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 they relate to plant-specific priorities.

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' importanceratings 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.

1

Facility: 1I,. ,e Mile Island - 1 PWR SR( amination Outline Printed: 02/2.(

ES - 401 - -

Emei ency and Abnormal Plant Evolutions Tier 1 I Group 1 Form ES-401-3 EIAPE #

00 1 EIAPE Name I Safety Function Continuous Rod Withdrawal / 1 K1 K2 -

42 X

KA Topic AA2.04 - Reactor power and its trend Imp.

4.3 Points 1

003 Dropped Control Rod / 1 X AK2.05 - Control rod drive power supplies and logic 2.8 1 circuits 003 Dropped Control Rod / 1 AA1.06 - RCS pressure and temperature 4.1 1 005 InoperablelStuck Control Rod / 1 X AKl.01 - Axial power imbalance 3.8 1 01 1 Large Break LOCA / 3 2.1.33 - Ability to recognize indications for system 1 operating parameters which are entry-level conditions for technical specifications.

01 1 Large Break LOCA 1 3 X EK 1.O1 - Natural circulation and cooling, including 1 reflux boiling 015 Reactor Coolant Pump (RCP) Malhnctions / 4 2.4.6 - Knowledge symptom based EOP mitigation 4.0 1 strategies.

017 Reactor Coolant Pump (RCP) Malfimctions (Loss of AA2.10 - When to secure RCPs on loss of cooling or 3.7 I RC Flow) / 4 seal injection 026 Loss of Component Cooling Water (CCW) / 8 AK3.03 - Guidance actions contained in EOP for Loss 4.2 1 of ccw 026 Loss of Component Cooling Water (CCW) / 8 -

AA 1.07 Flow rates to the components and systems that 3.0 1 are serviced by the CCWS; interactions among the I

Facility: 6 --iMile Island 1-PWR SRQ amination Outline Printed: 02/24 ES 401 Emergenc ind Abnormal Plant Evolutions - Tier 1 / G r o w 1 Form ES-401-3 E/APE #

029 EK2.06 - Breakers, relays, and disconnects 029 Anticipated Transient Without Scram (ATWS) I I EA1.05 - BIT outlet valve switches 3.6*

05 1 AK3.01 - Loss of steam dump capability upon loss of 3.1*

condenser vacuum 055 EK3.01 - Length of time for which battery capacity is 3.4 6 designed 067 Plant Fire on Site 1 9 2.1.32 - Ability to explain and apply all system limits 3.8 and precautions.

I 069 Loss of Containment Integrity / 5 AA2.01 - Loss of containment integrity 4.3 069 Loss of Containment Integrity / 5 X AK 1 .O 1 - Effect of pressure on leak rate 3.1 074 Inadequate Core Cooling 1 4 EA2.06 - Changes in PZR level due to PZR steam 4.6 1 bubble transfer to the RCS during inadequate core cooring 2

PWR SR( amination Outline Printed: 02124 ES - 401 Emergencv and Abnormal Plant Evolutions - Tier 1 I G r o w 1 Form ES-401-3 EIAPE # EIAPE Name I Safety Function A2 G KA Topic Imp. Points 1

~

A03 LOSSof NNI-Y / 7 AK2.2 - Facility's heat removal systems, including 3.3 1 primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility A03 Loss of "I-Y /7 AAI .1 - Components, and functions of control and 4.0 1 safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features A06 Shutdown Outside Control Room / 8 -

AK 1.3 Annunciators and conditions indicating signals, 3.4 1 and remedial actions associated with the (Shutdown Ouside Control Room)

E05 Excessive Heat Transfer / 4 EK2.1 - Components, and fimctions of control and 4.0 1 safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features E09 Natural Circulation Operations / 4 -

X 2.4.30 Knowledge of which events related to system 3.6 1 operations/status should be reported to outside agencies.

E09 Natural Circulation Operations / 4 EK3.2 - Normal, abnormal and emergency operating 3.8 1 procedures associated with (Natural Circulation uperations)

WACategory Totals: 4 4 4 4 4 4 Group Point Total: 24 3

PWR SR( amination Outline Printed: 0212(

Mile Island - 1 ES - 401 Form ES-401-3 I I I I

(/APE # E/APE Name / Safety Function K1 K2 K3 A1 A2 G KATopic Imp. Points 007 Reactor Trip I 1 X EK2.03 - Reactor trip status panel 3.6 1 007 Reactor Trip I 1 X EA1.03 - RCS pressure and temperature 4.1 1 008 Pressurizer(PZR) Vapor Space Accident (Relief AK2.01 - Valves 2.7 1 Valve Stuck Open) I 3 008 Pressurizer (PZR) Vapor Space Accident (Relief X AK3.03 - Actions contained in EOP for PZR vapor 4.6 1 Valve Stuck Open) 1 3 space accident/LOCA 009 Small Break LOCA 1 3 X 2.2.25 - Knowledge of bases in technical specifications 3.7 1 for limiting conditions for operations and safety limits.

I I I I EKI .O 1 - Natural circulation and cooling, including I I 009 Small Break LOCA 1 3 X I 4.7 I 1 II reflux boiling I I 022 027 Loss of Reactor Coolant Makeup 1 2 Pressurizer Pressure Control (PZR PCS) Malfunction X

X I- AKl.02 - Relationship of charging flow to pressure differential between charging and RCS AK2.03 - Controllers and positioners 3.1 2.8 1

1 033 I I I I I

Loss of lntermediate Range Nuclear lnstrumentation I I I 1 I 1x1 I

AK3.01 - Termination of startup following loss of 3.6 1 7 I I I I intermediate-range instrumentation 1

Facility: 1lhWeMile Island - 1 PWR SR( amination Outline Printed: 02/2(  ;

ES - 401 Emer :ncy and Abnormal Plant Evolutions - Tier 1 I Group 2 -

Form ES-401-3 EIAPE # E/APE Name / Safety Function KA Tonic

__._ ~- __ mp. Points 038 Steam Generator Tube Rupture (SGTR) / 3 EA1 .OS - Core cooling monitor 3.8* 1 038 Steam Generator Tube Rupture (SGTR) / 3 EA2.09 - Existence of natural circulation, using plant

- 4.2 1 parameters 06 1 Area Radiation Monitoring (ARM) System Alarms I 7 AA2.0 1 - ARM panel displays

- 3.7 1 06 1 Area Radiation Monitoring (ARM) System Alarms / 7 2.1.32 - Ability to explain and apply all system limits

- 3.8 and precautions.

AA 1.1 - Components, and functions of control and

- 3.7 A0 1 Plant Runback / 1 safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features E08 LOCA Cooldown / 4 EA2.2 - Adherence to appropriate procedures and

- 4.0 operation within the limitations in the facility's license and amendments EKI .3 - Annunciators and conditions indicating signals,

- 3.5 1 E08 LOCA Cooldown 1 4 and remedial actions associated with the (LOCA coolaown)

KIA Category Totals: 3 3 2 3 2 Group Point Total: 16 2

Facility: 1;1b ,e Mile Island - 1 PWR SR{ amination Outline Printed: 02/2( 3 E/APE # E/APE Name / Safety Function K1 K2 K3 A1 A2 G KATopic Imp. Points A08 Refuel Canal Level Decrease / 8 X AA2.1 - Facility conditions and selection of appropriate 4.0 1 procedures during abnormal and emergency operations E13 EOP Rules X 2.2.22 - Knowledge of limiting conditions for operations 4.1 1 and safety limits.

E13 EOP Rules X EKl.2 - Normal, abnormal and emergency operating 3.6 1 procedures associated with (EOP Rules)

WACategory Totals: 1 0 0 0 1 1 Group Point Total: 3 1

PWR SRO ( nination Outline Printed: 021 33 Facility:

\

'I lllae Mile Island -1 \

?S 401 rier 2 I ;mun 1 -Form S-401-SysIEv # System / Evolution Name K1 K2 K3 K4 K5 A4 G KA Topic Imp. Points 00 1 Control Rod Drive System / 1 K6.03 - Reactor trip breakers, including 4.2 1 controls 003 Reactor Coolant Pump System X K3.03 - Feedwater and emergency feedwater 3.1

- 1 (RCPS) / 4 003 Reactor Coolant Pump System A2.02 - Conditions which exist for an abnormal

- 3.9 1 (RCPS) / 4 shutdown of an RCP in comparison to a normal shutdown of an RCP 004 Chemical and Volume Control System X 2.4.4 - Ability to recognize abnormal

- 4.3 1 (CVCS) / 1 indications for system operating parameters which are entry-level conditions for emergency 004 Chemical and Volume Control System X and abnormal operating procedures.

A4.18 - Emergency borate valve

- 4.1 1 (CVCS) / 1 015 Nuclear Instrumentation System / 7 X K4.04 - Slow response time of SPNDs 3.6? 1 015 Nuclear Instrumentation System / 7 K6.04 - Bistables and logic circuits

- 3.2 1 022 Containment Cooling System (CCS) / X K2.01 - Containment cooling fans

- 3.1 1 5

026 Containment Spray System (CSS) / 5 X K1.O1 - ECCS 4.2 1 026 Containment Spray System (CSS) / 5 -

A3.01 Pump starts and correct MOV

- 4.5

- 1 positioning 06 1 Auxiliary / Emergency Feedwater X K5.02 - Decay heat sources and magnitude

- 3.6 1 (AFW) System / 4 -

1

PWR SRO ( qination Outline Printed: 02( 33 Facility: I tllee Mile Island - 1 ES 401 ;roup 1 Form ES-401-2 I I SysIEv # Svstem I Evolution Name IK1 IK2 I K ~I K ~ KA Topic Imp. Points 06 1 Auxiliary / Emergency Feedwater K6.01 - Controllers and positioners 2.8* 1 (AFW) System 1 4 063 2.1.14 - Knowledge of system status criteria which require the notification of plant 063 068 I

D.C. Electrical Distribution System 16 Liquid Radwaste System (LRS) 1 9 I I IX I I I I I K4.04 - Trips K1.07 Sources of liquid wastes for LRS 2.9 1 068 Liquid Radwaste System (LRS) 1 9 A3.02 - Automatic isolation 3.6 1 07 1 Waste Gas Disposal System (WGDS) X A2.02 - Use of waste gas release monitors, 3.6 1 19 radiation, gas flow rate, and totalizer 072 K3.02 - Fuel handling operations 3.5 1 072 Area Radiation Monitoring (ARM) X 41.01 - Radiation levels 3.6 1 KIA Category Totals: 2 1 2 2 3 1 2 2 1 2 Group Point Total: 19 2

PWR SRO ( nination Outline Printed: 02( '03 Facility: 'I llree Mile Island -1 ES - 401 ;roup 2 Form S-401-I System / Evolution Name KA Topic Imp. Points Reactor Coolant System (RCS) / 2 -

K3.02 Fuel 4.5 1 002 Reactor Coolant System (RCS) / 2 K5.18 - Brittle fracture 3.6 1 01 1 Pressurizer Level Control System A2.08 - Loss of level compensation 2.8 1 (PZR LCS) / 2 012 Reactor Protection System / 7 K4.05 - Spurious trip protection 2.9 1 012 Reactor Protection System / 7 K6.11 - Trip setpoint calculators 2.9 029 Containment Purge System (CPS) / 8 K1.01 - Gaseous radiation release monitors 3.7 033 Spent Fuel Pool Cooling System A3 .O 1 - Temperature control valves 2.7*

(SFPCS) / 8 034 Fuel Handling Equipment System A2.02 - Dropped cask 3.9 (FHES) / 8 035 Steam Generator System (S/GS) / 4 2.4.49 - Ability to perform without reference to 4.0 procedures those actions that require immediate operation of system components and controls.

039 Main and Reheat Steam System 2.4.6 - Knowledge symptom based EOP 4.0 (MRSS) / 4 mitigation strategies.

I 039 Main and Reheat Steam System A4.07 - Steam dump valves 2.9 (MRSS) / 4 I 1

Facility:

r; uree Mile Island -

'I 1 PWR SRO ( nination Outline Printed: 02( 03 3s - 401 S-401-2 Sys/Ev # System / Evolution Name (2 K3 KA Topic Imp. Points 055 Condenser Air Removal System -

A3.03 Automatic diversion of CARS exhaust 2.7* 1 (CARS) / 4 064 Emergency Diesel Generator (EDIG) X K2.03 - Control power 3.6 System 1 6 073 Process Radiation Monitoring (PRM) K4.01 - Release termination when radiation 4.3 System I 7 exceeds setpoint 073 Process Radiation Monitoring (PRM) A1.O1 - Radiation levels I I

3.5 System I 7 075 Circulating Water System I 8 X -

K3.07 ESFAS 3.5*

103 Containment System / 5 2.4.30 - Knowledge of which events related to 3.6 system oDerations7status should be reported to I WA Category Totals: 1 1 2 2 1 3 Group Point Total: 17 2

PWR SRO ( nination Outline Printed: 02( '03 Facility: 'I lllee Mile Island - 1 1

SS - 401 Plant rier 2 / ;roup 3 Form S-401-.

1

$ys/Ev# System / Evolution Name K3 K4 K5 KA Topic Imp. Points 005 Residual Heat Removal System K6.03 - RHR heat exchanger 2.6 1 (RHRS) 1 4 005 Residual Heat Removal System A 1.O 1 - Heatup/cooldown rates 3.6

- 1 (RHRS) 1 4 007 Pressurizer Relief TanWQuench Tank X -

K4.0 1 Quench tank cooling 2.9

- 1 System (PRTS) / 5 008 Component Cooling Water System 2.2.22 - Knowledge of limiting conditions for 4.1 1 (CCWS) 1 8 operations and safety limits.

WA Category Totals: 1 1 0 1 Group Point Total: 4 1

Generic Knowledger- 4 Abilities Outline (Tier 3)

Printed: 02124120(

PWR SRO Examination Outline Form ES-401-5 Facilitv: Three Mile Island - 1 Generic Category KA KA Topic Imp. Points Conduct of Operations 2.1.5 Ability to locate and use procedures and directives related to shift staffing and activities. 3.4 i 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating 4.4 1 characteristics, reactor behavior, and instrument interpretation.

2.1.10 Knowledge of conditions and limitations in the facility license. 3.9 1 2.1.34 Ability to maintain primary and secondary plant chemistry within allowable limits. 2.9 1 Category Total: 4 Equipment Control 2.2.1 Ability to perform pre-startup procedures for the facility, including operating those

-3.6 1

controls associated with plant equipment that could affect reactivity.

2.2.1 1 Knowledge of the process for controlling temporary changes. 3.4* 1 2.2.19 Knowledge of maintenance work order requirements. 3.1 1 2.2.26 Knowledge of refueling administrative requirements. 3.7 1 2.2.27 Knowledge of the refueling process. 3.5 1 Radiation Control I 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements. 3.0 1 2.3.3 Knowledge of SRO responsibilities for auxiliary systems that are outside the control room 2.9 (e.g., waste disposal and handling systems).

2.3.8 Knowledge of the process for performing a planned gaseous radioactive release. 3.2 I 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against oersonnel exuosure.

Category Total: 4 1

Generic Knowledge iy4Abilities Outline (Tier 3)

\ Printed: 02,241204 PWR SRO Examination Outline Emergency ProceduredPlan 2.4.10 Knowledge of annunciator response procedures. 3.1 1 2.4.1 1 Knowledge of abnormal condition procedures. 3.6 1 2.4.33 Knowledge of the process used track inoperable alarms. 2.8 1 2.4.44 Knowledge of emergency plan protective action recommendations. 4.0 1 CategoryTotal: 4 Generic Total: 17 2

ES-401 Record of Rejected K/As Form ES-401-10 I 11 I ier / Group R X , C ~ O I I Setecled

~IY WA I Reason for Rejection 2/ 1 061 A1.03 I This K / A needs t o be supressed, s i n c e i t a p p l i e s t o a m u l t i - u n i t f a l i l i t y .

I Tier 2 / G r o u p 1 K / A 061 K6.01 was randomly s e l e c t e d a s a replacement. I 212 035 2.4.49 This K / A needs t o be suppressed, s i n c e t h e r e a r e no "emergency e s s e n t i a l "

SWS Pumps a s s o c i a t e d with t h e Circulating Water System a t TMI.

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

ES-301 Administrative Topics Out1ine Form ES-301-1 Facility: Three Mile Island Unit 1 Date of Examination: May 12, 2003 Examination Level (circle one): RO / SRO Operating Test Number:

Administrative Describe method of evaluation:

Topic/Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions.

A. 1 Plant Perform Estimated Critical Boron Concentration Parameter Calculation. (JPM)

Ve rification Shift Staffing Minimum Shift Staffing, Control of Overtime. (JPM)

Requirements A.2 Use Of Station Predict Operational Impact Of Instrument Failure. (JPM)

Drawings A. 3 Radiation Liquid Radiation Release Approval. (JPM)

Release (Organ Dose prevents approval)

A.4 Emergency Classify Event And Complete Initial Notification Forms.

Classification (JPM)

5-301 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-:

Facility: Three Mile Island Unit 1 Date of Examination: Mav 12.2003 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.:

B.l Control Room Systems Type Code* Safety System/JPM Title Function

a. Chemical and Volume Control (OO4)/Perform an Emergency 1 Boration (Alt. Path - Backup Emergency Boration Required).

b. Engineered Safety Feature Actuation Systems (013)IRespond to D, s 2 inadvertent ES Actuation.

c. Emergency Core Cooling System (OOG)/Respond to a High N, A, S 3 Pressure Injection (HPI) initiation (Alt. Path - MU-V-14A fails).

d. Residual Heat Removal System (OOS)/Respond to a failure of N, A, S,L 4 Primary Low Pressure Injection (Alt. Path - DHV-6 Fails to Open).

e. Main Steam System (039)/Respond to inadvertent closure of a D, s 4 Secondary Main Steam Isolation Valve.

f. Containment Cooling System (022)/Return Reactor Building (RB) N, s 5 Emergency Cooling to Engineered Safeguards Standby.

g. Emergency Diesel Generator (EDG) System (064)/EDG D, A, S 6 Operation (Alt. Path EDG Fails to Auto Load).

8.2 Facility Walk-Through

a. Chemical and Volume Control System (004)/Manually Open RCP N, R 2 Seal Injection Isolation Valve (MU-V-26). Emergency

~ ~ ~ ~ ~ ~ ~ ~~ ~~ ~

b. Pressurizer Pressure Control System (010)TTransfer Pressurizer Heater Group 8 or 9 to an Engineered Safeguards Bus.

D l1 3 Emergency

c. Emergency Feedwater System (061)ILocal Reset of Emergency D 4 Secondary Feedwater Pump (EF-P-1). 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.: #I Op Test No.:

Facility: Unit 1 Examiners Operators CRS URO PRO 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.

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

Event Malf. Event Event No. No. Type* Description URO 1 ARO Secure feedwater pump (FW-P-1A) .

N us RDOI 17 C URO Dropped Control Rod in controlling group.

2 ARO c us R URO Manual power reduction due to dropped rod.

3 ARO us I/O OVERRIDE I URO Pressurizer level controller fails to 0% demand, closing makeup control 4 ARO valve (MU-V-17).

I us FWl5B M URO Loss of feedwater due to feedwater pump (FW-P-1B) failure.

5 M ARO . MAP - dont NOTES: Rapid control oil leak, auto start of oil pump, I M US I want,manualRxTrip.

I RD28 I I URO I 6 ARO ATWS - RPS Auto Trip Failure.

I us MU16A M URO Small Break LOCA, Loss of Subcooling Margin 7 M ARO M US MU23C C URO High pressure injection pump (MU-P-IC) fails to automatically start.

8 ARO c us

Scenario Outline Simulation Facility: Three Mile Island Scenario No.: #2 Op Test No.:

Unit 1 Examiners Operators CRS URO PRO Description 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 instrumentationfailure 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 1B 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.

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

Turnover See Attached "Shift Turnover" Sheet.

Event Malf. Event Event No. No. Type* Description URO Switch operating Condensate Pumps.

I N ARO N US N127B I URO RCS Loop A T-Hot transmitter failure (high), affecting ICs T-Ave 2 ARO indication.

I us 3 C URO Main Generator hydrogen gas leak.

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

4 ARO us ED02A C URO 1A Auxiliary Transformer fault (Technical Specifications).

5 C ARO c us MS02B M URO Main Steam Line Rupture Inside the RB with ESAS actuation.

6 M ARO M US MU08C C URO Stuck open high pressure injection valve.

7 ARO c us

Scenario Outline 1

%mulation Facility: Three Mile Island Scenario No.: Alternate Op Test No.:

Unit 1 Examiners Operators CRS URO PRO Description 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.

Initial Conditions 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-1B oil change.

-Turnover See Attached "Shift Turnover" Sheet.

1) Event No. I Malf. Event Event Type* Description I URO Pressurizer level instrument failure.

ARO I us URO Switch operating Makeup Pumps.

N ARO N US C URO MU-V-32, RCP seal injection valve, failure.

ARO c us C URO Small OTSG 1A tube leak.

ARO c us l 5 URO Initiation of plant shutdown.

ARO us I/R URO Control Rod Drive System automatic control failure.

II I I

TH16A M I

ARO us URO OTSG tube rupture.

ARO M US I URO ES Train failure.

8 I ES02B ARO I 10 Override I us (N)ormal, (R)eactivity, (I)n wment, (C)omponent, (M)ajor