ML19322B185

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Draft Written Examination and Operating Test Outlines (Folder 2)
ML19322B185
Person / Time
Site: Millstone 
Issue date: 04/03/2019
From: Peter Presby, Thomas Setzer
Operations Branch I
To:
Dominion Nuclear Connecticut
Shared Package
ML19052A319 List:
References
000500, 000956
Download: ML19322B185 (26)
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Text

I.,

.1 ES-201 Examination Outline Quality Checklist Fonn ES-201 *2 Facility:.

Date of Examination:

Initials Item Task Description a

b*

c-

1.

a, Verify that the ouUlne(s) fit(s) the appropriate model in accordance with E8-401 or ES-401 N.

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b. Assess whether* the outline was systematically and randomly prepared in accordance with R

Section 0.1 of ES-401 or ES-401 N and whether all KIA categories are appropriately sampled.

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C. Assess whether the outline overemphasizes any svstems, evolutions, or generic topics.

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d. Assess whether the justlficatlons for deselected or rejected KIA statements are appropriate.

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1 N

2.
a. Using Fonn ES-301-5, verify that the proposed scenario sets cover the required number of IL.

normal evolutions, instrument and component failures, technical specifications, and major

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transients.

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b. Assess whether there are enough scenario sets (and spares) to test the projected number and u

mix of applicants In accordance with the expected crew composition and rotation schedule *

'l- ~

L without COIT)prorrising exam Integrity, and ensure that each applicant can be tested using at f1 A

least one new or significantly modified scenario, that no scenarios are duplicated from the T

applicants' audit test(s), and that scenarios will not be repeated on subsequent days.

0 C. To the extent possible, assess whether the outllne(s) conforms with the qualllalive and

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. 'L--

quantitative criteria specified on Form ES-301-4 and described in AppendlX D and In

~

"(ti Section 0.5, "Specific Instructions for the 'Simulator Operating Test,'" of ES-301 (Including overlap).

3.
a. Verify that the systems walkthrough outiine meets the criteria spec11ied on Form ES-301-2:

(1) The outilne(s) contains the required number of control room and in-plant tasks distnbuted w

among the safety functions as specified on the. form.

A (2) Task repetition from the last two NRC examinations is within the limits specified on the form.

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L (3) No tasks are duplicated from the appficant's audit test(s).

K (4) The number of new or modllled tasks meets or exceeds the minimums specified on the form.

T (5) The number of altemate-path, low-power, emergency, and radiologically controlled area H

tasks meets the criteria on the form.

R 0

b. Verify that the administrative outitne meets tile criteria specified on *Form ES-301-1:

u (1) The tasks are distributed among the topics as specified on the fonn.

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(2) At least one task Is new or significantly modified.

H (3)

No more than one task is repeated from the last two NRC licensing examinations.

c. Determine whether there are enough different outlines to test the projected number and mix of ~ "'

I~

applicants and ensure that no Items are duplicated on subsequent days.

4.
a. Assess whether plant-specific priorities (Including probablllstlc rlsl< assessment and lndlvldual 1).- r,t-fl?

plant examination Insights} are covered in the appropriate exam sections.

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b. Assess whether the 1 O CFR 55.41, 55.43, and 55.45 sampllng is appropriate.

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c. Ensure that KIA Importance ratings (except for plant-specific priorities) are at least 2.5.

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d.
  • Check for duplication and overlap among exam sections and the last two NRC exams.

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e. Check the entire exam for balance of coverage.

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

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'II' PZ'ted Name/Signature tfl Ji,.. ~

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a. Author Wi j """ ~

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b. Facility Reviewer(*)

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~* NRC's Chief Examiner (#)

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d. NRC Supervisor

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  • Not applicable for NRC-prepared examination outlines.
  1. The independent NRC reviewer Initials Items In column "c"; the chief examiner's concurrence Is required.

ES-201, Page 30 of 32

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

Millstone 3 Date of Examination:

9/9/19 - 9/16/19 Examination Level: RO rgJ SRO D Operating Test Number:

2019 NRC Administrative Topic (see Note)

Type Describe activity to be performed Code*

Determine Reactivity Change for Rod Withdrawal Conduct of Operations D,R KIA 2.1.37 (Knowledge of procedures, guidelines, or RO A.1.1 limitations associated with reactivity management)

KIA Ratina: 4.3 / 4.6 Perform a Shutdown Margin for MODE 3 with Two Stuck Control Rods.

Conduct of Operations N,R KIA 2.1.37 (Knowledge of procedures, guidelines RO A.1.2 associated with reactivity management)

KIA Ratina:4.3 / 4.6 Recommend a Clearance Boundary for 3CCl*P1A.

Equipment Control P,D, R KIA 2.2.13 (Knowledge of tagging and clearance ROA.2 procedures.)

KIA Ratinq: 4.1 / 4.3 Perform Independent Verification Of ORMS Work Station Database Radiation Control N,R KIA 2.3.15 (Knowledge of radiation monitoring ROA.3 systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.)

KIA Ratinq: 2.9 / 3.1 Emergency Plan NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

  • Type Codes and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( ::;3 for ROs; ::;4 for SR Os and RO retakes)

(N)ew or (M)odified from bank ( ~1)

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

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

Millstone 3 Date of Examination:

9/9/19 - 9/16/19 Examination Level: RO D SRO

~

Operating Test Number:

2019 NRC Administrative Topic (see Note)

Type Code*

Describe activity to be performed Check Refueling Admin Requirements Conduct of Operations P,D, R KA: GEN. 2.1.40 (Knowledge of refueling administrative A.1.1 requirements)

KIA Ratinq: 2.8 / 3.9 Review a Shutdown Margin for MODE 3 with Two Stuck Control Rods Conduct of Operations N,R KIA 2.1.37 (Knowledge of procedures, guidelines A.1.2 associated with reactivity management)

KIA Ratinq:4.3 / 4.6 Review a clearance boundary for 3CCl*P1 B.

Equipment Control M,R KIA 2.2.13 (Knowledge of tagging and clearance A.2 procedures.)

KIA Ratinq: 4.1 / 4.3 Determine and Perform Actions Required to Remove a Radiation Monitor from Service Radiation Control M,R KIA 2.3.5 (Ability to use radiation monitoring systems, A.3 such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment etc.)

KIA Rating: 2.9 / 2.9 Emergency Plan Classification and Protective Action Recommendation Emergency Plan N,R KIA 2.4.41 / 2.4.44 (Knowledge of emergency action A.4 level thresholds, classification/ PARs)

KIA Rating: 2.9 / 4.6 ; 2.4 / 4.4 NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

  • Type Codes and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( s;3 for ROs; s;4 for SROs and RO retakes)

(N)ew or (M)odified from bank ( ~1)

(P)revious 2 exams ( s;1, randomly selected)

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

Millstone 3 Date of Examination:

9/9/19 - 9/16/19 Exam Level: RO ~ SRO-I D SRO-U D Operating Test Number:

2019 NRC Control Room Systems: 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Safety Code*

Function

a.

S.1/ Align Safety Grade Boration Path D,S,A 1-004 KIA Number: 004-A4.07 KIA Rating: 3.9/3.7 APE:024 M1.17; KIA Rating: 3.9/3.9

b.

S.2/ Transfer to Cold Leg Recirculation D,S,A 2-006 KIA Number: 006-A4.05; KIA Rating: 3.9/3.8 EPE 011-EA1.11 KIA Rating: 4.2/4.2 C.

S.3/ Respond to a Loss of All Charging Pumps (Overcurrent)

M,S,A 3-006 KIA Number: 004.A4.08 KIA Rating: 3.8/3.4 APE:022 M2.02; KIA Rating: 3.2/3.7

d.

S.4/ Shifting RHR System From Two Loop to Single Loop Operation N,S, L 4.1-005 KIA Number: 005-A4.01 KIA Rating: 3.6 / 3.4

e.

S.5/ Dump Steam Using Atmospheric Relief Valve N,S 4.2-039 KIA Number: 041-A4.06 KIA Rating: 2.9/3.1

f.

S.6/ Respond To An Inadvertent Containment Isolation Phase 'A' D, P, EN, 5-103 KIA Number: 103-A2.03; KIA Rating: 4.3/4.4 s

g.

S.7/ Energizing Bus 34A And 34C After Fault N,S,A 6-062 KIA Number: 062-A2.05; KIA Rating: 2.9/3.3 EPE 055 EA1.07 KIA Rating: 4.3/4.5

h.

S.8/ Respond to RMS-41/42 Alarm N,S, L 7-072 KIA Number: 072-A3.01; KIA Rating: 2.9/3.1 APE 061 M1.01 KIA Rating: 3.6/3.6 In-Plant Systems: 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i.

P.1/ Cross-Connect Service Water To East Switchgear Ventilation D,E,L, P 4.2- 076 KIA Number: 076-K1.19; KIA Rating: 3.6/3.7 APE 068 M1.21 KIA Rating: 3.9/4.1

j.

P.2/ Locally Starting An Emergency Diesel Generator D, E,A 6-064 KIA Number: EPE-055-EA1.024; KIA Rating: 4.3/4.4

k.

P.3/ Establish Feed and Bleed on SI Pump Cooling D,E,L,R 8

KIA Number: APE-056-M1.11; KIA Rating: 3.7/3.7 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for R /SR0-1/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank s9/s8/s4 (E)mergency or abnormal in-plant

~1,~1,~1 (EN)gineered safety feature

~1 / ~1 / ~1 ( control room system)

(L)ow-Power/Shutdown

~1,~1,~1 (N)ew or (M)odified from bank including 1 (A)

~/~/~1 (P)revious 2 exams s.3/ s.3/ s2 (randomly selected)

(R)CA

~1,~1,~1 (S)imulator

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

Millstone 3 Date of Examination:

9/9/19 - 9/16/19 Exam Level: RO D SRO-I ~ SRO-U D Operating Test Number:

2019 NRC Control Room Systems: 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Safety Code*

Function

a.

S.1/ Align Safety Grade Boration Path D,S,A 1-004 KIA Number: 004-A4.07 KIA Rating: 3.9/3.7 APE:024 AA1.17; KIA Rating: 3.9/3.9

b.

S.2/ Transfer to Cold Leg Recirculation D,S,A 2-006 KIA Number: 006-A4.05; KIA Rating: 3.9/3.8 EPE 011-EA1.11 KIA Rating: 4.2/4.2 C.

S.3/ Respond to a Loss of All Charging Pumps (Overcurrent)

M,S,A 3-006 KIA Number: 004.A4.08 KIA Rating: 3.8/3.4 APE:022 AA2.02; KIA Rating: 3.2/3.7

d.
e.

S.5/ Dump Steam Using Atmospheric Relief Valve N,S 4.2-039 KIA Number: 041-A4.06 KIA Rating: 2.9/3.1

f.

S.6/ Respond To An Inadvertent Containment Isolation Phase 'A' D, P, EN, 5-103 KIA Number: 103-A2.03; KIA Rating: 4.3/4.4 s

g.

S.7/ Energizing Bus 34A And 34C After Fault N,S,A 6-062 KIA Number: 062-A2.05; KIA Rating: 2.9/3.3 EPE 055 EA1.07 KIA Rating: 4.3/4.5

h.

S.8/ Respond to RMS-41/42 Alarm N,S, L 7-072 KIA Number: 072-A3.01; KIA Rating: 2.9/3.1 APE 061 AA 1.01 KIA Rating: 3.6/3.6 In-Plant Systems: 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i.

P.1/ Cross-Connect Service Water To East Switchgear Ventilation D,E,L, P 4.2- 076 KIA Number: 076-K1.19; KIA Rating: 3.6/3.7 APE 068 AA1.21 KIA Rating: 3.9/4.1

j.

P.2/ Locally Starting An Emergency Diesel Generator D,E,A 6-064 KIA Number: EPE-055-EA1.024; KIA Rating: 4.3/4.4

k.

P.3/ Establish Feed and Bleed on SI Pump Cooling D, E, L, R 8

KIA Number: APE-056-AA1.11; KIA Rating: 3.7/3.7 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for R /SR0-1/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (O)irect from bank

~9/~8/~

(E)mergency or abnormal in-plant

?
1/;?:1/;?:1 (EN)gineered safety feature
?
1 / ;?:1 / ;?:1 ( control room system)

(L)ow-Power/Shutdown

?
1/;?:1/;?:1 (N)ew or (M)odified from bank including 1 (A)
?
2/ ;?:2/ ;?: 1 (P)revious 2 exams

~3/ ~3/ ~ (randomly selected)

(R)CA

?
1/;?:1/;?:1 (S)imulator

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

Millstone 3 Date of Examination:

9/9/19 - 9/16/19 Exam Level: RO D SRO-I D SRO-U ~ Operating Test Number:

2019 NRC Control Room Systems: 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Safety Code*

Function

a.
b.

C.

S.3/ Respond to a Loss of All Charging Pumps (Overcurrent)

M,S,A 3-006 KIA Number: 004.A4.08 KIA Rating: 3.8/3.4 APE:022 M2.02; KIA Rating: 3.2/3.7

d.
e.
f.

S.6/ Respond To An Inadvertent Containment Isolation Phase 'A' D, P, EN, 5-103 KIA Number: 103-A2.03; KIA Rating: 4.3/4.4 s

g.
h.

S.8/ Respond to RMS-41/42 Alarm N,S, L 7-072 KIA Number: 072-A3.01; KIA Rating: 2.9/3.1 APE 061 M1.01 KIA Rating: 3.6/3.6 In-Plant Systems: 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i.
j.

P.2/ Locally Starting An Emergency Diesel Generator D, E,A 6-064 KIA Number: EPE-055-EA1.024; KIA Rating: 4.3/4.4

k.

P.3/ Establish Feed and Bleed on SI Pump Cooling D, E, L, R 8

KIA Number: APE-056-M1.11; KIA Rating: 3.7/3.7 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for R /SR0-1/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank

~9/~8/~

(E)mergency or abnormal in-plant

?
1/;?:1/;?:1 (EN)gineered safety feature
?
1/;?:1/;?:1 (control room system)

(L)ow-Power/Shutdown

?
1/;?:1/;?:1 (N)ew or (M)odified from bank including 1 (A)
?
2/ ;?:2/ ;?: 1 (P)revious 2 exams

~3/ ~3/ ~ (randomly selected)

(R)CA

?
1 / ;?:1 / ;?:1 (S)imulator

Appendix D Scenario Outline Form ES-D-1 Facility: Millstone 3 Scenario No.: 2K19 NRC-01 (Rev Ol Op-Test No.:

2K19 Examiners:

Operators:

Initial Conditions: IC-13, 100% Power, Beginning of life, Equilibrium Xe Turnover: The plant is at 100% power.. Additionally, Control Rods are in manual for repair of the automatic rod control circuitry (which is not functional due to Tavg / Tref circuit card failure).

'B' MDAFW pp is RTO.

Critical Tasks: 1. Manually trip turbine (CT-13) 2. Bleed and Feed (CT-44)

Event Malf.

Event Event No.

No Type*

Description 1

MS09C US C,TS

'C' Atmospheric Relief Valve Fails Open. (AOP 3581, AOP MS11C BOP C

3582) 2 CC01A US C,TS Loss of 'A' Reactor Plant Component Cooling Water RO C

(RPCCW) Pump requires several actions to avoid a reactor BOP C

trip. (AOP 3581, AOP 3561) 3 FW16C us R

'C' Heater Drain Pump trips requiring a 7% downpower RO R

(Annunciator response).

BOP N

4 FW39 ALL M

Main Feedwater Pipe Break in the Turbine Building FW18A generates a Reactor Trip with a loss of heat sink.

FW19 Complications include a transformer fault de-energizing all ED02B non-safety 4kv and 6kv buses. No Aux feed will present on the Reactor Trip (E-0, FR-H.1 ).

RP08A/ B 5

TC03 BOP C

On the Reactor Trip, the Main Turbine will not trip and the TC06A BOP must close the MSIV's.

TC07B 6

RC08B us C

During Bleed and Feed (FR-H.1 ), one PORV doesn't open RO C

requiring the RO to align head vent letdown.

(N)ormal, (R)eactivity, (I )nstrument, (C)omponent, (M)ajor D-41

SECTION 3 EXAM OVERVIEW Scenario 1

Title:

Loss of Heat Sink - Bleed and Feed The plant is at 100% power and at the beginning of life. Control Rods are in manual for repair of the automatic rod control circuitry (which is not functional due to Tavg / Tref circuit card failure).

Additionally, the 'B' MDAFW Pump is out of service to repair an oil leak.

Event 1: 'C' Atmospheric Relief Valve Fails Open. 3MSS-PV20C fails open and will not manually close. The Balance of Plant operator will perform AOP 3581 immediate actions and close the upstream isolation valve, 3MSS*MV1 BC. The US will enter the appropriate technical specification and perform follow-up actions in AOP 3582, Excessive Steam Demand.

Event 2: 'A' Reactor Plant Component Cooling Water (RPCCW) Pump Trips: The loss of 'A' train Reactor Plant Component Cooling Water (RPCCW) requires the crew to isolate charging and letdown using AOP 3581 Immediate Operator Actions. The crew will then use AOP 3561 Loss of RPCCW to provide cooling to the 'A' & 'D' RCP's, establish pressurizer level control, and re-establish containment cooling.

Event 3: 'C' Heater Drain Pump trips requiring a 7% downpower: The loss of a Heater Drain Pump will cause low suction pressure to the Main Feed Pumps and create an overpower condition (due to lowering feedwater temperature). The crew should use annunciator response procedure for guidance and lower Reactor Power to approximately 93%. Because control rods are in manual, the RO will need to manually insert Control Rods.

Event 4: Main Feedwater Pipe Break in the Turbine Building generates a Reactor Trip concurrent with a loss of heat sink: A large feedwater break develops in the Turbine Building.

As a result, Steam Generator levels will lower uncontrollably and the crew should trip the Reactor. On the trip, no Auxiliary Feedwater Flow (AFW) will be present. Additionally, a 'B' NSST transformer fault occurs on the trip leaving only 4kv emergency-busses 34C & 340 energized {slow transfer is successful). After carrying out actions in E-0, the crew transitions to FR-H.1, Response to Loss of Secondary Heat Sink.

Event 5 - Main Turbine Fails to trip: After the Reactor Trip, the BOP will identify a main turbine stop and control valve fail to close. The BOP will be successful in tripping the turbine by closing the MSIV's (Turbine Trip p/b and load limit pot will be unsuccessful). Auto MSI is failed to promote SG dry-out.

Event 6: During Bleed and Feed (FR-H.1), one PORV doesn't open The crew will mitigate this event by placing head vent letdown in service. After successfully establishing bleed and feed, the crew restores the TDAFW Pump to service and establishes minimum flow to one hot, dry SG.

2. EAL entry: Alert Charlie 1 (BAI-RCS Barrier lost)
3. Duration of Exam:

90 minutes

'K)

Facility: Millstone 3 Scenario No.: 2K19 NRC-02 (RevO)

Op-Test No.:

2K19 Examiners:

Operators:

Initial Conditions: IC-91, 74% Power, Beginning of life, Equilibrium Xe Turnover: The plant is at 7 4% power and at the beginning of life. The 'B' Emergency Diesel is out of service for emergent maintenance.

Additionally, the 'B' Condensate Pump is tagged for motor repair.

Critical Tasks: 1.Energize at least one ac emergency bus (CT-24) 2.Manually actuate SI (CT-2)

3. Establish Aux Feed Flow (CT-4)

Event Malt.

Event Event No.

No Type*

Description 1

RX16A us T/S, I Turbine Impulse pressure instrument (3MSS-PT505) fails RO I

low. (AOP 3581, AOP 3571)

BOP I

2 us R

ISO - NE requested emergency load reduction of 200 RO R

MWe. (AOP 3575, 3% per min).

BOP N

3 RC26 US T/S,C 16 gpm RCS leak (packing leak) to the Containment RO C

Drains Transfer Tank (AOP 3555) 4 RC02C ALL M

Loss of offsite power with a SBLOCA developing on the ED01 us C

Reactor Trip. Concurrent with this, the BOP must start 'A' EG13A&B RO C

EOG to power up a 4kv emergency bus. Additionally, the EG06B BOP C

RO must manually actuate SI.

FW19 FW20A RP07A/B 5

CV23A RO C

While performing actions of E-0, the RO responds to a failure of the 'A' Charging Pump to start on the SI ('B' Charging Pump has no power).

6 FW20A BOP C

While performing actions of E-0, the BOP starts the only FW19 available 'A' AFW Pump. Given the reduced aux feed, a short duration transition to FR-H.1 is needed to verify adequate heat sink.

7 RC26 In E-1, the crew mitigates the SBLOCA concurrent with the RC02C loss of offsite power. A transition is made to ES-1-2, Post LOCA Cooldown and Depressurization.

(N)ormal, ( R )eactivity, (l)nstrument, (C)omponent, (M)ajor 0-41

SECTION 3 EXAM OVERVIEW Scenario 2

Title:

Small Break LOCA with a Loss of offsite Power

1. The plant is at 74% power and at the beginning of life. The 'B' Emergency Diesel is out of service for emergent maintenance. Additionally, the 'B' Condensate Pump is tagged for motor repair.

Event 1: Turbine Impulse pressure instrument (3MSS-PT505) fails low. Control rods will automatically insert. The RO should note that no runback is in progress, respond to the rapid inward rod motion by taking immediate actions in accordance with AOP-3581, Immediate Actions, to place rod control in manual. The US will enter AOP 3581, confirm immediate actions are complete and transition to AOP 3571, Instrument Failure Response.

The crew will select the unaffected channel of turbine impulse pressure, place the Steam Dump system in the steam pressure mode and place rod control back in automatic. AMSAC will be placed in 'Bypass'.

Event 2: ISO - NE requested emergency load reduction of 200 MWe. ISO - NE will direct the crew to begin an Emergency Load Reduction decreasing unit electrical output by 200 MWe. The crew will use AOP 3575, Rapid Downpower to accomplish this down power at 3%/min.

Event 3: 16 gpm RCS leak (packing leak) to the Containment Drains Transfer Tank. This event simulates a 16 gpm packing leakage to the Containment Drains Transfer Tank. The crew will enter AOP 3555, RCS Leak, to stabilize Pressurizer level, identify the source of leakage, and enter the appropriate RCS leakage Tech Spec. Shortly after the crew determines a Containment entry is needed to isolate the leak, the next event will commence.

Event 4: A Loss of offsite power will occur next. There will be a momentary loss of all AC power (as 'A' EOG has an embedded auto start failure and 'B' EOG is tagged OOS). The BOP will mitigate this by starting 'A' EOG from MBB - Critical Task. Additionally, on the Reactor trip, a SBLOCA develops simulating the catastrophic failure of the leaking loop isolation valve's packing gland. The RO will need to identify SI has failed to actuate and manually actuate SI - Critical Task.

Event 5: While performing actions of E-0, the RO responds to a failure of the 'A' Charging Pump to start on the SI ('B' Charging Pump has no power).

Event 6: While performing actions of E-0, the BOP starts the only available 'A' AFW Pump

- Critical Task. Given the reduced aux feed, a short duration transition to FR-H.1 is needed to verify adequate heat sink.

Event 7: In E-1, the crew mitigates the SBLOCA concurrent with the loss of offsite power.

A transition is made to ES-1-2, Post LOCA Cooldown and Depressurization.

2. There are two possible EAL entries (both Alert Charlie 1) 1.) BA 1-RCS Barrier lost OR 2.)

PA2-Single AC Power Source

3. Duration of Exam:

90 minutes

Appendix D Scenario Outline Form ES-D-1 Facility: Millstone 3 Scenario No.: 2K19 NRC-03 (RevO)

Op-Test No.:

2K19 Examiners:

Operators:

Initial Conditions: IC-92, 60% Power, Beginning of life, Equilibrium Xe Turnover: The plant is at 60% power at the beginning of life. 'A' TDMFP has a 1 gpm outboard seal leak. The Motor Driven Feed Water Pump is tagged out to repair an oil leak. Additionally, the Control Rods are in manual for repair of auto circuitry.

Critical Tasks:

1. Insert negative reactivity into the core prior to dispatching operators to locally trip the reactor.

(CT-52)

2. Isolate faulted SG before transition out of E-2 (or FR-S.1) (CT-17)

Event Malf.

Event Event No.

No Type*

Description 1

RX10B us T/S, I Back-up Pressurizer Level Channel fails low causing RO I

letdown to isolate.

2 RX11D us T/S, I 'D' SG pressure channel (3MSS-PT544) fails low requiring BOP I

manual control of 'D' Feed Reg Valve and the Master Speed Controller.

3 CV10A us I

VCT level instrument, CHS*L T112, fails high causing a RO I

letdown divert to Boron Recovery. Annunciator response actions are required.

4 us R

Based upon worsening seal leakage on the 'A' Turbine RO R

Driven Main Feed Pump (TDMFP), the OMOC directs a BOP N

downpower to 45% power at 1 % / min. (AOP 3575) 5 MS02C ALL M

Steam line break in the Main Steam Valve Building RP09NB generates an A TWS (FR-S.1 ). 'C' SG remains faulted RP10NB following a Main Steam Isolation.

6 RP11K BOP C

While performing actions of E-0, the BOP identifies /

us C

corrects the failure of Containment Isolation Phase 'A' to actuate.

7 MS07C In E-2, the crew isolates the faulted 'C' SG.

(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor D-41

Scenario 3

Title:

ATWS SECTION 3 EXAM OVERVIEW

1. The crew takes the shift with the plant at 60% power and at the beginning of core life. 'A' TDMFP has a 1 gpm outboard seal leak. The Motor Driven Feed Water Pump is tagged out to repair an oil leak. Additionally, Control Rods are in manual for repair of the automatic rod control circuitry (which is not functional due to Tavg / Tref circuit card failure).

Event 1: Back-up Pressurizer Level Control Channel Fails Low. The back-up pressurizer level channel fails low causing letdown to isolate. In response to this, the RO will perform his/ her immediate actions and isolate Charging flow. After completing associated actions of AOP 3581, Immediate Operator Actions, the crew will transition to AOP 3571, Instrument Failure Response, to swap backup channels, trip bi-stables, and address Tech Specs.

Additionally, the RO will restore normal Charging/ Letdown iaw GA-13, Establish Normal Charging and Letdown.

Event 2: 'D' SG pressure channel (3MSS-PT544) fails low requiring manual control of 'D' Feed Reg Valve and the Master Speed Controller. The controlling channel of D' SG pressure (3MSS-PT544) will fail low. The BOP will perform immediate actions of AOP 3581, Immediate Actions, which include taking manual control the 'D' Feed Reg Valve and the Feed Pump Master Speed controller. Additionally, AOP 3571, Instrument Failure Response, will be entered to address the failed channel.

Event 3: CHS*LT112 Fails High. VCT level instrument, CHS*LT112, failing high causes a full letdown divert to Boron Recovery. Actual VCT level will lower, auto makeup will not function, and swap over to RWST will not occur at 4.4% VCT level. Annunciator Response Procedure actions will be necessary to correct.

Event 4: Emergent Down power (AOP 3575, 1 %/min) to remove the 'A' TDMFP from service. Based upon worsening seal leakage on the 'A' Turbine Driven Main Feed Pump (TDMFP), the OMOC directs a downpower to 45% power at 1 % / min. Control Rods will remain in manual.

Event 5: Steam line break in the Main Steam Valve Building generates an ATWS (FR-S.1 ).

ATWS / Load Center 32N fails to deenergize (FR-S.1 ). The crew enters FR-S.1, Response to Nuclear Power Generation/ATWS. Control rods are not in automatic (initial conditions),

requiring the RO to insert negative reactivity by manual insertion of control rods [Critical Task]. Ultimately, a PEO will successfully open the reactor breakers locally. 'C' SG remains faulted following a Main Steam Isolation. The crew will proceed through FR-S.1, transition to E-0, Reactor Trip and Safety Injection, and ultimately go to E-2, Faulted SG Isolation.

Event 6: Auto Act Fail: Phase 'A' CIA. While performing actions of E-0, the BOP identifies/

corrects the failure of Containment Isolation Phase 'A' to actuate.

Event 7: Faulted 'C' SG Isolation. While in E-2, the crew recognizes only the 'C' SG is faulted and the BOP isolates this SG [Critical Task].

2. There are two possible EAL entries and they may vary with operator/ plant response to the ATWS:
1. As a minimum, the SRO candidate (US) should classify this event as a Site Area Emergency-Charlie Two due to FR-S.1 being entered directly from E-0 (Criteria: ES1 ). -

OR--

2. A General Emergency - Bravo should be declared if the following conditions were met after entering FR-S.1: SG WR Levels <21% AND Rx Power >5% (Criteria: EG1)
3. Duration of Exam:

90 minutes

2019 NRC JPM Summary Simulator S.1 JPM task is Immediate Boration per step 4 of FR-S.1. JPM is considered an Alternate Path JPM as 3CHS*MV8106 and 3CHS*FCV121 will fail to open. Additional malfunction with the JPM is loss of Bus 34C which also results in using RNO column to align boration. 3CHS*HCV190A is used for the boration flowpath. Developed from JPM S001.

S.2 JPM task is to align for Transfer to Cold Leg Recirculation. JPM is considered Alternate Path JPM as the A RHR pump will not stop at 520,000 gallons in RWST. Additionally, the A train RSS pumps will not automatically start and 3SIH*MV8807B will not open. The JPM includes two TCOAs. Developed from JPM-S022A-3/2K13 NRC S.3.

S.3 JPM task is to respond to a loss of all charging. Immediate actions are performed from memory (AOP 3581 ). A transition is made to AOP 3580. The JPM ends when charging flow and control have been verified. Original JPM number "2K15 Audit S.2".

S.4 JPM task is shifting from two loop RHR operation to single loop operation. Currently, this JPM will be performed simultaneously with JPM S.8. The JPM is a MODE 6 JPM.

S.5 JPM task is to shift from condenser steam dump to Atmospheric Steam Dumps using GA-26 and then to start a cooldown for Immediate Cooldown, using the ADVs. This is a new JPM.

S.6 JPM task is to respond to an inadvertent actuation of the 'A' Train of CIA. This is a modified JPM from JPM 215 (2K15 S.6). This JPM was modified to only one train of CIA vice both trains of CIA (improves realism).

S.7 JPM task is to energize Bus 34A and 34C using GA-3. The cue is to energize both busses through the NSSA. The bus crosstie will not close, resulting in the need to energize Bus 34C from the RSSA. JPM is considered an Alternate Path JPM as the bus cross tie, 34C*1T2 will not close to energize 34C. This is a new JPM.

S.8 JPM task is to respond to radiation monitors RMS-41 and 42, Fuel Drop Monitors, coming into alarm. The response requires action at VP1. This JPM is new in MODE 6. Currently, this JPM will be performed simultaneously with JPM S.4.

In-Plant P.1 JPM is to align Equipment Ventilation for the East Switchgear Room in accordance with EOP 3509.1 Control Room, Cable Spreading Area or Instrument Rack Room Fire. JPM is classified as Previously used (P).

P.2 JPM is to start an EOG locally. Alternate path JPM based on needing to use air start motor valves. JPM is written for administration on the non-protected train.

P.3 JPM is to establish Alternate SI Pump Cooling for the CCI system using EOP 3501, Attachment N. JPM is written for administration on the non-protected train.

RO A.1.1 JPM task is to determine the amount of boric acid or primary water to add to the RCS based on rod withdrawal and xenon change. The JPM uses Cycle 19 Core Data. The JPM was previously used in 2013 (2K13 NRC RO A.1.1 ). There are slight modifications based on new Core Data book and the allowable bands for determining data. It is not significantly modified.

RO A.1.2 JPM task is to perform a SOM for MODE 3 with two stuck rods. The JPM uses Cycle 19 data. The JPM requires Xenon credit calculation be performed. The results of the SOM indicate that an Immediate Boration needs to be performed. This JPM is new.

RO A.2 JPM task is to recommend a clearance boundary for 3CCl*P1A. The JPM has the examinee determine the components to tag. This JPM was used in 2015 (2K15 NRC RO A.2).

RO A.3 JPM task is to perform an Independent Review of the ORMS Database for CCP31. The examinee must identify errors with the database and determine the rad monitor is not ready to be returned to service. It is classified as a new JPM (N).

SRO A.1.1 JPM task is to determine whether or not refueling Tech Spec/TRM admin requirements are met to commence moving fuel. JPM is classified as Previously used (P). This JPM has been modified to focus on Tech Spec/TRM actions (based on 2K17 NRC Exam).

SRO A.1.2 JPM task is to review and then correct a SOM calculation for MODE 3 with two stuck rods.

The JPM uses Cycle 19 data. The SRO will note an error in the calculation and be prompted to correct the calculation. This JPM is considered a Modified JPM (1)11) based on the core actions in RO A.1.2.

SRO A.2 JPM task is to review the clearance boundary for 3CCl*P1A. The SRO will note omissions in the clearance and be prompted to correct the boundary. This JPM includes a T/S entry determination.

SRO A.3 JPM task is to determine actions required to take 3HVR*RE10B, Ventilation Vent Stack monitor out of service. The JPM implements OP 3250.62, Removal and Restoration of Radiation Monitors. The JPM encompasses TRM, REMODCM, and EP-M-303 requirements. This is considered a Modified (M) JPM based on significant changes made.

SRO A.4 JPM task is to classify an event and develop a PAR. The JPM is considered New (N) based on new scenario data.

2

ES-401 PWR Examination Outline Form ES-401-2 Facility: Millstone Unit 3 Date of Exam: 2019 RO KIA Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

1 3

3 3

3 3

3 18 3

3 6

Emergency and Abnormal Plant 2

1 2

2 N/A 1

2 N/A 1

9 2

2 4

Evolutions Tier Totals 4

5 5

4 5

4 27 5

5 10 1

3 2

3 3

1 3

3 2

3 3

2 28 2

3 5

2.

Plant 2

1 1

1 1

1 1

1 1

1 1

0 10 0

1 2

3 Systems Tier Totals 4

3 4

4 2

4 4

3 4

4 2

38 3

5 8

3. Generic Knowledge and Abilities 1

2 3

4 10 1

2 3

4 7

Categories 2

2 3

3 1

2 2

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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the 'Tier Totals" in each KIA category shall not be less than two). (One Tier 3 radiation control KIA is allowed if it is replaced by a KIA from another Tier 3 category.)

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 outline. Systems or evolutions that do not apply at the facility should be deleted with justification. 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 K/As 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) Kl As 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 K/As.

8.

On the following pages, enter the KIA numbers, a brief description of each topic, the topics' 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 a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.

9.

For Tier 3, select topics from Section 2 of the KIA catalog and enter the KIA numbers, descriptions, I Rs, and point totals(#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic KIAs These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the KIA catalog.

These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan.

ES-401 2

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions-Tier 1/Group 1 (RO/SRO)

E/APE # / Name I Safety Function K1 K2 K3 A1 A2 G*

KIA Topic(s)

IR 000007 (EPE 7; BW E02&E10; CE E02) 5 Decay power as a function of time 3.3 1

Reactor Trip, Stabilization, Recovery/ 1 000008 (APE 8) Pressurizer Vapor Space 1

Thermodynamics and flow characteristics of 3.2 2

Accident/ 3 open or leaking valves 000015 (APE 15) Reactor Coolant Pump 9

RCS temperature detection subsystem 3.1 3

Malfunctions / 4 000025 (APE 25) Loss of Residual Heat 2

LPI or Decay Heat Removal/RHR pumps 3.2 4

Removal Svstem / 4 The conditions that will initiate the automatic 000026 (APE 26) Loss of Component 1

opening and closing of the SWS isolation valves 3.2 5

Cooling Water/ 8 to the CCWS coolers 000027 (APE 27) Pressurizer Pressure 1

PZR heaters, sprays, and PORVs 4.0 6

Control Svstem Malfunction / 3 000038 (EPE 38) Steam Generator Tube 9

Criteria for securing/throttling ECCS 4.1 7

Ruoture / 3 000054 (APE 54; CE E06) Loss of Main 2.1.32 Ability to explain and apply all system limits and 3.8 8

Feedwater /4 precautions.

000055 (EPE 55) Station Blackout/ 6 2.1.20 Ability to interoret and execute procedure steos.

4.6 9

000056 /APE 56) Loss of Offsite Power/ 6 18 Control room normal ventilation supply fan 3.2 10 000057 (APE 57) Loss of Vital AC 12 PZR level controller, instrumentation and heater 3.5 11 Instrument Bus/ 6 indications 000058 (APE 58) Loss of DC Power/ 6 1

Battery charqer equipment and instrumentation 2.8 12 000062 (APE 62) Loss of Nuclear Service 2.4.4 Ability to recognize abnormal indications for 4.5 13 Water I 4 system operating parameters which are entry-level conditions for emergency and abnormal operatinq procedures.

000065 (APE 65) Loss of Instrument Air / 8 5

When to commence plant shutdown if instrument 3.4 14 air pressure is decreasing 000077 (APE 77) Generator Voltage and 2

Actions contained in abnormal operating 3.6 15 Electric Grid Disturbances / 6 orocedures for voltage and grid disturbances (W E04) LOCA Outside Containment/ 3 2

Facility's heat removal systems, including 3.8 16 primary coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facilitv.

(W E 11) Loss of Emergency Coolant 1

Facility conditions and selection of appropriate 3.4 17 Recirculation / 4 procedures during abnormal and emergency operations.

(BW E04; W E05) Inadequate Heat 1

Components, and functions of control and safety 3.7 18 Transfer-Loss of Secondary Heat Sink/ 4 systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

000008 (APE 8) Pressurizer Vapor Space 2.4.49 Ability to perform without reference to 4.4 76 procedures those actions that require immediate Accident/ 3 ooeration of svstem comoonents and controls.

000011 (EPE 11) Large Break LOCA / 3 8

Conditions necessary for recovery when accident 3.9 77 reaches stable phase

ES-401 3

Form ES-401-2 000025 (APE 25) Loss of Residual Heat 1

Proper amperage of running LPl/decay heat 2.9 78 Removal Svstem / 4 removal/RHR oumo(sl 000056 (APE 56) Loss of Offsite Power I 6 2.4.45 Ability to prioritize and interpret the significance 4.3 79 of each annunciator or alarm.

000065 (APE 65) Loss of Instrument Air/ 8 4

Typical conditions which could cause a 2.7 80 comoressor trio (e.a. hiah temoeraturel (W E04) LOCA Outside Containment/ 3 2.4.35 Knowledge of local auxiliary operator tasks 4.0 81 during emergency and the resultant operational effects KIA Cateaorv Totals:

3 3

3 3

3/3 3/3 Grouo Point Total:

18/6

ES-401 4

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions-Tier 1/Grouo 2 (RO/SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*

KIA Topic(s)

IR 000001 (APE 1) Continuous Rod Withdrawal / 1 1

Manually driving rods into 3.2 19 position that existed before start of casualty 000036 (APE 36; BW/A08) Fuel-Handling Incidents/ 8 1

ARM system indications 3.2 20 000037 (APE 37) Steam Generator Tube Leak / 3 1

Use of steam tables 2.9 21 000051 (APE 51) Loss of Condenser Vacuum / 4 2.4.47 Ability to diagnose and 4.2 22 recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

AOP 3577, Loss of a 4kV Bus 1

Knowledge of the 3.5 23 interrelations between the Loss of a 4kV bus and following: Major svstem loads 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling/

3 Placing the plant in hot 3.4 24 4

standby status (W E15) Containment Flooding/ 5 1

Components and functions of 2.9 25 control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features.

(BW E08; W E03) LOCA Cooldown-Depressurization / 4 2

Facility's heat removal 3.7 26 systems, including primary coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the ooeration of the facilitv.

(CE A11**; W E08) RCS Overcooling-Pressurized Thermal 2

Adherence to appropriate 3.5 27 Shock/ 4 procedures and operation within the limitations in the facility's license and amendments.

000028 (APE 28) Pressurizer (PZR) Level Control 2.2.12 Knowledge of surveillance 4.1 82 Malfunction / 2 orocedures.

000068 (APE 68; BW A06) Control Room Evacuation / 8 6

RCS Pressure 4.3 83 000076 (APE 76) High Reactor Coolant Activity / 9 2.2.25 Knowledge of the bases in 4.2 84 Technical Specifications for limiting conditions for ooerations and safetv limits.

(BW E09; CE A13**; W E09 & E10) Natural Circulation/4 1

Facility conditions and 3.8 85 selection of appropriate procedures during abnormal and emerqencv operations.

KIA Category Point Totals:

1 2

2 1

2/2 1/2 Group Point Total:

ES-401 5

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant S,stems-Tier 2/Group 1 (RO/SRO)

Svstem # / Name K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G*

KIA Topic(s)

IR 003 (SF4P RCP) Reactor Coolant 1

Seal injection flow 3.3 28 Pumo 004 (SF1; SF2 CVCS) Chemical and 3

Ion exchange bypass 2.9 29 Volume Control 004 (SF1; SF2 CVCS) Chemical and 6

Letdown isolation and flow control valves 3.6 30 Volume Control 005 (SF4P RHR) Residual Heat 3

RHR heat exchanger 2.5 31 Removal 006 (SF2; SF3 ECCS) Emergency 14 Reactor vessel level 3.6 32 Core Coolina 007 (SFS PRTS) Pressurizer 3

Monitoring quench tank temperature 2.6 33 Relief/Quench Tank 007 (SFS PRTS) Pressurizer 1

Containment 3.3 34 Relief/Quench Tank 008 (SF8 CCW) Component Cooling 9

The "standby" feature for the CCW 2.7 35 Water I Pumps 010 (SF3 PZR PCS) Pressurizer 3

PZR sprays and heaters 3.2 36 Pressure Control 010 (SF3 PZR PCS) Pressurizer 4

PRT 2.9 37 Pressure Control 012 (SF7 RPS) Reactor Protection 4

Bistable, trips, reset and test switches 3.3 38 013 (SF2 ESFAS) Engineered 2

Safety system logic and reliability 2.9 39 Safetv Features Actuation 022 (SFS CCS) Containment Cooling Containment readings of temperature, 3.8 40 5

i pressure and humidity system 026 /SFS CSS} Containment Sorav 1

Pump starts and correct MOV oositionina 4.3 41 026 (SFS CSS) Containment Spray 2

Recirculation spray system 4.2 42 039 (SF4S MSS) Main and Reheat 2.1.31 Ability to locate control room switches, 4.6 43 controls and indications and to determine Steam that they are correctly reflecting the desired plant lineup.

039 (SF4S MSS) Main and Reheat 7

Reactor building isolation 3.4 44 Steam 059 (SF4S MFW) Main Feedwater 2

Automatic turbine/reactor trio runback 3.3 45 061 (SF4S AFW) 1 S/G level 3.9 46 Auxiliarv/Emeraencv Feedwater 061 (SF4S AFW) 2 AFW electric drive pumps 3.7 47 Auxiliarv/Emeraencv Feedwater 062 (SF6 ED AC) AC Electrical 1

Major system loads 3.3 48 Distribution 063 (SF6 ED DC) DC Electrical 1

Grounds 2.5 49 Distribution 063 (SF6 ED DC) DC Electrical 3

Battery charger and battery 2.9 50 Distribution Consequences of opening/closing 2.7 51 064 (SF6 EOG) Emergency Diesel 8

breaker between buses (VARS, out-of-Generator phase, voltage)

ES-401 6

Form ES-401-2 073 (SF7 PRM) Process Radiation 2.2.22 Knowledge of limiting conditions for 4.0 52 Monitoring operations and safety limits.

076 (SF4S SW) Service Water 5

Diesel Generator 3.8 53 078 (SF8 IAS) Instrument Air 2

Systems having pneumatic valves and 3.4 54 controls 103 (SF5 CNT) Containment 2

Containment isolation/containment 3.9 55 inteoritv 005 (SF4P RHR) Residual Heat 2

Pressure transient protection during cold 3.7 86 Removal shutdown 012 (SF7 RPS) Reactor Protection 2.4.50 Ability to verify system alarm setpoints 4.0 87 and operate controls identified in the alarm resoonse manual.

022 (SF5 CCS) Containment Cooling 2.2.37 Ability to determine operability and/or 4.6 88 availabilitv of safetv related eauioment.

003 (SF4P RCP) Reactor Coolant 2

Conditions which exist for an abnormal 3.9 89 Pump shutdown of an RCP in comparison to a normal shutdown of an RCP 062 (SF6 ED AC) AC Electrical 2.2.40 Ability to apply Technical Specifications 4.7 90 Distribution for a system.

KIA Cateaorv Point Totals:

3 2

3 3

1 3

3 2/2 3

3 2/3 Group Point Total:

28/5

ES-401 7

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant S istems-Tier 2/Grouo 2 (RO/SRO)

System# I Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

KIA Tooic(s)

IR 001 (SF1 CRDS) Control Rod Drive 11 Determination of SOM 3.5 56 011 (SF2 PZR LCS) Pressurizer 2

PZR heaters 3.1 57 Level Control 015 (SF7 NI) Nuclear 2

CRDS 3.3 58 Instrumentation 016 (SF7 NNI) Nonnuclear 1

Detector failure 3.0 59 Instrumentation 035 (SF 4P SG) Steam Generator 1

MSIVs 3.2 60 041 (SF4S SOS) Steam 2

Steam pressure 3.1 61 Dump/Turbine Bypass Control 071 (SF9 WGS) Waste Gas 4

Relationship of hydrogen/oxygen 2.5 62 Disposal concentrations to flammability 072 (SF7 ARM) Area Radiation 1

Changes in ventilation alignment 2.9 63 Monitorina SBO Diesel 1

Knowledge of the physical connections 3.0 64 and/or cause-effect relationships between the SBO Diesel System and the following system: SBO suooort svstem 086 Fire Protection 6

CO2 3.0 65 034 (SF8 FHS) Fuel-Handling 1

Dropped fuel element 4.4 91 Eauioment 072 (SF7 ARM) Area Radiation 2.1.7 Ability to evaluate plant performance and 4.7 92 Monitoring make operational judgments based on operating characteristics, reactor behavior and instrument interpretation.

079 (SF8 SAS**) Station Air 2.2.44 Ability to interpret control room 4.4 93 indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions KIA Cateaorv Point Totals:

1 1

1 1

1 1

1 1/1 1 1

0/2 Group Point Total:

10/3

ES-401 Generic Knowledge and Abilities Outline {Tier 3)

Form ES-401-3 Facility:

Date of Exam:

Category KIA#

Topic RO SRO-onlv IR IR 2.1.39 Knowledae of conservative decision makina practices.

3.6 66 2.1.43 Ability to use procedures to determine the effects on 4.1 67 reactivity of plant chanaes.

1. Conduct of Operations 2.1.3 Knowledge of shift or short term relief turnover 3.9 94 practices.

Subtotal 2

1 2.2.17 Knowledge of the process for managing maintenance 2.6 68 activities durinq power operations.

2.2.18 Knowledge of the process for managing maintenance 2.6 69 activities durinq shutdown operations.

2. Equipment Control 2.2.1 Ability to perform pre-startup procedures for the facility, 4.4 95 including operating those controls associated with plant equipment that could affect reactivitv.

2.2.15 Ability to determine the expected plant configuration 4.3 96 usinq desian and confiauration control documentation.

Subtotal 2

2 2.3.12 Knowledge of radiological safety principles pertaining to 3.2 70 licensed operator duties.

2.3.13 Knowledge of radiological safety procedures pertaining 3.4 71 to licensed operator duties.

2.3.4 Knowledge of radiation exposure limits under normal 3.2 72

3. Radiation and emerqencv conditions.

Control 2.3.11 Ability to control radiation releases.

4.3 97 2.3.7 Ability to comply with radiation work permit requirements 3.6 98 durina normal or abnormal conditions.

Subtotal 3

2 2.4.11 Knowledge of abnormal condition procedures.

4.0 73 2.4.31 Knowledge of annunciators alarms, indications or 4.2 74 response procedures.

2.4.46 Ability to verify that the alarms are consistent with the 4.2 75 plant conditions.

4. Emergency Procedures/Plan 2.4.35 Knowledge of local auxiliary operator tasks during 4.0 99 emergency and the resultant operational effects 2.4.38 Ability to take actions called for in the facility emergency 4.4 100 plan, including supporting or acting as emergency coordinator.

Subtotal 3

2 Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier/

Randomly Reason for Rejection Group Selected K/A 1/2 Rejected Emergency and Abnormal Plant Evolutions associated with Babcock and Wilcox (BW) and Combustion Engineering (CE) reactors - Millstone Unit 3 is a WestinQhouse desiQn.

2/1 Rejected 025 (SF5 ICE) Ice Condenser - Millstone Unit 3 does not have an ice condenser installed.

2/1 Rejected 053 (SF1; SF4P ICS*) Integrated Control - Millstone Unit 3 does not have ICS. This topic has no operational significance but similar topics are covered in MSRs.

1/2 RO Question #23: The licensee requested to add 6 items to the Tier 1, Group 2 list of E/APEs to be available for random selection in accordance with ES-401 Attachment 1. AOP 3577, "Loss of a 4kV Bus," was randomly selected for Question 23.

2/2 RO Question #64: The licensee added 2 items to the Tier 2, Group 2 list of Plant Systems to be available for random selection in accordance with ES-401 Attachment 1. "SBO Diesel," was randomly selected for Question 64.

ALL Rejected generics 2.2.3 and 2.2.4 as Millstone 3 is a single unit site

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