Text

February 50-325/2003-301 Exam Draft Simulator Scenarios
	ML030840299
Person / Time
Site:	Brunswick ; (DPR-062, DPR-071, NPF-037, NPF-066)
Issue date:	09/09/2002
From:	Ernstes M; Operator Licensing and Human Performance Branch
To:	Keenan J; Carolina Power & Light Co
References
	50-324/03-301, 50-325/03-301
	Download: ML030840299 (185)
	v • d • e

Draft Submittal BRUNSWICK EXAM 50-2003-301 50-325 & 50-324 FEBRUARY 10 - 14 & 19, 2003 Operating Test Simulator Scenarios

NUREG-1021, Revision 8, Supplement I Appendix D Operator Actions Form ES-D-2 Op-Test No.:

Scenario No.:

Event No.:

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Event Description: _The crew assumes the shift with power at 90% EOL. There is a severe demand for power due to a late heat wave. RCIC Inboard Steam Isolation valve E-51-F007 is Inoperable for breaker maintenance (valve is OPEN). TS 3.6.1.3 is satisfied with E-51 F045 closed and its breaker open. The previous shift initiated OGP-13 "Increasing Unit Capacity at End of Core Cycle", Section 5.1 (Bypassing Feedwater Heaters #4 and #5).

Night orders include direction to complete a Control Rod Operability Check on rod 42-39 per OPT-1 4.1 (all rods but 42-39 were already completed) and then to increase power to 100%.

Time Position Applicant's Actions or Behavior The RO completes OPT-14.1 and commences power increase to 100%.

Following power increase to 100% APRM 1 fails HI (MNI031 F).

"This is intended to simply create an instrument malfunction for the RO, but may be interpreted by the crew as being associated with OGP-13. The RO is expected to bypass the APRM and reset the half scram. SRO will check T One channel of MS Radiation fails low (MRM001 F). The crew is

expected to take appropriate actions per Alarm Response Procedures and TS and may again attribute the failure to actions taken for OGP-1 At this point the fuel failure is increased and is readily detectable on the three remaining MS Rad Monitors as well as the Off Gas Monitor(s). Note it is intended that the fuel failure be high enough to ensure the MSIVs remain closed, but not so high as to cause radiation levels in the Reactor Building to exceed "Maximum Safe" level The crew may attempt to reduce power, but MS radiation will continue to increase (MRM00 1 F through 13F). The increase is intended to be slow enough to allow the SRO to make the decision K

to manually Scram and close the MSIVs before automatic action occurs. With the feedwater heaters bypassed there are restrictions on how low the power may be reduced. This may complicate the decision and may "push" the SRO to manually scram earlier rather than late I1 I

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When the MSIV closure occurs a RCIC Steam Break occurs in the South RHR room upstream of E-51-F045 (MES025F). The Area Temperature Hi isolation signal is generated, but valve F008 motor trips on overload (MES026F). The SRO will direct entry into Secondary Containment Control OEOP-03-SCCP and will, ultimately result in Emergency RPV Depressurization since all attempts to isolate the RCIC Steam Leak will be unsuccessful.

NUREG-1021, Revision 8, Supplement 1

Additional work that needs to be done on Scenario # 1 Risk for CRD & HPCI ooc TS actions for APRM Water Leak vs Steam Leak Turnover Notes for Scenario Reactivity Plan for Scenario Need O-GP-12 Signed off to step in effec.

Ensure both RBM have been declared OP & on scale Get Good power rods, 00-12, Rings 1, 2, 3 and 4/lc to get good rods Steps to trip APRM Channel Page I of 20

Scenario No.:

Title:Medium Break LOCA inside containment with Loss of Offsite Power and failure of one EDG Examiners:

Operators:

Initial Conditions: The crew assumes the shift with the plant at 28% power, BOL. HPCI is OOC, the #1 APRM failed low and is bypassed. One CRD (22-19) is inoperable; stuck at position 48. CRD Pump "2B" is OOC for PM's. Severe weather has been reported in the area. The previous shift completed partial stroking of all MSIVs except "A" inboard Turnover: The previous shift has completed all Required Actions for the Stuck Rod per TS 3.1.3 (Rod 22-19 is disarmed). Night orders include direction to partial stroke the "A" inboard MSIV per PT-40.2.8 and then to increase power to 40%. The Reactor Engineer recommends using control rods per Sequence A-1 (continuous withdrawal acceptable) with a maximum power increase of 1% per minute.

Event Malf. N Event Event N Type*

Description

N/A N(BOP)

Partial stroke "A" Inboard MSIV

N/A R(RO)

Increase power from 28% to 40%.

(SRO)

MRD018F C (RO)

"A" CRD Pump suction filter plugge MNI037F I (RO)

2 APRM fails lo MRC007F C (RO)

"A" Recirc Pump Seal Lea MRC009F (SRO)

MCN017F C (BOP)

AOG Guard Bed Fire (SRO)

MEE032 M(ALL)

Loss of Offsite Power

MDG002F C(BOP)

One EDG fails to start I

I_ Note: This event is combined with Event 7 in one D-2

MNBO09F M (ALL)

1000 GPM leak inside drywell

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Op-Test No.:

Facility:

Brunswick

Scenario No.:

Event No.:

Event Description: Perform Partial Stroke Test for the "A" Inboard MSIV per PT-40. The surveillance has been completed for the remaininq MSIV's. The simulator instructor will provide relay and light status from scripted messages when asked by the BO Time Position Applicant's Actions or Behavior SRO Direct the BOP to oerform the MSIV functional surveillance for the

"A" Inboard MSIV per procedure PT-40.2.8. Note: The simulator operator will function as the relay/light observer at "back" panels BOP 1. Obtain and review Precautions and Limitations of PT-40. Request RO to monitor Rx Pressure and Main Steam Line Flow during valve strokin. Verify that MSIV DC & AC coil lights (P622/P623)

lights/indications are illuminate. Verify that and RPS Group 1-4 (P609/P61 1) are li. DEPRESS B21-F022 TEST pushbutton and confirm valve strokes to dual position indication and relays operat CUE: Simulator Operator will report "C72-K3A contacts 1-2 & 3-4 are open; 5-6 & 7-8 are closed at P609 and C72-K3B contacts 1-2

& 3-4 are open; 5-6 & 7-8 are closed on P611" 5.Release B21-F022A TEST pushbutton and confirm B21-F022A fully open CUE: If requested, Simulator Operator will report "C72-K3A contacts 1-2 & 3-4 are closed; 5-6 & 7-8 are open at P609 and C72-K3B contacts 1-2 & 3-4 are closed; 5-6 & 7-8 are open on P611" RO Monitor Main Steam Flow and Reactor Pressure Note: At this low power no significant pressure increase or flow change is expected Page 3 of 20 Op-Test No.:

Op-Test No.:

Scenario No.:

Event No.:

Page 4 of 20 Event Description: Perform Partial Stroke Test for the "A" Inboard MSIV per PT-40. The surveillance has been completed for the remainina MSIV's. The simulator instructor will provide relay and liqht status from scripted messages when asked by the BOP.

Time [ Position I Applicant's Actions or Behavior

Op-Test No.: __

Scenario No.:

Event No.:

Page 5 of 20 Event Description: Increase power from 28% to 40%. Note: The next event (CRD filter cloQQed) will be initiated when the Reactivity Manipulation has been satisfied or at the Lead Examiner's discretion Notes:1. The Reactor Engineer has specified power increase be accomplished using continuous withdrawal with Control Rod Sequence A-1 not to exceed 1% per minut.The RO will be given a copy of OGP-12 with sections 5.2.1 through 5.2.17 completed (signed off).

Time Position Applicant's Actions or Behavior SRO 1. Ensure power increase is acceptable to Load Dispatcher 2.Direct the RO to increase power in accordance with CGP-12. He will specify using continuous withdrawal of Control Rods in sequence A-1 at <1% power/mi. He will direct the BOP to monitor Turbine Operation in accordance with OP-26, Figure RO_

1) Obtain current copy of 20P-07 and OGP-10 2) Review Precautions and Limitations and Prerequisites in OGP 12, and OGP-1 ) Ensure steps 5.2.1 through 5.2.17 and Attachment 2 of OGP-12 have been completed 3) Withdraw control rods per sequence A-1 to effect <1%/min increase and monitor Reactor Parameter Select the desired rod

- Ensure ROD WITHDRAWAL PERMISSIVE is lit

- Hold EMERGENCY ROD IN NOTCH OVERRIDE to OVERRIDE and NOTCH OUT to effect continuous withdrawal

- Release both switches and ensure rod settles to desired position

- Monitor CRD position reactor power during withdrawal

- Check coupling integrity at 48

- Stop rod withdrawal one notch before desired position

- Notch rods with movement of 3 notches or less

- Place ROD SELECT POWER to OFF at end of rod movement 4) Swap Feedwater Level Control to 3-ELEM (if not previously done)

Op-Test No.: __

Scenario No.:

Event No.:

Page 6 of 20 Event Description: Increase power from 28% to 40%. Note: The next event (CRD filter cloooed) will be initiated when the Reactivity Manipulation has been satisfied or at the Lead Examiner's discretion Notes: 1. The Reactor Engineer has specified power increase be accomplished using continuous withdrawal with Control Rod Sequence A-1 not to exceed 1% per minut.The RO will be given a copy of OGP-12 with sections 5.2.1 through 5.2.17 completed (signed off).

Time Position Applicant's Actions or Behavior 5) Perform APRM Normal Trip Setpoint (if not previously done)

- CONFIRM FCBB is less than DEPRESS the NORMAL/SETUP pushbutton on FCTR card for each APRM

- CONFIRM NORMAL/SETUP LED is yellow 6) CONFIRM Turbine Stop Valve/Control Valve Fast Closure Reactor Scram is enabled

- Confirm C71 Relays are deenergized

- Confirm TURB CV FAST CLOS/SV TRIP BYPASS (A-05, 6-7) is clear 6) Confirm core thermal limits are per TS; Power/Flow is outside Restricted Region BOP 1 Obtain a copy of OP-2. Monitor Turbine Power Increase per OP-26, Figure Power change from 28% to 40 % can be done as a step increase

- Routinely monitor Turbine Parameters (Oil, vibration, vacuum)

3. Start a second Condensate Booster Pump 4. Notify Radwaste to put additional CDDs and CFDs in service

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Op-Test No.:

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Scenario No.:

Event No.:

Page 7 of 20 Event Description: Increase power from 28% to 40%. Note: The next event (CRD filter clogged) will be initiated when the Reactivity Manipulation has been satisfied or at the Lead Examiner's discretion Notes: 1. The Reactor Engineer has specified power increase be accomplished using continuous withdrawal with Control Rod Sequence A-1 not to exceed 1% per minut.The RO will be given a copy of OGP-12 with sections 5.2.1 through 5.2.17 completed (signed off).

TimeI Position Applicant's Actions or Behavior

Op-Test No.:

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Scenario No.:

Event No.:

Page 8 of 20 Event Description: "A" CRD Pump suction filter plugge Execute malfunction MDR01 8F per Lead Examiner directio Time Position Applicant's Actions or Behavior RO 1. Observe alarm "CRD Pump Inlet Filter DP High" alarm and monitor CRD parameter. Review Alarm Response Procedure A-05 5-. Stop any power increase 4. If more than one control rod is drifting THEN INSERT a manual scram SRO 1. Direct RO to complete ARP actions 2. Dispatch an AO to locally monitor filter DP and pump suction pressure CUE: AO will report that filter dp > 10 psid 3. Obtain a copy of OAOP-02.0 and direct RO actions as appropriate (pump failure)

4. Consult TS 3. Recognize that If/when one accumulator alarm comes in have 20 minutes to restore CRD pump (TS 3.1.5.B.1)

- Acknowledge/Specify 20 minute "clock" to get CRD pump ON CUE: Simulator provide second accumulator alarm in approximately 3 minutes RO 1. Observe running CRD pump trips and "CRD Pump 2A Lo Suct Press" alarm (Note: Other CRD alarms will come in as well)

2. Monitor core thermal parameters to keep within TS 3. Direct the AO to shift CRD Pump suction filters 4.When advised (simulator operator) that the CRD suction filters have been shifted, Re-Startthe 2A CRD Pump

Op-Test No.: __

Scenario No.:

Event No.:

Page 9 of 20 Event Description: "A" CRD PumD suction filter plugge Execute malfunction MDR018F per Lead Examiner directio Time Position Applicant's Actions or Behavior CUE: The AO will report back in 10 minutes that filter swap has been completed 5. Observe CRD parameters return to "normal"; specifically that charging water >940 psig within 20 minutes 6. Request the AO ensure the CRD suction filter is less than 3 PSI CUE: The AO will report suction filter DP is I psid SRO 1. Note shifting of CRD filters in log and initiate MR to replace filters in clogged uni. Ensure TS are satisfied in new configuration/exit TS 3.1.5

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Op-Test No.:

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Scenario No.:

Event No.:

Page 10 of 20 Event Description: "A" CRD Pump suction filter plugge Execute malfunction MDR018F per Lead Examiner directio Time Position Applicant's Actions or Behavior

Op-Test No.:

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Scenario No.:

Event No.:

Page 11 of 20 Event Description: #2 APRM Fails Low Insert malfunction MNI037F at direction from Lead Examiner Time Position Applicant's Actions or Behavior RO 1 Observe "APRM Downscale" alarm and notify SRO 2. Compare #2 APRM with #3 and #4 APRM channels (A-06 2-7)

3. Refer to TS 3.3.1.1 and TRMS 3.3 for APRM operability requirements 3. Confirm other APRM channels are functional (except #1)

4. Trip the affected APRM channel (Steps)

SRO 1. Consult TS 3.3.1.1 and TRMS. Ensure no other half scrams are present 3. Direct RO (or BOP) to place the appropriate APRM Channel to TRIP RO Ensure half scram-RPS Group Lights for Channel "A" out SRO 1. Ensure TS are satisfied and log information 2. Initiate MR to get APRM #2 repaired

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Op-Test No.: __

Scenario No.:

Event No.:

Page 11 of 20 Event Description: "A" Recirc PumD Seal Leak. Enter malfunctions MRC014F, MRC007F and MRC009F when directed by Lead Examine Time Position Applicant's Actions or Behavior RO 1.Observe "Recirc Pmp A Motor Vib High" alarm and notify SR Monitor Recirc Pump A motor bearing temperatures on recorder B32-R601-Attempt to reset the alarm to determine if it was spurious CUE: The recorder shows increasing motor bearing temperatures and the vibration alarm cannot be reset. MRCOO7F should be initiated after RO and SRO confer on vibration, but before any decision to Trip the pump is made 2.Observe "Outer Seal Leakage Flow Detection Hi" alarm and notify SR Note: This should be a clear indication that the seal has failed and that a discharge into containment is occurrin. Observe changing seal DPs and notify SRO that both seals appear to have failed or are failing 3. Request AO verify correct seal lineup and monitor parameters in reactor building 4. Monitor seal cavity temperatures on recorder B32-R601 at Panel P61 Note: Pump should be tripped and isolated prior to seal temperatures exceeding 200 F SRO 1. Direct RO to complete actions in ARPs (A-06 3-3 and A-06 5-3)

2. Direct BOP to monitor containment parameters BOP 1. Observe Drywell pressure and temperature and sump levels all increasing, refer to OAOP-14.0 and advise SR. Monitor drywell equipment drain sump pumps for frequency of operation and run tim. Ensure all drywell coolers are operating and RBCCW lined u Op-Test No.:

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Scenario No.:

Event No.:

Page 10 of 20 Event Description: "A" Recirc Pump Seal Leak. Enter malfunctions MRC014F. MRC007F and MRC009F when directed by Lead Examine Time Position Applicant's Actions or Behavior SRO 1. Direct RO to reduce power in anticipitation of tripping "A" Recirc Pump. Notes: Pump should be tripped and isolated if seal temperatures reach or exceed 200 F (A-06 5-3) SRO may direct to Transfer per OP-0. Call dispatcher and advise of power reduction/possible scram RO*

SRO may direct the BOP to Trip and Isolate the pump if the RO is driving rods 1. Reduce power as necessary per OGP-12 Note: Should insert rods per OGP-1 0, but may just transfer from two loop to one loop as specified in the following steps.

2. Place control switch for Seal Staging VIv B32-V14 to MAN/OPEN 3. Shutdown recirc Pump 2A by placing RECIRC MG SET 2A control switch to STOP Note:Shutdown of "A" Recirc Pump should be before 200F seal temperatures are reached. This should be done without a Turbine Trip from High RPV water level 5. Isolate "A" Recirc Pump (Critical Task)

- Close suction valve prior to closing discharge valve

- Close discharge valve before closing discharge bypass

- Isolate seal purge (from CRD) prior to closing bypass valve 6.Determine core flow using point WTCF 7.Raise speed on Recirc pump B if permitted by feedwater interlocks.

8. Ensure total core flow is<45mlb/hr 9. Ensure Precautions and Limitations of OGP-12 are satisfied following pump trip (E.G., Power/Flow)

10 CHANGE or REMOVE rods on SRI bus if approaching 30%

Op-Test No.:

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Scenario No.:

Event No.:

Page 11 of 20 Event Description:

"A" Recirc Pump Seal Leak. Enter malfunctions MRC014F, MRC007F and MRC009F when directed by Lead Examine Time Position Applicant's Actions or Behavior BOP/RO 1. If/when Drywell Pressure exceeds 1.7 psig advise SR. Start all available DW coolers-DO NOT vent DW (leak present)

SRO 1. Notify Operations Management and Reactor Engineer of power reductio. Ensure TS for single loop operation are satisfied 3. Log events 4. Enter OEOP-02-PCCP if/when drywell pressure exceeds psig

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Op-Test No.: __

Event Description:

and MRC009F wher Scenario No.:

Event No.:

Page 12 of 20

"A" Recirc PumD Seal Leak. Enter malfunctions MRC014F, MRC007F n directed by Lead Examiner.

Time I Position Applicant's Actions or Behavior

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+/- _______________________________________________________

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Op-Test No.: __

Scenario No.:

Event No.:

Page 13 of 20 Event Description: AOG Guard Bed Fire. Execute malfunction MCN01 7F when directed by Lead Examiner.

Time Position Applicant's Actions or Behavior BOP 1 Acknowledge receipt of "2-AOG-D1 Guard Bed Temperature High" alarm and advise SRO 2. Request AO to observe AOG parameters at local panel H2E Note: AO will report back that "guard bed temperatures have significantly increased but reheater temperatures are normal".

2. Ensure AOG HVAC equipment is normal 3. Bypass and Isolate guard bed-Open 2-AOG-V013 and 2-AOG-V014 THEN-Close 2-AOG-VO09, 2-AOG-V010, 2-AOG-V01 I and 2-AOG-V012 4. Monitor guard bed temperature If guard bed continues to rise, immediately purge with N2 per OP-33 CUE: Guard bed temperatures continue to rise even after isolating SRO 1.Based on hearing that the guard bed temperatures continue to rise after isolation, he should obtain a copy of OP-33 and direct an AO to "purge the guard bed with N2 in accordance with Section 8.5"

- Close AOG-V01 1 and V-012

- Disconnect switch ON

- Throttle open Nitrogen Purge supply Valve AOG-NP-V079

- Establish 50 SCFM N2 flow 2. Notify OPS Management and HP of the problem CUE: Once the order is given (to the AO) to purge the guard bed, MCNO17F will be removed. This event will be considered complete when temperatures of the guard bed begin to decrease.

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Page 14 of 20 Op-Test No.:

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Scenario No.:

Event No.:

Page 14 of 20 Event Description: AOG Guard Bed Fire. Execute malfunction MCN017F when directed by Lead Examiner.

Time Position Applicant's Actions or Behavior

Op-Test No.:

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Scenario No.:

Event No.:

Page 15 of 20 Event Description: Loss of Offsite Power (with failure of #1 EDG to start). Malfunctions MEE032 and MDG002F shall be inserted at the direction of the Lead Examine Time Position Applicant's Actions or Behavior SRO 1. Direct actions of EOP-01-RS. Obtain copy of AOP-36.1 and direct action Note: Although the loss of one EDG to start is considered a separate event in the D-I it is combined into this event with the BOP to take additional actions for manually restarting the ED. If/when suppression pool temperature exceeds 95 F, enter EOP-02-PCC RO 1. Complete actions of EOP-01-RSP "Reactor Scram".

-Mode switch to Shutdown-Verify power<5%

-Operate RCIC in level band +170" to +200"

-Open SRV to obtain 950 psig-Insert nuclear instrumentation 2. Recognize that not all rods are full in and advise the SRO 2. Monitor and control RPV level and pressure using RCIC and SRVs 3. Place RHR in suppression pool cooling without keepfill Note: Starting of RHR SW Pumps will be dependent on power availability 4.Start RPS MG Sets 5. Restart CRD pump(s) as necessary BOP 1. Recognize/announce that one EDG did not start 2. Attempt to start the failed EDG Note: This sequence is continued below 3. Ensure DC Oil Pumps start 4. Ensure NSW Pumps running, Start CSW Pumps to support RCC 5. Ensure battery chargers operating

Op-Test No.:

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Scenario No.:

Event No.:

Page 16 of 20 Event Description: Loss of Offsite Power (with failure of #1 EDG to start). Malfunctions MEE032 and MDG002F shall be inserted at the direction of the Lead Examine Time Position Applicant's Actions or Behavior 6. Start Control Room and battery Room HVAC 7. Restore Drywell Coolin. IF motor driven Fire Pump is ON, then Start Diesel Driven Fire Pump and shutdown motor driven (power conservation)

9. Trip Main Turbine (part of RSP) and monitor lube oil SRO 1. When advised of one rod full out he should recognize that the reactor will remain shutdown under all conditions and NOT go to

"Level/Power Control" 2. When advised that one EDG did not start he should direct the BOP to try a manual start per AOP-3 Note: It is likely that the BOP would do this on his own per OAOP 3 BOP 1. Manually start the failed EDG. Note: The BOP will be

"permitted" to manually start the EDG but it will immediately Trip on differential current when he attempts to close the output breaker 2. Attempt to close breaker (insert malfunction MDG024F)

3. Recognize that the Differential Fault is serious and probably means the EDG is lost for the duration of the even. Direct an AO to investigate the faul Note: Once the BOP recognizes the EDG is "gone" and actions of OAOP-36.1 have been completed, step into the next event (1000 GPM leak)

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Scenario No.:

I Event No.:

Page 17 of 20 Event Description: Loss of Offsite Power (with failure of #1 EDG to start). Malfunctions MEE032 and MDG002F shall be inserted at the direction of the Lead Examiner.

Time Position Applicant's Actions or Behavior

Op-Test No.: __

Scenario No.:

Event No.:

Page 18 of 20 Event Description:

1000 GPM leak inside the drvwell. Malfunction MNBO09F will be initiated as directed by the lead examiner Time Position Applicant's Actions or Behavior BOP 1. Observe containment parameters and identify that a leak inside containment is in progress. Note: The leak will start small and gradually increase to 1000 GPM. Initially RCIC and CRD will be able to maintain level, but level decrease will be apparent when leakage exceeds approximately 600 GPM SRO 1. Direct actions per EOP-01-RVCP. Execute RC/L and RC/P concurrently 2. Once it is apparent that Reactor Water cannot be maintained above TAF the SRO will direct the RO to initiate a cooldown to the point that either Core Spray or LPCI will makeup to the Reacto. When drywell pressure or suppression pool temperature exceeds entry conditions, he will enter OEOP-02-PCCP RO 1. Maximize RPV injection with available high pressure sources (RCIC and CRD)

2. Place Torus Sprays in service prior to exceeding 11.5 psi Note: The RO should observe power constraints with only one EDG available and use OAOP-36.1 when starting RHR pump(s)

3. Alternate SRVs to maintain/reduce RPV pressure (cooldown <

100 F/hr)

4. Inhibit ADS 5. Recognize/advise SRO that available high pressure sources will not be adequate to maintain RPV level above TAF 6. Ensure Core Spray is lined up for injection

7. Depressurize RPV using one or more SRVs until Core Spray is injecting Notes:

1. Shutoff head approximately 350 psig 2. The SRO may determine that water level cannot be maintained above TAF and proceed to Emergency Depressurization The scenario will be terminated once Core Spray flow has been established to the RP Event 8 LOCA exceeding high pressure makeup Time Position Applicants actions or behavior Event 8 LOCA exceeding high pressure makeup Time Position Applicants actions or behavior

_7771

Control Rod OPERABILITY 3..1 REACTIVITY CONTROL SYSTEMS 3. Control Rod OPERABILITY LCO 3. Each control rod shall be OPERABLE APPLICABILITY:

MODES 1 and ACTIONS

------------------------ NOTE ------------------------------------

Separate Condition entry is allowed for each control ro CONDITION REQUIRED ACTION COMPLETION TIME One withdrawn control

NOTE

N rod stuc Stuck control rod may be bypassed in the rod worth minimizer (RWM)

or RWM may be bypassed as allowed by LCO 3.3.2.1, "Control Rod Block instrumentation,"

if required, to allow continued operatio Verify stuck control Immediately rod separation criteria are me AND Disarm the associated 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months control rod drive (CRD).

AND (continued) (continued) Perform SR 3.1. hours from and SR 3.1.3.3 for discovery of each withdrawn Condition A OPERABLE control ro concurrent with THERMAL POWER greater than the low power setpoint (LPSP)

of the RWM AND Perform SR 3.1. hours 3.1-7 Brunswick Unit 1 Amendment N.1-7

Control Rod OPERABILITY 3.1.3 flL. +/-.2. JLN 0 CONDITION REQUIRED ACTION COMPLETION TIME Two or more withdrawn control rods stuc One or more control rods inoperable for reasons other than Condition A or B. (continued)

D. -----------

NOTE Not applicable when THERMAL POWER

> 8.75% RT Two or more inoperable control rods not in compliance with banked position withdrawal sequence (BPWS)

and not separated by two or more OPERABLE control rods.

B.I Be in MODE 3.

C.1---------

NOTE---------

Inoperable control rod may be bypassed in the RWM or RWM may be bypassed as allowed by LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable control rod.

AND C.2 D.1 OR D.2 Disarm the associated CRD.

Restore compliance with BPW Restore control rod to OPERABLE status.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 3 hours (continued)

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4 hours 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Amendment N Brunswick Unit 1 3.1-8

Control Rod Scram Accumulators 3.1.5 ACTIONS (continued)

CONDITION Required Action or C.1 and associated Completion Time not met.

REQUIRED ACTION NOTE --------

Not applicable if all inoperable control rod scram accumulators are associated with fully inserted control rods.

COMPLETION TIME Manually scram the Immediately reactor.

Amendment N Brunswick Unit 1 3.1-17

Control Rod Scram Accumulators 3.1.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME Two or more control Restore charging 20 minutes from rod scram accumulators water header pressure discovery of inoperable with to Ž 940 psi Condition B reactor steam dome concurrent with pressure Ž 950 psi charging water header pressure

< 940 psig AND B. NOTE--------

Only applicable if the associated control rod scram time was within the limits of Table 3.1.4-1 during the last scram time Surveillanc Declare the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated control rod scram time

"slow."

OR B. Declare the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated control rod inoperabl (continued) One or more control Verify all control Immediately upon rod scram accumulators rods associated with discovery of inoperable with inoperable charging water reactor steam dome accumulators are header pressure pressure < 950 psi fully inserte < 940 psig AND Declare the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated control rod inoperable.

Amendment N.1-16 Brunswick Unit 1

Control Rod Scram Accumulators 3.1.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME Two or more control Restore charging 20 minutes from rod scram accumulators water header pressure discovery of inoperable with to Ž 940 psi Condition B reactor steam dome concurrent with pressure Ž 950 psi charging water header pressure

< 940 psig AND B. NOTE--------

Only applicable if the associated control rod scram time was within the limits of Table 3.1.4-1 during the last scram time Surveillanc Declare the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated control rod scram time

"slow."

OR B. Declare the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated control rod inoperabl (continued) One or more control Verify all control Immediately upon rod scram accumulators rods associated with discovery of inoperable with inoperable charging water reactor steam dome accumulators are header pressure pressure < 950 psi fully inserte < 940 psig AND Declare the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated control rod inoperable.

Amendment N.1-3.6 Brunswick Unit I

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Required Action and Be in MODE hours associated Completion Time of Condition A, C,

or D not me OR Nine or more control rods inoperable.

Amendment N Brunswick Unit I 3.1-9

Control Rod OPERABILITY 3.1.3 REACTIVITY CONTROL SYSTEMS 3.1-3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.

APPLICABILITY:

MODES 1 and ACTIONS

-----------------------

NOTE ------------------------------------

Separate Condition entry is allowed for each control ro.

..

...

....-------------------------------------------------------------------

CONDITION REQUIRED ACTION COMPLETION TIME One withdrawn control

NOTE------

rod stuc Stuck control rod may be bypassed in the rod worth minimizer (RWM)

or RWM may be bypassed as allowed by LCO 3.3.2.1, "Control Rod Block Instrumentation,"

if required, to allow continued operatio Verify stuck control Immediately rod separation criteria are me AND Disarm the associated 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months control rod drive (CRD).

AND (continued) (continued) Perform SR 3.1. hours from and SR 3.1.3.3 for discovery of each withdrawn Condition A OPERABLE control ro concurrent with THERMAL POWER greater than the low power setpoint (LPSP)

of the RWM AND Perform SR 3.1. hours

'

e,..,ý,-'

Amendment N.1-7 Brunswick Unit I

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Two or more withdrawn Be in MODE hours control rods stuc One or more control NOTE--------

rods inoperable for Inoperable control reasons other than rod may be bypassed Condition A or in the RWM or RWM may be bypassed as allowed by LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operatio Fully insert 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months inoperable control ro AND (continued) (continued) Disarm the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CR D. ----------- NOTE ----.---- Restore compliance 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Not applicable when with BPW THERMAL POWER

> 8.75% RT OR Restore control rod 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Two or more inoperable to OPERABLE statu control rods not in compliance with banked position withdrawal sequence (BPWS)

and not separated by two or more OPERABLE control rods.

Amendment N.1-8 Brunswick Unit 1

EVENT 1 MSIV TESTING Instructor Activities o Establish communications with BOP operator; report you are on station to observe relay actuation during PT performanc [o If asked MSIV DC & AC coil lights (P622/P623) are lit and RPS Group 1-4 (P609/P611) are lit o Monitor MSIVs on panel mimic or cameras. When B21-F022A is closed, report relay C72-K3A contacts 1-2 & 3-4 are open, 5-6 & 7-8 are closed (P609) and that C72-K3B contacts 1-2 & 3-4 are open, 5-6 & 7-8 are closed (P611)

o When B21-F028A is attempted to be stroked, report no relay actuatio [o NOTE: The SCO has no way of knowing exactly why MSIV fails to stroke (could simply be a faulty test circuit)

Plant Response o] B21-F028A fails to stroke when attempted Operator Activities SCO o Direct Reducing Reactor Power to <85% using Recirculation flow o] Direct PT-40.2.8 be performed

[o Direct suspension of PT o Direct I&C to investigate failure of B21-F028A

[o Initiate tracking LCO on MSIV 28A and request engineering evaluation as to why valve did not stroke (01-01.08). (No TS requirement for valve to slow close, the potential operability issue is will the valve respond to an isolation)

RO

[o Reduce Reactor Power to <85% using Recirculation flow r-Monitor the plant during performance of PT-40. BOP o Verify prerequisites for test are met

[o Confirm no scram alarms are sealed in o Establish communications with individuals in back panels

[o Determine FFTR not in effect o Transfer Feedwater control to 1-elem per OP-32 oi Confirm scram group light are illuminated E] Depress B21-F022A test pushbutton, confirm valve strokes to dual position indication and relays operate ol Release B21-F022A pushbutton and confirm B21-F022A fully opens

[o Depress B21-F028A test pushbutton ol Recognize and report failure of B21-F028A to stroke 1 * *LT-IPO-203 Page 8 of 15 Rev.01

The bases associated with the MSIV stroke time limitations serve to both preserve the integrity of the Reactor Coolant System and to minimize radioactive releases. The minimum stroke time of 3 seconds allows the MSIV closure scram signal to reduce reactor power and pressure before the motion of the MSIV has caused pressure to start to rise. Since the MSIV Closure Scram signal is initiated by valve position less than 90% open and the valve does not significantly effect steam line flow or pressure until less than -25% open, it can be seen that, ideally, power and pressure are both decreasing before the MSIV closures can effect reactor power. The maximum stroke time of 5 seconds is twofold in that it serves to minimize the inventory loss from the Reactor Vessel and minimize the radioactive releases to the environmen Minimizing the inventory loss will reduce the likelihood of core uncovery, thus prevent fuel damage. Minimizing radioactive releases prevents exceeding the accident release limitations stated in 1O-CFR-100. Analysis has shown that excessive releases or core uncovery are not expected to occur for closing times up to 10.5 second Each MSIV operator contains two AC solenoids and one DC solenoid. One of the AC solenoids is used for valve stroke testing at power and is called the Slow Closure Test Solenoid. The other two solenoids (one AC and one DC)

determine the position of the MSIV by porting or venting the pneumatic source to or from the operator. Both of these solenoids must be deenergized for the MSIV to be closed. The AC solenoids are powered from the Reactor Protection System and the DC solenoids are powered from the Station Battery System. This arrangement prevents inadvertent MSIV closure unless redundant signals are received yet, is fail safe in that a loss of power will result in closure. MSIV position is controlled via CLOSE AUTO/OPEN control switches located on the P601 panel. The automatic function is associated with the AUTO/OPEN position and closure is caused by the Primary Containment Isolation System (PCIS) or a Low Vacuum condition in the Main Condense AI PRAO R

AIR CýlINDER CONTRaL V.lVES W.'AIR PISTON LOCATED HERE (NT SHOWN)

YOISRAULUC SPEED CONTROL CLOSURE SPRINGS-SPRING GUIDE (10% CLOSED I RIPS VALVS COUPLNGE SP*IN FLAGE

pE*

=*--LIMIT SWTCHE S ACTATOrC SUPPORT-(%

CLCSEDý)

PAD SPRING GUIDE

" B PACKING LEAK-OFF SHAFT TUBE)LINE

[

CE D)

E BONNETPACKING VALVE BODY SD-2e. 5 PEMaefC SD-2 Rev. 5 Page 16 of 87

MSIV isolation signal status is given by a group of white lights located on the P601 panel. These lights are arranged above the MSIV control switches as follows:

TABLE 25-3, MSIV ISOLATION SIGNAL STATUS Light INBD DC INBD AC OUTBD DC OUTBD AC Solenoid 125 VDC RPS "A" 125 VDC RPS "B" Power

"A"

"B" PCIS Logic B

A A

B A half isolation signal sensed by the "A" PCIS logic will result in extinguishing the two lamps in the center. The two outboard lamps will remain lit and no valve motion will occur. The two extinguished lamps represent the inboard AC and the outboard DC solenoids. The white lamps for the inboard valves are repeated on the P622 panel while those for the outboard valves are repeated on the P623. MSIV solenoid power status is provided by red LED indicators located on the back of P622 for the Inboard MSIVs and the back of P623 for the Outboard MSIVs. Each location contains eight RED LED lights, four for DC power and four for A Isolation capability to the inboard and outboard AC MSIV Pilot Solenoids circuitry is provided by a key locked isolation switch located on Panel P622 for the inboard valves and on Panel P623 for the outboard valves. In the event of a fire the key locked isolation switches provide the capability for total isolation of the AC power supplies to the pilot Solenoids. The isolation contacts will remove the ground path from the circuit in case of a "hot" short in the field cable which will prevent spurious valve operation.

2.5.1 Pneumatic Operator Operation of the MSIVs (Figure 25-6A thru 25-6C) is pneumatic to open and pneumatic with spring assist to close. Each MSIV is supplied with two pneumatic sources via a dual header and check valve arrangement. This arrangement allows for the loss of either source without effecting the other. The pneumatic sources for the outboard MSIVs are Reactor Building Non-Interruptible Air (RNA) System Division I and Division II. The pneumatic sources for the inboard MSIVs are Pneumatic Nitrogen System (PNS) Division I and Division II. Unlike the SRVs, a loss of PNS does not result in lining up the BU N2 System to the pneumatic operators.

SD-25 Rev. 5 Page 17 of 87

An air accumulator located between the MSIV air operator and the check valves provides backup operating air. The capacity of the accumulator is sufficient for the air operator to exercise the valve through one-half of a cycle (open-to-closed or closed-to-open) should the supply air to the operator be interrupte The MSIV air operator control unit is attached to the air cylinder and contains the pneumatic, AC, and DC solenoid valves and control valves for opening, closing, and slow speed exercising of the main valve. The control power for each MSIV is 120 volts AC, and 125 volts D During normal operation (MSIV open) either or both the AC and DC coils of the solenoid valve(s) are energized and instrument air is directed to the underside of the air operator piston. Thus, the closing force exerted by the springs is overcome by the air operator and the valve is maintained in the open positio When the solenoid operated control valve(s) are deenergized as in a two-channel trip or whenever the manual control switch is taken to the closed position, the air supply pressure is switched to pressurize the opposite side of the air operator piston and assists the spring to close the valv Air supply pressure acting on the operator piston or spring force is capable of independently closing the valve with the reactor vessel at full pressure. Thus if one fails, the other will successfully close the valve (provided area below the operator piston is vented off to atmosphere). Two vent valves provide a redundant means for bleeding off the under piston air in the event that a valve fails to operate on a valve closure signa A separate test pushbutton is provided for a manual test of slow closure of each MSIV from the Control Room. Slow Closure of a MSIV when testing should require 45 to 60 seconds (Figure 25-6C). Slow closure utilizes springs only and vents the area under the operator piston through an adjustable aperture.

SD-25 Rev. 5 Page 18 of 87

Fast closure time 2.5.2 is set between 3 and 5 seconds. These time limits are set in order to:

(1) prevent uncovering of the core through loss of inventory, (2)

reduce the amount of activity released to the environs in the event of a gross steam leak, and (3) minimize the pressure transients on the reactor vessel and fuel. Stroke time is obtained through adjustment of two timing control knobs located on the pneumatic operator. The upper speed control knob is to adjust opening speed and the lower control knob is to adjust closing speed. The air supply piping to the operator is sized such that no depressurization of the accumulator will take place during valve operatio Automatic Closure of MSIV's The main steam isolation valves are designed to close on any of the following primary containment isolation signals:

Reactor Water Low Level 3

Main steam line high flow

Reactor Building steam line tunnel high temperature

Turbine Building main steam tunnel high temperature

Low Condenser Vacuum - The low condenser vacuum trip bypassed when the turbine stop valves are less than 90% open and bypass switches (A71B-S34A-D) are in the bypass position and the reactor mode switch is not in the run positio * Main Steam Line Low Pressure - The low turbine inlet pressure trip is bypassed whenever the reactor mode switch is not in the run positio (See Table 25-5, Instrument and Control Setpoints.)

It is possible to completely isolate two steam lines by manually closing up to four MSIVs from the Unit RTGB without initiating a full reactor scram, provided the four MSIVs are only in two steam lines. Any attempt to isolate three main steam lines will cause a full scram through both RPS trip channel The automatic isolation signal is a one-out-of-two-taken-twice logic. A trip must occur in one or both trip channels of Trip System "A" and in one or both trip channels of Trip System "B", in order to initiate valve closure (Figure 25-7, 25-7A, and 25-7B).

SD-25 Rev. 5 Page 19 of 87 Reactor Protection System In the Figure 25-7 and 25-7A, we can see that the position of each MSIV is sensed by contact position from limit switches mounted on the valve. When the valve is >90% open the contacts (trip system A and B) will be closed indicating the valve is FULL OPEN. Likewise, if the valve is <90% open, the contacts are open and the appropriate relays are de-energized. This will be indicated by the MAIN STEAM ISOL VLV NOT FULL OPEN (A-05, 4-6) (any two of eight valves) annunciator. Shutting the second valve in the same steam line has no further effect on the relay Assuming that either valve in the A steam line has been shut, if a valve in either the B or C steam line is also shut then a Half-Scram will result in either RPS A or B respectively. Notice that having MSL A and D valves shut does not result in a Half-Scram. This is also true if only MSL B and C have valves shut. In either of these two cases, the first response of the RPS system will be a Half-Scram. The conclusion to be drawn from this is that any combination of valves may be shut in any two steam lines and the worst case is still only a Half-Scram. No matter the selected combination, any time a valve is closed in the third steam line a Full-Scram will resul Restated, the shutting of the three MSIVs may only result in a Half-Scram while the shutting off of any three steam lines will result in a Full-Scra.6 PClS Actuation/Bypasses Unless bypassed automatically or manually, any one of the isolation signals listed in Section 2.5.1 will result in full closure of the Group I Isolation Valves. In addition, this signal will block opening of these valves until reset by operator action. It should be noted that the isolation circuitry does not necessarily discriminate between sensed parameters. If an A or C channel signal exists from a Group I signal, a B or D channel signal from any other parameter will result in a Group I Isolatio The Unit 2 Main Steam Line High Flow (30%) Isolation is bypassed automatically when the Reactor Mode Switch is in the RUN position. This isolation is required at lower powers (out of RUN) to protect against pressure regulator failures which result in excessive Reactor Vessel cooldown caused by the opening of Turbine Bypass Valves. In this instance, the steam flow spike (thermal transit) resulting from the failure could be significant because of the 105% bypass valve capacity. On Unit 1, with only 25% bypass capacity, the failure is limited to a spike of 25% flow and the 30% isolation is not needed.

SD-25 Rev. 5 Page 29 of 87

Surveillance Testing During normal plant operation, the routine tests performed on the system include monthly Slow Closure Testing of the MSIVs, Full Closure Time Testing and Periodic Isolation Logic Instrumentation and Circuitry Testin MSIV Slow Closure Testing (PT-40.2.8) is performed to exercise the valve along its stroke and to verify that the Reactor Protection System recognizes that the valve is out of position. This test is performed at reduced Reactor Power (<80%) to avoid the high steam line flow isolation as sensed in the other three lines when the tested valve is shu The Full Closure Timing Test (PT-25.1) is the operability test of the Main Steam Line Isolation Valves. This test times the normal closure of the MSIV from switch selection to full close light indication. As discussed before, this stroking must be completed in a 3 to 5 second interva OPT-31.1, Non-interruptible Instrument Air System Valve Operability Test verifies proper operation of Division I and II check valves to the Inboard/Outboard MSIV and the SRV accumulators. This test is performed during periods of reactor shutdown when the drywell is accessible. During performance of this test for the MSIV's, Division 11(l) non-interruptible instrument air header is isolated. The associated MSIV accumulator drain valve is opened and depressurized. Continued air flow past the affected MSIV accumulator drain valve is verification that the associated Division I(11)

non-interruptible instrument air check valve goes to the open positio It should be noted that MSIV accumulator depressurization should take approximately one minute. After verification of check valve operation the affected air accumulator drain valve is closed and cappe Other less routine testing of the system includes measuring the containment isolation valve seat leakages, verifying the operability of other functions such as SRV lifting and ASSD functions, and post maintenance testing which restores reliability in the operability of the system/components. Most of this testing is associated with post refueling or post maintenance activities.

SD-25 Rev. 5 Page 34 of 87 4. NOTE:

Acc for FIGURE 25-6A Main Steam Line Isolation Valve Schematic Control Diagram MSIV Close Mode AC/DC Solenoids Deenergized

umulator Capacity sufficient one MSIV Operatio I I

Air Supply L_

I Air Supply TRIP Jr LOGIC TRIP LOGIC r-()1-A7I Ksl

"AC" TEST

I I

--

TO REACTOR PROTECTION SYSTEM

'Yr Open to Vent Operating Air MAIN STEAM LINE ISOLATION VALVE SCHEMATIC CONTROL DIAGRAM MSIV CLOSE MODE AC/DC SOLENOIDS DEENERGIZED NOTE:

"CS" in Schematic above is for Valve B21-F022A; for others, see table belo VALVES CS TRIP CS RE VALVES CS TRIP CS RE REFERENCE REF. LOGIC OTHER REFERENCE REF. LOGIC OTHER F022-A I-A71-SIA I-A71-S3A F028-A l-A71-S2A l-A71-S4A F022-B I-A71-SIB I-A71-S3B F028-B I-A71-S2B I-A71-S4B F022-C I-A71-SIC I-A71-S3D F028-C I-A71-S2C I-A71-S4D F022-D I-A71-S1D I-A71-$3C F028-D I-A71-S2D I-A71-S4C SD-25 Rev. 5 Page 79 of 87 P601

FIGURE 25-6C Main Steam Line Isolation Valve Schematic Control Diagram MSIV Slow Closure Test Mode Test Solenoid Energized NOTE; Accumulator Capacity sufficient for one MSIV Operation.

"Supply TRIP

--

LOGIC TRIP

"AC" TEST I'-.---TO REACTOR PROTECTION SYSTEM V

a)

0A7-1 P601 Open to Vent Operating Air MAIN STEAM LINE ISOLATION VALVE SCHEMATIC CONTROL DIAGRAM MSIV CLOSE MODE AC/DC SOLENOIDS DEENERGIZED VALVES CS TRIP CS RE VALVES CS TRIP CS RE REFERENCE REF. LOGIC OTHER REFERENCE REF. LOGIC OTHER F022-A I-A71-S1A I-A71-S3A F028-A I-A71-S2A I-A71-S4A F022-B I-A71-S1B I-A71-S3B F028-B I-A71-S2B I-A71-S4B F022-C I-A71-S1C I-A71-S3D F028-C I-A71-S2C I-A71-S4D F022-D I-A71-S1D I-A71-$3C F028-D I-A71-S2D I-A71-S4C SD-25 Rev. 5 Page 81 of 87 I

I.

-

FIGURE 25-7A MSIV Control Circuit Pool CLOSEDft Zj WHEN REE 'L-K7C K7A J

FiiS TRIP LOGIC 3-WAY CONTROL AR VALVE (DC SOL)

SIA CLOSED IN OPEN POSIT ION 3-WAY CONTROL AR VALVE IN BD MSIV 3-WAY CONTRO SF A.

uMSN'

5-2 I.

PC'S TRIP LOGIC CLOSED-K7--

-K WHEN RES*T*I-

1K15 0SIO-1 55A PUSH TO TEST 1K23 K13 & 1(21 RELAYS ENERGIZED WHEN PCIS TRIP LOGIC RESET K1S

KA231 RELAYS ENERGIZED WHEN PCIS TRIP LOGIC RESET 1(7 A/S/C/D CONTACT CLOSE WHEN GRI PCIS LOGIC FOR MSIVS RESET SD-25 Rev. 5 Page 83 of 87 PRECAUTIONS AND LIMITATIONS 3.17 Operation at power levels between 23% RTP (Unit 1) or 25% RTP (Unit 2)

and 90% RTP without a backup Main Turbine Pressure Regulator may be an unanalyzed condition. Operation at higher power levels is bounded by other transient analyses. Operation at low power levels has a large inherent margin that ensures MCPR is NOT exceeded. WHEN reactor thermal power is greater than or equal to 23% RTP (Unit 1) or 25% RTP (Unit 2) AND less than or equal to 90% RTP AND a main turbine pressure regulator is inoperable, THEN the inoperable pressure regulator must be restored to operable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . IF the pressure regulator can NOT be restored operable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , THEN a power change to less than 23%

RTP (Unit 1) or 25% RTP (Unit 2) OR greater than 90% RTP within the following four hours must be accomplished to avoid operation in an unanalyzed condition. The Nuclear Fuels (NFM&SA) Section should be notified immediately to analyze and recommend operation without a backup pressure regulator. It is important to note that IF reactor power is greater than 90% RTP WHEN a pressure regulator is found inoperable, THEN reactor power should NOT be reduced below 90% RTP. Operation in the permissible power ranges with an inoperable pressure regulator is a degraded condition which must have a time frame established for corrective actions to restore the pressure regulator to operable status. (GE SIL 614)

3.18 Unit 1 Only: Power operation of Unit 1 is limited as follows: Maximum Core Power (due to cross around relief valve capacity) Less than or equal to 2825 CMWt (2 nd Stage Reheat in Service). Less than or equal to 2752 CMWt (2 nd Stage Reheat NOT in Service). Main Generator Gross Output (Main Transformer Rating limit) MWe (Bus 10 AND Bus 1D fed from UAT) MWe (Bus 10 OR Bus 1D NOT fed from UAT) PREREQUISITES Reactor is in Mode I with Reactor Recirculation pumps above minimum spee.2 The Load Dispatcher concurs with loading plans.

IOGP-12 I

Rev. 20

Page 8 of 35

4t 1 o'0- 4-9-; PROCEDURAL STEPS Initials Power Increases Unit_-- Date/Time Started

-

-/

5. All applicable prerequisites as listed in Section 4.0 are me NOTE:

The following indications should be observed to verify proper response to increased speed demand from a recirculation pump speed controller: Recirculation pump speed increase.

Recirculation loop flow increase.

Reactor power increase NOTE:

Turbine load should be increased in accordance with 1(2)OP-26, Figure NOTE:

Procedural steps directing power increases may be performed concurrently with other steps of this procedur NOTE:

IF thermal power is increased more than 15% in one hour, THEN reactor coolant shall be sampled in accordance with TR 7.3.7.2 (ODOM Table 7.3.7-1, footnote c).

NOTE:

Process Computer Point B018 total core flow and H1P2-603 recorder 112B21-PDRIFR-R613 will read lower than Process Computer Point WTCF as the stability region is approached. Computer Point WTCF is the primary indication of total core flow and should be used for stability region complianc OGP-12 Rev. 20 Page 18 of 35 PROCEDURAL STEPS Initials 5. PERFORM Attachment 2, each 10% power change incremen R25 5. PERFORM power increases, as directed by the Unit SCO, by withdrawing control rods in accordance with 1 (2)OP-07 in the sequence designated by OGP-1 0, Attachment I and increasing recirculation flow in accordance with Reactor Engineer's recommendatio. IF Digital Feedwater Level Control System is in 1-ELEM control, THEN swap to 3-ELEM control in accordance with 1(2)OP-3. IF operating using FEEDWATER RECIRC TO CONDENSER VLV, FW-FV-177, to stabilize feedwater flow, THEN CLOSE FEEDWA TER RECIRC TO CONDENSER VLV, FW-FV-17.

WHEN FEEDWA TER RECIRC TO CONDENSER VLV, FW-FV-177, is closed, THEN CLOSE FW-FV-177 ISOL VLV, FW-VI OGP-12 Rev. 20 Page 19 of 35 CAUTION Reactor recirculation pumps should be operated in accordance with the Flow Control Operation Map. Care should be taken to avoid the regions of possible core thermal hydraulic instability, as specified in the COL Unit 1 Only: The OPRM system monitors the LPRMs foriridiation of thermal-hydraulic, instability when greater thanor equal to 25% thermal power AND less than or equal to 60% recirculation flow. This system provides alarms and auto matic trips as applicabl IF the OPRM system is ifnoprale AND operation is within Region A, THEN an immediate manual scram is required IF the OPRM systemeis i6operable AND ications of therma ui c i bility are present with operation within Region B, 5% Buffer Region, or theOPRMEnabled Region of the applicable Flow Control Operation Map, THEN an immediate manual sctram i's required Unit 2 Only: IF the Exclusion Region is entered, anutomatic reactor scram will occu iF operations r

n theMonitored Region ;or theRestricted Region, additional

'controls are required to plaper Technical Specifications (Fraction of Core,

,,Boiling Boundary3.2.3 ador, Period Basedi Dete~tion Syistem i.3.1.3ý).7 Initials PROCEDURAL STEPS PROCEDURAL STEPS Initials 5. PERFORM OPT-13.1, Reactor Recirculation Jet Pump Operability, prior to exceeding 25% reactor powe.2.7'

Unit 1 Only: WHEN reactor power is between 23%

and 28%, THEN CONFIRM APRM GAFs are less than or equal to 1.0. IF reactor power was decreased to less than 23%

(Unit 1) or 25% (Unit 2), THEN PERFORM 1(2)PT-01.11, Core Performance Parameter Check, within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reaching or exceeding 23% RTP (Unit 1) or 25% RTP (Unit 2).

NOTE:

Heater drains recirculation should be conducted such that the system will be ready for forward pumping of the heater drains when turbine load reaches 200 MW. IF secured, THEN PLACE heater drains in the recirculation mode in accordance with 1(2)OP-3.2.10 Unit 2 Only: IF recommended by the Reactor Engineer, AND the additional controls of Technical Specifications 3.2.3 and 3.3.1.3 for operating in the Restricted Region have been implemented, THEN PERFORM APRM Normal Trip Setpoint setup as follows: CONFIRM FCBB is less than or equal to 1.0 using Core Monitor edit progra.

DEPRESS the NORMALISETUP push-button on the FCTR card for each APR.

CONFIRM NORMAL/SETUP LED is yello.2.11 IF SJAE Condensate Recirculation Valve, CO-FV-49, is open, THEN THROTTLE CLOSED as necessary to maintain Condensate Pump discharge pressure between 150 psig and 190 psi.2.12 IF Condensate Booster Pump suction pressure approaches 114 psig during power increase AND it is desired to maintain HWC supply on liquid, THEN CONTROL HWC on liquid supply in accordance with 1(2)OP-5 OGP-12 I

Rev. 20 Page20of35 PROCEDURAL STEPS 5.2.13 NOTIFY radwaste to perform the following: PLACE CDDs, CFDs, and Master Flow Controllers in service as require.

PLACE Hotwell level control in feed and bleed in accordance with 1(2)OP-32, as desire.2.14 ADJUST Low Load Valve Panel Loaders at IR-TB-13 and IR-TB-14, as main turbine load increases, to increase second stage tube bundle pressure at 15 minute intervals in accordance with 1(2)OP-36, Figure.2.15 WHEN turbine load increases to between 200 MWe and 360 MWe, THEN COMMENCE forward pumping of the heater drains in accordance with 1 (2)OP-3.2.16 WHEN turbine load reaches approximately 240 MWe, THEN ENSURE HP TURB 7TH STAGE EXHAUST DRAIN VLVS MVD-MOV-CA-4131112 are close.2.17 Unit 2 Only: WHEN turbine load reaches approximately 240 MWe, THEN PERFORM the following: OBTAIN Select Rod Insert (SRI) control rod list from the Reactor Engineer.

NOTE:

The following step must be performed by a licensed person and independently verified by a Senior Reactor Operator. PLACE the SRI control rods on the SRI bus recommended by the Reactor Engineer in accordance with 20P-0.

RESET the APRM setdown.

IOGP-12 I

Rev. 20 Page 21of35 Initials

/ IV SRO PROCEDURAL STEPS 5.2.18 PRIOR to 26% RTP (760 MWT on Unit 1) or 30% RTP (767 MWT on Unit 2), CONFIRM Turbine Stop Valve/Control Valve Fast Closure Reactor SCRAM is enabled by performing the following for the applicable Unit: Unit I Only: CONFIRM TURB CV FAST CLOS/SV TRIP BYPASS (A-05, 6-7) is clea.

Unit 2 Only: CONFIRM TURB CV FAST CLOSISV/RPT TRIP BYPASS (A-05, 6-7) is clear.

NOTE:

The K9A-D relays are deenergized when they are at the stop screw.

CONFIRM relay C71A(72A)-K9A on Panel H12-P609 is deenergize.

CONFIRM relay C71A(72A)-K9C on Panel H12-P609 is deenergize.

CONFIRM relay C71A(72A)-K9B on Panel H12-P611 is deenergize.

CONFIRM relay C71A(72A)-K9D on Panel H12-P611 is deenergize OGP-12 I

Rev. 20 Page 22 of 35 NOTE:

The Turbine Stop Valve/Control Valve Fast Closure Reactor Scram MUST be enabled PRIOR to exceeding 26% RTP (Unit 1) or 30% RTP (Unit 2). This may be accomplished by annunciator and relay confirmation of automatic enabling OR by manually enabling this function by removing fuses.

Initials PROCEDURAL STEPS 5.2.32 Unit 2 Only: ENSURE APRM setdowns are reset.

NOTE:

Control rod withdrawal to the Full Out position in a sequence other than that

called for in OGP-10 shall be documented on Attachment 1.

5.2.33 Unit 1 Only: INCREASE reactor power as directed by the Unit SCO, in accordance with the Reactor Engineer's recommendation, to the most limiting of the values stated in Step 3.1.2.34 Unit 2 Only: INCREASE reactor power to 100% as directed by the Unit SCO, in accordance with the Reactor Engineer's recommendatio.2.35 WHEN unit is at 100% maximum achievable reactor power, THEN ENSURE reactor pressure is at rated pressure of 1030 psig utilizing narrow range indications (preferably Computer Point B01 5 if available).

5.2.36 CONFIRM core thermal limits are limits of Technical Specifications.

within the prescribed Date/Time Completed Performed By (Print)

Reviewed By:

Initials Unit SCO COMMENTS:

OGP-12 Rev. 20 Page 28 of 35

FIGURE 3 Page 1 of 1 Time to Make Load Change Saturated Nuclear Units

r-Co

U,

"I

10

V~

C 100

80

20

40 100 LOWER LOAD (% RATED)

EXAMPLE:

POWER CHANGE FROM 10% TO 50% SHOULD BE ACCOMPLISHED AT A RATE SUCH THAT IT WILL REQUIRE AT LEAST 20 MINUTES TO REACH 50% LOAD.

120P-26 I

Rev. 92 Page 135 of 149 1

)b

____TO MAKE CHANGE iI

/,'

STEP CHANGE PERMISSIBLE THIS SIDE OF "0" CURVE w

w D

50

30

10

0

&-&c 4 t SYSTEM OPERATION During normal full loading of the main turbine, the be routinely monitored: Operating oil pressure TO-PI-621 Turbine front standard Bearing header pressure TO-PI-623 Turbine front standard Pump suction pressure TO-PI-622 Turbine front standard TBCCW to main turb oil cooler TCC-TIC-615 Panel XU-2 Journal bearing metal temperatures TSI-TR-642 Panel XU-4 Thrust bearing metal temperatures TSI-TR-642 Panel XU-4 Turbine vibration TSI-XR-640 Panel XU-4 Condenser Vacuum OG-PR-23 Panel XU-2 following parameters should 220 to 250 psig 23 to 29 psig 20 to 30 psig 1100 to 120'F 170 to 190OF 130 to 150'F Less than 5 mils Less than or equal to 30% load, > 26" Hg Vac Greater than 30% load, > 25" Hg Vac 120P-26 Rev. 92 Page 46 of 149 I

Information Use

/

FIGURE 3 Page 1 of I Time to Make Load Change Saturated Nuclear Units

80

60

40

20

0

.

-.-

i-n r

r -r TT T

0/0

N./CD/

0/)C

0 CE) y, 0:Q I

ss

//* /*/**

-*

*'*0 (STEP CHANGE)

5

//

' /*/*TIME IN MINUTES TO MAKE CHANGE STEP CHANGE PERMISSIBLE V/000THIS SIDE. OF "0"CURVE

10

30

100 LOWER LOAD (% RATED)

EXAMPLE:

POWER CHANGE FROM 10% TO 50% SHOULD BE ACCOMPLISHED AT A RATE SUCH THAT IT WILL REQUIRE AT LEAST 20 MINUTES TO REACH 50% LOA P-26 Rev. 92 Page 135 of 149 C

100i CI

,0,

-J

I REFERENCES 2.16 GE Service and Information Letters: Serials 52; 139, Supplement 2; and 407 2.17 EER 88-0259 2.18 001-53, Rod Worth Minimizer (NUMAC-RWM) Operating Instruction 2.19 Control Rod Drive, GEI-924184B R20 2.20 INPO SER 17-92, Inadequate Control of Testing Results In An Unintended Reactor Power Transien.21 INPO SO ER 84-2, Control Rod M ispositioning 2.22 0-FP-50012 Sheet 3, Units 1 & 2 Reactor Manual Control System Elementary Diagram 2.23 2-FP-50012 Sheet 13, Reactor Manual Control Elem Diag. Unit No. 2 2.24 0-FP-50012 Sheet 14, Elem. Diag. Reactor Manual Control Sys 2.25 2-FP-50472 Sheet 2, CRD Sel. Relay IR H12-P616 Conn Diag. Unit No. 2 2.26 0-FP-50472 Sheet 3, CRD Sel. Relay IR Conn Diag. H12-P616 Units I &.27 OPT-14.2.1, Single Rod Scram Insertion Times Test PRECAUTIONS AND LIMITATIONS During a hot startup following a reactor Scram from power, extremely high rod notch worths can be encountered due to peak xenon with no moderator void.2 The reactor should NOT be operated with a stable period of less than 100 second.3 IF single notch withdrawals result in reactor periods approaching 20 seconds, the control rod(s) should be inserted to achieve a stable period of greater than or equal to 100 seconds and the rod withdrawal sequence discontinued until a thorough assessment has been performed by the Reactor Engineer and approved by the Unit SC L20P-07 Rev.68 Page5of96L PRECAUTIONS AND LIMITATIONS IF a reactor period of less than or equal to 12 seconds is reached, the reactor shall be shut down until a substantial shutdown margin is achieve IF this is done, at least ten control rods shall be fully inserted past the step in GP at which the short period was experienced. The reactor startup shall be discontinued until a thorough assessment as to the cause/recommendation to prevent recurrence has been made by the Reactor Engineer and approved by the Unit SC.5 Reactor Coolant System pressure and temperature shall meet the Technical Specification limits prior to withdrawing control rods for an approach to criticality and during critical operations thereafter (see Technical Specification 3.4.9). The Reactor Engineer should be present to monitor power flux shaping during non-routine power ascension.7 Heatup of the Reactor Coolant System with reactor heat should be coordinated with the BOP Operator to prevent the Reactor Coolant System from being heated at a faster rate than the BOP can be placed in servic.8 During reactor shutdown, plant cooldown should be coordinated with control rod drive insertion to prevent an inadvertent criticalit.9 Coupling integrity of a control rod shall be checked anytime a control rod is fully withdrawn by verifying that the rod does NOT reach the overtravel position (see Technical Specification SR 3.1.3.5).

3.10 All rod select push buttons should be deselected whenever rod movement has stabilized to minimize select switch damage from overheatin.11 Any time a control rod has been determined immovable or untrippable, determination of the shutdown margin shall be made within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (see Technical Specification 3.1.3.A).

2oP-07 Rev. 68 Page6of996 PRECAUTIONS AND LIMITATIONS 3.12 Continuous control rod withdrawal should NOT be utilized when approaching criticalit.13 The RWM shall be operable when reactor power is less than or equal to 10% rated power. IF the RWM is bypassed or a control rod is bypassed in the RWM, the Reactor Engineer should be notified prior to further rod movements and a second licensed operator or other qualified member of the technical staff shall verify that the rod sequence is correctly followed in accordance with the applicable GP (see Technical Specification 3.3.2.1.C, 3.3.2.1.D).

R21 3.14 Any deviation from the original withdrawal sequence should be recommended by the Reactor Engineer, authorized by the Unit SCO, and documented on the proper rod sequence checkoff shee.15 IF an uncoupled control rod is recoupled and coupling integrity verified, the control rod should be restricted to the "Notch Out" mode of withdraw operation. It should be removed for filter screen inspection and replacement at the next scheduled outag.16 The following Technical Specification requirements shall be observed for the Reactor Manual Control System:

3.1 Section 2.0, Safety Limits (SLs)

3.1 Section 3.1.1, Shutdown Margin (SDM)

3.1 Section 3.1.2, Reactivity Anomalies 3.1 Section 3.1.3, Control Rod Operability 3.1 Section 3.1.6, Rod Pattern Control 3.1 Section 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR)

3.1 Section 3.2.2, Minimum Critical Power Ratio (MCPR)

3.1 Section 3.3.1.1, Reactor Protection System (RPS) Instrumentation 3.1 Section 3.3.2.1, Control Rod Block Instrumentation I20P07 Rev. 68

Page 7 of 96 PRECAUTIONS AND LIMITATIONS 3.16.10 Section 3.4.9, Reactor Pressure and Temperature (P/T) Limits 3.16.11 Section 3.6.1.1, Primary Containment 3.16.12 Section 3.9.1, Refueling Equipment Interlocks 3.16.13 Section 3.9.3, Control Rod Position 3.16.14 Section 3.10.3, Single Control Rod Withdrawal - Hot Shutdown 3.16.15 Section 3.10.4, Single Control Rod Withdrawal - Cold Shutdown 3.16.16 Section 3.10.6, Multiple Control Rod Withdrawal - Refueling 3.16.17 Section 3.10.8, Shutdown Margin (SDM) Test-Refueling 3.17 The single rod scram test panel was NOT designed for multiple single rod scrams for rapid power reduction. IF used in this manner, the risk of incurring unauthorized rod patterns is greatly increased, as are potential challenges to the scram discharge volume high water level automatic scra.18 Plant parameters should be closely monitored when control rod movement is performed with the Reactor at or near rated power since such movement can cause thermal limits or rated thermal power to be exceede.19 Use of hydraulic modules "JMC" test jacks on Panel H12-P610 will disable the seal in logic for the rod drift circuitry for the associated set of rods, causing rod drift alarms, and rod drift indication on matrix display to automatically rese.20 To minimize the possibility of inadvertent control rod misposition when inserting or withdrawing a control rod to an intermediate position (notch positions '02' through '46'), and the control rod is to be moved more than one notch, the following practices SHOULD be adhered to:

NOTE:

This guidance is waived during emergency condition.

When moving a control rod four notches or more, the control rod SHOULD be stopped one notch prior to reaching the intended position and then single notched into the final intended position. This guidance does NOT supersede any other requirement to single notch control rod P-07 Rev. 68 Page8of96 FR270 PRECAUTIONS AND LIMITATIONS When moving a control rod three notches or less, the control rod SHOULD be single notched for the entire mov.21 WHEN moving control rods, wait a minimum of 3 seconds after settle function to select another rod for movement to preclude inadvertent rod movemen.0 PREREQUISITES Reactor Protection System is in operation in accordance with 20P-0.2 Control Rod Drive Hydraulic System is in operation in accordance with 20P-0.3 Neutron Monitoring System is in operation in accordance with 20P-0.4 Radiation Monitoring System is in operation in accordance with 20P-1.5 120 Volt AC UPS, Emergency and Conventional Electrical Systems are in operation in accordance with 20P-5.6 Reactor Manual Control System Electrical Lineup is complete in accordance with Attachment.7 Reactor Manual Control System Panel Lineup is complete in accordance with Attachment 2.

20P-07 I

Rev. 68 Page 9 of 96 STARTUP I

Information Continuous Control Rod Withdrawal Use 5. Initial Conditions All applicable prerequisites as listed in Section 4.0 are me.

Reactor power is greater than 25%.

OR Continuous control rod withdrawal is desired AND Reactor Engineer approval has been obtained to continuously withdraw the control ro. Procedural Steps ENSURE ROD SELECT POWER control switch is in O.

SELECT the desired control rod by depressing its CONTROL ROD SELECT push butto.

ENSURE the backlighted CONTROL ROD SELECT push button is brightly illuminated AND the white indicating light on the full core display is also illuminate.

ENSURE the ROD WITHDRAWAL PERMISSIVE indication has illuminated.

120P-07 Rev. 68

Page 10 of 96 ;

Procedural Steps CONTINUOUSLY WITHDRAW the control rod to the position designated on GP pull sheets by holding EMERGENCY ROD IN NOTCH OVERRIDE, C12A-CS-Z9-SI, to OVERRIDE, while simultaneously holding ROD MOVEMENT, C12A-CS-Z8-S1, to NOTCH OU.

MONITOR control rod position AND nuclear instrumentation while withdrawing the control rod.

NOTE:

Normal operation of the RMCS should be monitored in accordance with Section 6.0. IF the control rod fails to withdraw, THEN GO TO Sections 8.1, 8.2, 8.6, or 8.7 to free the control rod AND RETURN TO Step 5.1..

PERFORM the following for control rods to be withdrawn to an intermediate position: BEFORE the control rod reaches the position designated on GP pull sheets, RELEASE ROD MOVEMENT, C12A-CS-Z8-SI, and EMERGENCY ROD IN NOTCH OVERRIDE, C12A-CS-Z9-SI, control switche ENSURE the control rod settles into the desired positio ENSURE the rod settle light extinguishes.

2P-07I Rev. 68 Page 11 of 96 FR21]

FR21]

5. Procedural Steps NOTE:

IF the rod is uncoupled, THEN the four rod display indication will go out for the uncoupled rod AND the ROD OVER TRAVEL (A-05 4-2) annunciator will illuminate. PERFORM the following for control rods to be fully withdrawn: WHEN the control rod reaches position 48, THEN MAINTAIN the continuous withdraw signal for approximately 3 to 5 seconds OR apply a separate notch withdraw signa ENSURE the control rod does NOT retract beyond position 48. (ref. SR 3.1.3.5) RELEASE ROD MOVEMENT, C12A-CS-Z8-S1, and EMERGENCY ROD IN NOTCH OVERRIDE, C12A-CS-Z9-SI, if use ENSURE the control rod settles at position 48 AND the rod settle light extinguishe ENSURE the control rod reed switch position indicators agree with the FULL OUT indication on the full core displa.

WITHDRAW the remaining control rods, as necessary, utilizing the GP pull sheets and repeating Section 5.1 or.

IF movement of control rods is no longer required, THEN DESELECT the rods by placing ROD SELECT POWER to OFF.

120P-07 Rev. 68 Page 12 of 96 FR271 5. Procedural Steps MONITOR control rod position AND nuclear instrumentation while withdrawing the control rod.

NOTE:

Normal operation of the RMCS should be monitored in accordance with Section 6.0. IF the control rod fails to notch out, THEN GO TO Section 8.1, 8.2, 8.5, 8.6, or 8.7, to free the control rod AND RETURN TO Step 5.2..

PERFORM the following for control rods to be withdrawn to an intermediate position: WHEN the control rod reaches the position designated on GP pull sheets, THEN ENSURE the control rod settles into the desired positio ENSURE the rod settle light extinguishes.

NOTE:

IF the rod is uncoupled, THEN the four rod display indication will go out for the uncoupled rod AND the ROD OVER TRAVEL (A-05 4-2) annunciator will illuminat.

PERFORM the following for control rods to be fully withdrawn: WITHDRAW the control rod to position 48 using either single notch or continuous withdra MAINTAIN the continuous withdraw signal for approximately 3 to 5 seconds OR apply a separate notch withdraw signa ENSURE the control rod does NOT retract beyond position 48. (ref. SR 3.1.3.5)

20P-07 Rev. 68 Page 14 of 96 FR271 5. Unit 2 APP A-05 3-1 Page 1 of 2 CRD PUMP 2A LO SUCT PRESS AUTO ACTIONS If low suction pressure condition exceeds approximately 3 seconds, CRD Pump 2A will tri CAUSE Plugged suction filte.

Low condensate storage tank leve.

Improper CRD pump suction valve lineu.

Circuit malfunctio OBSERVATIONS CRD Pump 2A is of.

CRDHS flow rate decreasing tq zero (Cl2-FI-R606) Cooling water flow rate decreasing to zero (C12-FI-R605). Charging water pressure decreasing to zero (C12-FI-R601). Cooling water differential pressure decreasing to zero (C12-PDI-R603). Drive water differential pressure decreasing to zero (C12-PDI-R602). CRD Pump 2A suction pressure less than 18 inches Hg absolute (C12-PI-RO17A as read locally). CRD pump suction filter differential pressure greater than 10 psid (CO-PDIS-1490 as read locally). CST level less than 11 feet (CO-LI-I60A).

1 CRD PUMP 2B LO SUCT PRESS (A-05 5-1) alar. CRD PUMP INLET FILTER AP HIGH (A-05 6-1) alar.

CRD HYD TEMP HIGH (A-05 1-2) alar.

CRD ACCUM LO PRESS/HI LEVEL (Ar07 6-1) alar ACTIONS C

j

!Reactor, power and geertood shoulbe maintained donstafit during_

tie hed~~sCA (nob CRD: ýpdmps running)ý. Refer to OAOP-0.

If CRD pump suction filter differential pressure is greater than 10 psid, shift CRD pump suction filters and start up the CRDHS per OP-0.

If the CST level is low, perform the following steps: If available, shift the CRD suction from the CST to the Condensate and Feedwater System and startup the CRDHS per OP-0 APP-Ak-05 Rev. 43 Page 3 f9 I

Unit 2 APP A-0S 3-1 Page 2 of 2 ACTIONS (Continued) If the CRD suction was not shifted to the Condensate and Feedwater System, fill the CST per OOP-31.2 and start up the CRDHS per OP-0.

If CST level and CRD pump suction filter differential pressure are normal, verify that the CRDHS valve lineup is correct and startup the CRDHS per OP-0 DEVICE/SETPOINTS Pressure Switch Cl2-PSL-NO01A 18 inches Hg absolute POSSIBLE PLANT EFFECTS If the reactor is in operation and the CRDHS cannot be returned to operation, the CRD temperatures will increas.

If the reactor is in operation and the CRDHS cannot be returned to operation, the CRD accumulator pressures will decrease, which may require the reactor to be shutdow.

Loss of the CRDHS may result in a technical specification LC.

If reactor pressure is less than 800 psig, improper reactor Scram from control rod REFERENCES 1. LL-9364 -

73 Technical Specification 3..

OP-08, Control Rod Drive Hydraulic System OOP-31.2, Condensate and Demineralized Water Storage and Transfer System OAOP-02.0, Control Rod Malfunction/Misposition APP A-05 5-1, CRD PUMP 2B LO SUCT PRESS APP A-05 6-1, CRD PUMP INLET FILTER AP HIGH APP A-05 1-2, CRD HYD TEMP HIGH APP A-07 6-1, CRD ACCUM LO PRESS/HI LEVEL 12APP-A-05 Rev. 43

Page 35 of 93 1 SYMPTOMS Abnormal flux patter RI Control Rod found out of intended positio.3 Thermal limits approaching or exceeding Technical Specification limi.4 CRD PUMP IA(2A) LO SUCT PRESS (A-05 3-1) annunciator in alar.5 CRD PUMP IB(2B) LO SUCT PRESS (A-05 5-1) annunciator in alar.6 CRD PUMP INLET FILTER AP HIGH (A-05 6-1) annunciator in alar.7 CRD ACCUM LO PRESS/HI LEVEL (A-07 6-1) annunciator in alar.8 ROD DRIFT (A-05 3-2) annunciator in alar.9 Low control rod drive water pressure (normal pressure is reactor pressure

+ 260 psig to 275 psig) Cll(C12)-PDI-R60.10 Low cooling water pressure (normal pressure is reactor pressure + 10 psig to 26 psig) C11 (C12)-PDI-R60.11 Low charging water header pressure (955 psig @ accumulator)

C11(C12)-PI-R60.12 Failure of CRD to operate on normal withdraw or insert signa.13 Loss of one or more control rod position indicators in the four-rod group displa.14 Loss of one or more control rod indicators on the full core displa.15 Double image on control rod position indicator.0 AUTOMATIC ACTIONS Possible rod block or select block from a failed reed switch or a loss of powe.2 CRD pumps trip after a 3 second delay on low suction pressure.

IOAOP-0 Rev. 11

Page 2 of 7 OPERATOR ACTIONS Immediate Actions 3..1.2 STOP any power changes in progres IF more than one control rod is drifting, THEN INSERT a manual scra.2 Supplementary Actions 3.2.1...2.*

MONITOR core thermal parameters to keep within Tech Spec limits.

CONTACT the Reactor Engineer for further control rod movement instruction MONITOR off-gas radiation AND NOTIFY E&RC to take coolant samples if fuel element failure is suspecte IF a control rod is drifting, THEN GO TO 1(2)APP-A-05 Window 3-2 for respons IF unable to move control rods, THEN PERFORM the following: IF the operating CRD Pump has failed, THEN RESTART the CRD Hydraulic System following loss of a CRD Pump in accordance with1 (2)OP-0 IF reactor pressure is below 800 psig (e.g., during startup or shutdown evolutions), AND CRD pressure CANNOT be restored with either CRD Pump, THEN INSERT a manual reactor SCRA.

REFER to Technical Specification 3.1.5 for any control rod scram accumulator required action.

TRY to move control rods in accordance with 1(2)OP-0.

CHECK the following circuits:

-

Unit 1:

120-volt UPS:

120-volt Inst Power:

Panel V7A, CKT 12 (49' Control Bldg)

Panel lAB, CKT 2 (Cable Spread)

IOAOP-02.° I

Rev. 11 Page3of71 FRI1 Supplementary Actions

-

Unit 2:

120-volt UPS:

Panel V8A, CKT 12 (49' Control Bldg)

120-volt Inst Power:

Panel 2AB, CKT 2 (Cable Spread) RESET any electrical circuit that is trippe.

MONITOR the following CRD System parameters for possible system leakage or flow control valve failures: CRD Drive Water Pressure, C11(C12)-PDI-R60 CRD Cooling Water Pressure, C1I(C12)-PDI-R60 CRD Drive Temperature, CII(C12)-TR-R01 CRD Charging Water Header Pressure, C'I(C12)-PI-R601 CHECK CRD Pump suction and drive water filters for high differential pressur.

CHECK for air in the drive water header by venting the system in accordance with 1(2)OP-0.

IF single control rod failure is indicated, THEN CHECK the following: CRD directional control valve filters for plugging (W/R). Individual HCU power and control logi. IF a loss of control rod position has occurred, THEN PERFORM the following: OBTAIN control rod positions from the plant process computer or ERFI.

CHECK the following electrical circuits:

-

Unit 1:

120-volt UPS Panel V9A, CKT 1 (49' Control Bldg)

-

Unit 2:

120-volt UPS Panel V10A, CKT 1 (49' Control Bldg)

IOAOP-0 I Rev. 11 Page 4 of 7

/*

Unit 2 APP A-06 2-7 Page 1 of 2 APRM DOWNSCALE AUTO ACTIONS Rod withdrawal block (bypassed when reactor mode switch is not in RUN). Reactor half-Scram (if companion IRM channel is upscale or inop and the reactor mode switch is in RUN). Computer printou CAUSE APRM channel(s) indicate downscal.

Circuit malfunctio OBSERVATIONS APRM recorder(s) on RTGB Panel P603 indicates less than or equal to 4.7 on the 0 to 125 scal.

APRM channel(s) downscale (DNSC) white indicating light on RTGB Panel P603 illuminate.

ROD OUT BLOCK (A-05 2-2) alar.

The rod withdrawal permissive indicating light will be of.

The following alarms may be activated: NEUT MON SYS TRIP (A-05 4-7) alar REACTOR AUTO SCRAM SYS A (A-05 1-7) or REACTOR AUTO SCRAM SYS B (A-05 2-7) alar LPRM DOWNSCALE (A-06 1-7) alar.

LPRM downscale white indicating lights, for LPRMs associated with affected APRM channel(s),

are illuminate.

On Control Panel H12-P608, observe the following: Affected APRM indicates less than or equal to 4.7 on the 0 to 125 scal LPRM downscale amber indicating lights for LPRMs associated with affected APRM channel(s) are illuminate ACTIONS Compare affected APRM channel indication with the other APRM channel.

If the affected APRM channel indication differs from other channels or is erratic, perform the following:

NOTE:

The APRM downscale and companion IRM upscale/inop scram channels are:

Channel Instruments Al APRM A AND IRM A, OR APRM E AND IRM E A2 APRM C AND IRM C, OR APRM E AND IRM G B1 APRMBANDIRMB, F MFAND IRM F B2 APRM D AN-D IRM D, OR APRM F AN--D IRM H Refer to Technical Specification 3.3.1.1 and TRMS 3.3 for the APRM operability requirement Notify the Unit SC Bypass the affected APRM channe APP-A-06 I

Rev. 36 Page 28 of 82

Unit 2 APP A-06 2-7 Page 2 of 2 ACTIONS (Continued If the APRM downscale condition is caused by LPRM input failure, perform the following: Bypass the failed LPRMs at Control Panel H12-P60 Verify that the remaining LPRM inputs to the affected APRM channel meet the minimum requirements of Technical Specifications 3.3.1.1 and TRMS Return the affected APRM channel to service by placing the APRM bypass switch to NEUTRA.

If necessary, reset the half-Scram signa DEVICE/SETPOINTS APRM channel A-F, downscale trip unit POSSIBLE PLANT EFFECTS 4.7 on the 0 to 125 scale Reactor Scram if one APRM in each RPS trip system is downscale and their companion IRMs are upscale or inop (Run mode only). If an APRM channel is bypassed or inoperable, a technical specification LCO may resul.

APRM inoperabl REFERENCES....

LL-9364 -

FP-5851 -

Technical Specification 3.3.1.1 and TRMS APP A-05 2-2, ROD OUT BLOCK APP A-05 4-7, NEUT MON SYS TRIP APP A-05 1-7, REACTOR AUTO SCRAM SYS A APP A-05 2-7, REACTOR AUTO SCRAM SYS B APP A-06 1-7, LPRM DOWNSCALE 2APPA-06I Rev. 36 Page 29 of82

(iýxý It)

RECIRC PMP A MOTOR VIB HIGH AUTO ACTIONS NONE CAUSE Pump motor imbalanc Pump motor bearing wea Circuit malfunction.

OBSERVATIONS Recirculation Pump A motor bearing temperatures increasing (Recorder B32-R601 on Control Panel H12-P614).

ACTIONS Monitor Recirculation Pump A motor bearing temperatures on Recorder B32-R60.

Attempt to reset the alarm to determine if it was spuriou.

If motor bearing temperatures are increasing and the vibration alarm cannot be reset, notify Shift Supervisor.

DEVICE/SETPOINTS Vibration Detector B32-CO01A-XSH2833 POSSIBLE PLANT EFFECTS 3 mils Loss of Recirculation Pump A.

REFERENCES LL-9364 -

12APP-A-06 I

Rev. 36 Page 36 of 82 1.

3.

S Unit 2 APP A-06 3-3 Page 1 of 1

Unit 2 APP A-06 5-3 Page 1 of 2 OUTER SEAL LEAKAGE FLOW DETECTION HI AUTO ACTIONS NONE CAUSE Failure of Seal N (upper seal). Failure of both seal.

Closure of seal staging valv.

Circuit malfunctio OBSERVATIONS If Seal N fails: Seal N cavity pressure decrease PUMP A SEAL STAGING FLOW HIGH/LOW (A-06 6-3) alar.

If both seals fail: Both Seal Cavity N and Seal Cavity N pressures decreas PUMP A SEAL STAGING FLOW HIGH/LOW (A-06 6-3) alar Drywell pressure and temperature may increas.

If seal staging valve is closed, Seal N cavity pressure approaches Seal N cavity pressur ACTIONS Verify that Seal Injection Valve, B32-V22, is ope.

Verify that Seal Staging Valve, B32-V14, is ope.

Observe the Recirc Pump A outer seal leakage flow indicator to determine leakage rate (20'

El. Reactor Building adjacent to H21-PO09). Monitor seal cavity pressures for indication of seal failur.

Monitor seal temperatures on Recorder B32-R601 on Control Panel H12-P614 for increasing temperature.

If seal temperatures reach or exceed 2000 F, trip Recirculation Pump A and isolate the loo.

Monitor the drywell pressure and temperature to determine if increasin.

Monitor the drywell equipment drain sump pumps for frequency of operation and pump run tim.

If drywell pressure and temperature are increasing and/or the drywell equipment drain sump pump operations indicate excessive leak rates, refer to OAOP-14.0 and notify the Unit SC DEVICE/SETPOINTS Flow Switch B32-FIS-N5514A 0.5 gpm 12APP-A-06 Rev. 36 Page 63 of 82

Unit 2 APP A-06 5-3 Page 2 of 2 POSSIBLE PLANT EFFECTS.

3.

REFERENCES.

3.

Loss of Recirculation Pump Increased drywell pressure, temperature and activit Reactor Scra LL-9364 -

AOP-14.0, Abnormal Primary Containment Conditions APP A-06 6-3, PUMP A SEAL STAGING FLOW HIGH/LOW 12APP-A-06 I

Rev. 36 Page 64 of 821

I

§*

Transfer From Two Loop/Two Pump Operation To One Loop/One Pump Operation, Loss Of Running Pump.

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless denoted in the Comment Step I - Obtain a current revision of OP-02, Section Current Revision of OP-02, Section 8.1 obtained and verified via NRCS, if applicabl SAT/UNSAT*

PROMPT: Irf asked' a"OUit SCOc y uconcr that dpeed be mnis~natchedý between the 2A and

""2B Recdiro:Pumtps§by tloýe maxiu afl~,bie misiatch prior to sh'sttingdw Recirc JPumhp-2A.:'

Step 2 - Slowly raise Pump B speed while slowly lowering Pump A speed until a 20%

mismatch in speed is achieve Recirc Pump B speed 20% above Recirc Pump A spee SAT/UNSAT*

Step 3 - Place the control switch for Seal Staging Vlv B32-V14 to MAN/OPE B32-V14 switch in MAN/OPE ** CRITICAL STEP ** SAT/UNSAT*

LOT-SIM-JP-002-A08 Page 4 REV. 0

Transfer From Two Loop/Two Pump Operation To One Loop/One Pump Operation, Loss Of Running Pump.

Step 4 - Shutdown Recirc Pump 2A by placing Recirc Pump 2A control switch in RECIRC MG SET 2A control switch to STO STO ** CRITICAL STEP ** SAT/UNSAT*

Step 5 - Close Disch Bypass VIv B32-F032 B32-FO32A is full closed.

- CRITICAL STEP ** SAT/UNSAT*

Step 6 - Close Pump A Disch VIv B32-FO31A B32-FO31A is full closed.

SATIUNSAT*

Step 7 - Determine core flow using point WTC Core flow determined using point WTCF.

SAT/UNSAT*

LOT-SIM-J P-002-AO8

?~~g~~~~T:

A.~i~C4r6.....

i

~t rais oo"mh tpat Iasf3 77=77771b/h<

REV. 0 Page 5

Transfer From Two Loop/Two Pump Operation To One Loop/One Pump Operation, Loss Of Running Pump.

Step 8 - Raise speed on Recirc Pump B to raise core flow above 30.8 mlb/h Core flow indicates >30.8 mlb/hr on WTCF.

SATIUNSAT*

Step 9 - Ensure total core flow is <45 mlb/h Total core flow >30.8 but less than 45 mrb/hr.

SAT/UNSAT*

Step 10 -Within 5 minutes, open Pump A Disch VIv B32-FO31A and Disch Bypass Valve B32 F032 B32-FO32A placed to OPEN

- CRITICAL STEP ** SAT/UNSAT*

LOT-SIM-JP-002-A08 Page 6 REV. 0

A Cý

ý t /

Unit 2 APP UA-49 2-7 Page I of 2 2-AOG-DI GUARD BED TEMPERATURE HIGH AUTO ACTIONS NONE APPLICABLE CAUSES..

5.

AOG Building HVAC System malfunctio Reheater outlet temperature hig Fire in guard be Guard bed temperature element failur Circuit failure.

OBSERVATIONS High guard bed temperature indicated by 2-AOG-TI-015 (Local Control Panel H2E) or UR-152 (XU-80). High reheater outlet temperature alarm at Local Control Panel H2.

Low moisture indicated on 2-AOG-MI-013 (Local Control Panel H2E)

or UR-155 (XU-80, RED).

ACTIONS Check operation of AOG Building HVAC equipment in accordance with OP-37.8, Section.

Check operation of Reheater, 2-AOG-HT.

Bypass and isolate guard bed by opening 2-AOG-V013 and 2-AOG-V014, then close 2-AOG-V009, 2-AOG-VOO, 2-AOG-011, and 2-AOG-V01.

If guard bed temperature continues to rise, a guard bed fire is possibl Immediately purge the guard bed with N2 in accordance with OP-33, Section Notify the Control Roo.

If cause of alarm is 2-AOG-TE-015 failure, place 2-AOG-TE-016 in service in accordance with OP-3 If both sensing elements fail, monitor process gas temperature by downstream charcoal adsorber bed temperature indicator.

If an equipment or circuit malfunction exists, ensure a WR is submitted.

DEVICE/SETPOINTS 2-AOG-TSH-015 95 0 F POSSIBLE PLANT EFFECTS Purging the guard bed requires bypass and isolation of the AOG Syste.

Bypassing the AOG System may cause high stack gas radiation and may result in ODCM Required Compensatory Measures.

I2APP-UA-49 I

Rev. 9 Page 23 of 31 Bypassing the Charcoal Guard Bed R

Reference Use 8. Initial Conditions All applicable prerequisites as listed in Section are me. Procedural Steps Operating the AOG 'Charcoal Ado'riber, Sysetem with the::charcoal guard bed bypassed may carry over mois tur 6hadioactielon ived' prticuat daughterpproducts: intot te charcoa dobrtns OPEN the following: GUARD BED BYPASS VALVE, AOG-V014 GUARD BED BYPASS VALVE, AOG-VO13 CLOSE the following: GUARD BED OUTLET ISOLATION VALVE, AOG-V01I GUARD BED OUTLET ISOLATION VALVE, AOG-V012 GUARD BED INLET ISOLATION VALVE, AOG-VO09 GUARD BED INLET ISOLATION VALVE, AOG-V010 20P-33 Rev. 38 Page 20 of 53 Drying Charcoal Guard Bed with Nitrogen C

Continuous Use 8. Initial Conditions CHARCOAL GUARD BED DEW POINT, AOG-UR-155, on Panel XU-80, and GUARD BED OUTLET DEW POINT, AOG-MI-018, is greater than 40 0.

AOG Charcoal Adsorber System is shut down in accordance with Section. Procedural Steps CAUTION Operating the AOG Charcoal Adgorber ýSystem with the cha rcoal guard bed bypassed m y

cr over moisture and radioactive log lved particulate dagte rducts into'the chjarcoal ýadsorber tanks.:, CLOSE the following to isolate the charcoal guard bed: GUARD BED INLET ISOLATION VALVE, AOG-VO09 GUARD BED INLET ISOLATION VALVE, AOG-V010 GUARD BED OUTLET ISOLATION VALVE, AOG-V011 GUARD BED OUTLET ISOLATION VALVE, AOG-V012 ENSURE AOG SYS VLV CONT SEL SW, AOG-CS-3161, is in LOCAL on Panel XU-8 P-33 I

Rev. 38 Page 24 of 53

Procedural Steps OPEN the following: OFFGAS SYS OUTLET PRIMARY ISO VALVE, AOG-XCV-143 OFFGAS SYS OUTLET SECONDARY ISO VALVE, AOG-XCV-141 ENSURE AOG PURGE GAS HEATER OUTLET VALVE, AOG-NP-V99, is ope.

PLACE AOG PURGE N2 GAS HTR CONTROL SWITCH 2-AOG-CS-3638, located on Panel HOM in AOG Building 21' N-S hallway next to AOG-FI-150 rotameter, in O.

OPEN NITROGEN PURGE GAS ISOL FOR 1&2 AOG-CV-2981, using control switch located behind Panel H I.

OPEN the following: CHARCOAL ADSORBERS AOG-V039 CHARCOAL ADSORBERS AOG-V061 BYPASS VALVE, BYPASS VALVE, IF the Condenser Air Removal System is shut down, THEN CLOSE AOG SYSTEM BYPASS VALVE, AOG-HCV-10.

PLACE DISCONNECT SWITCH, 2-AOG-EHT-1-DISC-SW, located on AOG N2 PURGE GAS HEATER CONTROL PANEL on AOG roof, in O.

THROTTLE OPEN NITROGEN PURGE SUPPLY VALVE TO CHARCOAL GUARD BED D1, AOG-NP-V080, to maximize nitrogen flow, NOT to exceed 50 scfm, as indicated by OFFGAS SYS OUTLET FLOW, AOG-FI-035.

120P-33 I

Rev. 38

Page 25 of 531 8. Procedural Steps CLOSE the following: GUARD BED OUTLET ISOLATION VALVE, AOG-V011 GUARD BED OUTLET ISOLATION VALVE, AOG-V012 IF the Condenser Air Removal System is shut down, THEN CLOSE AOG SYSTEM BYPASS VALVE, AOG-HCV-10.

PLACE DISCONNECT SWITCH, 2-AOG-EHT-1-DISC-SW, located on AOG N2 PURGE GAS HEATER CONTROL PANEL on AOG roof, in O.

THROTTLE OPEN NITROGEN PURGE SUPPLY VALVE TO CHARCOAL ADSORBERS, AOG-NP-V079, until nitrogen purge flow is 50 scfm as indicated by OFFGAS SYS OUTLET FLOW, AOG-FI-035 OR NITROGEN ROTAMETER, 112 AOG-FI-150.

NOTE:

Charcoal adsorbers are being purged with nitrogen purge ga.

WHEN nitrogen purging is complete, THEN CLOSE NITROGEN PURGE SUPPLY VALVE TO CHARCOAL ADSORBERS, AOG-NP-V07.

PLACE AOG PURGE N2 GAS HTR CONTROL SWITCH 2-AOG-CS-3638, located on Panel HOM in AOG Building 21' N-S hallway next to AOG-FI-150 rotameter, in OF.

CLOSE NITROGEN PURGE GAS ISOL FOR 1&2 AOG-CV-298 P-33 I

Rev. 38 Page 30 of 53 8.,Wo L Loss 2....2 Loss 2.....3 Loss 2.....

4.

of Off-Site Power The following DC oil pumps start on low header pressure:

RFPTs Reactor Recirc. M-G Sets Main Turbine Hydrogen Seal Oil of E Bus Partial Groups 2, 3, 6, 8, and 10 isolation. Loss of Bus E2(E4) will result in a Div I and Div II Group 3 isolatio One Diesel Generator starts and energizes its respective 4160V E Bu Reactor Building HVAC isolate Standby Gas Treatment initiate of BOP Bus One or two Diesel Generators start and energize their respective 4160V E Bus(es).

Hydrogen Water Chemistry isolates on loss of Bus 1(2) The following DC oil pumps start on low header pressure:

RFPTs Reactor Recirc. M-G Sets Main Turbine Hydrogen Seal Oil OAOP-36. I Rev. 22 Page3of77 flý

-7 OPERATOR ACTIONS Immediate Actions None Supplementary Actions 3.2.1 Initial Actions Determination NOTE:

Attachments 1 (Unit 1) and 2 (Unit 2) contain a listing of instrumentation, and the associated power supply, that will be available for use in this procedure. IF ANY Diesel Generator has NOT started and loaded, THEN START the Diesel Generator AND TIE it to its respective 4160V E Bu.

DETERMINE AND PERFORM the appropriate Supplementary Actions Section from Table 1 below:

Table I DEENERGIZED BUS SUPPLEMENTARY ACTIONS SECTION Loss of Off-Site Power GO TO Section 3.2.2 (Page 6)

BOP Bus 1C GO TO Section 3.2.3 (Page 16)

BOP Bus 2C GO TO Section 3.2.4 (Page 19)

BOP Bus 1 D GO TO Section 3.2.5 (Page 22)

BOP Bus 2D GO TO Section 3.2.6 (Page 23)

E Bus El or E5 GO TO Section 3.2.7 (Page 24)

E Bus E2 or E6 GO TO Section 3.2.8 (Page 27)

E Bus E3 or E7 GO TO Section 3.2.9 (Page 29)

E Bus E4 or E8 GO TO Section 3.2.10 (Page 32) IF EOP actions require cross-tieing 4160V GO TO Section 3.2.11 (Page 35).

or 480V E Buses, THEN IOAOP-3 I Rev. 22

Page4of77 FRI

Loss of Off-Site Power NOTE:

Attachment 4 (Page 71) contains a listing of loads supplied from E Buses El through E4.

...

..-

CAUTION.

Ifoln isejsiece pwer restrict ions may prevent restarting all

systems red b

etn of the procedute. The Unit SCO must use iwhatload to roestar ldepding on existing plant condti -tios. Maiu islG rtrlad:n i385 KW r P IF while executing this procedure, ALL AC power is lost to EITHER unit, THEN BOTH units GO TO OAOP-3.

IF the SAT was lost due to loss of power on the CP&L System, THEN PLACE the AUTO RECLOSE switches in MANUAL, AND TRIP all transmission line PCB.

IF the SAT was lost due to a fault and is unavailable, AND the switchyard is energized, THEN ESTABLISH a Unit Auxiliary Transformer Backfeed in accordance with 1(2)OP-50, AND PERFORM CONCURRENTLY with this procedure.

NOTE:

Remote CST level indicators will fail downscale if off-site power is lost to both Units or Common A and B Busses do not cross-tie on loss of off-site power to one uni.

IF CST level indication is not available in the Control Room or Radwaste, THEN MONITOR CST level locally while HPCI or RCIC is running with suction from the CST, AND REPORT the level every hou.

IF only one Diesel Generator per unit is operating, AND the Motor Driven Fire Pump is operating to maintain ring header pressure, THEN PERFORM the following: At the Diesel Driven Fire Pump local control panel, PLACE the Diesel Mode Switch in the MANA or MAN B positio DEPRESS AND HOLD the MANUAL START push button until the Diesel Driven Fire Pump start OAOP-3 I Rev. 22 Page 6 of 77 m

3. Loss of Off-Site Power At the Motor Driven Fire Pump local control panel, TRIP the Motor-Driven Fire Pump by depressing the MANUAL RELEASE push butto NOTE:

Each Nuclear Service Water Pump should automatically start once power is restored to its respective 4160V E Bu.

For each 4160V E Bus that is energized, ENSURE that its associated Nuclear and Conventional Service Water Pumps are operating as appropriat NOTE:

Steps 7 through 19 can be performed concurrently or in any sequence that supports existing plant conditions and manpower availabilit NOTE:

Battery Charger operation can be verified by proper voltage indication at the RTG.

For each 480V E Bus that is energized, PERFORM the following steps on the affected unit: ENSURE the appropriate 125V DC Battery Chargers return to service by checking the voltage between 130 and 140 volt ENSURE the appropriate 24V DC Battery Chargers return to service by checking the voltage between 24 and 28 volt.

MONITOR 125VDC Batteries and remove loads from the batteries prior to battery terminal voltage reaching 105 volt.

IF any 125 VDC Battery terminal voltage reaches 105 volts, THEN REMOVE the battery from servic.

MONITOR 24 VDC Batteries and remove loads from the batteries prior to battery terminal voltage reaching 21 volt.

IF any 24 VDC Battery terminal voltage reaches 21 volts, THEN REMOVE the battery from service.

OAOP-3 I Rev. 22

Page 7 of77 3. Loss of Off-Site Power 1 START the Control Building Ventilation System on the affected unit as follows: VERIFY the Control Building Instrument Air Compressors are functioning properl NOTE:

Restarting the following Control Building Ventilation System components requires approximately 100 KW. ENSURE at least one of the following Control Room A/C and Supply Fan units is operating:

-

CTL ROOM AIC & SUPPLY FAN, 1D-CU-CB and ID-SF-C (Sub E5/MCC 1CA)

-

CTL ROOM AIC & SUPPLY FAN, 2D-CU-CB and 2D-SF-C (Sub E7/MCC 2CA)

-

CTL ROOM AIC SPARE FAN, 2E-SF-C (Sub E8/MCC 2CB) IF available, THEN START the following Battery Room Ventilation Fans as required:

-

BATTERY ROOM 1A VENT FANS, IC-EF-CB. (Sub E5/MCC ICA)

-

BATTERY ROOM 1B VENT FANS, 1B-EF-CB. (Sub E6/MCC 1CB)

-

BATTERY ROOM 2A VENT FANS, 2C-EF-CB. (Sub E7/MCC 2CA)

-

BATTERY ROOM 2B VENT FANS, 2B-EF-CB. (Sub E8/MCC 2CB)

IC-SF-CB and 1B-SF-CB and 2C-SF-CB and 2B-SF-CB and SOAOP-3 I Rev. 22

Page 8 of 77 3. Loss of Off-Site Power 1 RESTORE Drywell Cooling in accordance with the following steps:

NOTE:

Deenergized RBCCW pumps will start when power is restored by the Diesel Generators if their control switches are in AUTO and a low pressure is sensed, or if their control switches are in the ON position NOTE:

Operating each RBCCW pump will require approximately 48 K IF all three RBCCW pumps are running, THEN STOP one RBCCW pump, AND PLACE its control switch in AUT IF only one RBCCW pump is running, THEN, if available, START a second RBCCW pum NOTE:

Each Drywell Cooler will start automatically when its MCC is energized while a SCRAM signal is presen ENSURE that all available Drywell Coolers on the affected unit are operatin NOTE:

The Nuclear Service Water header supply valves to RBCCW, RBCCW HXS SW INLET VLV, SW-V103, and RBCCW HXS SW INLET VLV, SW-V106 will be closed due to a LOOP signa NOTE:

The following steps transfer RBCCW heat exchanger cooling water from the Nuclear Service Water header to the Conventional Service Water heade VERIFY that the Conventional Service Water header is availabl CLOSE the NUC HDR TO RBCCWHXS SPLY VLV, SW-V19 OPEN the CONVHDR TO RBCCWHXS SPLY VLV, SW-V14 OAOP-3 Rev. 22 Page 9 of 77 3. Loss of Off-Site Power

-

1 IF the opposite unit has power to its Instrument and Service Air Compressors, THEN OBTAIN the opposite Unit SCO's permission AND CROSS-TIE air by opening UNIT I CROSSTIE VALVE, 2-SA-V.

IF opening the service air receiver cross-tie valve adversely affects the opposite unit, THEN CLOSE UNIT I CROSSTIE VALVE, 2-SA-V.

IF Suppression Pool Cooling has been directed by the EOP AND only one Diesel Generator is available on the unit, THEN INITIATE Suppression Pool Cooling as follows:

NOTE:

Placing a loop of RHR in Suppression Pool Cooling requires approximately:

- 1100 KW without an RHRSW Booster Pum KW with an RHRSW Booster Pum CLOSE EITHER SW TO TBCCWHXS OTBD ISOL, SW-V3 (MCC 1 (2)XB), OR SW TO TBCCW HXS INBD ISOL, SW-V4 (MCC 1(2)XA).

NOTE:

RHR Service Water Booster Pumps may be started depending on power availabilit `d-posb(,dale ddagetb ýpjoing ýan'

dsupport Cotisideration shoul be given to

....

S..

....

.......

y c u * s r

..............

.

....

filli*

and ventig athese ritr to returning them seti'ice g

_!nq,

ýsyit i p esyste i, p ito 1 START RHR Service Water in accordance with 1(2)OP-4 START Suppression Pool Cooling without keepfill in accordance with 1(2)OP-1 OAOP-3 Rev. 22Page lOof 77 3. Loss of Off-Site Power 1 IF Sub E5 is deenergized, THEN SHIFT the Manual Bus Transfer (MBT) devices for the following panels to the alternate power source:

(Sub E6/DP I E6)

-

j Panel lAB (Control Building 23')

- Panel 31AB (Control Building 49')

- Panel 1AB-RX (Reactor Building 20')

1 IF Sub E7 is deenergized, THEN SHIFT the Manual Bus Transfer (MBT) devices for the following panels to the alternate power source:

(Sub E8/DP 2E8)

- Panel 2AB (Control Building 23')

- Panel 32AB (Control Building 49')

a Panel 2AB-RX (Reactor Building 20')

NOTE:

Restarting RPS will require approximately 45 K.

START RPS MG Sets A and B in accordance with 1(2)OP-0 NOTE:

Restarting CRD requires approximately 200 K.

START the CRD System in accordance with 1(2)OP-0 OAOP-3 Rev. 22 Page 11 of 77 3. Loss of Off-Site Power NOTE:

Restarting Reactor Building ventilation requires approximately 365 K NOTE:

Both RPS buses must be energized to allow restoration of Reactor Building ventilatio NOTE:

In order to open the Reactor Building isolation dampers, and to restart Reactor Building HVAC, it may be necessary to cross-tie E Buse.

START Reactor Building HVAC as follows: IF PROCESS OG VENT PIPE RAD HI-HI (UA-03 5-4)

annunciator is in alarm, AND is NOT the result of a valid high radiation signal, THEN PLACE CAC PURGE VENT ISOL OVRD, CAC-CS-5519, in OVERRID RESET the following Reactor Building Ventilation Radiation Monitors on Panel H12-P606:

-

PROCESS REACTOR BLDG VENTILATION RADIATION MONITOR "A", D12-RM-K609A

-

PROCESS REACTOR BLDG VENTILATION RADIATION MONITOR "B", D12-RM-K609B DEPRESS the following Isolation Reset Groups push buttons:

-

ISOLATION RESET GROUPS 1, Z 3, 6, 8, A71-$32

-

ISOLATION RESET GROUPS 1, 2, 3, 6, 8, A71-$33 ENSURE Instrument Air header pressure is greater than 95 psig.

IOAOP-3 I Rev. 22

Page 12 of 77i 3. Loss of Off-Site Power i.atches ar*,nglad wd, *ila~ag~the latchld~niper ass~emb*i.

ISATC a AuilaryOpratr o ESUE he latching DISPATCH an Auxiliary Operator to ENSURE the latching mechanisms are disengaged for the following valves:

-

RB VENT INBD SUPPLY ISOL VALVE, A-BF/V-R RB VENT OTBD SUPPLY ISOL VALVE, B-BFIV-R RB VENT INBD EXHAUST /SOL VALVE, C-BFIV-R RB VENT OTBD EXHAUST ISOL VALVE, D-BFIV-R OPEN RB VENT INBD ISOL VALVES, A-BFIV-RB and C-BFIV-R OPEN RB VENT OTBD /SOL VALVES, B-BFIV-RB and D-BFIV-RB.

NOTE:

Operating one (1) set of Reactor Building Ventilation Fans requires approximately 120 KW of loa START three (3) sets of Reactor Building Ventilation Fans in accordance with 1(2)OP-37.1 to maintain Reactor Building static pressure negativ IF HPCI or RCIC is not operating, THEN SHUT DOWN the running SBGT trains in accordance with 1(2)OP-1 NOTE:

Restarting RWCU requires approximately 40 K NOTE:

Both RPS buses must be energized to allow the restart of RWC.

START RWCU in accordance with 1(2)OP-1 IAOP-36.1 Rev. 22

Page 13 of 77 3. Loss of Off-Site Power NOTE:

Starting a Fuel Pool Cooling pump will require approximately 50 K.

START Fuel Pool Cooling in accordance with 1(2)OP-1.

DIRECT an Auxiliary Operator to close the Main Turbine Lube Oil Reservoir SIGHT OVERFLOW ISOLATION VALVE, LO-V23, to prevent a conditioner oil spil NOTE:

Turbine Building Sump and Radwaste Building Sump pumps will NOT be available until Off-site Power is restore.

MONITOR sump levels locally, AND IMPLEMENT appropriate actions to control level in the sump NOTE:

Transfer Switch (LJ4) is located in the Unit 2 Control Building Cable Spread Area El. 23' on the west wall behind the ERFIS Power Distribution Panels next to the Lighting & Communications UPS Inverte.

IF the chimney obstruction (stack) lights were deenergized from the Loss of Off-site Power, THEN DIRECT an Auxiliary Operator to place TRANSFER SWITCH FOR FILTER HOUSE & STACK LIGHTING (LJ4) in the AUX FEED position (Sub E8/DP E12) to reenergize the chimney obstruction (stack) light.

IF the meteorological (microwave) tower lights were deenergized from the Loss of Off-site Power, THEN DIRECT an Auxiliary Operator to the Switchyard Relay House to verify that the Microwave Tower Power Transfer Switch (1 K1) shifted to the ALT position (Sub 1SY/MCC SYA).

OAOP-3 I Rev. 22 I

Page 14 of 77 3. Facility:

Brunswick Scenario No.:

I Op-Test No.:

Title:Medium Break LOCA inside containment with Loss of Offsite Power and failure of one EDG Examiners:

Operators:

Initial Conditions: The crew assumes the shift with the plant at7Z/6 power, BOL. HPCI is OOC the #APRM failed low and is bypassed. One CRD is inoperable: stuck at position 48, Severe weather has been reported in the area. The previous shift completed stroking of all MSIVs except "B" and "C" inboards Turnover:

Night orders include direction to slow stroke the "B" and "C" inboard MSIVs (full closed) and then to increase power to 100%. The Reactor Engineer recommends using Recirculation Flow for the power increase.

Event Malf. N Event Event N Type*

Description

N/A N(BOP)

Slow stroke "B" and "C" Inboard MSIV's (full closed)

N/A R(RO)

Increase power to 100%.

(SRO)

MRD018F C (RO)

"A" CRD Pump suction filter plugge MNI037F I (RO)

2 APRM fails lo MRC007F C (RO)

"A" Recirc Pump Seal Lea MRC009F (SRO)

MCN017F I (BOP)

GO dUBedF (SRO)

do V-, V4

MEE032 M(ALL)

Loss of Offsite Power 1VC,4

' MDG002F C(BOP)

One EDG fails to start MNBO09F M (ALL)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 19

si.?

SID* ?

- co1C. ;A UMo.'4LA

j10"i

~l Jk~to Lu Op-Test No.:

Facility:

Brunswick Scenario No.:

C10*OGPM l~eak inside d~w-e1--

Scenario No.:

Event Description: Perform SlowFull Stroke Test for the "B" and "C" Inboard MSIV IThe surveillance is comDlete for the remainina MSIV's. -UJe..,*: s r,,k In, t'Q*

b,,s,.,

-.......

..

r.

.

OW Cr sl Time Position Applicant's Actions or Behavior SRO Direct the BOP to perform the Quarterly MSIV functional surveillance for the "B" and "C" Inboard MSIV' reJesA )(,(-

XXt -AXX BOP Review Precautions and Limitations. Instruct RO to monitor Rx Pressure and Main Steam Line Flow during valve strokin BOP Confirm that applicable relays are energized. _p,,

4(Cr etqtays S~~~d

V CZM61 (,o*.,,

Confirm that MSIV lights/indications are illuminate BOP Direst SSS to check relay status on Panel XXX & Panel XXX Place/Confirm applicable MSIV control switch in OPEN.SLOW TEST Take MSIV VIv Test Switch to TEST and HOL sVi

./

'laICLV ci Q Q 441

.

l ".,,) ",.Asw

'.5k C

I BOP With RO monitoring Reactor pressure and MSL flow and SSS monitoring relay status at PXXX and PXXX, the BOP will release the Test Switch immediately upon any one of the following:

1) MSL flow change observed

"

.

_____

_______

2) Reactor pressure change observe MISL. *-(.

When MSIV is completely closed OR precaution XXX is met, THEN release test switch

'I 4'

A-x Atd IA.'

... L..dI 4b Event No.:

Page 2 of 19 Op-Test No.:

Op-Test No.:

__

Scenario No.:

Event No.:

Page 3 of 19 Event Description: Perform SlowFull Stroke Test for the "B" and "C" Inboard.MSIVs. The surveillance is complete for the remaining MSIV' *

,ne ft Time Position Applicant's Actions or Behavior BOP Verify appropriate relays are energized. /M6,,

RO If required, reset the Half Scram. ( toil 4-k"s 6o-stpucdt SRO Review completed surveillance Acceptance Criteri NOTE: During the scenario, the applicant will only be required to perform/valve ____

~~As stoaeJ bae-acc 4Atcsat scfW aL u~J1atpa

__2-ottca 4.rc CacA" VioJ4 Hflbýe-L Yoo

____t.,

6e-__

i Q~aIt et 4o ge 4L4ý& VI.,,3

ýk

s Sr

___

____L~

.Iy I

II).

S it.&'aAssk La%"_

__

CI N

&br'C

Op-Test No.:

__

Scenario No.:

Event No.:

Page 4 of 19 Event Description: Increase power to 100%. Note: The next event (CRD filter clocqced),wilL be initiated when the Reactivity Manipulation has been satisfied or at the Lead Examineresjii.)

discretion

1 0,g, 0 j Metal It

%b¶SItI2LL4#

fln S tbtArtr/flo.J i

BOP I Check Precautions and Limitations in OP-2 slep - one. Monitor Turbine Power Increase per OP-2.,foM ;,

RO After power increase has been completed, complete filling out page 28 of OGP-12).

..

I A

I",

.1*%

Ir-*--

^ -f-

-L

-

.L

,.

F S

SRO Review and sign OGP-12, page 28

_____

I tauovij nut gcwc.+ AcI e rramr htý gv 1r!

all lo

,

is__

0,____

46r~e L

i7!s-0 U

4 axflt..aa 1An M Time j Position j

Applicant's Actions or Behavior SRO 1. Ensure power increase is acceptable to Load Dispatcher

t,,.ip 2.Direct the RO to increase power in accordance with OGP-12. He will specify using the Reactor Recirculation Flo *.F >/5% P

3. He will direct the CRO to monitor Turbine Operation in c/ag.A.n-accordance with OP-26, Figure RO 1) Review Precautions and Limitations and Prerequisites in OGP 1 ) Ensure steps 5.2.1 through 5.2.34 have been completed ? *V 3) Increase Recirculation Pump speed(s) and monitor Rea r

_Parameters. /MIaM s*

5,ee or" (Jba/

Powtr renJLK IL 4) When at 100% power ensure reactor pressure is 1030 psi ) Confirm core thermal limits are per TS

Op-Test No.:

__

Scenario No.:

Event No.:

Page 5 of 19 Event Description: Increase power to 100%. Note: The next event (CRD filter clogged) will be initiated when the Reactivity Manipulation has been satisfied ner at the Lead Examiner's discretion Time~ Position Applicant's Actions or Behavior

Op-Test No.:

__

Scenario No.:

Event No.:

Page 6 of 19 Event Description: "A" CRD Pump suction filter plugge Execute malfunction MDR018F per Lead Examiner directio Time Position Applicant's Actions or Behavior RO 1. Observe alarm "CRD Pump Inlet Filter DP High" alarm and monitor CRD parameter. Review Alarm Response Procedure A-05 5- _____

Ae S

~ps we-r igctaey"&

SRO 1. Direct RO to completeA c,

hacA-tans 2. Dispatch an AO to locally monitor filter DP and pump suction pressure RO 1. Observe running CRD pump trips and "CRD Pump 2A Lo Suct Press" alarm (Note: Other CRD alarms will come in as well)

2. Refer to OAOP-02-Oý Q.cjiat. !pL./,ygt,. 0-1b 0 t.

Tr.s. LLO?.

?

3. Direct the AO to shift CRD Pump suction filters (

i. n,

.sw 4.When advised (simulator operator) that the CRD suction filters have been shifted, Re-Start the 2A CRD Pump. (',,

5. Observe CRD parameters return to "normal".

6. Request the AO ensure the CRD suction filter is less than 3 PSI SRO Note shifting of CRD filters in log and initiate MR to replace filters in clogged unit.

_____________________________

__________________________________________________

I

_,'6 7 I Y

Trý

Op-Test No.:

__

Scenario No.:

Event No.:

Page 7 of 19 Event Description: "A" CRD Pump suction filter plugge Execute malfunction MDROI 8F per Lead Examiner direction.

Position Applicant's Actions or Behavior,

_____________ I ______________________

iI

____

~~~sl kr4 204 a1(~

Ia ____

-3. AcwAujqjor-dajaL&

In.

SRO_ *rs

% aCwwOV Aary A/V 4f y8 T.n4 d coil s al JA a~w-s,

___T-S

.LC.o etkw-4' (Sky-j4 dvn 6kptt LLO'\\

l~

+Lu -6a~

oc.&L

"r Scs-o.. Air pA

__

JA#

i,, i-cc (,32)

Time

Op-Test No.:

__

Scenario No.:

Event No.:

Page 8 of 19 Event Description: #,(APRM Fails Low Insert malfunction MNI037F at direction from Lead Examiner Time Position [

Applicant's Actions or Behavior RO 1 Observe "APRM Downscale" alarm and notify SRO

______ C

.....to.patien in,ARP A-06 2-7

.

.EC-,%

3. Observe other APRM channels are functional (except #1)

___________

tI.Er I0 a-j

ý

-

4. Bypass the effeeted APRM channel

-

U-

-.5

-,

I.

L1 SRO 1. Consult TS 3.3.. Ensure no other half scrams are present 3. Direct RO (or BOP) to place the-op'pr-rite APRM Channel to TRIP /C..J6,,

I

'

-

. --

--

jp

w-w RO Observe half scram OF.,

"

08d

\\J_-_

d.e.j--.

SRO 1. Ensure TS are satisfied and log information 2. Initiate MR to get APRM #g repaired IS AkUS AVflM 01 c INWC SalktC CXLIcuVnucQ?

L(ý So A

Ccb1A V~k byp...

4 ot44 ctAut CSoJSIAj

AL

+6%kf eQ o-. scAt.. cLa, j 61be

--

a.,lA you 'rý 'r-d oJ vtdlN you

. iCte Iwb-ypcas ? Mb

~t~MaA~

a&

rz-~~

i~no

-- I IL 49"

-

-' L.

Op-Test No.:

__

Scenario No.:

Event No.:

Page 9 of 19 Event Description: "A" Recirc Pump Seal Leak. Enter malfunctions MRC007F and MRC009F when directed by Lead Examine Time Position Applicant's Actions or Behavior RO

/1'. Observe changing seal DPs and notify SRO that both seals appear to have failed or are failin X. Observe "Outer Seal Leakage Flow Detection Hi" alarm and notify SRO. Note: This should be a clear indication that the seal has failed and that a discharge into containment is occurring SRO 1. Direct RO to complete actions in ARP (A-06 5.3)

2. Direct BOP to monitor containment parameters BOP Observe Drywell pressure and temperature and sump levels all increasing and advise SR SRO 1. Direct RO to reduce power in anticipitation of tripping "A" Recirc Pump. Notes: Pump should be tripped and isolated if seal temperatures reach or exceed 200 F (A-06 5-3) SRO may direct to Transfer per OP-02. If so refer to LOT-SIM-JP-002 Steps 1-10 2. Call dispatcher and advise of power reduction/possible scram XXX RO*

1. Reduce power as necessary per OGP-1 2 Note: Should insert rods per OGP-10, but may lust increase Recirc Pump "B" to max before reducing flow on "A".

(-I*, S. 'I

?) syle.lb

%a.,t.-V

SRO may 2. Trip "A" Recirc Pump before 200F seal temperatures are direct the reached. This should be done without a Turbine Trip from High BOP to Trip RPV water level (Critical Task)

and Isolate L-..Aw cr'COca?

the pump if the RO is driving rods 3. Isolate "A" Recirc Pump (Critical Task)

-

Ye- -,-p lea. Ensure Precautions and Limitations of OGP-12 are satisfied 14)1CC 0&

MQ18

%~ Op-Test No.:

__

Scenario No.:

Event No.:

Page 10 of 19 Event Description: "A" Recirc Pump Seal Leak. Enter malfunctions MRC007F and MRC009F when directed by Lead Examiner.

Time

____________

Iý Position BOP/RO Applicant's Actions or Behaviomn 1. If Containment parameters exceed EOP entry conditions advise SRO.

- 346C4b ceao%5 /,j~.64-SRO 1. Notify Operations Management and Reactor Engineer of power reductio Ensure TS are satisfied at new plant configuration T. 5. 0 3. Log events 4. Check if rey i i-r uro.: No

......

...........

__n

___'_

bo rlguirod at tI

.*,.

,t

________________

I nip'

AW sVr,(

Op-Test No.:

__

Scenario No.:

Event No.:

Page 11 :of 19 Event Description:

"A" Recirc Pump Seal Leak. Enter malfunctions MRC007F and MRC009F when directed by Lead Examine Time Position Applicant's Actions or Behavior,

Op-Test No.:

Scenario No.:

I Event No.:

Page 12 of 19 Event Description:

AOG Guard Bed Fire. Execute malfunction MCN017F when directedb.,

Lead Examiner.

Time Position Applicant's Actions or Behavior, t

BOP 1 Acknowledge receipt of "2-AOG-D1 Guard Bed Temperature High" alarm and advise SRO 2. Complete observation steps <-7-7ý"

3. Bypass and Isolate guard bed Cc ion3 of UA-49 2-7 b

4. Monitor guard bed temperature. Monitor Off Gas and Stack Radiation Monitors SRO 1.Based on hearing that the guard bed temperatures continue to rise after isolation, he should direct an AO to "purge the guard bed with N2 in accordance with.P-33, Section 8.5) i,--s 2. Notify OPS Management and HP of the problem Note: Once the order is given (to the AO) to purge the guard bed, MCN01 7F will be removed. This event will be considered complete when temperatures of the guard bed begin to decrease.

Page 12 of 19

_________

______________ I____________

I Page 12 of 19 Op-Test No.:

Scenario No.:

Event No.:

&ýVf Page 13 of 19 Op-Test No.:

__

Scenario No.:

Event No.:

Page 13 of 19 Event Description: AOG Guard Bed Fire. Execute malfunction MCN017F when directed b Lead Examine Time Position Applicant's Actions or Behavior

Op-Test No.:

__

Scenario No.:

Event Description: Loss of Offsite Power (with failure of #1 ED&teistaft). Malfunctions;,5,c MEE032 and MDG002F shall be inserted at the direction of the Lead Examiner.

5. Start Control Room and battery Room HVAC 6. Restore Drywell Cooling.

SRO 1. When advised of one rod full out he should recognize that the reactor will remain shutdown under all conditions and NOT go to

"Level/Power Control"-(G1it tej9 ?

2. When advised that one EDG did not start he should direct the BOP to try a manual start 066-M Page 14 of 19 Event No.:

Time Position Applicant's Actions or Behavior SRO 1. Direct actions of EOP-01 -RSP. __,P AOP-3r,. k 2. Direct actions of AOP-36. 1. Note: Although the loss of one EDG to start is considered a separate event in the D-1 it is combined into this event with the BOP to take additional actions for manually restarting the ED. If/when suppression pool temperature exceeds 95 F, enter EOP-02-PCC RO 1. Complete actions of EOP-01-RSP. Note: The RO should recognize that not all rods are full in and should advise the SRO 2. Monitor and control RPV level and pressure using RCIC and SRVs 3. Place RHR in suppression pool cooling 4. Start CRD per OP-08 BOP 1. Recognize/announce that one EDG did not start 2. Complete Actions P. Ensure NSW Pumps running, Start CSW Pumps to support RCC 4. Ensure battery chargers operating

Op-Test No.:

__

Scenario No.:

I Event No.:

Page 15 of 19 Event Description: Loss of Offsite Power (with failure of #1 EDG to start). Malfunctions MEE032 and MDG002F shall be inserted at the direction of the Lead Examihe Time Position Applicant's Actions or Behavior, BOP 1. Manually start the failed EDG. Note: The BOP will be

"permitted" to manually start the EDG but it will immediately Trip on differential current when he attempts to close the output breaker 2. Attempt to close breaker (insert malfunction MDG024F)

3. Recognize that the Differential Fault is serious and probably means the EDG is lost 4. Direct an AO to investigate the faul Note: Once the BOP recognizes the EDG is "gone" step into the next event (1000 GPM leak) L"...... I

¢eh

.. I Mo Cotcrusjt6

________________ I

____________________________

_______

I ____________

Op-Test No.:

Scenario No.:

Event No.:

Page 16 of 19 Event Description: Loss of Offsite Power (with failure of #1 EDG to start). Malfunctions;:

MFF032 and MDG002F shall be inserted at the direction of the Lead Examine.r.

Time Position Applicant's Actions or Behavior C[*

Op-Test No.:

Scenario No.:

Event No.:

Page 17 of 19 Event Description:

1000 GPM leak inside the drywell. Malfunction MNB009F'will:be:, -,ak :

initiated as directed by the lead examiner

'

-U"

'.,

4. Recognize/advise SRO that available high pressure sources will not be adequate to maintain RPV level above TAF ?

5. Prevent RPV injection from LPCI. -

v

-

Q C.,r.,

.ra (nra Cnr.n, it, IinoA u.n Mr iniottinn EnIIs UIre

.J I3 IIICore r.ay s II u

I

,

IJ 'LIj '

C_-7'-,

7. Depressurize RPV until Core Spray is injecting The scenario will be terminated once Core Spray flow has been established to the RPV.

)

kv Time Position Applicant's Actions or Behavior BOP 1. Observe containment parameters and identify that a leak inside containment is in progress. Note: The leak will start small and gradually increase to 1000 GPM. Initially RCIC and CRD will be able to maintain level, but level decrease will be apparent when leakage exceeds approximately 600 GPM W,'

C-

"

SRO 1. Direct actions per EOP-01-RVCP. Execute RC/L and RC/P concurrently 2. Once it is apparent that Reactor Water cannot be maintained above TAF the SRO will direct the RO to initiate a cooldown to the point that Core Spray will makeup to the Reactor. He will also direct that injection from LPCI be prevented 3. When drywell pressure or suppression pool temperature I exceeds entry conditions, he will enter OEOP-02-PCCP RO 1. Maximize RPV injection with available high pressure sources (RCIC and CRD)

(6Oo 2. Place Torus Sprays in service. [ (ftn-s :V,

3. Alternate SRVs to maintain/reduce RPV pressure (cooldown <

100 F/hr)

o._lr Qrc 1-In qglt,)ý ok A

oV

Need additional work for Scenario #.

Turnover Notes for Scenario Reactivity Plan for Scenario What ever else is necessary.

Page 1 of 18

Facility:

Brunswick Scenario No.:

Op-Test No.:

Title:Steam leak outside containment with ATWS Examiners:

Operators:

Initial Conditions: The crew assumes the shift with the plant at 100% power, Mid Life. The

"A" Condensate Booster Pump is OOC.. HPCI is "available" but OOC due to I&C maintenence being conducted for the past two hour Turnover: Night orders include direction to shift RWCU pumps.

Event Malf. N Event Event N Type*

Description

N/A N(RO)

Shift RWCU Pumps

RD005M C(RO)

Rod 18-19 Drift Out

MRW010F C(BOPand Inadvertent HPCI initiatio RO*)

Assume RO stays at RX Pni

N/A R (RO)

Reduce power in preparation for removing "A" RFP from service

MCF01OF C (BOP)

"B" Condensate Booster Pump Sheared Shaft and "B" and RFP Trip (Low Suction Pressure).

MCF036F

MCF027F C(BOP)

Manual shutdown/Trip of "A" RFP, Steam Leak from (increasing)

C(RO*)

RCIC

RCIC Manual Scram (RO)

trips upon initiation

MRP009F M(ALL)

ATWS (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Op-Test No.:

Facility:

Brunswick Scenario No.:

(N)ormal,

Op-Test No.:__ Scenario No.:

Event No.:

Page 3 of 18 Event Description:

Shift RWCU pump Time Position Applicant's Actions or Behavior SRO Direct the RO to shift RWCU pumps per 20P-14, Section 8.1 RO 1. Obtains copy of 20P-14 and Reviews Precautions and Limitation. Dispatches an AO to Ensure at least one F/D A(B) Flow Controller is in MAN (20P-14, Step 8.11.2.1).

3. Request the AO to decrease each manually controlled F/D approximately 10 GPM. (20P-14, Step 8.11.2.2)

-Observe this occurs on FI-R605A/B 4. Start the non operating RWCU Pump 5. Stop the running RWCU pump 6. Request the AO to return F/Ds to desired flow rate (70 to 170 GPM on each F/D) and control mode 7. Monitor normal operation per Section 6.0 of 20P-14-Observe system flow 140-214 GPM on FI-R605A/B

- System inlet pressure equal to reactor pressure

- F/D inlet temperature <130 F

________________

+/-

Op-Test No.:

__

Scenario No.:

Event No.:

Page 4 of 18 Event Description: Shift RWCU pump Time [Position I Applicant's Actions or Behavior

________________ ____________________________

I

4 I

4 i i I

i

__________________________________________________

J

Scenario No.:

Event No.:

Event Description:

Control Rod 18-19 drifts ou Note: Once rod is driven to 00. malfunction RD005M will be removed and rod will remain at 00. If asked to disarm 18-19 hydraulically, acknowledge the request.

Time Position Applicant's Actions or Behavior SRO 1. Direct the RO to complete actions of APP 2.Direct entry into AOP-0. Direct rod 18-19 be inserted to position 00 4. Refer to TS 3.1.3 for Control Rod Operability RO 1. Determine rod drift on 18-19 drifting ou. Attempt arrest rod using RMC. Enter and announce AOP-0. Insert rod 18-19 to position 00

.1 Page 5 of 18 Op-Test No.:

Op-Test No.:

__

Scenario No.:

Event No.:

Page 6 of 18 Event Description: Control Rod 18-19 drifts ou Note: Once rod is driven to 00, malfunction RD005M will be removed and rod will remain

=f nnl If =*I-r1 tn*,'li~rrn 1 R.IQ hvcirnaiPi.*llv. sn~knnwle~doe the. re~nue~st.

Time I Position IApplicant's Actions or Behavior

___________ If askd t

diarm18-9 hdraillvakowdnetherene_

Op-Test No.:

__

Scenario No.:

Event No.:

Page 7 of 18 Event Description: Inadvertant HPCI initiatio Note: Malfunction MRWO1OF will be inserted at direction of lead examiner Time Position Applicant's Actions or Behavior SRO 1. Direct entry into AOP-0. Direct termination of HPCI operatio. Direct reduction of reactor power as necessary to prevent Scram 4. Contact I&C to determine cause of inadvertent start Note: SRO will be told by I&C that the instrument they were working on shorted out and that HPCI will not work manually or I automaticall. Refer to 01-01.07 and determine reportability requirement Note: No reporting will be actually done as part of scenario RO 1. Diagnose and report inadvertant HPCI initiation 2. Enter and announce AOP-0. Reduce reactor power as required to prevent a reactor scram and/or to maintain reactor power< 100%

BOP 1. Verify by two independent methods that the initiation is not valid and terminate HPCI operatio Note: Once HPCI is shutdown, initiate malfunction MES016F to incapacitate HPCI for further use.

I

________________.1

Op-Test No.: __

Scenario No.:

Event No.: -3 Page 8 of 18 Event Description: Inadvertant HPCI initiatio Note: Malfunction MRWOI OF will be inserted at direction of lead examiner Time I Position IApplicant's Actions or Behavior

_________________ ____________________________

I I. I I. *

I

________________.1 ____________________________

Op-Test No.:

Scenario No.:

Event No.:

Page 9of 18 Event Description: Oil leak on "A" RF Execute malfunction MCFC27F per Lead Examiner direction.

Note: It is assumed that the first indication in the simulator of an oil leak will be "RFP A Turb Oil Tank Level HI-LO" alarm. If no action is taken, bearing oil drain temperatures are expected to increase followed by RFP Trip on Low Lube Oil Pressure.

Time Position Applicant's Actions or Behavior BOP 1. Observe alarm "RFP A Turb Oil Tank Level HI-LO" alarm and advise SR. Review Alarm Response Procedure UA-04 4-SRO 1. Direct BOP to complete ARP actions 2. Dispatch an AO to locally monitor RFP "A" lube oil reservior leve. Develop MR to get oil leak repaire. Consider actions if RFP must be shutdown/discuss with OPS IVIMGMT AO 1. Advise SRO that there is a small oil leak and that reservior level (Simulator is approximately 3" below normal)

Operator)

BOP/SRO 1. Direct the AO to transfer oil from the Lube Oil and Conditioning system to the "A" RFP reservior per OP-4. Contact Chemistry to initiate oil spill control measures (optional).

3. If/when bearing oil drain temperature increases to 225 F, Trip RFP "A" and refer to AOP-23.0

__

I

___

I

______________

Page 9 of 18 Op-Test No.:

Scenario No.:

Event No.:

Scenario No.:

Event No.:

Event Description:

Oil leak on "A" RF Execute malfunction MCF027F ner Lead Examiner direction.

Note: It is assumed that the first indication in the simulator of an oil leak will be "RFP A Turb Oil Tank Level HI-LO" alarm. If no action is taken, bearing oil drain temperatures are expected to increase followed by RFP Trip on Low Lube Oil Pressure.

Time [Position Applicant's Actions or Behavior

4

4 Page 10 of 18 Op-Test No.:

Op-Test No.:

__

Scenario No.:

Event No.:

Page 11 of 18 Event Description: Oil leak on "A" RF Execute malfunction MCF027F cer Lead Examiner direction.

Note: It is assumed that the first indication in the simulator of an oil leak will be "RFP A Turb Oil Tank Level HI-LO" alarm. If no action is taken, bearing oil drain temperatures are expected to increase followed by RFP Trip on Low Lube Oil Pressure.

Time I Positionj Applicant's Actions or Behavior

Op-Test No.: _____

Scenario No.:

Event No.:

Page 12 of 18 Op-Test No.:

__Scenario No.:

Event No.:

Page 12 of 18 Event Description: Reduce Power in preparation for removing "A" RFP from service Increase malfunction MCF027F at direction from Lead Examiner Note: The AO will report that the oil leak on the "A" RFP is getting worse and that he doesn't believe the reservior level can be maintained for more than about 20 minutes. Oil has entered the floor drain.

Time Position Applicant's Actions or Behavior SRO 1 Obtain feedback from AO on extent of oil leak (increasing)

2. Direct the RO to start reducing power to 60% per OGP-12 3. Direct the BOP to start actions to shutdown "A" RFP and to review AOP-23 "Condensate/Feedwater System Failure" 4. Notify Dispatcher and OPS Management 5. Notify Chemistry of oil spill into floor drains (optional)

RO 1. Obtain copy of OGP-12 and review Precautions and Limitations 2. Reduce power to 60% using OGP-1 Note: Following reduction to 90%, the next event will be starte Make small changes (2 to 4%) to keep recirc loop flows within 10%

- Observe recirc pump speed, loop flow, and reactor power all decrease

- Ensure not in unstable region of power/flow map BOP Start shifting control to "B" RFP in anticipitation of potential "A" RFP trip using 20P-32. BOP will complete as much of this as is possible with the existing power leve Adjust level setpoint on MSTR RFPT SP/RX LVL CTL to 190"

- Place FW-FV-V46 to OPEN

- DEPRESS ANM pushbutton and check A/M indicator in "M" (manual)

- Return level setpoint to 187"

- Lower RFP A MSC to low speed stop and trip turbine

- Ensure "A RFPT on turning gear

Scenario No.:

Event No.:

Event Descriotion: "B" Condensate Booster Pump Sheared Shaft. RFP "B" trip (low suction pressure) RCIC trip on initial start plus RCIC Steam Leak after re-start. Enter malfunction MCF01OF when directed by Lead Examiner. Malfunction MCF 036F is initiated approximately 15 seconds later to trip the "B" RFP.Malfunction MES027F will be initiated on first start and then removed for second start. Malfunction MES025F will be initiated when RO re-starts RCIC. Malfunction MRP009F will be inserted in case a manual scram (or automatic scram) is initiated. Once the "B" RFP is tripped it will not be available for the remainder of the scenario.

Notes:

1. Since "A" Condensate Booster Pump is OOC it is expected that the RO will need to rapidly reduce reactor power (per AOP-23) to avoid a SCRAM when the "B Condensate Booster Pump and "B" RFP fail. The "A" RFP may automatically trip on low lube oil pressure at some point in the even. The RCIC steam leak will be initiated after the restart (following reset from trip). This leak is intended to be small enough to NOT cause immediate automatic RCIC isolation nor to exceed 4 MR in exhaust radiatio. If/when a SCRAM occurs, this event will be performed coincident with Event #7, ATWS Time Position Applicant's Actions or Behavior BOP 1. Recognize and announce failure of "B" Condensate Booster Pump and (15 seconds later) the "B" RF. Adjust feed pump "A" flow as necessary to restore normal RPV leve. Perform actions in AOP-23

- IF level increases to RFP trip setpoint, then TRIP "A" RFPT

- Maintain hotwell level +7" and -7"

- CLOSE any open feed system recirc valves SRO 1. Direct RO to complete actions in AOP-23 (rapidly reduce recirc flow to 45%)

Notes:

1. Auto runback may occur before operator action 2. Considering the oil leak on the "A" RFP, the SRO may decide to manually SCRAM the reactor and to manually Trip the "A" RFP at this poin. Direct BOP to monitor condensate/feedwater parameters 3. Call dispatcher and advise of power reduction/possible scram 4. Conduct shift brief to discuss ramifications of losing the "A" RFP Page 10 of 18 Op-Test No.:

Op-Test No.: __

Event Description:

pressure) RCIC trin Scenario No.:

Event No.:

Page 11 of 1'

"B" Condensate Booster Pumn Sheared Shaft. RFP "B" trip (low suction on initial start olus RCIC Steam Leak after re-start. Enter malfunction

MCF01OF when directed by Lead Examiner. Malfunction MCF 036F is initiated approximately 15 seconds later to trip the "B" RFP.Malfunction MES027F will be initiated on first start and then removed for second start. Malfunction MES025F will be initiated when RO re-starts RCIC. Malfunction MRP009F will be inserted in case a manual scram (or automatic scram) is initiated. Once the "B" RFP is tripped it will not be available for the remainder of the scenario.

Notes:

1. Since "A" Condensate Booster Pump is OOC it is expected that the RO will need to rapidly reduce reactor power (per AOP-23) to avoid a SCRAM when the "B Condensate Booster Pump and "B" RFP fail. The "A" RFP may automatically trip on low lube oil pressure at some point in the even. The RCIC steam leak will be initiated after the restart (following reset from trip). This leak is intended to be small enough to NOT cause immediate automatic RCIC isolation nor to exceed 4 MR in exhaust radiatio. If/when a SCRAM occurs, this event will be performed coincident with Event #7, ATWS Time Position Applicant's Actions or Behavior RO 1. Rapidly reduce recirc flow to SPEED DEMAND of 45% to avoid reactor scra Note: Power will need to be reduced to approximately 60%.

2. Monitor power/flow and ensure PBDS is operable if in Monitored Region (Refer to 2AOP-04.0)

SRO may 3. If/when a Reactor SCRAM occurs (manual or automatic),

direct the complete actions in EOP-01 -RSP. Must recognize that a reactor BOP to Trip power is >5% and that a significant number of rods (114) have not the "A" RFP inserted (Critical Task)

Note: Event 6 will be conducted concurrently with Event 5 if/when the SCRAM occurs BOP 1.Continue to monitor "A" RFP parameters and respond to subsequent alarms as additional lube oil system failures occu NOTE: The RFP should be manually tripped before total lube oil failure occurs. Per APP UA-04 4-2 the pump must be tripped before bearing drain temperatures exceed 225. Trip the "A" RFP when directed by SRO 3. Trip the Main Turbine as part of Reactor Scram Procedure

Scenario No.:

Event No.:

Event Description:

"B" Condensate Booster Pumo Sheared Shaft. RFP "B" trip (low suction pressure) RCIC trip on initial start plus RCIC Steam Leak after re-start. Enter malfunction MCF01OF when directed by Lead Examiner. Malfunction MCF 036F is initiated approximately 15 seconds later to trip the "B" RFP.Malfunction MES027F will be initiated on first start and then removed for second start. Malfunction MES025F will be initiated when RO re-starts RCIC. Malfunction MRP009F will be inserted in case a manual scram (or automatic scram) is initiated. Once the "B" RFP is tripped it will not be available for the remainder of the scenario.

Notes:

1. Since "A" Condensate Booster Pump is OOC it is expected that the RO will need to rapidly reduce reactor power (per AOP-23) to avoid a SCRAM when the "B Condensate Booster Pump and "B" RFP fail. The "A" RFP may automatically trip on low lube oil pressure at some point in the even. The RCIC steam leak will be initiated after the restart (following reset from trip). This leak is intended to be small enough to NOT cause immediate automatic RCIC isolation nor to exceed 4 MR in exhaust radiatio. If/when a SCRAM occurs, this event will be performed coincident with Event #7, ATWS Time Position Applicant's Actions or Behavior RO 1.Once "A" RFP has tripped, manually start RCIC (using the hard card) to Maintain RPV water level as directed by SR Note: Malfunction MES027F (turbine trip) on initial star. Respond to RCIC trip

- recognize and announce trip

- run trip throttle valve closed and hold closed to reset overspeed trip-restart RCIC (MES027F removed)

Note:MES025F will be inserted (Steam Leak in RHR Room from RCIC) as soon as RCIC is re-started.

3. Respond to RCIC Steam leak per APP A-02 5-7

- Enter OEOP-03-SCCP when directed by SRO

- Confirm which leak area temperature is high (Monitors on P614)

- Before area exceeds Max Safe Temperature Manually Scram Note: EOPs will take precedence over AOP, and RCIC should not be isolated to stop the leak (since it is required for RPV makeup)

Enter EOP-03-SCCP "Secondary Containment Control" and direct actions Execute SC/T, SC/R and SC/L concurrently Page 12 of 18 Page 12 of 18 Op-Test No.:

Scenario No.:

Event No.:

Event Description: "B" Condensate Booster Pump Sheared Shaft, RFP "B" trip (low suction pressure) RCIC trip on initial start plus RCIC Steam Leak after re-start. Enter malfunction MCF01 OF when directed by Lead Examiner. Malfunction MCF 036F is initiated approximately 15 seconds later to trip the "B" RFP.Malfunction MES027F will be initiated on first start and then removed for second start. Malfunction MES025F will be initiated when RO re-starts RCIC. Malfunction MRP009F will be inserted in case a manual scram (or automatic scram) is initiated. Once the "B" RFP is tripped it will not be available for the remainder of the scenari Notes:

1. Since "A" Condensate Booster Pump is OOC it is expected that the RO will need to rapidly reduce reactor power (per AOP-23) to avoid a SCRAM when the "B Condensate Booster Pump and "B" RFP fail. The "A" RFP may automatically trip on low lube oil pressure at some point in the even. The RCIC steam leak will be initiated after the restart (following reset from trip). This leak is intended to be small enough to NOT cause immediate automatic RCIC isolation nor to exceed 4 MR in exhaust radiatio. If/when a SCRAM occurs, this event will be performed coincident with Event #7, ATWS Time I Position I

Applicant's Actions or Behavior BOP Perform actions in EOP-03-SCCP

- Monitor and Control Reactor Building Temperature, Level and Radiation

- Align Service Water to Vital header and start RHR room coolers

- Start Reactor Building HVAC fans

- Advise SRO if/when any area exceeds Max Safe Temperature Page 13 of 18 Op-Test No.:

Page 13 of 18

Scenario No.:

Event No.:

Event Description:

"B" Condensate Booster Dressurel RCIC trio on initial start plus RCIC Pump Sheared Shaft. RFP "B" trio (low suction Steam Leak after re-start. Enter malfunction MCF01OF when directed by Lead Examiner. Malfunction MCF 036F is initiated approximately 15 seconds later to trip the "B" RFP.Malfunction MES027F will be initiated on first start and then removed for second start. Malfunction MES025F will be initiated when RO re-starts RCIC. Malfunction MRP009F will be inserted in case a manual scram (or automatic scram) is initiated. Once the "B" RFP is tripped it will not be available for the remainder of the scenario.

Notes:

1. Since "A" Condensate Booster Pump is OOC it is expected that the RO will need to rapidly reduce reactor power (per AOP-23) to avoid a SCRAM when the "B Condensate Booster Pump and "B" RFP fail. The "A" RFP may automatically trip on low lube oil pressure at some point in the even. The RCIC steam leak will be initiated after the restart (following reset from trip). This leak is intended to be small enough to NOT cause immediate automatic RCIC isolation nor to exceed 4 MR in exhaust radiatio. If/when a SCRAM occurs, this event will be performed coincident with Event #7, ATWS Time I Position I

Applicant's Actions or Behavior Page 14 of 18 Page 14 of 18 Op-Test No.:

Scenario No.:

Event No.:

Event Description: ATWS Initiate malfunction MRP009F at reauest of lead examiner or upon Scra Note: The intent of the 114 rods failure is to result in Dower level >5% but less than 20%

after scram Time Position Applicant's Actions or Behavior RO If/when a Reactor SCRAM occurs, complete actions in EOP-01 RSP. Must recognize that power level is above 5% and that a significant number of rods (114) did not insert and announce to SRO (Critical Task)

-Insert manual scram, place mode switch in Shutdown

- Verify no SRVs are cycling

- Start RCIC

- Enter EOP-01-RVCP at direction from SRO

- Insert IRMs to bring on scale SRO 1. Once the SCRAM occurs must recognize that the reactor is above 5% and enter EOP-01-LP (Critical Task)

Note: It is possible the SRO will determine the reactor can not be shutdown before suppression pool temperature exceeds 110 F and elect to initiate SLC at the onset (RC/Q-10).

2. Direct RO to trip Recirc Pumps (RC/Q-7)

3. Direct the RO and BOP to intentionally lower reactor water level while leaving RCIC, SLC and CRD on. (RC/L-17)(Critical Task)

Note: Per Table 1 of LPC if RCIC isolates on steam leak, the isolation will be bypassed and RCIC restarte. Direct BOP to stabilize reactor pressure at 1050 psig using Bypass Valves (RC/P-20)

4. IF/when more than one area in Secondary Containment exceeds Max Safe Temperature, direct RO to Emergency Depressurize (SCCP-26)

BOP 1. Maintain RPV pressure using Bypass Valves as directed by SRO Page 15 of 18 Page 15 of 18 Op-Test No.:

Op-Test No.: __

Scenario No.:

Event No.:

Event Description: ATWS Initiate malfunction MRP009F at request of lead examiner or upon Scra Note: The intent of the 114 rods failure is to result in power level >5% but less than 20%

after scram Time Position Applicant's Actions or Behavior 2. Inhibit condensate/feedwater flow as necessary to lower RPV water level to +90 inches as directed by SR Note: It is expected that the "A" RFP will have been tripped (manually or automatically) by this time and that RCIC and CRD will be the only high pressure sources of reactor makeu. Monitor turbine lube oil system following turbine trip RO*

1. Initiate EOP-01-LPC RC/L, RC/P as directed by SR *Some 2. Begin Alternate Control Rod Insertion (Critical Task)

Alternate Note: The ATWS is a hydraulic lock and rod insertion will be Rod

"allowed" in order to reduce power to <5%. Any attempt at insertion manually inserting rods from the control room will be successful may be done and all rods will be able to be inserted to 0 by BOP as directed by SRO 3. Initiate SLC at direction from SRO The scenario is terminated when all rods are at 00, when Emergency Depressurization is initiated or at the discretion of the lead examiner.

Page 16 of 18 Page 16 of 18

Page 17 of 18 Op-Test No.:

__

Scenario No.:

Event No.:

Page 17 of 18 Event Description: ATWS Initiate malfunction MRP009F at request of lead examiner or upon Scra Note: The intent of the 114 rods failure is to result in Dower level >5% but less than 20%

after scram Time IPositionj Applicant's Actions or Behavior

4)

-7 C- 'e,ýeA SYSTEM OPERATION CAUTION IF Process Computer Points B074 or B075, or NUMAC LDM B21-XY-5949B is NOT available, THEN the PPC heat balance may be up to 3 CMWt-V lower than actual reactor power because cdmputer pbints only get input from RWcU filter demin flow. Raising reactor power to rated will exceed 2558 CMWT if either of the following exist:

S RWCU is in service with G31-FO44 ope RWCU filter demin(s) is in servjce with RWCU reject flowin progres During normal operation of the RWCU System, the operator should routinely observe the following Control Room parameters: System flow One demineralizer 70 - 107 gp G31-FI-R605A/B Two demineralizers 140 - 214 gp NOTE:

It is desired to obtain a maximum flowrate of 107 gpm through each I

filter as often as possible to obtain the optimum vessel exchange rate. System inlet pressure F/D inlet temperature Equal to reactor pressure.

Less than 130'F.

120P-14 Rev. 107

Page 56 of 148 1 Information Use C?

I (L--'ý

'2-SIMULATOR SETUP Initial Conditions IC Rx Pwr Core Age Event Triggers

/

, ;- L Z"

100%

MOC Event Trigger Description El Manually Initiated (Rod 18-19 Drift)

E2 Manually Initiated (Fuel Failure)

E3 Manually Initiated (Recirc Pump B Speed Control Fails)

E4 Manually Initiated (Gross Fuel Failure)

E5 Manually Initiated (Main Steam Line Break Turbine Building)

E6 Manually Initiated (LEP-02 Jumpers)

Malfunctions Event System Tag Title Value Activate Deactivate (Ramp)

Time Time E2 RP RP009F ATWS #2

0 SEC NA A

NI N1032F APRM 4 Fails Low NA 0 SEC NA A

MS MS042F Inboard MSIV D Fail To NA 0 SEC NA Close A

MS MS046F Outboard MSIV D Fail To NA 0 SEC NA Close El RD RD005M Control Rod 18-19 Withdraw NA 0 SEC NA Drift E2 NB NBO05F Fuel Failure 10%

0 SEC NA 10 MIN E4 NB NBO10F Gross Fuel Failure 100%

0 SEC NA 0 SEC E5 MS MS002F MSL Break In Turbine Bldg 1%

0 SEC NA 60 SEC E6 RP RP005F Auto Scram Defeat NA 0 SEC NA LOT-EOP-045

Rev.02

L EVENT 1 ROD OUT DRIFT Instructor Activities

"o When the crew has the watch, initiate trigger El to activate rod out drift

"o If contacted as NE, report you will check thermal limits

"o When rod 18-19 has been driven to 00, delete rod drift malfunction

"o If asked to disarm 18-19 hydraulically, acknowledge the request Plant Response o Rod 18-19 drifts full out and will respond to RMCS Operator Activities SS o Monitor crew activities and maintain overvie SCO o Conduct shift turnover shift briefin o Direct actions of APP for rod drift

"o Direct entry into AOP-0 "o Direct rod 18-19 inserted to position 00

"o Refer to Tech Spec 3.1.3 for Control rod Operability RO/BOP

"o Determine rod drift on 18-19 drifting out

"o Attempt arrest rod using RMCS o Enter and announce AOP-0 El Insert rod 18-19 to position 00 STA o As directed by the SCO/SS WCC o As directed by the SCO/SS LOT-EOP-045

Rev.02

EVENT 1 HPCI INITIATION Instructor Activities o Provide Shift Briefing sheet to the SCO o When the crew has the watch, initiate TRIGGER 1 to initiate the ol If asked as I&C to investigate, acknowledge the request o If crew response is not quick enough to prevent a scram, allow c immediate actions, then freeze discuss expected actions and ba crew to perform again.

Plant Response o HPCI Initiation causes power to rise. Power stabilizes at allowed to continue.

HPCI system rew to perform cktrack to allow

ý1 14% if injection is Operator Activities SS

[] Monitor crew activities and maintain overvie SCO E] Direct entry into AOP-0 [] Direct termination of HPCI operation

"o Direct reduction of reactor power as necessary to prevent a scram

"C Contact I&C to investigate HPCI failure El Refer to Tech Spec 3.5.1. Determine Actions D & E apply since RHR Pump B is under clearance o Refer to 01-01.07 and determine reportability requirements RO/BOP Eo Diagnose and report inadvertent HPCI initiation

[o Enter and announce AOP-0 [o Verify by two independent methods the initiation is not valid and terminate HPCI operation

[o Reduce reactor power as required to prevent a reactor scram and/or maintain reactor power<1 00%

STA o] Refer to 01-01.07 and determine WCC o] As directed by the SCO/SS reportability requirements.

LOT-AOP-104

I 9 of 14 Rev.03

7-Z/

ue C

C-ll P&*L CAROLINA POWER & LIGHT COMPANY C

C L Continuous BRUNSWICK NUCLEAR PLANT Use PLANT OPERATING MANUAL VOLUME IV GENERAL PLANT OPERATING PROCEDURE

OGP-12 R19*

OGP-12 POWER CHANGES REVISION

IGP-12 Rev. 20

Page 1 of 35

SUMMARY Revision 17 incorporates EC 46861, Extended Power Uprate setpoint changes and EC 49319, Condensate Pumps Setpoint/Auto-Start Logic Change, both for Unit Revision 16 incorporates ESR 00-00442, which modified Unit I Neutron Monitoring System and Power/Flow Map operating region Revision 15 incorporates recommended action in accordance with A/R 25864-09, to add operator actions to reduce reactor power, within appropriate limitations, to mitigate the consequences of a single reactor feed pump trip from high powe Revision 14 revised to add necessary instruction to meet the changes to the plant from ESRs 95-00080, 95-00081, and 96-00499 which implement Reactor Stability Long Term Solution: Option 1-A. Upgrade to 0AP-005 format per the writers guide.

IOAOP-2 I Rev. 17

Page 7of7 7

Rich&rd Baldw in - scenario3.w pd

..

......

.

..

....

.

Page 1 Scenario 3-ATWS with MSIV closure Description:

The crew assumes the shift with power at 90%? The previous shift initiated OGP-13 "Increasing Unit Capacity at End of Core Cycle", Section 5.1 (Bypassing Feedwater Heaters #4 and #5).

(t{)

Night orders include direction to complete a Control Rod Operability Check on rod 42-39 per OPT-1 4.1 (all rods but 42-39 were already completed) and then to increase power to 100%. A, t-LC"j

(0%

The RO completes OPT-14.1 and commences power increase to 100%.

Following power increase to 100% APRM 1 fails HI (MN1031 F). This is intended to simply create an instrument malfunction for the RO, but may be interpreted by the crew as being associated with OGP-13. The RO is expected to bypass the APRM and reset the half scra SRO will check T (%) One channel of MS Radiation fails low (MRM001 F). The crew is expected to take appropriate actions per Alarm Response Procedures and TS and may again attribute the failure to actions taken for OGP-1 py*e -n-*-

A'/,-- A-cun The SDV vent and drain valves are failed closed to provide a high level alarm on Scram Discharge Volume(s). (Malfunction MDR036F). This is intended to be a precussor to an ATWS following SCRAM from MS Line Hi Red (MNBOO5F). While a AO is sent to investigate the SDV problem (vent and drain valves fail closed) the fuel failure is increased and is readily detectable on the three remaining MS Rad Monitors as well as the Off Gas Monitor(s).

The crew may attempt to reduce power, but MS radiation will continue to increase (MRM0011 F through 13F). The increase is intended to be slow enough to allow the SRO to make the decision to manually Scram and close the MSIVs before automatic action occurs. With the feedwater heaters bypassed there are restrictions to how low the power may be reduced. This may complicate the decision and may "push" the SRO to manually scram earlier rather than late When the SCRAM occurs MRP009F will keep sufficient rods out to result in power between 5%

and 10%. The crew is expected to enter Power/Level control and may enter EOP-02-PCCP if suppression pool temperature exceeds 95 F. Once power is <1%, rod motion will be allowed by driving. Once rod motion is started the scenario is terminate Rich Bard Baldwin - scenario3.wpd

_____________

_________________.............Page

Note:

SRO(I) in RO position Event Malf. N Event Description N Type

N/A N(RO)

Control Rod Operability Check (42-39 only)

N/A R(RO)

Increase power to 100%

MRM001F I(BOP)

One channel of MS Red Monitors fails low

MNI031F I(RO)

APRM 1 Fails HI

MDR036F C(RO)

Hi level on Scram Discharge Volume

MRM0011F C(BOP)

Fuel failure causes increasing radiation levels on

,0012F and remaining MS Rad Monitors 0013F

MRP009F M(ALL)

MSIV closure on Hi Rad (may be manually initiated)

with ATWS following scram

Rich,)rd Baldwin - scenario3.wpd Page 3 Time Position Applicant's Actions or Behavior (Minutes)

T=0 All Accept shift, review panels. The shift may review OGP-13 to determine configuration/limit T=5 RO Performs operability check on 42-39 per OPT-1 "T=110 RO Commences power increase to 100% per 0GP-12. RO should coordinate with BOP to ensure Feedwater and Turbine-Generator are following the reactor power increase. SRO may consult OGP-13 and call the Reactor Engineer to ensure thermal-hydraulic limits are met with the feedwater heaters bypasse T=30 RO Observes APRM HI, consults with SRO and bypasses the APR Then he will reset the half scram. SRO will ensure applicable TS are satisfied. There may be some discussion on the possible effects from 0GP-1 T=35 RO Observes MS Rad low alarm and takes appropriate action per Alarm Response Procedure. The BOP may be used to take actions on the

"back panels" (may have to be simulated). There may be some discussion on the possible effects from OGP-1 T=40 RO Will respond to Hi SDV level alarm (or may observe the vent/drain valves going closed). If an AO is dispatched he will report that "the valve solenoids are cool to the touch and the SDV level is slowly increasing but everything else looks normal".

T=45 BOP Observes MS rad levels are increasing (or responds to alarms). The SRO may order the power reduced. However, rad levels will continue to increase (will reach scram setpoint in 5 minutes).

T=50 RO Following Scram he will observe (and announce) that all rods are NOT at 04 and power is above 5% (should be between 5% and 10%). He will continue in the Reactor Scram procedure (EOP-01-RSP) until directed to exit by the SR T=51 SRO Will exit RSP and enter EOP-01-LPC "Level/Power Control". The BOP may be used to get suppression pool cooling on. If/when the suppression pool approaches 95 F the SRO will enter OEOP-02-PCCP, "Primary Containment Control" T=55 SRO Will either direct the RO to initiate SLC (if power remains above 4%

when recirc pumps are tripped) or attempt to drive rods per EOP-01-LEP-02 (if power is <4%). RO will be inserting IRMs or using other means to determine power is <4%. BOP will be assisting RO to maintain containment parameters within limits while attempting to get rods inserted.

-__

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1 -1 __j

IR 05000325/2003301

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Person / Time
ML030840299
Site:	Brunswick (DPR-062, DPR-071, NPF-037, NPF-066)
Issue date:	09/09/2002
From:	Ernstes M Operator Licensing and Human Performance Branch
To:	Keenan J Carolina Power & Light Co
References
50-324/03-301, 50-325/03-301
Download: ML030840299 (185)
v • d • e

v t e Inspection Report - Brunswick - 2003301 (Operator Exams)
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2003 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) Security Operator Exams 301(0) 301(0) Emergency Planning Other

IR 05000325/2003301

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