ML072960427

July-August Exam 50-325, 324/2007301 Draft Simulator Scenarios (1 of 4)

ML072960427
Person / Time
Site:	Brunswick
Issue date:	01/31/2007
From:	- No Known Affiliation
To:	Office of Nuclear Reactor Regulation
References
50-324/07-301, 50-325/07-301 50-324/07-301, 50-325/07-301
Download: ML072960427 (67)

Text

Draft Submittal (Pink Paper)

SIMULATOR SCENARIOS BRUNSWICK JULY-AUG EXAM - 325,324/2007-301 DRAFT SIMULATOR SCENARIOS (4)

PROGRESS ENERGY CAROLINAS BRUNSWICK TRAINING SECTION 2007 NRC EXA BRUNSWICK JULY-AUG EXAM - 325,324/2007-301 DRAFT SIMULATOR SCENARIO 1 OF 4 2007 NRC Examination Scenario #1

SCENARIO DESCRIPTION Unit 2 is operating at maximum power, End Of Cycle with level transmitter N026B and ERFIS out of service.

A swap of RB Supply & Exhaust Fans will be required to support maintenance activities.

Following the swap of RB fans, RBCCW Pump 2C will trip and 2B RBCCW Pump will fail to Auto-Start on pressure, but will be able to be manuaUy started.

After restart of the 2B RBCCW Pump, Reactor Recirculation 2A will runback to Limiter #2. After addressing the Technical Specifications cussions with I&C, the 2A Recirculation Pump Limiter #2 signal will be reset an .r-' eturned to the pre-event level.

Reactor Instrument Penetration line break occur instrument N026B for Remote Shutdown Pan Specifications must be addressed.

. DG3 will auto start and II be unavailable. E1 and E3 crew attempts to crosstie E7 7 results in loss of level for RPV level control.

Ive will fail to auto open will be shifted to

, 2-ABRX and 32AB will be transferred olers will trip and the RHR Loop "B" can ned causing drywell temperature to rise cy depressurization (CRITICAL TASK). Following tor pressure and drywell reference leg temperature will aturation limit The onl ments (N004A, N004C, N036 and N027B) will begin to exhibit indi e leg flashing. With no valid indication of RPV level, the crew will ente looding Procedure.

ailable injection to maximum until at least 5 SRVs are open and Reactor pressur- is at least 50 psig above suppression chamber pressure (Minimum Reactor Flooding Pressure) (CRITICAL TASK). Once these conditions are established the crew will throttle flow to maintain at least the required 50 psig differential but as low as possible.

When RPV flooding conditions have been established, the scenario may be term*inated.

2007 NRC Examination Scenario #1 2

SIMULATOR SETUP Initial Conditions IC 188 Scenario #1 ENP 24 for IC 14 Rx Pwr 1000/0 Core Age EOC EVENTS Event Trigger Trigger Description Number 1 NA NA Swap RBCCW PU~,

2 1 Manual 2C RBCCW Puml ,i~:t:J~r 3 2 Manual Runback of 2A 4 NA NA 5 3 Manual ,B21-N026B fails downscale 6 4 Manual .. Differential Fault, Reactor Scram 7 NA NA 8

5 9

10 NA NB006F A MSL BREAK BEFORE FLOW o 4.00 00:05:00 5 RESTRICTOR RC026F RECIRC PMP "A" RUNBACK TO False True 00:00:05 2 LIMITER #2 2007 NRC Examination Scenario #1 3

Remotes Summary Remf ID Mult Description Current Target Rmptime Actime Trig ID Value Value EP_IACS993P DW CLR A & D OVERRIDE - NORMAL STOP 00:05:00 6 NORMAUSTOP EP_IACS994P DW CLR A & D OVERRIDE - NORMAL 00:04:00 6 NORMAUSTOP U1-U2 WET HDR X-TIE VLV v5063 10 SW-V193 MAN ISOL NSW TO RBCCW 8 CONY SW TO RBCCW HSX V146 8 RP_IARPSB RESTART RPS MG SET B 9 RP~IAEPAMGB PRS M-G SET B EPA BKRS 9 PNL 2AB PWR (E7=NORM/E8=ALT) 7 7

ED~ZIEDHXO PNL 32AB PWR (E7=; 7 2007 NRC Examination Scenario #1 4

Override Summary Trig RBCCW PMP C AUTO RBCCW PMP C AUTO RBCCW PMP C AUTO RBCCW PMP A AUTO Annunciator Summary Window Descri tion Tri NONE Batch Files File Trigger Special Instructions 2007 NRC Examination Scenario #1 5

SHIFT BRIEFING Plant Status The plant is operating at maximum power, End of Cycle.

Equipment Out of Service No equipment is out of service Plan of the Day Maintain current power.

Following shift turnover, Place the 20 RB S secure 2C Fans. Maintenance personne clearance is required.

2007 NRC Examination Scenario #1 6

SCENARIO INFORMATION Examiner Notes Procedures Used in Scenarios:

EVENT 1 EVENT 2

• Annunciator 2-UA-3, Window 2-5 (RBCCW Pump Trip)

• OAOP-16 (RBCCW Pump Trip)

EVENT 3/4

• Annunciator 2-A-6, Window 3-2 (Recirc Pum,;,

EVENT 5 OR L CONTROL PROCEDURE)

• Y CONTAINMENT CONTROL PROCEDURE)

EVENT 10

• 2EOP-01-RV ACTOR VESSEL CONTROL PROCEDURE)

• OEOP-01-RXFP (REACTOR FLOODING PROCEDURE)

Critical Tasks Perform emergency depressurization when drywell average temperature cannot be restored and maintained below 300°F.

Establish and maintain RPV pressure at least 50 psig above suppression chamber pressure with at least 5 SRVs open.

2007 NRC Examination Scenario #1 7

EVENT 1 SHIFT TURNOVER, SWAPPING OF RB SUPPLY & EXHAUST FANS The crew swaps RB HVAC Fans per SCQ direction Malfunctions required - None Objectives:

sea - Directs BOP to shift from the 2C to the 2D RB Suppl haust Fans to support Maintenance BOP - Starts 20 RB HVAC Fans in service & remove 37.1 Section 8.9 Success Path:

RB HVAC Supply & Exhaust Fans fans 2A, B, D Simulator Operator Activities:

• ~!!ator, report that pre-ted

• When aske* t the fans appear to be operating normally.

2007 NRC Examination Scenario #1 8

EVENT 1 SHIFT TURNOVER/SWAPPING OF RB SUPPLY & EXHAUST FANS Required Operator Actions SRO Normal 0 eration - Swa of RB Su

• Directs BOP to shift from the 2C to the 2D RB Sup Maintenance BOP Normal 0 eration -Swa

• Starts 2D RB HVAC Fans in service &

37.1 Section 8.9 APPLICANT'S ACTIONS OR BEH~':"

2007 NRC Examination Scenario #1 9

EVENT 2 RBCCW PUMP TRIP/PUMP in AUTO FAILS TO AUTO START The crew responds to a trip on one of the operating RBCCW Pumps Malfunctions required:

• 2B RBCCW Pump will fail to auto-start on a low RBCC pressure Objectives:

sea - Directs BOP to enter and execute OAOP-1 BOP - Enters OAOP-16.0 to respond to the.

start Success Path:

28 RBCCW Pump is manually start ' witch to ON) and RBCCW is returned to normal operation (norm e 2C RBCCW Pump motor is hot to report that the power supply breaker for 2C agnetics.

report that the power supply breaker for 2C RBCCW agnetics.

2007 NRC Examination Scenario #1 10

EVENT 2 RBCCW PUMP TRIP Required Operator Actions Transient Response - Entry into OAOP-16.0: Failure of RBCCW System SRO

• Directs BOP operator to enter and execute OA

• Directs start of 2B RBCCW pump BOP

• Refers to Annunciator Respo

• Enter OAOP-16

• Manually starts 2B R 2007 NRC Examination Scenario #1 11

EVENT 3 2A REACTOR RECIRCULATION PUMP RUNBACK TO LIMITER #2 The crew responds to a spurious runback of the 2A Recirculation Pump speed controller to Speed Limiter #2. The scoop tube lockup circuit is defeated. (can not be locked up)

Malfunction required:

• 2A Reactor Recirlation MG Set will experience a spuri . nback signal to the Limiter #2 setpoint

• Lockout circuit on scoop tube will be defeated (

Objectives:

SCQ Directs RO to enter and execute 2A rious, direct reset of the urn of power to the pre-RO eactor Recirculation MG runback nback signal per 20P-02, Section 8.3 The runba 2A Reactor Recirculation Pump is identified, 2AOP-04.0 is entered an er level is stabilized, Technical Specification requirements are evaluated f circulation Loop operation, and the Runback Signal is successfully reset P-02.0 following resolution.

Simulator Operator Activities:

WHEN directed by the lead examiner, activate TRIGGER 2.

WHEN asked, as the Turbine Building Auxiliary Operator (TBAO), wait 3 minutes and report that there are no apparent problems at the 2A Reactor Recirculation Motor Generator.

2007 NRC Examination Scenario #1 12

Simulator Operator Activities (continued)

WHEN asked, as I&C, wait 5 minutes and report that the cause of the Runback to Limiter #2 was due to an error in installation of a jumper during the performance of a surveillance currently in progress.

WHEN asked, as I&C, report that the lockout circuit has been r retested next shift. The runback may be reset.

WHEN asked, as I&C, communicate with the Reactor balance/verify the demand vs. actual speed control signal, in support of e runback

[Instructor Aids - Panels - Recirc MG Set Bail.. '

WHEN asked, as NE, provide guidance on pattern does not need to be symmetrical.

2007 NRC Examination Scenario #1 13

EVENT 3 2A REACTOR RECIRCULATION PUMP RUNBACK TO LIMITER #2 Required Operator Actions Transient response - Entry into 2AOP-04.0: Low Core Flow SRO

• Directs RO to enter and execute 2AOP-04.0:

• Contacts I&C for support

• Evaluates Technical Specification (T.S. 3.4 or single Recirculation Loop operation (flow mism, of the loop with the lower speed Recirculaton Pump a

• Following determination of the Runbac purious, direct reset of the 2A Reactor Recirculation Pump Limit I to return of power to the pre-event power level

• Contacts Nuclear Eng. for gui RO eactor Recirculation MG runback nback signal per 20P-02, Section 8.3 2007 NRC Examination Scenario #1 14

EVENT 4 INCREASING POWER FOLLOWING RUNBACK SIGNAL RESET The crew will take action to restore reactor power to the pre-runback level.

Malfunctions required:

None Objectives:

seo Directs RO to raise reactor power to 90% by rais*,

Loop RO Raises reactor power per 20P-02.0 by r .

Success Path:

Simulator Operator Activities:

be raised without ramp 2007 NRC Examination Scenario #1 15

EVENT 4 INCREASING POWER FOLLOWING RUNBACK SIGNAL RESET Required Operator Actions Normal Operating Procedures - 20P-02 SRO

• Contacts Load dispatcher regarding power incr

• Direct RO to raise reactor power to 900/0 RO o When directed, raises reacto Reactor Recirculation Flow per 20P-02. .

APPLICANT'S ACTIONS OR BE 2007 NRC Examination Scenario #1 16

EVENT 5 INSTRUMENT LINE PENETRATION FAILS The crew will observe and report the parameter changes impacted by the instrument failure. The SCO will diagnose the failure and evaluate the impact to plant operation, including Technical Specification action statement(s).

Malfunctions required:

• Penetration X49A Level instrument B21-NO Objectives:

Success Path:

SCO correctly evaluate correct Technical Sp 2007 NRC Examination Scenario #1 17

EVENT 5 INSTRUMENT LINE PENETRATION FAILS Required Operator Actions:

seQ

• Evaluate the plant impact and Technical Specifica . quirements for the instruments affected. (3.3.3.2 - Remote Shutdo nitoring Instrumentation- 30 days)

APPLICANT'S ACTIONS OR BEHAVIOR:

2007 NRC Examination Scenario #1 18

EVENT 6 REACTOR SCRAM, LOSS OF OFF-SITE POWER, DG #3 FAILURE The crew will respond to a loss of off-site power resulting in a reactor scram and a corresponding failure of Emergency Diesel Generator #3 due to an electrical fault on Bus E3.

Malfunctions required:

• Off-site power will be lost due to a grid disturba ulting in, a reactor scram. Immediately following the starting an nization of #3 Emergency Diesel Generator, an electrical op on Emergency Bus E3, resulting in a tripping of #3 Die availability of the Emergency Bus.

Objectives:

It parameters re band of 800-1000 psig nd of 170" to 200" AOP-36.1 in response to the loss of off-

• Iure er to determine Distribution Grid status RO am and takes actions to control level and pressure ds using HPCI, RCle, and SRVs BOP Plant status ite power ded to Bus E4 o ripped due to overcurrent trip on bus Places CB auto-reclosers to OFF Directs AO to cross-tie air Ensures 2B Nuclear Service Water Pump is running, starts 28 and 2C Conventional Pumps Starts Control Room and Battery Room HVAC Direct Transfer of RBCCW Cooling to Conventional Service Water Start 2B CRD Pump Direct RPS to be restarted 2007 NRC Examination Scenario #1 19

Success Path:

sea successfully enters 2EOP-01, 2EOP-01-RSP, Reactor Scram Procedure and, subsequently, enters 2EOP-01-RVCP, Reactor Vessel Control Procedure and directs activities relating to reactor vessel control (RPV pressure and level) and directs activities relating to the loss of electrical power. RO takes actions to control reactor level and pressure (HPCI, RCIC, SRV operation). BOP enters OP-36.1 and takes actions as directed to address the loss of electrical power.

7 2007 NRC Examination Scenario #1 20

EVENT 6 REACTOR SCRAM, LOSS OF OFF-SITE POWER, DG #3 FAILURE Simulator Operator Activities WHEN directed by lead examiner, activate TRIGGER 4 IF contacted as Load Dispatcher, report that there has be damage to the Transmission Grid and that there is, currently, not a pro* e for return to service.

IF asked as Unit 1 for permission to cross-tie air, WHEN it is requested to cross tie air, report t open.

IF asked to transfer RCC to CSW, wait 5 minute IF asked to restart RPS, wait 3 min IF requested to transfer 2AB, 2AB-R ait 2 minutes and activate TRIGGER 7.

IF asked by I&C to* inutes and report the EDG trip is due to an electrical lock 2007 NRC Examination Scenario #1 21

EVENT 6 REACTOR SCRAM, LOSS OF OFF-SITE POWER, DG #3 FAILURE Required Operator Actions SRO

• Executes the Emergency Operating Procedures: 2EOP-01-RSP, Reactor Scram Procedure and, subsequently, e EOP-01-RVCP, Reactor Vessel Control Procedure (if Rx Pre o psig or Rx Level 100") .

• Enters and executes EOP-PCCP wh degrees.

• Directs the RO to control reactor an

• Control pressure using

• Directs BOP t*

power and etermine Distribution Grid status o Conventional Service Water RO

• ram the reactor per 2EOP-01-RSP, Reactor Scram s to control Reactor Vessel Level and Pressure using HPCI, afety Relief Valves BOP ReportsElectric Plant status

-Loss of off-site power

-DG #4 loaded to Bus E4 2007 NRC Examination Scenario #1 22

-DG #3 tripped due to overcurrent trip on bus Places PCB auto-reclosers to OFF Directs AO to cross-tie air Ensures 2B Nuclear Service Water Pump is running, sta Conventional Pumps Starts Control Room and Battery Room HVAC Transfers of RBCCW Cooling to Conventio Starts 2B CRD Pump Execute applicable steps of OAOP-36. the loss of a - ite power and failure of the #3 Emergency Diesel G < nd Emergency Bus E3 APPLICANT'S ACTIONS OR BE 2007 NRC Examination Scenario #1 23

EVENT 7 HPCIINJECTION VALVE FAILS TO OPEN The crew will respond to a failure of the HPCI injection valve to fail to automatically open on a valid initiation signal.

Malfunctions required:

• The 2-E41-F006, HPCI Injection Valve, will fail to atically open on a valid initiation signal Objectives:

SCO/RO Identifies that the HPCI Injection initiation signal RO Opens the HPCI Injection Valve position and commences injection Success Path:

rator action to open the 2007 NRC Examination Scenario #1 24

EVENT 7 HPCIINJECTION VALVE FAILS TO OPEN Required Operator Actions:

SRO/RO Identifies the HPCllnjection Valve, 2-E41-F006, has failed to open on a valid initiation signal RO Opens the 2-E41-F006 by taking the control s* 0 the "OPEN" position and establishing injection flow to the reactor PCI.

APPLICANT'S ACTIONS OR BEHAVIOR:

2007 NRC Examination Scenario #1 25

EVENT 8/9 STEAM LEAK IN DRYWELL REQUIRING EMERGENCY DEPRESSURIZATION The crew will respond to a steam leak in the drywell in conjunction with a failure of the ability to spray the containment, subsequently leading to a requirement to Emergency Depressurize the reactor due to high drywell temperature.

Malfunctions required:

• A steam leak will occur in the drywell, resulti temperatures

• A failure will be inserted, preventing the Spray injection valve

• Cross tie of E7 to E8 will not functi Objectives:

SCO Recognize condition temperatures and pre Direct execution of app Containment Control Pr at the Reactor Saturation n during the depressurization in control of Reactor Water Level and Pressure, as e

trip :C when High level trip setpoint is reached (206")

, of RH in Suppression Pool Cooling when directed "8" Loop in Suppression Chamber and Drywell Sprays when directed eport failure of "8" Loop Drywell Spray valve to open and s to attempt to open the valve When Irected, Emergency Depressurize the reactor (high drywell temperature)

Control injection from Low Pressure systems to maintain reactor water level during depressurization.

Success Path:

When 300°F is exceeded in the drywell, the reactor is Emergency Depressurized and level is restored/maintained in the normal band (170" to 200")

2007 NRC Examination Scenario #1 26

EVENT 8/9 STEAM LEAK IN DRYWELL REQUIRING EMERGENCY DEPRESSURIZATION Simulator Operator Activities:

WHEN directed by the lead examiner, activate TRIGGER 5 IF asked, report that the breaker for the 2-E11-F016B apR indication and that the thermal overload appears to be .

IF asked, report that the 2-E11-F016B is mechani.

WHEN to lock-out drywell coolers, activate T IF requested to support E7 to E8 cross-tie, ac 2007 NRC Examination Scenario #1 27

EVENT 8/9 STEAM LEAK IN DRYWELL REQUIRING EMERGENCY DEPRESSURIZATION Required Operator Action:

SRO

• Recognize conditions of the steam leak in the drywell ted temperatures and pressures) and provide direction to the RQ an

• Direct execution of applicable steps of 2EQP-0 Control Procedure).

• When drywell temperature cannot be Depressurization of the reactor RO/BOP

• Continue to main by the SCQ

• Must man I trip setpoint is reached (206")

001 Cooling when directed in Suppression Chamber and Drywell Sprays per SEP-

• ure of "8" Loop Drywell Spray valve to open and take pen the valve

• Control i Low Pressure systems to maintain reactor water level during depr zation.

• When directed, Emergency Depressurize the Reactor by opening 7 ADS valves

• Control Low Pressure Injection Systems to prevent Reactor Vessel overfeed on re-flood following Emergency Depressurization 2007 NRC Examination Scenario #1 28

APPLICANT'S ACTIONS OR BEHAVIOR:

2007 NRC Examination Scenario #1 29

EVENT 10 LEVEL INSTRUMENT FAILURE DUE TO REFERENCE LEG FLASHING

- REACTOR FLOODING REQUIRED The crew will respond to indications of Reactor Pressure Vessel Level reference leg flashing, resulting in a loss of all level instrumentation.

Objectives:

SCO As the eactor depressurizes, recognize indicati reference leg flashing and it impact, being no level instru n being available Enter and execute EOP-01-RxFP,React direction to the RO/BOP operators RO/BOP Implement directions given by Flooding Conditions to ensur Success Path:

Reactor Pressure Ve established resulting in at least 5 safety relief val ith at least 50 psid (but as low as possible) betwee ssion Chamber Pressure.

2007 NRC Examination Scenario #1 30

EVENT 10 LEVEL INSTRUMENT FAILURE DUE TO REFERENCE LEG FLASHING

- REACTOR FLOODING REQUIRED Required Operator Actions:

EOP Action - Entry into and Execution of Reactor Flooding Procedure seQ

• As the reactor depressurizes, correctly evaluate in . s to determine Level Instrument Reference Leg Flashing is occurring

• Enter and execute EQP-01-RxFP,Reactor adequate core cooling RO/BOP

• Observe and report indications of Reacto flashing ing conditions (5 or more ssion pool and reactor.)

2007 NRC Examination Scenario #1 31

Simulator Operator Activities:

WHEN directed by the lead examiner, place the simulator in FREEZE.

CAUTION DO NOT RESET THE SIMULATO TO RECEIPT OF CONCURRENCE TO DO LEAD EXAM 2007 NRC Examination Scenario #1 32

8.9 Swapping Reactor Building Ventilation Fans R Reference Use 8.9.1 Initial Conditions

1. Reactor Building Ventilation System is in service in D accordance with Section 5.1 or 8.1.

8.9.2 Procedural Steps

1. PERFORM the following to swap a Reactor Building Exhaust Fan:

a. PLACE the selected fan control switch in START D AND HOLD.

b. ENSURE the selected fan discharge damper D opens.

c. RELEASE the selected fan control switch. D

d. ENSURE the selected fan is running by observing D the control switch red fan light is on.

e. PLACE the selected fan control switch in STOP. D

2. PERFORM the following to swap a Reactor Building Supply Fan:

a. PLACE the selected fan control switch in STOP. D

b. PLACE the selected fan control switch in START D AND HOLD.

c. ENSURE the selected fan discharge damper D opens.

d. RELEASE the selected fan control switch. D

e. ENSURE the selected fan is running by observing D the control switch red fan light is on.

120P-37.1 Rev. 50 Page 33 of 551

8.9.2 Procedural Steps

3. ENSURE REACTOR BLDG NEG PRESSURE, D VA-PI-1297, at a minimum of 0.25 inches of water.

4. ENSURE MSIV PIT EXHAUST AIR CHECK DAMPER, D VA-2A-CV-RB, did NOT close.

\ZOP-37.1 Rev. 50 Page 34 of 551

BRUNSWICK NUCLEAR PLANT PLANT OPERATING MANUAL VOLUME XXI ABNORMAL OPERATING PROCEDURE UNIT o

OAOP-16.0 RBCCW SYSTEM FAILURE REVISION 18 IOAOP-16.0 Rev. 18 Page 1 of 11 I

1aO SYMPTOMS 181 RBCCW PUMP DISCH HEADER PRESS LOW (UA-03 2-5) in alarm 1.2 RBCCW HEAD TANK LEVEL HI/LO (UA-03 1-5) in alarm 1.3 PUMP A SEAL CLOSED CLG WTR FLOW LO (A-06 1-4) in alarm 1.4 PUMP B SEAL CLOSED CLG WTR FLOW LOW (A-O? 6-5) in alarm IRISI 1.5 RBCCW HX OUTLET HDR TEMP HI (UA-03 1-3) in alarm 1.6 UNIT 2 Only: DRYWELL CHILLER TRIP (UA-05 5-10) in alarm 1.? High temperature alarms on equipment supplied by RBCCW.

IRISI 1.8 High NSW or CSW header pressure approaching pump shutoff head (approximately 90 psig).

280 AUTOMATIC ACTIONS 2.1 IF system pressure decreases to 65 psig, THEN the standby RBCCW pump will start.

D 2.2 IF non-regenerative heat exchanger outlet temperature increases to greater than 135°F, THEN RWCU will isolate.

D 2.3 IF Drywell Equipment Drain Sump OR Reactor Building Equipment Drain Tank temperature increases to 180°F, THEN D

recirculation of the affected system initiates.

3.0 OPERATOR ACTIONS 3.1 Immediate Actions None IOAOP-16.0 Rev. 18 Page 2 of 11 I

3.0 OPERATOR ACTIONS 3.2 Supplementary Actions NOTE: High drywell pressure and temperature alarms should be anticipated.

3.2.1 PERFORM the foHowing as necessary to maintain RBCCW discharge header pressure greater than 60 psig:

1. START available RBCCW pumps.

D

2. ISOLATE any identified leaks due to pipe rupture.

D 3.2.2 IF 2D RBCCW Pump is in service to either drywell THEN PERFORM the following:

1. IF 2D RBCCW Pump is the source of the leakage, THEN PERFORM the following:

a. SECURE 20 RBCCW Pump.

D

b. ISOLATE the unit from the leak.

D

2. IF a loss of heat sink (Unit 1 RB Chiller) has occurred, THEN ENSURE 20 RBCCW Pump is tripped.

D NOTE: A complete loss of RBCCW is defined as discharge header pressure below 60 psig, high temperature alarms on components supplied by RBCCW, and all available RBCCW Pumps running.

3.2.3 IF there is a complete loss of RBCCW, THEN PERFORM the following:

1. TRIP all RBCCW pumps (including 20 RBCCW Pump if operating on the affected unit).

D

2. CLOSE the following valves:

RBCCW TO OW ISOL VLVS, RCC-V28 D

- RBCCW TO OW ISOL VLVS, RCC-V52 D

3. TRIP RWCU pump(s).

D IOAOP-16.0 Rev. 18 Page 3 of 11 I

3.0 OPERATOR ACTIONS System~

4~ ISOLATE the RWCU D

5~ REDUCE the speed of both reactor recirculation pumps to minimum~

D 6~ MANUALLY SCRAM the reactor AND ENTER 1(2)EOP-01-RSP~

D 7~ pumps~

TRIP both reactor recirculation D

NOTE: CRD pumps may NOT be operated for greater than 20 minutes without cooling water except as directed by the Unit sca under the following conditions:

- A CRD pump is available AND alternate control rod insertion is required OR

- CRD pump operation is required for reactor vessel level control 8~ IF CRD pumps are NOT needed for control rod insertion OR reactor vessel level control, THEN TRIP both CRD D

pumps~

3.2.4 IF there is a partial loss of RBCCW pressure or service water, THEN PERFORM the following:

1. IF any of the following conditions exist, THEN REFER to OAOP-18.0 or OAOP-19.0:

High temperatures on equipment cooled by RBCCW D NSW or CSW header pressure approaching pump D shutoff head (approximately 90 psig)

RBCCW HX OUTLET HDR TEMP HI (UA-03 1-3) in D alarm IOAOP-16.0 Rev. 18 Page 4 of 11 I

3.0 OPERATOR ACTIONS

2. MONITOR recirculation pump seal temperature on RECIRC. PUMP TEMP recorder, 832- TR-R601.

D

3. IF either of the following conditions exist, THEN SHUT DOWN the affected reactor recirculation pump(s):

Seal heat exchanger inlet temperature for Seal 1 or Seal 2 exceeds 200°F D

RBCCW to the recirculation pump seal heat exchangers is lost for more than 10 minutes.

D

4. IF the Reactor Mode Switch is in RUN AND both reactor recirculation pumps have been shut down, THEN D

INSERT a manual reactor scram .

5. REDUCE system heat load by removing the following systems from service:

RWCU D

Fuel Pool Cooling D

Drywell Equipment Drain Cooler D

Reactor Building Equipment Drain Cooler D

Reactor Building, PASS, and Radwaste Sample Stations D

6. MONITOR drywell temperature and pressure.

D 7.. IF abnormal primary containment condition occurs, THEN REFER to OAOP-14.0.

D

8. IF entry conditions are reached, THEN ENTER OEOP-02-PCCP.

D

9. IF RBCCW can NOT be restored, THEN COMMENCE a plant shutdown in accordance with OGP-05.

D

10. IF necessary to maintain spent fuel pool water temperature below 125°F, THEN REFER to OAOP-38.0.

D IOAOP-16.0 Rev. 18 Page 5 of 11 I

3.0 OPERATOR ACTIONS 3.2.5 IF in-leakage from components cooled by RBCCW is suspected, THEN PERFORM the following:

NOTE: In-leakage from a recirculation pump seal cooler may cause high recirculation pump motor temperature, low seal number 1 pressure, low recirculation pump seal staging flow, as well as high activity in the RBCCW system.

1. INITIATE actions to identify and isolate the source of the in-leakage.

D

2. IF source of the in-leakage can NOT be determined from available indications, THEN NOTIFY E&RC to sample RBCCW from the following locations:

Reactor Recirc Pump A Cooler Outlet D

Reactor Recirc Pump B Cooler Outlet D

Cleanup NRHX A Shell Side Outlet D

Cleanup NRHX B Shell Side Outlet D

Fuel Pool HX A Shell Side Outlet D

Fuel Pool HX B Shell Side Outlet D

3.2.6 IF RBCCW HEAD TANK LOW LEVEL (UA-03 1-5) is in alarm, AND low level is confirmed, THEN PERFORM the following:

1. MONITOR RBCCW Head Tank level.

D

2. FILL the RBCCW Head Tank in accordance with 1(2)OP-21 as necessary.

D

3. WALK DOWN accessible system piping to locate leakage.

D

4. CHECK drywell floor drain sump leakage rate.

D

5. ISOLATE any source of leakage.

D IOAOP-16.0 Rev. 18 Page 6 of 11 I

3.0 OPERATOR ACTIONS IR191 NOTE: Loss of AC power to the unit could result in the inability to monitor drywell temperature from the Control Room. During the time when the Control Room indication is NOT available, CAC-TR-778, located at the RSDP, can be used to ensure peak local drywell temperature history is accurately known.

NOTE: Maximum drywell temperatures allowed below the 75' elevation are as follows:

- Control Room temperature recorder CAC-TR-4426 greater than or equal to 260°F

- RSDP temperature recorder CAC- TR-778 greater than or equal to 258°F, Points 1, 3, and 4 3.2.7 IF all RBCCW pumps are off, THEN PERFORM the following:

1. IF drywell temperature has previously exceeded or is currently greater than the maximum temperature allowed, THEN PERFORM the following:

a. PLACE all RBCCW pump control switches in OFF.

D

b. RESTART RBCCW in accordance with the infrequent operation section of 1(2)OP-21 for high D

drywell temperature.

2. IF drywell temperature has NOT exceeded the maximum temperature AND the RBCCW pumps have lost electrical power or will NOT start, THEN PERFORM the following:

a. IF the affected unit's 4KV E buses are D deenergized, THEN REFER to OAOP-36.1.

b. IF RBCCW pump breakers local thermal or D magnetic trips have activated, THEN RESET the tripped device.

c. RESTART RBCCW in accordance with 1(2)OP-21 D when available.

IOAOP-16.0 Rev. 18 Page 7 of 11 I

3.0 OPERATOR ACTIONS

d. WHEN RBCCW is returned to service, THEN RESTORE plant systems to operation in D

accordance with their respective operating procedures.

3.2.8 WHEN directed by the EOPs, THEN OPEN the following valves to restore service water to RBCCW following a LOCA closure:

- RBCCW HX SERVICE WATER INLET VALVE, D SW-V103

- RBCCWHX SERVICE WATER OUTLET VALVE, D SW-V106.

IOAOP-16.0 Rev. 18 Page 8 of 11 I

480 GENERAL DISCUSSION RBCCW failure interrupts cooling water supply to the following components:

(1) Reactor Recirculation Pumps (2) CRD Pumps (3) RWCU Recirculation Pumps, Precoat Pump, and Nonregenerative Hx (4) DrywelJ Coolers (5) Penetration Coolers (6) Fuel Pool Heat Exchangers (7) Drywell Equipment Drain Heat Exchanger (8) RBEDT Heat Exchanger (9) Reactor Building Sample Station (10) Radwaste Building Sample Station (11) Postaccident Sample System Sample Coolers (12) Drywell Plate and Frame Heat Exchanger (when 2D RBCCW Pump in service)

RBCCW system failure could be due to a pipe rupture, pump failure, or loss of service water. This procedure attempts to reduce the system heat load on a partial loss of pressure due to pump failure, or high temperature due to service water failure. In the case where no RBCCW pumps are running and any drywell local temperature below the 75' elevation has exceeded 260 degrees, restart of the RBCCW pumps will be controlled using Infrequent Operation sections in 1(2)OP-21 in order to protect RBCCW piping integrity inside the drywell. If a loss of all pumps has occurred, and elevated drywell temperature existed, system piping may contain voids, which could lead to waterhammer upon an uncontrolled restart.

IOAOP-16.0 Rev. 18 Page 9 of 11 I

5.0 REFERENCES

5~ 1 OAOP-14.0, Abnormal Primary Containment Conditions 5.2 OAOP-18.0, Nuclear Service Water System Failure 5.3 OAOP-19.0, Conventional Service Water System Failure 5.4 OAOP-36.1, Loss of Any 4160V Buses Or 480V E-Buses 5.5 OAOP-38.0, Loss Of Fuel Pool Cooling 5.6 1(2)APP-UA-03, Annunciator Panel Procedure for Panel UA-03 5.7 1(2)APP-A-06, Annunciator Panel Procedure for Panel A-06 5.8 1(2)APP-A-07, Annunciator Panel Procedure for Panel A-07 5.9 2APP-UA-05, Annunciator Panel Procedure for Panel UA-05 5.10 OEOP-02-PCCP, Primary Containment Control Procedure 5.11 OENP-24.0, Reactor Engineering Guidelines 5.12 1(2)EOP-01-RSP, Reactor Scram Procedure 5.13 OGP-05, Unit Shutdown 5.14 1(2)OP-21, Reactor Building Closed Cooling Water System Operating Procedure IRISI 5.15 IER #92-21-03(IFI); FACTS #9389034 5.16 EWR #09588, Operation of CRD Pumps without RBCCW 5.17 GE SIL No. 459, S2 5.18 Technical Memorandum No. 8/M-90-001 IR191 5.19 NRC Generic Letter 96-06, Assurance of Equipment Operability and Containment Integrity During Design Basis Conditions 5.20 ESR 01-00400, AST Implementation for Fuel Handling 6.0 ATTACHMENTS None IOAOP-16.0 Rev. 18 Page 10 of 11 I

REVISION

SUMMARY

Revision 18 is an editorial correction to replace the Desdemona font used for check-off boxes and to enhance wording in Notes~

Revision 17 clarifies conditions under which a CRD pump may be operated for greater than 20 minutes without cooling.

Revision 16 adds clarifying information in a NOTE above steps required to be performed for a complete loss of RBCCW.

Revision 15 - Added new Step 3.2.4.4 to instruct the Operator to scram the reactor upon loss of forced recirculation (Mode Switch in RUN) as required by 1(2)AOP-04.0.

Revision 14 - Format changes to meet the requirements of OAP-005 and Microsoft Word XP. This revision makes non-intent changes to steps for clarity. This change does NOT implement an intent change. Additional administrative changes classified as "editorial": are bolding action verbs, italicizing components, change of cover page logo, removal of the "bar code" from the cover page, and adding place keeping aids.

Revision 13 incorporates EC 47025, Permanent RB/DW Chiller Installation. New Supplementary Action steps 2 and 3 have been added to address actions required if the 2D RBCCW Pump is in service to either drywelL Revision 12 incorporates actions required in response to NRC Generic Letter 96-06.

This action added a step to prevent RBCCW pump restart if all pumps were lost and any local drywell temperature below the 75' elevation exceeded 260 degrees Fahrenheit.

Revision 11 switches the order of steps in the Supplementary Actions section to clarify the actions to be taken for a complete loss of RBCCW and corrects procedure references.

IOAOP-16.0 Rev. 18 Page 11 of 11 I

Unit 2 APP UA-03 2-5 Page 1 of 1 RBCCW PUMP DISCH HEADER PRESS LOW AUTO ACTIONS NONE CAUSE

1. RBCCW pump trip due to any of the following:

a. Overload device.

b. Load shed sequence for applicable emergency bus.

c. Circuit malfunction.

2. Gross leakage or piping failure.

3. Improper valve lineup.

4. Increased heat load.

S. Circuit malfunction.

OBSERVATIONS

1. RBCCW pump indicates tripped or associated emergency 4160 volt bus has received an undervoltage or loss of off-site power signal.

2. RBCCW Pump Discharge And Header Pressure Indicator, RCC-PI-671-1, indicates less than 68 psig.

3. If RBCCW header pressure reaches 65 psig as indicated on RCC-PI-671-1, then the standby RBCCW pump should start.

ACTION

1. For RBCCW pump trip, start the standby pump if auto start has not occurred.

2. If pressure cannot be restored, refer to AOP-16.0, Reactor Building Closed Cooling Water System Failure.

3. If a circuit malfunction is suspected, ensure that a Trouble Tag is prepared.

DEVICE/SETPOINTS Pressure switch RCC-PS-673 68 psig POSSIBLE PLANT EFFECTS

1. Loss of RBCCW cooling capacity could result in a unit shutdown.

REFERENCES

1. LL-9353 - 35

2. AOP-16.0, RBCCW System Failure 1 2APP-UA-03 Rev. 42 Page 21 of 631

BRUNSWICK NUCLEAR PLANT PLANT OPERATING MANUAL VOLUME XXI ABNORMAL OPERATING PROCEDURE UNIT 2

2AOP-04.0 LOW CORE FLOW REVISION 16

\2AOP-04.0 Rev. 16 Page 1 of 131

1.0 SYMPTOMS 1&1 Reduction in core flow, reactor recirculation pump motor amps, reactor power, generator output, etc.

1.2 The following alarms may also appear in particular instances:

- RECIRC FLOW A LIMIT (A-06 3-2)

- RECIRC FLOW B LIMIT (A-07 2-4)

- SPEED CONTROL A SIGNAL FAIL (A-06 5-1)

- SPEED CONTROL B SIGNAL FAILURE (A-O? 4-3)

- RECIRC M-G A DRIVE MTR TRIP (A-06 2-3)

- RECIRC M-G B DRIVE MTR TRIP (A-O? 1-5)

- OPRM PBAlCDA ALARM (A-OS 5-8)

- OPRM UPSC TRIP (A-05 6-8)

- APRM UPSCALE (A-06 2-8) 2.0 AUTOMATIC ACTIONS 2.1 Reactor scram if OPRM detects instability when it is enabled 3.0 OPERATOR ACTIONS 3.1 Immediate Actions 3.1.1 IF the Reactor Mode Switch is in RUN AND both reactor D recirculation pumps have tripped, THEN INSERT a manual reactor scram.

3.1.2 IF reactor recirculation pump speed is lowering AND a D recirculation runback has NOT occurred, THEN PLACE the affected pump(s) SCOOP TUBE A(B) LOCK switch to TRIP.

!2AOP-04.0 Rev. 16 Page 2 of 131

380 OPERATOR ACTIONS 3.2 Supplementary Actions NOTE: Reactor recirculation pump speed mismatch and jet pump loop flows should be maintained within the following limits:

- 20% speed and jet pump loop flows within 10% (maximum indicated 6 6 difference 7.5 x1 0 Ibs/hr) with total core flow less than 58 x10 Ibs/hr

- 10 % speed and jet pump loop flows within 5% (maximum indicated 6

difference 3.5 x10 Ibs/hr) with total core flow greater than or equal to 6

58 x10 Ibs/hr NOTE: Process Computer Point U2CPWTCF, when validated, is the primary indication of total core flow, and should be used for stability region compliance. U2CPWTCF is invalid, U2NSSWDP or Attachment 1 may be used as an alternate indication for total core flow.

NOTE: As the stability region is a*pproached, Process Computer Point B018, Total Core Flow, and recorder 2B21-PDR/FR-R613, located on H12-P603, will read lower than Process Computer Point U2CPWTCF.

NOTE: The following computer screens may be used for reference:

- 802, Power/Flow - OPRM Operable - TLO

- 803, Power/Flow - OPRM Inoperable - TLO

- 804, Power/Flow - OPRM Operable - SLO

- 805, Power/Flow - OPRM Inoperable - SLO

- 806, Power/Flow - OPRM Operable - FWTR

- 807, Power/Flow - OPRM Inoperable - FWTR.

3.2.1 PERFORM the following to determine the current operating point on the applicable Power-Flow Map:

1. IF reactor recirculation pump speed AND jet pump loop D flow mismatch is within the allowable limits, THEN DETERMINE the current operating point using the applicable Power-Flow Map, as specified by OENP-24.0.

12AOP-04.0 Rev. 16 Page 3 of 131

3.0 OPERATOR ACTIONS

2. IF reactor recirculation pump speed OR-jet pump loop flow mismatch is NOT within the allowable limits OR the plant is in single loop operation, THEN PERFORM the foHowing:

NOTE: To compensate for signal noise, an average of several core DP readings should be used. Process Computer Point 8017 or ERFIS point B21DA014 is the preferred method for obtaining this average.

a. IF a valid total core flow from U2CPWTCF OR D U2NSSWDP is NOT available, THEN DETERMINE total core flow using Attachment 1.

b. DETERMINE the current operating point using the D applicable Power-Flow Map, as specified by OENP-24.0.

3.2.2 IF OPRM System is operable AND the current operating point is in the Scram Avoidance Region, THEN use one of the following methods to immediately exit the region:

NOTE: When raising core flow with two reactor recirculation pumps operating, pump speeds and jet pump loop flow mismatch should be maintained within the allowable limit.

6 NOTE: Total core flow should NOT exceed 45 x 10 lbs/hr (58%) in single loop operation.

RAISE core flow D INSERT control rods in accordance with D OENP-24.0, Form 2, Immediate Reactor Power Reduction Instructions.

12AOP-04.0 Rev. 16 Page 4 of 131

3.0 OPERATOR ACTIONS fR1l 3.2.3 IF the temperature differential between the coolant within D

~ the dome and the bottom head drain can NOT be maintained less than 145°F during the performance of this procedure, THEN INSERT a manual reactor scram.

3.2.4 IF OPRM System is inoperable, THEN PERFORM the following:

1. IF either of the following conditions are met, THEN INSERT a manual reactor scram:

The current operating point is in Region A D NOTE: Instability may be indicated by any of the following:

- OPRM PBNCDA ALARM (A-05 5-8) is in alarm

- OPRM UPSCALE TRIP (A-OS 6-8) is in alarm

- A rise in baseline APRM noise level. SRM power level and period meters may also be oscillating at the same frequency

- LPRM and/or APRM upscale or downscale alarms being received

- Sustained reactor power oscillations with a peak to peak duration of less than 3 seconds.

Indications of thermal hydraulic instability exist D AND the current operating point is in Region B, the 50/0 Buffer Region, or the OPRM Enabled Region.

12AOP-04.0 Rev. 16 Page 5 of 131

3.0 OPERATOR ACTIONS

2. IF the current operating point is in Region B, THEN use one of the following methods to exit the region:

6 NOTE: Total core flow should NOT exceed 45 x 10 lbs/hr (580/0) in single loop operation.

NOTE: When raising core flow with two reactor recirculation pumps operating, pump speeds and jet pump loop flow mismatch should be maintained within the allowable limit.

RAISE core flow D INSERT control rods in accordance with D OENP-24.0, Form 2, Immediate Reactor Power Reduction Instructions.

NOTE: Operating time in the 5% Buffer Region should be minimized.

3. IF the current operating point is in the 5% Buffer Region, D THEN INCREASE monitoring nuclear instrumentation for thermal hydraulic instability.

3.2.5 IF both reactor recirculation pumps have tripped, THEN PERFORM the following:

1. REDUCE CRD flow to 30 gpm. D

2. IF the Reactor Mode Switch is NOT in RUN, THEN D PLACE the plant in Mode 3 with 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3.2.6 IF the plant is in single loop operation, THEN PERFORM the following:

1. REDUCE CRD flow to 30 gpm. D 12AOP-04.0 Rev. 16 Page 6 of 13\

3.0 OPERATOR ACTIONS 6

NOTE: Total core flow should be maintained greater than 30.8 x 10 Ibs/hr to prevent the idle loop from cooling down and possibly exceeding the 100°F per hour cooldown rate.

6

@] 2. IF total core flow is less than 30.8 x 10 Ibs/hr, THEN RECORD the following at 15 minute intervals:

Bottom head drain temperature D Idle loop temperature D 3.2.7 NOTIFY the duty Reactor Engineer. D 3.2.8 MONITOR individual LPRM bar graphs from RBM ODAs D or PPC for reactor power oscillations.

[E] 3.2.9 MONITOR the following for reactor power oscillations:

APRMs D SRMs D SRM period meters D 3.2.10 MONITOR core thermal parameters AND ADJUST the following per the Reactor Engineer's recommendations:

Rod position D Reactor recirculation pumps speeds D 3.2.11 MONITOR plant parameters including the following:

Off-gas activity D Stack gas activity 0 Reactor recirculation pump variables D Recirculation loop temperatures D 12AOP-04.0 Rev. 16 Page 7 of 131

390 OPERATOR ACTIONS 3.2.12 IF OPRM System is inoperable, AND entry into the 5% D Buffer Region is required, THEN INCREASE monitoring nuclear instrumentation for thermal hydraulic instability.

3.2.13 IF both reactor recirculation pumps are operating, THEN PERFORM the following:

1. IF OPRM System is inoperable, THEN ENSURE Region D B is NOT entered.

2. ADJUST reactor recirculation pump speed as necessary D to maintain pump speed and jet pump loop flow mismatch within required limits.

3. ENSURE thermal limits are NOT violated. D

/2AOP-04.0 Rev. 16 Page 8 of 131

3.0 OPERATOR ACTIONS 3.2.14 IF all of the following conditions occur, THEN DETERMINE total core flow from U2NSSWDP OR Attachment 1 AND NOTIFY the Reactor Engineer for computer point substitution:

The plant is in single loop operation D Reactor power is greater than or equal to 23% D Computer point U2CPWTCF is NOT available D 3.2.15 CONFIRM all systems and components are operating D within the Precautions and Limitations Section of 20P-02.

3.2.16 IF 2AOP-04.0 entry was due to reactor recirculation D pump trip OR runback, THEN NOTIFY NIT within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to back up OPRM data for evaluation.

3D2D17 NOTIFY Chemistry to sample for iodine within two to six D hours following a change of thermal power of more than 150/0 in one hour.

3.2D18 IF entry condition for this procedure was a reactor recirculation pump trip, THEN PERFORM the following:

1D REVIEW OGP-14 for applicability. D 2D PERFORM the following to facilitate recovery from loss D of a recirculation loop:

NOTE: An idle reactor recirculation pump should NOT be started with the discharge valve open.

a. CONTINUE plant operation with the idle reactor D recirculation pump discharge open.

bD MAINTAIN total core flow between 30D8 x 10 6

D Ibs/hr (40 % ) and 45 x 10 fbs/hr (58 % ) to provide 6

adequate backflow through the idle loop.

c. IF desired to keep the loops differential D temperature less than or equal to 50°F, THEN RAISE the operating reactor recirculation pump speed AND REDUCE the seal purge flow to a minimum of 3 gpm.

/2AOP-04.0 Rev. 16 Page 9 of 131

4~O GENERAL DISCUSSION Several varieties of recirculation flow system malfunctions can cause a decrease in core coolant flow. The Reactor Recirculation System creates forced circulation of reactor coolant through the core. It is a piping system designed primarily to provide driving flow for the reactor jet pumps which, in turn, provide the coolant flow through the reactor core. The system is comprised of two separate and parallel recirculation loops. The tripping of one recirculation pump will reduce core flow from 1000/0 to approximately 600/0. In this case, flow would reverse through the 10 idle jet pump diffusers, and the other 10 jet pumps would continue to function.

If both recirculation pumps trip, natural circulation will provide approximately 300/0 of rated core flow and a gradual reduction in flow is the only result. However, due to core thermal hydraulic instability uncertainties, the reactor must be manually scrammed in response to a dual recirculation pump trip with the reactor mode switch in RUN. Recent problems identified with core flow measurement in single loop operation, and with recirculation pump speeds outside the allowable mismatch, have created the need (under some circumstances) to use core differential pressure to determine entry into the region of thermal hydraulic instability. Core differential pressure was chosen as a means to estimate core flow due to its relationship to core flow. Recirculation pumps will automatically trip on low water level +105" or high pressure 1137 psig.

The OPRM system provides alarms and automatic trips as applicable. If the OPRM System is inoperable, then Tech Specs require an alternate method to detect and suppress thermal hydraulic instability oscillations in accordance with BWR Owners Group Guidelines for Stability Interim Corrective Action, June 6 1994. This requires three stability monitoring regions (Region A - manual scram, Region B - immediate exit, and 5 % Buffer).

\2AOP-04.0 Rev. 16 Page 10 of 131

5.0 REFERENCES

em 5.1 NEDO-32465-A, Licensing Topical Report, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applicability GE Nuclear Energy, August 1996.

5.2 Technical Specifications 5.3 20P-02, Reactor Recirculation System Operating Procedure

[EJ 5.4 General Electric Service Information Letter No. 251/251, Supplement 1 5.5 General Electric Service Information Letter No. 517 5.6 Core Operating Limits Report (COLR) 5.7 OENP-24, Reactor Engineering Guidelines 5.8 Off-Site Dose Calculation Manual (ODCM)

~ 5.9 LER 1-99-002 (Insertion of Manual Reactor Trip Due to Reactor Vessel Bottom Head Stratification) 5.10 OGP-14, Extended Single Recirculation Loop Operation 6.0 ATTACHMENTS 1 Estimated Total Core Flow vs.Core Support Plate Delta-P

!2AOP-04.0 Rev. 16 Page 11 of 131

ATTACHMENT 1 Page 1 of 1 Estlmlated'TotaJ COfe Flow VS~ 'Core Support'PIO'te Delta P'for'B2C18 j *PerCellt 27.0 11/'/11 26:0 /fl/I//

lltt/IIIJ 25.0 'flllllL

/,111/1;'7/*

24.0 Ii I/II/rl

'/i 1111/

i 23.0 22.0 21.0

)'1/ ' / f (I i ))'//.Il/I/

20.0 Iii 1/'111 19.0

IJIJ/I, /

i~

1.0:~

0.0 i 20 25 30 35 40 45 50 55 60 85 70 75 80 85 Core Flow (Mlb/hr) 12AOP-04.0 Rev. 16 Page 12 of 131

REVISION

SUMMARY

Revision 16 incorporates EC 62929 by updating the title of Attachment 1 to reflect B2C18.

Revision 15 adds jet pump loop flow limit to recirc speed mismatch criteria.

Revision 14 incorporates EC 46653 (child 62488) by adding PPC point U2NSSWDP as an alternate for determining total core flow. This revision also adds a caution that an idle recirc pump may not be restarted to exit the scram avoidance region.

Revision 13 incorporates EC 50100 by updating annunciator A-05 5-8 noun name to

'OPRM PDA/CDA ALARM' and EC 56472 by updating the core flow-core dIp figure for the current fuel cycle and update map numbers. Added a step to notify NIT for backing up OPRM data.

Revision 12 - Incorporated EC 55156 which adds computer screens 811 and 812 to Note prior to Step 3.2.1 and corrected nomenclature for screens 806, 807,808, and 809.

Revision 11- Format changes to meet the requirements of OAP-005 and Microsoft Word XP. Steps that are not time dependant have been re-ordered to provide an easier transition. Other steps that have common sub-steps have been grouped together. These changes do NOT implement an intent change or a change in procedure methodology. Additional administrative changes classified as "editorial": are bolding action verbs, italicizing components, change of cover page logo, removal of the "bar code" from the cover page, and adding place keeping aids.

Revision 10 incorporates EC 46730 'Power Range Neutron Monitoring', EC 47907 'EPU Implementation' and EC 49331 'B2C16 Reload Core Design (figure 1).

Revision 9 incorporates ESR 00-00260 by updating Figure 1 (Core Flow vs Core Plate Differential Pressure) for cycle B2C15. This revision also deletes confusing wording in a caution and deletes a caution that was duplicated on the same page.

Revision 8 provides additional instructions for obtaining and documenting FCBB to ensure Tech Spec Compliance.

\2AOP-04.0 Rev. 16 Page 13 of 131

8.3 Recovery from Reactor Recirculation Pump Runback C Continuous Use 8.3.1 Initial Conditions

1. Reactor Recirculation Pump operation was previously in D accordance with Section 5.2.

NOTE: Recirculation Pump runback to approximately 49% speed occurs when reactor water level is less than or equal to 182 inches AND feedwater flow A or B is less than or equal to 16.40/0 of rated flow. A Recirculation Pump speed runback to 34% will occur when the Recirculation Pump discharge valve is NOT fully open or total feedwater flow is less than 16.40/0 of the rated flow. Both of these conditions will require a manual reset of the runback.

2. The conditions that caused the runback have cleared. D

3. The system operation has stabilized. D 120P-02 Rev. 126 Page 57 of 1531

8.3.2 Procedural Steps

1. ADJUST the potentiometer on RECIRC PUMP 2A(B) D CONTROL lowering the speed demand signal until the speed signal shows a slight decrease in pump speed using multiple indicators.

2. MONITOR Recirculation Pump speed and be prepared D to manually lock out the scoop tube if speed increases rapidly.

3. RESET the recirculation runback for Reactor Recirculation Pump A(B) as follows:

a. DEPRESS the RECIRC RUNBACK RESET push D button for Reactor Recirculation Pump A(B).

b. ENSURE reactor power and flow are stabilized. D

4. ADJUST flow as directed by the Unit SCQ. D 120P-02 Rev. 126 Page 58 of 1531

5.3 Speed/Power Increases Using the Recirculation Pump A(B) Speed R Reference Control Use 5.3.1 Initial Conditions

1. Reactor Recirculation Pumps in operation in accordance D with Section 5.2.

2. Feedwater flow is greater than 16.4% AND Recirculation D Pump flow limits are cleared.

5.3.2 Procedural Steps NOTE: Recirculation Pump speed changes are performed when directed by OGP-04 and OGP-12. Other operating procedures are used simultaneously with this procedure as directed by OGP-04, OGP-12, or the Unit seo. Speed changes are accomplished by slowly turning the potentiometer clockwise for increases and counterclockwise for decreases.

NOTE: Speed limiters number 1 and 2 must be manually reset prior to increasing pump speed above the respective speed limit setpoint.

NOTE: The following indications should be observed to ensure proper response to increased speed demand from a Recirculation Pump speed controller:

a. Recirculation Pump speed increases.

b. Recirculation loop flow increases.

c. Reactor power increases.

120P-02 Rev. 126 Page 34 of 1531

5.3.2 Procedural Steps

1. INCREASE Recirculation Pump speed in increments as directed by the Unit SeQ by slowly turning the RECIRC o

PUMP 2A(2B) SPEED CONTROL potentiometer in the clockwise direction.

120P-02 Rev. 126 Page 35 of 1531

Unit 2 APP UA-24 1-4 Page 1 of 1 PEN X 49B ELEV 86'-0" AZIMUTH 315 0 AUTO ACTIONS

1. The following excess flow check valve for the broken instrument line will close:

a. B21-IV-2455 (X49B-A) - excess flow check valve on Instrument Line B21-7013 for B21-LT-N026A.

CAUSES

1. Pipe break in the above instrument line.

2. Circuit malfunction.

OBSERVATIONS

1. Amber line break indicating light for penetration X49B-A on RTGB Panel XU-2 (Control Switch Module RIP-CS-1200) will be on for the broken instrument line.

2. Valve closed indicating light for penetration X49B-A on RTGB Panel XU-2 (Control Switch Module RIP-CS-1200) will be on for the excess flow check valve on the broken instrument line.

ACTIONS

1. Investigate the abnormal condition per 001-44.

2. If the cause of the annunciator is an instrument line break or a circuit malfunction is suspected, ensure that a WRlwo is prepared.

DEVICE/sETPOINTS Excess flow check valve B21-IV-2455 1.5 - 3.0 gpm POSSIBLE PLAN~ EFFECTS

1. Release of radioactivity into the secondary containment.

2. Invalid initiation signals and indications from the instruments supplied by the affected instrument line.

3. Excess flow check valve closure may result in a technical specification LCO.

REFERENCES

1. LL-9361 - 23

2. Technical Specification 3.6.1.3

3. 001-44, Excess Flow Check Valve position Indication Evaluation 12APP-UA-24 Rev. 33 Page 80f791

Remote Shutdown Monitoring Instrumentation 3.3.3.2 3.3 INSTRUMENTATION 3.3.3.2 Remote Shutdown Monitoring Instrumentation LCO 3.3.3.2 The Remote Shutdown Monitoring Instrumentation Functions shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS

---

~---------------------------------------~()TE Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required Function 30 days Functions inoperable. to OPERABLE status.

8. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel that is normally energized.

(continued)

Brunswick Unit 2 3.3-30 Amendment No. 260

Remote Shutdown Monitoring Instrumentation 3.3.3.2 SURVEILLANCE REQUIREMENTS (continued SURVEILLANCE FREQUENCY SR 3.3.3.2.2 Perform CHANNEL CALIBRATION for each required 24 months instrumentation channel.

Brunswick Unit 2 3.3-31 Amendment No. 233