ML031710462

Draft - Outlines

ML031710462
Person / Time
Site:	Crane
Issue date:	05/14/2003
From:	Gumbert R AmerGen Energy Co
To:	Conte R NRC/RGN-I/DRS/OSB
Conte R
References
50-289/03-301 50-289/03-301
Download: ML031710462 (20)

Text

ES-401 PWR SRO Examination Outline Printed: 02/24/2003 Facility:

Three Mile Island - 1 Form ES-40 1-3 Exam Date: 05/12/2003 Exam Level: SRO IUA Category Points Note: I. Ensure that at least two topics from every IUA category are sampled within each teir (Le., the "Tier Totals" in each K/A category shall not be less than two).

2. Actual point totals must match those specified in the table.

3. Selecttopics from many systems; avoid selecting more than two or three K/A topics from a given system unless

4. Systems/evolutions within each group are identified on the associated outline.

5. The shaded areas are not applicable to the categoryltier.

6. The generic WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

7. On the following pages, enter the IUA numbers, a brief description of each topic, the topics' importance ratings for the RO license level, and the point totals for each system and category. WAS below 2.5 should be justified on the basis of plant-specific priorites. Enter the tier totals for each category in the table above.

they relate to plant-specific priorities.

1

Facility: 1 I,.,e Mile Island - 1 KA Topic AA2.04 - Reactor power and its trend AK2.05 - Control rod drive power supplies and logic circuits ES - 401 Imp.

4.3 2.8 EIAPE #

00 1 003 003 005 01 1 01 1 015 017 026 026 K1 X

X Emei K2 X

EIAPE Name I Safety Function Continuous Rod Withdrawal / 1 AA1.06 - RCS pressure and temperature Dropped Control Rod / 1 Dropped Control Rod / 1 4.1 InoperablelStuck Control Rod / 1 AKl.01 - Axial power imbalance Large Break LOCA / 3 3.8 Large Break LOCA 13 2.4.6 - Knowledge symptom based EOP mitigation strategies.

AA2.10 - When to secure RCPs on loss of cooling or seal injection AK3.03 - Guidance actions contained in EOP for Loss of ccw AA 1.07 - Flow rates to the components and systems that are serviced by the CCWS; interactions among the Reactor Coolant Pump (RCP) Malhnctions / 4 4.0 3.7 4.2 3.0 Reactor Coolant Pump (RCP) Malfimctions (Loss of RC Flow) / 4 Loss of Component Cooling Water (CCW) / 8 Loss of Component Cooling Water (CCW) / 8 PWR SR(

amination Outline ency and Abnormal Plant Evolutions - Tier 1 I Group 1 42 X -

Printed:

02/2.(

Form ES-401-3 2.1.33 - Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

EK 1.O 1 - Natural circulation and cooling, including reflux boiling Points 1

1 1

1 1

1 1

I 1

1 I

Facility: 6

--i Mile Island - 1 ES - 401 E/APE #

029 029 05 1 055 067 069 069 074 Emergenc Anticipated Transient Without Scram (ATWS) I I 6

Plant Fire on Site 1 9 Loss of Containment Integrity / 5 Loss of Containment Integrity / 5 X

Inadequate Core Cooling 14 PWR SRQ amination Outline ind Abnormal Plant Evolutions - Tier 1 / G r o w 1 Printed:

02/24 Form ES-401-3 EK2.06 - Breakers, relays, and disconnects EA1.05 - BIT outlet valve switches 3.6*

AK3.01 - Loss of steam dump capability upon loss of condenser vacuum 3.1*

EK3.01 - Length of time for which battery capacity is 3.4 designed 2.1.32 - Ability to explain and apply all system limits and precautions.

3.8 I

AA2.01 - Loss of containment integrity 4.3 AK 1.O 1 - Effect of pressure on leak rate 3.1 EA2.06 - Changes in PZR level due to PZR steam bubble transfer to the RCS during inadequate core cooring 4.6 1 -

2

ES - 401 EIAPE #

A03 A03 A06 E05 E09 E09 PWR SR(

amination Outline Emergencv and Abnormal Plant Evolutions - Tier 1 I Grow 1 A2 EIAPE Name I Safety Function LOSS of NNI-Y / 7 G

X Loss of "I-Y

/ 7

~ KA Topic primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 1 AK2.2 - Facility's heat removal systems, including AAI. 1 - Components, and functions of control and AK 1.3 - Annunciators and conditions indicating signals, and remedial actions associated with the (Shutdown Ouside Control Room)

EK2.1 - Components, and fimctions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 2.4.30 - Knowledge of which events related to system EK3.2 - Normal, abnormal and emergency operating operations/status should be reported to outside agencies.

procedures associated with (Natural Circulation uperations)

Shutdown Outside Control Room / 8 Excessive Heat Transfer / 4 Imp.

Points 3.3 1

4.0 1

3.4 1

4.0 1

3.6 1

3.8 1

Natural Circulation Operations / 4 Natural Circulation Operations / 4 Printed:

02124 Form ES-401-3 WACategory Totals:

4 4 4 4 4 4 Group Point Total:

24 3

PWR SR(

amination Outline Mile Island - 1 A2 ES - 401 G KATopic Imp.

Points EK2.03 - Reactor trip status panel 3.6 1

(/APE #

007 007 008 008 009 009 022 027 033 E/APE Name / Safety Function Reactor Trip I 1 I

I I

I K1 K2 K3 A1 X

EA1.03 - RCS pressure and temperature AK2.01 - Valves AK3.03 - Actions contained in EOP for PZR vapor space accident/LOCA Reactor Trip I 1 4.1 1

2.7 1

4.6 1

X X

Pressurizer(PZR) Vapor Space Accident (Relief Valve Stuck Open) I 3 2.2.25 - Knowledge of bases in technical specifications 3.7 1

for limiting conditions for operations and safety limits.

Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) 13 Loss of Reactor Coolant Makeup 12 X

X Small Break LOCA 13 AKl.02 - Relationship of charging flow to pressure differential between charging and RCS Small Break LOCA 13 3.1 1

X Pressurizer Pressure Control (PZR PCS) Malfunction X I-AK2.03 - Controllers and positioners AK3.01 - Termination of startup following loss of intermediate-range instrumentation 2.8 1

3.6 1

I I I I I

I 1

I Loss of lntermediate Range Nuclear lnstrumentation I I I 1 x 1 7

I I

I I

Printed:

0212(

Form ES-401-3 I

I I

I I

I EKI.O 1 - Natural circulation and cooling, including I 4.7 I 1

reflux boiling I I I

I 1

Facility: 1 lhWe Mile Island - 1 ES - 401 EIAPE #

03 8 03 8 06 1 06 1 A0 1 E08 E08 Emer E/APE Name / Safety Function Steam Generator Tube Rupture (SGTR) / 3 Steam Generator Tube Rupture (SGTR) / 3 Area Radiation Monitoring (ARM) System Alarms I 7 Area Radiation Monitoring (ARM) System Alarms / 7 Plant Runback / 1 LOCA Cooldown / 4 LOCA Cooldown 14 PWR SR(

amination Outline Printed:

02/2(

ncy and Abnormal Plant Evolutions - Tier 1 I Group 2 Form ES-401-3 KIA Category Totals:

3 3

2 3

KA Tonic

~- _ _

EA1.OS - Core cooling monitor EA2.09 - Existence of natural circulation, using plant parameters AA2.0 1 - ARM panel displays 2.1.32 - Ability to explain and apply all system limits and precautions.

AA 1.1 - Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features EA2.2 - Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments EKI.3 - Annunciators and conditions indicating signals, and remedial actions associated with the (LOCA coolaown) mp.

3.8* -

4.2 3.7 -

3.8 3.7 4.0 -

3.5 -

2 Group Point Total:

Points 1

1 1

1 16 2

PWR SR{

amination Outline Facility: 1; 1b,e Mile Island - 1 E/APE #

E/APE Name / Safety Function K1 K2 K3 A1 A2 G A08 Refuel Canal Level Decrease / 8 X

E13 EOP Rules X

E13 EOP Rules X

Printed:

02/2(

3 KATopic Imp.

Points procedures during abnormal and emergency operations AA2.1 - Facility conditions and selection of appropriate 4.0 1

2.2.22 - Knowledge of limiting conditions for operations 4.1 1

EKl.2 - Normal, abnormal and emergency operating 3.6 1

and safety limits.

procedures associated with (EOP Rules)

WACategory Totals:

1 0

0 0

1 1

Group Point Total:

3 1

Facility:

K3 K4 X

X

?S - 401 SysIEv #

00 1 003 003 004 004 015 01 5 022 026 026 06 1 K5 X

\\

'I lllae Mile Island - 1 rier A4 X

System / Evolution Name Control Rod Drive System / 1 2 I G

X Reactor Coolant Pump System (RCPS) / 4 Reactor Coolant Pump System (RCPS) / 4 K1 X

Chemical and Volume Control System (CVCS) / 1 K2 X

Chemical and Volume Control System (CVCS) / 1 Nuclear Instrumentation System / 7 Nuclear Instrumentation System / 7 Containment Cooling System (CCS) /

5 Containment Spray System (CSS) / 5 Containment Spray System (CSS) / 5 Auxiliary / Emergency Feedwater (AFW) System / 4 PWR SRO (

nination Outline Printed: 021 33

mun 1 KA Topic K6.03 - Reactor trip breakers, including controls K3.03 - Feedwater and emergency feedwater A2.02 - Conditions which exist for an abnormal shutdown of an RCP in comparison to a normal shutdown of an RCP 2.4.4 - Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

A4.18 - Emergency borate valve K4.04 - Slow response time of SPNDs K6.04 - Bistables and logic circuits K2.01 - Containment cooling fans K1.O1 - ECCS A3.01 - Pump starts and correct MOV positioning K5.02 - Decay heat sources and magnitude Form Imp.

4.2 -

3.1 -

3.9 4.3 4.1 3.6? -

3.2 -

3.1 4.2 -

4.5 -

3.6 -

\\

S-401-Points 1 -

1 1

1 1

1 1

1 1 -

1 1

1

Facility:

KA Topic K6.01 - Controllers and positioners ES - 401 SysIEv #

06 1 063 063 068 068 07 1 072 072 Imp.

Points 2.8*

1 I tllee Mile Island - 1 K1.07 - Sources of liquid wastes for LRS Svstem I Evolution Name IK1 IK2 I K

~

I K

~

2.9 1

Auxiliary / Emergency Feedwater (AFW) System 14 A3.02 - Automatic isolation D.C. Electrical Distribution System 16 I 3.6 1

I I

I I

Liquid Radwaste System (LRS) 19 I X I I I A2.02 - Use of waste gas release monitors, radiation, gas flow rate, and totalizer Liquid Radwaste System (LRS) 19 3.6 1

Waste Gas Disposal System (WGDS) 19 K3.02 - Fuel handling operations Area Radiation Monitoring (ARM) 3.5 1

KIA Category Totals:

2 1

2 2

41.01 - Radiation levels PWR SRO (

qination Outline Printed: 02(

33 3.6 1

X X

roup 1 Form ES-401-2 I

I 2.1.14 - Knowledge of system status criteria which require the notification of plant K4.04 - Trips 3

1 2

2 1

2 Group Point Total: 19 2

Facility:

KA Topic K3.02 - Fuel ES - 401 Imp.

4.5 002 K5.18 - Brittle fracture 01 1 3.6 012 012 A2.08 - Loss of level compensation 029 2.8 033 K4.05 - Spurious trip protection K6.11 - Trip setpoint calculators 034 2.9 2.9 035 K1.01 - Gaseous radiation release monitors 039 039 3.7

'I llree Mile Island - 1 A3.O 1 - Temperature control valves A2.02 - Dropped cask 2.4.49 - Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

2.4.6 - Knowledge symptom based EOP System / Evolution Name Reactor Coolant System (RCS) / 2 2.7*

3.9 4.0 4.0 Reactor Coolant System (RCS) / 2 A4.07 - Steam dump valves Pressurizer Level Control System (PZR LCS) / 2 2.9 Reactor Protection System / 7 Reactor Protection System / 7 Containment Purge System (CPS) / 8 Spent Fuel Pool Cooling System (SFPCS) / 8 Fuel Handling Equipment System (FHES) / 8 Steam Generator System (S/GS) / 4 Main and Reheat Steam System (MRSS) / 4 Main and Reheat Steam System (MRSS) / 4 PWR SRO (

nination Outline Printed:

02(

'03

roup 2 Form I

I mitigation strategies.

I S-401-Points 1

1 1

1 1

Facility:

(2 X

r; K3 X

3s - 401 Sys/Ev #

055 064 073 073 075 103

'I uree Mile Island - 1 System / Evolution Name Condenser Air Removal System (CARS) / 4 Emergency Diesel Generator (EDIG)

System 16 Process Radiation Monitoring (PRM)

System I 7 Process Radiation Monitoring (PRM)

System I 7 Circulating Water System I 8 Containment System / 5 WA Category Totals:

1 1

2 PWR SRO (

nination Outline Printed: 02(

03 KA Topic Imp.

A3.03 - Automatic diversion of CARS exhaust 2.7*

K2.03 - Control power 3.6 K4.01 - Release termination when radiation exceeds setpoint 4.3 I

A1.O1 - Radiation levels I

3.5 K3.07 - ESFAS 3.5*

2.4.30 - Knowledge of which events related to 3.6 system oDerations7status should be reported to I S-401-2 Points 1

2 1

3 Group Point Total: 17 2

Facility:

K3 SS - 401

$ys/Ev #

005 005 007 008 K4 K5 X

'I lllee Mile Island - 1 KA Topic K6.03 - RHR heat exchanger System / Evolution Name Residual Heat Removal System (RHRS) 14 Residual Heat Removal System (RHRS) 14 Imp.

2.6 Pressurizer Relief TanWQuench Tank System (PRTS) / 5 A 1.O 1 - Heatup/cooldown rates Component Cooling Water System (CCWS) 1 8 3.6 WA Category Totals:

K4.0 1 - Quench tank cooling PWR SRO (

2.9 2.2.22 - Knowledge of limiting conditions for operations and safety limits.

Plant 1

1 4.1 nination Outline Printed: 02(

'03 rier 2 /

1 0

1

roup 3 Form 1

S-401-.

Points 1 -

1 -

1 1

Group Point Total:

4 1

Facilitv:

Three Mile Island - 1 Generic Category Conduct of Operations 2.1.5 Equipment Control Ability to locate and use procedures and directives related to shift staffing and activities.

Generic Knowledger-4 Abilities Outline (Tier 3) 2.1.34 PWR SRO Examination Outline Ability to maintain primary and secondary plant chemistry within allowable limits.

Printed: 02124120(

Form ES-401-5 Knowledge of 10 CFR: 20 and related facility radiation control requirements.

KA KA Topic 3.0 1

2.1.7 2.1.10 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Knowledge of conditions and limitations in the facility license.

2.2.1 2.2.1 1 2.2.19 2.2.26 2.2.27 Radiation Control I

2.3.1 2.3.3 2.3.8 I 2.3.10 Imp.

Points 3.4 4.4 3.9 2.9 i

1 1

1 Category Total:

4 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

Knowledge of the process for controlling temporary changes.

Knowledge of maintenance work order requirements.

Knowledge of refueling administrative requirements.

Knowledge of the refueling process.

3.6 3.4*

3.1 3.7 3.5 1

1 1

1 1

Knowledge of SRO responsibilities for auxiliary systems that are outside the control room (e.g., waste disposal and handling systems).

Knowledge of the process for performing a planned gaseous radioactive release.

2.9 3.2 Ability to perform procedures to reduce excessive levels of radiation and guard against oersonnel exuosure.

Category Total:

4 1

Generic Knowledge iy4 Abilities Outline (Tier 3)

\\

PWR SRO Examination Outline Emergency ProceduredPlan 2.4.10 Knowledge of annunciator response procedures.

2.4.1 1 2.4.33 2.4.44 Knowledge of abnormal condition procedures.

Knowledge of the process used track inoperable alarms.

Knowledge of emergency plan protective action recommendations.

Printed: 02,241204 3.1 1

3.6 1

2.8 1

4.0 1

CategoryTotal:

4 Generic Total: 17 2

ES-401 Record of Rejected K/As Form ES-401-10 035 2.4.49 I ier / Group 2/ 1 This K/A needs t o be suppressed, since there are no "emergency essential" SWS Pumps associated with the Circulating Water System a t TMI.

212 I

11 R X, C ~ O I I ~ I Y Setecled WA I Reason for Rejection 061 A1.03 I This K/A needs t o be supressed, since i t applies t o a m u l t i - u n i t f a l i l i t y.

I Tier 2/Group 1 K/A 061 K6.01 was randomly selected as a replacement.

II NUREG-1021, Revision 8, Supplement 1 46 of 46

ES-301 Administrative Topics Out1 ine Form ES-301-1 A. 1 Facility: Three Mile Island Unit 1 Examination Level (circle one): RO / SRO Date of Examination: May 12, 2003 Operating Test Number:

Plant Parameter Ve r if i cat i o n Administrative Topic/Su bject Description A.2 Use Of Station Drawings Shift Staffing Requirements A. 3 Radiation Release A.4 Emergency Classification Describe method of evaluation:

1. ONE Administrative JPM, OR

2. TWO Administrative Questions.

Perform Estimated Critical Boron Concentration Calculation. (JPM)

Minimum Shift Staffing, Control of Overtime. (JPM)

Predict Operational Impact Of Instrument Failure. (JPM)

Liquid Radiation Release Approval. (JPM)

(Organ Dose prevents approval)

Classify Event And Complete Initial Notification Forms.

(JPM)

5-301 Facility: Three Mile Island Unit 1 Exam Level (circle one): RO / SRO(I) / SRO(U)

Control Room Systems and Facility Walk-Through Test Outline Form ES-301-:

Date of Examination: Mav 12.2003 Operating Test No.:

b. Engineered Safety Feature Actuation Systems (01 3)IRespond to inadvertent ES Actuation.

B.l Control Room Systems D, s System/JPM Title

c. Emergency Core Cooling System (OOG)/Respond to a High Pressure Injection (HPI) initiation (Alt. Path - MU-V-14A fails).

Type Code*

N, A, S Safety Function

d. Residual Heat Removal System (OOS)/Respond to a failure of Low Pressure Injection (Alt. Path - DHV-6 Fails to Open).

a. Chemical and Volume Control (OO4)/Perform an Emergency Boration (Alt. Path - Backup Emergency Boration Required).

N, A, S, L 1

e. Main Steam System (039)/Respond to inadvertent closure of a Main Steam Isolation Valve.

f.

Containment Cooling System (022)/Return Reactor Building (RB)

Emergency Cooling to Engineered Safeguards Standby.

g. Emergency Diesel Generator (EDG) System (064)/EDG Operation (Alt. Path - EDG Fails to Auto Load).

D, s N, s D, A, S 2

a. Chemical and Volume Control System (004)/Manually Open RCP Seal Injection Isolation Valve (MU-V-26).

N, R 3

4 Primary 8.2 Facility Walk-Through 4 Secondary 5

6 2

Emergency

~

~

~

~

~

~

~

~~

~~

~

b. Pressurizer Pressure Control System (01 0)TTransfer Pressurizer Heater Group 8 or 9 to an Engineered Safeguards Bus.

D l

3 1

Emergency

c. Emergency Feedwater System (061)ILocal Reset of Emergency Feedwater Pump (EF-P-1 ).

D 4 Secondary Emergency

• Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA.

Scenario Outline Simulation Three Mile Island Scenario No.:

1. I Op Test No.:

Facility:

Unit 1 Event No.

1 2

3 4

5 Examiners Malf.

Event Event No.

Type*

Description URO ARO Secure feedwater pump (FW-P-1A).

N us RDOI 17 C

URO Dropped Control Rod in controlling group.

ARO c

us R

URO Manual power reduction due to dropped rod.

ARO us I/O OVERRIDE I

URO Pressurizer level controller fails to 0% demand, closing makeup control ARO valve (MU-V-17).

I us M

ARO NOTES: Rapid control oil leak, auto start of oil pump, MAP - dont FWl5B M

URO Loss of feedwater due to feedwater pump (FW-P-1 B) failure.

Operators CRS 6

7 URO ARO ATWS - RPS Auto Trip Failure.

I us M

ARO MU16A M

URO Small Break LOCA, Loss of Subcooling Margin PRO 8

Description Evaluate the ability of the crew to perform normal operations (secure FWP) and forced power reduction (due to dropped control rod) with ICs in manual. Following the power reduction, a controller failure requires the operator to implement manual Pressurizer level control. When the only operating Feedwater Pump trips, the Main Turbine trips, but the reactor does not (ATWS). The operator is required initiate a manual reactor trip. Following reactor trip, an RCS piping break results in loss of Reactor Coolant, and saturated liquid conditions. The overall scenario provides the opportunity to demonstrate ability to utilize normal, emergency, and accident mitigation procedures, and compliance with Technical Specifications requirements.

ICs in manual due to SG-Rx Master controller failure (Malfunction IC23 OTSG Reactor Master output fails to zero volts). Plant is at 68% power, ready to secure FW-P-1A to enable coupling repair.

Initial Conditions M

US MU23C C

URO High pressure injection pump (MU-P-IC) fails to automatically start.

ARO c

us I M US I want,manualRxTrip.

I RD28 I I URO I

Scenario Outline Simulation Facility:

Three Mile Island Scenario No.:

1. 2 Op Test No.:

Event No.

I 2

3 4

5 Examiners Description Initial Conditions Turnover Malf.

Event Event No.

Type*

Description URO Switch operating Condensate Pumps.

N ARO N

US N127B I

URO RCS Loop A T-Hot transmitter failure (high), affecting ICs T-Ave ARO indication.

I us C

URO Main Generator hydrogen gas leak.

C ARO c

us R

URO ARO us Manual load reduction to 800 MW due to ICs controller failures.

ED02A C

URO 1A Auxiliary Transformer fault (Technical Specifications).

C ARO Unit 1 Operators CRS URO PRO 6

7 This scenario provides operational situations to evaluate the ability of crew members to implement plant procedures to perform normal operations (switch Condensate Pumps) and to respond abnormal and emergency conditions. While operating at full power, a control system instrumentation failure upsets the balance of plant control. After re-establishing plant stability, a hydrogen gas leak lowers Main Generator gas pressure, requiring a forced load reduction to protect the generator. Because of an Integrated Control System malfunction the load reduction must be performed manually. Following the load reduction, protective relay operation transfers loads off the 1A Auxiliary Transformer to 1 B Auxiliary Transformer and Emergency Generator EG-Y-I B. A major steam line rupture inside the Containment Building causes the reactor to be tripped. Excessive OTSG heat transfer results in a core overcooling event, and ESAS actuation.

Following isolation of feedwater sources to the affected OTSG, crew members are required take actions to prevent RCS reheat and re-pressurization. RCS pressure and temperature are required to be stabilized to protect OTSG and RCS components from excessive stresses that could lead to material failure and fission product release. Control and termination of HPI flow is complicated by a stuck open injection valve.

Plant is at 100% power, with ICs if full automatic. Dispatcher ordered +200 MVAR k20. Ready to switch Condensate Pumps.

See Attached "Shift Turnover" Sheet.

c us M

ARO M

US MS02B M

URO Main Steam Line Rupture Inside the RB with ESAS actuation.

MU08C C

URO Stuck open high pressure injection valve.

ARO c

us

Scenario Outline 1

%mulation Facility:

Three Mile Island Scenario No.:

Alternate Op Test No.:

Unit 1 Event Type*

I URO ARO I

us URO N

ARO N

US C

URO ARO c

us C

URO ARO c

us URO ARO us I/R URO ARO I

us M

URO ARO M

US I

URO ARO I

us wment, (C)omponent, Examiners Description Event Description Pressurizer level instrument failure.

Switch operating Makeup Pumps.

MU-V-32, RCP seal injection valve, failure.

Small OTSG 1A tube leak.

Initiation of plant shutdown.

Control Rod Drive System automatic control failure.

OTSG tube rupture.

ES Train failure.

(M)ajor Initial Conditions

-Turnover Operators CRS URO PRO This scenario provides operational situations to evaluate the ability of crew members to implement plant procedures to perform normal operations, and to respond to abnormal and emergency conditions. While operating at full power, the operators are required to mitigate the effects of a controlling instrument failure by establishing manual control and then selecting alternate (valid) input signals. After re-establishing automatic RCS inventory control, the operators implement normal operating procedures to switch operating Makeup Pumps. An additional control system malfunction requires the operators to establish manual flow control for RCP seal injection. A small OTSG tube leak (greater than Technical Specification limits) forces the operators to implement an emergency operating procedure that includes plant shutdown. The power reduction is performed in manual due to an automatic control problem in the Control Rod Drive System. During the shutdown, a large OTSG Tube rupture develops, requiring the operators to initiate High pressure Injection and trip the reactor. One of the two ES Trains will not actuate automatically or manually at the Train level, requiring the operator is to initiate individual components. Following reactor trip, actions are performed to ensure the reactor is shutdown properly, establish radiological controls and isolate potential secondary release paths, prevent inadvertent operation of the Main Steam Safety Valves, and reduce RCS leakage through the OTSG tubes in order to limit off-site doses.

Plant is at 100% power, with ICs if full automatic. EF-P-2A is out of service for bearing replacement. MU-P-1A is operating, cooled by NSCC, to support MU-P-1 B oil change.

See Attached "Shift Turnover" Sheet.

1)

Event No. I Malf.

l 5

II I

I II TH16A 8

I ES02B I

10 Override (N)ormal, (R)eactivity, (I)n