low Power
  • 2/3-EOI Appendix-170, RHR System Operation Shutdown Cooling
  • 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode) A4.01 Ability to manually operate and/or monitor in the control room: SDC/RHR Pumps IMPORTANCE: RO 3.7 SRO 3.7
        *Operator places Loop I of RHR in SOC IAW EOI Appendix 170.

ILT 1306 Control Room/In-Plant Systems JPM Narrative

d. 2/3-SR-3.3.1.1.8(5) MSIV Closure- RPS Trip Channel Functional Test/MSIV Partial Closure At Power (Unit 2 or 3)
        *New I Alternate Path I Simulator
  • 2/3-SR-3.3.1.1.8(5) MSIV Closure- RPS Trip Channel Functional Test
  • 2/3-ARP-9-5B; 2/3-ARP-9-4A; 2/3-AOl-1-3, MSIV Closure at Power
  • 212000 Reactor Protection System A2.11 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Main steamline isolation valve closure IMPORTANCE: RO 4.0 SRO 4.1
  • Operator directed to perform control room operations for MSIV Closure- RPS Trip Channel Functional Test IAW 2-SR-3.3.1.1.8(5). Subsequently, the operator will respond to a Partial MSIV Closure at Power IAW with 2/3-AOl-1-3 by reducing power to <66% and placing the associated MSIV in CLOSE..
e. 2/3-EOI Appendix-13, Emergency Vent (Unit 2 or 3)
  • Alternate Path /ENgineered Safety Feature /Direct from Bank /Simulator
  • 2/3-EOI Appendix-13, Emergency Venting Primary Containment
  • 223001 Primary Containment System and Auxiliaries A2. Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.07 High drywell pressure IMPORTANCE RO 4.2 SRO 4.3

  • Operator is directed to emergency vent Primary Containment to restore and maintain Drywell Pressure below 55 psig as directed by 2/3-EOI Appendix-13, Emergency Venting Primary Containment. Emergency Venting of the Suppression Chamber through the Hardened Wetwell Vents will be unsuccessful and the operator will vent the Drywell to Secondary Containment via Primary Containment vent duct failure.
f. Restore Offsite Power to 4KV Shutdown Bd (Unit 2 or 3)
  • Modified I Simulator I Low Power
  • 0/3-01-82, Standby Diesel Generator System
  • 264000 Emergency Generators (Diesel/Jet) A2.01 Ability to (a) predict the impacts of the following on the EMERGENCY GENERATORS (DIESEL/JET); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Parallel operation of emergency generator IMPORTANCE: RO 3.5 SRO 3.6
  • Operator performs operations necessary to restore offsite power to 4kV SD BO A/3EA.

ILT 1306 Control Room/In-Plant Systems JPM Narrative

g. Restore Fuel Pool Level IAW Fuel Pool Failure
  • Direct from bank I Simulator
  • 2/3-AOl-78-1, Fuel Pool Cleanup System Failure
  • 233000 Fuel Pool Cooling and Clean-up A2.02 Ability to (a) predict the impacts of the following on Fuel Pool Cooling and Clean-up; and (b) based on those predications, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low pool level IMPORTANCE: RO 3.1 SRO 3.3
        *Operator commences to restore Fuel Pool level IAW 2/3-AOl-78-1.
h. Cross-Tie CAD to Drywell Control Air (Unit 2 and 3)
  • Direct from Bank I Simulator
  • 2/3-EOI Appendix-BG Crosstie CAD to Drywell Control Air
  • 21BOOO Automatic Depressurization System A2. Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A2.03 Loss of air supply to ADS valves: IMPORTANCE: RO 3.4 SRO 3.6
        *Operator crossties CAD to Drywell Control Air IAW 2/3-EOI Appendix-BG.

In-Plant Systems:

i. Vent the SCRAM pilot Air Header IAW 1/2/3-EOI Appendix-18 (U1/U2/U3)
  • Direct from Bank I Emergency or Abnormal In-Plant I RCA Entry
  • EOI Appendix-1 B, Venting and Repressurizing the SCRAM Pilot Air Header
  • 201003 Control Rod and Drive Mechanism Ability to (a) predict the impacts of the following on the CONTROL ROD AND DRIVE MECHANISM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A2.05 Reactor Scram IMPORTANCE RO 4.1SRO4.1
  • Simulate component manipulations required to vent and subsequently re-pressurize the scram pilot air header as directed by EOI Appendix-1 B.

ILT 1306 Control Room/In-Plant Systems JPM Narrative

j. Stuck Open SRV 1-22 (Unit 3)
  • Direct from Bank I RCA Entry I Emergency or Abnormal In-Plant I Alternate Path
  • 3-AOl-1-1 Relief Valve Stuck Open
  • 239002 Relief I Safety Valves A2.03 Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Stuck open SRV IMPORTANCE: RO 4.1 SRO 4.2
  • Operator attempts to close a stuck open SRV from the Remote Shutdown Panel, when that fails the operator opens the supply breakers or removes fuses to remove power from the SRV. The SRV will close when power is removed.
k. Reset the Unit 1 Excess Flow Check Valve
  • New I RCA Entry
  • 0-01-32, Control Air System
  • 300000 Instrument Air System (IAS) A2. Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: A2.01 Air dryer and filter malfunctions IMPORTANCE RO 2.9 SRO 2.8
  • Simulates field actions to Reset the Unit 1 Excess Flow Check Valve in accordance with 0-01-32.

Appendix D Scenario Outline Form ES-D-1 Facility: BFN Scenario Number: NRC-1 Op-Test Number: 1501 Examiners: Operators: SRO: ATC: BOP: Initial Conditions: Unit 1, 2, and 3 Reactor Power is approximately 100%. 2A CCW pump was removed from service for breaker PMs and is ready to be returned to service. Turnover: Return the 2A CCW pump to service IAW 2-01-27 section 8.3. Southwest Load Dispatch is anticipating a power system alert starting this evening due to a cold front moving through the area. Event Malfunction Event Event Description Number Number Type* N-BOP Return the 2A CCW pump to service IAW 2-01-27 1 N/A N-SRO section 8. 3 (T-Mod-BFN-2-2013-010). I-BOP Respond to an LPRM downscale alarm and bypass 2 nm07 

                               ~ ' the failed detector.

The U 1 US reports that 1-XA-55-22C window 5 Start of Strong Motion Accelerograph is in alarm. M-3 N/A O-AOl-100-5 Earthquake is entered and the SM and precursor U 1 US are evaluating 4.2[1] through [1 O] to determine if shutting down is required The SM directs initiating a Reactor Shutdown IAW R-ATC 2-GOl-100-12A. The Reactor Engineer recommends 4 N/A R-SRO using the urgent load reduction RCP initially to lower core flow to 60%. C-ATC 5 rd01a CRD pump 2A trips respond IAW 2-AOl-85-3 C-SRO rc09 C-BOP 6 RCIC steam leak failure of RCIC to auto isolate rc10 TS-SRO 7 fw19 M-ALL FW A line break-SCRAM 8 override C-ATC 2C RFP discharge valve fails to close 9 N/A C-BOP HPCI is manually secured due to pumping out the feedwater break 

      * (N)ormal,  (R)eactivity,   (l)nstrument,    (C)omponent,   (M)aJor

Appendix D Scenario Outline Form ES-D-1 Events

1. The crew will conduct a pre-job brief for placing the 2A CCW pump in service. The BOP operator will place 2A CCW pump in service per 2 27 section 8.3 which has incorporated T-MOD-BFN-2-2013-010. Once the pump is in service the scenario may continue.
2. The crew will respond to 2-XA-55-5A window 5, LPRM Downscale. The BOP operator will bypass the failed LPRM IAW 2-0l-92B. The SRO will determine that the affected APRM/OPRM and RBM remain operable.

Once the Tech Spec Call has been made and the LPRM bypassed, the scenario may continue.

3. The U1 US reports that 1-XA-55-22C window 5; Start of Strong Motion Accelerograph is in alarm. O-AOl-100-5 Earthquake is entered and the SM and U1 US are evaluating steps 4.2[1] through [1 O] to determine if shutting down is required. The crew will monitor the plant for abnormal conditions and as soon as the lead examiner is ready to move on, the next event can be initiated.
4. The crew will enter 2-GOl-100-12A and the Urgent Load Reduction RCP.

The ATC will lower Reactor Recirc flow to 60% IAW step 1A of the RCP. When the Lead Examiner is satisfied with the Reactivity manipulation the scenario may continue.

5. The 2A CRD pump will trip. The ATC operator will perform the immediate operator actions of 2-AOl-85-3 to place the 1B CRD pump in service. The ATC operator will verify 1B CRD pump is in service and perform the subsequent actions of the AOI. Once the CRD system flows and pressures have been restored to their pre-trip conditions the scenario may continue.
6. RCIC will receive an isolation signal due to a steam leak in the RCIC pump room with a failure to automatically isolate. The US will enter EOl-3 and 2-AOl-64-2C, and the BOP operator will perform actions necessary to isolate RCIC (manually closing 2-FCV-71-2 and 3). The SRO will determine RCIC system inoperable and RCIC isolation valves inoperable. (TS 3.5.3 Condition A, 3.6.1.3 Condition B). Once the steam leak has been terminated (i.e. Steam Valves closed) and the Tech Spec call has been made the scenario may continue.
7. Once the plant is stable, the 'A' Feedwater line will break in the Steam Tunnel. A scram will be inserted due to loss of feedwater and lowering reactor water level. The crew will isolate the feedwater system. The crew will respond IAW EOl-1 and EOl-3. Reactor water level will not be able to

Appendix D Scenario Outline Form ES-0-1 be maintained above (-) 162 inches and the US will enter C-1. When Reactor Water Level lowers to (-) 162 inches the crew will verify Low Pressure injection systems aligned and running then perform an ED and recover Reactor Water Level using Low pressure systems.

8. Following the 'A' Feedwater line break the 2C RFP discharge valve will fail to close. The ATC operator will close the 2C RFP suction valve to complete the Feedwater isolation.
9. The BOP operator will manually trip and lock out HPCI due to pumping out the feedwater line break.

The Scenario ends when the crew has performed an emergency depressurization and re-established Reactor Water Level above the top of active fuel (-) 162 inches.

Appendix D Scenario Outline Form ES-D-1 Critical Tasks 4

1. With a primary system discharging into secondary containment, take action to manually isolate the leak.
1. Safety Significance Isolating high energy sources can preclude failure of secondary containment and subsequent radiation release to the public
2. Cues Procedural compliance Secondary Containment Area Temperature and Radiation indications
3. Measured by Operator action to isolate the leak
4. Feedback Valve position indication and lowering Area Temperature
2. To prevent an uncontrolled RPV depressurization when Reactor Water Level cannot be restored and maintained above -162 inches, inhibit ___
          .,._____" ADS .
1. Safety Significance Maintain adequate core cooling Prevent degradation of fission product barrier
2. Cues Procedural compliance
3. Measured by ADS logic inhibited prior to automatic initiation
4. Feedback RPV Pressure and Water level trend ADS LOGIC BUS AJB INHIBITED annunciators

Appendix D Scenario Outline Form ES-D-1

3. With a(n) injection system(s) operating and lined up for injection, before RPV water level reaches -180", initiate Emergency Depressurization.
1. Safety Significance:

Maintain adequate core cooling Prevent degradation of fission product barrier

2. Cues:

Procedural compliance Reactor Water level trend

3. Measured by:

Observation: US direct Emergency Depressurization before RPV level lowers to -180 inches.

4. Feedback:

Reactor Pressure trend SRV status indications

4. With RPV pressure below the Shutoff Head of the available Low Pressure system(s), operate available Low Pressure system(s) to restore RPV water level above TAF.
1. Safety Significance:

Maintain adequate core cooling Prevent degradation of fission product barrier.

2. Cues:

Procedural compliance Pressure below low pressure ECCS system(s) shutoff head

3. Measured by:

Operator manually starts or initiates at least one low pressure ECCS system and injects into the RPV to restore water level above TAF

4. Feedback:

Reactor pressure trend Reactor water level trend

Appendix D Scenario Outline Form ES-D-1 Facility: BFN Scenario Number: NRC-2 Op-Test Number: 1501 Examiners: Operators: SRO: ATC: BOP: Initial Conditions: Reactor Power is 100%. The Steam Tunnel booster fan is tagged out for lubrication PMs. Suppression Pool Cooling is in service due to a HPCI flow rate test on the previous shift. Turnover: Secure Suppression Pool Cooling. MIG signed on and will be performing 2-SR-3.3.6.1.5(4A/A) Core and Containment Cooling Systems RCIC Turbine Steam Line High Flow Instrument Channel A Calibration. Event Malfunction Event Event Description Number Number Type* N-BOP Secure from Suppression Pool Cooling using 1 N/A N-SRO 2-01-74 MIG reports that 2-RLY-071-13A-K12 did not I-SRO 2 N/A energize when 2-PDT-71-1A was pressurized and TS-SRO that they have stopped at step 7.4[7]8. Override 28 C-BOP The 2A Steam Packing Exhauster will trip and the 28 3 SPE Auto C-SRO Steam Packing Exhauster will not auto start. C-ATC 4 SW10A The 2A Fuel Pool Cooling pump will Trip. C-SRO _CC-ATC) 5 TH12B o / ~C-SRO Recirc Pump 28 vibration high I - R-ATC 6 TH108 C-BOP Recirc Pump 28 seal failure/2-AOl-68-1A TS-SRO TH22 7 RH01A&C M-All LOCA/Scram with inability to spray the Drywell/C4 RH068 SRO directs cool down or rapid depressurization of \' C-BOP I 8 TC02 the RPV using Turbine bypass valves however they C-SRO fail closed and ED will be required. C-80P Core Spray Loop II injection valve will fail to open on 9 CS02A&8 initiation signal but can be manually opened. 

         * (N)ormal,   (R)eactivity,   (l)nstrument,      (C)omponent,     (M)ajor
                                               ) 3,

Appendix D Scenario Outline Form ES-D-1 Events

1. The BOP operator will secure from Suppression Pool Cooling using 2-01-74. When Suppression Pool Cooling is secured the scenario may continue.
2. The Instrument Mechanic Foreman will call the SRO and report that during performance of 2-SR-3.3.6.1.5(4A/A), Core and Containment Cooling Systems RCIC Turbine Steam Line High Flow Instrument Channel A Calibration, 2-RLY-071-13A-K12 did not energize when 2-PDT-71-1A was pressurized and that they have stopped at step 7.4[7]B. The SRO will evaluate Tech Spec 3.3.6.1 and table 3.3.6.1-1 to determine that function 4a is the affected function and that the Tech Spec requires placing the channel in trip within 24 hours. Once the Tech Spec call is completed the scenario may continue.
       ~

(I 3 . The running steam packing exhauster will trip and the standby exhauster will fail to auto start. The BOP operator will place the standby steam packing exhauster in service IAW 2-0l-47C section 6.3. While the BOP operator is adjusting to obtain 10-12 inches of H20 vacuum the scenario may continue.

4. The 2A Fuel Pool Cooling pump will trip. The SRO will direct the ATC to enter 2-AOl-78-1 and the 2B Fuel Pool Cooling pump will be placed in service IAW 2-AOl-78-1 step 4.2[3]. When the 2B Fuel Pool Cooling pump is in service and the lead examiner is ready the scenario may continue.
5. 2-XA-55-4B window 20 RECIRC Pump Motor B Vibration High alarms.

The BOP operator will dispatch an AUO to 2-LPNL-925-0712 and he/she will report that 2-Xl-68-71 D and E are in alarm reading 12.0 mils and rising slowly. The SRO will direct lowering 2B Recirc speed to attempt to clear the alarm. The ATC operator will lower 2B Recirc pump speed. The AUO will report that 2-Xl-68-71 D and E lowered to -10 mils and will reset the alarm locally. When the lead examiner is ready the scenario may continue.

6. 2-XA-55-4B window 20, RECIRC Pump Motor B Vibration High alarms again and 2-XA-55-4B, window 25 Recirc Pump B no. 1Seal Leakage ABN, alarms. The number 2 seal pressure will rise to approximately Reactor Pressure. The SRO will direct tripping the 2B Recirc Pump and entering 2-AOl-68-1A. The ATC operator will lower Reactor Power IAW the RCP and 2-AOl-68-1A. The BOP operator will carry out the subsequent actions of the AOI. The SRO will address Tech Spec 3.4.1.

When conditions have stabilized and the lead examiner is ready the scenario may continue.

Appendix D Scenario Outline Form ES-D-1

7. A leak in the Drywell will develop causing Drywell Temperature and Pressure to rise. The SRO will set a trigger value for a Reactor Scram and when that value is reached a manual Scram will be inserted or the Reactor will Scram at 2.45 psig Drywell Pressure. All control rods will be inserted on the scram. The SRO will direct entry into 2-AOl-100-1. The SRO will direct Suppression Chamber spray per EOl-2 Appendix 17C.

The BOP operator.will attempt to spray the. Suppre§sion Chamber however the Eirst loqQJ!tt~mpted the RHR pumps wilTfrip\ The BOP operator will attempt to sprayJfieSupp.ressiorrChamoefwith the Second 

                                          **************~**--*~

loop of RHR however the:.:§ele.cU09ic. ~ill fail. JMhe SRO directs spraying the Suppression Chamber with Standbycoolant or Fire Protection using RHR loop 1, the breakerfor 2-FCV-074-0100 (480V RMOV BO 1B compartment 19A) or RHR loop II the breaker for 2-FCV-074-0101 (480V RMOV BO 3B compartment 19E) will not close. The SRO/BOP operator will determine that neither the Suppression Chamber nor the Drywell can be sprayed. The SRO may attempt to cool down or anticipate that an ED will be required and attempt to rapidly depressurize the Reactor using the bypass valves however the bypass valves will fail closed. An ED will be required based on Drywell Temperature or the PSP curve. As the Reactor depressurizes the action required area of curve 8 RPV Saturation Temp will be entered and Reactor Water Level indication will be lost. The SRO will direct entry into C-4 and the crew will inject using available systems until the Main Steam Lines are flooded.

8. When the SRO directs a cool down or rapid depression of the RPV using the main turbine bypass valves, the operator will determine and report that the bypass valves have failed closed. This will lead to an ED being required.
9. With an accident signal present the Core Spray loop II injection valve will fail to automatically open, the BOP operator will manually open the injection valve.

The Scenario ends when the crew has performed an emergency depressurization and flooded the RPV to the Main Steam Lines. I l D\

Appendix D Scenario Outline Form ES-D-1 Critical Tasks 2

1. When Suppression Chamber pressure cannot be maintained within the safe area of Curve 6 (PSP) the SRO determines that Emergency Depression is required and is initiated as directed by the SRO.
1. Safety Significance Precludes failure of Primary Containment
2. Cues Procedural compliance High Suppression Chamber or Drywall pressure
3. Measured by Observation-SRO updates or briefs the crew that ED is required based on exceeding PSP curve AND the operator opens 6 ADS/MSRVs
4. Feedback MSRV open indications RPV Pressure lowering OR
1. When Drywell temperature cannot be restored and maintained below 280°F the SRO determines that Emergency Depression is required and is initiated as directed by the SRO.
1. Safety Significance Precludes failure of Primary Containment
2. Cues Procedural compliance High Drywall temperature
3. Measured by Observation-SRO updates or briefs the crew that ED is required based Drywall temperature AND the operator opens 6 ADS/MSRVs
4. Feedback MSRV open indications RPV Pressure lowering

Appendix D Scenario Outline Form ES-D-1

2. With Reactor Water Level unknown, inject into the RPV with available sources until there is indication that the Main Steam Lines are flooded (C-4 Note 7) and maintain them flooded.
1. Safety Significance Prevent fuel damage by establishing adequate core cooling
2. Cues Procedural compliance Loss of all RPV level indications
3. Measured by Observation-Indications that the Main Steam Lines are flooded are listed in C-4 Note 7
4. Feedback MSRV tail pipe temperature MSRV acoustic monitor RPV Pressure trend

Appendix D Scenario Outline Form ES-D-1 Facility: BFN Scenario No.: NRC - 3 Op-Test No.: 1501 Examiners: Operators: SRO: ATC: BOP: Initial Conditions: Reactor Power is 5%. Unit1 and Unit 2 are at 100% power. Turnover: Perform EHC Auto Pump Start Test & Weekly Pump Alternation IAW 3-01-47 A, Sec. 6.2. Continue plant startup IAW 3-GOl-100-1A section 5.4, mode change from Mode 2 to Mode 1. Event Malfunction Event Event Description Number Number Type* N-BOP Perform EHC Auto Pump Start Test & Weekly Pump 1 N/A N-SRO Alternation IAW 3-0l-47A, Sec. 6.2 R-ATC 2 N/A Power increase with Control Rods to 8% IAW GOI R-SRO C-ATC RD07R0239 3 TS- Control Rod Drift in RD06R0239 SRO C-BOP 4 OG04A Loss of SJAE 'A' I Swap to STBY SJAE 'B' C-SRO I-ATC 5 NM05 IRM 'C' Failure Upscale/Half Scram TS,;SBO 

                                 ~AT~-
                                  """----~!

Loss of 4KV Shutdown Board 3ED, 3D DIG fails to 6 DG03D C-BOP AUTO tie TS-SRO FW14C C-ATC _,___ 7 Trip of RFP 3C/ recover with already warm RFP 3B C-SRO PC 14 (e20 0) 100 Loss of Torus Water level /SCRAM (ATWS)and ED on 8 300 75 M-ALL Torus water level 9 FW30 C-ATC Failure of RFP 3B governor/pump needs tripped. AD01D C-BOP 10 ADS SRV Failures AD01E C-SRO 

     *    (N)ormal,   (R)eactivity,    (l)nstrument,     (C)omponent,    (M)ajor

Appendix D Scenario Outline Form ES-D-1 Events

1. The BOP Operator will perform EHC Auto Pump Start Test & Weekly Pump Alternation IAW 3-01-47A, Section 6.2. Once the EHC Pumps have been swapped the scenario may continue.
2. ATC will withdraw control rods in order to raise power to 8% for a mode change from 2 to 1. Once the Chief Examiner has seen an appropriate amount of power rise he may choose to continue with the next event which will halt control rod withdrawal.
3. Control Rod xx-xx will begin to drift in to approximately position 10, and the ATC will respond IAW 3-AOl-85-5 and bypass the RWM and insert Control Rod xx-xx to position 00. The SRO will declare Control Rod xx-xx inoperable and refer to Tech Spec 3.1.3 condition C. The SRO will also refer to Tech Spec 3.1.6 condition A for BPWS and 3.3.2.1 condition C for RWM. Once the rod has been inserted and the Tech Spec call has been made the scenario may continue.
4. Loss of SJAE A, BOP operator swaps to B SJAE IAW 3-AOl-47-3, Loss of Condenser Vacuum or IAW 3-01-66 or the hardcard. After the standby SJAE has been placed in service and Main Condenser vacuum has recovered, the scenario may continue.
5. The ATC will respond to a failure of IRM 'C' upscale. IRM 'C' will be bypassed IAW 3-AOl-92-A, section 6.1. The SRO will r~l9-_r to Tech Spec 3.3.1.1 and enter an Information Only LCO due to 7/8 IRM's,remaining operable. After the BOP Operator bypasses the failed LPRM and the SRO has completed the Tech Spec call the scenario may continue.
6. 3ED 4KV Shutdown Board will lose power and the 30 Diesel Generator will fail to automatically tie to the Shutdown Board. The BOP will manually tie the Diesel to the board. SRO will refer to Tech Specs and determine TS 3.8.1 condition A, B, and G, and TS 3.8.7.A. The ATC will be reseting
       ~ and PCIS. After the BOP Operator ties the Dlesel to the Buss and the SRO has completed the Tech Spec call the scenario may continue.
7. The ATC will respond to a trip of the 3C RFP IAW 3-AOl-3-1 by raising the speed of the warm RFP 3B to feed the RPV. Once the ATC has entered AOl-3-1 and raised the speed on the standby Reactor Feedpump to maintain RPV Water Level the scenario may continue.
8. At the cue of the Chief Examiner initiate the next event. An unisolable leak will develop on the suppression chamber. The US will direct entry into EOl-3 on secondary containment area flood alarms and EOl-2 on suppression pool water

Appendix D Scenario Outline Form ES-D-1 level. Prior to 12.75 ft, in the Suppression Pool, the US will direct HPCI to be secured and locked out. Prior to 11.5 ft in the Suppression Pool the US will transition to EOl-1 and direct a SCRAM. An ATWS will exist on the SCRAM. The crew will work through EOl-1 and C-5 to insert control rods, maintain reactor water level, and reactor pressure. The US will transition to C-2 to emergency depressurize before Suppression Pool water level lowers to 11.5 feet.

9. The US will direct terminating and preventing IAW EOI appendix 4, and the 3B RFP governor will fail as is. The ATC/BOP will Trip the 3B RFP.
10. The BOP will report that two of the ADS SRV's failed for Emergency Depressurization. Two additional non ADS SRV's will be opened at the direction of the SRO.

The Scenario ends when Emergency Depressurization and Reactor Water Level is restored and maintained within the assigned band or upon request of Lead Examiner.

Appendix D Scenario Outline Form ES-D-1 Critical Tasks 4

1. When Suppression Pool Level cannot be maintained above 12.75 feet HPCI secured to prevent damage.
1. Safety Significance:

Prevent failure of Primary Containment from pressurization of the Suppression Chamber

2. Cues:

Procedural compliance Suppression Pool Level indication

3. Measured by:

Observation - HPCI Auxiliary Pump placed in Pull to Lock

4. Feedback:

HPCI does not Auto initiate No RPM indication on HPCI

2. When Suppression Pool level cannot be maintained above 11.5 feet the US determines that Emergency Depressurization is required, RO initiates Emergency Depressurization as directed by US.
1. Safety Significance:

Precludes failure of Containment.

2. Cues:

Procedural compliance. Suppression Pool Level Trend.

3. Measured by:

Observation - US determines (indicated by announcement or observable transition to C-2) that Emergency Depressurization is required before Suppression Pool level drops below 11.5 feet. Observation - RO opens at least 6 SRV's during performance of Emergency Depressurization actions.

Appendix D Scenario Outline Form ES-D-1

4. Feedback:

RPV pressure trend. SRV status indications. Suppression Pool temperature trend.

3. During an ATWS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection from ECCS and Feedwater until reactor pressure is below the MARFP as directed by US.
1. Safety Significance:

Prevention of fuel damage due to uncontrolled feeding.

2. Cues:

Procedural compliance.

3. Measured by:

Observation - No ECCS injection prior to being less than the MARFP. Observation - Feedwater terminated and prevented until less than the MARFP.

4. Feedback:

Reactor power trend, power spikes, reactor short period alarms. Injection system flow rates into RPV.

Appendix D Scenario Outline Form ES-D-1

4. With RPV pressure <MARFP, slowly increase and control injection into RPV to restore and maintain RPV level above TAF as directed by US.
1. Safety Significance:

Maintaining adequate core cooling and preclude possibility of large power excursions.

2. Cues:

Procedural compliance. RPV pressure indication.

3. Measured by:

Observation - Injection not commenced until less than MARFP, and injection controlled such that power spikes are minimized, level restored and maintained greater than TAF.

4. Feedback:

RPV level trend. RPV pressure trend. Injection system flow rate into RPV.

Appendix D Scenario Outline Form ES-D-1 -=acility: BFN Scenario No.: NRC-4 Op-Test No.: 1501 Examiners: Operators: SRO: ATC: BOP: Initial Conditions: Reactor Power is 100%. EECW Pump A3 and Steam Packing Exhauster 3A are out of service. Turnover: Lower reactor power to <95% to conduct Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9). The Spare RBCCW pump is in service and the 3B RBCCW pump will be tagged out later this shift for an oil change. Event Malfunction Event Event Description No. Number Type* R-ATC 1 N/A Lowers Reactor Power to <95% IAW 3-GOl-100-12 R-SRO Conducts Turbine Control Valve Fast Closure, or Turbine Trip and 2 N/A N-BOP RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9). '7~ EG13A C-BOP Bus Duct Cooling Fan 3A trip C-ATC RBCCW 3A Pump trip and failure of sectionalizing valve to auto-4 SW02A TS-SRO close C-BOP 3 fa' ED10B TS:,~RQ~, Loss of 480V SID Board 3B (~~C-BOP ' 

                        '~~~~

6 MC04 Loss of Condenser Vacuum C-SRO 7 ED01 M-ALL Loss of Offsite Power 8 DG01A C-BOP DG 3EA Fails to Auto start 9 TH21 M-All LOCA 10 HP04 C-BOP HPCI Steam Supply Valve fails to auto open. 

 *     (N)ormal,   (R)eactivity,   (l)nstrument,     (C)omponent,      (M)ajor

Appendix D Scenario Outline Form ES-D-1 Events

1. ATC will lower reactor power to <95% at the direction of the SRO. Once power is approximately 95% the scenario may continue.
2. BOP will conduct Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9). When 3-SR-3.3.1.1.8(9) is completed or at the direction of the lead examiner the scenario may continue.
3. Bus Duct Cooling Fan 3A will trip and annunciator 3-XA-55-7A window 31, GEN BUS DUCT FAN FAILURE, will be received. The BOP operator will place the 3B Bus Duct Cooling Fan in service. When the 3B Bus Duct Cooling Fan is in service the scenario may continue.
4. The crew will respond to a trip of RBCCW Pump 3A IAW 3-AOl-70-1, The spare /

RBCCW pump is aligned and will be the only one running until the crew places RBCCW Pump 3B in service. The RBCCW sectionalizing valve will fail to auto close on the trip of RBCCW Pump 3A, the ATC will close the sectionalizing valve. The SRO will evaluate TRM 3.4.1 and take actions for failure to meet surveillance requirement TSR 3.4.1.1. When the Tech Spec call is completed and the lead examiner is ready the scenario may continue.

5. The crew will respond to a loss of 480V Shutdown Board 3B. This will cause a loss of RPS B, loss of 480V RMOV BO 3B, 3C and 3E. The crew will need to restore power to the 480V RMOV Boards, reset RPS, reset PCIS and restore systems. The SRO will refer to Technical Specification 3.4.5 and determine conditions A, B, and D are required for inoperable containment atmospheric and drywell Floor Drain sump monitoring equipment. Loss of the Shutdown Board will result in entry into Tech Spec 3.8.7 condition B to restore the board in 8 hours and 3.8.7 condition C to declare the affected RHR subsystem (RHR Loop II) inoperable immediately. This will also require entry into Tech Spec 3.8.7.G for U1 and U2 due to the loss of power to GREV B. When power has been restored to the RMOV boards and the Tech Spec call is complete or as directed by the Lead Examiner the scenario may continue.
6. Condenser Vacuum will begin to degrade the SRO will initially enter 3-AOl-47-3 and direct reducing reactor power in an attempt to maintain condenser vacuum. Condenser Vacuum will continue to degrade. The SRO will set a trigger value to trip the main turbine and scram the reactor before an automatic turbine trip occurs at approximately 24.3 inched Hg.
7. After the Reactor Scram on vacuum a Loss of Offsite Power will occur. The crew will respond to the Reactor Scram IAW 3-AOl-100-1 and O-AOl-57-1A.
                                                               /
8. During the LOOP DG 3EA will fail to automatically start and will have to be manually started and after it starts auto tie to the buss.
9. Sometime after the LOOP a LOCA will develop requiring the crew to utilize systems to maintain Reactor Level and Containment parameters.
10. The HPCI Steam Supply Valve, 3-FCV-73-16, will fail to OPEN on an automatic HPCI initiation signal.

The scenario ends when Drywell Sprays have been initiated and Reactor Level is maintained above TAF (- 162 inches) or upon request of Lead Examiner.

Appendix D Scenario Outline Form ES-D-1 Critical Tasks 3

1. RPV Level maintained above TAF (-162 inches)
1. Safety Significance:

Maintaining adequate core cooling

2. Cues:

RPV level indication

3. Measured by:

Reactor level indication above -162 inches

4. Feedback:

RPV level trend HPCl/RCIC injection valve open indication

2. When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve and prior to exceeding the PSP limit.
1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation - US directs Drywell Sprays IAW with EOI Appendix 178 AND Observation - RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment OR

2. Before Drywell temperature rises to 280°F, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve.
1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation - US directs Drywell Sprays IAW with EOI Appendix 178 AND Observation - RO initiates Drywell Sprays

Appendix D Scenario Outline Form ES-D-1

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment}}

Retrieved from "https://kanterella.com/w/index.php?title=ML15079A074&oldid=2511143"