ML063060395

October 2006 Exam - Draft Section C Operating Exam (Folder 2)

ML063060395
Person / Time
Site:	Oyster Creek
Issue date:	09/05/2006
From:	AmerGen Energy Co
To:	D'Antonio J Operations Branch I
Sykes, Marvin D.
References
Download: ML063060395 (136)

Text

Scenario Outline Appendix D Event No.

Malf. No.

Event Type*

1 R

RO 2

N BOP 3

TS SRO 4

I BOP 5

C RO 6

TS SRO 7

C RO 8

C BOP 9

M Crew 10 M

Crew Facility: Ovster Creek Event Description Withdraw control rods.

Transfers from the LFRV to the MFRV.

Respond to inoperability of Wide Range Drywell Pressure Monitor (PT-53).

Respond to failed acoustic monitor for Electromatic Relief Valve (EMRV)

Respond to Control Rod Drive Flow Control Valve (CRD FCV) failed closed.

Respond to loss of DC control power to Containment Spray Pump 51 A.

Respond to Intermediate Range Monitor (IRM) 11, which fails low.

Respond to Electronic Pressure Regulator (EPR) fluctuations.

Respond to failure of the Mechanical Pressure Regulator (MPR) and Anticipated Trip Without Scram (ATWS) (electric).

Respond to a LOCA.

Scenario No.: NRC 1 Op Test No.: NRC 2006-1 Examiners:

Operators:

Initial Conditions:

0 The reactor is starting up, after a 5-day forced outage, with the MODE SWITCH in STARTUP.

NRC 1 Page 1 of 19

W Event Type Position 1

R RO 5

C RO 7

C RO 8

1 4

2 2

0 2

Total Malfunctions (5-8):

Malfunctions after EOP entry (1 -2)

Abnormal Events (2-4)

Major Transients (1 -2)

EOPs entered requiring substantive actions (1 -2)

EOP Contingencies w/ substantive actions (0-2)

Critical Tasks (2-3)

Description Withdraw control rods Failed closed of CRD FCV Drifting low IRM Event Type Position 2

N BOP 4

C BOP 8

C BOP

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Position Description 9

Crew MPR failure plus ATWS Description Transfer from the LFRV to the MFRV EMRV acoustic monitor failure EPR Fluctuations Event Type Position 3

TS SRO 6

TS SRO (3.4. C)

Description Loss of DW wide range pressure transmitter (3.13.E)

Loss of Containment Spray Pump 51A control power NRC 1 Page 2 of 19

Scenario Summary The scenario will begin with Instrument Maintenance calibrating the drywell wide range pressure transmitters PT-53 (IAW 604.3.01 8, Wide Range Drywell Pressure Calibration).

%W

1. The RO will withdraw control rods IAW procedure 201, Plant Startup.

(REACTIVITY MANIPULATION)

2. The BOP will successfully transfer from the LFRV to the MFRV per procedure (This will be directed by procedure 201, step 6.57, and performed in procedure 317, step 6.3.15) (NORMAL EVOLUTION)

3. The SRO will receive a call from the Instrument Technician Foreman that drywell wide range pressure transmitter (PT-53) cannot be left in-spec per the calibration procedure (604.3.01 8). The SRO will declare the instrument inoperable and apply TS 3.13. E. (TS)

4. The BOP will respond to a failure of the acoustic monitor for EMRV NR-108A.

Annunciator B4g, SV/EMRV NOT CLOSED, will alarm and the effected EMRV will still indicate closed. The BOP will defeat the alarm IAW procedure 413, Operation of the Safety Valve/EMRV Acoustic Monitoring System.

(INSTRUMENT FAILURE) (ABN)

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5. The in-service CRD FCV (NC30A) will fail closed. Several panel indications are available to diagnose the problem. The RO will swap to the alternate FCV IAW procedure 302.1 (Control Rod Drive System), step 4.3.3. ABN-6, Control Rod Drive System may be entered. (COMPONENT FAILURE) (ABN)

6. The TB Operator will notify the Control Room that there is no valid breaker position indicating lights at the breaker for Containment Spray Pump 51 A. There will also be a loss of proper indication on the Control Room panel, and the SRO will declare the Containment Spray Pump 51A inoperable and apply the appropriate TS 3.4.C. (TS)

7. IRM 11 will fail low causing a rodblock (RAP-Gle, G2e). The IRM can be bypassed, and the rodblock cleared. The SRO will verify TS, and that no actions are required. (TS 3.1.l) (COMPONENT FAILURE)

NRC 1 Page 3 of 19

8. Small step changes in the EPR begins, which affect reactor power and pressure. ABN-9, EPR Malfunctions, should be entered. It is expected that the BOP will transfer to the Mechanical Pressure Regulator (MPR) IAW 315.4, Transferring Pressure Regulators. (COMPONENT FAILURE) (ABN)

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9. The MPR will fail causing all turbine bypass valves (TBVs) to go closed, causing RPV pressure to increase. The RO may attempt to scram prior to the RPV high pressure scram setpoint, but an electrical ATWS will occur and no control rods will insert. The crew will enter RPV Control -With ATWS. The crew will insert control rods by venting the scram air header (IAW SP-21, Alternate Insertion of Control Rods). RPV pressure control will be available with both loops of isolation condensers and electromatic relief valves (EMRVs). (EOP) (MAJOR)

10. A LOCA in the primary containment will require the Crew to enter the Primary Containment Control EOP. The operator will spray the containment (IAW SP-29, Initiation of the Containment Spray System for Drywell Sprays). (EOP) (MAJOR)

Critical Tasks

1.

With a scram signal present and the reactor not shutdown, initiate alternate control rod insertion methods to bring the reactor shutdown.

These actions act to shutdown the reactor given that a reactor scram setpoint has been reached (high reactor pressure) and the reactor did not automatically shutdown.

Sprays the Drywell when DrywelliTorus pressure exceeds 12 psig.

This action ensures the continued operability of the Primary Containment as a viable fission product barrier.

2.

NRC 1 Page 4 of 19

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Scenario No.: NRC 1 Event No.: 1 Event

Description:

Raise reactor Dower bv withdrawina control rods IAW 302.2, Control Rod Drive Manual Control Svstem. steD 3.3/4.3. (as directed from Drocedure 201, Plant StartuD.1. Complete the withdrawal of control rods in the current group.

Initiation: Following the shift turnover and the Crew assumption of the shift.

Cue: As directed by SRO Time Position SRO RO BOP

~~~

Applicant's Actions or Behavior 0

Direct control rod withdrawals to raise power IAW procedure 201, Plant Startup, with control rod manipulations performed IAW procedure 302.2, Control Rod Drive Manual Control System, step 3.3.

0 Withdraws control rods IAW procedure 302.2, Control Rod Drive Manual Control System, step 3.3 (attached). (Panel 3F)

The individual performing the peer check shall verify that the correct control rod has been selected by comparing the control rod selected with the governing procedure.

The individual performing the peer check shall state agreement with the control rod selection, its initial position, its target position, the method (Le. single notch or continuous) and the direction of movement.

Observe the actions of the RO to verify movement of the correct control rod to its target position.

0 Terminus: I Control rods in the current group have been withdrawn to their target position.

Notes/Comments I

1 NRC 1 Page 5 of 19

I OYSTER CREEK GENERATING Number STATION PROCEDURE I

302.2 an rueio.'r,n ~ompt-y i_/

Title Control Rod Drive Manual Control System Revision No.

29 3.2.8.6 The individual performing the peer check shall observe the actions of the licensed operator to verify movement of the correct control rod to its target position. The individual performing the peer check shall immediately identify any mistake.

3.2.9 If a control rod is inadvertently withdrawn beyond its intended position or a control rod is inadvertently inserted one notch beyond its intended position and the deviation is identified immediately, then notify the US and immediately restore the control rod to its intended position prior to selection of the next control rod.

3.2.10 If a control rod is inadvertently moved one notch beyond its intended position (e.g. double notching) and not restored to its intended position prior to selecting the next control rod or the control rod is moved more than one notch beyond its intended position (SOER 84-02), then consider the control rod to be mispositioned.

If a control rod was mispositioned due to a Control System failure, then refer to ABN-6, Control Rod Drive System.

e' 3.2.1 I

--+-

3.3 Instructions 3.3.1 Notch Override Emergency Rod In switch is used to prevent a double notch, THEN RECORD event in accordance with LS-AA-I 25, Corrective Action Program (CAP) Procedure.

3.3.2 3.3.3 VERIFY the PERMIT light is illuminated.

CONFIRM the ROD POWER Switch ON.

3.3.4 SELECT the control rod to be withdrawn by rnornentarilv depressing the pushbutton on the CONTROL ROD SELECT pushbutton matrix.

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I 8.0

I ii Title Control Rod Drive Manual Control System 3.3.4.1 VERIFY the following:

Revision No.

29

1. The pushbutton light and the rod select indicator above the position display for selected rod are illuminated.

3.3.6

2.

the selected rod pushbutton and rod select indicator above the desired rod are illuminated.

Nuclear Instrumentation shall be continuously monitored during rod movement to avoid unexpected power response and prevent 3.3.5 NOTE The following verification shall be performed by 2nd Operator (Peer Checker).

r r

9.0

h m OYSTER CREEK GENERATING STATION PROCEDURE Av Exebn Company Number 302.2 3.3.7 3.3.8 3.3.9 I

u Title Control Rod Drive Manual Control System NOTE If the handle is not released in the following step, the timer will complete its cycle but will not reset until released.

Revision No.

29 Immediately RELEASE the switch allowing it to return to OFF by spring action.

3.3.7.1 VERIFY the following:

1. The red WITHDRAW light is illuminated approximately 2 seconds following switch movement and remains on for approximately 1.5 seconds.

2. The rod latches in the next even-numbered position before the SETTLE light is extinguished.

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I IF a rod double notches out (moves to next latching posit i o n ),

THEN RETURN the rod to its desired position and RECORD per LS-AA-125, Corrective Action Program (CAP)

Proced we.

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I NOTE If a control rod and drive become uncoupled, the rod position display will go dark and the ROD OVERTRAVEL alarm (H-5-a) will annunciate.

WHEN THEN 3.3.9.1 3.3.9.2 a control rod reaches position "48",

PERFORM a coupling check as follows:

HOLD the ROD CONTROL switch in ROD OUT NOTCH and simultaneouslv PLACE the NOTCH OVERRIDE Switch in NOTCH OVERRIDE.

AFTER the timer times out, THEN VERIFY the rod position display indicates a continuous digital readout of "48" with red backlighting.

10.0

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Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Transfers from Low Flow Reaulatina Valve (LFRV) to the Main Feed Reaulating Valve (MFRV) for Feedwater Pump A. IAW procedure 317, Feedwater Svstem, section 6.3.15.

Scenario No.: NRC 1 Event No.: 2 Initiation: Directly following control rod withdrawals.

Cue: As directed by SRO following control rod withdrawals.

Time Terminus:

ER in AUTO IAW procedure 317, BOP Transfers from the Low Flow Regulating Valve (LFRV) to the Main Feed Regulating Valve (MFRV) and places on the MASTER FEEDWATER LEVEL CONTROLLER in AUTO IAW procedure 317, Feedwater System, section 6.3.1 5. (attached).

Report the MFRV for Fw Pump A is in service in AUTO.

Feedwater flow control has been transferred from the LFRV to the MFRV and placed on the MASTER FEEDWATER LEVEL CONTROLLER in AUTO.

NotesKomments NRC 1 Page 6 of 19

&neam_

1 OYSTER CREEK GENERATING Number An $x&n Company STATION PROCEDURE I

317 I

U' Title Feedwater System (Feed Pumps to Reactor Vessel)

Revision No.

74

3.

WHEN the S display digital readout and the P display digital readout are

equal, THEN PLACE the LRFV FLOW CONTROLLER in AUTO.

6.3.13.1 I MONITOR Reactor level for any changes.

6.3.13.12 CONTROL RPV water level between 155 and 165 in TAF as follows:

1. ADJUST the LFRV FLOW CONTROLLER V-ID12A or V-ID12C setpoint as required.

2. II CAUTION Operation with the low flow valve less than 10% open should be minimized in order to prevent erosion of the valve disk and seat.

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I ESTABLISH letdown flow through the RWCU System in accordance with Procedure 303.

3. ADJUST the letdown flow rate to open the Low Flow Regulating Valve approximately 10% stroke.

6.3.13.13 Locally STOP the auxiliary oil pump by UNLOCKING and RELEASING the local pushbutton.

6.3.14 WHEN feedwater flow reaches approximately 0.5 X I O 6 Ibm/hr, 1

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I THEN VERIFY the MIN FLOW VALVE V-2-18 or V-2-20 indicates CLOSED.

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6.3.1 5 WHEN feedwater flow reaches approximately 0.6 X 1 O6 Ibm/hr, THEN COMPLETE the following:

1 6.3.15.1 VERIFY that the MFRV is CLOSED.

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1 OYSTER CREEK GENERATING Number nr irebn Company STATION PROCEDURE 1

317 I

Title Feedwater System (Feed Pumps to Reactor Vessel)

Revision No.

74 6.3.qt5.2 OPEN the MFRV BLOCK VALVE V-2-740 or V-2-741 as follows:

1.

I. PLACE the selected string block valve control switch in the OPEN position.

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I NOTE In order to maintain RPV water level constant, the opening of the MFRV and closing of the LFRV should be performed concurrently and in small increments.

2.

CONFIRM the block valve indicates OPEN.

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3. VERIFY the selected string LFRV throttles closed as required to compensate for MFRV leakage.

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I PLACE the selected string LFRV in MAN.

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2. Slowly OPEN the MFRV.

3. Slowlv CLOSE the LFRV.

4. REPEAT Steps 2. and 3. until the LFRV is fully closed.

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I 6.3.16 PLACE the selected string Main Feedwater Regulating Valve in MASTER MANUAL as follows:

6.3.16.1 CONFIRM the MASTER FEEDWATER LEVEL CONTROLLER in MAN.

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OYSTER CREEK GENERATING Number 1

STATION PROCEDURE I

317 AmerGeaw An Exchn Company Title Feedwater System (Feed Pumps to Reactor Vessel)

Revision No.

74 6.3.16.2 I NOTE The manual adjustment knob is responsive when the V-display is selected.

MATCH the S display with the V display on the selected string MFRV FLOW CONTROLLER by rotating the manual adjustment knob on the MASTER FEEDWATER LEVEL CONTROLLER as required while monitoring deviation (Y display) on the MFRV FLOW CONTROLLER.

6.3.16.3 1

NOTE The Y display on the MFRV FLOW CONTROLLER provides an indication of deviation (S-V).

WHEN THEN the S and V displays are approximately equal (zero deviation on the Y display)

AND the S Bar is stable, PLACE the selected string MFRV FLOW CONTROLLER in AUTO.

6.3.16.4 6.3.16.5 MONITOR RPV water level and feedwater flow for any changes.

CONTROL RPV water level between 155 and 165 in TAF using the MASTER FEEDWATER LEVEL CONTROLLER.

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I 37.0

OYSTER CREEK GENERATING Number 1

317 I

STATION PROCEDURE AmerEen.

All &elon CO7lpdil~

L' Title Feedwater System (Feed Pumps to Reactor Vessel)

Revision No.

74 6.3.17 If the MASTER FEEDWATER LEVEL CONTROLLER is to be placed in MASTER AUTO, THEN PERFORM the following:

6.3.17. I CONFIRM the LEVEL TRANSMITTER SELECTOR switch is in the AUTO position.

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I 6.3.17.2 SELECT the S display on the MASTER FEEDWATER LEVEL CONTROLLER.

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I 6.3.17.3 MATCH the S display digital readout to the P display digital readout on the MASTER FEEDWATER LEVEL CONTROLLER.

6.3.17.4 WHEN the S display digital readout and the P display digital readout are equal,

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I THEN PLACE the MASTER FEEDWATER LEVEL CONTROLLER in AUTO.

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I 6.3.17.5 MONITOR Reactor level and feedwater flow for any changes.

6.3.17.6 MAINTAIN Reactor level at 160 in TAF or as directed by the OS by adjusting the MASTER FEEDWATER LEVEL CONTROLLER setpoint.

38.0

Appendix D Required Operator Actions FO~TYI ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Failure of D w e l l Pressure Instrument (PT-53) to calibrate.

Scenario No.: NRC 1 Event No.: 3

~~

Initiation: Following the shift from the LFRV to the MVRV.

Cue: Notified by Instrument Technician Forman by phone call to the Control Room.

Time ROLE PLAY ROLE PLAY Terminus:

Applicants Actions or Behavior As the Instrument Technician Fonan, call the Control Room and report: I am reviewing procedure 604.3.01 8, Wide Range Drywell Pressure Calibration, and note that the drywell wide range pressure transmitter (PT-53) could not be left in-spec per this procedure.

SRO 0

Review Tech Specs 3.1 3.E (attached)

Declares drywell pressure instrument PT-53 inoperable and applies TS:

o With the number of OPERABLE accident monitoring channels less than the total Number of Channels shown in Table 3.13.1 restore the inoperable channel@) to OPERABLE status within 7 days or place the reactor in the SHUTDOWN CONDITION within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

0 Updates the Crew Notifies Work Week manager for repair If asked, report that drywell wide range pressure transmitter PT-54 was calibrated without any problems. (This pressure transmitter is also calibrated by the same procedure as PT-53)

The SRO has applied TS 3.13.E, updated the Crew, and requested repair.

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Notes/Comments I

NRC 1 Page 7 of 19

'U' C.

In the event that any of these monitoring channels become inoperable, they shall be made OPERABLE prior to startup following the next COLD SHUTDOWN.

D.

Wide Ranae Torus Water Level Monitor

1.

Two wide range torus water level monitor channels shall be continuously indicated in the control room during POWER OPERATION.

2.

With the number of OPERABLE accident monitoring channels less than the total Number of Channels shown in Table 3.13.1, restore the inoperable channel(s) to OPERABLE status within 7 days or place the reactor in the SHUTDOWN CONDITION within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels operable requirements of Table 3.1 3.1, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or place the reactor in the SHUTDOWN CONDITION within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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E.

Wide Range D w e l l Pressure Monitor

1.

Two Wide Range Drywell Pressure monitor channels shall be continuously indicated in the control room during POWER OPERATION.

2.

With the number of OPERABLE accident monitoring channels less than the total Number of Channels shown in Table 3.13.1, restore the inoperable channelts) to OPERABLE status within 7 days or place the reactor in the SHUTDOWN CONDITION within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels operable requirements of 3.13.1, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or place the reactor in the SHUTDOWN CONDITION within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

F.

DELETED OYSTER CREEK 3.13-2 Amendment No.: a4,5?,

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123,

246

TABLE 3.13.1 INSTRUMENT ACC I D ENT MO N ITOR IN G I N STRUM E NTAT ION

1. Relief Valve Position Indicator (Primary Detector)

Relief Valve Position Indicator (Backup Indications")

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2. Wide Range Drywell Pressure Monitor (PT/PR-53 & 54)

3. Wide Range Torus Water Level (LT/LR-37 & 38)

4. DELETED

5. Containment High Range Radiation

6. High Range Radioactive Noble Gas Effluent Monitor

a. Main Stack

b. Turbine Building Vents TOTAL NO. OF MINIMUM CHANNELS CHANNELS OPERABLE 1 /valve 1 /valve 2

2 2

1 1

1 I

1 Acoustic Monitor Thermocouple I*

Thermocouple TE 65A can be substituted for thermocouple TE210-43V, W, or X Thermocouple TE 65B can be substituted for thermocouple TE210-43Y or Z OYSTERCREEK 3.1 3-5 Amendment No.: 5 4 4 3 4 8,

??e, ?E' 246

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Acoustic monitor for EMRV NR-108A fails.

Scenario No.: NRC 1 Event No.: 4 Initiation: Following Crew update of failed DW Pressure Transmitter PT-53.

Cue: Annunciator B4g, SV/EMRV NOT CLOSED, is in alarm.

o Verifies RPV pressure, checks for Auto-Depressurization and checks Valve Open position indication (1 F/2F) o Reports that EMRV NR-108A shows open by acoustic monitor but no other indications show that the valve is open o Defeats the alarm IAW procedure 413, section 4.3.6 (Panel 15R)

Switches the HI-ALARM switch to DEFEAT Switches the LO-BIAS switch to DEFEAT Presses the associated Alarm Reset push-button Terminus:

SRO 0

0 0

0 Directs BOP to defeat the acoustic monitor alarm Directs EMRV tailpiece temperature monitoring oncekhift Notifies Work Week manager for repair.

Reviews Tech Spec 3.13, Accident Monitoring Instrumentation (no actions required)

The acoustic monitor has been defeated for EMRV NR-108A.

____L P 3 NotesKomments NRC 1 Page 8 of 19

Appendix D Required Operator Actions F

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ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

In-Service CRD FCV (NC30A) fails closed.

Scenario No.: NRC 1 Event No.: 5 Initiation: Following defeat of acoustic monitor for EMRV NR-108A Cue: COOLING WATER FLOW, CLG WTWREACTOR AP, DRV WTWREACTOR AP indicators show downscale; position indicator for NC30A indicates closed; (4F) CRD TEMP HI annunciator H5c(5 F/6 F)

Time I Position I Applicants Actions or Behavior RO 0

Reports off-normal CRD indications 0

0 Responds to annunciator H5c, CRD TEMP HI (5F/6F)

Reports CRD FCV NC30A indicates closed Dispatches NLO to investigate (Role Play)

Places alternate CRD FCV in service IAW procedure 302.1, Control Rod Drive System, section 4.3.3, (Panel 4F) (attached)

Reports CRD parameters have returned to normal with CRD FCV NC30B in service.

Sim.

Operator

/ Role When requested to place CRD FCV B in service, insert LOA-CRD012 to 1 (open) for the FCV inlet valve and LOA-CRDO11 to 1 (open) for the FCV outlet valve. No need to isolate the failed CRD FCV. When complete, ROLE PLAY as the NLO and report the CRD FCV B has been lined-BOP Verifies CRD temperatures at CRD Temperature recorders (Panel 8R)

NOTE:

1 The CRD Temperature recorders (8R) are NON-FUNCTIONAL in the simulator. If the BOP goes back to read these recorders, state that several control rods indicate greater than 250 F (this is the alarm setpoint).

SRO 0

Directs RO to place the alternate CRD FCV in service IAW procedure 302.1, Control Rod Drive System (section 4.3.3).

Notifies Work Week manager for FCV air line repair.

Role Play As the NLO directed to investigate the CRD FCV A, report to the Control Room: There is a leak in the air line to CRD FCV A, and that you have isolated it. (FCV fails closed on loss of air)

NRC 1 Page 9 of 19

Appendix D Required Operator Actions FOMI ES-D-2 I

Notes/Comments NRC 1 Page 10 of 19

OYSTER CREEK GENERATING STATION PROCEDURE

- AmerGm.,

A,- !xela Company I --

Title Control Rod Drive System

3.

Slowly OPEN the B CRD filter vent valve V-I 5-50 to vent the filter.

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I Number 302.1 Revision No.

92

4.

OPEN the filter drain valve V-I 5-51 after the filter is vented.

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5.

The filter is now available for maintenance. If desired, hang tags using the WORKER TAGOUT PROCESS.

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6.

After maintenance is complete, place the filter in standby as follows:

6.1 CLOSE the B filter drain valve V-I 5-51.

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I 6.2 CLOSE the B CRD filter vent valve V-I 5-50.

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I 6.3 Slowlv OPEN the B CRD filter inlet valve V-I 5-1 3.

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I 6.4 Slowlv CRACK OPEN the B CRD filter vent valve V-I 5-50 until water issues from the vent.

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I 6.5 CLOSE the filter vent valve V-I 5-50.

The filter is now in standby.

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I 4.3.3 Flow Control Valve 4.3.3.1 PERFORM the following to place the alternate FCV in-service:

1.

TAKE manual control of the in-service Flow Control Valve (FCV) (NC30A or NC30B).

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I 28.0

OYSTER CREEK GENERATING STATION PROCEDURE AmerGent-An Ewm Company I

Title U

Control Rod Drive System Number 302.1 Revision No.

92

4.

2.

OPEN the Inlet and Outlet Valves for the alternate FCV to be placed into service:

a. For FCV NC30A:

OPEN V-15-16 0

OPEN V-I 5-17

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OPEN V-15-18

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OPEN V-15-19

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b. For FCV NC30B:

3.

PLACE the CRD FLOW CONTROL VALVES Switch to the position for the alternate FCV being placed in service.

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NOTE V-6-421, 4 Way Valve for V-6-422423 is a four position plug valve that can be rotated 360 degs and has a square head post operator on it. Positions are:

1. Air supplied to A control (a normal position).

2. Air supplied to A and B control (position to rotate through when changing FCVs).

3. Air supplied to B control (a normal position).

4. Air supplied isolated (A and B control lines are connected but have no air supply to either/position only to go to if no control air is needed to either FCV).

Refer to Attachment 302.1-9 for V-6-421 positioning.

PLACE the 4 Way Valve V-6-421 (CRD Valving Area, RB 23 elev) in the correct position to supply air to the alternate controller in accordance with Attachment 302.1-9.

1 29.0

OYSTER CREEK GENERATING STATION PROCEDURE her-..

An I*c'rm Company Number 302.1

5.

VERIFY all pressures and flows are normal in accordance with values specified on Attachment 302.1-6.

I Title

'U Control Rod Drive System

6. RETURN control of the FCV to automatic.

Revision No.

92

7.

CLOSE the Inlet and Outlet Valves for the FCV to be taken out of service.

4.3.4.1

a. For NC30A:

NOTES

1. The differential pressure in the drive water header (local DP indicator) should remain stable while NC25 is opened and V-I 5-21 is closed.

2. Isolation of Drive Water Pressure Control valve (PCV) NC18 (V-15-122) requires manual adjustment of PCV NC25 (V-I 5-1 27) to maintain drive water pressure in the CRD Hydraulic System.

CLOSE V-I 5-1 6 CLOSE V-I 5-1 7

b. For NC30B:

0 CLOSE V-I 5-1 8 0 CLOSE V-I 5-1 9 4.3.4 Drive Water Pressure Control Valve PERFORM the following to transfer from PCVNCI8 TO PDVNC25:

1.

Simultaneouslv OPEN manual control valve NC25 (V-I 5-1 27) and CLOSE PCV NC18 inlet valve (V-15-21) (CRD Valve Area).

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I 30.0

Appendix D Required Operator Actions FOITTI ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Loss of Containment Serav Pumr, 5 i A control Dower Scenario No.: NRC 1 Event No.: 6 Initiation: Following swap of CRD FCVs and CRD parameters return to normal Cue: Notified by in-plant NLO while performing building Rounds Volt room, there is no valid breaker indication for Containment Spray Pump 51A (Bus 1A2). I changed the light bulbs but it made no difference.

I Reviews Tech Spec 3.4.C.4 (attached) o o

Declares Containment Spray Pump 51 A inoperable If a pump in the containment spray system or emergency service water system becomes inoperable, the reactor may remain in operation for a period not to exceed 15 days provided the other similar pump is verified daily to be operable.

Verifies redundant Containment Spray Pump is operable.

o Notifies Work Week Manager for repair 0

0 Updates the Crew Verifies no valid breaker indication on Panel 1 F/2F for Containment Spray Pump 51 A.

Notes/Comments NRC 1 Page 11 of 19

B.

Automatic Depressurization System

1.

Five electromatic relief valves, which provide the automatic depressurization and pressure relief functions, shall be operable when the reactor water temperature is greater than 212°F and pressurized above 110 psig, except as specified in 3.4.B.2 and during Reactor Vessel Pressure Testing consistent with Specifications 1.39 and 3.3.A.( i).

If at any time there are only four operable electromatic relief valves, the reactor may remain in operation for a period not to exceed 3 days provided the motor operated isolation and condensate makeup valves in both isolation condensers are verified daily to be operable.

2.

3.

If Specifications 3.4.B.1 and 3.4.B.2 are not met; reactor pressure shall be reduced to I I O psig or less, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.

The time delay set point for initiation after coincidence of low-low-low reactor water level and high drywell pressure shall be set not to exceed two minutes.

C.

Containment Spray System and Emergency Service Water System NOTE: LCO 3.0.C.2 is not applicable to the Containment Spray System and Emergency Service Water System 4

1.

2.

3.

4.

5.

The containment spray system and the emergency service water system shall be operable at all times with irradiated fuel in the reactor vessel, except as specified in Specifications 3.4.C.3, 3.4.C.4,3.4.C.6 and 3.4.C.8.

The absorption chamber water volume shall not be less than 82,000 ft3 in order for the containment spray and emergency service water system to be considered operable.

If one emergency service water system loop becomes inoperable, its associated containment spray system loop shall be considered inoperable. If one containment spray system loop andor its associated emergency service water system loop becomes inoperable during the run mode, the reactor may remain in operation for a period not to exceed 7 days provided the remaining containment spray system loop and its associated emergency service water system loop each have no inoperable components and are verified daily to be operable.

If a pump in the containment spray system or emergency service water system becomes inoperable, the reactor may remain in operation for a period not to exceed 15 days provided the other similar pump is verified daily to be operable. A maximum of two pumps may be inoperable provided the two pumps are not in the same loop. If more than two pumps become inoperable, the limits of Specification 3.4.C.3 shall apply.

During the period when one diesel is inoperable, the containment spray loop and emergency service water system loop connected to the operable diesel shall have no inoperable components.

OYSTER CREEK

-LJ 3.4-5 Amendment No.: 75, ! 5 ?,

?44,247

6.

If primary containment integrity is not required (see Specification 3.5.A), the containment spray system may be made inoperable.

7, If Specifications 3.4.C.3, 3.4.C.4, 3.4.C.5 or 3.4.C.6 are not met, the reactor shall be placed in cold shutdown condition. If the containment spray system or the emergency service water system becomes inoperable, the reactor shall be placed in the cold shutdown condition and no work shall be performed on the reactor or its connected systems which could result in lowering the reactor water level to less than 4'8" above the top of the active fuel.

8.

The containment spray system may be made inoperable during the integrated primary containment leakage rate test required by Specification 4.5, provided that the reactor is maintained in the cold shutdown condition and that no work is performed on the reactor or its connected systems which could result in lowering the reactor level to less than 4'8" above the top of the active fuel.

D.

Control Rod Drive Hydraulic System

1.

The control rod drive (CRD) hydraulic system shall be operable when the reactor water temperature is above 212°F except as specified in 3.4.D.2 and 3.4.D.3 below.

2.

If one CRD hydraulic pump becomes inoperable when the reactor water temperature is above 212"F, the reactor may remain in operation for a period not to exceed 7 days provided the second CRD hydraulic pump is operating and is checked at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. If this condition cannot be met, the reactor water temperature shall be reduced to less than 212°F.

3.

Core Spray and Containment Spray Pump Compartments Doors During reactor vessel pressure testing, at least one CRD pump shall be operable.

i/

E.

The core spray and containment spray pump compartments doors shall be closed at all times except during passage in order to consider the core spray system and the containment spray system operable.

F.

Fire Protection System

1.

The fire protection system shall be operable at all times with fuel in the reactor vessel except as specified in Specification 3.4.F.2.

2.

If the fire protection system becomes inoperable during the run mode, the reactor may remain in operation provided both core spray system loops are operable with no inoperable components.

OYSTER CREEK 3.4-6 Amendment No.: 355-r;3e,-r53,247 I

Appendix D Required Operator Actions Fort~ ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Intermediate Ranae Monitor (IRM) 11 fails low Scenario No.: NRC 1 Event No.:

~~

Initiation: Following Crew update by SRO on Containment Spray Pump 51A Cue: Annunciator G4e, IRM DSCL, and H7a, ROD BLOCK Time Terminus:

Position RO BOP SRO Applicant's Actions or Behavior 0

Responds to annunciator G4e (3F), IRM DSCL and H7a, ROD BLOCK (5F/6F) o Bypasses IRM 11 IAW procedure 402.4, IRM Bypass Operation (4F) o o

Verify alarms clear o

o Reports IRM 11 bypassed Reports IRM 11 is downscale and all other operable IRMs show normal indications Place IRM BYPASS joystick in the CH 11 position Verify IRM 11 HI-HI, HIGH, and DN SCL OR INOP lights illuminated Checks IRM 11 drawer (3R) and reports IRM 11 is downscale and all other operable IRMs show normal indications Verify IRM BYPASS light is illuminated on the SRM-IRM AUXILIARIES drawer (when bypassed) (3R) 0 Verifies Tech Spec 3.1.1, and directs IRM 11 bypassed Notifies Work Week Manager for repair IRM 11 is bypassed Notes/Comments NRC 1 Page 12 of 19

d L-J Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Scenario No.: NRC 1 Event

Description:

Electronic Pressure Reaulator (EPR) fluctuations Event No.: 8 Initiation: Following IRM 11 bypass Cue: Changes in reactor pressure and power (4F), Changes in EPR relay position indication (7F),

Changes in Turbine Bypass valve positions (7F)

I Time Terminus:

Position RO BOP SRO Applicants Actions or Behavior 0

Report change in reactor power and pressure 0

Reports change in EPR Relay position Enters ABN-9, Electronic Pressure Regulator Malfunction o

Transfers RPV pressure control to the Mechanical Pressure Regulator (MPR) IAW procedure 31 5.4, Transferring Pressure Regulators (7F)

. Lower the MPR setpoint by placing the MPR RELAY POSITION switch to the lower position (7%) for approximately 1 second periods until the MPR indicator moves in the direction of and reaches the EPR setting When the MPR is in control, TURN OFF the EPR power switch Raise RPV pressure to normal, as directed by the SRO Directs entry into ABN-9, Electronic Pressure Regulator Malfunction, to swap to the MPR Directs raising RPV pressure back to normal with the MPR Notifies Work Week Manager for repair The MPR has been placed in control of RPV pressure.

NoteslComments NRC 1 Page 13 of 19

Appendix D Required Operator Actions FOITYI ES-D-2 I Op-Test No.: NRC 2006-1 Scenario No.: NRC 1 Event No.: 9 Event

Description:

Failure of the MPR causina all Turbine Bvoass Valves to ao closed and electric ATWS I Initiation: After the MPR has been placed in control of RPV pressure

. Scrams the reactor 9

InsertsSRMs Reports control rods not inserted and reports reactor power Initiates Alternate Rod Insertion (ARI) (4F)

Bypasses ROPS (Reactor Overfill Protection System) (4F)

Maintain RPV water level in the band 138 - 175 TAF IAW Support Procedure 19 (attached)

Insert control rods IAW Support Procedure 21 ICT)

(attached) o Vent the Scram Air Header o

Manual Control Rod Insertion (4F)

Reports all control rods inserted 0

0 0

0 0

l 0

Sim.

Operator To vent the scram air header, insert LOA-CAS021 to 0 to close the scram air inlet valve and LOA-CAS024 to 1 to open the scram air vent valve.

To close the CRD charging header supply valve V-15-52, insert LOA-CRD024 to 0.

BOP 0

May report TBV indicate closed 0

0 0

0 0

0 Confirms EOP automatic actions IAW Support Procedure 1 (attached)

Prevents ADS by placing ADS Timer switches in Bypass (1F/2F)

Bypass MSlV Low-Low water level isolation IAW Support Procedure 16 (attached)

Bypass RBCCW Drywell Isolation IAW Support Procedure 18 (attached)

Trips recirc. pumps, as directed (3F)

Controls RPV pressure 800-1000 psig with Support Procedure 11/12 o EMRVs: Cycle EMRV control switch to MAN/AUTO to open/close the valve (1 F/2 F) o Isolation Condensers: Cycle IC Condensate Return Valves (1 F/2F)

NRC 1 Page 14 of 19

Appendix D Required Operator Actions Form ES-D-2 Terminus:

~

SRO May directs reactor scram and ABN-I, Reactor Scram, prior to reactor pressure scram setpoint (1045 psig)

Directs entry into RPV Control -With ANVS 0

0 0

0 0

0 0

0 0

0 Directs ARI initiation Directs ROPS bypassed Directs confirmation of automatic actions IAW Support Procedure 1 Directs ADS Timer switches in Bypass Directs bypass MSlV Low-Low water level isolation IAW Support Procedure 16 Directs bypass of RBCCW isolation IAW Support Procedure 18 Directs RPV water level band 138 - 175 TAF IAW Support Procedure 19 Directs Crew to insert control rods IAW Support Procedure 21 Directs 800 - 1000 psig RPV pressure with Isolation Condensers IAW Support Procedure 11 and EMRVs IAW Support Procedure 12 Directs recirc. pumps tripped Directs entry into RPV Control - No ATWS (following control rod insertions) o May direct ADS placed back to AUTO AI1 control rods have been inserted, and RPV water level and pressure are under control.

Notes/Comments W

NRC 1 Page 15 of 19

OYSTER CREEK GENERATING AmerGen-An E~eloll tcmp;lrcj STATION PROCEDURE Title

\\W REACTOR SCRAM 2.3 Other Indications Number ABN-1 Revision No.

2

1.

2.

Red scram lights lit on full core display.

Control rod positions indicate blank with green backlighting on the full core display.

Group scram solenoid lights noJ lit on 4F, 6R, 7R.

3.

3.0 OPERATOR ACTIONS If then EXECUTE this procedure concurrently with the appropriate EOP.

3.1 while executing this procedure, an entry condition for any EOP occurs, If then PERFORM the following:

a manual scram is to be performed and time permits, a

REDUCE recirculation flow to approximately 8.5 x I O4 gpm.

off line.

[

I

[

I a

NOTIFY the System Dispatcher that Oyster Creek will be taken

\\

4.0

OYSTER CREEK GENERATING STATION PROCEDURE AmerGen...

an fxelon Cornpry Title L'

REACTOR SCRAM 3.8 CONFIRM the following:

Main Turbine Tripped

[

I 0

GDI Open

[

I GCI Open

[

I Field breaker Open

[

I 3.9 If off-site power is available to both startup transformer, then EXIT this procedure and enter ABN-36, Loss of Off-site Power. [ ]

Number ABN-1 Revision No.

2

4.

When directed by the Unit Supervisor, PLACE the LFRV in automatic.

[

I 5.0

OYSTER CREEK GENERATING STATION PROCEDURE AmerGen++

An Exelon Ccqpany Title w

REACTOR SCRAM

6.

RESET the scram in accordance with step 3.12, if allowed.

Otherwise, CLOSE CRD Charging Header Supply, V-15-52 [ ]

Number ABN-1 Revision No.

2 6.0

OYSTER CREEK GENERATING AmerGen-An Exelor! Cornpafly STATION PROCEDURE Title L-'

REACTOR SCRAM RWCU system in the recirculation mode.

[

Number ABN-I Revision No.

2 RWCU system in the let down mode.

[

NOTE: Main Condenser must be available and ICs must not be isolated to utilize IC tube side vents for pressure control.

Isolation Condenser tube side vents.

[

NOTE: Resetting the scram will minimize the injection of cold water into the reactor bottom head from the CRD system and will relieve pressure from the control rod drives. (CM-I) 7.0

Procedure EMG-3200.01B Support Proc-19 Rev. 14 Attachment F Page 1 of 2 SUPPORT PROCEDURE 19 FEEDWATER/CONDENSATE AND CRD SYSTEM OPERATION 1.0 PREREQUISITES Directed to maintain RPV level with the Feed, Condensate and CRD Systems by the Emergency Operating Procedures.

2.0 PREPARATION None 3.0 PROCEDURE CAUTION A large power excursion and substantial core damage could result from a rapid increase in injection into the RPV.

3.1 Feedwater/Condensate 3.1.1 Confirm one Condensate Pump and one Feedwater Pump operating.

3.1.2 Control RPV Water level using the following:

Feedwater Regulating Valves MFRV Block Valves Feedwater Low Flow Valves Heater Bank Outlet Isolation Valves Feedwater and Condensate Pumps 3.1.3 9 RPV water level cannot be restored and maintained below 160 in.,

THEN close the Heater Bank Outlet Isolation Valve f o r the operating pump.

3.1.4 RPV water level reaches 170 in.,

THEN trip the operating feedwater pump.

OVER (2

0 001B/S 8 )

E6-1

Procedure EMG-3200.01B Support Proc-19 R e v. 14 Attachment F Page 2 of 2 3. 1. 5 E

RPV water l e v e l reaches 170 i n.,

AND RPV p r e s s u r e i s below 350 psig ( s h u t o f f head of t h e Condensate Pumps),

THEN c l o s e Heater Bank O u t l e t I s o l a t i o n Valves.

3.1.6 E

RPV water l e v e l continues t o i n c r e a s e,

THEN t r i p t h e o p e r a t i n g Condensate Pumps.

3.2 3. 2. 1 Control Rods cannot be driven u s i n g CRD CRD bypass flow i s d e s i r e d for RPV water l e v e l c o n t r o l,

THEN

1. S t a r t a l l a v a i l a b l e CRD Pumps.

2.

Open CRD Bypass I s o l a t i o n Valve, V-15-237 (RB 23 S E ).

3.

Monitor flow on Flow Gauge FI-225-2 (RB 23 SE) t h r o t t l e open CRD Bypass Valve V-15-30 (RB 23 SE) so a s not t o exceed 150 gpm flow f o r one o r two pump o p e r a t i o n.

3. 2. 2 E

RPV water l e v e l cannot be maintained less than 160 in.,

THEN 1. T h r o t t l e closed CRD Bypass Valve, V-15-30 (RB 23 SE), if opened.

2.

T h r o t t l e c l o s e d CRD Flow Control Valve.

3. 2. 3 E

RPV water l e v e l reaches 170 i n.,

AND t h e CRD System i s NOT r e q u i r e d f o r rod i n s e r t i o n,

THEN t r i p any o p e r a t i n g CRD pump.

E6-2

1.0

~~~~

~

~

~

~

SYMPTOMS ELECTRICAL HYDRAULIC INDICATOR Group scram solenoids ENERGIZED DE-ENERGIZED 2.0 Scram air header Individual scram lights (4F)

Scram Valves (at HCU) 3.0

--7 DEPRESSURIZED PRESSURIZED OFF ON CLOSED OPEN Procedure EMG-3200.01B Support Proc-21 Rev. 14 Attachment N Page 2 of 2.

SUPPORT PROCEDURE 21 ALTERNATE INSERTION OF CONTROL RODS PREREQUISITES Insertion of Control Rods utilizing alternate methods has been directed by the Emergency Operating Procedures.

PREPARATION None PROCEDURE 3.1 3.2 Manual Rod Insertion 3.1.1 Confirm all available CRD pumps are running.

3.1.2 Close Charging Header Supply valve V-15-52 (RB 23 SEI.

3.1.3 Place Reactor Mode Switch in REFUEL.

3.1.4 Place the ROD WORTH MINIMIZER keylock in BYPASS.

3.1.5 Close CRD DRIVE WATER PRESSURE CONTROL NC18 (Panel 4F) to maximize CRD drive water differential pressure.

3.1.6 CRD Bypass valve V-15-30 is open for RPV injection, THEN close CRD Bypass valve V-15-30 (RB 23 SE).

Identify whether an ELECTRICAL ATWS or an HYDRAULIC ATWS is in progress.

ELECTRICAL ATWS - Continue in Section 4.0 HYDRAULIC ATWS - Continue in Section 5.0 (20001B/S16)

OVER E14-1

u 4.0 Procedure EMG3200.01B Support Proc-21 Rev. 14 Attachment N Page 2 of 1 ELECTRICAL ATWS NOTE A l t e r n a t e methods of c o n t r o l rod i n s e r t i o n f o r e l e c t r i c a l ATWS may be performed i n any o r d e r o r c o n c u r r e n t l y as a p p l i c a b l e.

Vent t h e Scram A i r Header 4. 1 4. 2 4.3 4. 1. 1 Close Scram A i r Header i s o l a t i o n v a l v e V-6-175 (RB 23 SE).

4. 1. 2 Open Scram A i r Header d r a i n v a l v e V-6-409 (RB 23 SEI.

4.1.3 WHEN c o n t r o l rods a r e no longer moving i n,

THEN

1.

Close Scram A i r Header d r a i n v a l v e V-6-409.

2.

Open Scram A i r Header i s o l a t i o n v a l v e V-6-175.

De-energize t h e Scram Solenoids 4. 2. 1 E

MSIVs a r e OPEN, THEN p l a c e RPS Channel I and I1 Subchannel Test Keylocks I A, I B, I I A and I I B i n t h e T r i p p o s i t i o n (Panels 6R/7R).

WHEN t h e c o n t r o l rods a r e no l o n g e r moving i n,

THEN confirm t h e RPS Channel I and I1 Subchannel T e s t keylocks i n t h e NORMAL p o s i t i o n.

4. 2. 2 MSIVs are CLOSED, THEN p l a c e both 100 amp Main RPS breakers i n OFF (Panels 6R/7R).

t h e c o n t r o l rods are no l o n g e r moving i n,

confirm both 100 amp Main RPS b r e a k e r s i n ON.

THEN I n c r e a s e CRD Cooling Water D i f f e r e n t i a l Pressure 4.3.1 Confirm a l l a v a i l a b l e CRD pumps a r e running.

4.3.2 Close Charging Header Supply v a l v e V-15-52 (RB 23 SEI.

4. 3. 3 I n c r e a s e CRD c o o l i n g water d i f f e r e n t i a l p r e s s u r e by c l o s i n g CRD Cooling Water PCV NC40 (Panel 4 F ).

4.3.4 I F CRD Bypass v a l v e V-15-30 i s open f o r RPV i n j e c t i o n,

THEN c l o s e CRD Bypass valve V-15-30 (RB 23 SE).

E14-2

L-1.0 2.0 3.0 Procedure EMG-3200.01A Support Proc-1 Rev. 12 Attachment B Page 1 of 3 SUPPORT PROCEDURE 1 CONFIRMATION OF AUTOMATIC INITIATIONS AND ISOLATIONS PREREQUISITES Confirmation of automatic initiations and isolations has been directed by the Emergency Operating Procedures.

PREPARATION None PROCEDURE Confirm the following isolations/starts not required to by bypassed by the Emergency Operating Procedures:

SYSTEM

~

Reactor Isolation Scram Discharge Volume Isolation

~

320001A/S4)

OPERATING DETAILS IF -

THEN -

Any of the following conditions exist:

RPV water level at or below 86 i n. and not bypassed Steam tunnel temperature at or above 180F Any steam line flow at or above 4. 0 ~ 1 0 ~

l b d h r Reactor mode switch in RUN and RPV pressure at or below 850 psig Main steam line radiation at or above 800 units no ATWS condition exists Confirm closed the following:

IC VENTS RX SAMPLE MSIVs NS03A V-14-1, -19 V-24-30 (11F)

NSO4A V-14-5, -20 V-24-29 (11F)

NS03B DW AIR SUPPLY NS04B V-6-395 (11F)

~

IF A Reactor Scram is initiated AND SDV HI-HI LVL SCRAM switch is not in BYPASS, THEN Confirm closed the following:

NORTH SDV Vents & Drains SOUTH SDV Vents & Drains OVER E2-1

L-'

SYSTEM IC-A Isolation IC-B Isolation Cleanup System I sol at ion Shutdown Cooling System Isolation Isolation Condenser Initiation Core Spray System Start 32 00 01A/S4 )

Procedure EMG-3200.01A Support Proc-1 Rev. 12 Attachment B Page 2 of 3 OPERATING DETAILS IC-A is ruptured, 3

Confirm closed the following:

IC ISOLATION VALVES IC VENTS V-14-30

__ V-14-34 V-14-5 V-14 V-14-36 V-14 -2 0 I F IC-B is ruptured, THEN Confirm closed the following:

IC ISOLATION VALVES V-14-32

__ V-14-35 V-14 V-14-37 IC VENTS V-14-1 V-14-19

-~

~

~

I F Any of the following conditions exist:

0 RPV water level at or below 86 in. and not bypassed 0

Drywell pressure at or above 3.0 psig RWCU HELB Alarms THEN Confirm closed the following Cleanup Isolation valves:

V-16-1 V-16-14 V-16-2 V-16-61 IF Any of the following conditions exist:

RPV water level at o r below 86 in.

Drywell pressure at o r above 3.0 psig THEN Confirm closed the following SDC Isolation Valves:

V 54 V-17-19 IF RPV water level is at or below 8 6 in.,

THEN confirm initiation of both Isolation Condensers.

IF Any of the following conditions exist:

RPV water level at or below 86 in. and not bypassed.

w Drywell pressure at or above 3.0 psig and not bypassed.

THEN Confirm start of one Main Pump and one Booster Pump in each Core Spray System if not bypassed.

E2-2

4 SYSTEM Srimary Iontainment Isolation

( 3 2 0001A/S4 )

P r o c e d u r e EMG-3200.01A S u p p o r t Proc-1 Rev. 1 2 Attachment B Page 2 of 3 OPERATING DETAILS E

Any of t h e following c o n d i t i o n s e x i s t :

RPV water level a t or below 86 i n. and n o t bypassed.

Confirm closed t h e following valves t h a t are n o t r e q u i r e d t o be open by t h e Emergency Operating Procedures:

Drywell p r e s s u r e a t or above 3.0 psig and n o t bypassed.

E System DW Vent/Purge Torus Vent T o r u s 2" Vent Bypass DWEDT DW Floor Sump Torus/Rx Bldg.

Vacuum Breakers T I P Valves DW 2" Vent Bypass N2 Purge N2 Makeup v-21-2 V-27-3 V-21-4 V-28-17 V-28-18 V-28-47 Valve N o.

V-27-1 (Panel 11F)

I, 7 1 I

Panel 11F) 1 Panel 11F)

(Panel 11F) v-22-1 v-22-2 V-22 -2 8 (Panel 11F) 11 V-22-29 11 V-26-16 (Panel 11F)

V-26-18 11 Common Ind.

(Panel 1 l F )

V-23-21 (Panel 12XR)

V-2 3-2 2 t1 V-23-13 (Panel 12XR)

V-23-14

,I V-23-15

,7 V-23-16

,I V-23-17 (Panel 12XR)

V-23-18 1

V-23-19 11 V-23-20 r

E2-3

1.0 2.0 3.0

'U' procedure EMG-3200.01B Support Proc-16 Rev. 14 Attachment C Page 1. of 1.

SUPPORT PROCEDURE 16 BYPASSING MSIV LO-LO LEVEL ISOLATION INTERLOCKS PREREQUISITES Bypassing of the MSIV Lo-Lo level isolation interlocks has been directed by the Emergency Operating Procedures.

PREPARATION None PROCEDURE 3.1 3.2 3.3 3. 4 3. 5 Obtain four (4) bypass plugs from the EOP Tool Box in t h e Control Room.

Open the EOP BYPASS PLUGS panel in the rear of Panel 6R.

3.2.1 Insert a bypass plug in position BP1.

3.2.2 Insert a bypass plug in position BP2.

Open the EOP BYPASS PLUGS panel in the rear of Panel IR.

3.3.1 Insert a bypass plug in position BP1.

3.3.2 Insert a bypass plug in position BP2.

Place the ISOL SIGNAL BYPASS V-6-395 switch in BYPASS position (Panel 11F).

Inform the LOS that the MSIV LO-LO Level Isolation Interlock has been bypassed.

( 2 0 001B/S5 )

E3-1

.-\\---

2. 3 Procedure EMG-3200.01B Support Proc-18 Rev. 14 Attachment E Page 1 of 2 NOTE After completing the following steps, Valves V-5-147, V-5-166 and V-5-167 will only operate from the Control Room. All automatic operation is removed.

SUPPORT PROCEDURE 18 BYPASSING RBCCW ISOLATION INTERLOCKS 1.0 PREREQUISITES Bypassing of the RBCCW isolation has been directed by the Emergency Operating Procedures AND the RBCCW system is NOT isolated due to a LOCA or Main Steam Line Break (MSLB) in the Drywell.

2.0 PREPARATION 2.1 CAUTION Reinitiating RBCCW flow to the Drywell following a LOCA or MSLB in the Drywell may cause a water hammer to occur and subsequent piping failure.

Verify that the RBCCW System is not isolated due to high Drywell pressure/low RPV water level conditions.

2.2 E the RBCCW System is isolated due to high Drywell pressure/low RPV water level, In the rear of Panel 2R, open the EOP BYPASS PLUGS panel 2.3.1 Remove the bypass plug from position BP1.

2.3.2 Remove the bypass plug from position BP2.

( 2 0 0 OlB/S 7 )

OVER E5-1

>4 3.0 PROCEDURE 3. 1 Confirm open RBCCW Isolation Valves (Panel 1F/2F)

V-5-147 V-5-166 V-5-167 V-5-148 3.2 Start all available DW RECIRC FANS by placing their respective control switches in ON (Panel 11R).

3. 3 Place the ISOL SIGNAL BYPASS V-6-395 switch in BYPASS position (Panel 11F).

Procedure EMG3200.01B Support Proc-18 Rev. 14 Attachment E Page 2 of 2 E5-2

Required Operator Actions Form ES-0-2 Appendix D 3p-Test No.: NRC 2006-1 Event

Description:

Recirc Leak in the Primaw Containment Initiation: When all control rods are fully inserted Cue: Annunciator C3f, DW PRESS HVLO, Hld, H2d, DW PRESS HI-HI I & II, drywell temperature and Scenario No.: NRC 1 Event No.: 10 0

0 0

RPV water level control Report Primary Containment Control EOP entry and RPV Control - No ATWS EOP at 3 psig OW pressure Confirms EOP automatic actions IAW Support Procedure 1 (attached)

Lines-up Containment Spray IAW Support Procedure 29 (attached)

Initiates Drywell Spray IAW Support Procedure 29 (CT) (attached) o Uses FW (IAW Support Procedure 2, Feed and Condensate System Operation), and Core Spray (Support Procedure 9, Lineup for Core Spray System Injection) to control RPV water level above TAF (attached) o Lowers RPV pressure to allow Core Spray injection Terminus:

Notes/Co I I

NRC 1 ontrol-No ATWS lnd, Core Spray for RPV water level control IAW

' and Support Procedure 9 Containment Control EOP irmed IAW Support Procedure 1

'essure, directs lineup of Drywell Sprays IAW 3

tceeds 12 psig, direct initiation of Drywell Sprays Ire 29 pressure to allow low pressure injection systems pressure.

i; Page 16 of 19

1.0 2.0 3.0 Procedure EMG-3200.01A Support Proc-2 Rev. 12 Attachment C Page 1 of 1 SUPPORT PROCEDURE 2 FEED AND CONDENSATE SYSTEM OPERATION PREREQUISITES Directed to maintain RPV level with the Feed and Condensate Systems by the Emergency Operating Procedures.

PREPARATIONS None PROCEDURE 3. 1 3.2 3. 3 3.4 3. 5 3.6 IF -

THEN RPV water level is increasing, select one Feedwater Pump to be the operating pump and trip the Feedwater Pumps NOT selected.

Control RPV water level using the following:

Feedwater Regulating valves.

MFRV Block Valves.

Feedwater Low Flow valves.

Heater Bank Outlet Isolation valves.

Feedwater and Condensate pumps.

IF THEN IF THEN IF THEN IF THEN RPV water level cannot be restored and maintained below 160 in.,

close the Heater Bank Outlet Isolation valves for the operating pumps.

RPV water level reaches 170 in.,

trip any operating Feedwater pump.

RPV water level reaches 170 in.

AND RPV pressure is below 350 psig (the shutoff head of the Condensate pumps),

close Heater Bank Outlet Isolation valves.

RPV water level continues to increase, trip the operating Condensate Pumps.

(32 00 01A/S5)

E3-1

Procedure EMG-3200.01A S u p p o r t Proc-9 Rev. 12 Attachment J Page 1 of 3 SUPPORT PROCEDURE 9 LINEUP FOR CORE SPRAY SYSTEM INJECTION 1. 0 PREREQUISITES Confirmation of t h e l i n e u p f o r Core Spray System i n j e c t i o n has been d i r e c t e d by t h e Emergency Operating Procedures.

2.0 PREPARATION None 3. 0 PROCEDURE CAUTION Core Spray s u c t i o n s t r a i n e r plugging may occur due t o d e b r i s i n t h e Primary Containment and r e s u l t i n a loss of Core Spray System Flow.

3. 1 3. 2 3.3 3.4 Confirm open Core Spray System 1 and 2 Pump Suction Valves (Panel 1F/2F).

V-20-32 V-20-3 V-20-33 V-20-4 Confirm c l o s e d Core Spray System 1 and 2 Test Flow Return Valves i n each system (Panel 1F/2F).

V-20-27 V-20-26 CAUTION NPSH problems w i l l develop on a l l o p e r a t i n g pumps if more than 4 Containment Spray/Core Spray Main pumps a r e operated a t t h e same t i m e.

I F 4 Containment Spray/Core Spray Main pumps a r e i n o p e r a t i o n,

THEN s e c u r e Containment Spray pumps a s necessary t o run Core Spray pumps.

Confirm one Core Spray System Main Pump o p e r a t i n g i n each system (Panel 1F/2F).

SYSl SYS2

( 3 2 000 1A/S12

)

OVER E10-1

3.2 3.3 3.4 Procedure EMG-3200.02 Support Proc. 29 Rev. 17 Attachment G Page 2 of 3 W H E N THEN complete the following:

directed to initiate Drywell sprays,

1. Confirm all Reactor Recirculation Pumps tripped.

2. Confirm the Drywell Recirc Fans tripped (Panel 11R).

CAUT I ON NPSH problems will develop on all operating pumps if more t.han 4 Containment Spray/Core Spray Main pumps are operated at the same time.

IF 4 Containment Spray/Core Spray Main pumps are in operation, THEN do start additional Containment Spray pumps until Core Spray Main pumps can be secured.

Start a Containment Spray Pump as follows:

3.4.1 3.4.2 Select a Containment Spray Pump to be started.

Place and hold the System Pump Start Permissive Keylock for the selected pump in the appropriate position (Panel 1F/2F).

Start the selected Containment Spray Pump using its control switch (Panel 1F/2F).

3.4.3 3.5 Start an associated ESW Pump using its control switch (Panel 1F/2F).

3.6 (320002/9)

CAUTION 3peration of Containment Spray pumps with flow above the NPSH or vortex limits may result in equipment damage. When operating beyond any flow limits, periodic evaluations should be made to verify that continued operation beyond these limits is still required.

Monitor System parameters for expected performance.

E7-2

3.7 3.8 Procedure EMG-3200.02 Support Proc. 29 Rev. 17 Attachment G Page A of 2 Valves V-5-147, 166 and 167 do not seal in when the control switch is taken to CLOSE. The control switch must be held in CLOSE until the valve indicates closed.

Confirm the following RBCCW Isolation valves closed (Panel 1F/2F):

V-5-147 V-5-148 V-5-166 V-5-167 CAUTION Diesel generator overload will result if a Containment Spray pump and ESW pump are started with a Diesel Generator load of greater than 2160 KW.

Start additional Containment Spray and ESW Pumps in accordance with Steps 3.3 through 3.6 as directed by the LOS.

3.9 IF while performing the following steps, Containment Spray Pumps fail to trip, THEN place the respective system MODE SELECT switch in TORUS CLG position.

3.10 Maintain primary containment pressure in a band of 4 to 12 psig unless otherwise directed by the LOS as follows:

3.10.1 Secure Drywell Sprays when Drywell pressure drops to 4 psig.

3.10.2 WHEN Torus or Drywell pressure increases to 12 psig, THEN initiate Drywell Sprays in accordance with Steps 3.3 through 3.6.

3.11 -

IF any Core Spray Booster pump is running AN D Torus Drywell pressure drops to 2 psig, THEN confirm termination of Drywell Sprays due to NPSH concerns.

OVER (320002/9)

E7-3

Required Operator Actions Form ES-D-2 Appendix D Malfunction List Presets:

0 0

Cleanup Pump B out-of-service: PTL control switch IRM 17 failed: MAL-NISOlOH (and bypassed)

Event 1: Raise reactor power with control rods None Event 2: Transfer from LFRV to MFRV None Event 3: Failure of DW Pressure Instrument (PT-53) to calibrate None Event 4: Failure of EMRV NR-108A Acoustic Monitor 0

MAL-NSS026A to 120% EMRV NRO108A acoustic monitor fails upscale Event 5: In-service CRD flow control valve NC30A fails closed 0

MAL-CRDOOlA to 0 fails CRD FCV closed To place the alternate CRD FCV in service:

0 0

0 LOA-CRDO12 to 1 to open standby CRD FCV NC30B inlet valve LOA-CRDO1 1 to 1 to open standby CRD FCV NC30B outlet valve Note: No need to isolate the failed FVC Event 6: Loss of DC control power to Containment Spray Pump 51A BKR-CNS007 to FAIL CNTL mJSE Event 7: IRM 15 fails low MAL-NIS009E to 0 Event 8: EPR fluctuations 0

MAL-TCSO10 to 980 psig over 15 second ramp MAL-TCSO10 to 970 psig over 15 second ramp (one-two minutes after initial TCSOlO malfunction)

NRC 1 Page 17 of 19

Appendix D

~~

~

Required Operator Actions F0t-m ES-DP Event 9: Failure of MPR causing increased reactor pressure and TBVs fail closed plus Electric ATWS MAL-TCS008 to 1084 psig (raises reactor pressure) (MAY NOT BE NECESSARY)

MAL-TCS006A through TCSOO6I to 0 (individual malfunctions to close each TBV) BUT fail 1 and 2nd TBV (MAL-TCS006A and MAL-TCS006B over 12 minutes.

-&--A 0

Electric ATWS (CAED ATWS)

To vent the scram air header:

LOA-CAS021 to 0 (close air inlet valve) AND LOA-CAS022 to 1 (open vent valve)

To close the CRD charging header supply valve V-15-52 to insert control rods during ATWS:

LOA-CRD024 to 0 Event 10: Recirc LOCA inside primary containment 0

MAL-NSS004A recirc suction break to 5% over 10 minutes Set-UP Notes - IC-174

1.

2.

IRM-17 is bypassed

3.

4.

Have a marked-up copy of the startup procedure, 201 Cleanup Pump B is in PTL with a clearance tag on the panel control switch The EPR is in control NRC 1 Page 18 of 19

i i

J

References:

NRC 1 Page 19 of 19

Appendix D Event No.

Malf. No.

Event Type*

1 2

TS SRO 3

C BOP R

RO 4

C BOP TS SRO Ro 5

I 6

C RO 7

M Crew crew 8

9 C

Crew Scenario Outline Event Description Restore the amplidyne to service and transfer the main generator from manual voltage control to automatic voltage control.

Respond to Drywell-Torus vacuum breaker fail to close.

Respond to indications of low TBCCW cooling.

Respond to Reactor Recirculation Pump alarms.

Respond to failure of RPV water level input to Feed Water Level Control System (LT ID1 3A and LT ID1 3C).

Respond to failure of Average Power Range Monitor 7 (APRM).

Respond to loss of stator cooling.

Respond to steam leak in the primary containment with failure of the Containment Spray system.

Respond to failure of primary containment to automatically isolate.

Facility: Ovster Creek Scenario No.: NRC 2 Op Test No.: NRC 2006-1 Examiners:

Operators:

NRC 2 Page 1 of 19

Total Malfunctions (5-8):

Malfunctions after EOP entry (1 -2)

Abnormal Events (2-4)

Major Transients (1 -2)

EOPs entered requiring substantive actions (1 -2)

EOP Contingencies w/ substantive actions (0-2)

Critical Tasks (2-3)

Event Type Position 4

R RO 5

I RO Description Respond to recirculation pump alarms Respond to failure of RPV level input to FWLC (LT 6

ID13A and LT ID13C)

C RO Respond to failure of APRM 7 Event F Event 1

3 4

Type I Position I Description Type Position Description N

BOP Restore amplidyne to service C

BOP Respond to indications of low TBCCW cooling C

BOP Respond to recirculation pump alarms M

M C

Crew Crew Crew Respond to loss of stator water cooling Respond to steam leak in the drywell with failure of containment spray Respond to failure of drywell equipment drain and floor sump isolation valves to auto close on the primary containment isolation signal (V-22-1, V-22-2, V-22-28, and V-22-29)

NRC 2 Event 2

4 Page 2 of 19 Type Position Description TS SRO Respond to vacuum breaker failure (open) (TS TS SRO Respond to recirculation pump alarms 3.5.A.5)

Scenario Summarv The scenario will begin with field personnel continuing the drywell-torus vacuum breaker surveillance test. No actions are taken by control room operators except communicating with the field personnel and annunciation acknowledgement. The last valve is to be tested. Field personnel will continue with the surveillance after being notified by the control room that the test may reconvene (after placing the amplidyne in service).

1. The BOP will swap from main generator manual voltage control to automatic voltage control IAW 336.1, 24 KV Main Generator Electrical System. (NORMAL EVOLUTION)

2. Operators in the field are performing the DW/Torus vacuum breaker exercise test. The last vacuum breaker tested (V-26-14) opens with a little more resistance than the others, makes a strange noise when fully opened, then remains open after being allowed to close. Alarms of the open valve are present in the control room. The SRO will declare the valve inoperable and applies TS 3.5.A.5 (TS)

3. One of the operating TBCCW pumps will trip, and the low system pressure switch to start the standby pump will not function. The H2 System Trouble annunciator will alarm. The BOP will deduct the reduced system pressure and start another TBCCW pump IAW ABN-20, TBCCW Failure Response.

(COMPONENT FAILURE) (ABN)

4. The operators will respond to alarms for a recirculation pump. Alarms for low oil and high vibrations (RAP-E2d, RAP-E6d) will require an immediate pump trip by the BOP. The RO will make changes to power to ensure recirc pump limitations and power/flow considerations are met (recirc flow and/or CRAM rods). The SRO will apply TS 3.3.F, Recirc Loop Operability, and 3.1 O.A, Core Limits (as required by the Core Operating Limits Report). (COMPONENT FAILURE) (REACTIVITY MANIPULATION) (TS) (ABN)

5. The RO will respond to a leak in the common leg to RPV water level transmitters ID1 3A and ID1 3C which input to feed water level control. The crew will enter ABN-17, Feedwater System Abnormal Conditions. The RO will take manual control of feed water and return water level to the normal band. The RO will swap level transmitters to Feedwater Level Control IAW procedure 31 7, Feedwater System. (INSTRUMENT FAILURE)

6. The next event is an INOP failure of APRM 7 causing a '/2 scram. The RO will bypass the APRM and reset the Y2 scram. The SRO will verify compliance with Tech Specs. (COMPONENT FAILURE)

NRC 2 Page 3 of 19

7. The crew will receive annunciator STATOR CLG TROUBLE (R6c) and indications of the loss of one pump and auto start of the standby stator water pump. The standby pump will then trip causing a turbine runback. The Crew will implement ABN-11, Loss of Generator Stator Cooling, and will scram the reactor and perform ABN-1, Reactor Scram. (MAJOR) (ABN)

8. A steam leak in the primary containment occurs and the crew will enter the Primary Containment Control EOP. Because a drywell-to-torus vacuum breaker is open, the pressure suppression function of the torus is Iostlreduced. The Containment Spray System 1 will not function in the Containment Spray mode (Some of System 2 will function). The SRO will direct emergency depressurization (IAW Emergency Depressurization - No ATWS) as primary suppression pressure limits are approached. The SRO may also anticipate emergency depressurization and order a rapid RPV depressurization with TBVs and ICs. (MAJOR) (EOP) (COMPONENT FAILURE) (EOP CONTINGENCY)

9. The Operators will recognize that the drywell equipment drain isolation valves and drywell sump isolation valves failed to close on a primary containment isolation signal and will manually close the valves (V-22-1,V-22-2, and V-22-28, V-22-29) (COMPONENT FAILURE) (CT)

Critical Tasks

1.

With a failure of primary containment isolation on a valid isolation signal, manually isolate the primary containment.

With a Primary Containment isolation signal present and the failure of the Primary Containment to isolate, manual actions should occur to complete the isolation.

2.

Initiate Emergency Depressurization with EMRVS prior to exceeding Primary Suppression Pressure Limit (PSP), or when the Primary Containment bulk temperature cannot be maintained below 281 O F; or, anticipating ED and performing a rapid RPV depressurization before the limits are exceeded.

Performing an Emergency Depressurization (or performing an anticipatory ED rapid depressurization) ensures the Primary Containment does not fail.

NRC 2 Page 4 of 19

L-'

Appendix D Required Operator Actions FOIIII ES-D-2 Op-Test No.: NRC 2006-1 Scenario No.: NRC 2 Event No.: 1 Event

Description:

Shift from manual aenerator voltaqe control to automatic qenerator voltaae control IAW procedure 336.1 24 KV Main Generator Electrical Svstem, Section 6.0 Initiation: Following shift turnover Cues: As directed by the SRO Applicant's Actions or Behavior 0

Direct the BOP to shift from manual generator voltage control to automatic generator voltage control IAW procedure 336.1, 24 KV Main Generator Electrical System, Section 6.0.

Shift from manual generator voltage control to automatic generator voltage control IAW procedure 336.1,24 KV Main Generator Electrical System, Section 6.0 (8F/9F)

(attached) 0 0

Makes plant announcement prior to starting the amplidyne Shifts voltage control IAW procedure 336.1, section 6.0. (attached)

After placing the amplidyne in auto, reports the amplidyne is in service Role Play I As the NLO, when directed to verify the amplidyne running, state the amplidyne is running.

I I

Terminus: I The amplidyne is in automatic control.

Notes/Comments NRC 2 Page 5 of 19

OYSTER CREEK GENERATING STATION PROCEDURE AmerGen.

hn lxclon Company Number 336.1

,-?

6.0 SHIFTING FROM MANUAL TO AUTOMATIC VOLTAGE REGULATION 6.1 START AMPLIDYNE motor as follows:

I Title

'4' 24 KV Main Generator Electrical System 6.1.I NOTIFY PECO Generation Dispatcher that the Main Generator to be place in Automatic Voltage Control.

NOTIFY the Power Team that the Main Generator is to be placed in Automatic Voltage Control.

[

I

[

I 6.1.2 Revision No.

45 6.1.3 Simultaneously PLACE the AMPLIDYNE CONTROL 43CS switch in TEST and OBSERVE 1Al bus ammeter for an amp deflection (Panel 8F/9F).

[

I

[

I 6.1.4 CONFIRM amplidyne motor started as indicated by the amp deflection on I A I bus (Panel 8F/9F).

[

I 6.1.5 Locally VERIFY the AMPLIDYNE motor is running.

6.2 ZERO the AMPLIDYNE Voltmeter as follows (Panel 8F/9F):

6.2.1 6.2.2 6.2.3 E the needle is deflected,

[

I

[

I PRESS and HOLD the AMPLIDYNE VOLTS Low Range pushbutton.

OBSERVE the AMPLIDYNE Voltmeter for a zero or center scale reading.

THEN ADJUST the voltage using the AMPLIDYNE ADJUST rheostat to return the regulator output to a zero or center scale reading.

[

]

6.2.4 RELEASE the AMPLIDYNE VOLTS Low Range pushbutton.

[

]

6.2.5

[

I VERIFY the AMPLIDYNE Voltmeter reads zero or center scale.

6.2.6 a zero or center scale reading is

obtained,

[

I THEN REPEAT Substeps 6.2.1 through 6.2.5 above as necessary.

9.0

OYSTER CREEK GENERATING I

STATION PROCEDURE AmerGen.

An Fxcion Company Number 336.1 6.3 E

a zero or center scale reading cannot be obtained in AMPLIDYNE Volts High range, THEN LEAVE the AMPLIDYNE regulator out of service and manually CONTROL excitation.

PLACE the AMPLIDYNE CONTROL 43CS switch to ON, which leaves the AMPLIDYNE controlling excitation.

REMOVE VOLTAGE CONTROL IN MANUAL operator aids installed per Section 5.0.

[

I

[

I

[

I 6.4 6.5 Title 24 KV Main Generator Electrical System Revision No.

45 CONTROL terminal voltage or excitation level of the generator by adjusting the AMPLIDYNE ADJUST rheostat as required in accordance with Section 3.0.

[

I NOTE Raise direction of the 70M switch will move the indicator in the

" B U C K d i recti on.

6.6 I

I NOTE Terminal voltage of the main generator is limited to the range of 23.3 to 24.7 KV (Reference TDR 1136).

6.7 MAINTAIN a slight BUCK reading by adjusting EXCITER FIELD RHEOSTAT CONTROL 70M switch as needed, which ensures the Main Field will pick up control should the AMPLIDYNE trip.

I 1

10.0

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

DW/Torus vacuum breaker V-26-14 will not close after beina manuallv oDened locallv for surveillance.

Scenario No.: NRC 2 Event No.: 2.

Initiation: The amplidyne has been placed in automatic.

Cues: Notification from in-plant Operator.

V-26-14). IAW the surveillance dure, Step 6.3.12, you will receive two S/DW I VAC BRKR OPEN and C5f, Notifies ln-Plant Operator when annunciators C4f and C5f (TORUWDW VAC BRKR OPEN I and I I ) are received Sim.

When the Role Play in-plant operator is ready to open the last vacuum breaker, V-26-14, then open Operator the valve.

Role Play As the in-plant Operator, call the Control Room and report: vacuum breaker V-26-14 open, but will not close and cannot be closed.

Torus/DW Vacuum Breaker V-26-14 will not close Declares the vacuum breaker inoperable Review Tech Spec 3.5.A.5.a (attached) o 3.5.A.5.a: When primary containment is required, all suppression chamber-drywell vacuum breakers shall be OPERABLE except during testing and as stated in Specification 3.5.A.5.b and c, below.

3.5.A.5.b: Five of the fourteen suppression chamber - drywell vacuum breakers may be inoperable provided that they are secured in the closed position.

3.5.A.5.d: If Specifications3.5.A.5(a), (b) or (c) can not be met, the reactor shall be PLACED IN the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Refers to OP-OC-100, Oyster Creek Conduct of Operations o

o o

Notifies Work Week manager for repair 0 Notifies plant management about TS required shutdown Updates the Crew NRC 2 Page 6 of 19

Appendix D Required Operator Actions Form ES-D-2 Notes/Comments I

'W' NRC 2 Page 7 of 19

5.

Pressure Suppression Chamber - D w e l l Vacuum Breakers

a. When primary containment is required, all suppression chamber-drywell vacuum breakers shall be OPERABLE except during testing and as stated in Specification 3.5.A.5.b and c, below. Suppression chamber - drywell vacuum breakers shall be considered OPERABLE if (I)

The valve is demonstrated to open from closed to fully open with the applied force at all valve positions not exceeding that equivalent to 0.5 psi acting on the suppression chamber face of the valve disk.

(2) The valve disk will close by gravity to within not greater than 0.1 0 inch of any point on the seal surface of the disk when released after being opened by remote or manual means.

(3) The position alarm system will annunciate in the control room if the valve is open more than 0.10 inch at any point along the seal surface of the disk.

b. Five of the fourteen suppression chamber - drywell vacuum I

breakers may be inoperable provided that they are secured in the closed position. With one of the nine required suppression chamber - drywell vacuum breakers inoperable, restore one vacuum breaker to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

c. One position alarm circuit for each OPERABLE vacuum breaker may be inoperable, provided that each OPERABLE suppression chamber - drywell vacuum breaker with one defective alarm circuit, and associated remaining position alarm circuit are verified to be OPERABLE immediately, and monthly in accordance with 4.5.F.5.a. Additionally, a daily verification using the OPERABLE position alarm circuit that the affected vacuum breaker is closed shall be performed.

d. If Specifications3.5.A.5(a), (b) or (c) can not be met, the reactor shall be PLACED IN the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

6.

After completion of the startup test program and demonstration of plant electrical output, the primary containment atmosphere shall be reduced to less than 4.0% O2 with nitrogen gas within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the reactor mode selector switch is placed in the RUN MODE.

Primary containment deinerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a scheduled shutdown.

7.

Deleted.

OYSTER CREEK 3.5-4 Amendment No.:

,230

Required Operator Actions Form ES-D-2 Appendix D

'W'

\\-e-Op-Test No.: NRC 2006-1 Event

Description:

The runninq TBCCW pumD trips (without trip annunciation). combined with the failure of standbv TBCCW Durn13 to auto start (PS-116 fails to trip)

Initiation: TS 3.5.A.5 has been reviewed and the Crew has been updated.

Cues: Annunciator R6b, H2 SYSTEM TROUBLE (8F/9F)

Scenario No.: NRC 2 Event No.: 3 Time Role Play Sim.

Operator Terminus:

Position BOP SRO Applicant's Actions or Behavior Responds to Annunciator R6b, H2 SYSTEM TROUBLE e

Request NLO investigate the Hydrogen Cooling Control Cabinet (TB Basement)

Reports TBCCW Pump 1-1 has tripped Reports the standby TBCCW Pump did not auto start on low TBCCW discharge pressure; and, starts the standby TBCCW Pump 1-2 (13R)

. Places TBCCW Pump 1-2 switch to START

. Verifies TBCCW pump start and system pressure Enters ABN-20, TBCCW Failure Response, if directed.

e e

c May direct entry into ABN-20, TBCCW Failure Response Notifies Work Week Manager about the failed TBCCW pump and the failure of the standby TBCCW to auto start As the NLO investigating the Hydrogen Cooling Control Cabinet, report: Machine gas temperature is 57" C and steady, and machine gas pressure is 48 psig and steady, and hydrogen purity is 99%

and steady.

As the NLO, if asked to check the Stator Water Cooling Panel, report that there are no alarms on that panel.

As the NLO, if asked to check TBCCW Pumps, report that only TBCCW Pump 1-3 is running.

A minute after the standby TBCCW Pump is started, delete the annunciator for H2 System Trouble (R6b).

The standby TBCCW Pump has been manually started.

Notes/Comments LJ' NRC 2 Page 8 of 19

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Reactor Recirculation Pump A abnormal oDeration rewiring pump shutdown (low oil level and hiah vibration]

Scenario No.: NRC 2 Event No.: 4 Initiation: Following TBCCW evolution Cues: Annunciators E6d, OIL LEVEL HVLO, and E2d, VIBRATION HI A Time Position BOP RO SRO Applicant's Actions or Behavior Responds to Annunciators VIBRATION HI A E2d, and OIL LEVEL HVLO E6d, 0

Annunciator VIBRATION HI A, E2d o

Attempts to reset vibration alarm (3F) 0 Annunciator OIL LEVEL HI/LO, E6d o

Reports that immediate pump trip is required IAW the RAP-E6d 0

Manually trips Reactor Recirc. Pump A o

Place Recirc Pump A DRIVE MOTOR switch to STOP position (3F)

Refers to ABNQ, Recirculation System Failures, Section 3.1.2 (attached) o Close the Pump A DISCHARGE valve (3F) o Selects an operating recirc loop temperature point (3F) o Monitors for fuel failures Changes reactor pressure, if directed, by changing the EPR setpoint (7F)

(3F) 0 0

Refers to ABN-2, Recirculation System Failures Verifies operation NOT within the Exclusion Zone Reports recirc. flow > 8.5 E4 GPM and pump speed > 33 Hz 0

Reduces reactor power IAW ABN-2, Section 3.1.2.E (attached) o Lowers reactor power with recirc flow to 8.5 E4 GPM OR until pump speed e 33 Hz (4F) o With recirc speed still > 33 Hz, lowers reactor power to about 55% power by inserting CRAM rods (4F) o Once at about 55% power, lower recirc pump speed to e 33 Hz Verifies position on Power Operations Curve Reports having entered the Buffer Zone on the Power Operations Curve o Maintains heightened awareness of plant parameters 0

Directs trip of Reactor Recirc. Pump A, IAW RAPS 0

Directs reactor power reduction IAW ABN-2 Refers to Tech Spec 3.3.F, 3.10.A (3.3.F attached)

NRC 2 Page 9of 19

Appendix D Required Operator Actions FOITTI ES-D-2 Notify System OwnedDispatcher, Chemistry, Reactor Engineering Notes/Comments NRC 2 Page 10 of 19

1 OYSTER CREEK GENERATING STATION PROCEDURE An Exelmi Compaly Number ABN-2

2.

If 4 or 5 recirculation loops were operating when a pump trips, I

Title Recirculation System Failures Then PERFORM the following:

Revision No.

6 A.

CONFIRM open the DlSCH BYPASS valve for the CLOSE the DISCHARGE valve for the tripped CONFIRM the DISCHARGE valve for the tripped tripped pump.

[

I pump.

[

I pump closes in approximately 2 minutes.

[

I B.

C.

NOTE: Step 3.1.2.D may be performed concurrently with Steps 3.1.2.E through 3.1.2.H.

D.

If the DISCHARGE valve can not be closed, Then PERFORM the following:

1.

CLOSE the pump SUCTION valve.

2.

CLOSE the DISCHARGE valve using Attachment ABN-2-1.

3.

If Discharge valve closure is successful Then REOPEN the affected pump suction valve to place the loop in an IDLE condition.

[

I

[

I

[

I 9.0

Amerkk 1 OYSTER CREEK GENERATING STATION PROCEDURE hn fxrioii Company Number ABN-2 I

I Title Recirculation System Failures r

Revision No.

6 I

CAUTION Prolonged operation of Reactor Recirculation pumps above E.

I NOTE Procedure 301.2, Section 9.0 contains guidance for local manual operation of a Recirculation Pump MG-Set Pneumatic Drive Control, including operation with a disconnected linkage rod.

~~

~

~

If 3 loops are in operation following the recirc pump trip, Then PERFORM the following:

1.

If the Exclusion Zone has been

entered, Then EXIT the Exclusion Zone by raising pump speed to a maximum of 33 HZ andlor inserting the CRAM array per the CRAM Rod Move Sheet of the approved Control Rod Sequence Package to lower power to ~ 2 5 %.

2.

If total recirc flow >8.5 E4 gpm and recirc pump speed >33 Hz, Then LOWER total recirc flow to 8.5 E4 gpm or until recirc pump speed <33 Hz.

3.

If recirc pump speed >33 Hz, Then PERFORM the following:

[

I 11.0

I OYSTER CREEK GENERATING STATION PROCEDURE AI7 fXe?G:l CGmpdny Number ABN-2

a.

LOWER reactor power to approximately 55% using the CRAM array per the CRAM Rod Move Sheet of the approved Control Rod Sequence Package or as directed by Reactor Engineering.

[

I I

Title Recirculation System Failures

b.

LOWER recirc pump speed to ~ 3 3 Hz.

[

I Revision No.

6 F.

VERIFY the plotted point on the Power Operation Curve.

[

I I.

If operating in the Buffer Zone of the Power Operations Curve (CM-I),

Then PERFORM the following:

a.

VERIFY THE Exclusion Zone is not entered.

[

I

b.

MAINTAIN a heightened awareness of plant parameters.

[

I

2.

If the Exclusion Zone has been entered, (CM-I)

Then EXIT immediately using rods or flow.

[

I

3.

REFER to Technical Specifications Sections 3.3.F., 3.1 O.A.

[

I G.

CONFIRM at least one of the RECIRC PUMP SUCTION TEMPS indicators is selected to an operating loop.

[

I 12.0

E.

Reactor Coolant Oualitv

'Lc' 1,

The reactor coolant quality during power operation with steaming rates to the turbine-condenser of less than 100,000 pounds per hour shall be limited to:

conductivity chloride ion 0.1 ppm 2 us/cm[s=mhos at 25°C (77"FI

2.

When the conductivity and chloride concentration limits given in 3.3.E.I are exceeded, an orderly shutdown shall be initiated immediately, and the reactor coolant temperature shall be reduced to less than 212°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The reactor coolant quality during power operation with steaming rates to the turbine-condenser of greater than or equal to 100,000 pounds per hour shall be limited to:

3.

conductivity 10 uS/cm

[S=mhos at 25OC (77"F)I chloride ion 0.5 ppm When the maximum conductivity or chloride concentration limits given in 3.3.E.3 are exceeded, an orderly shutdown shall be initiated immediately, and the reactor coolant temperature shall be reduced to less than 2 12°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.

5.

During power operation with steaming rates on the turbine-condenser of greater than or equal to 100,000 pounds per hour, the time limit above 1.O uS/cm at 25OC (77OF) and 0.2 ppm chloride shall not exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for any single incident.

6.

When the time limits for 3.3.E.5 are exceeded, an orderly shutdown shall be initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

F.

Recirculation LOOP Operability

1.

During POWER OPERATION, all five recirculation loops shall be OPERATING except as specified in Specification 3.3.F.2.

POWER OPERATION with a maximum of two IDLE RECIRCULATION LOOPS or one IDLE RECIRCULATION LOOP and one ISOLATED RECIRCULATION LOOP is permitted. The reactor shall not operate with two ISOLATED RECIRCULATION LOOPS.

2.

a.

With one ISOLATED LOOP the following conditions shall be met:

1.

The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) as a function of average planar exposure, at any axial location shall not exceed 98% of the limits specified in

3. I O.A. The action to bring the core to 98% of the APHLGR limits shall be completed prior to isolating the recirculation loop.

OYSTER CREEK AmendmentNo: 42,93, 135, 140,212

\\

3.3-3 Corrected Letter dated 8/7/2000

u

2.

The circuit breaker of the recirculation pump motor generator set associated with an ISOLATED RECIRCULATION LOOP shall be open and defeated from operation.

3.

An ISOLATED RECIRCULATION LOOP shall not be returned to service unless the reactor is in the COLD SHUTDOWN condition.

b.

When there are two inoperable recirculation loops (either two IDLE RECIRCULATION LOOPS or one IDLE RECIRCULATION LOOP and one ISOLATED RECIRCULATION LOOP) the reactor core thermal power shall not exceed 90% of rated power.

3.

If Specifications 3.3.F.1 and 3.3.F.2 are not met, an orderly shutdown shall be initiated immediately until all operable control rods are fully inserted and the reactor is in either the REFUEL MODE or SHUTDOWN CONDITION within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.

With reactor coolant temperature greater than 2 12°F and irradiated fuel in the reactor vessel, at least one recirculation loop discharge valve and its associated suction valve shall be in the full open position.

5.

If Specification 3.3.F.4 is not met, immediately open one recirculation loop discharge valve and its associated suction valve.

6.

With reactor coolant temperature less than 212°F and irradiated fuel in the reactor vessel, at least one recirculation loop discharge valve and its associated suction valve shall be in the full open position unless the reactor vessel is flooded to a level above 185 inches TAF or unless the steam separator and dryer are removed.

OYSTER CREEK Amendment No: 135, 140,212

.l../-,

3.3-3a Corrected Letter dated 8/7/2000

Required Operator Actions Form ES-D-2 Appendix D Op-Test No.: NRC 2006-1 Scenario No.: NRC 2 Event No.: 5 Event

Description:

Leak in variable leq to RPV water level transmitters ID1 3A & ID1 3C (inwts into Feedwater Level Control Svstem, FWLC)

~

~

initiation: Reactor power is stable and notifications of the down-power event are made.

Cues: RPV water level indicators ID13A and ID1 3C both lowering (5F/6F), and other RPV water level indicators rising (5F/6F) rising and feedwater flow rising (5F/6F); followed by annunciator J ~ c,

FCWRFCS; followed by annunciator H7e, RX LVL HI/LO (5F/6F);

Time Terminus:

Position RO BOP SRO Applicants Actions or Behavior Reports RPV water level indicators ID1 3A and ID1 3C are lowering and other RPV water level indicators are rising and FW flow is rising (5F/6F)

Refer ABN-17, Feedwater System Abnormal Conditions, Section 3.2.1 0

Changes the Level Select into Feedwater Level Control IAW procedure 317, section 11.8 (attached) (4F) o When the RPV water level input has been changed:

Monitor feedwater flow and RPV water level Reports Level Control is selected to B and that RNIRPV water level are rewondina Responds to annunciator J8c (FCS/RFCS) and H7e(RX LVL HI/LO)

Checks RPV water level indication (5F/6F) o 0 Checks DW bulk temperature (PPC)

Checks DW pressure (4F, 1F/2F) 0 Refer to ABN-17, Feedwater System Abnormal Conditions 0

Directs entry into ABN-17, Feedwater System Abnormal Conditions 0

Directs changing the Level Select input into Feedwater Level Control IAW procedure 317, Section 11.8 (from ABN-17)

Notifies Work Week manager for repair Reports RPV water level indicators ID1 3A and ID1 3C are rising and other RPV water level indicators lowering ID13B is the selected RPV water level input to FWLC and RPV water level and FW flow are athear normal values.

Notes/Comments I

1 NRC 2 Page 11 of 19

OYSTER CREEK GENERATING SJAJJON PROCEDURE

AmerGen, A r Exem Company

'-?-

Title Feedwater System (Feed Pumps to Reactor Vessel)

Revision No.

74 11.8.1 NOTE In a transient, the feed system continues to use the last Auto Signal seen until a manual adjustment is made.

PLACE the MASTER FEEDWATER LEVEL CONTROLLER in MAN.

[

I 11.8.2 PLACE the LEVEL TRANSMITTER SELECTOR in the desired position (A, AUTO OR B).

[

I 11.8.3 SELECT the S display on the MASTER FEEDWATER LEVEL CONTROLLER.

[

I 11.8.4 MATCH the S display digital readout to the P display digital readout on the MASTER FEEDWATER LEVEL CONTROLLER. [ ]

I I

.8.5 WHEN the S display digital readout and the P display readout are equal, THEN PLACE the MASTER FEEDWATER LEVEL CONTROLLER in AUTO.

[

I 11.8.6 MONITOR Reactor level and feedwater flow for any changes 11.8.7 MAINTAIN Reactor level at 160 in TAF or as directed by the OS by adjusting the MASTER FEEDWATER LEVEL CONTROLLER setpoint.

64.0

Appendix D Required Operator Actions Form ES-D-2 u

J Op-Test No.: NRC 2006-1 Event

Description:

INOP failure of APRM 7 Scenario No.: NRC 2 Event No.: 6 Initiation: 2 minutes after ID1 36 is the selected RPV water level input to FWLC and RPV water level and FW flow are athear normal values.

Cues: Annunciators GZf, APRM HI-HI/INOP II (3F); Glc, SCRAM CONTACTOR OPEN (3F)

I Time Terminus:

Position RO BOP SRO Applicants Actions or Behavior Responds to annunciators GZf, APRM HI-HVINOP II (3F); Glc, SCRAM CONTACTOR OPEN (3F) o Reports DN SCL INOP light ON for APRM CH 7 and normal reading on the other operable APRMs Places APRM BYPASS joystick to the CH 7 position Presses SCRAM SYSTEM RESET pushbutton Bypasses APRM 17 (3F) o Resets the /2 scram (3F) o Checks APRM cabinets (5R) 0 Reviews Tech Spec 3.1 0

Directs APRM 17 bypassed Directs /2 scram reset Notifies Work Week Manager for repair o

Declares APRM 7 inoperable APRM 7 is bypassed.

Notes/Comments NRC 2 Page 12 of 19

Appendix D

~~

~

~

Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Loss of Stator Water Coolina PumDs Initiation: 2-3 minutes after APRM 7 is placed in bypassed.

Scenario No.: NRC 2 Event No.: z Cues: Annunciator R6c, STATOR CLG TROUBLE (7F)

Time -

Role Play Terminus:

P Position BOP I

Applicant's Actions or Behavior Respond to annunciator R6c, STATOR CLG TROUBLE o

o Monitors stator temperature o

o Following the scram, maintains reactor pressure 800-1000 with TBVs (or as directed)

Dispatch NLO to the Stator Water Cooling Panel (Role Play)

Enters ABN-11, Loss of Generator Stator Cooling Reports generator runback (TBVs opening)

As the NLO sent to the Stator Water Cooling Panel, report that the running Stator Water Cooling Pump has tripped, and the standby stator cooling pump has started. (About 2 minutes later, trip the second pump as the SIM OPERATOR) and report that the second Stator Water Cooling Pump also tripped and that neither pump can be started.

SRO RO BOP Directs entry into ABN-11, Loss of Generator Stator Cooling Directs the RO to scram the reactor and perform ABN-1, Reactor Scram, due to generator runback Directs entry into RPV Control - No ATWS EOP on low RPV water level o Directs RPV water level 138-175" with feedwaterhondensate IAW Support Procedure 2 (attached) o Directs RPV pressure 800-1000 psig with TBVs May direct a reactor cooldown at below 100" F/Hr Manually scrams the reactor and carries out ABN-1, Reactor Scram, Section 3.2 through 3.7, and 3.10 (attached) o Scrams the reactor o Inserts SRMs and IRMs Performs ABN-1, Reactor Scram actions, Section 3.8, and 3.1 1 (attached)

The reactor is scrammed; RPV Control - No ATWS EOP has been entered and RPV water level and RPV pressure are being controlled in the desired band.

NRC 2 Page 13 of 19

Appendix D Required Operator Actions F

o

~

ES-D-2 NotedComments L../*'

I NRC 2 Page 14 of 19

AmerGen.

OYSTER CREEK GENERATING 1

STATION PROCEDURE Number ABN-1 I

Revision No.

Title u

2 REACTOR SCRAM 2.3 Other Indications

1.

Red scram lights lit on full core display

2.

Control rod positions indicate blank with green backlighting on the full core display.

3.

Group scram solenoid lights not lit on 4F, 6R, 7R.

3.0 OPERATOR ACTIONS If while executing this procedure, an entry condition for any EOP occurs, then EXECUTE this procedure concurrently with the appropriate EOP.

3. I If a manual scram is to be performed and time permits, then PERFORM the following:

REDUCE recirculation flow to approximately 8.5 x I O 4 gpm.

[

I NOTIFY the System Dispatcher that Oyster Creek will be taken off line.

[

I 4.0

OYSTER CREEK GENERATING STATION PROCEDURE AmerGen*

.ln ixelori Co?lpay Title ii REACTOR SCRAM 3.8 CONFIRM the following:

0 Main Turbine Tripped

[

I 0

GDI Open

[

I 0

GCI Open

[

I 0

Field breaker Open

[

I 3.9 If then EXIT this procedure and enter ABN-36, Loss of Off-site Power. [ ]

off-site power is not available to both startup transformer, Number ABN-1 Revision No.

2

4.

When directed by the Unit Supervisor, PLACE the LFRV in automatic.

[

I 5.0

OYSTER CREEK GENERATING I

STATION PROCEDURE Number ABN-1

6.

RESET the scram in accordance with step 3.12, if allowed.

Otherwise, CLOSE CRD Charging Header Supply, V-15-52 [ ]

I Title REACTOR SCRAM 6.0 Revision No.

2

OYSTER CREEK GENERATING AmerGen..

Pn ix&n Compm STATION PROCEDURE Title

..-.-.J RWCU system in the recirculation mode.

RWCU system in the let down mode.

NOTE: Main Condenser must be available and ICs must not be isolated to utilize IC tube side vents for pressure control.

Isolation Condenser tube side vents.

[

[

NOTE: Resetting the scram will minimize the injection of cold water into the reactor bottom head from the CRD system and will relieve pressure from the control rod drives. (CM-1)

Number ABN-I Revision No.

2 7.0

1.0 2.0 3.0 Procedure EMG-3200.01A Support Proc-2 Rev. 12 Attachment C Page 1. of 1 SUPPORT PROCEDURE 2 FEED AND CONDENSATE SYSTEM OPERATION PREREQUISITES Directed to maintain RPV level with the Feed and Condensate Systems by the Emergency Operating Procedures.

PREPARATIONS None PROCEDURE 3.1 3.2 3.3 3.4 3.5 3.6 IF RPV water level is increasing, THEN select one Feedwater Pump to be the operating pump and trip the Feedwater Pumps NOT selected.

Control RPV water level using the following:

Feedwater Regulating valves.

MFRV Block Valves.

Feedwater Low Flow valves.

Heater Bank Outlet Isolation valves.

Feedwater and Condensate pumps.

IF THEN IF THEN IF THEN IF THEN RPV water level cannot be restored and maintained below 160 in.,

close the Heater Bank Outlet Isolation valves f o r the operating pumps.

RPV water level reaches 170 in..

trip any operating Feedwater pump.

RPV water level reaches 170 in.

AND RPV pressure is below 350 psig (the shutoff head of the Condensate pumps),

close Heater Bank Outlet Isolation valves.

RPV water level continues to increase, trip the operating Condensate Pumps.

( 3 2 0 0 0 1A/S5)

E3-1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Steam leak in the Primaw Containment with failure of Containment Sorav Svstem

1. 1 and failure of Containment SDrav Pumo 51 C, which leads to Emerqencv Depressurization of the RPV due to PSP Concerns Scenario No.: NRC 2 Event No.: 8/9 Initiation: The reactor and turbine have tripped; RPV water level is inhear the normal band Cues: Increasing DW pressure and temperature; Annunciators C3f DW PRESS HVLO, followed by Hld, H2d, DW PRESS HI-HI I & 11 Time

~~~

Position SRO Applicant's Actions or Behavior Responds to annunciator C3f, DW PRESS HVLO Reports hi containment pressure and entry into Primary Containment Control EOP and RPV Control - No ATWS Monitors/reports Primary Containment parameters Confirms automatic actions, IAW Support Procedure 1 (attached) o Reports that Drywell Equipment Drain Isolation Valves, V-22-1 and V-22-2, and DW Sump valves V-22-28 and V-22-29 did not close, and closes the valves Place EQUIP SUMP switches to the CLOSE position (11F) (CT)

Places DW SUPM switches to the CLOSE position (11F) (CT)

Lines-up and sprays the DW IAW Support Procedure 29 (attached)

Reports that PSP is rising For Emergency Depressurization: (CT2) o Stops injection with Core Spray not required for adequate core cooling IAW Support Procedure 10 (1 F/2F) (attached) o Bypass ROPS (Reactor Overfill Protection) (4F) o Report Torus water level > 9 0 (1 F/2F) o Open 5 EMRVs by placing AUTO DEPRESS VALVE switches in the MAN position (1 F/2F)

Initiates torus cooling, as directed, IAW Support Procedure 25 (attached)

(1 F/2F)

Directs entry into Primary Containment Control EOP and re-enters RPV Control o

Directs automatic actions confirmed IAW Support Procedure 1 o

Before DW/Torus reaches 12 psig, directs lineup of Drywell Sprays IAW Support Procedure 29 o

When DW/Torus exceeds 12 psig, or before bulk DW temperature reaches 281" F, directs initiation of Drywell Sprays IAW Support Procedure 29 o

When it has been determined that bulk DW temperature cannot be restored/maintained below 281 O F or it has been determined that Torus pressure cannot be maintained below Pressure Suppression Pressure, then

- NO ATWS NRC 2 Page 15 of 19

Appendix D Required Operator Actions F O I ~

ES-D-2 Terminus:

direct Emergency Depressurization (CT)

Enters EMG-3200-04A, Emergency Depressurization - No ATWS

=

. Directs ROPS bypass

=

Directs 5 EMRVs open May direct torus cooling be placed into service IAW Support Procedure 25 o

Direct stopping injection with Core Spray not required for adequate core cooling IAW Support Procedure 10 Directs verification of Torus water level o

The RPV has been emergency depressurized with EMRVs.

Notes/Comments NRC 2 Page 16 of 19

Procedure EMG-3200.02 Support Proc. 1 Rev. 17 Attachment B Page 1 of 3

.W 1.0 2.0 3.0 SUPPORT PROCEDURE 1 CONFIRMATION OF AUTOMATIC INITIATIONS AND PREREQUISITES SOLAT I ONS Confirmation of automatic initiations and isolations has been directed by the Emergency Operating Procedures.

PREPARATION None PROCEDURE Confirm the following isolations/starts not required to by bypassed by the Emergency Operating Procedures:

SYSTEM Reactor I solation Scram Discharge Volume I solation OPERATING DETAILS IF -

THEN Any of the following conditions exist:

RPV water level at or below 86 in. and not bypassed Steam tunnel temperature at or above 180F Any steam line flow at or above 4.0~106 lbm/hr Reactor mode switch in RUN and RPV pressure at or Main steam line radiation at or above 800 units below 850 psig no ATWS condition exists Confirm closed the following:

MSIVs IC VENTS RX SAMPLE NS03A V-14-1, -19 V-24-30 (11F)

NS04A V-14-5,-20 V-24-29 (11F)

NS03B DW AIR SUPPLY NS04B V-6-395 (11F)

IF A Reactor Scram is initiated AND SDV HI-HI LVL SCRAM switch is not in BYPASS, THEN Confirm closed the following:

NORTH SDV Vents & Drains SOUTH SDV Vents & Drains OVER E2-1

( 3 2 0 0 0 2 / 4 )

Procedure EMG-3200.02 Support Proc. 1 Rev. 17 Attachment B Page 2 of 2 1

OPERATING DETAILS

\\--

SYSTEM IF IC-A is ruptured, THEN Confirm closed the following:

IC ISOLATION VALVES IC VENTS V-14 V-14-34 V-14-5 V-14 V-14-36 V-14-20 IF IC-B is ruptured, IC-B Isolation C 1 e anup System Isolation I solat ion Condenser shutdown Cooling System 1 so la t ion V-17-54 V-17-19 IF RPV water level is at or below 86 in.,

Initiation Core Spray THEN Confirm closed the following:

IC ISOLATION VALVES IC VENTS V-14 V-14 -35 V-14-1 V-14 V-14-37 V-14 -1 9 I F Any of the following conditions exist:

RWCU HELB Alarms RPV water level at or below 86 in. and not bypassed Drywell pressure at or above 3.0 psig THEN -

I F confirm initiation of both Isolation Condensers.

Any of the following conditions exist:

THEN Confirm closed the following Cleanup Isolation valves:

V-16-1 V-16-14 V-16-2 V-16-61 I F Any of the following conditions exist:

RPV water level at or below 86 in.

Drywell pressure at or above 3.0 psig THEN Confirm closed the following SDC Isolation Valves:

l-System Start RPV water level at or below 86 in.

Drywell pressure at or above 3.0 psig and not bypassed.

and not bypassed.

THEN Confirm start of one Main Pump and one Booster Pump in each Core Spray System if not bypassed.

L/-- I (320002/4)

E2-2

SYSTEM Primary Containment I solation

( 3 2 0 0 0 2 / 4 )

Procedure EMG-3200.02 Support Proc. 1 Rev. 17 Attachment B Page 3 of 3 OPERATING DETAILS IF Any of the following conditions exist:

RPV water level at or below 86 in. and not bypassed.

Drywell pressure at or above 3.0 psig and not bypassed.

THEN Confirm closed the following valves that are not required to be open by the Emergency Operating Procedures:

System Valve No.

DW Vent/Purge V-27-1 (Panel 11F)

Torus Vent Torus 2" Vent Bypass DWEDT DW Floor Sump Torus/Rx Bldg.

Vacuum Breakers TIP Valves DW 2" Vent Bypass N2 Purge N2 Makeup v-27-2 V-27-3 V-27-4 V-28-17 V-2 8 -1 8 V-28-47 v-22-1 v-22-2 V-22-28 V-22-29 V-26-16 V-26-18 Common Ind

"-23-21 V-23-22 V-2 3-13 V-23-14 V-23-15 V-2 3-16 V-2 3-17 V-23-18 V-2 3-1 9 V-23-20 11

,I 11 Panel 11F)

I1 Panel 11F)

Panel 11F) r Panel 11F)

,I Panel 11F) 11 Panel 11F)

Panel 12XR 11 Panel 12XR I,

I, 11 (Panel 12XR) 11 I,

11 E2-3

Procedure EMG-3200.02 Support Proc. 29 Rev. 17 Attachment G Page 1. of 4 SUPPORT PROCEDURE 29 INITIATION OF THE CONTAINMENT SPRAY SYSTEM FOR DRYWELL SPRAYS 1.0 PREREQUISITES Manual initiation of Drywell Sprays has been directed by the Emergency Operating Procedures.

2.0 PREPARATION 2.1 Select the Containment Spray System to be used by confirming either SYSTEM 1 MODE SELECT or SYSTEM 2 MODE SELECT switch in DW SPRAY position Verify that the system TORUS CLG DISCHARGE valve closes and DW SPRAY DISCHARGE valve opens (Panel 1F/2F).

(Panel 1F/2F).

2.2 3.0 PROCEDURE II CAUTION Containment Spray suction strainer plugging may occur due to debris in the Primary Containment and result in a loss of Containment Spray System Flow.

3*1 (I CAUTION Diesel Generator overload will result if a Containment Spray pump and ESW pump are started with a Diesel Generator load of greater than 2160 KW.

IF Bus 1C or 1D are being supplied by an Emergency Diesel Generator, verify that adequate load margin is available so as NOT to exceed EDG load limit when starting THEN Containment Spray and ESW pumps OVER (320002/9)

E7-1

ij 3.2 3.3 3.4 3.5 3. 6 Procedure EMG-3200.02 Support Proc. 29 Rev. 17 Attachment G Page 2 of 3 WHEN directed to initiate Drywell sprays, THEN complete the following:

1. Confirm all Reactor Recirculation Pumps tripped.

2. Confirm the Drywell Recirc Fans tripped (Panel 11R).

CAUTION NPSH problems will develop on all operating pumps if more than 4 Containment Spray/Core Spray Main pumps are operated at the same time.

IF 4 Containment Spray/Core Spray Main pumps are in operation, THEN do not start additional Containment Spray pumps until Core Spray Main pumps can be secured.

Start a Containment Spray Pump as follows:

3.4.1 Select a Containment Spray Pump to be started.

3.4.2 Place and hold the System Pump Start Permissive Keylock for the selected pump in the appropriate position (Panel 1 F / 2 F ).

3.4.3 Start the selected Containment Spray Pump using its control switch (Panel 1 F / 2 F ).

Start an associated ESW Pump using its control switch (Panel l F / Z F ).

CAUTION Operation of Containment Spray pumps with flow above the NPSH or vortex limits may result in equipment damage. When operating beyond any flow limits, periodic evaluations should be made to verify that continued operation beyond these limits is still required.

Monitor System parameters for expected performance.

( 3 2 0 0 0 2 / 9 )

E7-2

CAUTION Diesel generator overload will result if a Containment Spray pump and ESW pump are started with a Diesel Generator load of greater than 2160 KW.

c Procedure EMG-3200.02 Support Proc. 29 Rev. 17 Attachment G Page 3 of 3 NOTE 3.7 Valves V-5-147, 166 and 167 do not seal in when the control switch is taken to CLOSE. The control switch must be held in CLOSE until the valve indicates closed.

I I

Confirm the following RBCCW Isolation valves closed (Panel 1F/2F):

3. 8 v-5-147 V-5-148 V-5-166 V-5-167 unless otherwise directed by the LOS as follows:

3.10.1 Secure Drywell Sprays when Drywell pressure drops to 4 psig 3.10.2 WHEN Torus or Drywell pressure increases to 12 psig, THEN initiate Drywell Sprays in accordance with Steps 3.3 through 3.6.

3.11 -

IF any Core Spray Booster pump is running AN D Torus Drywell pressure drops to 2 psig, THEN confirm termination of Drywell Sprays due to NPSH concerns.

( 3 2 0 0 0 2 / 9 )

E7-3

Procedure EMG-3200.04A Support Procedure - 10 Rev. 4 Attachment B Page 1 of 1 SUPPORT PROCEDURE 10 STOPPING INJECTION FROM THE CORE SPRAY SYSTEM

t. O PREREQUISITES Stopping Core Spray injection has been directed by the Emergency Operating Procedures.

2. 0 PREPARATION None 3.0 PROCEDURE 3.1 Depress the OVERRIDE switches for all sensors that are lit (Panel 1 F / 2 F ).

Depress $J& ACTUATED switches whether lit or unlit (Panel 1F/2F).

3. 2 3.3 Confirm closed Core Spray Parallel Isolation Valves NOT required to be open to assure adequate core cooling (Panel 1F/2F).

3.4 Secure Core Spray Booster Pumps NOT required to assure adequate core cooling by placing their respective control switch in STOP (Panel 1F/2F).

3.5 Secure Core Spray Main Pumps required to assure adequate core cooling by placing their respective control switch in STOP (Panel 1F/2F).

3.6 Verify Core Spray KEEP FILL TROUBLE (B-3-d) alarm is NOT illuminated.

( 320004Al S4)

E2-1

1.0 2.0 3. 0 Procedure EMG-3200.02 Support Proc. 25 Rev. 17 Attachment C Page 1 of 2 SUPPORT PROCEDURE 25 INITIATION OF THE CONTAINMENT SPRAY SYSTEM IN THE TORUS COOLING MODE PREREQUISITES Initiation of the Containment Spray System in the Torus Cooling Mode has been directed by the Emergency Operating Procedures.

PREPARATION None PROCEDURE CAUTION I

Containment Spray suction strainer plugging may occur due to debris in the Primary Containment and result in a loss of Containment Spray System Flow.

CAUTION 3.1 Diesel Generator overload will result if a Containment Spray Pump and ESW pump are started with a Diesel Generator load of greater than 2150 KW.

3. 2 IF Bus 1C or 1D are being supplied by an Emergency Diesel Generator, THEN verify that adequate load margin is available so as NOT to exceed EDG load limit when starting Containment Spray and ESW Pumps.

CAUTION NPSH problems will develop on all operating pumps if more than 4 Containment Spray/Core Spray Main pumps are operated at the same time.

IF 4 Containment Spray/Core Spray Main pumps are in operation, THEN do not start additional Containment Spray pumps until Core Spray Main pumps can be secured.

( 3 2 0 0 0 2 / 5 )

OVER E3-1

Procedure EMG-3200.02 Support Proc. 25 Rev. 17 Attachment C Page 2 of 2 I

---

e 3. 3 Start a Containment Spray Pump as follows:

3. 3. 1 Select the Containment Spray System to be used by confirming either SYSTEM 1 MODE SELECT or SYSTEM 2 MODE SELECT switch in TORUS CLG position (Panel 1 F / 2 F ).

3. 3. 2 Select a Containment Spray Pump to be started.

3. 3. 3 Place and hold the System Pump Start Permissive Keylock for the selected pump in the appropriate position (Panel l F / Z F ).

3.3.4 Start the selected Containment Spray Pump using its control switch (Panel 1 F / 2 F ).

3. 4 Start an associated ESW Pump using its control switch (Panel 1F/2F).

3. 5 CAUTION Operation of Containment Spray pumps with flow above the NPSH or vortex limits may result in equipment damage. When operating beyond any flow limits, periodic evaluations should be made to verify that continued operation beyond these limits is still required.

Monitor system parameters for expected performance.

3. 6 Start additional Containment Spray and ESW Pumps as directed by the LOS in accordance with Steps 3. 1 thru 3.5.

3.1 -

IF all Containment Spray pumps that are running are not required for Torus cooling, THEN inform the LOS and secure the Containment Spray pumps that are not required for Torus cooling.

(320002/5)

E3-2

Appendix D Required Operator Actions Form ES-D-2 Malfunction List Presets:

0 IRM 14 failed: MAL-NISOlOD 0

Cleanup Pump B 00s - place control switch in PTL Event 1 : Shifting to automatic generator voltage control None Event 2: DW/Torus vacuum breaker V 14 opens MAL-PCNOOlN to 100%

Event 3: Trip of TBCCW Pump 1-1 combined with failure of auto-start of the standby TBCCW Pump, failure of TBCCW Pump 1-1 Trip annunciator (Q3f) 0 0

0 0

MAL-TBCOOlA, Trip of TBCCW Pump 1-1 PSW-TBCOOlA to Fail to Trip (prevents auto start of standby TBCCW pump 1-2)

ANN-Q-3f to OFF to fail TBCCW PUMP 1 TRIP annunciator (Q3f)

ANN-R-6b to ON for H2 System Trouble alarm Event 4: Low oil level in Recirc Pump A, followed by Recirc Pump A high vibration LOA-RCPOO1 to TRUE 0

ANN-E-6D Hi/Lo oil level annunciator for Recirc Pump A (delayed 1-2 minutes from high vibration annunciator)

Event 5: Variable leg leak in common leg to RPV level transmitters ID13A and ID13C (indicators on 5F/6F and input into F W level control)

MAL-NSS012E to 5% over 12 minutes Event 6: High trip of APRM 7 0

MAL-NIS021G INOP trip of APRM 7 Event 7: Failure of stator water pumps 0

0 MAL-GEAOOSA Trip of Stator Water Cooling Pump A MAL-GEAOO5B Trip of Stator Water Cooling Pump B o Delay trip of the second stator cooling pump by 10-20 seconds Erom NLO role play at the stator cooling control panel verifying auto start of the standby stator cooling pump NRC 2 Page 17 of 19

Appendix D

~~

~

Required Operator Actions Form ES-D-2 Event 8: Steam leak in DW with failure of Containment Spray System 1 (torus cooling valve stays open) and Containment Spray 52C will not start 0

0 0

MAL-NSS017A to 25% over 20 minutes steam leak in DW SWI-CNSO11C to OFF Containment Spray System 1 fails in torus cooling mode SWI-CNSOO4C to ON Disables Containment Spray Pump 52C o PLACE Event 9 malfunctions in at this time also Event 9: Failure of DWEDT isolation valves and DW Sump isolation valves to auto close on high DW pressure [valves V 1 and V-22-2 (equipment drains), V-22-28 and V-22-29 (sump drains)]

MAL-RPS007A to Fail to Actuate Setup Notes - IC-172

1.

2.

3.

4.

5.

6.

Cleanup Pump B control switch is in PTL with a clearance tag applied Isolation Condenser A is isolated (steam and condensate return valves closed and tagged), with vents open IRM 14 is failed and bypassed with a clearance tag The main generator is in manual voltage control with yellow magnetic operator aids applied on the panel Have a copy of 604.4.016, Torus to Drywell Vacuum Breaker Operability and In-Service Test available for crew reference Have a copy of Attachment 403-2, LPRM and APRM Status Information Sheet filled out with inoperable LPRMs/APRMs

%----a NRC 2 Page 18 of 19

Procedure ##

Procedure Name 1

336.1 24 KV Main Generator Electrical System 2

Revision 45 I 604-4-016 Torus to Drywell Vacuum Breaker Operability and In-Service Test 34 3

4 5

1 Tech Specs RAP-C4f Torus/DW 1 Vac Brkr Open 0

RAP-C5f Torus/DW 2 Vac Brkr Open 0

6 7

ABN-2 Recirculation System Failures 5

ABN-17 Feedwater System Abnormal Conditions 3

8 9

31 7 Feedwater 74 RAP-R6b H2 System Trouble 0

10 11 ABN-20 TBCCW Failure Response 2

RAP-G2f APRM HI-HVINOP 2

12 13 RAP-G 1 c SCRAM CONTACTOR OPEN 1

RAP-R6c STATOR CLG TROUBLE 0

20 14 15 1 ABN-11 ABN-1 Reactor Scram 2

EMG-3200.01 A RPV Control - No ATWS 12 Loss of Generator Stator Cooling 16 17 10 EMG-3200.02 Primary Containment Control 17 EMG-3200.04A Emergency Depressurization - No ATWS 4

21 I RAP-E2d 18 19 I VIBRATION HI A RAP-H7e RX LVL HVLO 0

202.1 Power Ope ration 98 I O 22 I RAP-E6d I OIL LEV L HVLO A l o NRC 2 Page 19 of 19

Scenario Outline 1

2 3

Facility: Ovster Creek N

BOP Swaps Instrument Air compressors.

TS SRO C

Bop Ro single control rod scram.

TS SRO Respond to field report of loss of oil in Core Spray backup pump.

Respond to trip of Reactor Protection System (RPS) MG Set 1 and Scenario No.: NRC 3 4

5 6

Op Test No.: NRC 2006-1 C

RO Responds to control rod high temperature.

R RO Respond to loss of feedwater heating.

Crew Respond to notification of ground motion by plant personnel.

Examiners:

Operators:

8 9

Initial Conditions:

0 0

The plant is in 5-loop operation at 95% power.

Cleanup Pump B is out of service for motor repair, and is expected to return to service tomorrow.

Isolation Condenser System A was removed from service two hours ago and is isolated, due to motor operated valve torque switch replacement. Technical Specifications applicability has been reviewed.

The system is expected to return to an operable status in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

IRM 14 failed during the control rod withdrawal to critical, and is BYPASSED. Technical Specifications applicability has been reviewed and an IR has been generated.

Procedure 634.2.004, 24 Volt DC Battery Weekly Surveillance, is in progress.

Place Instrument Air Compressor #2 in LEAD and place Instrument Air compressor #1 in LAG IAW procedure 334. A Non Licensed Operator and Field Supervisor are already in the field and will be controlling the evolution. Control manipulations of the air compressors will be from the Control Room.

0 0

Turnover:

0 M

Crew crew to automatically initiate SGTS.

Respond to break in cleanup system with failure to isolate.

Respond to failure of Reactor Building ventilation radiation monitors I

I I

Event No. I Malf. No. I EventType*

Event Description 7

Respond to turbine high vibrations and turbine thrust bearing high annunciator with failure of turbine to auto trip.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Transient, (TS) Tech Specs NRC 3 Page 1 of 23

Total Malfunctions (5-8):

Malfunctions after EOP entry (1 -2)

Abnormal Events (2-4)

Major Transients (1 -2)

EOPs entered requiring substantive actions (1 -2)

EOP Contingencies w/ substantive actions (0-2)

Critical Tasks (2-3)

L./

Event 3

4 5

Player Type Description RO RO C

Responds to control rod high temperature RO R

Respond to loss of Nv heating C

Respond to trip of RPS MG Set 1 and single control rod scram I Event I Player 1 Type I Description 1

3 7

BOP N

Swaps Instrument Air compressors BOP BOP C

Respond to failure of turbine to auto trip C

Respond to trip of RPS MG Set 1 and single control rod scram Event Player Type 7

Crew M

8 Crew M

9 Crew C

after EOP Description Respond to turbine high vibrations and turbine thrust bearing high annunciator Respond to break in cleanup system with failure to isolate Respond to failure of RB vent radiation monitors NRC 3 Event 2

3 Page 2 of 23 Player Type Description SRO SRO TS TS Respond to field report of loss of oil in Core Spray backup pump (TS 3.4.A)

Respond to trip of RPS MG Set 1 and single control rod scram (TS 3.2)

i, Scenario Summarv

1. The BOP will swap Instrument Air compressors IAW procedure 334, Instrument and Service Air System. (NORMAL EVOLUTION)

2. A Non Licensed Operator will call the Control Room to explain that while on Rounds, he's found a considerable amount of oil on the floor surrounding the core spray main backup pump NZOlC, and that the glass oil cup is broken. It is expected that the SRO will declare the pump inoperable and apply Tech Specs 3.4.A. (TS)

3. The crew will respond to the trip of RPS MG Set 1 and a single control rod scram to position 04 (RAP-G~c, ABN-6, CRD Failures, ABN-50, Loss of VMCC 1A2). The BOP will re-power the RPS Bus and the RO will reset '/2 scram and '/2 isolations. The SRO will review TS 3.2.8.4 and declare the control rod inoperable, and valve-out the control rod at position at 04. (COMPONENT FAILURE) (ABN) (TS)

4. The RO will respond to a control rod high temperature alarm (RAP-H5c). The RO will apply stall flows IAW 61 7.4.002 (CRD Exercise and Flow TesVlST Cooling Water Header Check Valve), which will clear the alarm. (COMPONENT FAILURE) (ABN)

5. The crew will respond to a loss of feedwater heating. The SRO will direct a power reduction (ABN-1 7, Feedwater System Abnormal Conditions). (COMPONENT

\\-.---

FAILURE) (ABN) (REACTIVITY MANIPULATION).

6. The control room is then verbally notified of ground motion felt in the plant and the crew will review ABN-38, Station Seismic Event.

7. Turbine vibration annunciators (Q3b) will alarm, and the turbine thrust bearing high alarm (Q2b) (causes auto turbine trip). The turbine will fail to auto trip and can be successfully tripped by the operator. The scram will be successful and ABN-1 (Scram), and ABN-10 (Turbine Trip) will be entered. (COMPONENT FAILURE)

(ABN) (MAJOR)

8. A cleanup system break occurs with failure of automatic isolation, and it cannot be isolated. Secondary Containment Control EOP is entered. Two areas (cleanup and cleanup pump/heat exchanger areas) will exceed MAX SAFE parameters and the SRO will direct Emergency Depressurization (IAW Emergency Depressurization -

No ATWS) or may anticipate ED and direct a rapid RPV depressurization with ICs and TBVs. (EOP) (MAJOR) (EOP CONTINGENCY)

NRC 3 Page 3 of 23

9. As Reactor Building radiation levels rise, the RB vent radiation monitors will increase to the setpoint of auto isolation of RB HVAC and auto start of SGT. Not all of these automatic actions will occur, but are expected to be performed by the operators. (COMPONENT FAILURE)

Critical Tasks

1.

Initiate Emergency Depressurization when two areas exceed the MAX SAFE levels for radiation or temperature, or rapidly depressurize the RPV with TBVsACs when ED is anticipated.

This places the RPV in the lowest energy state to minimize the amount of energy deposited outside of Secondary Containment (radioactivity barrier).

With the RB HVAC vent exhaust radiation monitors above the high setpoint for automatic initiation of SGTS and SGTS not running, manually initiate SGT.

This automatic action minimizes the off-site radiological dose. To ensure protection of the public, a manual system initiation will fulfill this function.

2.

NRC 3 Page 4 of 23

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Place Instrument Air ComDressor #2 in LEAD and place Instrument Air comrxessor

1. 2 in LAG IAW Drocedure 334.

Scenario No.: NRC 3 Event No.: 1 Initiation: Following shift turnover.

Cues: When notified by In-Plant OperatorlField Supervisor.

Time ROLE PLAY Sim.

Operator ROLE PLAY As the Field Supervisor, call the Control Room and report: Steps 5.4.1.1 through 5.4.1.4 of procedure 334 have been completed. The local display indicates ready for start local or remote for air compressor #2. Start air compressor #2 per step 5.4.1.5 of procedure 334.

Ensure #2 air compressor is selected as LEAD locally (LOA-CAS038 to LEAD).

SRO BOP Allows continuing placing air compressors #2 in the LEAD, and #1 air compressor in LAG, IAW procedure 334, Section 5.4 (start will be from the Control Room) (attached) 0 Makes plant page regarding starting #2 air compressor Places #2 air compressor as the LEAD by: (7F) o o

Places #1 air compressor as the LAG o

o Verify #2 air compressor loads and unloads to maintain system pressure 85-105 psig (7F)

Reports air compressor status to SRO Place COMPRESSOR 2 switch to the START position for 3-5 seconds Reports to the Field Supervisor that step 5.4.1.5 is complete.

Place the COMPRESSOR 1 switch to the START position for 3-5 seconds Confirm #1 air compressor runs unloaded for 10 minutes and then auto shuts down 0

I As the Field Supervisor, after completion of step 5.4.1.5 (#2 air compressor running), call the Control Room and report: Step 5.4.1.7 of procedure 334 has been completed satisfactorily.

Compressor settings for air compressor #1 have been confirmed IAW step 5.4.1.9. Place air compressor #1 in LAG IAW step 5.4.1.9.2.

NRC 3 Page 5 of 23

Appendix D Required Operator Actions Form ES-D-2

e Terminus:

1. 2 air compressor is running as the LEAD, and #I air compressor is running as the LAG (do not need to wait for auto shutdown of #1 air compressor)

Notes/Comments NRC 3 Page 6 of 23

- AmerGen-An Exdon Company OYSTER CREEK GENERATING Mmber STATION PROCEDURE 1

334 e

Low control air header pressure may indicate a fouled prefilter and/or post filter, the filters should be swapped or bypassed if they are fouled, and if necessary changed.

L-.

Title Instrument and Service Air System 0

The discharge air temperature out of the in-service drying towers should be below 125*F, as indicated locally.

Revision No.

93 The SA Isolation Bypass switch should be in the NORMAL position.

0 The dew point analyzer (hygrometer} on the control air system should be less than 0°F. (This dew point temperature will fluctuate depending on the cycle time position of the dryers. Lower dewpoint temperatures are achievable at the beginning of the drying cycle.) Insure that there is minimum flow (approx. 5 scfh) through the sample cell as indicated on local flow indicator.

Swapping Air Compressors 5.4.1 START an out of service or Lag compressor as follows:

5.4.1.1 1

NOTE After the operator setpoint of Lead is selected, but before the compressor is started, the compressor may start if air pressure drops below the auto start IF

1. I or #2 Air Compressor is to be placed in

LEAD, THEN CONFIRM the compressors settings are as specified in Attachment 334-9, Guidelines for Changing Operator Setpoints or Options, for Lead compressor operation.

19.0

OYSTER CREEK GENERATING STATION PROCEDURE AmerGen An txron Company Title Instrument and Service Air System 5.4.1.2 5.4.1.3 5.4.1.4 5.4.1.5 5.4.1.6 Number 334 Revision No.

93 IF

1. 3 Air Compressor is to be placed in LEAD, THEN PLACE the local select switch to the CONFIRM cooling water is correctly lined up to the CONFIRM the compressor oil level is as follows:

1. I and #2 Air Compressors oil level is between 0 #3 Air Compressor oil level is about the SELECT position.

I 1

backup compressor.

[

I the top and bottom of the sightglass I

1 centerline of the sightglass I

1 IF -

AND THEN IF AND THEN IF THEN

1. I or #2 Compressor is to be started, the compressor local display indicates ready for start local or remote.

PLACE the Panel 7F Control Switch to start and hold for 3-5 seconds, or press the local start (green) button.

[

I

1. I or #2 Compressor is to be started, the compressor has stopped in Auto Restart displayed on the local screen and auto start had not occurred, START the compressor as follows:

1. PLACE the Panel 7F Control Switch to STOP, and allow the switch to spring return to normal after stop.

2. PLACE the Panel 7F Control Switch to START, and hold for 3-5 seconds, or press the local start (green) button.

[

I

1. 3 Air Compressor is to be started, PLACE the #3 Compressor 7F Control Switch to START.

[

I

[

I 20.0

AmerGm OYSTER CREEK GEN ERATl NG Number 1

STATION PROCEDURE I

334 An Excim Company Title ii Instrument and Service Air System Revision No.

I 93 5.4.1.7 VERIFY satisfactory start of compressor by observing the following:

5.4.1.8 0

Lube oil pressure is 40-50 psig, for #I and #2 Air Compressors, 45 to 55 psig for #3 Air Compressor.

[

0 2 stage inlet pressure (interstage pressure) is 25-35 psig.

0 TBCCW is observed flowing in the sight glass, and TBCCW leaving the compressor is less than 125°F.

[

NOTE If stopping # I or #2 Air Compressors, the compressor may continue to operate for about 10 seconds after giving a stop signal. This is to allow the compressor to automatically unload and depressurize prior to stopping.

STOP the operating compressor by placing the control switch to the STOP.

(Panel 7F)

OR PRESS the local STOP (red) button.

[

1. VERIFY OFF indication.

5.4.1.9 E

1. I or #2 Air Compressor will be placed in LAG I THEN PERFORM the following to setup the compressor to auto start at 90 psig:

1. CONFIRM the compressors settings are as specified in Attachment 334-9, Guidelines for Changing Operator Setpoints or Options, for Lag

2. START the compressor by pressing the local (green) START pushbutton.

compressor operation.

[

OR PLACE the start switch in the START position for 3-5 seconds.

(Panel 7F).

[

I 21.o

W

.L--

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Reoort of oil leak on Core Smav Main Backup Pump NZOl C Scenario No.: NRC 3 Event No.: 2 Initiation: #2 air compressor is running as the LEAD, and #1 air compressor is running as the LAG Cues: Call from in-plant Non Licensed Operator Time ROLE PLAY Terminus:

Position Applicants Actions or Behavior As the NLO, call the Control Room with the following report: While on Rounds, I found a considerable amount of oil on the floor surrounding Core Spray Main Pump NZOl C, and the glass oil cup is broken. I have contained the oil and none has reached a floor drain. I will continue cleaning up the oil.

If requested to check the other core spray pumps, report that there are no visible deficiencies with the other core spray pumps.

SRO Note Reviews Tech Spec 3.4.A (attached) o o

o CONDITION: Any active loop component becomes inoperable REQUIREMENT: The Reactor may remain in operation for a period not to exceed 15 Days.

PROVIDED: Both Emergency Diesel Generators are OPERABLE. The Redundant active loop components within the same loop as the inoperable components are verified OPERABLE on a daily basis.

Specification 3.4.A.3 is met unless only a core spray booster pump is inoperable

. 3.4.A.3: APLHGR shall not exceed 90% of the limits (attached)

Declares Core Spray Main Pump NZOl C inoperable o

Notifies Work Week manager for repair Protects the other Core Spray Pumps and EDGs 0

Updates the Crew Unexpected Plant Change Checklist Redundant System Verification Form 0P-0c-101-1000 Core Spray Main Backup Pump NZOl C has been declared inoperable and the Crew has been updated.

NRC 3 Page 7 of 23

Appendix D Required Ope rator Actions Form ES-D-2 N otes/Com me n ts L/-

NRC 3 Page 8 of 23

3.4 EMERGENCY COOLING Requirement A~~licabilitv: Applies to the operating status of the emergency cooling systems.

Provided:

Obiective:

To assure operability of the emergency cooling systems.

One Emergency Diesel Generator is inoperable.

The Reactor may remain in operation for a period not to exceed 7 Days.

(Refer to Section 3.7.C.2)

SDecifications:

booster pump is inoperable.

All core spray equipment connected to the OPERABLE emergency diesel generator is OPERABLE.

A.

Core Spray System NOTE: LCO 3.0.C.2 is not applicable to the Core Spray System

1.

The Core Spray System shall be OPERABLE at all times with irradiated fuel in the reactor vessel with an absorption chamber water volume of at least 82,000 ft3 except as specified Table 3.4.1 ) or as noted below.

in

2.

If Specification 3.4.A.1 is not met the reactor shall be PLACED IN the COLD SHUTDOWN CONDITION and no work shall be performed on the reactor or its connected systems which could result in lowering the reactor water level to less than 48above TOP OF ACTIVE FUEL.

Table 3.4. I Run or Startup Mode (except for low power physics testing)

Condition Any active loop component becomes inoperable.

Two or more active loop components in the same loop (System 1 or System 2) are inoperable provided no two components are redundant.

-0R-The Reactor may remain in operation for a period not to exceed 15 Days.

Both Emergency Diesel Generators are OPERABLE.

The Redundant active loop components within the same loop as the inoperable components are verified OPERABLE on a daily basis.

Specification 3.4.A.3 is met unless only a core spray OYSTER CREEK 3.4-1 AmendmentNo.:

Corrected by letter of 10/13/04

Run or Startup Mode (except for low power physics testing)

Condition One core spray loop (System 1 or System 2) or its core spray header delta-P instrumentation becomes inoperable.

Both of the redundant components in a loop (System 1 or System 2) are inoperable.

Two of the four redundant active loop components in the core spray system not in the same loop (System 1 or System

2) are inoperable.

Two or more non-redundant active loop components are inoperable in both loops (System 1 and System 2).

-0R-

-0R-Condition Maintenance or modifications of core spray systems, their power supplies, or water supplies.

Requirement The Reactor may remain in operation for a period not to exceed 7 Days.

The Reactor may remain in operation for a period not to exceed 7 Days.

Shutdown or Refuel Mode Requirement Maintain reduced core spray system availability as follows:

1.At least one core spray pump, and system components necessary to deliver rated core spray to the reactor vessel, must remain OPERABLE to the extent the pump and any necessary valves can be started or operated from the control room or from local control stations.

2.The Fire protection system is OPERABLE to the extent that one diesel driven fire pump is capable of providing water to the core spray system.

3.Verifjr the systems in 1 & 2 above are OPERABLE on a weekly basis.

Provided:

Both Emergency Diesel Generators are OPERABLE.

The remaining loop (System 1 or System 2) has no inoperable components and is verified daily to be OPERABLE.

Specification 3.4.A.3 is met.

Both Emergency Diesel Generators are OPERABLE.

The Redundant active loop components within the same loop as the inoperable components are verified OPERABLE on a daily basis.

Specification 3.4.A.3 is met.

Provided:

The Reactor is maintained in the COLD SHUTDOWN CONDITION or in the REFUEL MODE with the reactor coolant system maintained at less than 2 12°F and vented.

-AND-No work is performed on the reactor vessel and connected systems that could result in lowering the reactor water level to less than 4'8" above the TOP OF ACTIVE FUEL.

OYSTER CREEK 3.4-2 Amendment No.: 75, !53, I-&$

Zl-+-L, 247

Condition Maintenance or modifications of core spray systems, their power supplies, or water supplies while work is in progress having the potential to lower reactor water level below 48 TAF.

The Reactor is in the startup mode for low power physics testing.

-0R-The requirements for maintenance or modification can not be met.

Shutdown or Refuel Mode Reauirement Maintain reduced core spray system availability as follows:

1.At least one core spray pump in each loop, and system components necessary to deliver rated core spray to the reactor vessel, must remain OPERABLE to the extent that the pump and any necessary valves in each loop can be started or operated from the control room or from local control stations.

2.Fire protection system is OPERABLE to the extent that one diesel driven fire pump is capable of providing water to the core spray system.

3.Verify the systems in 1 & 2 above are OPERABLE everv 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Initiate work to meet the requirements.

Provided:

The Reactor is:

In the REFUEL MODE with the reactor coolant system maintained at less than 2 12°F.

In the STARTUP MODE for the purpose of low power physics testing.

-0R-Specification 3.4.A.2 is met.

..+

3. In the event of inoperable active loop components the APLHGR of all the rods in any fuel assembly, as a function of average planar exposure, at any axial location shall not exceed 90% ofthe limits given in Specification 3.10.A. The action to bring the core to 90% of the APLHGR Limits must be completed within two hours after the component has been determined to be inoperable.

4. The core spray system is not required to be operable when the following conditions are met:

a. The reactor mode switch is locked in the Refuel or Shutdown position.

b.

(1)

There is an operable flow path capable of taking suction from the condensate storage tank and transferring water to the reactor vessel, and (2)

The fire protection system is OPERABLE to the extent that one diesel driven fire pump is capable of providing water to the core spray system, and (3)

These systems are verified to be OPERABLE on a weekly basis.

OYSTER CREEK 3.4-3 Amendment No.: 75, ! 5 ?, !e?,

2-44, a, 247

Amendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Trir, of RPS MG 1 and sinale control rod scram (control rod 14-43)

Scenario No.: NRC 3 Event No.: 3 Initiation: Core Spray Main Backup Pump NZOlC has been declared inoperable and the Crew has been updated Cues: Annunciator G2c, RPS MG SET 1 TRIP; G1 c, SCRAM CONTACTOR OPEN; H6a, ROD DRIFT; 9XF3a, PROT SYS PNL PWR LOST (plus RPS A /2 scram annunciators)

Time Position ROLE PLAY NRC 3 RO BOP Applicants Actions or Behavior 0

Responds to annunciators G1 c, SCRAM CONTACTOR OPEN and H6a, ROD DRIFT o

o o

Following restoration of RPS A, resets the following: (4F) o Half scram Depresses SCRAM SYSTEM RESET pushbutton o

Main Steam isolation Depresses MAIN STEAM ISOLATION RESET pushbutton o

Associated annunciators Reports loss of RPS A Reports single control rod scram 14-43: did not scram full-in Refers to ABN-6, Control Rod Drive System 0

0 Refers to annunciator G2c, RPS MG SET 1 TRIP and 9XF3a, PROT SYS PNL PWR LOST o

o o

Restores System Panel 1 (PSP-I) from Transformer PS-1 IAW procedure 408.12, Operation of Reactor Protection System Panel 1-1 and Transfonner PS-1, Section 5.4 (6R) (attached)

Following restoration of RPS A, resets components IAW procedure 408.12:

o Monitor 4160V level for Transformer PS-1 hourly (8F/9F)

Checks MG status lights and voltage (6R)

Verifies loss of power to RPS A components Dispatches NLO to investigate RPS MG 1 0

0 Steps 5.4.8 - 5.4.1 2 (attached) 0 As the NLO, when requested to investigate RPS MG 1, report the following: RPS A MG Set input breaker is open (EPA breakers EPA-1 and EPAQ are open also??), with no indications of a fault.

SRO 0

Directs the rod scram event o

o Directs entry into ABN-6, Control Rod Drive System Declare the control rod inoperable and directs to isolate the control rod Page 9 of 23

Appendix D Required Operator Actions Form ES-D-2 o

Applies TS 3.2.8.4 (attached) o Applies TS 3.2.A (attached) o May also apply TS 3.13, Accident Monitoring Instrumentation, while RPS A is de-energized (due to loss of one channel of Wide Range Torus Water Level indication and one channel of Containment High Range Radiation instrument).

Notify Reactor Engineering of control rod event Directs restoration of Protection System Panel 1 (PSP-1) from Transformer PS-1 IAW procedure 406.1 2, Operation of Reactor Protection System Panel 1-1 and Transformer PS-1 Notifies Work Week manager of RPS MG set and scrammed control rod May direct shutdown of the RPS A MG Set o

Directs the loss of RPS event o

o o

Sim Operator When the scram is reset, DELETE the control rod scram malfunction I

ROLE PLAY As the NLO, when requested to isolate the scrammed control rod, report back a few minutes later that the control rod is isolated IAW 302.1, Control Rod Drive System Terminus:

RPS A has been transferred to Transformer PS-1 and alarms/isolations reset (not required to wait for MG Set shutdown)

Notes/Comments NRC 3 Page 10 of 23

amfarm. 1 OYSTER CREEK GENERATING STATION PROCEDURE An !*rho Company I

~.-'

Title Operation of Reactor Protection System Panel 1-1 and Transformer PS-1 Number 408.12 Revision No.

12 ActionNerifv 5.3.6 VERIFY Power In Lamp illuminated on Electrical Protection Assembly (EPA) #5.

f 5.3.6.1 RESET EPA #5 by closing circuit breaker.

f 5.3.7 VERIFY EPA #5 Power Out Lamp and EPA #6 Power In 1

Lamp On.

5.3.8 RESET EPA #6 by closing Circuit Breaker on EPA #6.

f 5.3.8.1 VERIFY EPA #6 Power Out Lamp On.

I 5.3.9 VERIFY Green Trans Output lights on Panel 6R and 7R are On.

f

.-p 5.4 Transferring RPS Panel 1-1 Power from MG Set to Transformer PS-1 5.4.1 Reactor is >850 psig (as indicated by MN STM PRESS LO alarm J-5(6)-a clear)

THEN VERIFY the following RPS 2 relays are energized:

-2K51 -2K52A -2K73 -2K76

-2K51A -2K71

-2K74 -2K77

-2K52 -2K72

-2K75 (4081 2) 9.0

her-.

1 OYSTER CREEK GENERATING An Exdon Company STATION PROCEDURE Number 408.12 ActionNerifv 5.4.2 Reactor is 5 600 psig (as indicated by RPS 600#/SD BYPASS alarm G-4-c annunciated),

I

~

Title Operation of Reactor Protection System Panel 1-1 and Transformer PS-1 AND the mode switch is in REFUEL gr SHUTDOWN, Revision No.

12 THEN PERFORM the following:

5.4.2.1 VERIFY following RPS 2 relays are energized:

-2K51 -2K52A -2K73 -2K76

-2K51A -2K71

-2K74 -2K77

-2K52 -2K72

-2K75 5.4.2.2 CONFIRM jumpers installed on the following IAW Procedure 203.2:

-2K117 -2K11 -2K12

-2K118 -2K17 -2K18 5.4.3 OBSERVE Green Trans Output light illuminated.

(Panel 6R)

/

5.4.4 VERIFY RPS Panel 1-1 Main Breaker Closed.

(Panel 6R) 5.4.5 PLACE Power Select Switch to OFF for two seconds, (Panel 6R)

I then to TRANSFORMER position.

5.4.6 VERIFY green Trans Output Lamp extinguishes and Red Trans Output Lamp illuminates.

1-5.4.7 VERIFY DCC-Y computer (DCS-Pnl. 9R) is not controlling.

1-(4081 2) 10.0

AmerGen..

An Fxim Company Number 408.12 OYSTER CREEK GENERATING STATION PROCEDURE Revision No.

12 ActionNerifv 5.4.8 RESET the following:

a Half scram signal.

Main steam isolation.

0 APRM lights on Panel 3R.

I l

a APRM flow converters on Panel 3 R and 5 R.

Main Steam Line Rad Monitor RNOGA and B drawers.

I Annunciator alarms.

1 5.4.9 VERIFY Status LED on FCTR card is Green.

I 5.4.9.1 E Status LED on FCTR card is solid THEN PERFORM the following:

flashing red.

1. PUSH reset.

I

2. VERIFY LED is Green.

I (Refer to Attachment 408.12-1) 5.4.9.2 E Status LED cannot be reset to Green, THEN REFER to Procedure 403, LPRM-APRM System Operations, Attachments 403-5 and 403-6.

I (4081 2) 11.0

3.2 REACTIVITY CONTROL Applicabilitv:

Applies to core reactivity and the operating status of the reactivity control systems for the reactor.

To assure reactivity control capability of the reactor.

L Obiective:

Specification:

A.

Core Reactivity

1.

The SHUTDOWN MARGIN (SDM) under all operational conditions shall be equal to or greater than:

(a)

(b)

If one or more control rods are determined to be inoperable as defined in Specification 3.2.6.4 while in the STARTUP MODE or the RUN MODE, then a determination of whether Specification 3.2 A. is met must be made within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If a determination cannot be made within the specified time period, then assume Specification 3.2 A.l is not met.

If Specification 3.2.A.1 is not met while in the STARTUP Mode or the RUN MODE, meet Specification 3.2.A.1 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in the SHUTDOWN CONDITION within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

If Specification 3.2.A.1 is not met while in the SHUTDOWN CONDITION, or the COLD SHUTDOWN CONDITION, then:

(a)

(b)

If Specification 3.2.A.1 is not met while in the REFUEL MODE, then:

(a) 0.38% delta klk, with the highest worth control rod analytically determined; or 0.28% delta k/k, with the highest worth control rod determined by test.

2.

3.

4.

Fully insert all insertable control rods within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, AND Comply with the requirements of Specifications 3.2.C and 3.5.B.

5.

Immediately suspend CORE ALTERATIONS except for fuel assembly removal, AND Immediately initiate action to fully insert all insertable control rods in control cells containing one or more fuel assemblies, AND Comply with the requirements of Specifications 3.2.C and 3.5.6.

(b)

(c)

OYSTER CREEK 3.2-1 Amendment No: 75, 113, 178

B.

Control Rod Svstem ij

1.

The control rod drive housing support shall be in place during power operation and when the reactor coolant system is pressurized above atmospheric pressure with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3.2.A is met.

2.

The Rod Worth Minimizer (RWM) shall be operable during each reactor startup until reactor power reaches 10% of rated power except as follows:

(a)

Should the RWM become inoperable after the first 12 rods have been withdrawn, the startup may continue provided that a second licensed operator verifies that the licensed operator at the reactor console is following the rod program.

(b)

Should the RWM be inoperable before a startup is commenced or before the first twelve rods are withdrawn, one startup during each calendar year may be performed without the RWM provided that the second licensed operator verifies that the licensed operator at the reactor console is following the rod program and provided that a reactor engineer from the Core Engineering Group also verifies that the rod program is being followed. A startup without the RWM as described in this subsection shall be reported in a special report to the Nuclear Regulatory Commission (NRC) within 30 days of the startup stating the reason for the failure of the RWM, the action taken to repair it and the schedule for completion of the repairs.

Control rod withdrawal sequences shall be established with a banked position withdrawal sequence so that the rod drop accident design limit of 280 callgrn is not exceeded. For control rod withdrawal sequences not in strict compliance to BPWS, the maximum in sequence rod worth shall be S I.O% AK.

3.

The average of the scram insertion times of all operable control rods shall be no greater than:

Rod Length Insertion Time Inserted (YO)

(Seconds) 5 20 50 90 OYSTER CREEK 0.375 0.900 2.00 5.00 3.2-2 Amendment No: 75, 113, 178

The average of the scram insertion times for the three fastest control rods of all groups of four control rods in a two-by-two array shall be no greater than:

Rod Length Insertion Time Inserted (%)

(Seconds) 5 20 50 90 0.398 0.954 2.120 5.300 Any four rod group may contain a control rod which is valved out of service provided the above requirements and Specification 3.2.A are met. Time zero shall be taken as the de-energization of the pilot scram valve solenoids.

4.

In service control rods which cannot be moved with control rod drive pressure shall be considered inoperable. If a partially or fully withdrawn control rod drive cannot be moved with drive or scram pressure, the reactor shall be brought to a shutdown condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> unless investigation demonstrates that the cause of the failure is not due to a failed control rod drive mechanism collet housing. Inoperable control rods shall be valved out of service, in such positions that Specification 3.2.A is met. In no case shall the number of inoperable control rods valved out of service be greater than six during the power operation. If this specification is not met, the reactor shall be placed in the shutdown condition.

5.

Control Rods shall not be withdrawn for approach to criticality unless at least two source range channels have an observed count rate equal to or greater than 3 counts per second.

C.

Standbv Liauid Control Svstem

1.

The standby liquid control system shall be operable at all times when the reactor is not shut down by the control rods such that Specification 3.2.A is met and except as provided in Specification 3.2.C.3.

2.

The standby liquid control solution shall have a Boron-I 0 isotopic enrichment equal to or greater than 35 atom YO, be maintained within the cross-hatched volume-concentration requirement area in Figure 3.2-1 and at a temperature not less than the temperature presented in Figure 3.2-2 at all times when the standby liquid control system is required to be operable.

3.

(a)

If one standby liquid control system pumping circuit becomes inoperable during the RUN mode and Specification 3.2.A is met, the reactor may remain in operation for a period not to exceed 7 days, provided the pump in the other circuit is verified daily to be operable, otherwise be in the Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

OYSTER CREEK 3.2-3 Amendment No: ?5, 124, 157, ?78,253

Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Hiah temperature control rod 02-1 9 Scenario No.: NRC 3 Event No.: 4 Initiation: RPS A has been transferred to Transformer PS-1 and alarms/isolations reset Cues: Annunciator H5c, CRD TEMP HI (4F)

Time ROLE PLAY Sim.

Operator CUE:

Terminus:

Position RO Applicant's Actions or Behavior 0

Responds to annunciator H5c, CRD TEMP HI o

o o

Confirm CRD cooling water differential pressure/flow within limits of 302.1, Control Rod Drive System (15 psid at 30-45 GPM)

Direct NLO to check leaky scram discharge valve for control rod 02-19 Attempt to clear the alarm:

Apply stall flow signals to control rod 02-1 9 IAW procedure 61 7.4.002, CRD Exercise and Flow Test/lST Cooling Water Header Check Valve (attached)

As NLO, when requested to check leaky scram discharge valve for control rod 02-19, report that the scram discharge piping is no warmer than other control rods in the area.

When the RO is applying stall flow, DELETE the CRD high temperature malfunction (it will take about 30 seconds to clear the annunciator).

BOP 0

Responds to annunciator H5c, CRD TEMP HI o

Determines which control rod is effected (02-1 9) and reports (8R)

Verifies control rod 02-19 temperature normal following stall flow (8R) 0 Control Rod Drive Temperature recorders on Panel 8R are SIMULATED. When the BOP investigates, tell him that the only alarming control rod is Control Rod 02-1 9 indicates 255" F and steady.

Following the stall flow, when the BOP goes to monitor CRD temperatures, tell the BOP that control rod 02-19 indicates 235" F and steady.

CRD 02-1 9 shows normal temperature.

\\

NRC 3 Page 11 of 23

Appendix D

~

~

~

~~~~

Required Operator Actions Form ES-D-2 Notes/Comments NRC 3 Page 12 of 23

OYSTER CREEK GENERATING STATION PROCEDURE AmerGen.

An fxelon Company Title CRD Exercise and Flow TestllST Cooling Water Header Check Valve Number 61 7.4.002 Revision No.

44

~~

Perform N e r i fv 6.0 PROCEDURE 6.1 VERIFY all Section 3.0 prerequisites met.

I 6.2 VERIFY Operating Stabilizing Valve flows in acceptable range by performing the following:

6.2.1 IF Stabilizing Valves NC-19NNC-19B are in

service, THEN PERFORM Attachment 617.4.002-I.

f 6.2.2 IF Stabilizing Valves NC-1 9ElNC-19F are in

service, THEN PERFORM Attachment 617.4.002-2.

I 6.2.3 RECORD stabilizing valve flow from Attachment 617.4.002-1 or Attachment 617.4.002-2 on Data Sheet, 1 7.4.002-3.

I 6.3 STALL FLOW MEASUREMENT OF CRDS 7-6.3.1 PERFORM the following for fullv withdrawn CRDs:

6.3.1.1 TURN ROD POWER Switch to ON.

I 6.3.1.2 SELECT a CRD at position 48.

I 6.3. I

.3 HOLD NOTCH OVERRIDE Switch in NOTCH OVERRIDE position.

I 6.3.1.4 HOLD ROD CONTROL Switch to ROD OUT NOTCH position.

f 10.0

ArnerGem 1

OYSTER CREEK GENERATING A? lrelon Company STATION PROCEDURE umber 61 7.4.002 6.3.1.5 WHEN CRD returns to position 48, I

L4=

Title CRD Exercise and Flow TestllST Cooling Water Header Check Valve AND drive pressure stabilizes, Revision No.

44 THEN RECORD the withdraw stall flow as indicated on FT-RD10, Drive Water Flow on Data Sheet, 1 7.4.002-3.

(Panel 4F)

I 6.3.1.6 RELEASE ROD CONTROL Switch.

I b-6.3.1.7 RELEASE NOTCH OVERRIDE Switch 6.3.1.8 PERFORM the following for withdraw stall flows >5 gpm:

I

1.

IDENTIFY each CRD with a withdraw stall flow >5 gpm in Comments section on the Data Sheet, Attachment 617.4.002-3.

I

2.

CONVERT stall flows >5 gpm from differential pressure on DPI-234, Drive Water Flow Element Diff Pressure Indicator to gallons per minute using 1 7.4.002-5.

(DPI-234 Behind CRD Filters)

I 6.3.1.9 PERFORM the following for each remaining CRD at position 48:

I. REPEAT Steps 6.3.1.2 through 6.3.1.7 I

2.

PERFORM Step 6.3.1.8 for stall flows

>5 gpm.

I

3.

RECORD stall flows.

I 6.3.1.10 RETURN ROD POWER Switch to OFF.

11.0

~

Appendix D Required Ope rator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Loss of Feedwater Heating Scenario No.: NRC 3 Event No.: s Initiation: CRD 02-1 9 shows normal temperature Cues: Annunciator N3d, HP A3 LEVEL HVLO (7F)

Time ROLE PLAY Sim.

Operator Terminus:

Position BOP Applicants Actions or Behavior Responds to annunciator N3d, HP A3 LEVEL HI/LO; N2d, HP A3 MRV OPEN; Nld, HP A3 REV CK VLV TRIP (and similar for FWH A1 and o

Monitors feedwater temperature (5F/6F) and reports FW temperature val ue/t rend o

Refers to ABN-17, Feedwater System Abnormal Conditions A2) (7F)

Monitor off-gas activity (1 R)

Monitor Main Steam Line radiation (1 R)

Monitor FLLLP (PPC)

As the NLO and when requested, notify the Control Room that HP (IP, LP) feedwater heater level is high.

RO SRO Monitors reactor power Reduces reactor power as directed to 20% less than the pre-trip value with recirc. flow o

Rotates the MASTER RECIRC SPEED CONTROLLER knob in the counter-clockwise direction (4F) o Monitors the Power Operations Curve Directs entry into ABN-17, Feedwater System Abnormal Conditions o

o o

o Makes notification for down-power Direct reactor power reduction with recirculation flow to maintain 20%

below the pre-trip power level or until 8.5 x 1 O4 GPM Maintains plant load less than 502.5 MWe ISW ABN-17 with 1 bank (LP, IP and HP) heaters lost Notifies Work Week Manager for repair The intent is to trip all three FW heaters in string A. If they do not all trip as designed, then activate the associated annunciators to simulate the trip (annunciators N1 d through N8d).

Reactor power has been lowered with recirc. flow NRC 3 Page 13 of 23

Appendix D Required Operator Actions Form ES-D-2 NotesEomments

--....-=

NRC 3 Page 14 of 23

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Seismic event Scenario No.: NRC 3 Event No.: 6 P

P Initiation: Reactor power has been lowered with recirc. flow following the partial loss of feedwater heating.

Cues: Phone notification by in-plant Engineer.

SRO Direct entry into ABN-38, Station Seismic Event o

Direct NLOs to perform plant walk-downs to inspect for damage (includes IFSFl Facility)

NOTE:

1 Notes/Comments NRC 3 Page 15 of 23

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Scenario No.: NRC 3 Event No.: 7 Event

Description:

Hiah main turbine vibrations, hiah main turbine thrust bearina wear and failure of main turbine to auto trip Initiation: 3 minutes after the NLOs have been directed to perform plant walk-downs to inspect for damage Cues: Annunciator Q3b, VIBRATION HI, followed by annunciator QZb, THRUST BRG WEAR HI Time Terminus:

Position BOP RO SRO Applicants Actions or Behavior 0

Respond to Annunciator Q3b, VIBRATION HI, and Q2b, THRUST BRG WEAR HI o

Verify thrust bearing indication (7F) and report indication is above the turbine trip setpoint o

Report that scram and turbine trip required IAW RAP-Q2b (Following the scram) Depresses TURBINE EMERGENCY TRIP pushbuttons (7F)

. Confirm the following:

0 Main Stop valves closed Turbine Control valves closed Turbine Reheat and Intercept valves closed HWC H2 inlet Isolation valve, V-567-005, closed 230 KV breakers GC1 and GD1 open Plant electrical loads are transferred to the Startup Transformers SA, SB Maintain RPV pressure 800-1 000 psig with TBVs IAW RPV Control - No ATWS Manually scrams the reactor and carries out ABN-1, Reactor Scram (attached)

(4F) o Scrams the reactor o

Inserts SRMs/lRMs o

Maintains RPV water level 0

Direct reactor scram and turbine trip Directs entry into RPV Control -No ATWS o

Directs auto actions confirmed o

o Directs RPV water level 138 - 175 with feedwater Directs RPV pressure 800 - 1000 psig with turbine bypass valves The reactor has been scrammed and main turbine tripped, and ABN-1 immediate actions have been performed.

NRC 3 Page 16 of 23

Appendix D Required Operator Actions Form ES-D-2 u'

Notes/Co m me n ts NRC 3 Page 17 of 23

AmerGm OYSTER CREEK GENERATING Number I

Revision No.

Title 2

REACTOR SCRAM 2.3 Other Indications

1.

Red scram lights lit on full core display.

2.

Control rod positions indicate blank with green backlighting on the full core display.

3.

Group scram solenoid lights not lit on 4F, 6R, 7R.

3.0 OPERATOR ACTIONS If while executing this procedure, an entry condition for any EOP occurs, then EXECUTE this procedure concurrently with the appropriate EOP.

3.1 If then PERFORM the following:

a manual scram is to be performed and time permits, REDUCE recirculation flow to approximately 8.5 x I O4 gpm.

[

I NOTIFY the System Dispatcher that Oyster Creek will be taken off line.

[

I 4.0

OYSTER CREEK GENERATING AmerGm An fxelvn Company STATION PROCEDURE

.L/

Title REACTOR SCRAM 3.8 CONFIRM the following:

0 Main Turbine Tripped

[

I 0

GDI Open

[

I 0

GCI Open

[

I Field breaker Open

[

I 3.9 If then EXIT this procedure and enter ABN-36, Loss of Off-site Power. [ ]

off-site power is not available to both startup transformer, Number ABN-1 Revision No.

2

4.

When directed by the Unit Supervisor, PLACE the LFRV in automatic.

[

I 5.0

AmerGen.

OYSTER CREEK GENERATING STATION PROCEDURE Number ABN-1 Title ii REACTOR SCRAM

6.

Revision No.

2 RESET the scram in accordance with step 3.12, if allowed.

Otherwise, CLOSE CRD Charging Header Supply, V-15-52 [ ]

6.0

~

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Event

Description:

Non-isolatable leak from the CleanuD Svstem into the Secondarv Containment lwith failure of svstem isolation valves to close) and fuel failures Scenario No.: NRC 3 Event No.: 8 Initiation: After ABN-1 immediate actions have been performed.

Cues: Annunciators Dld, RWCU HELB I and D2d RWCU HELB II; followed by annunciator D8d, CU ROOM TEMP HI (3F); Report of steam from the Cleanup Cage area Time NRC 3 Position RO/BO P Applicant's Actions or Behavior Respond to annunciators Dld, RWCU HELB I and D2d RWCU HELB II o

Reports failure of Cleanup System to automatically isolate and manually attempts to close the following valves by taking the respective switch to the CLOSE position:

... V-16-14, CU Inlet IV V-16-1, CU Inlet IV from RPV V-16-2, Inlet IV to CU Aux. Pump V-16-61, Heat Exchanger Outlet to RPV Reports that the valves V-16-1 and V-6-14 will not close Dispatches NLO to manually close the valves Re-established RB HVAC IAW Support Procedure 50 (following RB HVAC trip on high RB AP from the leak, followed by alarm RB AP LO, L6c)

Respond to radiation alarms: AREA MON HI (1OF1 k), CU SYS AREA HI (10F3k), VENT HI (10Flf) (1OF) o Verify radiation monitors (2R) o Evacuate the Reactor Building o

Confirm Secondary Containment initiations and isolations IAW Support Procedure 49 (after VENT HI alarm)

=

Initiates SGT by placing EXHAUST FAN 1-8 (1-9) switch to the HAND position (1 1 R) (CT)

Verifies SUPPLY FAN 1-12 and 1-14 OFF (Place SUPPLY FAN 1-12, 1-14 to the OFF position)

For Emergency Depressurization:

o Stops injection with Core Spray not required for adequate core cooling IAW Support Procedure 10 (1 F/2F)

Depress the OVERRIDE switches for all sensors that are lit

. Depress all ACTUATED switches

. Confirm closed Core Spray Parallel Isolation Valves not required to be open for adequate core cooling Place PARALLEL ISOL switches to the CLOSE position for Core Spray pumps not required Secure Core Spray Booster pumps not required for adequate core cooling Place BOOSTER PUMP switches to the STOP position for Page 18 of 23

Appendix D Required Operator Actions Form ES-D-2 Role Play Terminus:

Core Spray Booster pumps not required

. Secure Core Spray Main pumps not required for adequate core cooling Bypass ROPS (Reactor Overfill Protection) (4F)

Report Torus water level > 90 Open 5 EMRVs (1F/2F) (CT) 0 Place MAIN PUMP switches to the STOP position for Core Spray Main pumps not required

=.

Place AUTO DEPRESS VALVE switches in the MAN position When the RB AP Low annunciator alarms, as the NLO call the Control Room that steam is visible coming from the Reactor Water Cleanup cage and that you have exited the RB.

SRO Directs entry into EMG-3200.11, Secondary Containment Control o

o Directs isolation of systems discharging into Secondary Containment Directs Emergency Depressurization when 2 area radiation levels or 2 area temperatures exceed the MAX SAFE level (CT)

. Directs entry into EMG-3200.04A, Emergency Depressurization - No A W S Directs ROPS bypass 0

Directs 5 EMRVs open Direct stopping injection with Core Spray not required for adequate core cooling IAW Support Procedure 10 Directs verification of Torus water level Directs evacuation of Reactor Building from high radiation Directs initiation of SGT (CT)

The RPV has been emergency depressurized II No tes/Co m men ts I

I NRC 3 Page 19 of 23

Appendix D

~

~~~~

~

~~

Required Operator Actions Form ES-D-2 Op-Test No.: NRC 2006-1 Scenario No.: NRC 3 Event No.: 9 Page -

of -

Event

Description:

Failure of Standby Gas Treatment (SGT) System to auto-initiate Initiation: During the steam leak into the secondary containment Cues: Annunciator lOFlf, VENT HI (1OF)

Responds to annunciator lOFlf, VENT HI (1OF) o o

o Verifies radiation monitor VENT MANIFOLD 1,2 readings (2F)

Reports failure of SGT to auto initiate Manually starts SGT 1 (2) (1 1 R) 9 Confirm STANDBY GAS SELECT in SYS 1 (SYS 2)

Start exhaust Fan EF-1-8 (9) by placing EXHAUST FAN 1-8 (1 -9) to the HAND position Verify exhaust fan starts and valves open Verify RB Main Supply valves close Verify RB Containment Isolation valves close Verify RB Supply Fans trip and Exhaust Fans trip and Supply valves close SRO Directs manual start of SGT IAW Support Procedure 49 (CT)

Terminus: I SGT has been initiated Notes/Comments NRC 3 Page 20 of 23

Appendix D Required Operator Actions Form ES-D-2 Malfunction List Presets:

0 IRM 14 failed: MAL-NISOIOD Cleanup Pump B 00s - Pump switch in PTL Event 1 : Swap air compressors LOAXAS038 to LEAD to place air compressor #2 in Lead LOAXAS037 to LAG to place air compressor #1 in Lag Event 2: Report of oil leak on Core Spray Backup Main Pump NZOlC None Event 3: Trip of RPS MG Set A and scram of control rod 14-43 (from position 48)

Note: The 3 malfunctions below all go in toaether.

LOA-RPS001; RPS MG Set 1-1 Supply Breaker Trip MAL-CRDO11-1443 (with an 8 second delay) Scram of control rod 14-43 MAL-CRD007-1443 (with a 12 second delay) Stuck Rod 14-43 o it takes about 8 seconds for Y2 scram on loss of AC input breaker to RPS MG Set o The additional time delay lets the control rod scram part way in before becoming stuck Event 4: High temperature on control rod 02-19 MAL-CRDOl3-0219: Plugged cooling orifice for control rod 02-1 9 o it takes about 30-45 seconds for high temp. alarm to come in, and about 30-45 seconds to clear once deleted o this malfunction must be deleted when the RO takes expected actions to do stall flow for control rod 02-19 Event 5: FW Heater A3, A2, and A1 high level and fuel failures CNH-001 B to 0 (high level): High level in A3 FW Heater o shortly after high level annunciator, steam admission reverse check valve closes and moisture removal valve will auto open (both annunciated).

CNH-FWH004B to 0 delayed 15 seconds (high level): High level in A2 FW heater (and eventual heater trip)

CNH-FWH007B to 0 delayed 30 seconds (high level): High level in A1 RN heater (and eventual heater trip) o These FW heater malfunctions can all be placed in at the same time MAL-RXS001 to 0.001 over 15 minutes (added in Event 8)

Event 6: Reported seismic event ii No actions NRC 3 Page 21 of 23

ADDendix D Rea u i red Ope rat0 r Actions Form ES-D-2

.iJ Event 7: Main turbine high vibrations with high thrust bearing vibration with failure of main turbine to auto trip on high thrust bearing vibration 0

MAL-TS1002A thru TS1002J to 1 1-13 mils (all bearings high vibrations are not required)

To simulate a turbine thrust bearing trip:

0 ANN-Q-2b to ON: this simulates a high thrust bearing with failure of main turbine to auto trip (manual and other auto trips still function)

STL-TS1002 to ON:

this activates the thrust trip amber light on Panel 7F Event 8: Leak in the Cleanup System with failure of the Cleanup System isolation to close (failure of both auto and manual) Also put in Event 9 malfunction now.)

i-0 MAL-RCU013 to 20% over 15 minutes Cleanup System leak 0

Fail CU IVs:

o VLV-RCU001 to Mech Seize (V-16-1) o VLV-RC0004 to Mech Seize (V-16-14) 0 Fuel Failures o MAL-RXS001 to 0.001 over 15 minutes o ICH-RMS028A to 1.85 5-minute ramp with 2-minute delay for SDC ARM o ICH-RMS025A to 2.5 5-minute ramp with 2-minute delay for RWCU ARM o ICH-RMS027A to 1.5 5-minute ramp with 2-minute delay for IC ARM o ICH-RMS035A to 1.25 6-minute ramp with 2-minute delay for #I vent rad. monitor o ICH-RMS036A to 1.15 6-minute ramp with 2-minute delay for #2 vent rad. monitor Failure of SGTS to auto start o MAL-SCN005 Event 9: Failure of SGT to auto start on high radiation in RB HVAC discharge NOTE: This malfunction may not be required. On high RB pressure, normal HVAC trips and isolates. Secondary Containment EOP states that if RBHV isolates or shuts-down, it is restarted by Support Procedure 50. The jumper used to allow manual start of RB HVAC also over-rides an auto shutdown of RB HVAC from high RB Building HVAC radiation monitors. When these radiation monitors do go high, normal HVAC and SGTS remains unaffected. The Operator must manually start SGTS and secure the other valves.

MAL-SCN005 NRC 3 Page 22 of 23

Appendix D Required Operator Actions Form ES-D-2 Procedure ##

334 Procedure Name Revision Instrument and Service Air 93 408.12 RAP-H~c Operation of Reactor Protection System Panel 1-1 and Transformer PS-1 CRD TEMP HI 0

12 CRD Exercise and Flow Test/lST Cooling Water Header Check Valve 44 RAP-N 1 d ABN-17 RAP-N2d HP A3 REV CK VLV TRIP 0

Feedwater System Abnormal Conditions 3

HP A3 MRV OPEN 0

ABN-1 0 RAP-Dld

~~

Turbine Generator Trip 2

RWCU HELB I 2

RAP-D2d EMG-3200.11

~

RWCU HELB II 2

Secondary Containment Control 12 EMG-3200-04A RAP-1 OF1 f Emergency Depressurization - No ATWS 4

VENT HI 0

330 31 7.1 Standby Gas Treatment System 43 Feedwater Heaters 36 1

2 RAP-G~c I RPS MG SET 1 TRIP l o RAP-G 1 c I SCRAM CONTACTOR OPEN 3

4 RAP-H6a 1 ROD DRIFT 5

ABN-6 I Control Rod Drive System 12 6

7 8

302.1 I Control Rod Drive System I92 9

61 7.4.002 RAP-N3d I HP A3 LEVEL HVLO l o 10 11 12 13 14

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ABN-38 1 Station Seismic Event 13 15 RAP-Q3b I VIBRATION HIGH RAP-Q2b 1 THRUST BRG WEAR HI l o 16 17 ABN-1 1 Reactor Scram 12 18 19 20 21 22 23 24 25 NRC 3 Page 23 of 23