ML102520178

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Draft - Outlines (Folder 2)
ML102520178
Person / Time
Site: Beaver Valley
Issue date: 06/30/2010
From:
FirstEnergy Nuclear Operating Co
To: D'Antonio J
Operations Branch I
Hansell S
Shared Package
ML100560049 List:
References
TAC U01821
Download: ML102520178 (21)
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Text

ES-401 PWR Examination Outline Form ES-401-2 Facility:

Beaver Valley Unit 2 Date Of Exam: Weeks of 8/16 & 8/23 2010 RO KIA Category Points SRO-Only Points Tier Group K1 K2 IK3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G*

1.

1 3

3 3

3 3

3 18 0

0 Emergency 2

I 2

2 I

2 I

9 0

0 N/A N/A Abnormal Tier Plant Totals 4

5 5

4 5

4 27 0

0 Evolutions 1

3 3

3 3

2 I

3 3

2 2

3 28 0

0

2.

2 I

I 0

2 I

I I

1 I

0 I

10 0 I 0 0

Plant Systems Tier 4

4 3

5 3

2 4

4 3

2 4

38 0

0 Totals

3. Generic Knowledge And 1

2 3

4 1

2 3

4 10 Abilities Categories 2

3 2

3 0

0 0

0 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/evolutions 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-onlyexams.

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.

Total 0

0 0

0 0

0 0

NUREG-1021, Revision 9 Supplement 1 Page 1 of 8 FENoe Facsimile Rev. 0

PWR RO Examination Outline Facility:

Beaver Valley Unit 2 ES*401 Emergency and Abnormal Plant Evolutions* Tier I I Group I Form ES*401*2 I E/APE # I Name I Safety Function KI K2 K3 Al A2 G

KA Topic Imp_

Points Ql 000007 Reactor Trip* Stabilization*

X EK 1.02. Shutdown margin 3.4 1

Recovery II Q2 000008 P'ess"dz" V",o, Sp,e, ACdd"'tffilli AK2.03

  • Controllers and positioners 2.5 I

13 Q3 000009 Small Break LOCA 13 EK2.03. SIGs 3.0 I

Q4 000015/000017 RCP Malfunctions 1 4 X

AA 1.03

I and indicators Q5 000022 Loss of Rx Coolant Makeup 12 X

AA2.0 I. Whether charging line leak 3.2 I

exists Q6 000025 Loss of RHR System I 4 X

AK3.02* Isolation ofRHR low-pressure 3.3 I

piping prior to pressure increase above specified level Q7 000027 Pressurizer Pressure Control X

2.2.44

  • Ability to interpret control room 4.2 I

System Malfunction 13 indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Q8 000029 A TWS I 1 X

EK 1.05

  • definition of negative 2.8 1

temperature coefficient as applied to large PWR coolant systems Q9 000040 Steam Line Rupture - Excessive X

AK 1.03 - RCS shrink and consequent 3.8 Heat Transfer 1 4 depressurization QI0 000054 Loss of Main Feedwater 1 4 X

AA2.07 - Reactor trip first*out panel 3.4*

I indicator QII 000055 Station Blackout 16 X

EA2.02

  • RCS core cooling through 4.4 I

natural circulation cooling to S/G cooling Q12 000056 Loss ofOff-site Power / 6 AK3.01 - Order and time to initiation of 3.5 I

power for the load sequencer Ql3 000062 Loss of Nuclear Svc Water 14 X

AK3.04 Effect on the nuclear service 3.5 I

water discharge flow header ofa loss of CCW Q14 000065 Loss of Instrument Air / 8 X

AA 1.02 - Components served by 2.6 I

instrument air to minimize drain on system Q15 000077 Generator Voltage and Electric X

AA 1.03 - Voltage regulator controls 3.8 I

Grid Disturbances 1 6 Ql6 W/E04 LOCA Outside Containment 13 X

2.1.20 - Ability to interpret and execute 4.6 I

procedure steps.

I NUREG-1021, Revision 9 Supplement 1 Page 2 of 8 FENoe Facsimile Rev. 0

PWR RO Examination Outline Facility:

Beaver Valley Unit 2 ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 I Group 1 Form ES-401-2 I E/APE # I Name I Safety Function Kl K2 K3 Al A2 G

KA Topic Imp.

Points Ql7 W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink 14 X

EK2.1 - Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 3.7 I

Ql8 W/Ell Loss of Emergency Coolant Recirc./4 X

2.4.6 - Knowledge of EOP mitigation strategies.

3.7 I

KIA Category Totals:

3 3

3 3

3 3

Group Point Total:

18 NUREG-1 021, Revision 9 Supplement 1 Page 3 of 8 FENOC Facsimile Rev. 0

PWR RO Examination Outline Facility:

Beaver Valley Unit 2 ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 I Group 2 Form ES-401-2 I E/APE # I Name I Safety Function Kl K2 K3 At A2 G

KA Topic Imp.

Points Qt9 000003 Dropped Control Rod I I X

AK1.l5 - Definition and application of 2.8 1

power defect Q20 000036 Fuel Handling Accident I 8 X

AK3.02 - Interlocks associated with fuel 2.9 handling equipment Q21 000060 Accidental Gaseous Radwaste X

AA2.02 - The possible location of a 3.1 J

ReI. 19 radioactive-gas leak, with the assistance of PEa, health physics and chemistry personnel Q22 000067 Plant Fire On-site I 9

! X I AA2.16 - Vital equipment and control 3.3 I

systems to be maintained and operated during a fire Q23 W/E02 SI Termination I 3 X

EA 1.1 - Components, and functions of 4.0 I

control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Q24 W IE03 LOCA Cooldown - Depress. I 4 X

2,4.18 - Knowledge of the specific bases 3.3 I

for EOPs.

Q25 W/E071nad. Core Cooling 14 X

EK3,4 - RO or SRO function within the 3.3 I

control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated Q26 WIE09 Natural Circ. I 4 X

EK2.2 - Facility's heat removal systems, 3.6 I

including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Q27 W IE 16 High Containment Radiation 19 X

EK2.2 - Facility's heat removal systems, 2.6 I

including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation ofthe facility KIA Category Totals:

1 2

2 I

2 1

Group Point Total:

9 NUREG-1021, Revision 9 Supplement 1 Page 4 of 8 FENOC Facsimile Rev. 0

PWR RO Examination Outline Facility:

Beaver Valley Unit 2 Plant Systems - Tier 2 / Group 1 Form ES-401-2 ES - 401 Sys/Evol # / Name Kl K2 K3 K4 K5 K6 At A2 A3 A

KA Topic Imp.

Points Q28 003 Reactor Coolant Pump X

K2.01 - RCPS 3.1 i Q29 003 Reactor Coolant Pump X 2.4.31 - Knowledge of 4.2 annunciator alarms, indications, or response procedures.

Q30 004 Chemical and Volume KI.22 - BWST 3.4 Q31 OOS Residual Heat Removal KS.O I - Nil ductility 2.6 transition temperature (brittle fracture)

Q32 006 Emergency Core Cooling X

- Loss of heat tracing 2.8 Q33 007 Pressurizer Relief/Quench X

I - Quench tank cooling 2.6 Tank Q34008 Component Cooling Water X 2.1.28 - Knowledge of the 4.1 purpose and function of major system components and controls.

Q35 008 Component Cooling Water X

A2.04 - PRMS alarm 3.3 Q36 0 to Pressurizer Pressure

- PZR heaters 3.0 Q37 010 Pressurizer Pressure

- RCS 3.8 Q38 012 Reactor Protection X

- Bistables and bistable 2.8 uipment Q39 013 Engineered Safety Features

- Fuel 4.4 Actuation Q40 013 Engineered Safety Features X 2.1.7 - Ability to evaluate 4.4 Actuation plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Q41 022 Containment Cooling K1.04 - Chilled water 2.9*

Q42 026 Containment Spray K 1.02 - Cooling water 4.1 Q43 039 Main and Reheat Steam KS.OS - Bases for RCS 2.7 wn limits Q44 OS9 Main Feedwater X

A4.12 - Initiation of 3.4 automatic feedwater isolation Q45 059 Main Feedwater X

K4.13 - Feedwater fill for S/G 2.9 upon loss of RCPs Q46 061 Auxiliary/Emergency X

A2.07 - Air or MOV failure 3.4 Feed water Q47 062 AC Electrical Distribution X

A1.0 I - Significance of D/G

3.

load limits Q48 063 DC Electrical Distribution X

A4.02 - Battery voltage 2.8*

indicator NUREG-1 021, Revision 9 Supplement 1 Page 5 of 8 FENoe Facsimile Rev. 0

PWR RO Examination Outline Facility:

Beaver Valley Unit 2 Plant Systems - Tier 2 I Group 1 Form ES-401-2 ES - 401 I

I I

I I

I I

Sys/Evol # I Name Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp.

Points Q49 063 DC Electrical Distribution X

K3.02 - Components using DC control power 3.5 I

Q50 064 Emergency Diesel X

K2.02 - Fuel oil pumps 2.8*

I Q51 064 Emergency Diesel

=

I X

A3.06 - Start and stop 3.3 I

Q52 073 Process Radiation X

K4.01 - Release termination when radiation exceeds setpoint 4.0 I

Q53 076 Service Water X

A 1.02 - Reactor and turbine building closed cooling water temperatures 2.6*

I Q54 078 Instrument Air I

X A3.0 I - Air pressure 3.1 I

Q55 103 Containment X

A1.0 I - Containment pressure, temperature, and humidity 3.7 I

KIA Category Totals:

3 3

3 3

2 1

3 3

2 2

3 Group Point Total:

28

PWR RO Examination Outline Facility:

Beaver Valley Unit 2 Plant Systems - Tier 2 / Group 2 Form ES-401-2 ES - 401 I

I Sys/EvoJ # / Name K3 K4 K5 K6 Al A2 A3 A4 G IKA Top;'

Imp.

Points.

Q56 001 Control Rod Drive X

d height 3.7 I

Q57 028 Hydrogen Recombiner and X

A2.03 - The hydrogen air 3.4 1

Purge Control concentration in excess of limit flame propagation or detonation with resulting equipment damage in containment Q58 029 Containment Purge X

A 1.02 - Radiation levels I

Q59 035 Steam Generator X

KLOI - MFW/AFW systems 4.2 1

I Q60 045 Main Turbine Generator X

K5.17 - Relationship between 2.5*

1 moderator temperature coefficient and boron concentration in RCS as T/G load increases Q61 068 Liquid Radwaste X

K6.10 - Radiation monitors 2.5 1

Q620TI Waste Gas Disposal X

K4.04 - Isolation ofwaste gas 2.9 I

I release tanks Q63 072 Area Radiation Monitoring X 2.4.31 - Knowledge of 4.2 1

annunciator alarms, indications, or response I

procedures.

Q64 075 Circulating Water X

K2.03 - Emergency/essential 2.6*

I SWS pumps Q65 086 Fire Protection

~

K4.06 - CO2 3.0 1

KIA Category Totals:

I I

0 2

I I

1 1

1 0

I Group Point Total:

10

Generic Knowledge and Abilities Outline (Tier 3)

PWR RO Examination Outline Facility:

Beaver Valley Unit 2 Form ES-401-3 Generic Cate!!orv KA KA Topic Points Conduct of Operations 2.1.39 Q66 Knowledge of conservative decision making practices.

3.6 I

2.1.43 Q67 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

4.1 1

Category Total:

2 Equipment Control 2.2.25 Q68 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

3.2 J

2.2.38 Q69 Knowledge of conditions and limitations in the facility license.

3.6 I

2.2.41 Q70 Ability to obtain and interpret station electrical and mechanical drawings.

3.5 I

Category Total:

3 I Radiation Control 2.3.7 Q71 Ability to comply with radiation work permit requirements during normal or abnormal conditions.

3.5 1

2.3.15 Q72 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

2.9 1

Category Total:

2 Emergency Procedures/Plan 2.4.23 Q73 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

3.4 I

2.4.25 Q74 Knowledge of fire protection procedUres.

3.3 I

2.4.45 Q75 Ability to prioritize and interpret the significance of each annunciator or alarm.

4.1 I

Category Total:

3 Generic Total:

10

d f R* t d KIA Form ES 401-4 ES-401 Recor 0 eJec e s

Facility: Beaver Valley Unit 2 Date of Exam Weeks of 8/16 & 8/23 2010 Operating Test No.: NRC 2LOT7 (jer I Randomly Reason for Rejection Group Selected KIA RO OUTLINE 1/1 0272.2.18

  • ES-401 D.1.b requires exclusion of generic KlAs for Tier 1 &2. These generic KlAs were not suppressed and therefore were not automatically omitted using PWROG Random Generator (manually reselected from same generic group) 1/1 W/E04 2.1.42
  • Same as above.

1/1 W/E112.4.12

  • Same as above.

1/2 W/E032.4.16

  • Same as above.

2/1 0032.4.32

  • Same as above.

211 0082.1.36

  • Same as above.

2/1 0132.1.8

  • Same as above.

2/1 025.K6.01 Ice Condensers are not applicable to BVPS 2/1 073 K4.02 Letdown Isolation on High RCS Activity is not applicable to BVPS.

2/1 022 K1.02 SEC/Remote Monitoring Systems are not applicable to Containment Cooling at BVPS 2/2 027 K1.01 Containment Iodine Removal System is no longer applicable to BVPS Unit 2.

I During last refueling outage a plant modification was completed to use a passive sodium tetraborate system.

2/2 0722.4.27

  • Same as above.

2/2 0722.4.34 The manually reselected KIA has no RO tasks associated with the Area Radiation Monitoring System performed outside of the control room, during an emergency.

212 086 A4.01 Fire Water Pumps are operated from BVPS Unit 1 and therefore are not operated or monitored from BVPS Unit 2. With concurrence from Chief Examiner this KIA was deselected on the basis Unit 2 RO discrimination is not applicable.

SRO OUTLINE 1/1 0572.4.25

  • Same as above.

1/2 0742.4.40

  • Same as above.

2/1 0222.2.6

  • Same as above.

2/2 011 2.4.29

  • Same as above.

II 2/1 025 A2.05 Ice Condensers are not applicable to BVPS NUREG-1021, Revision 9 Supplement 1 FENoe Facsimile Rev, 0

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

Beaver Valley Unit 2 Date of Examination:

Weeks of 8/16 & 8/232010 Examination Level RO [g]

SROD Operating Test Number:

2LOT7 Administrative Topic (See Note)

(A1.1) Conduct of Operations (A 1.2) Conduct of Operations (A2)

Equipment Control Type Code*

RN RN RN Describe activity to be performed 2.1.25 3.9 Ability to interpret reference materials, such as graphs, curves, tables, etc.

2AD-036 Perform SDM Calculation for At power condition and ONE inoperable Rod.

2.1.7 4.4 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

2AD-030 Calculate the RCS initial void volume and final void volume (lAW 20M-6.4.T, Response to Voids in the reactor vessel) 2.2.13 4.1 Knowledge of tagging and clearance procedures.

2AD-031 Prepare a clearance Tagout for Quench Spray Pump 2QSS*P21B (A3)

Radiation Control RD 2.3.11 3.8 Ability to control radiation releases.

2AD-010 Determine GW Discharge Bleed Flowrate Emergency Procedures/Plan N/A NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative i

topics, when ali 5 are required, I *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)

Ii (P)revious 2 exams (:::: 1; randomly selected)

NUREG-1021, Revision 9 Supplement 1 FENOC Facsimile Rev, 0

ES-301 Administrative Topics Outline Form ES-301-1 II Facility:

Beaver Valley Unit 2 Date of Examination:

Weeks of 8/16 & 8/23 2010 Examination Level RO 0 SRO IRl Operating Test Number:

2LOT7 Administrative Topic (See Note)

Type Code*

Describe activity to be performed (A1.1) Conduct of Operations R, N 2.1.25 4.2 Ability to interpret reference materials, such as graphs, curves, tables, etc.

2AD-035 Perform SDM Calculation for At power condition and ONE inoperable Rod; Determine Tech Spec Applicability.

(A1.2) Conduct of Operations R,M 2.1.3 3.9 Knowledge of shift or short-term relief turnover practices.

2AD-024 Determine Availability for call-in (3 ROs)

(A2)

Equipment Control i

R, N 2.2.37 4.6 Ability to determine operability and/or availability of safety related equipment.

2AD-033 Determine Compensatory Actions for Low CO2 Tank Level (A3) Radiation Control R,D 2.3.4 3.7 Knowledge of radiation exposure limits under normal or emergency conditions.

2AD-014 Approve Emergency Exposure i

(A4)

Emergency Procedures/Plan S, N 2.4.41 4.6 Knowledge of the emergency action level thresholds and classifications.

2AD-034 Classify an E-Plan event (Scenario Specific) and Complete Initial Notification Form NOTE All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • Type Codes &Criteria (C)ontrol Room, (S)imulator, or Class(R)oom (D)irect from bank (::; 3 for ROs; :::. 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (:;: 1)

(P)revious 2 exams (::: 1; randomly selected)

NUREG-1 021, Revision 9 Supplement 1 FENOC Facsimile Rev.O

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Beaver Valley Unit 2 Date of Examination:

Weeks of 8/16 & 8/23 2010 Exam Level: RO [E] SRO(I) []

SRO(U) 0 Operating Test No.:

2LOT7 I

Control Room Systems'91 (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Safety Code*

Function

a. (S1) RCS Dilution (2CR-089)

S,D 1

I

b. (S2) Transfer from Hot Leg to Cold Leg Recirculation (2CR-560)

S, D,A, E, 3

EN S,D 2

c. (S3) Place Excess Letdown in Service (2CR-56)

I

d. (S4) Respond to a Reactor Coolant Pump #1 Seal Failure (2CR-040)

S,D,L 4P

e. (S5) Transfer from Bypass to Main Feed Regulating Valve (2CR-520)

S,N,A 4S

f. (S6) Synchronize and Load 2-1 EDG (2CR-524)

S,D,A,EN 6

g. (S7) Primary Component Cooling Water Pump (2CCP*21A) Test (2CR-157) 8 S, N,L 9

S, N,E,A If h. (S8) Verify CREVS Actuation (2CR-599)

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

i. (P1) Transferring Powerfor2RHS-MOV702A (2PL-061)

D,E 4P

j. (P2) Locally Throttle AFW Valves during ECA - 0.0 (2PL-150)

N, E,R 4S i

k. (P3) Test The EDG UV Start Relay (2PL-549) 6 0, E,A All RO and SRO 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.

...~.

(A)lternate Path 4-6 14-6 12-3 (C)ontrol room (D)irect from bank

9/
:;8/::;4 (E)mergency or abnormal in-plant
1/;
;1/2:1 (EN)gineered safety feature

- I-I ~ 1 (Control room system)

(L)ow-power 1Shutdown 2:1/2:1/2:1 (N)ew or (M)odified from bank including 1 (A) 2:2/2:2/2:1 (P)revious 2 exams

3/
:; 3/::; 2 (randomly selected)

(R)CA I

2:1/2:1/2:1 (S)imulator NUREG-1 021, Revision 9 Supplement 1 FENoe Facsimile Rev.O

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

Beaver Valley Unit 2 Date of Examination:

Weeks of 8/16 & 8/23 2010 Exam Level: RO 0 SRO(I) lID SRO(U) 0 Operating Test !\\lo.:

2LOT7 Control Room Systems<gi (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Type Code*

Safety Function System I JPM Title

a. (S1) RCS Dilution (2CR-089) 1 S, D
b. (S2) Transfer from Hot Leg to Cold Leg Recirculation (2CR-560)

S, D, A, E, 3

EN 2

S, D

c. (S3) Place Excess Letdown in Service (2CR-56)
d. (S4) Respond to a Reactor Coolant Pump #1 Seal Failure (2CR-040) 4P S, D,L
e. (S5) Transfer from Bypass to Main Feed Regulating Valve (2CR-520) 4S S, N,A
f. (S6) Synchronize and Load 2-1 EDG (2CR-524) 6 S, D,A,EN
g.
h. (S8) Verify CREVS Actuation (2CR-599)

S, N,E,A 9

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

i. (P1) Transferring Power for 2RHS-MOV702A (2PL-061)

D, E 4P

j. (P2) Locally Throttle AFW Valves during ECA - 0.0 (2PL-150) 4S N,E, R
k. (P3) Test The EDG UV Start Relay (2PL-549) 6 All RO and SRO 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.

D, E,A

  • Type Codes Criteria for RO 1SRO-I 1SRO-U (A)lternate Path 4-6 14-6 12-3 (C)ontrol room (D)irect from bank
59/:58/:54 (E)mergency or abnormal in-plant
1/::::1/::::1 (EN)gineered safety feature I::: 1 (Control room system)

(L)ow-power 1Shutdown

1/::::1/::::1 (N)ew or (M)odified from bank including 1 (A)
2/::::2/::::1 (P)revious 2 exams
5 31:5 31:5 2 (randomly selected)

(R)CA

1/::::1/::::1 (S)imulator NUREG-1 021, Revision 9 Supplement 1 FENOC Facsimile Rev,O

" 1 0 tr A~ppend"IX 0 Scenarlo U Ine Form ES-0 1 II Facility:

Beaver Valley Unit 2 Scenario No.:

1 Op Test No.:

2LOT7NRC Examiners:

Candidates:

SRO ATC BOP Initial IC 211: 61 % power, MOL, Equ. Xe Conditions, CB "D" @ 152 steps, RCS boron - 1007 Conditions:

ppm.

Turnover:

Maintain current power level.

Critical Tasks:

E-O.A, Auto Rx Trip failure E-O.H, Start LHSI Pumps E-O.E, Manually Initiate CIB Event Malf. No.

Event Type Event Description No.

1 XMT-RCS054A I(ATC/SRO)

Loop 1 Tcold RTD fails low SROTS 2

PMP-CFWOO4 C(BOP/SRO) 2FWS-P21A pump trip.(Power reduction required) 3 R(ATC)

Power reduction to <50%

N(SRO/BOP) 4 C(ALL)

FLX-CCP34 CCP supply leak to 2RCS*P21B (10 minute ramp to 450 gpm leads to an automatic RCP trip).

PMP-RCP003 SROTS 5

I C(ATC)

PPL01A Auto Reactor Trip failure (manual available)

PPLOIB (SRO) 6 RCS03B M(ALL)

B Loop Large break LOCA 7

C(ATC)

PPL07A Both low head SI pumps fail to auto start (manual start available).

PPL07B (SRO) 8 PPL09A C(BOP)

Auto CIB failure (manual available)

PPL09B (SRO)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG*1021, Revision 9 Supplement 1 FENoe Facsimile Rev 0

Appendix D Scenario 1 Outline Form ES-D-1 "he crew will take the shift at 61 % power, MOL at Equilibrium Xe, Bank D control Rods at 152 steps and RCS ooron is at 1007 ppm. Crew instructions are to maintain current stable plant conditions.

Loop I Tcold RTD will fail low, requiring implementation of20M-6.4.1F, attachment 4. The ATC will have to accurately determine failed channel and select a non-failed channel and the SRO will evaluate TS.

2FWS-P21A will then trip requiring the crew to enter AOP 2.24.1 and rapidly reduce power to less than 50%.

When plant conditions have begun to stabilize, a leak develops on the CCP supply line to 2RCS-P2IB, "B" RCP. The leak will ramp to 450 gpm over ten minutes. First indication is annunciator A2-2B for UIL HIGH, ARP 20M-9.4.AAL will be entered at which point dropping CCP surge tank level indication will require entry into AOP 2.15.1. The US will evaluate TS 3.7.7 applicability. The "B" RCP will degrade and eventually trip with a failure of the reactor to trip automatically, The Unit Supervisor should direct a manual reactor trip and enter E-O Immediately following the reactor trip, a large break LOCA occurs. After completing E-O immediate actions, all RCPs should be manually tripped due to the loss of component cooling.

Additional failures that occur following the LOCA, the A TC operator will identify that both low head SI pumps failed to start automatically and will manually start them. During the performance ofEOP attachment A-O.II, the BOP will identify that Containment Isolation Phase "B" (CIB) failed to actuate automatically and will manually initiate CIB. The Unit Supervisor will progress through E-O to E-l at the "Check if the RCS is Intact"

<=;tep, E-l will be implemented until the crew determines that ES-I.2 transition is not appropriate in E-I step 19 Jased upon RCS pressure being less than 225 psig.

The drill is terminated at step 20 of E-I when the crew evaluates if Transfer to Cold Leg Recirculation is required.

Expected procedure flow path is EO ---+ E I.

NUREG*1021, Revision 9 Supplement 1 FENOC FaCSimile Rev 0

d* D S

. 20 r Alppen IX cenarlo utme Form ES D 1 II Facility:

Beaver Valley Unit 2 Scenario No.:

2 Op Test No.:

2LOT7NRC Examiners:

Candidates:

SRO ATC BOP Initial IC 212: 10% power, BOL, Equ. XE Conditions, CB "D" @ 118 steps, RCS boron - 1884 Conditions:

ppm. SG 21A Bypass Feed Reg Valve (BPFRV) is in Manual, Auto control is erractic.

Turnover:

Raise power to 15% to S/U main turbine.

Critical Tasks:

E-3.A, Isolate Ruptured SG E-3.B, Cooldown RCS E-3.C, Depressurize RCS Event Event Type Event Description No.

1 Malf. No.

R(ATC)

Normal power increase to 15% lAW 20M-52.4.A N(SRO/BOP) 2 RCS04B C(ALL)

SG 21B Tube Leak SROTS 3

CNHMSS03A C (ATC/SRO)

SG 21 B atmospheric dump valve fails open.

SROTS 4

NIS08A I(ALL)

N41 Power Range Instrument fuse blown SRO TS) 5 RCP06B C (ATC/SRO) 21 B RCP high vibration (Manual RCP trip required)

RCPOIB Manual reactor trip 6

RCS04B M (ALL) 21 B SG Tube Rupture 7

CNHPCS07A I (BOP/SRO)

Condenser steam dumps fail closed.(Requires cooldown with 21 A and 21 C Atmospheric Steam Dumps 8

VLV-MSS057A C (BOP/SRO) 2SDS-AOVI29A failed open, requires RNO actions for S/G isolation.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 9 Supplement 1 FENoe Facsimile Rev 0

Appendix D Scenario 2 Outline Form ES-D-1

'\\fter taking the shift, the crew will continue the startup to raise power to 15% in accordance with 20M-52.4.A,

~Jart IV A, Plant Startup, currently the plant is at 10% power and Step 7 is in progress. The A TC will withdraw control rods lAW the reactivity plan while the BOP will be required to manually control the "A" S/G level due to the AUTO level control function being OOS.

A SG Tube leak will occur as evidenced by Rad monitor indications, the Unit Supervisor will direct actions per AOP 2.6.4 and refer to TS 3.4.13 for Primary to Secondary leakage.

2SVS*PCVI0IB, 21B SG atmospheric steam dump valve will fail open and stick open, the valve will not close in Auto, Manual or Locally, requiring the crew to direct the local isolation. The Unit Supervisor will refer to TS 3.7.4 for Inoperable Atmospheric Steam Dumps.

An Instrument power fuse will blow for Power range Channel N41, requiring the use of AOP 2.2.1C. The crew will take actions per AOP 2.2.1C and the SRO will evaluate TS 3.3.1.

When the AOP actions are completed for N41, a 21B RCP high vibration condition develops, the crew will respond using AOP 2.6.8. The high vibration will increase at>1 millhr which requires the crew to trip the reactor, enter E-O and take the reactor coolant pump out of service.

Following the reactor trip, a 500 gpm SGTR occurs on the 21B SG resulting in a Safety Injection actuation.

The crew will transition to E-3 at the "Check if SG Tubes are Intact" step of E-O.

When the cooldown is attempted in E-3 the condenser steam dumps will fail closed, requiring the BOP operator

,0 use manual control of the atmospheric steam dump valves and/or 2SVS-HCV104.

The drill will be terminated when HHSI flow is isolated in E-3.

Expected procedure flow path is EO ----> E3.

NUREG-1 021, Revision 9 Supplement 1 FENoe Facsimile Rev 0

d" 0

" 30 r A~ppen IX Scenarlo utme Form ES-01 II Facility:

Beaver Valley Unit 2 Scenario No.:

J Op Test No.:

2LOT7NRC Examiners:

Candidates:

SRO ATC BOP Initial IC-213 MOL, 100 % power Equ Xe, Rods Bank D @ 230 steps, RCS Boron - 883 PPM Conditions:

Turnover:

Maintain current plant conditions Critical Tasks:

FR-S.l.A, Crew isolates the main turbine from the SG's FR-S.1.C, Crew inserts negative reactivity into the core by inserting RCCAs E-2.A, Crew isolates/directs-isolation of faulted SG Event Malf. No.

Event Type Event Description No.

1 XMT-MSS043A I(ALL) 2MSS*PT447 fails LOW, Rods Auto insert, Rx power rises due to cold Feedwater, Power reduction required.

SRO T.S.

2 R(ATC)

Emergency Power reduction N(BOP/SRO) 3 I(ATC/SRO)

XMT-RCS030A 2RCS*PT444 drifts HIGH, pzr pressure decreases, manual SROT.S.

control of Pzr pressure required.

4 M(ALL)

FLX-CFW31 4500 gpm Feedwater leak inside cnmt on "A" S/G 5

M(ALL)

PPLOIA PPLOIB ATWS Failure of auto/manual Rx trip 6

C(BOP/SRO)

EHC03B Incomplete turbine trip, requires manual Steamline isolation actuation EHCOIB 7

VLV-MSS003A C(BOP/SRO)

Auto MSLI Isolation actuation failure with 2MSS-AOVI01A PPLlOA PPLIOB failing to close on a manual MSLI actuation, manual isolation required.

8 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 9 Supplement 1 FENoe Facsimile Rev 0

Appendix D Scenario 3 Outline Form ES-D-1 The crew will assume the shift at 100% power, Equilibrium MOL conditions with instructions to maintain vurrent power level.

2MSS*PT447 will fail low, this will cause 2CNM-AOVI00 to open resulting in an overpower condition, causing the control rods to insert. The crew will respond by performing an Emergency Shutdown lAW AOP 2.51.1. Once the overpower condition has been corrected, 20M-24.4.IF, attachment 5 will be implemented to respond to 2MSS*PT447 failure. The SRO will evaluate TS 3.3.1 applicability and initiate required actions.

When the plant has stabilized, 2RCS*PT444 will drift high, requiring use of20M-6.4.IF, Attachment 2, the ATC will be required to manually control RCS pressure. The SRO will evaluate TS 3.4.1 applicability.

A 4500 gpm Feedwater leak on the "A" loop inside containment will begin to ramp in, 1 st indication of leak will be containment sump pumpout and Incore sump alarms. Based upon the sump alarms, the crew may enter AOP 2.6.7 to diagnose leak location. Feed flow to the "A" S/G will increase, however, S/G level will begin to drop.

Level will continue to drop to the Reactor trip setpoint, however, Auto and Manual Reactor trips have been inhibited. When the crew recognizes a reactor trip is necessary a manual reactor trip will be directed by the SRO and E-O will be entered.

When it is recognized that the reactor will not trip, the SRO will direct entry into FR-S.l, this will be complicated by TV -2 and GV -2 sticking open on a turbine trip. An automatic MSLI actuation is inhibited requiring the BOP to manually actuate a MSLI, this is further complicated by 2MSS-AOV101A failing to automatically close on the manual MSLI actuation, which will require manual closure. The A TC will be inserting control rods and initiating emergency boration. An operator will be dispatched to locally trip the cactor, which will not occur until the crew reduces power to <5% at which time the crew will transition back to E-O. The crew will identify the "A" S/G as faulted and transition to E-2 to isolate.

When the crew has isolated the "A" S/G, in E-2, the drill will be terminated.

Expected Procedure flow path is E-O -

FR-S.l -

E-O-E-2 NUREG*1 021, Revision 9 Supplement 1 FENoe Facsimile Rev 0

lppend"IX D

" 40 r Form ES D 1 A

Scenarlo ut Ine Facility:

Beaver Valley Unit 2 Scenario No.:

1 Op Test No.:

2LOT7NRC Examiners:

Candidates:

SRO ATC BOP Initial IC 214: Reactor power E-6 amps, BOL, Equ. Xe Conditions, CB "D" @ 101 steps, RCS Conditions:

boron - 1851 ppm.

Turnover:

Raise power to above the POAH (3-5%)

Critical Tasks:

E-1.C, Trip RCPs E-O.I, Manually Start HHSI Pump ECA-1.1.B, Makeup to RWST Event Malf. No.

Event Type Event Description No.

1 R(ATC)

Raise Reactor power to 3-5%

N(BOP/SRO) 2 2RSS-P21D, Recirculation Spray Pump Seal failure BST-CSS035 C(ATC/SRO) 2RSS-P21D seal tank level low BST-CSS036 SROTS 2RSS-P21D seal tank level 10-10 3

SIS01 C(BOP/SRO) 500 gpm suction leak to the "B" Quench Spray Pump.

SROTS 4

PMP-CAS003 C(BOP/SRO)

Station Air compressor Trip/auto start failure of standby FLX-CAS10 C(ALL) 5 Instrument Air Header leak - requires manual reactor trip 6

RCS02C M(ALL) 5000 gpm SBLOCA on Loop C upon reactor trip BKR-HIV08 Inadvertent trip of 2DF feeder Bkr on Rx trip.

DSG01B 7

C(BOP/SRO) 2-2 EDG Auto start failure with subsequent trip 8

PPL07A C (ATC/SRO)

Standby HHSI pump fails to auto start (manual start required).

9 LOA-LOVOn C(ALL)

Loss of2MCC-E11, (Both Trains of Transfer to Cold Leg Recirculation are unavailable - ECA -1.1 entry required)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 9 Supplement 1 FENoe Facsimile Rev 0

Appendix D Scenario 4 Outline Form ES-D-1

'he Crew will continue the reactor startup, raising reactor power to above the POAH, ~3%. The A TC operator will withdraw control rods and observe startup rate while the BOP controls RCS temperature by manually adjusting 2SVS-HCVI04.

After the Crew has raised power to above the POAH, AI-3H will annunciate due to a low level in the seal tank for 2RSS*P21D. AI-3H will reflash in 5 minutes due to the seal tank for 2RSS-P21D level being LO-LO which indicates a seal failure for 2RSS *P21 D. lAW the ARP for A 1-3H, the crew will dispatch an operator to fill the seal lAW 20M-13.4.F. The SRO will evaluate TS 3.5.2 & 3.6.7 A 500 gpm isolable leak develops on the suction line to 2QSS*P21B in North Safeguards. The North Safeguards sump Hi level annunciator will alarm followed by annunciator A6-1 D due to a low level in the RWST, the crew will monitor RWST level and initiate actions to locate the leak and makeup to the RWST. The SRO will evaluate TS 3.5.4 (RWST level) and TS 3.6.6 for 2QSS*P2lB when leak is located)

The running station air compressor will trip requiring manual start ofthe standby station air compressor.

A slowly developing Instrument Air leak will occur. The air leak will continue to worsen, requiring the crew to determine that a reactor trip is appropriate lAW AOP 2.34.1.

Upon the reactor trip the crew will enter E-O, a 5000 gpm SBLOCA will occur on Loop C. When the reactor trips, one of the 4kv tie breakers, between the 44D" and "DF" buses will trip open, 2D 1 O. The automatic start of the 2-2 EDG will fail and requires the BOP to manually start, it will trip 90 seconds after starting. Since the

~'l3" CHS/HHSI pump was in service it will no longer be running, the "A" CHS/HHSI pump will require the

~TC to manually start it due to an automatic start failure. The crew will Transition to E-l based upon containment conditions when checking ifthe RCS is intact. 5 minutes after the reactor trip, the feeder breaker to 2MCC -Ell will trip When the crew performs EOP attachment A-0.6 to verify Cold Leg Recirculation capability, it will be determined that at least one train ofrecirculation capability cannot be verified requiring the SRO to transition to ECA-l.l. The crew will remain in ECA-l.l; after the crew has initiated actions to add makeup water to the RWST (step 9) the drill will be terminated.

Expected procedure flow path is E-O ~ E-l ~ ECA-l.l.

NUREG-1021, Revision 9 Supplement 1 FENoe Facsimile Rev 0

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