ML14154A393

Initial Exam 2014-301 Final Simulator Scenarios

ML14154A393
Person / Time
Site:	Browns Ferry
Issue date:	06/02/2014
From:	NRC/RGN-II
To:	Tennessee Valley Authority
Shared Package
ML14155A059	List: ML14154A324 ML14154A337 ML14154A346 ML14154A365 ML14154A369 ML14154A373 ML14154A377 ML14154A383 ML14154A390 ML14154A393 ML14154A396 ML14154A399
References
50-259/OL-14, 50-260/OL-14, 50-296/OL-14
Download: ML14154A393 (95)

Text

NRC Scenario 6 Facility: Browns Ferry NPP Scenario No.: NRC -6 Op-Test No.: 1404 Examiners: Operators: SRO:_

ATC:

BOP:

Initial Conditions: 1.3% power, operating in 2-G0I-l00-1A Section 5.4 steps 63.3 and 65.

Turnover: Warm RFPT B lAW 2-01-3, section 5.6 and then Continue to pull rods for Mode Change.

Event Maif. No. Event Type* Event Description No.

N-BOP 1 Warm RFPT B lAW 2-01-3, section 5.6 NSRO R-ATC 2 Raise power with Control Rods R-SRO C-ATC Control Rod will difficult to withdraw, control rod at position other 3 rd05r3435 C-SRO than 00, and then control rod stuck TS-SRO C-BOP 4 ed07a Loss of 480V Unit Board 2A, failure of EHC Pump 2B to auto start C-SRO C-BOP 5 rcOl RCIC inadvertently starts TS-SRO C-ATC 6 thO3b Reactor Recirculation Pump 2B trip TS-SRO 7 batch M-ALL SSI Fire 25-1 8 hpO3 I HPCI Flow controller will not operate in Auto (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor q

I

NRC Scenario 6 Critical Tasks Three CT 1- Within 10 minutes of recorded time in SSI an Operator has placed Path A Vent Flow Controller, 2-FIC-84-20, in MANUAL and 0 SCFM, at Panel 2-9-55.

1. Safety Significance:

Maintaining adequate RHR Pump NPSH.

2. Cues:

Procedural compliance.

Containment Pressure indication.

3. Measured by:

Observation 2-FIC-84-20 in manual and set at 0 SCFM.

Observation 2-FCV-84-20 closed.

4. Feedback:

Containment Pressure trend.

No flow through A vent path.

CT 2- Within 10 minutes of recorded time in SSI an Operator has initiated a controlled 100°F per hour cooldown rate using HPCI and relief valves as required.

1. Safety Significance:

Prevent Drywell Temperature from exceeding design basis temperature.

2. Cues:

Procedural compliance.

Reactor Pressure indication.

3. Measured by:

Observation HPCI in Pressure Control Mode.

Observation SRVs opened to lower pressure.

4. Feedback:

Reactor Pressure trend.

2

NRC Scenario 6 Critical Tasks Three CT 3- Within 10 minutes of recorded time in SSI an Operator has placed the following switches in Test/Inhibit, at Panel 2-9-3: ECCS SYS I HI DW PRESS Test/Inhibit, 2-HS-75-59 AND ECCS SYS II HI DW PRESS Test/Inhibit, 2-HS-75-60.

1. Safety Significance:

Prevent CAS initiation due to actual high Drywell Pressure, and minimize the number of subsequent additional actions (to secure/realign both credited and non-credited pumps).

2. Cues:

Procedural compliance.

No AUTO initiation of ECCS when Drywell Pressure exceeds 2.45 psig.

3. Measured by:

Observation 2-HS-75-59 and 60 in Test/Inhibit.

Observation No AUTO initiation on high drywell pressure.

4. Feedback:

ECCS Pumps green lights ON and Red Lights Off.

3

NRC Scenario 6 EVENTS

1. BOP Operator warms RFPT B lAW 2-01-3 Feedwater System, section 5.6.

2. ATC Continues Power ascension with control rods.

3. During power ascension Control Rod 34-3 5 will fail to withdraw. The crew will respond lAW 2 85. Once Drive water pressure is at 350 psig or greater the control rod will triple notch to position 14 which is one notch beyond the banked position of 12. The Unit Supervisor should enter 2-A0I-85-7 for a mispositioned control rod. All attempts to insert the control rod to the correct position will fail.

The control rod will be declared stuck and the SRO will enter Tech Specs and determine TS 3.1.3 condition A.

4. A loss of 480V Unit Board 2A will occur. EHC Pump 2A will trip due to loss of power and the standby pump will not auto start, BOP operator will start EHC Pump 2B to prevent a loss of EHC pressure and closure of Turbine Bypass Valves.

5. RCIC inadvertent initiation. The BOP Operator will respond lAW ARPs. BOP Operator will verify that level is in normal band and secure RCIC. The SRO will evaluate Technical Specification 3.5.3 Condition A.

6. Reactor Recirculation Pump 2B will trip, the crew will respond lAW 2.-AOI-68-1A. The SRO will enter Tech Specs and determine TS 3.4.1 condition A.

7. The crew will respond to a fire and enter 0-AOI-26-l and SSI 25-1, Intake Pumping Station Pump El. 550, Cable Tunnel to Fire Door 440, RHRSW Room B, RHRSW Pump Room D. The SRO will also enter E0I-l and 2 and perform actions that do not conflict with the SSI guidance.

8. Shortly after entering the SSI the crew will commence a controlled cooldown lAW the SSI utilizing HPCI and SRVs, the HPCI flow controller will fail in Auto but will operate in manual.

4

NRC Scenario 6 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:

Reactor Level is maintained Controlled Cooldown in progress SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 6 9 Total Malfunctions Inserted: List (4-8) 3 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 0 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No) 5

NRC Scenario 6 Scenario Tasks TASK NUMBER RO Q Warm REPT 2B JAW 2-01-3 ROU-003-N0-23 259001 A4.02 3.9 3.7 Raise Power with Control Rods RO U-085-N0-07 SRO S-000-AD-31 2.2.2 4.6 4.1 Control Rod difficult to withdraw from a position other than 00 ROU-085-NO-19 201001 A4.04 3.1 3.1 SRO S-000-AD-3 1 Control Rod Mispositioned RO U-085-AB-07 201002 A2.02 3.2 3.3 SRO S-085-AB-07 Reactor Recirculation Pump Trip RO U-068-AB-1 202001A2.03 3.6 3.7 SRO S-068-AB-1 Loss of 480V Unit BD 2A RO U-57B-AL-06 226001 A2.04 3.8 4.2 SRO S-57B-N0-07 RCIC Inadvertent Start ROU-071-N0-5 217000A2.01 3.8 3.7 SRO S-000-AD-27 SSI FIRE RO U-000-EM-85 600000 AA2.16 3.0 3.5 RO U-000-SS-30 RO U-000-N0-32 SRO S-000-EM-30 SRO S-000-SS-30 SRO S-000-SS-31 6

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NRC Scenario 6 Simulator Instructor IC 97 Batch File NRC/l4O4nrc-6 imf cu06b RWCU valve failure trg e5 NRC/ehc trg e5 bat NRC/ehcpumptrip-i trg elO NRC/ads 1-179 trg elO = dmfad0lm imf rd06r343 5 stuck control rod imf rd26b triple notch Preference File NRC/l4O4nrc-6 pfk0i tog pfk 02 ann silence pfk 03 bat NRC/l4O4nrc-6 pfkO4 imfth03b 2B Reactor Recirc Pump trip pflc 05 imfed07a 480V unit bd 2A loss pfk 06 imfrc02 RCIC start pfk 07 imfth30d 28 level instrument fails low pfk 08 ior zdihs2388a start RHRSW Pump start B3 pflc 09 ior zdihs23 la start RHRSW Pump start Al pfk 10 ior zdihs238a start RHRSW Pump start Cl pfk 11 dor zdihs23 1 a pflc 12 dor zdihs238a pflcsl pflc s2 imfrdO6r3435 Stuck Control Rod pfk s3 bat ulu3 scram pflc s4 bat NRC/i 404-25-1 pfk s5 bat NRC/1404-25-la pflc s6 pflcs7 pfk s8 ior z1ohs6749a2[i] off RHRSW Cl Supply valve pflc s9 ior zlohs6749a2[2] off pfk slO mrf edl6a emerg_on pfk si I mrfedl6b emerg_on pfk s12 mrfedl6c emergon 8

NRC Scenario 6 Batch File NRC/1404-25-1 imfth30h 100 LT 3-208d failed high imfth3lg 100 PT 3-207 failed high imf th24d condensing pot 3-822 equalization imf ad02g SRV 1-4 imf ad02k SRV 1-41 imf ad02e SRV 1-31 imf ad02b SRV 1-19 imfad0lm35 SRV 1-179 open ior zdihs0 11 8a close/auto SRV 1-18 ior zdihs0 123 close/auto SRV 1-23 ior zdihs0 11 80a close/auto SRV 1-180 ior zdihs0l 142 close/auto SRV 1-42 imfhpo3 10 HPCI flow controller imf sw03f RHRSW Pump B3 trip irff04a start Fire Pump A start irff,04b irffp04c Batch File NRC/1404-25-la imf th03a imf tho3b SSI Attachment action ior zlohs68aii[2j off ior zlohs685aii[1j off 9

NRC Scenario 6 Scenario 6 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 97 restorepref NRC/i 4O4nrc-6 Simulator Setup Load Batch F3 NRC/l4O4nrc-6 Simulator Setup Verify file loaded Verify Rod Worth Minimizer Working, Reset RWM_screen_alarms RCP required (1% 8%),

- Marked up copy of 2-GOI-100-1A Unit Startup 10

NRC Scenario 6 Simulator Event Guide:

Event 1 Normal: Warm RFPT 2B lAW 2-01-3 SRO Directs Warm RFPT 2B lAW 2-01-3 section 5.6 BOP Warm RFPT 2B lAW 2-01-3 section 5.6 5.6 Warming the Second and Third RFPIRFPT

[3.21] PERFORM RFPT Trip Test on Panel 2-9-6 as follows:

• DEPRESS RFPT 2B TRIP, using 2-HS-3-151A, and VERIFY HP and LP Stop Valves close by using the position lights.

[3.221 CHECK Turning Gear automatically engages or RFP rolling on minimum flow.

[3.23] PERFORM RFPT Trip Reset:

B. DEPRESS RFPT 2B TRIP RESET, 2-HS-3-150A and CHECK the following:

• Blue light extinguishes.

• HP Stop Valves open.

• LP Stop Valves open.

CAUTION Do NOT raise RFP discharge pressure to greater than Reactor Pressure to prevent J) injection to vessel.

NOTE Normal operating range for RFP lube oil to bearings is 110°F to 120°F. Illustration 7 has instructions to control Raw Cooling Water through RFP lube oil cooler.

[3.24] START RFPT from Panel 2-9-6 as follows:

• PLACE RFPT 2B START/LOCAL ENABLE, 2-HS-46-138A, in START and OBSERVE RFPT accelerates to approximately 600 rpm on RFPT 2B SPEED, 2-SI-46-9A.

11

NRC Scenario 6 Simulator Event Guide:

Event 1 Normal: Warm RFPT 2B lAW 2-01-3 BOP Warm RFPT 2B JAW 2-01-3 section 5.6

[3.25] IF the lube oil to the bearings is below 110°F, perform the following:

ADJUST RFPT 2B SPEED CONT RAISE/LOWER switch, 2-HS-46-9A, as necessary to RAISE RFPT speed to approximately 1100 rpm.

[3.26] WHEN lube oil to the bearings reaches 110°F, perform the following:

• REDUCE REPT speed using RFPT 2B SPEED CONT RAISE/LOWER switch, 2-HS-46-9A, until RFP 2B discharge pressure, 2-PI-3-9A on Panel 2-9-6 is less than Reactor pressure.

12

NRC Scenario 6 Simulator Event Guide:

Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36 SRO Direct Power increase using Control Rods per 2-GOT-i 00-lA, Section 5.4 5.4 Withdrawal of Control Rods while in Mode 2

[64] VERIFY IRMIAPRM overlap by operator visual observation before exceeding 5% power.

[66] CONTINUE to withdraw control rods to raise Reactor power lAW 2-01-85 ATC Raise Power with Control Rods JAW 2-01-85, Section 6.6 Group 33 10-35,26-51, 34-51, 50-35, 50-27,34-11,26-11, 10-27 from 08 to 12 Group 34 18-43, 42-43, 42-19, 18-19 from 08 to 12

=

Group 35 26-3 5, 34-3 5, 34-27, 26-27 from 08 to 12

=

Group 36 02-35, 26-59, 34-59, 58-35, 58-27, 34-03, 26-03, 02-27 from 00 to 12 6.6.1 Initial Conditions Prior to Withdrawing Control Rods

[2] VERIFY the following prior to control rod movement:

. CRD POWER, 3-HS-85-46 in ON.

. Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing (not required with no fuel in RPV).

- 6.6.2 Actions Required During and Following Control Rod Withdrawal

[1] IF control rod fails to withdraw, THEN Refer to Section 8.15 for additional methods to reposition control rod.

[2] IF control rod double notches, or withdraws past its correct/desired position, THEN Refer to Section 6.7 for inserting control rod to its correct/desired position.

[3] IF at any time while driving a selected rod during the performance of this section, the Control Rod moves more than one notch from its intended position, THEN Refer to 2-AOI-85-7, MISPOSITIONED CONTROL ROD.

[4] OBSERVE the following during control rod repositioning:

• Control rod reed switch position indicators (four rod display) agree with indication on Full Core Display.

• Nuclear Instrumentation responds as control rods move through the core (This ensures control rod is following drive during Control Rod movement.)

[5] ATTEMPT to minimize Automatic RBM Rod block as follows:

• STOP Control Rod Withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM Displays on Panel 9-5 and perform step 6.6.2[6].

13

NRC Scenario 6 Simulator Event Guide:

Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36

[61 IF Control Rod movement was stopped to keep from exceeding a RBM Setpoint or ATC was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REiNITIALIZE the RBM:

[6.1] PLACE the CRD Power, 2-HS-85-46 to the OFF position to deselect the control Rod.

[6.2] PLACE the CRD Power, 2-HS-85-46 to the ON position.

[6.3] IF desired, THEN CONTINUE to withdraw Control Rods and PERFORM applicable section for Control Rod withdraw.

6.6.3 Control Rod Notch Withdrawal

[1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 2-XS-85-40.

[2] OBSERVE the following for selected control rod:

• CRD ROD SELECT pushbutton is brightly ILLUMiNATED.

• White light on the Full Core Display iLLUMINATED

• Rod Out Permit light ILLUMINATED.

[3] VERIFY ROD WORTH MINIMIZER operable and LATCHED in to correct ROD GROUP when Rod Worth Minimizer is enforcing.

[4] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.

[5] OBSERVE control rod settles into desired position AN]) ROD SETTLE light extinguishes.

14

NRC Scenario 6 Simulator Event Guide:

Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36 ATC [6] IF control rod is notch withdrawn to rod notch Position 48, THEN PERFORM control rod coupling integrity check as follows:

[6.1] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.

[6.2] CHECK control rod coupled by observing the following:

• Four rod display digital readout AND full core display digital readout AND background light remain illuminated.

• CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window

14) does not alarm.

[6.3] CHECK control rod settles into Position 48 and ROD SETTLE light extinguishes.

[6.41 IF control rod coupling integrity check fails, THEN Refer to 2-AOI-85-2.

6.6.4 Continuous Rod Withdrawal NOTES

1) Continuous control rod withdrawal may be used when a control rod is to be withdrawn greater than three notches.

2) When in areas of high notch worth, single notch withdrawal should be used instead of continuous rod withdrawal. Information concerning high notch worth is identified by Reactor Engineering in Control Rod Coupling Integrity Check, 2-SR-3 .1.3 .5A.

3) When continuously withdrawing a control rod to a position other than position 48, the CRD Notch Override Switch is held in the Override position and then the CRD Control Switch is held in the Rod Out Notch position.

• Both switches should be released when the control rod reaches two notches prior to its intended position. (Example: If a control rod is to be withdrawn from position 00 to position 12, the CRD Notch Override Switch and the CRD Control Switch would be used to move the control rod until reaching position 08, then both switches would be released.)

• If the rod settles in a notch prior to the intended position, the CRD Control Switch should be used to withdraw the rod to the intended position. (using the above example; If the control rod settles at a notch prior to the intended position of 12, the CRD Control Switch would be used to withdraw the control rod to position 12.)

15

NRC Scenario 6 Simulator Event Guide:

Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36 ATC [1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 2-XS-85-40.

[2] OBSERVE the following for selected control rod:

• CR]) ROD SELECT pushbutton is brightly ILLUMINATED.

• White light on the Full Core Display ILLUMINATED

• Rod Out Permit light ILLUMINATED.

[3] VERJZFY ROD WORTH MiNIMIZER operable and LATCHED in to correct ROD GROUP when Rod Worth Minimizer is enforcing.

[4] VERIFY Control Rod is being withdrawn to a position greater than three notches.

[5] IF withdrawing the control rod to a position other than 48, THEN PERFORM the following: (Otherwise N/A)

[5.1] PLACE and HOLD CR1) NOTCH OVERRIDE, 2-HS-85-47, in OVERRIDE.

[5.2] PLACE and HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.

[5.3] WHEN control rod reaches two notches prior to the intended notch, THEN RELEASE both CR]) NOTCH OVERRIDE, 2-HS-85-47, and CR])

CONTROL SWITCH, 2-HS-85-48.

[5.4] IF control rod settles at notch before intended notch, ThEN PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.

[5.5] WHEN control rod settles into the intended notch, THEN CHECK the following:

• Four rod display digital readout and full core display digital readout and background light will remain illuminated.

• CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window

14) does not alarm.

[5.6] CHECK control rod settles at intended position and ROD SETTLE light extinguishes.

ATC During power ascension IRM B fails to respond to continuous steady counts, ATC reports to Unit Supervisor SRO Directs Startup to continue, have all needed IRM instruments. Need 6 of 8 IRMs.

16

NRC Scenario 6 Simulator Event Guide:

Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36 ATC NOTE When continuously withdrawing a control rod to position 48, the control rod coupling integrity check can be performed by one of the two following methods:

1) Coupling integrity check while maintaining the CRD Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position. If this method is selected, perform Step 6.6.4[6] and N/A Step 6.6.4[7].

2) Coupling integrity check after releasing the CR1) Notch Override Switch and the CRD Control Switch. If this method is selected, perform Step 6.6.4[7] and N/A Step 6.6.4[6}.

[6] IF continuously withdrawing the control rod to position 48 and performing the control rod coupling integrity check in conjunction with withdrawal, THEN PERFORM the following: (Otherwise N/A)

[6.1] PLACE and HOLD CR1) NOTCH OVERRIDE, 2-HS-85-47, in OVERRIDE.

[6.2] PLACE and HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.

[6.3] MAINTAIN the CRD Notch Override Switch in the Override position and the CR1) Control Switch in the Rod Out Notch position with the control rod at position 48.

[6.4] CHECK control rod coupled by observing the following:

• Four rod display digital readout and full core display digital readout and background light will remain illuminated.

• CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window

14) does not alarm.

[6.5] RELEASE both CRD NOTCH OVERRIDE, 2-HS-85-47, and CR1)

CONTROL SWITCH, 2-HS.85-48.

[6.6] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.

[6.7] IF control rod coupling integrity check fails, ThEN Refer to 2-AOI-85-2.

17

NRC Scenario 6 Simulator Event Guide:

Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36 ATC [7] IF continuously withdrawing the control rod to position 48 control rod coupling integrity check is to be performed after the CR1) NOTCH OVERRIDE, 2-HS-85-47, and CRD CONTROL SWITCH, 2-HS-85-48 are to be released, THEN PERFORM control rod coupling integrity check as follows (otherwise N/A):

[7.11 PLACE and HOLD CR1) NOTCH OVERRIDE, 2-HS-85-47, in OVERRIDE.

[7.2] PLACE and HOLD CR1) CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.

[7.3] WHEN position 48 is reached, THEN RELEASE CR1) NOTCH OVERRIDE, 2-HS-85-47, and CRD CONTROL SWITCH, 2-HS-85-48.

[7.4] VERIFY control rod settles into position 48.

[7.5] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.

[7.6] CHECK control rod coupled by observing the following:

• Four rod display digital readout and full core display digital readout and background light will remain illuminated.

• CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window

14) does not alarm.

[7.7] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.

[7.8] IF control rod coupling integrity check fails, THEN Refer to 2-AOI-85-2.

6.6.5 Return to Normal after Completion of Control Rod Withdrawal

[1] WHEN control rod movement is no longer desired AND deselecting control rods is desired, THEN:

[1.1] PLACE CR1) POWER, 2-HS-85-46, in OFF.

[1.2] PLACE CRD POWER, 2-HS-85-46, in ON.

18

NRC Scenario 6 Simulator Event Guide:

Event 3 Component: Control Rod 34-35 difficult to withdraw 8.15 Control Rod Difficult to Withdraw ATC [1] VERiFY the control rod will not notch out. Refer to Section 6.6.

[2] REVIEW all Precautions and Limitations in Section 3.0

[3] IF RWM is enforcing, THEN VERIFY RWM is operable and LATCHED in to the correct ROD GROUP.

NOTES

1) Steps 8.15[4] through 8.15[6] should be used when the control rod is at Position 00 while Step 8.15 [7] should be used when the control rod is at OR between Positions 02 and 46.

2) Double clutching of a control rod at Position 00 will place the rod at the overtravel in stop, independent of the RMCS timer, allowing maximum available time to establish over-piston pressure required to maintain the collet open and prevent the collet fingers from engaging the 00 notch.

3) Step 8.15 [4] may be repeated as necessary until it is determined that this method will not free the control rod.

[7] IF the control rod is at or between Positions 02 and 46, THEN PERFORM the J following to withdraw the control rod using elevated drive water pressure:

[7.1] RAISE the CR1) DRiVE WTR HDR DP, 2-PDI 1 7A, to 300 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A.

[7.2] ATTEMPT to withdraw the Control Rod using CR1) CONTROL SWITCH, 2-HS-85-48.

[7.3] IF the control rod successfully notches out, THEN LOWER CRD DRIVE WTR HDR DP, 2-PDI-85-17A, to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A, AND PROCEED TO Section 6.6.

CAUTION To prevent a drive from double notching in a high rod worth region and to reduce exposure of drive seals and directional control valves to excessive pressures, the CRD DRIVE WTR HDR DP should be returned to between 250 psid and 270 psid as soon as possible.

19

NRC Scenario 6 Simulator Event Guide:

Event 3 Component: Control Rod 34-3 5 difficult to withdraw

[7.4] IF the control rod still fails to NOTCH OUT, THEN RAISE CRD DRiVE WTR HDR DP, 2-PDI-85-17A, to 350 psid, using CR1) DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A.

[7.5] ATTEMPT to withdraw the Control Rod using CRD CONTROL SWITCH, 2-HS-85-48.

[7.6] LOWER CR1) DRIVE WTR HDR DP, 2-PDI-85-17A, to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A.

[7.7] IF the control rod still fails to NOTCH OUT using elevated CRD DRiVE WTR HDR DP, THEN CONTACT Reactor Engineer and NOTIFY Unit Supervisor for further instructions.

ATC Will raise drive water pressure to 350 psid to successfully move Control Rod 34-3 5 from position 8 Driver when ATC raises drive water pressure to 350 psid, delete stuck rod 34-35 NOTE When the control rod moves with drive water pressure at 350 psid the control will triple notch Driver After control rod triple notches RWM Alarm will come in when that alarm energizes insert Shift F2 to stick control rod 34-3 5 (rd06r3435)

ATC Report Control Rod 34-35 triple notched to position 14 SRO Enter 2-AOI-85-7, Mispositioned Control Rod Mispositioned Control Rod 2-AOI-85-7 4.0 OPERATOR ACTIONS 4.1 Immediate Actions None 4.2 Subsequent Actions

[1] STOP all intentional control rod movement.

[2] IF Control Rod is determined to be mispositioned, THEN NOTIFY the following:

• Reactor Engineer (RE)

• Shift Technical Advisor (STA)

• Unit Supervisor

• Shift Manager (SM)

• Operations Superintendent.

20

NRC Scenario 6 Simulator Event Guide:

Event 3 Component: Control Rod 34-35 difficult to withdraw

[3] fi? the Control Rod is> 2 notches from the intended position, THEN ATC PERFORM the following: (Otherwise N/A)

CAUTION NRC/C] Operations outside of the allowable regions shown on the Recirculation System Operating Map could result in thermal-hydraulic power oscillations and subsequent fuel damage. Monitoring to be performed during a power decrease and required actions are contained in 2-GOI-100-12. [NCO 940245010]

NOTE

1) Rod moves to recover from mispositioned rod will NOT be any different from normal rod moves.

2) Power level at which recovery is performed and movement of other rods to support recovery are as allowed by 0-TI-248.

3) Rod movement during recovery is governed by 2-GOI-100-1A, 2-GOI-100-12, 2-GOI-100-12A and 2-01-85.

[4] IF the Control Rod is 2 notches from the intended position, THEN PERFORM the following:

[4.1] OBTAIN recommendation from the Reactor Engineer.

[4.2] IF a Reactor Startup or Shutdown is not in progress, THEN VERIFY 2-GOl- 100-12, Power Maneuvering has been entered if a power change is anticipated. (Otherwise N/A)

(1

. [4.3] IF required to allow rod movement to correct the rod position error, THEN REDUCE core thermal power as recommended by the Reactor Engineer/Reactivity Control Plan. (Otherwise N/A)

[4.4] MOVE Control Rods with the Unit Supervisor permission and Shift Manager concurrence to recover from the rod positioning error as recommended by the Reactor Engineer/Reactivity Control Plan.

[6] iF evidence of fuel damage, THEN REFER TO EPIP-1 for emergency classification.

[7] INTIATE a Service Request/PER for Control Rod error or mispositioned Control Rod.

Driver When called as RE concur with the Unit Supervisors recommendation Ensure Applicant makes_a recommendation do NOT prompt.

SRO Direct ATC to insert control rod to position 12 ATC Attempts to insert Control Rod 34-35 to position 12, reports control rod failed to move.

21

NRC Scenario 6 Simulator Event Guide:

Event 3 Component: Control Rod 34-3 5 difficult to INSERT 8.16 Control Rod Difficult to Insert

[1] VERIFY the control rod will not notch in, in accordance with Section 6.7 or 8.19.

[2] REVIEW all Precautions and Limitations in Section 3.0.

[3] IF RWM is enforcing, THEN VERIFY RWM operable and LATCHED in to the correct ROD GROUP.

ATC

[4] ChECK CRD SYSTEM FLOW is between 40 gpm and 65 gpm, indicated by 2-FIC-85-1 1.

[5] CHECK CR1) DRIVE WTR HDR DP, 2-PDI-85-17A is between 250 psid and 270 psid.

[6] IF the CRD SYSTEM FLOW or CRD DRIVE WTR HDR DP had to be adjusted, THEN PROCEED TO Section 6.7.

[7] IF control rod motion is observed, but the CRD fails to notch-in with normal operating drive water pressure, THEN: NA

[8] IF the control rod problem is believed to be air in the hydraulic system, THEN

[9] IF the control rod begins to insert normally, THEN

( [10] IF the control rod still fails to notch in AND the control rod problem is believed to

- be air in the hydraulic system, THEN

[12] IF the control rod still fails to notch in, THEN:

[12.1] NOTIFY the Unit Supervisor and Reactor Engineer to Refer to section Stuck Control Rod-Test to distinguish a Hydraulic Problem from Mechanical Binding, TI-20, and RETURN to Section 8.16.

[12.2] REQUEST the Unit Supervisor and Reactor Engineer to evaluate the control rod operability. Refer to Tech Spec 3.1.

SRO Contact Reactor Engineering and Evaluate Tech Spec 3.1 Driver Acknowledge RE communication will commence working on a course of action Driver When directed by the NRC insert preference key F5 for loss of 480V Unit BD 2A 22

NRC Scenario 6 Simulator Event Guide:

Event 3 Component: Control Rod 34-3 5 difficult to iNSERT SRO Contact Reactor Engineering and Evaluate Tech Spec 3.1 3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.

APPLICABILITY: MODES 1 and 2 Condition A: One withdrawn control rod stuck Required Action A. 1: Verify stuck rod separation criteria are met Completion Time: Immediately AND Required Action A.2: Disarm the associated control rod drive (CRD)

Completion Time: 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> AND Required Action A.3: Perform SR 3.1.3.3 for each withdrawn OPERABLE control rod Completion Time: 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of Condition A concurrent with Thermal power greater than the low power setpoint (LPSP) of the RWM AND fl Required Action A.4: Perform SR 3.1.1.1 Completion Time: 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> triver Acknowledge RE communication will commence working on a course of action Driver When directed by the NRC insert preference key F5 for loss of 480V Unit BD 2A 23

NRC Scenario 6 Simulator Event Guide:

Event 4 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay.

BOP Responds to the following alanus; 7A-31, 8C-3, 8C-lO, 8C-15, 8C-16, 8C-25, 8B-16, 7B-1 and 7B-15.

7A-3 1 GEN BUS DUCT FAN FAILURE A. VERIFY Main Bus Cooling Fans, 2-HS-262-1A or 2-HS-262-2A, indicates running on Panel 2-9-8 AN]) START GEN BUS DUCT HX FAN A(B) using 2-HS-2-6-2-1A(2A), on panel 2-9-8 to start the standby fan.

Starts Bus Duct Cooling Fan 2B 8B-16 480V UNIT BD 2A UV ORXFR A. VERIFY automatic action has occurred.

B. INSPECT 480V Unit Bd A for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.

C. REFER TO O-OI-57B to re-energize board.

D. REFER TO appropriate 01 for recovery or realignment of equipment.

Reports trip of 480V Unit Bd 2A, dispatches operators 8C-3 480V RX BLDG VENT BD 2A UV OR XFR A. VERIFY automatic action has occurred.

B. CHECK or START refuel floor and reactor zone exhaust fans 2A or 2B.

C. CHECK board for abnormal condition: relay targets, smoke, burned paint, breaker position, etc.

D. REFER TO O-0I-57B to re-energize or transfer board.

E. REFER TO appropriate 0! for recovery or realignment of equipment.

24

NRC Scenario 6 Simulator Event Guide:

Event 4 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay.

BOP Responds to the following alarms; 7A-22, 8C-3, 8C-1O, 8C-15, 8C-16, 8C-25, 8B-16, 7B-1 and 7B-15.

8C-1O 480V RX BLDG VENT BD 2B UV OR XFR A. VERIFY automatic action has occurred.

B. CHECK or START refuel floor and reactor zone fans 2A or 2B.

C. CHECK 480V Reactor Bldg Vent Bd 2B for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.

D. REFER TO O-OI-57B to re-energize board.

E. REFER TO appropriate 01 for recovery or realignment of equipment.

8C-15 480V TB VENT BD 2A UV OR XFR A. VERIFY alarm by checking the following:

• Associated annunciator, TURBINE BLDG VENTILATION ABNORMAL, (2-XA-55-3D, Window 4) in alarm.

• Power light on MTOT vapor extractor and EHC fluid heaters, Panel 2-9-7.

B. DISPATCH Personnel to 480V Turb Bldg Vent Bd 2A to CHECK equipment and board status for abnormal conditions.

• Mechanical spaces supply fan 2A, 2B and exhaust fan.

• Turb room supply fan 2A, 2B and exhaust fans 2A, 2B, 2C, 2D.

• EHC fluid transfer and filtering pump.

C. REFER TO O-0I-57B to re-energize or transfer the board.

D. REFER TO appropriate 01 for recovery or realignment of equipment.

If called to verii if the following boards transferred report that these boards have auto transferred to their alternate supply. 480V RX BLDG VENT BD 2A, 480V RX BLDG Driver VENT BD 2B, 480V TB VENT BD 2A, 480V TURB MOV BD 2A, and 480V CNDS DEMIN BD 2 25

NRC Scenario 6 Simulator Event Guide:

Event 4 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay.

BOP Responds to the following alanns; 7A-22, 8C-3, 8C-1O, 8C-15, 8C-16, 8C-25, 8B-16, 7B-1 and 7B-15.

8C-16 480V TURB MOV BD 2A UV OR XFR A. VERIFY alarm by checking light indication to the following equipment:

• RFW heaters (2B1,2B2,2C1,2C2) extraction isolation valves.

• RFPT 2B 2B2 Main Oil Pump.

• RFPT 2C 2C1 Main Oil Pump.

B. CHECK board inspected for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.

C. REFER TO ICS screen VFDAAL or VFDBAL and verify PROCESS ALARM Internal HX Fan Power status is OK.

D. REFER TO O-OI-57B to re-energize or transfer the board.

E. REFER TO appropriate 01 for recovery or realignment of equipment.

8C-25 480V CNDS DEMIN BD 2 UV OR XFR A. VERIFY automatic transfer by dispatching personnel to 480V Cnds Demin Bd 2 to check for the following:

• Power available lights illuminated.

• Normal disconnect switch 1A open and alternate disconnect 2A closed.

• Any abnormal conditions such as breaker trips.

B. NOTIFY Radwaste Operator.

C. IF power NOT available, THEN OBTAIN status of the following from Radwaste:

• Condensate demins precoat operation.

• Condensate backwash transfer operation.

• Backwash receiver pit floor drains.

D. REFER TO O-0I-57B for power restoration or transfer instructions.

If called to verify if the following boards transferred report that these boards have auto

. transferred to their alternate supply. 480V RX BLDG VENT BD 2A, 480V RX BLDG Driver VENT BD 2B, 480V TB VENT BD 2A, 480V TURB MOV BD 2A, and 480V CNDS DEMIN BD 2 26

NRC Scenario 6 Simulator Event Guide:

(3 Event 4 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key ES for imfedO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay.

BOP Responds to the following alarms; 7A-22, 8C-3, 8C-l0, 8C-15, 8C-16, 8C-25, 8B-16, 7B-1 and 7B-15.

7B-1 EHC HYD FLUID HDR PRESS LOW A. VERIFY Standby EHC PUMP 2B(2A), 2-HS-47-2A(1A) running.

B. ChECK EHC HEADER PRESSURE indicator, 2-PI-47-7 between 1550 and 1650 psig.

C. DISPATCH personnel to inspect EHC pump unit.

D. IF EHC Hydraulic system fails, THEN VERIFY turbine trips at or below 1100 psig.

7B-15 STANDBY EHC PUMP FAILED A. On Panel 2-9-7:

VERIFY alarm by checking EHC HEADER PRESSURE indicator, 2-PI-47-7.

1.

2. VERIFY EHC PUMP 2B, 2-HS-47-2A and/or EHC PUMP 2A, 2-HS-47-1A running.

3. CHECK EHC PUMP 2B PUMP MTR CURRENT 2-EI-47-2 and/or EHC PUMP 2A PUMP MTR CURRENT 2-EI 1.

NOTE Lights extinguish at 1300 psig lowering and illuminate at 1500 psig rising.

4. CHECK lights above EHC PUMP 2A TEST pushbutton 2-HS-47-4A and EHC PUMP 2B TEST pushbutton 2-HS-47-5A.

B. DISPATCH personnel to pumping unit to check for abnormal conditions.

C. IF EHC Hydraulic System fails, THEN VERIFY turbine trips at or below 1100 psig.

BOP Starts Standby EHC Pump 2B and verifies EHC pressure returns to normal Driver When directed by NRC insert preference key F7 followed 5 seconds later by preference key F6, and when asked to investigate in aux instrument room report LIS 3-58B is failed at

(-) 95 inches.

27

NRC Scenario 6 Simulator Event Guide:

Event 5 Instrument: RCIC inadvertent start BOP Responds to RCIC inadvertent start Report RCIC has initiated If RCIC initiates on an invalid initiation signal, it is expected that the operator report the initiation to the US, verify the initiation signal is not valid, and with concurrence from the US trip RCIC.

BOP Trip RCIC Driver When sent to the breaker for the RCIC Minimum Flow Valve 250V DC RMOV BD 2B compartment 5D wait two minutes and when told to de-energize the 7 1-34 valve ior ypovfcv7 134 fail_now Crew Determines RCIC minimum flow valve being open is adding water o the suppression pool SRO Directs RCIC minimum flow valve isolated BOP Coordinates with plant operator to de-energize 71-34 when the valve is closed.

SRO Evaluate Technical Specifications 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure

> 150 psig.

Condition A: RCIC System inoperable.

Required Action A. 1: Verify by administrative means HPCI is OPERABLE.

Completion Time: Immediately Required Action A.2: Restore RCIC System to OPERABLE status.

Completion Time: 14 Days Driver When directed by Lead Evaluator insert F4 for RR Pump 2B trip 28

NRC Scenario 6 Simulator Event Guide:

Event 6 Component: RR Pump 2B Trip ATC Responds to numerous alarms and report trip of RR 2B Pump SRO Enter 2-AOI-68-1A, Recirc Pump Trip/Core Flow Decrease OPRMs Operable

[2] IF a single Recirc Pump tripped, THEN CLOSE tripped Recirc Pump discharge valve.

ATC [3] IF Region I or II of the Power to Flow Map is entered, THEN IMMEDIATELY take actions to INSERT control rods to less than 95.2% loadline. Refer to O-TI-464, Reactivity Control Plan Development and Implementation.

ATC Range IRMs Down

[9] NOTIFY Reactor Engineer to PERFORM the following:

[10] WHEN the Recirc Pump discharge valve has been closed for at least five minutes (to prevent reverse rotation of the pump), THEN OPEN Recirc Pump discharge valve as necessary to maintain Recirc Loop in thermal equilibrium.

SRO Tech Spec 3.4.1 Recirculation Loops Operating

} LCO 3.4.1 Two recirculation loops with matched flows shall be in operation.

APPLICABILITY: MODES 1 and 2.

Condition A: Requirements of the LCO not met.

Required Action A. 1: Satisfy the requirements of the LCO.

Completion Time: 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

29

NRC Scenario 6 Simulator Event Guide:

Event 7 MAJOR: SSI Fire Driver When directed by NRC or if the crew scrams the reqtor insert mrf f0 1 alarm for the fire alarm.

Commence Radio traffic between the Driver and theshift manager Report a fire reported at RHRSW Pump Rooms Insert Preference Key F8 start of RHRSW Pump B3 Two minutes after initial report, report as the Incident Commander that there was a fuel oil truck accident at the RHRSW Pump Rooms, RHRSW Pump Room B is fully involved and the fire appears to be spreading to RHRSW Pump Room D Insert Preference Key F9 start of RHRSW Pump Al Has incident Commander request off site Fire fighting support.

Insert Preference Key F 10 start of RHRSW Pump Cl Insert Preference Key Fl 1 and F12 after RHRSW Pump Cl start Two minutes later inform crew Fire is still active but silencing fire alarm Driver When Reactor is scrammed insert preference key shiftF3 to scram Unit 1 and 3 and Shift F4 for bat file NRC/1404 1, after batch file 1404-25-1 is started go to remote function summary (RFS) and start the three fire pumps.

Crew Respond to Fire Alarm

) SRO Enter O-AOI-26-l, Fire Response 4.0 OPERATOR ACTIONS 4.1 Immediate Actions None 4.2 Subsequent Actions

[1] CONFIRM fire area and ANNOUNCE over the PA.

[2] WHEN requested by the Incident Commander, ThEN TRIP fans, air handling units, or cooling unit for affected area. REFER TO Attachment 1.

[3] REFERTOEPLP-17.

NOTES

1) The Shift Manager will remain in communication with the Incident Commander and reference O-SSI-OO 1 for applicability based on the severity of the fire.

2) Each Safe Shutdown Instruction contains illustrations which depict the credited plantlunit equipment and instrumentation for that specific Fire Area.

3) The AUOs are assembled in the Control Rooms to ensure SSI manual actions can be completed within the required time.

4) To ensure that in the event of a APP R fire, containment pressure is not vented below that which is needed to maintain RHR pump NPSH, maintain I (2,3)-FIC 19 in normal position of Manual and 0 scfm.

30

NRC Scenario 6 Simulator Event Guide:

Event 7 MAJOR: SSI Fire

[4] IF directed by the Unit Supervisor, THEN PERFORM the following:

[4.1] NOTIFY AUOs to report to their assigned Control Room(s), all other AUOs will report to Unit 2 MCR.

[4.2] REVIEW applicable SSI for the fire area.

[4.3] DISTRIBUTE SSI attachments to assigned AUOs and review the Section 1.0 time critical actions for the Fire Area.

[4.4] NOTIFY the AUOs to:

• OBTAIN an Appendix R Radio.

AND

• STANDBY until fire is out OR determination is made to enter applicable SSI.

[4.5] NOTIFY the Unit Operators to OBTAIN an Appendix R Radio.

[5] MONITOR Control boards indications of equipment failures or spurious operation.

Crew

[6] VERIFY PATH B VENT FLOW CONT, 1(2,3)-FCV-84-19 in Manual and 0 SCFM.

Crew Report spurious start of RHRSW Pumps

[7] IF directed by the US, THEN PERFORM the following:

)

[7.1] DISPATCH AUOs to location of initial time critical actions contained in respective SSI Attachment Section 1.0 (some attachments are long term actions and may not require immediate dispatch of the AUO).

[7.2] DIRECT AUOs to standby until notification is received that fire is out or notification to perform assigned SSI Attachment.

SRO Enters 0-551-25-1 O-SSI-25-1 2.0 UNIT 2 CONTROL ROOM OPERATOR ACTIONS NOTE When 0-SSI-25-1 is in use, other plant procedures such as Ols, AOIs, and EOIs will be used concuffently with this SSI as plant conditions warrant. In the case where direction conflicts between this SSI and another plant procedure, this SSI will take precedence.

31

NRC Scenario 6 Simulator Event Guide:

Event 7 MAJOR: SSI Fire Driver When directed by NRC or if the crew scrams the reactor insert mrf fpO 1 alarm for the fire alarm.

Commence Radio traffic between the Diver and the shift manager Report a fire reported at RHRSW Pump Rooms Insert Preference Key F8 start of RHRSW Pump B3 Two minutes after initial report, report as the Incident Commander that there was a fuel oil truck accident at the RHRSW Pump Rooms, R1{RSW Pump Room B is fully involved and the fire appears to be spreading to RHRSW Pump Room D Insert Preference Key F9 start of RHRSW Pump Al Has incident Commander request off site Fire fighting support.

Insert Preference Key FlO start of RHRSW Pump Cl Insert Preference Key Fil and F12 after RIIRSW Pump Cl start Two minutes later inform crew Fire is still active but silencing fire alarm Driver When Reactor is scrammed insert preference key shift F3 to scram Unit 1 and 3 and Shift F4 for bat file NRC/1404-25-1, after batch file 1404-25-1 is started go to remote function_summary (RFS)_and_start the three_fire pumps.

SRO TBD-2

[1] DIRECT Unit 3 Unit Supervisor to PERFORM Section 3.0 of 0-SSI-.25-l to Scram Unit 3, AND PROCEED TO cold shutdown.

[2] DIRECT Unit 1 Unit Supervisor to PERFORM Section 4.0 of 0-SSI-25-1 to Scram Unit_1,_AND PROCEED_TO cold_shutdown.

32

NRC Scenario 6 Simulator Event Guide:

Event 7 MAJOR: SSI Fire Crew NOTE The following instruments are those which have been credited for safe shutdown, and must be referenced when executing actions for this fire zone:

2-LI-3-58A and 2-PI-3-74A for reactor level and pressure.

2-XR-64-50 CH 2 (PT-64-50) for drywell pressure.

2-XR-64-50 CH 1 (TE-64-52C) for drywell temperature 2-FI-74-50, RHR SYS I Flow 2-FI-23-42, RHR HX 2C RHRSW Flow TBD-81 2-LI-64-66 and 2-TI-64-161 for suppression pool level and temperature. (0 Mm)

(0 Mm)

[3] DIRECT Unit 2 Operator to perform the following:

TBD-3 TBD- 1

[3.1] VERIFY reactor Scram AND RECORD current time (SSI time of entry).

Time SRO TBD-13

[4] DIRECT all operators to perform the following:

(90Mm)

( A. Operator 1 Section 1.0 of Attachment 1 (Places Fire Pump B local controls to EMERG at 4KV Shutdown Board)

(90 MinIl2O Mm)

B. Operator 2 Section 1.0 of Attachment 2 (Places Fire Pumps A & C local controls to_EMERG_at 4KV_Shutdown Boards,_trips_Unit 2_RPT_breakers)

SRO NOTE Diesel Generator loading should be closely monitored and maintained within the limits of 0-01-82. Prompt action to secure non-Appendix R designated loads should be taken to prevent overloading a Diesel Generator/4KV Shutdown Board prior to an RHR pump start.

Some 480V non-essential loads may require load shedding to keep Diesel Generator AJ4KV Shutdown Board A within load limits.

(0 Mm)

[5] DIRECT Unit 2 Operator to perform the following:

TBD-4

[5.1] VERIFY Main Steam Isolation Valves closed.

[5.2] VERIFY Reactor Recirculation Pumps tripped.

ATC/BOP Close MSIV and Trip Reactor Recirc Pumps 33

NRC Scenario 6 Simulator Event Guide:

Event 7 MAJOR: SSI Fire SRO (0mm)

[6] DIRECT Unit 2 Operator to verify the following pumps are NOT operating, at Panel 2-9-3:

• RHRSW PUMP B2, 0-HS-23-19A12

• RHRSW PUMP D2, 0-HS-23-27A/2 ATC/BOP Verifies RHRSW Pumps B2 and D2 are NOT operating SRO CAUTION TBD-87 To prevent loss of the HPCI systems, reactor pressure should NOT be allowed to decrease below 150 psig until low pressure system injection is available.

NOTES

1) FIPCI system injection and relief valve operation will be used to maintain reactor water level and a controlled 100°F per hour cooldown rate until the proper alignments for LPCI injection can be performed.

2) The HPCI system flow controller should be left in the AUTO position if functioning properly.

3) Credited relief valves for this unit as listed on Illustration 1 are, 2-PCV-0O 1-0005, 2-PCV-00 1-0022, 2-PCV-00 1-0030, and 2-PCV-00 1-0034.

D (10 Mm)

CT-2

[7] DIRECT Unit 2 Operator to initiate a controlled 100°F per hour cooldown rate using HPCI_(pressure_control_mode)_and relief valves.

ATCIBOP Places HPCI in pressure control within 10 minutes of the scram and commences a CT 2 controlled cooldown not to exceed 100°F per hour cooldown rate. Report HPCI controller failed in automatic but is controlling in manual.

SRO TBD-101 CTI (10 Mm)

[8] DIRECT Unit 2 Operator to place PATH A VENT FLOW CONT, 2-FIC-84-20, in MANUAL_and_0_SCFM,_at Panel 2-9-55.

CTi ATC/BOP Places 2-FIC-84-20 in Manual and 0 SCFM within 10 minutes 34

NRC Scenario 6 Simulator Event Guide:

Event 7 MAJOR: SSI Fire ATC/BOP Places HPCI in pressure control within 10 minutes of the scram and commences a CT 2 controlled cooldown not to exceed 100°F per hour cooldown rate. Report BPCI controller failed_in_automatic_but_is_controlling_in_manual.

8.4 Placing HPCI in CST to CST Operation

[5] PLACE SGTS in operation. REFER TO 0-01-65.

[6] ESTABLISH communication with the AUO locally in the HPCI room.

[7] DEPRESS and HOLD HPCI AUX OIL PUMP, 2-HS-73-47B START, push button_(local)_for_approximately_2_minutes_to_prime_the_oil_system.

[8] IF an initiation signal occurs while HPCI is operating in CST-to-CST Recirc Mode, THEN:

[9] PLACE HPCI AUXILIARY OIL PUMP, 2-HS-73-47A, in START.

[10] START HPCI STEAM PACKING EXHAUSTER using 2-HS-73-1OA.

[11] OPEN FIPCI/RCIC CST TEST VLV, 2-FCV-73-36.

[12] OPEN HPCI PUMP CST TEST VLV, 2-FCV-73-35.

[13] OPEN HPCI PUMP MIN FLOW VLV 2-FCV-73-30.

( CAUTION HPCI Turbine overspeed can occur if 2-FCV-73 -16 fails to fully open and is given a second open signal with the Aux oil pump running and the Turbine Control Valve in the full open/demand position.

[14] OPEN HPCI TURBINE STEAM SUPPLY VLV, 2-FCV-73-16.

[15] VERIFY the following automatic actions occur:

A. HPCI TURBiNE STOP VALVE, 2-FCV-73-18, opens.

B. HPCI TURBINE CONTROL VALVE, 2-FCV-73-l9, opens.

C. HPCI TURBINE SPEED, 2-SI-73-5 1, rises.

D. HPCI STEAM LINE 1NBD DRAiN VLV, 2-FCV-73-6A, and HPCI STEAM LINE OUTBD DRAIN VLV, 2-FCV-73-6B, close.

E. HPCI HOTWELL PUMP INBD ISOL VLV, 2-FCV-73-17A, and HPCI HOTWELL PUMP OUTBD ISOL VLV, 2-FCV-73-17B, close.

F. HPCI PUMP M1N FLOW VALVE, 2-FCV-73-30, closes as flow rises above 1255 gpm.

35

NRC Scenario 6 Simulator Event Guide:

Event 7 MAJOR: SSI Fire ATC/BOP Places HPCI in pressure control within 10 minutes of the scram and commences a CT 2 controlled cooldown not to exceed 100°F per hour cooldown rate. Report HPCI controller failed in automatic but is controlling in manual.

8.4 Placing FIPCI in CST to CST Operation NOTE If system is operating for RPV pressure control, RPV pressure can be varied by changing either HPCI pump discharge pressure or flow rate.

[16) WHILE maintaining less than 5300 gpmHPCI flow, THROTTLE HPCI PUMP CST TEST VLV, 2-FCV-73-35, to establish:

• HPCI SYSTEM FLOW/CONTROL indication, 2-FIC-73-33, as high as achievable but less than 5300 gpm.

• HPCI MAiN PUMP PRESS, 2-PI-73-31A, 60 to 100 psi above RPV pressure.

[17] VERIFY HPCI Auxiliary Oil pump stops and Shaft Driven Oil pump is operating properly.

[17.11 WHEN HPCI Auxiliary Oil Pump stops, THEN PLACE HPCI AUXILIARY OIL PUMP, 2-HS-73-47A, in AUTO.

[18] REFER TO Section 6.0 to control and monitor HPCI Turbine operation.

36

NRC Scenario 6 Simulator Event Guide:

Event 7 MAJOR: SSI Fire SRO TBD-89 (10 Mm)

[9] DIRECT Unit 2 Operator to place the following switches in TEST/INHIBIT, at Panel 2-9-3:

A. ECCS SYS I HI DW PRESS TEST/INHIBIT, 2-HS-75-59.

CT -3 B. ECCS SYS II HI DW PRESS TEST/INHIBIT, 2-HS-75-60.

CT - 3 ATCIBOP Place 2-HS-75-59 and 2-HS-75-60 in Test/Inhibit within 10 minutes SRO CAUTIONS

1) RHRSW flow may be diverted to EECW by 0-FCV-067-0049 which may also be potentially affected by the fire and has the power removed at 480V Diesel Aux Board by Operator 3 during performance of Attachment 1.

2) To keep from overloading the Diesel Generator, the RHR pump should be started prior to starting the RHRSW pump.

3) RHRSW Pumps Al and Cl are potentially affected by the fire and could spuriously start, if so the pumps can be secured from the Main Control Room to prevent dead heading. The Al pump is credited on Unit 3 for decay heat removal, when required, and can be restarted from MCR.

(10Mm)

IF Cl RHRSW Pump has spuriously started, THEN STOP Cl RI{RSW PUMP,

[10] using 0-HS-23-8A/2.

ATC/BOP Stops Cl RHRSW Pump after it had spuriously started SRO (10Mm)

[11] IF B3 RHRSW Pump has spuriously started, THEN STOP B3 RHRSW PUMP, using 0-HS-23-88A/2.

(10 Mm)

[12] IF Al RHRSW Pump has spuriously started, THEN STOP Al RHRSW PUMP, using 0-HS-23-1A/2.

ATCIBOP Stops RHRSW Pumps B3 and Al after they spuriously started TBD-76 (10 mm)

[13] DIRECT Unit 2 Operator to verify R}{RSW PUMP Cl SPLY TO EECW FCV-67-49, 0-HS-67-49/A2, is closed.

SRO

[14] DIRECT Unit 2 Operator to verify the following pumps are NOT operating at Panel 2-9-3:

• RHRSW PUMP C2, 0-HS-23-12AJ2.

• RHRSW PUMP Al, 0-HS-23-1A/2.

37

NRC Scenario 6 Simulator Event Guide:

Event 7 MAJOR: SSI Fire i)river After RHRSW B3 and Al are stopped call control room and report Attachment I Section 1 complete. Insert Preference Key Shift F5 Bat NRC/1404-25-la SRO NOTE When suppression pooi cooling is required, then RHR Pump 2C and RHRSW Pump C2 are the credited pumps for this unit. RHRSW Pump C2 can be operated from the Control Room as needed.

TBD-l 06 (90 miii)

[151 WHEN RHRSW Pump C2 is required to be started for suppression pool cooling, ThEN DIRECT Unit 2 Operator to start the pump using RHRSW PUMP C2, handswitch 0-HS-23-( 12A/2),

[16] VERIFY the following complete:

A. Attachment 1 Section 1.0 (Placed Fire Pump B local controls to EMERG at 4KV Shutdown Board)

B. Attachment 2 Section 1.0 (Placed Fire Pumps A & C local controls to EMERG_at 4KV_Shutdown Boards,_tripped Unit 2_RPT breakers)

.\ Driver When contacted report Attachment 2 Section 1 complete.

( SRO [17] NOTIFY site Operations personnel of the following:

A. Safe Shutdown procedure 0-SSI-25-1 is in effect.

TBD-10 B. Spurious equipment operation may occur.

[18] NOTIFY Security Shift Supervisor of the following

[18.1] When implementing safeguard measures NOT to lock down doors.

[18.2] DISPATCH Security personnel to ensure the following Electrical Board Room doors to Unit 1(2,3) Rx Buildings are unlocked to allow Operator ingress/egress:

• Door 634 (El. 621, Electrical Board Rni 1A)

• Door 648 (El. 621, Electrical Board Rm 2A)

• Door 657 (El. 621, Electrical Board Rm 3A)

• Door 538 (El. 593, Electrical Board Rm 1B)

• Door 540 (El. 593, Electrical Board Rm 2B)

• Door 513 (El. 593, Electrical Board Rni 3B).

38

NRC Scenario 6 Simulator Event Guide:

Event 7 MAJOR: SSI Fire SRO TBD-18 (30 Mm)

[19] DIRECT Unit 2 Operator to close 2-FCV-069-0002, using RWCU OUTBD SUCT ISOLATION VALVE, 2-HS-69-2A, at Panel 2-9-4.

ATC/BOP Close 2-FCV-069-2 TBD-37 (60 Mm)

[20] DIRECT Operator ito perform Section 2.0 of Attachment ito reset Battery Chargers i, 2A, and 3.

NOTE Steps 2.0[2 1] and 2.0[22] may be performed concurrently.

TBD-105 (90 Mm)

[21] DIRECT Operator Ito perform Section 3.0 of Attachment ito de-energize and verify closed 0-FCV-067-0049.

TBD-30 TBD-31 TBD-32 TBD-33 (240 Mm)

[22] DIRECT Operator 2 to perform Section 2.0 of Attachment 2 to align and start Unit and 2 ventilation systems and electric board room AHUs.

TBD-82 (540 Mm)

[23] DIRECT Unit 2 Operator to verify closed 2-FCV-001-0055 using MN STM LiNE DRAIN INBD ISOLATION VLV, 2-HS-i-55A, on Panel 2-9-3.

TBD-100

[24] IF drywell temperature approaches 281°F, THEN INITiATE shutdown cooling.

When directed to perform Section 2 of attachment I wait 4 minutes and insert Preference Key Shift Fl 0, Shift Fl 1 and Shift Fl 2 and report Attachment 1 section 2 complete Driver When directed to perform section 3 of attachment 1 wait 5 minutes and insert Preference Key Shift F8 and Shift F9 and report Attachment 1 section 3 complete When directed to perform section 2 of attachment 2 acknowledge direction 39

NRC Scenario 6 SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:

None OperationslMaintenance for the Shift:

1.3% power. 2-G0I-lOO-1A Section 5.4 Step 63.3 and 65.

Warm RFPT 2B lAW 2-01-3 Section 5.6 at step [3.21].

Continue to pull rods in accordance with RCP.

Units 1 and 3 are 100% Power Unusual ConditionsiProblem Areas:

Severe Thunderstorm Warnings are in affect for the entire area for the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

40

NRC Scenario 5 acility: Browns Ferry NPP Scenario No.: NRC 5 Op-Test No.: 1404 Examiners: Operators: SRO:_____

ATC:___

BOP:

Initial Conditions: 100% power, DG D and RBCCW 2B Pump are tagged out. Spare RBCCW Pump is aligned for operation.

Turnover: Return LPRM 8-49B to Operate from a Bypassed Condition JAW 2-OI-92B. Lower Power with flow to 90% for Main Turbine Valve Testing.

Event Maif. No. Event Type* Event Description No.

N-BOP 1 Return LPRM 8-49B to Operate JAW 2-OI-92B N-SRO R-ATC 2 Commence power decrease with flow to 90%

R-SRO C-BOP 3 edl8a Loss of I&C Bus A TS-SRO R-ATC 4 adOic TS-SRO ADS SRV 1-22 leaking C-BOP C-ATC VFD Cooling Water Pump 2A trips with failure of the standby 5 thl8a C-SRO pump to auto start C-ATC LOCA Recirculation Pump 2A Inboard and Outboard seal 6 R-ATC thl 0/11 a failure TS-SRO Two Level instruments fail high tripping Feedwater and FIPCI /

7 Batch File M-ALL LOCA / ED on Reactor Level 8 edl0a C Loss of 480V SD Board 2A RHR and Core Spray Division 2 Injection Valves will not Auto 9 Batch I open 10 rcO8 C RCIC Steam Valve fails to Auto open

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Lk 1

NRC Scenario 5 Critical Tasks Three CT 1 -With RPV pressure below the Shutoff Head of the available Low Pressure system(s),

operate available Low Pressure system(s) to restore RPV water level above T.A.F. (-162 inches).

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance.

Pressure below low pressure ECCS system(s) shutoff head.

3. Measured by:

Operator manually starts initiates at least one low pressure ECCS system and injects into the RPV to restore water level above -162 inches.

4. Feedback:

Reactor water level trend.

Reactor pressure trend.

CT 2 -With an injection system(s) operating and the reactor shutdown and at pressure, after RPV water level drops to -162 inches, transition to Emergency Depressurization before RPV level lowers to -180 inches.

1. Safety Significance:

Maintain adequate core cooling, prevent degradation of fission product barrier.

2. Cues:

Procedural compliance.

Water level trend.

3. Measured by:

Observation SRO directing actions in accordance with 2-C-2 Emergency Depressurization

4. Feedback:

RPV pressure trend.

SRV status indications.

2

NRC Scenario 5 Critical Tasks Three CT 3 -To prevent an uncontrolled RPV depressurization when Reactor level cannot be restored and maintained above -162 inches, inhibit ADS.

1. Safety Significance:

Maintain adequate core cooling, prevent degradation of fission product barrier.

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation.

4. Feedback:

RPV pressure trend.

RPV level trend.

ADS ADS LOGIC BUS AJB INHIBITED annunciator status.

3

NRC Scenario 5 Events

1. BOP operator will return LPRM 8-.49B to Operate lAW 2-OI-92B.

2. ATC lowers power to 90% using recirculation flow.

3. The crew will respond to a momentary loss of I&C Bus A. The in-service SJAE (A) will isolate and numerous alarms will come in. The BOP operator will shift SJAEs to B or reset SJAE A and return to service lAW 2-01-66 or 2-AOI-47-3. Reactor Zone Differential pressure low will alarm and the operator will have to reset Refuel and Reactor Zone fans. When one of the SJAEs are restored high H2 may result in Off Gas, the SRO may evaluate TRM 3.7.2 and enter Condition A.

4. During I&C Bus A loss, Main Steam Relief Valve open will alarm. When power is restored to I&C Bus A the alarm will clear but ADS SRV 1-22 will be leaking by and the acoustic monitor will indicate the leak by. SRO should enter 2-AOl-i-i, the ATC will lower power to less than 90%. When power is below 90% the BOP operator will perform 2-AOl-i-i actions to attempt to close the SRV. SRO will refer to Tech Specs and determine TS 3.5.1 condition F is applicable.

5. The VFD Cooling Water Pump for the A Reactor Recirc VFD will trip and the standby pump will fail to start. The ATC will start the standby VFD Cooling Water Pump to restore cooling water preventing a VFD and Reactor Recirc Pump trip.

6. #1 and #2 recirc pump seal failure ATC will note alarm and report #2 seal carrying full pressure. A short time later seal #2 will fail ATC will note that a small LOCA exists. ATC will trip and isolate A RR Pump lAW with 2-AOI-68-1A. ATC will insert control rods to exit Region 2 of the power to flow map. SRO will determine Technical Specification 3.4.1 Condition A, is applicable again with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to establish single loop conditions.

7. Level instruments 208A and 208D will fail high, causing a high level trip of Main Turbine, RFPTs and HPCI. RCIC will be the only major source of high pressure injection and the steam supply valve to RCIC will fail to auto open. The Crew will maintain reactor level until the LOCA is beyond the ability of RCIC to control. The SRO will determine ED is required in order to restore level with available low pressure systems.

8. After the scram 480V Shutdown Board 2A will fail due to a lockout, this will prevent operation of Core Spray Division 1 System for injection. RHR Loop 1 may be used for injection but no throttle capability will exist. RHR Loop 1 will not be available for Containment cooling operation. However, if RHR loop 1 was previous in suppression pool cooling due to the leaking SRV it will remain in suppression pool cooling.

9. With Division 2 Accident logic bypassed RHR and Core Spray will not auto start on any accident signals. The crew will have to manually start pumps and open injection valves. RHR Loop 2 will be available for Containment Cooling functions until required for injection.

4

NRC Scenario 5 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:

Emergency Depressurization complete Reactor Level is restored SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 5 10 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 2 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No) 5

NRC Scenario 5 Scenario Tasks TASK NUMBER IL iQ $JQ Restore an LPRM from Bypass RO U-92B-NO-05 215005A4.04 3.2 3.2 Lower Power with Recire Flow RO U-068-NO-03 SRO S-000-AD-31 2.1.23 4.3 4.4 Loss of I&C Bus A RO U-57C-AB-03 26200 1A2.04 3.8 4.2 SRO S-57C-AB-03 ADS SRV leaking RO U-001-AB-O1 239002A2.03 4.1 4.2 SRO S-0O1-AB-01 VFD Cooling Water Pump Failure RO U-068-AL-19 202001A2.22 3.1 3.2 SRO S-068-AB-01 RR Pump Seal Failure RO U-068-AL09 203000A4.02 4.1 4.1 SRO 5-068-AB-Ol Loss of 480V SD BD 2A RO U-57B-AL-06 22600 1A4.05 3.3 3.3 SRO S-57B-NO-07 LOCA/Low Level ED RO U-003-AL-24 29503 1EA2.04 4.6 4.8 RO U-000-EM-01 RO U-000-EM-13 SRO S-000-EM-14 SRO S-000-EM-1 5 SRO 5-000-EM-Ol 6

NRC Scenario 5 Procedures Used/Referenced:

Procedure Number [ Procedure Title 2-0I-92B Average Power Range Monitoring 2-GOI-lOO-12 Power Maneuvering 2-01-68 Reactor Recirculation System 2-A0I-57-5A Loss of I&C Bus A 2-ARP-9-8C Panel 9-8 2-XA-55-8C 2-ARP-9-7A Panel 9-7 2-XA-55-7A 2-ARP-9-6C Panel 9-6 2-XA-55-6C 2-ARP-9-7C Panel 9-7 2-XA-55-7C 2-ARP-9-3C Panel 9-3 2-XA-55-3C 2-ARP-9-3D Panel 9-3 2-XA-55-3D 2-ARP-9-53 Panel 9-53 2-XA-55-53 2-E0I-3 Secondary Containment Control 2-A0I-64-2D Group 6 Ventilation System Isolation

. Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 2-E0I Appendix-8F Isolation 2-01-66 0ff-Gas System 2-A0I-66-1 0ff-Gas H2 High Technical Specifications Technical Requirements Manual 2-AOl-i-i Relief Valve Stuck open 2-01-74 Residual Heat Removal System 2-E0I-2 Primary Containment Control 2-E0I Appendix-i 8 Suppression Pool Water Inventory Removal and Makeup 2-ARP-9-4A Panel 9-4 2-XA-55-4A 2-A0I 1 A Recirc Pump Trip/Core Flow Decrease OPRMs Operable 2-A0I-100-1 Reactor Scram 2-E0l-1 RPV Control 2-E0I Appendix-8B Reopening MSIVs / Bypass Valve Operation 2-E0I- i-C-i Alternate Level Control 2-EOI Appendix-6A Injection Subsystems Lineup Condensate 2-EOI Appendix-i 7C RHR System Operation Suppression Chamber Sprays 2-EOI Appendix-i7B RHR System Operation Drywell Sprays 2-E0I-3-C-2 Emergency RPV Depressurization 2-EOI Appendix-5B Injection System Lineup CRD 2-EOI Appendix-7B Alternate RPV Injection System Lineup SLC System 2-EOI Appendix-6B Injection Subsystems Lineup RHR System I LPCI Mode 2-EOI Appendix-6C Injection Subsystems Lineup RFIR System II LPCI Mode 2-EOI Appendix-6E Injection Subsystems Lineup Core Spray System II EP1P-1 Emergency Classification 7

NRC Scenario 5 Console Operator Instructions A. Scenario File Summary Batch File NRC/l3O6nrc-5 ior zloOil2l ld2Ob[1] off ior zloOil2l ld2Ob[2j off ior zloOhs2l lOd2Oa[1] off Tag DG D ior zlo0hs2 11 Od2Oa[21 off ior zloOhs2l 10d20a[2] off mrfdg0ld open ior zdihs708a null ior zlohs7o8a[1] off Tag RBCCW 2B ior zlohs708a[2] off ior zlohs708a[3] off ior zlohs682a2a[l] on br zlohs682a2a[2J off A VFD Cooling Pump Trip mrfthl8b trip trg 1 NRC/avfd trg 1 bat NRC/130605-1 A VFD Cooling Pump Trip imfth30f (e5 0)100 imfth30h (eS 60) 100 45 55 Level 8 instrument failures imfrc08 RCIC steam supply valve failure imfthl0a(e3 0)100 imfthl la (e3 60) 100 90 0 RR 2A Pump seal mrf cs09b inhibit mrfrhl 5 inhibit Div 2 accident logic bypassed ior z1oi17556a off ior zloil74lS4a off mrf edi 3 open momentary loss of I&C Bus A Batch File NRC/l3O6nrc-5-1 imfth2l (none 330) .6 600 .1 LOCA imfedl0a (none 370) Loss of 480V SD BD 2A 8

NRC Scenario 5 Preference File NRC/l3O6nrc-5 pfk0l tog pfk 02 ann silence pfk 03 mrf swO2 align align spare RBCCW Pump pflc 04 bat NRC/l3O6nrc-5 pflc 05 imfedl8a Loss of I&C Bus A pfk 06 ior zdihs682ala[1] off VFD A Cooling Pump trip pfk 07 imfadOic 10 ADS SRV Leak by pfko8trg! e3 RR Pump A Seal Failure pflc 09 trg! e5 Loss of Feedwater pfk 10 bat NRC/l3O6nrc-5-1 LOCA and Loss of 480V SD BD 2A pflc 11 mrfad0lc out pflc 12 ior zdixs0 122 Back up control panel switch to emergency pfksl pfk s2 pfk s3 pflc s4 ior zdihs0l22c open Back up control panel srv 1-22 pfk s5 ior zdihs0l22c close Back up control panel srv 1-22 pfk s6 bat appl8rhra pflc s7 bat appl8rhrb pfk s8 mrfedl3 close Scenario 5 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 28 manual Bypass LPRM 8-49B restorepref NRC/i 3O6nrc-5 mrf swO2 align RBCCW wait one F3 minute and turn off RBCCW Pump 2B Simulator Setup Load Batch F4 bat NRC/l3O6nrc-5 Simulator Setup manual Tag DG D and RBCCW Pump 2B Simulator Setup Verify file loaded, Clear alarms for Reactor_Recirc RCP required (100% -90% with flow) and RCP for Urgent Load Reduction 9

NRC Scenario 5 Simulator Event Guide:

Event 1 Normal: Return LPRM 8-49B to Operate from a Bypassed Condition JAW 2-OI-92B Driver At NRC direction call the control room as Reactor Engineer and request LPRM 08-49B be returned to service SRO Directs BOP to return LPRM 8-49B to Operate JAW 2-OI-92B BOP Return LPRM 8-49B to Operate lAW 2-OI-92B 6.4 Returning an LPRM to Operate From a Bypassed Condition

[1] REVIEW all precautions and limitations. REFER TO Section 3.0.

[2] REFERENCE Illustration 4 to find the APRM/LPRM Channel associated with the desired LPRM to be returned to normal.

[3] At Panel 2-9-14, DEPRESS any softkey to illuminate the display on the desired APRM/LPRM channel chassis. (Chassis 2)

[4] DEPRESS the ETC softkey until BYPASS SELECTIONS illuminates on the bottom row of the display.

[5] DEPRESS BYPASS SELECTIONS softkey, enter the password, and DEPRESS ENT. (Password 1-2-3-4)

[6] SELECT the desired LPRM to be returned to service by using the left or right arrows on the softkey board until the inverse video illuminates the correct LPRM.

[7] DEPRESS the OPERATE softkey.

[8] CHECK the BYP/HV OFF is replaced by OPERATE below the selected LPRM.

[9] DEPRESS EXIT softkey to return display to the desired bargraph.

[10] VERIFY, as a result of returning this LPRM to operate, that any alarms received on_Panel_2-9-5_or_on the_APRM/LPRM_channel_are_reset.

10

NRC Scenario 5 Simulator Event Guide:

Event 2 Reactivity: Power decrease with Recirc Flow SRO Notifies BA of power decrease.

Directs Power decrease using Recirc Flow, JAW 2-GOI-100-12.

[1] REVIEW all Precautions and Limitations listed in Section 3.0.

[2] VERIFY Prerequisite listed in Section 4.0 is satisfied.

[3] NOTIFY the Balancing Authority System Operators (BA) (5-751-4134) of impending power reduction.

[4] NOTIFY Radiation Protection of purpose for power reduction, the target power level (see above note), and RECORD time Radiation Protection notified in NOMS Narrative Log.

[6] IF power is being reduced (less than 10%) for any of the following reasons:

• Weekly Control Rod Exercise

• Main Turbine Valve Testing

• Ultimate heat Sink temperature> 92.5°F

[6.1] REDUCE Recirculation flow. REFER TO 2-01-68.

[6.2] MAINTAIN Reactor thennal power within the limits shown on ICS and 0-TI-248, Station Reactor Engineer, as appropriate.

[10] PERFORM the following while reducing Reactor power:

[10.1] WHEN Reactor power is at approximately 90%, THEN REFER TO 2-01-3 and START a RFP Injection Water Pump.

ATC Lower Power wlRecirc, lAW 2-01-68, Section 6.2 Driver When directed by NRC, insert preference key F5 imf cdl 8a Loss of I&C Bus A, followed by F7 imfadOic 10 and after 5 seconds Shift F8 mrfedl3 close NRC Two additional power decreases in Scenario, can continue when ready. will have to mismatch speeds at 1300 rpm.

11

NRC Scenario 5 Simulator Event Guide:

Event 2 Reactivity: Power decrease with Recire Flow ATC Lower Power w/Recirc, lAW 2-01-68, Section 6.2 D. Individual pump speeds should be mismatched by -60 RPM during dual pump operation between 1200 and 1300 RPM to minimize harmonic vibration (this requirement may be waived for short periods for testing or maintenance).

[1] IF desired to control Recire Pumps 2A and/or 2B speed with Recirc Individual Control, THEN PERFORM the following:

• RAISE Recire Pump 2A using RAISE SLOW (MEDIIJM), 2-HS-96-15A(15B).

(Otherwise N/A)

• LOWER Recirc Pump 2A using SLOW(MEDIUM)(FAST),

2-HS 1 7A( 1 7B)( 1 7C). (Otherwise N/A).

AND/OR

• RAISE Recirc Pump 2B using RAISE SLOW (MEDIUM), 2-HS-96-16A(16B).

(Otherwise N/A)

• LOWER Recirc Pump 2B using SLOW(MEDIUM)(FAST),

2-HS 1 8A( 1 8B)( 1 8C). (Otherwise N/A).

[2] WHEN desired to control Recirc Pumps 2A and/or 2B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump Speed 2A & 2B using the following pushbuttons as required.

RAISE SLOW, 2-HS-96-3 1 RAISE MEDIUM, 2-HS-96-32 LOWER SLOW, 2-HS-96-33 LOWER MEDIUM, 2-HS-96-34 LOWER FAST, 2-HS-96-35 Driver When directed by NRC, insert preference key F5 imf ed 1 8a Loss of I&C Bus A, followed by F7 imfadOic 10 and after 5 seconds Shift F8 mrfedl3 close.

When dispatched wait two minutes and report Failure of 9-9 Throwover Switch, switch tripped to alternate.

NRC Two additional power decrease in Scenario, can continue when ready 12

NRC Scenario 5 Simulator Event Guide:

Event 3 Component: Loss of I&C Bus A Crew Respond to numerous alarms when I&C Bus A deenergizes. The most significant of these are 8C-21, 6C-12, 3C-25, 7C-22 and 3D-3, 19, and 32.

ATC Announces Power, Level, and Pressure are stable BOP Alarm 8C-21, I&C BUS A VOLTAGE ABNORMAL A. VERIFY alarm by checking the following:

• Loss of instrument power and remote position indication to Core Spray Div I and RHR Div I (Panel 9-3).

• RWCU Filter Demin A isolation.

• Reactor Building/Refuel Zone Ventilation isolation.

B. NOTIFY Unit 3 Unit Supervisor.

C. REFER TO 2-AOI-57-5A and O-GOI-300-2.

SRO Announce entry to 2-AOI-57-5A, Loss of I&C Bus A.

ATC Alarm 6C-12, RFPT GOVERNOR POWER FAILURE OR GOV ABNORMAL

-. A. VERIFY RFPT/RFPs continue to control Reactor Water Level.

(& B. IF a RFPT/RFP has tripped, THEN VERIFY other RFPTs in Automatic operation raise or lower output flow to maintain reactor water level.

C. DISPATCH personnel to UNIT 2 Auxiliary Instrument Room to PERFORM the following at Panels 2-9-48,49,50:

• CHECK Power Supply lights illuminated.

• CHECK display screens for Governor abnormal conditions.

Announced RPV Level stable, dispatches personnel BOP Alarm 3C-25, MAIN STEAM RELIEF VALVE OPEN A. CHECK MSRV DISCHARGE TAILPIPE TEMPERATURE, 2-TR-l-1, on Panel 2-9-47 and SRV Tailpipe Flow Monitor on Panel 2-9-3 for raised temperature and flow indications.

B. REFERTO2-AOI-1-l.

BOP Announce Main Steam Relief Valve Open alarm cleared, but have indication on acoustic monitor of SRV partially open or leaking by. ADS SRV 1-22 3C-25 alarms on a loss of I&C Bus A, when the bis re-energizes ADS SRV will show NOTE acoustic monitoring indication of leaking by. BOP operator should report to SRO and SRO enter 2-AOl-i-i. These events will occur under event four.

13

NRC Scenario 5 Simulator Event Guide:

Event 3 Component: Loss of I&C Bus A Crew Respond to numerous alarms when I&C Bus A deenergizes. The most significant of these are 8C-21, 6C-12, 3C-25, 7C-22 and 3D-3, 19, and 32.

BOP Alarm 7C-22, DRYWELL/SUPPR CHAMBER H202 ANALYZER FAILURE A. CHECK Panel 2-9-54 and 2-9-55 for abnormal indicating lights such as low flow, H2 or 02 downscale, pump off, etc.

B. IF sample pump is NOT running, THEN ATTEMPT to start pump using 2-HS-76-1 l0/S5.

C. IF sample pump will NOT start OR H2/02 analyzer malfunction, THEN PLACE H2/02 Analyzer in Service per 2-01-76 section 5.4.

D. REFER TO TRM 3.3.11_and TRM Section 3.6.2.

BOP Resets 112/02 ANALYZER ISOLATION RESET, 2-HS-76-91 Resets_Alarm_on_2-MON 110,_touch_screen.

BOP Alarm 3D-19, DRYWELL LEAK DETECTION RADIATION DNSC A. DETERMINE cause of alarm by performing the following:

1. CHECK AIR PARTICULATE MONITOR CONTROLLER, 2-MON-90-50 on Panel 2-9-2 for condition bringing in alarm

2. DISPATCH personnel to determine which alarm is annunciating using the HELP button (REFER TO 2-01-90 for complete annunciator list).

E. REFER TO Tech Specs 3.4.4, 3.4.5, and TRM 3.3.10 for CAM LCO requirements and iMPLEMENT appropriate TS/TRM actions as required.

F. WHEN conditions permit, THEN RESET alarm per 2-01-90, Section 6.5.

BOP Determines DW Radiation Monitor Cam isolated, resets the following to restore to operation.

UPPER 1NBD SUPPLY ISOL VALVE RESET, 2-HS-90-254A-A LOWER 1NBD SUPPLY ISOL VALVE RESET, 2-HS-90-254B-A OUTBD RETURN ISOL VALVE RESET, 2-HS-90-257A-A OUTBD SUPPLY ISOL VALVE RESET, 2-HS-90-255A INBD RETURN ISOL VALVE RESET, 2-HS-90-257B-A 14

NRC Scenario 5 Simulator Event Guide:

Event 3 Component: Loss of I&C Bus A Crew Respond to numerous alarms when I&C Bus A deenergizes. The most significant of these are 8C-21, 6C-12, 3C-25, 7C-22 and 3D-3, 19, and 32.

BOP Alarms 3D-3, RX BLDG VENTILATION ABNORMAL A. IF PCIS group 6 isolation exists, ThEN REFER TO 2-AOI-64-2d.

B. NOTIFY Unit Supervisors, Unit 1 and Unit 3.

C. VERIFY standby fans start.

D. DISPATCH personnel to check Bldg AP (PDIC 64-2, El 639, Rx Bldg.)

E. IF AP is at or above -0.17 in. H20 THEN ENTER 2-EOI-3 Flowchart, 2-XA-55-3D, window 32.

BOP Alarms 3D-32, REACTOR ZONE DIFFERENTIAL PRESSURE LOW D. IF alarm is valid, THEN INFORM Unit Supervisor of 2-EOI-3 entry condition.

E. REQUEST personnel to check fans locally for any apparent problems.

( F. REFER TO 2-OI-30B and PLACE standby fan in service to restore normal differential pressure.

SRO Enters 2-EOI-3, Secondary Containment Control and 2-AOI-64-2D, Group 6 Ventilation System Isolation Directs Reactor and Refuel Zone Ventilation returned to service by either 2-EOI Appendix-8F, Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 Isolation or 2-AOI-64-2D The above procedures for restoring ventilation are basically the same will describe NOTE Appendix-SF below. The only action in EOI-3 is to restore ventilation.

15

NRC Scenario 5 Simulator Event Guide:

Event 3 Component: Loss of I&C Bus A ATCIBOP Appendix 8F Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 Isolation VERIFY PCIS Reset.

2. PLACE Refuel Zone Ventilation in service as follows (Panel 2-9-25):

a. VERIFY 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch is in OFF.

b. PLACE 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch to SLOW A (SLOW B).

c. CHECK two SPLYIEXH A(B) green lights above 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch extinguish and two SPLY/EXH A(B) red lights illuminate.

d. VERIFY OPEN the following dampers:

• 2-FCO-64-5, REFUEL ZONE SPLY OUTBD ISOL DMPR

• 2-FCO-64-6, REFUEL ZONE SPLY INBD ISOL DMPR

• 2-FCO-64-9, REFUEL ZONE EXH OUTBD ISOL DMPR

• 2-FCO-64-lO, REFUEL ZONE EXH INBD ISOL DMPR.

3. PLACE Reactor Zone Ventilation in service as follows (Panel 2-9-2 5):

a. VERIFY 2-HS-64-l 1A, REACTOR ZONE FANS AND DAMPERS, control switch is in OFF.

b. PLACE 2-HS-64-l 1A, REACTOR ZONE FANS AND DAMPERS, control switch in SLOW A (SLOW B).

c. CHECK two SPLY/EXH A(B) green lights above 2-HS-64-1 1A, REACTOR ZONE FANS AND DAMPERS, control switch extinguish and two SPLY/EXH A(B) red lights illuminate.

d. VERIFY OPEN the following dampers:

• 2-FCO-64-13, REACTOR ZONE SPLY OUTBD ISOL DMPR

• 2-FCO-64-14, REACTOR ZONE SPLY 1NBD ISOL DMPR

• 2-FCO-64-42, REACTOR ZONE EXH INBD ISOL DMPR

• 2-FCO-64-43, REACTOR ZONE EXH OUTBD ISOL DMPR.

5. IF Reactor Zone Fan fast speed is desired following 5 minutes of slow speed operation, THEN PLACE 2-HS-64-l IA, REACTOR ZONE FANS AND DAMPERS, control switch in FAST A (FAST B).

6. IF Refuel Zone Fan fast speed is desired following 5 minutes of slow speed operation, THEN PLACE 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch in FAST A (FAST B).

h 16

NRC Scenario 5 Simulator Event Guide:

Event 3 Component: Loss of I&C Bus A SRO Enters 2-AOI-57-5A 4.2 Subsequent Actions

[1] VERIFY Automatic Actions have occurred.

[2] IF a Reactor Scram occurs, THEN PERFORM 2-AOl- 100-1 concurrently with this procedure.

[3] VERIFY a flow path for Condensate System, or STOP the condensate pumps/booster pumps. REFER TO 2-01-2.

[4] START Standby Gas Train(s) and ChECK Reactor Building pressure at or below 0.25 H20 vacuum (PDIC 64-1, Panel 25-215; PDIC 64-2, Panel 25-213). REFER TO 0-01-65, Section Standby Gas Treatment System Manual Initiation.

[5] VERIFY SJAE B in service to maintain condenser vacuum. REFER TO 2-01-66.

[6] IF Auto Transfer of Panel 2-9-9, Cabinet 2, failed THEN (otherwise N/A)

N [7] WHEN Reactor water level is normal, THEN RESET PCIS Group 6 inboard isolation and RETURN the affected systems to service or standby readiness.

REFER TO 2-A0I-100-1, if a Reactor Scram occurred, otherwise REFER TO 2-AOI-64-2D.

SRO Directs restoration of Reactor Building DP, should restore Ventilation JAW Appendix-8F or 2-A0I-64-.2D. May call Unit 1 to start Standby Gas Fans SRO Directs restoration of SJAE, lAW 2-01-66 hard card BOP Restores SJAE to service, Standby SJAE System Lineup Hard Card 17

NRC Scenario 5 (47 Simulator Event Guide:

Event 3 Component: Loss of I&C Bus A BOP Restores SJAE to service, Standby SJAE System Lineup Hard Card

[1] VERIFY RESET Off-Gas isolation using 2-HS-90-155, OG OUTLET/DRAIN ISOLATION VLVS.

NOTE With power back to I&C Bus A, once RO resets 2-HS-90-155, can place SJAE A back in service or can transfer to SJAE B. All steps are listed below for either.

[2] VERIFY OPEN the following valves:

• 2-HS-66-l 1(15), SJAE 2A(2B) INLET VALVE.

• 2-HS-1.-155A(156A), STEAM TO SJAE 2A(2B).

[3] VERIFY in AUTO/OPEN 2-HS-66-14(18), SJAE 2A(2B) OG OUTLET VALVE.

[4] PLACE 2-HS-l-150(152), SJAE 2A(2B) PRESS CONTROLLER, in CLOSE and then in OPEN.

[5] VERIFY OPEN the following valves (red light illuminated):

• 2-PCV-l-151/166 (153/167), STEAM TO SJAE 2A(2B) STAGES 1,2, AND 3.

• 2-FCV-1-150(152), SJAE 2A(2B) 1NTMD CONDENSER DRAIN.

[6] MONITOR hotwell pressure as indicated on recorder 2-XR-2-2, HOTWELL TEMP AND PRESS, on Panel 2-9-6.

) [7] FOR the SJAE not being placed in service, VERIFY CLOSED the following valves:

• 2-HS-66-18(14), SJAE 2B(2A) OG OUTLET VALVE.

• 2-HS-1-152(150), SJAE 2B(2A) PRESSURE CONTROLLER.

• 2-HS-1-156A(155A) STEAM TO SJAE 2B(2A)

NRC Depending on the time to restore SJAEs High H2 may occur in OG BOP Acknowledge Panel 2-9-53 Alarms, Report high hydrogen levels 53-3 and 13, HIGH OFFGAS % 112 TRAIN A, and HIGH OFFGAS % H2 TRAIN BB 18

NRC Scenario 5 Simulator Event Guide:

Event 3 Component: Loss of I&C Bus A BOP Report high hydrogen levels 53-3 and 13, HIGH OFFGAS % H2 TRAIN A, and HIGH OFFGAS % H2 TRAIN BB A. CHECK Off-gas Hydrogen Analyzer, 2-H2R-66-96 (CH 1) on Panel 2-9-53 to verify H2 concentration.

B. IF alarm is valid, THEN REFER TO 2-AOI-66-1.

SRO Enters 2-AOI-66-1, Off-Gas H2 High BOP/ATC [1] PLACE both OFFGAS TRAIN A(B) AUTO CHANNEL CHECK /

BYPASS control switches, 2-HS-066-1007 and 1008, on OFFGAS SAMPLE PANEL, 2-LPNL-925-0588, in BYPASS to assure continuous availability of hydrogen monitoring.

[2] IF HWC System injection is in service, THEN (otherwise N/A)

[3] VERIFY proper operation of in service SJAE.

[4] IF hydrogen concentration is greater than or equal to 4%, THEN REFER TO TRM 3.7.2.

[10] MONITOR the following parameters at Control Room Panel 9-53 and 9-8:

t J

• RECOMBINER 2A/2B TEMPERATURE, 2-TRS-66-77, for abnormal trend; either rising or lowering.

OFF GAS HYDROGEN ANALYZER, 2-H2R-66-96, for hydrogen concentration.

NOTE H2 concentration may rise to 8 to 12% and return to a normal value of less than 1%

Restores PSC Pump Suction Inboard Isolation Valve, 2-FCV-75-57, by -re-opening the 2-BOP FCV-75-57 19

NRC Scenario 5 Simulator Event Guide:

Event 3 Component: Loss of I&C Bus A SRO - Directs restoration of FPC system lAW 2-01-78 NRC Note- FPC demin isolated on loss of I&C A, with demin bypass open BOP/ATC 5.2 Placing Filter-Demineralizer in Service

[6] At Control Room Panel 2-9-4, PERFORM the following:

[6.1] VERIFY OPEN FILTER DEMIN 11JBD ISOL VALVE, 2-FCV-78-63, using handswitch 2-HS-78-63A.

[6.2] VERIFY OPEN FILTER DEM1N OUTBOARD ISOL VALVE, 2-FCV 64, using handswitch_2-HS-78-64A CAUTION Rapidly opening FUEL POOL FID B(D) EFFLUENT VLV, 2(0)-FCV-78-26, before closing FILTER DEMIN BYPASS VALVE A(B), 2-FCV-78-66(65), may cause flow transients that could damage filter-demineralizer precoat.

[10] PLACE filter-demineralizer in service as follows:

[10.1] ESTABLISH communications between Main Control Room and 0-LPNL-925-0017.

[10.2] VERIFY in OPEN at 0-LPNL-925-OO1OA, FUEL POOL F/D B(D)

BOP/ATC EFFLUENT VLV, 2(0)-HS-78-26B.

[10.3] SIMULTANEOUSLY CLOSE FILTER DEMIN BYPASS VALVE A(B), using handswitch 2-HS-78-66B(65B), on JB 4108, EL. 621, RI 1-S Line, AND RAISE flow using FPC VESSEL B(D) EFFL FLOW recorder controller, 2(0)-FRC-078-0024, in MANUAL on 0-LPNL-925-0017.

CAUTION Maximum flow is 600 gpm for each filter-demineralizer.

[10.41 RAISE flow through demineralizer until desired flow (approximately 500 gpm) has been established, using FPC VESSEL B(D) EFFL FLOW recorder controller, 2(0)-FRC-078-0024, in MANUAL on 0-LPNL-925-00 17.

[10.5] IF it is desired to place flow controller in automatic, THEN:

[10.5.1] BALANCE FPC VESSEL B(D) EFFL FLOW recorder controller, 2(0)-FRC-078-0024, manual signal with automatic process signal till zero BOP/ATC deviation exists on vertical meter on 0-LPNL-925-0017.

[10.5.2] PLACE FPC VESSEL (B)D EFFL FLOW recorder controller, 2(O)-FRC-078-0024, in AUTO on 0-LPNL-925-0017

[14] RECORD applicable data on Demin Operating Log.

[15] [NER/C] NOTIFY Chemistry that filter-demineralizer has been placed in service, AN]) REQUEST a sample of effluent. [INPO SOER 82-0 13]

Below is event 4 which started with the failure of I&C Bus A, depending on SRO priorities NOTE may have addressed SRV first and then restoration from Bus loss.

20

NRC Scenario 5

ç. Simulator Event Guide:

Event 4 Component: ADS SRV 1-22 leaking Event 4 started with the failure of I&C Bus A, depending on SRO priorities may have NOTE addressed SRV first and then restoration from Bus loss.

SRO Enters 2-AOl-i-i BOP 4.1 Immediate Action

[1] IDENTIFY stuck open relief valve by OBSERVING the following:

• SRV TAILPIPE FLOW MONITOR, 2-FMT-1-4, on Panel 2-9-3, OR

• MSRV DISCHARGE TAILPIPE TEMPERATURE recorder, 2-TR-l-1 on Panel 2-9-47.

ATC [2] IF relief valve transient occurred while operating above 90% power, THEN REDUCE reactor power to 9O% RTP with recirc flow.

BOP [3] WHILE OBSERVING the indications for the affected Relief valve on the Acoustic Monitor; CYCLE the affected relief valve control switch several times as required:

• CLOSE to OPEN to CLOSE positions

[4] IF all SRVs are CLOSED, THEN CONTINUE at Step 4.2.4. (N/A) 4.2 Subsequent Action 4.2.2 Attempt to close valve from Panel 9-3:

[1] PLACE the SRV TAILPIPE FLOW MONITOR POWER SWITCH in the OFF position.

[2] PLACE the SRV TAILPIPE FLOW MONITOR POWER SWITCH in the ON position.

[3] IF all SRVs are CLOSED, THEN CONTINUE at Step 4.2.4. (N/A)

[4] PLACE MSRV AUTO ACTUATION LOGIC INHIBIT, 2-XS-1-202 in iNHIBIT:

[5] IF relief valve closes, THEN OPEN breaker or PULL fuses as necessary using Attachment 1 (Unit 2 SRV Solenoid Power Breaker/Fuse Table).

[6] PLACE MSRV AUTO ACTUATION LOGIC INHIBIT 2-XS-1-202, in AUTO.

[7] IF the SRV valve did not close, THEN PERFORM the appropriate section from table below.

RELIEF STEP Switch Breaker Fuse VALVE Number Location Location Location SRV 1-22 Step 4.2.3[2] Panel 25-32 Multiple Panel 25-32 21

NRC Scenario 5 Simulator Event Guide:

Event 4 Component: ADS SRV 1-22 leaking Actions for SRV 1-22, wait two minutes and for taking control at panel 25-32 Preference key Fl 2, for cycling SRV preference key shift F4 open, shift F5 close to 10, shift F4 open, shiftF5 closeto 10.

D river Contact control room and determine if valve closed. When told to remove power preference key Fli. When back to normal at panel 25-32 delete override for annunciator xa553e10.

When told to power back up srv 1-22 mrf adO Ic in.

Driver [2] IF 2-PCV-l-22 is NOT closed, THEN PERFORM the following:

[2.1] On Panel 2-25-3 2 PLACE the transfer switch associated MAIN STM LINE B RELIEF VALVE XFR, 2-XS-l-22 in EMERG position.

[2.2] IF the SRV does NOT close, THEN PERFORM the following while OBSERVING the indications for the 2-PCV-1-22 on the Acoustic Monitor:

CYCLE the MAIN STM LINE B RELiEF VALVE, 2-HS-1-22C to the following positions several times. CLOSE/AUTO to OPEN to CLOSE/AUTO

[2.3] IF the SRV does NOT close, THEN PERFORM the following:

A. VERIFY the MAIN STM LINE B RELIEF VALVE, 2-HS-l-22C, in the CLOSE/AUTO position.

B. PLACE the transfer switch associated MAIN STM LINE B RELIEF VALVE XFR, 2-XS-1-22 in NORM position.

Driver

[2.4] IF the SRV does NOT close, THEN REMOVE the power from 2-PCV-l-22 by performing one of the following:

A. OPEN the following breakers (Preferred method)

[2.5] IF the valve does NOT close, THEN CLOSE the breakers or REINSTALL fuses removed in Step 4.2.3 [2.4].

BOP [2.6] CONTINUE at Step 4.2.4.

22

NRC Scenario 5 Simulator Event Guide:

Event 4 Component: ADS SRV 1-22 leaking BOP [2.6] CONTINUE at Step 4.2.4.

4.2.4 Other Actions and Documentation

[1] NOTIFY Reactor Engineering of current conditions.

[2] IF ANY EOI entry condition is met, THEN ENTER the appropriate EOI(s).

[3] REFER TO Technical Specifications Sections 3.5.1 and 3.4.3 for Automatic Depressurization System and relief valve operability requirements.

[4] INITIATE suppression pool cooling as necessary to maintain suppression pool temperature less than 95°F.

[5] IF the relief valve can NOT be closed AND suppression pooi temperature CANNOT be maintained less than or equal to 95°F, THEN PLACE the reactor in Mode 4 in accordance with 2-GOI-100-12A.

[6] DOCUMENT actions taken and INITIATE Work Order (WO) for the valve.

SRO Directs Suppression Pool Cooling JAW 2-01-74 BOP Initiates Pool Cooling as directed SRO Refers to Tech Specs 3.5.1 ECCS Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

Condition E: One ADS valve inoperable.

Required Action E. 1: Restore ADS valve to OPERABLE status.

Completion Time: 14 days.

BOPIATC Inform SRO when Suppression Pool Level meets EOI-2 entry requirements SRO Enter EOI-2 on Suppression Pool Level, directs suppression pool cooling lAW EOl- APPX 1 7A NOTE One RHR Pump will almost maintain pooi temperature depending on reactor power, Do NOT expect pool temperature to exceed bulk 95°F.

23

NRC Scenario 5 Simulator Event Guide:

Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Cooling as directed 8.5 Initiation of Loop 1(11) Suppression Pool Cooling CAUTION PSA concerns with RHR in Suppression Pool Cooling Mode with a LOCA and a LOSP identify that severe water hammer may occur during the pump restart. Therefore, the following guidelines should be used to try and maintain the system below the PSA Risk Assessment goals:

• RHR in suppression pool cooling should be minimized.

• Two Loops of RHR in suppression pooi cooling should be minimized.

• Use two pumps per loop, if needed, to minimize total time spent in suppression pool cooling.

• Suppression pool cooling run times are tracked in 2-SR-2 to ensure risk assessment goals are not exceeded.

NOTES

1) Suppression Pool Cooling is required to be initiated whenever necessary to maintain suppression pool temperature less than 95°F or when directed by other procedures.

[1] VERIFY RHR Loop 1(11) is in Standby Readiness. REFER TO Section 4.0

[2] REVIEW the precautions and limitations in Section 3.0.

[3] NOTIFY other units of placing Loop 1(11) of RHR in suppression pool cooling, the subsequent start of common equipment (i.e., RHRSW pumps) and associated alarms are to be expected.

[4] NOTIFY Radiation Protection for impending action to initiate Suppression Pool Cooling. RECORD name and time of Radiation Protection representative notified in NOMS narrative log

[5] IF possible, THEN BEFORE placing RHRSW in service, NOTIFY Chemistry that RHRSW sampling is to be initiated (RHRSW sampling requirements).

[6] VERIFY at least one R}{RSW Pump is operating on each EECW Header.

NOTE One RHR Pump will almost maintain pool temperature constant depending on reactor power,_Do NOT expect pooi temperature to exceed 95°F.

24

NRC Scenario 5 Simulator Event Guide:

Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Cooling as directed

[7] PLACE RHR Pump and Heat Exchanger A(C) in service as follows:

[7.1] START an RHRSW Pump to supply RHR Heat Exchanger A(C).

[7.2] ESTABLISH RHRSW flow by performing one the following:

[7.2.1] REQUEST another unit establish minimum flow for Pump which will be utilized for Suppression Pool Cooling, (RHRSW Pump A(C) and establish minimum flow. (between 4000 and 4500 gpm RHRSW flow) REFER TO 0-01-23.

OR

[7.2.2] THROTTLE OPEN RHR HX 2A(2C) RHRSW OUTLET VLV, 2-FCV-23 -34(40), as required for cooling (if another is maintaining minimum flow) and/or to maintain between 4000 and 4500 gpm R}IRSW flow as indicated on 2-FI-23 -36(42), R}IR HTX 2A(2C)

RHRSW FLOW. [Z

[7.3] VERIFY CLOSED RHR SYS I LPCI INBD INJECT VALVE, 2-FCV-74-53.

• [7.4] IF NO RHR PUMP (1A OR 1C) is operating in Suppression Pool Cooling, k THEN VERIFY CLOSED RI-IR SYS I SUPPR POOL CLG/TEST VALVE, 2-FCV-74-59.

[7.5] VERIFY CLOSED RHR SYS I SUPPR CHBR SPRAY VALVE, 2-FCV-74-58.

[7.6] VERIFY CLOSED RHR SYS I DW SPRAY OUTBD VLV, 2-FCV-74-60.

[7.7] VERIFY OPEN RHR SYS I SUPPR CHBRIPOOL ISOL VLV, 2-FCV-74-57.

[7.8] IF desired to operate without the Drywell DP Compressor, THEN:

[7.9] START RFIR PUMP A(C) using 2-HS-74-5A(16A).

[7.10] THROTTLE RHR SYS I SUPPR POOL CLG/TEST VLV, 2-FCV-74-59, to maintain RHR flow within limits, as indicated on RHR SYS I CTMT SPRAY FLOW, 2-FI-74-56:

RHR Pumps in 1 2 Operation Loop Flow 7,000 to <13,000 gpm &

10,000 gpm & Blue Blue light light illuminated illuminated 25

NRC Scenario 5 Simulator Event Guide:

Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Cooling as directed

[7.11] IF desired to raise Suppression Pool Cooling flow and only one Loop I pump is in service, THEN PLACE the second Loop I RHR Pump and Heat Exchanger in service by REPERFORMING Step 8.5 [7] for the second pump.

[8] CHECK pump motor breaker charging spring recharged for all 4160 Volt pump motors operated in this section, as follows:

• Amber breaker spring charged light on,

• Closing spring target indicates charged.

[10] PLACE RHR Pump and Heat Exchanger B(D) in service as follows:

[10.2] ESTABLISH RFIRSW flow by one of the following methods:

[10.2.1] REQUEST another unit establish minimum flow for Pump which will be utilized for Suppression Pool Cooling, and establish minimum flow. (between 4000 and 4500 gpmRHRSW flow)

REFER TO 0-01-23.

OR

[10.2.2] THROTTLE OPEN RHR HX 2B(2D) RHRSW OUTLET VLV, 2-FCV-23-46(52), as required for cooling (if another is maintaining minimum flow) and/or to maintain between 4000 and 4500 gpm RHRSW flow as indicated on 2-FI-23-48(54), RHR HX 2B(2D)

RHR.SW FLOW.

[10.3] VERIFY CLOSED RHR SYS II LPCI 1NBD INJECT VALVE, 2-.FCV-74-67.

[10.4] IF NO RHR PUMP (lB or ID) is operating in Suppression Pool Cooling, THEN VERIFY CLOSED RHR SYS II SUPPR POOL CLG/TEST VLV, 2-FCV-74-73.

[10.5] VERIFY CLOSED RHR SYS II SUPPR CHER SPRAY VALVE, 2-FCV-74-72.

[10.6] VERIFY CLOSED RHR SYS II DW SPRAY OUTBD VLV, 2-FCV-74-74.

[10.7] VERIFY OPEN RHR SYS II SUPPR CFIBR/POOL ISOL VLV, 2-FCV-74-7 1.

26

NRC Scenario 5 Simulator Event Guide:

Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Cooling as directed

[10.8] IF desired to operate without the Diywell DP Compressor, THEN:

[10.9] START RHR PUMP 2B(2D) using 2-HS-74-28A(39A).

[10.10] THROTTLE RHR SYS II SUPPR POOL CLG/TEST VLV, 2-FCV-74-73, to maintain RHR flow within limits, as indicated on RHR SYS H CTMT FLOW, 2-FI-74-70.

RHRPumpsin 1 2 Operation Loop Flow 7,000 to <13,000 gprn &

10,000 gpm & Blue Blue light light illuminated illuminated

[11] IF desired to RAISE Suppression Pool Cooling flow and only one Loop II pump is in service, THEN PLACE the second Loop II R}IR Pump AND Heat Exchanger in service. REPERFORM Step 8.5 [10] for the second pump.

[12] CHECK pump motor breaker charging spring recharged for all 4160 Volt pump

, motors operated in this section, as follows:

Amber breaker spring charged light on, Closing_spring target indicates_charged.

SRO Tech Spec 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be -6.25 inches with and -7.25 inches without differential pressure control and -1.0 inches.

APPLICABILITY: MODES 1,2, and 3.

Condition A: Suppression pool water level not within limits.

Required Action A. 1: Restore suppression pool water level to within limits.

Completion Time: 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Note AS the SRV remains open adding inventory to suppression pooi, pool level spec will be appropriate.

27

NRC Scenario 5 Simulator Event Guide:

Event 4 Component: ADS SRV 1-22 leaking SRO Enter EOI-2 on Suppression Pool Level SRO PC/H Verify H202 analyzer in service (APP 19)

When H2 is detected in PC (2.4% on control room indicators continue, does not continue SPIT MOMTOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX I 7A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, does not continue PC/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI-64-1),

PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, does not continue DW/T Monitor and control Drywell temperature below 1 60F using available Drywell cooling Can Drywell Temperature be maintained below 160F, YES SPIL MOMTOR and CONTROL suppr p1 lvi between-i in. and -6 in. (APPX 18)

Can suppr p1 lvl be maintained above -6 in., YES Can suppr p1 lvl be maintained below -1 in., YES SRO Direct Appendix 18, Suppression Pool Water Inventory Removal And Makeup BOP Calls for Operator to perform field action of Appendix 18 28

NRC Scenario 5 Simulator Event Guide:

Event 4 Component: ADS SRV 1-22 leaking BOP Calls for Operator to perform field action of Appendix 18

3. IF Directed by SRO, THEN REMOVE water from Suppression Pool as follows:

a. DISPATCH personnel to perform the following (Unit 2 RB, El 519 ft, Torus Area):

1) VERIFY OPEN 2-SHV-074-0786A(B), RHR DR PUMP 2A(2B)

DISCH TO MN CNDR/RW SOy.

2) OPEN the following valves:

• 2-SHV-074-0564A(B), RHR DR PMP 2A(2B) SEAL WATER SUPPLY SOV

• 2-SHV-074-0529A(B), RHR DRAIN PUMP A(B) SHUTOFF VLV.

3) UNLOCK and OPEN 2-SHV-074-0765A(B), RHR DR PUMP 2A(2B)

DISCH SOy.

4) NOTIFY Unit Operator that RUE. Drain Pump 2A(2B) is lined up to remove water from Suppression Pool.

5) REMAIN at torus area UNTIL Unit 2 Operator directs starting of R1{R Drain Pump 2A(2B).

b. IF Main Condenser is desired drain path, THEN OPEN 2-FCV-74-62, RUE.

MAIN CNDR FLUSH VALVE.

c. IF Radwaste is desired drain path, THEN PERFORM the following:

1) ESTABLISH communications with Radwaste.

2) OPEN 2-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE

d. NOTIFY personnel in Unit 2 RB, El 519 ft, Torus Area to start RHR Drain Pump 2A(2B).

e. THROTTLE 2-FCV-74-108, RUE. DR PUMP 2AJB D1SCH HDR VALVE, as necessary.

Driver When dispatched to remove water from the suppression pool, wait 10 minutes and call and report aligned step 4 above, when the operator calls you back to start the RHR Drain Pump Shift F6 for bat app 1 8rhra and Shift F7 for bat appi 8rhrb 29

NRC Scenario 5 Simulator Event Guide:

Event 4 Component: ADS SRV 1-22 leaking BOP Appendix 18

4. WHEN Suppression Pool level reaches -5.5 in., THEN SECURE RHR Drain System as follows:

a. DISPATCH personnel to STOP the Drain System as follows (Unit 2 RB, El 519 ft, Torus Area):

1) STOP RHR Drain Pump 2A(2B).

2) CLOSE the following valves:

• 2-SHV-074-0564A(B), R}IR DR PMP 2A(2B) SEAL WATER SUPPLY SOV

• 2-SHV-074-0529A(B), RHR DRAIN PUMP A(B) SHUTOFF VLV.

3) CLOSE and LOCK 2-SHV-074-0765A(B), RHR DR PUMP 2A(2B)

DISCH SOy.

b. CLOSE 2-FCV-74-108, RHR DR PUMP 2A/B DISCH HDR VALVE.

C. VERIFY CLOSED 2-FCV-74-62, R}IR MAIN CNDR FLUSH VALVE.

d. VERIFY CLOSED 2-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE.

Driver When directed by NRC for VFD Cooling Pump trip, Preference Key F6 30

NRC Scenario 5 Simulator Event Guide:

Event 5 Component: VFD Cooling Water Pump 2A trip Driver When directed by NRC for VFD Cooling Pump trip, Preference Key F6 NRC NOTE: VFD will trip if a hi Transformer (or Cell) temperature condition develops in conjunction with both cooling pumps off.

ATC Respond to the following alarms, 4A-12, 4A-28 and 4A-32 ATC Report Trip of Recirc Drive 2A Cooling Pump 2A1, and failure of standby pump to start Alarm 4A-12, RECIRC DRIVE 2A COOLANT FLOW LOW Automatic Action Standby RECII{C DRIVE cooling water pump will auto start.

A. VERIFY RECIRC DRIVE cooling water pump running.

B. DISPATCH personnel to the RECIRC DRIVE to check the operation of the Recirc Drive cooling water system.

Alarm 4A-28, RECIRC DRIVE 2A PROCESS ALARM A. IF 2-XA-55-4B Window 28 is also in alarm, THEN (N/A)

B. Refer to ICS screen VFDAAL and determine cause of alarm Alarm 4A-32, RECIRC DRIVE 2A DRIVE ALARM A. REFER TO ICS Group Display GD @VFDADA and DETERMINE cause of alarm.

B. IF a problem with the cooling water system is indicated, THEN VERIFY proper operation of cooling water system.

Start Standby Recirc Drive 2A Cooling Pump 2A2, dispatches personnel to investigate ATC Wait 4 minutes after dispatched, THEN report tripped VFD Pump 2A1 is hot to the

. touch, internal bkr closed, 480 volt bkr tripped (480 V SD BD 2A 5C).

Driver When directed by NRC initiate RR Pump 2A Seal Failure Preference Key F8 31

NRC Scenario 5 Simulator Event Guide:

Event 6 Component: LOCA Recirculation Pump 2A Inboard and Outboard seal failure Driver When directed by NRC initiate RR Pump 2A Seal Failure Preference Key F8 ATC Respond to alarm 4A-25, RECIRC PUMP A NO. 1 SEAL LEAKAGE ABN A. DETERMINE initiating cause by comparing No. 1 and 2 seal cavity pressure indicators on Panel 2-9-4 or ICS.

• Plugging of No. 1 RO No. 2 seal cavity pressure indicator drops toward zero.

• Plugging of No. 2 RO No. 2 seal pressure approaches no. 1 seal pressure.

• Failure of No. 1 seal No. 2 seal pressure is greater than 50% of the pressure of No. 1.

• Failure of No. 2 seal no. 2 seal pressure is less than 50% of the No. 1 seal.

B. RECORD_pump_seal parameters hourly on Attachment_1, ATC Report of failure of number 1 seal or inner seal Respond to alarm 4A-18, RECIRC PUMP A NO.2 SEAL LEAKAGE HIGH A. COMPARE No. 2 cavity pressure indicator (2-PI-68-63A) to No. 1 cavity pressure indicator (2-PI-68-64A). No. 2 seal degradation is indicated if the pressure at No. 2 seal is less than 50% of the pressure at No. 1 seal.

ATC Reports the second seal is failed both pressure indicators trending toward zero psig.

C. IF dual seal failure is indicated, THEN

1. SHUTDOWN Recirc Pump 2A by depressing RECIRC DRIVE 2A SHUTDOWN, 2-HS 19.

2. VERIFY TRIPPED, RECIRC DRIVE 2A NORMAL FEEDER, 2-HS-57-17.

3. VERIFY TRIPPED, RECIRC DRIVE 2A ALTERNATE FEEDER, 2-HS-57-15.

4. CLOSE Recirculation Pump 2A suction valve.

5. CLOSE Recirculation Pump 2A discharge valve.

6. REFER TO 2-AOI-68-1A or 2-AOI-68-1B AND 2-01-68.

7. DISPATCH personnel to_secure Recirculation Pump_2A_seal water.

ATC Trips RR Pump 2A and closes suction and discharge valves Reports rising Drywell Pressure, reports DW Pressure stable once valves are closed SRO Enters 2-A0I-68-1A, Recirc Pump Trip/Core Flow Decrease OPRMs Operable 32

NRC Scenario 5 Simulator Event Guide:

Event 6 Component: LOCA Recirculation Pump 2A Inboard and Outboard seal failure SRO Enters 2-AOI 1A, Recirc Pump Trip/Core Flow Decrease OPRMs Operable 4.2 Subsequent Actions

[1] IF both Recirc Pumps are tripped in modes 1 or 2, ThEN (N/A),

[21 IF a single Recirc Pump tripped, THEN CLOSE tripped Recirc Pump discharge valve.

[3] IF Region I or II of the Power to Flow Map is entered, THEN IMMEDIATELY take actions to INSERT control rods to less than 95.2% loadline. Refer to 0-TI-464, Reactivity Control Plan Development and Implementation.

[4] RAISE core flow to greater than 45%. REFER TO 2-01-68.

[5] INSERT control rods to exit regions if not already exited. Refer to 0-TI-464, Reactivity Control Plan Development and Implementation.

[6] MAINTAIN operating Recirc pump flow less than 46,600 gpm. Refer to 2-01-6 8.

[7] WHEN plant conditions allow, THEN, MAINTAIN operating jet pump loop flow greater than 41_x_106_lbmlhr (2-FI-68-46_or_2-FI-68-48).

SRO Direct inserting control rods lAW Urgent Load Reduction and Rod Shove Sheets ATC Inserts Control Rods to Exit Region II of the Power to Flow Map Driver When dispatched to isolate seal water wait 5 minutes and then mrf rdO3 close and report closed 33

NRC Scenario 5 Simulator Event Guide:

Event 6 Component: LOCA Recirculation Pump 2Alnboard and Outboard seal failure ATC Inserts Control Rods to Exit Region II of the Power to Flow Map Inserts all of the following Control Rods to lower rod line to < 95%:

Control Rods 30-39, 38-31, 30-23,22-31 from 08 to 00 Control Rods 14-39, 46-39, 46-23, 14-23 from 18 to 00 Control Rods 22-47, 38-47, 38-15, 22-15 from 18 to 00 Control Rod 14-31,30-47,46-31,30-15 from 48 to 00 ATC Raise Speed of RR Pump B until core flow is 46 to 50% and ensure RR Pump B drive flow is below 46,600 gpm Report Exit from Region II of Power to Flow Map NRC Due to the seal failure the crew may choose to vent the drywell due to rising DW press SRO Directs venting the drywell lAW 2-01-64 BOP/ATC Vents DW JAW 2-01-64 6.1 Venting the Drywell with Standby Gas Treatment Fan 6.1.1 Venting Lineup

[1] REVIEW all Precautions and Limitations. REFER TO Section 3.0.

[2] VERIFY all Prestartup/Standby Readiness requirements in Section 4.0 are satisfied.

.-*. [3] VERIFY Stack Dilution Fans in operation. REFER TO 2-01-66.

) [4] CHECK Group 6 Isolation Signal (Ventilation Systems) NOT present.

[5] VERIFY CLOSED 2-FCV-64-29 using DRYWELL VENT INBD ISOL VALVE, 2-HS-64-29.

[6] VERIFY CLOSED, PATH B VENT FLOW CONT, 2-FIC-84-19, (Panel 2-9-55).

[7] VERIFY PATH A VENT FLOW CONT, 2-FIC-84-20, in AUTO and set at 100 SCFM (Panel 2-9-55).

[8] IF the Drywell DP Compressor is in operation, THEN STOP the compressor using DRYWELL DP COMP AND VALVES CONTROL, 2-HS-64-142A. (Otherwise N/A)

[9] VERIFY CLOSED, DW DP COMP SUCTION ISOL VLV, 2-FCV-64-139, using 2-ZI-64-139.

[10] NOTIFY Unit 1 and 3 Control Room that Unit 2 Drywell venting with SGT is about to start.

[11] PLACE a Standby Gas Treatment Train in service as follows: (N/A this section if SBGT is already in service.)

[11.1] CONTACT Unit 1 to determine which SBGT Train to be started.

[11.2] START the desired Standby Gas Treatment Train using the appropriate hand-switch listed below:

A. On Panel 1-9-25

• SGTS TRAIN A FAN, 0-HS-65-18A/1

• SGTS TRAIN B FAN, 0-HS-65-40A/1 B. On Panel 2-9-25

• SGTS TRAIN C FAN, 0-HS-65-69A12.

[11.3] RECORD in the Narrative Logs the SBGT Train started and the Appropriate Filter DP.

34

NRC Scenario 5 Simulator Event Guide:

Event 6 Component: LOCA Recirculation Pump 2A Inboard and Outboard seal failure Driver If called as unit 1 to start SBGT Train Fan A or B-start appropriate train BOP/ATC 6.1.2 Venting Drywell

[1] VERIFY Section 6. 1.1 has been completed. (N/A if venting has not been completely secured.)

[2] RECORD initial data for Drywell Venting in 2-SI-4.7.A.2.a if required.

[3] CLOSE, 2-FCV-64-34, using SUPPR CHBR INBD ISOLATION VLV, 2-HS-64-34.

[4] VERIFY OPEN, 2-FCV-64-3 1, using DRYWELL INBD ISOLATION VLV, 2-HS 31.

[5] OPEN 2-FCV-84-20 using 2-FCV-84-20 CONTROL DW/SUPPR CHBR VENT, 2-HS-64-3 5.

[6] VERIFY flow at approximately 100 SCFM on PATH A VENT FLOW CONT 2-FIC-84-20 (Panel 2-9-55).

[7] MONITOR Drywell pressure.

[8] WHEN the desired pressure OR the lower DP limit is reached as indicated by DRYWELL DP CPRSR DISCH VALVE, 2-HS-64-140 opening, THEN CONTINUE in this procedure.

[9] CLOSE 2-FCV-84-20 by using 2-FCV-84-20 CONTROL DW/SUPPR CHBR VENT, 2-HS-64-3 5.

[10] RECORD Final Data for Drywell Venting in 2-SI-4.7.A.2.a if required.

[11] OPEN 2-FCV-64-34, using SUPPR CHBR INBD ISOLATION VLV, 2-HS-64-34.

[12] IF Step 6.1.1 [8] was performed, THEN VERIFY DRYWELL DP COMP AND

) VALVES CONTROL, 2-HS-64-142A, in AUTO. (Otherwise N/A)

[13] IF it is desired to continue venting the Drywell, THEN RE-PERFORM Section 6.1.2

[14] WHEN Venting is no longer required, ThEN CONTINUE with Section 6.1.3.

SRO Tech Spec 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation.

OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:

a. LCO 3.2.1, AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR), single loop operation limits specified in the COLR;

b. LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR),

single loop operation limits specified in the COLR;

c. LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation, Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation; APPLICABILITY: MODES 1 and 2.

Condition A: Requirements of the LCO not met.

Required Action A. 1: Satisfy the requirements of the LCO.

Completion Time: 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Driver When directed by NRC, Preference Key F9, Level Instruments Fail high When mode switch is out of run or NOT in run Preference Key Fl 0 35

NRC Scenario 5 Simulator Event Guide:

Event 7 Major: Loss of Feedwater and HPCI ATC Report Trip of Main Turbine and RFPTs and Reactor Scram ATC 4.1 Immediate Actions

[1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5A/S3A and 2-HS-99-5AJS3B, on Panel 2-9-5.

[2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds. (Otherwise N/A), Step is NA

[3] REFUEL MODE ONE ROD PERMISSIVE light check:

[3.1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in REFUEL.

[3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-85-46.

[3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-85-46, is not illuminated, ThEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (Otherwise N/A) Step is NA

[4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in SHUTDOWN.

[5] REPORT the following status to the US:

• Reactor Scram

• Mode Switch is in Shutdown

• All rods in or rods out

• Reactor Level and trend (recovering or lowering).

• Reactor pressure and trend

• MSIV position (Open or Closed)

• Power level

[1] ANNOUNCE Reactor SCRAM over PA system.

[3] DRIVE in all IRMs and SRMs from Panel 2-9-5 as time and conditions permit.

[3.1] DOWNRANGE IRMs as necessary to follow power as it lowers.

[5] MONITOR and CONTROL Reactor Water Level between +2 and +51 , or as directed by US, as follows:

ATC/BOP Open RCIC Steam Supply Valve to start RCIC for Level Control, RCIC has received an Auto Start signal but the Steam Supply Valve failed to Open.

Driver When mode switch is in out of run Preference Key FlO 36

NRC Scenario 5 Simulator Event Guide:

Event 7 Major: Loss of Feedwater and HPCI SRO Enters EOI-1 on RPV Water Level SRO EOI- 1 (Reactor Pressure)

Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? NO IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate? NO IF Emergency Depressurization is or has been required THEN exit RCIP and enter C2 Emergency Depressurization? NO -

IF RPV water level cannot be determined? NO -

Is any MSRV Cycling? - No IF Steam cooling is required? NO Suppression Pool level and temperature cannot be maintained in the safe area of Curve IF J 3?-NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? NO IF Drywell Control air becomes unavailable? NO-IF Boron injection is required? NO Stabilize RPV pressure below 1073 psig with the main turbine bypass valves (APPX 8B)

SRO Direct a pressure band, may direct a cooldown LAW Appendix 8B ATC/BOP Maintain directed pressure with Bypass Valves lAW Appendix 8B, Reopening MSIVs I Bypass Valve Operation 37

NRC Scenario 5 Simulator Event Guide:

Event 7 Major: Loss of Feedwater and HPCI ATCIBOP Maintain directed pressure with Bypass Valves lAW Appendix 8B, Reopening MSIVs I Bypass Valve Operation

1. IF pressure control with bypass valves is desired and MSIVs are open, THEN proceed to step 10.

10. Verify Condenser Vacuum is greater than 7

11. IF manual opening of Bypass Valves is desired, THEN perform the following step:

a. Depress the Bypass Valve Opening Jack Raise Pushbutton, 2-HS 13 OB to slowly open the Bypass Valves.

b. Adjust BPV Positio0n as necessary by using the raise, 2-HS-47-130B and Lower 2-HS-47-130A pushbuttons to maintain desired cooldown rate.

12. IF EHC Auto Cooldown is desired, THEN perform the following steps:

a. Verify EHC is in Pressure Control using 2-HS-47-204

b. Verify Bypass Valve Demand is set at ZERO

c. On the EHC Work Station on Panel 2-9-7:

1) Select Main Menu from the toolbar at bottom of the screen.

2) Select Log In on Display Screen and Enter OPS for name and OPS for password.

3) Select Auto Cooldown from list of function on the screen.

d. On the Auto Cooldown Display Screen

1) Check the following are displayed.

• Turbine Tripped or All Valves Closed indicates reset

• RX Press Ctrl indicates reset

2) Select the block above the FINAL PRESSURE TARGET

3) Enter the desired pressure using the display screen or keyboard

4) Select OK

5) Depress the START button

6) When Are You Sure You Want to Initiate Auto Cooldown? appears, Select YES

7) Check the following:

• EHC PRESSURE SETPOINT, 2-Pl-47-162, is lowering

• EHC AUTO COOLDOWN displays IN PROCESS 38

NRC Scenario 5 Simulator Event Guide:

Event 7 Major: Loss of Feedwater and HPCI SRO EOI- 1 (Reactor Level)

Monitor and Control Reactor Water Level.

Directs_Verification_of PCIS_isolations.

ATC/BOP Verifies PCIS isolations.

SRO IF It has NOT been determined that the reactor will remain subcritical without boron under all condition THEN EXIT RC/L NO -

RPV water level CANNOT be determined NO PC water level CANNOT be maintained below 105 feet NO -

Restore and Maintain RPV water level between +2 inches and +51 inches with RCIC (APPX 5C)

ATC/BOP RCIC failed to auto start, Opens RCIC Steam Supply Valve and verifies RCIC operation.

1. VERIFY 2-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller in AUTO with setpoint at 600 gpm.

7. OPEN 2-FCV-71-8, RCIC TURBiNE STEAM SUPPLY VLV, to start RCIC Turbine.

8. ChECK proper RCIC operation by observing the following:

a. RCIC Turbine speed accelerates above 2100 rpm.

b. RCIC flow to RPV stabilizes and is controlled automatically at 600 gpm.

c. 2-FCV-71-40, RCIC Testable Check Vlv, opens by observing 2-ZI-7 1 -40A, DISC POSITION, red light illuminated.

d. 2-FCV-71-34, RCIC PUMP MIN FLOW VALVE, closes as flow rises above 120 gpm.

39

NRC Scenario 5 Simulator Event Guide:

Event 8 Component: Loss 480V SD BD 2A and LOCA ATCIBOP Report rising Drywell Pressure and Temperature ATC/BOP Report loss of 480V SD BD 2A and 480V RMOV BD 2A ATC/BOP Dispatch personnel to investigate loss of Board SRO Re-Enter EOI-2 on High DW Pressure and Temperature ATCIBOP IF RHR Loop 1 was in Pool Cooling for leaking SRV, then operators report that RHR Loop 1 remains in Pool cooling.

NOTE RHR Loop 1 has lost power to almost all valves but NO valves reposition on board loss SRO EOI-2 on High Drywell Pressure DWIT Monitor and control Drywell temperature below 1 60F using available Drywell cooling Can Drywell Temperature be maintained below 1 60F, NO Operate all available drywell cooling

(

Before Drywell Temperature rises to 200F enter EOI- 1 and Scram Reactor, Completed Before Drywell Temperature rises to 280F continue Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recire Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B)

Driver When dispatched for Board loss, wait 4 minutes and report overcuffent trip of supply breaker on 480V SD BD 2A. If requested to energize 480V RMOV BD 2A from alternate supply, wait 3 minutes and report that unable to restore power to Board 40

NRC Scenario 5 Simulator Event Guide:

Event 8 Component: Loss 480V SD BD 2A and LOCA SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI 1),

PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig continue, Continues Initiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7C), Direct Appendix 17C When suppression chamber pressure exceeds 12 psig, Continues Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7B)

When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue SRO Enters EOI-2 on High Drywell Pressure PC/H VerifS H2O2 analyzer in service (APP 19)

When 112 is detected in PC (2.4% on control room indicators continue, does not continue 41

NRC Scenario 5 Simulator Event Guide:

Q Event 8 Component: Loss 480V SD BD 2A and LOCA SRO Enters EOI-2 on High Drywell Pressure SPIT MONITOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, does not continue Enters EOI-2 on High Drywell Pressure SPIL MONITOR and CONTROL suppr p1 lvi between -1 in. and -6 in. (APPX 18)

Can suppr p1 lvi be maintained above -6 in., YES Can suppr p1 lvi be maintained below -1 in., YES SRO Direct Suppression Chamber Sprays and Drywell Sprays on RHR Loop II ONLY 42

NRC Scenario 5 Simulator Event Guide:

Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP 2-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays

1. BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.

2. IF Adequate core cooling is assured, OR Directed to spray the Suppression Chamber irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary by PLACING 2-HS-74-155B, LPCI SYS II OUTBD NJ VLV BYPASS SEL in BYPASS.

3. IF Directed by SRO to spray the Suppression Chamber using Standby Coolant Supply, THEN CONTINUE in this procedure at Step 7.

4. IF Directed by SRO to spray the Suppression Chamber using Fire Protection, THEN CONTINUE in this procedure at Step 8.

5. INITIATE Suppression Chamber Sprays as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header.

b. IF EITHER of the following exists:

• LPCI Initiation signal is NOT present, OR

• Directed by SRO, THEN PLACE keylock switch 2-XS-74-130, RHR SYS II LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 2-XS-74.-129, RHR SYS II CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 2-FCV-74-67, RHR SYS II INBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-66, RHR SYS II OUTBD INJECT VALVE.

e. VERIFY OPERATING the desired RHR System II pump(s) for Suppression Chamber Spray.

f. VERIFY OPEN 2-FCV-74-71, RHR SYS II SUPPR CHBR!POOL ISOL VLV.

g. OPEN 2-FCV-74-72, RHR SYS II SUPPR CRBR SPRAY VALVE.

ATC/BOP Aligns RHR Loop II Pumps in Suppression Chamber Sprays 43

NRC Scenario 5 Simulator Event Guide:

Event 8 Component: Loss 480V SD BD 2A and LOCA ATCIBOP 2-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays

h. IF RHR System II is operating ONLY in Suppression Chamber Spray mode, THEN CONTINUE in this procedure at Step 5.k.

i. VERIFY CLOSED 2-FCV-74-30, RHR SYSTEM II M1N FLOW VALVE.

j. RAISE system flow by placing the second RHR System II pump in service as necessary.

k. MONITOR RHR Pump NPSH using Attachment 2.

1. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s).

m. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm flow:

• 2-FCV-23-46, RHR HX 3B RHR.SW OUTLET VLV

• 2-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV.

n. NOTIFY Chemistry that RHRSW is aligned to in-service RHR Heat Exchangers.

ATC/BOP Aligns RHR Loop II Pumps in Suppression Chamber Sprays 44

NRC Scenario 5 Simulator Event Guide:

Event 8 Component: Loss 480V SD BD 2A and LOCA

= ATC/BOP 2-EOI APPENDIX-17B, RHR System Operation Drywell Sprays

1. BEFORE Drywell pressure drops below 0 psig, CONTINUE in this procedure at Step 7.

2. IF Adequate core cooling is assured, OR Directed to spray the Drywell irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary by PLACING 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

3. VERIFY Recirc Pumps and Drywell Blowers shutdown.

4. IF Directed by SRO to spray the Drywell using Standby Coolant supply, THEN CONTINUE in this procedure at Step 8.

5. IF Directed by SRO to spray the Drywell using Fire Protection, THEN CONTINUE in this procedure at Step 9.

6. INITIATE Drywell Sprays as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header.

b. IF EITHER of the following exists:

• LPCI Initiation signal is NOT present, OR

• Directed by SRO, THEN PLACE keylock switch 2-XS-74l30, RHR SYS II LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 2-XS-74-129, RHR SYS II CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 2-FCV-74-67, RHR SYS II LPCI INBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE.

e. VERIFY OPERATING the desired System II RI{R pump(s) for Drywell Spray.

f. OPEN the following valves:

• 2-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV

• 2-FCV-74-75, RHR SYS II DW SPRAY 1NBD VLV.

ATC/BOP Aligns RHR Loop II Pumps in Drywell Sprays 45

NRC Scenario 5 Simulator Event Guide:

Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP 2-EOI APPENDIX-17B, RNR System Operation Drywell Sprays

g. VERIFY CLOSED 2.-FCV-74-30, RHR. SYSTEM II M1N FLOW VALVE.

h. IF Additional Drywell Spray flow is necessary, THEN PLACE the second System II RHR Pump in service.

i. MONITOR RHR Pump NPSH using Attachment 2.

j. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s).

k. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:

• 2-FCV-23-46, RHR HX 3B RHRSW OUTLET VLV

• 2-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV.

1. NOTIFY Chemistry that R}IRSW is aligned to in-service RHR Heat Exchangers.

7. WHEN EITHER of the following exists:

• Before drywell pressure drops below 0 psig, OR

• Directed by SRO to stop Drywell Sprays, THEN STOP Drywell Sprays as follows:

a. VERIFY CLOSED the following valves:

• 2-FCV-74-100, RI-JR SYS I U-2 DISCH XTIE

• 2-FCV-74-74, R}IR SYS II DW SPRAY OUTBD VLV

• 2-FCV-74-75, RI-JR SYS II DW SPRAY INBD VLV.

b. VERIFY OPEN 2-FCV-74-30, RHR SYSTEM II MIN FLOW VALVE.

c. IF RHR operation is desired in ANY other mode, THEN EXIT this EOI Appendix.

d. STOP RHR Pumps.

ATC/BOP Aligns RHR Loop II Pumps in Drywell Sprays 46

NRC Scenario 5 Simulator Event Guide:

Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP Report lowermg RPV water level unable to maintain with RCIC SRO EOI-1 Reactor Level RPV water level drops below -120 inches OR The ADS timer has initiated NO IF RPV water level CANNOT be restored and maintained between +2 and +51 inches, THEN Restore and maintain RPV water level above -162 inches Augment RPV water level control as necessary with any of the following SRO Directs additional level control systems:

SLC (boron tank) APPX-7B CR1) APPX-5B ATCIBOP Aligns CRD and SLC JAW Appendix 5B and 7B ATC CRD Appendix 5B

2. IF BOTH of the following exist:

. CRD is NOT required for rod insertion, AND N

• Maximum injection flow is required,

(

THEN LINE UP ALL available CR1) pumps to the RPV as follows:

a. IF CR1) Pump 2A is available, THEN VERIFY RUNNING CRD Pump 2A.

b. IF CRD Pump lB is available, THEN PERFORM the following:

1) NOTIFY Unit 1 Operator to verify closed 1-FCV-85-8, CRD PUMP B DISCHARGE VALVE (Unit 1, Panel 9-5).

2) START CR1) Pump lB.

3) OPEN 2-FCV-85-8, CR1) PUMP lB DISCH TO U2.

c. OPEN the following valves to increase CR1) flow to the RPV:

. 2-PCV-85-23, CRD DRIVE WATER PRESS CONTROL VLV

. 2-PCV-85-27, CRD CLG WATER PRESS CONTROL VLV

. 2-FCV-85-50, CRD EXH RTN LINE SHUTOFF VALVE.

d. ADJUST 2-FIC-85-1 1, CR1) SYSTEM FLOW CONTROL, on Panel 9-5 to control injection WHILE maintaining 2-PI-85-13A, CRD ACCUM CHG WTR H1)R PRESS,_above_1450_psig,_if possible.

e. IF Additional flow is necessary to prevent or mitigate core damage, THEN DISPATCH personnel to fully open the following valves as required:

• 2-THV-085-0527, PUMP DISCH THROTTLING (RB NE, el 565) 2-BYV-085-0551, PUMP TEST BYPASS (RB NE,_el 565).

Driver When called as unit one operator FCV-85-8, CR1) PUMP B DISCHARGE VALVE is closed 47

NRC Scenario 5 Simulator Event Guide:

Event 8 Component: Loss 480V SD BD 2A and LOCA ATCIBOP Aligns CRD and SLC JAW Appendix 5B and 7B ATC SLC Appendix 7B

2. IF RPV injection is needed immediately ONLY to prevent or mitigate fuel damage, THEN CONTINUE at Step_10 to inject SLC Boron Tank to RPV.

10. UNLOCK and PLACE 2-HS-63-6A, SLC PUMP 2A12B, control switch in START-A or START-B (Panel 9-5).

11. CHECK SLC injection by observing the following:

• Selected pump starts, as indicated by red light illuminated above pump control switch.

. Squib valves fire, as indicated by SQUIB VALVE A and B CONTiNUITY blue lights extinguished,

• SLC SQUIB VALVE CONTINUITY LOST Annunciator in alarm (2-XA-5 5-SB, Window 20).

• 2-PI-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.

• System flow, as indicated by 2-IL-63-1 1, SLC FLOW, red light illuminated, 2 . SLC iNJECTION FLOW TO REACTOR Annunciator in alarm (2-XA 5B, Window 14).

12. IF Proper system operation CANNOT be verified, THEN RETURN to Step 10 and START other SLC pump.

48

NRC Scenario 5 Simulator Event Guide:

Event 8 Component: Loss 480V SD BD 2A and LOCA SRO EOI- 1 Reactor Level Can RPV water level be restored and maintained above 162 inches NO SRO Announces_entry to_EOI-C- 1_Alternate_Level_Control RPV water level CANNOT be determined NO PC water level CANNOT be maintained below 105 feet NO -

IF RPV water level can be restored and maintained above 162 inches NO CT-3 Inhibit ADS CT-3 ATC/BOP Inhibits ADS SRO Restore and maintain RPV water level above -162 inches using any of the following:

Condensate APPX 6A LPCI System I APPX 6B LPCI System II APPX 6C Core Spray System II APPX 6E SRO Directs 2 or_more_of the above_systems_lined_up_for_injection ATC/BOP Aligns the directed systems for Injection 49

NRC Scenario 5 Simulator Event Guide:

Event 8 Component: Loss 480V SD BD 2A and LOCA SRO EOI-C-1 Alternate Level Control SRO Can 2 or more Condensate, LPCI or Core Spray injection subsystems be lined up YES -

When RPV Water level drops to -162 inches Proceeds at TAF or -162 inches Is any Condensate, LPCI or Core Spray injection subsystems lined up for injection with at least one pump running YES-Is any RPV injection source lined up with at least one pump running YES BEFORE RPV water level drops to -180 inches CONTiNUE Continues-Emergency Depressurization is required Inject into the RPV with any available sources CT-2 SRO Enters EOI-C-2 Emergency Depressurization Will the reactor remain subcritical without boron under all conditions YES Is DW pressure above 2.4 psig YES Prevent injection from only those Core Spray and LPCI pumps not required NO Is suppression pool level above 5.5 feet YES Open all ADS Valves Directs ADS valves open ATC/BOP Opens all 6 ADS valves, reports all ADS valves open CT-i When pressure is below the shutoff head of the available injection systems direct injection SRO to restore level to +2 to +51 inches CT-i ATCIBOP Injects with available systems to restore level SRO Emergency Classification EPIP-1 1.1-S 1 Reactor water level can NOT be maintained above -162 inches. (TAF) 50

NRC Scenario 5 Simulator Event Guide:

Event 8 Component: Loss 480V SD BD 2A and LOCA WT= ATC Aligns Condensate JAW Appendix 6A

1. VERIFY CLOSED the following feedwater heater return valves:

. 2-FCV-3-71, HP HTR 2A1 LONG CYCLE TO CNDR

. 2-FCV-3-72, HP HTR 2B1 LONG CYCLE TO CNDR

. 2-FCV-3-73, HP HTR 2C1 LONG CYCLE TO CNDR.

2. VERIFY CLOSED the following RFP discharge valves:

. 2-FCV-3-19, RFP 2A DISCHARGE VALVE

. 2-FCV-3.-12, RFP 2B DISCHARGE VALVE

. 2-FCV-3-5, RFP 2C DISCHARGE VALVE.

3. VERIFY OPEN the following drain cooler inlet valves:

. 2-FCV-2-72, DRAIN COOLER 2A5 CNDS INLET ISOL VLV

. 2-FCV-2-84, DRAiN COOLER 2B5 CNDS INLET ISOL VLV

. 2-FCV-2-96, DRAIN COOLER 2C5 CNDS INLET ISOL VLV.

4. VERIFY OPEN the following heater outlet valves:

. 2-FCV-2-124, LP HEATER 2A3 CNDS OUTL ISOL VLV

. 2-FCV-2-125, LP HEATER 2B3 CNDS OUTL ISOL VLV

• 2-FCV-2-126, LP HEATER 2C3 CNDS OUTL ISOL VLV.

5. VERIFY OPEN the following heater isolation valves:

• 2-FCV-3-38, HP HTR 2A2 FW INLET ISOL VLV

• 2-FCV-3-3 1, HP HTR 2B2 FW INLET ISOL VLV

• 2-FCV-3-24, HP HTR 2C2 FW INLET ISOL VLV

• 2-FCV-3-75, HP HTR 2A1 FW OUTLET ISOL VLV

• 2-FCV-3-76, HP HTR 2B1 FW OUTLET ISOL VLV

• 2-FCV-3-77, HP HTR 2C1 FW OUTLET ISOL VLV.

6. VERIFY OPEN the following RFP suction valves:

• 2-FCV-2-83, RFP 2A SUCTION VALVE

• 2-FCV-2-95, REP 2B SUCTION VALVE

• 2-FCV-2-108, RFP 2C SUCTION VALVE.

7. VERIFY at least one condensate pump running.

8. VERIFY at least one condensate booster pump running.

9. ADJUST 2-LIC-3-53, RFW START-UP LEVEL CONTROL, to control injection (Panel 2-9-5).

10. VERIFY RFW flow to RPV.

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NRC Scenario 5 Simulator Event Guide:

Event 9 Component: RHR and Core Spray Division Injection Valves will not Auto open CT-i Aligns injection systems LPCI Loop I and II if directed and Core Spray Loop I, JAW BOP Appendix 6B, 6C and 6E Although most valve power is lost for RHR Loop I, injection is still available, the pumps have power, the Outboard Injection Valve does not have power but is normally open and the only valve with power is the Inboard Injection Valve which can be opened.

NOTE If RHR Loop I is used the only to control injection is to turn pumps on and off. In addition if it was aligned for Pool Cooling those valves will still be open, so the injection pressure to the vessel will be much lower.

BOP Appendix 6B

1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THENPLACE 2-HS-74-l 55A, LPCI SYS I OUTBD 1NJ VLV BYPASS SEL in BYPASS.

2. VERIFY OPEN 2-FCV-74-l, RHR PUMP 2A SUPPR POOL SUCT VLV.

3. VERIFY OPEN 2-FCV-74-12, RHR PUMP 2C SUPPR POOL SUCT VLV.

4. VERIFY CLOSED the following valves:

. 2-FCV-74-6l, RHR SYS I DW SPRAY 1NBD VLV

. 2-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV

. 2-FCV-74-57, RHR SYS I SUPPR CHBRJPOOL ISOL VLV 2-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

. 2-FCV-74-59, R}IR SYS I SUPPR POOL CLG/TEST VLV

5. VERIFY RHR Pump 2A and/or 2C running.

6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 2-FCV-74-53, RHR SYS I LPCI INBD iNJECT VALVE.

7. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 2-FCV-68-79, RECIRC PUMP 2B DISCHARGE VALVE.

8. THROTTLE 2-FCV-74-52, RHR SYS I LPCI OUTBD INJECT VALVE, as necessary to control injection.

Can inject but cannot throttle 2-FCV-74-52 and will have to open 2-FCV-74-53 with the BOP handswitch 52

NRC Scenario 5 Simulator Event Guide:

Event 9 Component: RHR and Core Spray Division Injection Valves will not Auto open Aligns injection systems LPCI Loop I and II if directed and Core Spray Loop I, lAW

= BOP Appendix 6B, 6C and 6E BOP Appendix 6C

1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THENPLACE 2-HS-74-155B, LPCI SYS II OUTBD NJ VLV BYPASS_SEL in BYPASS.

2. VERIFY OPEN 2-FCV-74-24, RHR PUMP 2B SUPPR POOL SUCT VLV.

3. VERIFY OPEN 2-FCV-74-35, RHR PUMP 2D SUPPR POOL SUCT VLV.

4. VERIFY CLOSED the following valves:

• . 2-FCV-74-75, RHR SYS II DW SPRAY INBD VLV

• . 2-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV

• . 2-FCV-74-71, RHR SYS II SUPPR CHI3R!POOL ISOL VLV

. 2-FCV-74-72, RHR SYS II SUPPR CHBR SPRAY VALVE

. 2FCV-74-73, RHR SYS II SUPPR POOL CLG/TEST VLV r 5. VERIFY RHR Pump 2B and/or 2D running.

...) 6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 2-FCV-74-67, RNR SYS II LPCI INBD iNJECT VALVE.

7. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 2-FCV-68-3, RECIRC PUMP 2B DISCHARGE VALVE.

8. THROTTLE 2-FCV-74-66, RNR SYS II LPCI OUTBD iNJECT VALVE, as necessary to control injection.

BOP Will have to open 2-FCV-74-67 with the handswitch 53

NRC Scenario 5 Simulator Event Guide:

Event 9 Component: RHR and Core Spray Division Injection Valves will not Auto open CT-i Aligns injection systems LPCI Loop I and!! if directed and Core Spray Loop I, JAW BOP Appendix 6B, 6C and 6E BOP Appendix 6E

1. VERIFY OPEN the following valves:

. 2-FCV-75-30, CORE SPRAY PUMP 2B SUPPR POOL SUCT VLV

. 2-FCV-75-39, CORE SPRAY PUMP 2D SUPPR POOL SUCT VLV

. 2-FCV-75-51, CORE SPRAY SYS II OUTBD INJECT VALVE.

2. VERIFY CLOSED 2-FCV-75-50, CORE SPRAY SYS II TEST VALVE.

3. VERIFY CS Pump 2B and/or 2D running.

4. WHEN RPV pressure is below 450 psig, THEN THROTTLE 2-FCV-75-53, CORE SPRAY SYS II 1NBD iNJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

Coordinate RPV Level Control to restore and maintain Level +2 to +51 inches.

N BOP/ATC Condensate and Core Spray will restore and maintain level. When RPV pressure is low J enough Condensate System will maintain directed level band.

BOP Will have to open 2-FCV-75-53 with the handswitch 54

NRC Scenario 5 SHIFT TURNOVER SHEET Equipment Out of ServicefLCOs:

100% power, DG D and RBCCW 2B Pump are tagged out. Spare RBCCW Pump is aligned for operation.

Temporary DGs are NOT provided.

OperationsfMaintenance for the Shift:

Lower Power with flow to 91% for Main Turbine Valve Testing.

Unit 1 and 3 are at 100% Power Unusual Conditions/Problem Areas:

Severe Thunderstorms are forecast for today, currently no watches or warnings are in effect.

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