Text

Feb-March 05000259/2008301 Exam Final Scenarios
	ML081370161
Person / Time
Site:	Browns Ferry
Issue date:	04/08/2008
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
References
	50-259/08-301
	Download: ML081370161 (437)
	v • d • e

Final Submittal

(Blue Paper)

FINAL SIMULATOR SCENARIOS

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Appendix D Scenario Outline Form ES-D-l

Facility: BFN Scenario Number: HLTS-3-1 Op-Test Number: HLT0610

Examiners: - - - - - - - - - - Operators: _

Initial Conditions:

Unit 3 has been operating for 192 days. Unit 2 has been operating for 56 days. Unit 1 has been

operating for 274 days. 3ED Diesel Generator is tagged for water jacket leakage repair. Day 2

of the LCO. Expected to be returned to service this shift. Fuelleakers on U3 are currently at

RFI 60,000. Thunderstorms are passing through the region, but no watches are in effect for the

immediate area. The 3C RFP was oscillating approximating 30 RPM during last shift, but is

now working properly and being monitored. The 3C RFP Pump is operating in automatic

in order to collect data for the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . A trouble shooting plan is being developed.

Turnover:

Support scheduled maintenance and testing activities. Alternate Stator Cooling Water Pumps per 3-01-

35A, Sect 6.3 per scheduled OPA.

Event Malfunction Event Event

Number Number Type* Description

1 mrf anO 1b reset N-ATC The crew will alternate Stator Cooling Water Pumps using 3-

N-BOP OI-35A.

N-SRO

1 N/A I-BOP The crew will respond to a HPCI Rupture Diaphragm pressure

TS-SRO switch PS-73-20B failure.

2 imf fw05b 100 R-ATC The crew will respond to a 3B HP FW heater isolation using

8:00 C-BOP 3-AOI-6-1.

R-SRO The crew will reduce power to r--<91 % using a recirc flow

reduction.

The crew will isolate feedwater to the 3B FW heater string.

The crew will further reduce power to <79% using a recirc

flow reduction.

3 imf swl0a C-BOP The crew will respond to a trip of the 3A Fuel Pool Cooling

C-SRO pump using 3-AOI-78-1.

TS-SRO

4 imffw13b C-ATC The crew will respond to a trip of the 3B Reactor Feedwater

C-BOP Pump (RFP) using 3-AOI-3-1 and 3-01-3.

C-SRO

5 bat M The crew will respond to a total loss of feedwater and reactor

NRCrfpactrip All scram.

6 bat M The crew will respond to a RCIC steam leak into secondary

HLTS04-1 All containment.

The crew will anticipate Emergency Depressurization or

perform Emergency Depressurization due to secondary

containment high radiation.

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

Appendix D Scenario Outline Form ES-D-l

Facility: BFN Scenario Number: HLTS-3-2 Op-Test Number: HLT0610

Examiners:- - - - - - - - - Operators: _

Initial Conditions:

Unit 3 is at 79% power. 3C RHRPump is out of service. T.S 3.5.I.A.I, 3.6.2.3, 3.6.2.4, 3.6.2.5

have been entered. Unit 3 is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months into a seven day LCO. Appendix R LCO addressed and in LCO

tracking. Loop II ofRHR has been vented within the hour in preparation for placing Torus cooling

in service. Valve 3-FCV-73-36 seal-in circuit has been disabled per step 7.6 of 3-SR-3.5.I.7

Turnover:

Continue with 3-SR-3.5 .1.7 which is in progress and is complete up to Step 7.11 (HPCI Main and

Booster Pump Set Developed Head and Flow Rate Test at Rated Reactor Pressure).

Alternate EHC Pumps per section 6.3 of 3-01-47A. Increase reactor power to 90% using Recirc flow (3-

GOI-IOO-I2, step 5.132) at 8 Mwe per minute.

Event Malfunction Event Event

Number Number Type* Description

N/A N-BOP The crew will alternate EHC pumps using 3-0I-47A.

2 N/A R-ATC The crew will continue with power ascension using 3-GOI-I2

R-SRO and 3-01-68.

3 imfhp08 C-BOP The crew will recognize and respond to a HPCI steam line

C-SRO break. HPCI will fail to auto isolate and must be manually

TS-SRO isolated.

The SRO will enter and execute EOI-3.

4 imfrdOla C-ATC Recognize and respond to a 3A CRD pump trip using 3-AOI-

85-3.

5 imfadOlg 40 C-BOP The crew will recognize and respond to a stuck open SRV

C-SRO using 3-AOI-I-I.

TS-SRO

6. batRRPAVIB M The crew will recognize and respond to a recirc pump high

imf cr02a 75 All vibration, dual seal failure, trip, core power oscillations and

3:00 scram.

The crew will carry out actions using EOI-I & 2 and 3-AOI-

100-1.

7 imfth22 100 M The crew will recognize and respond to a MSIV Closure and

1:30 All LOCA using EOI-I & 2.

The crew will monitor and control primary containment until

reactor water level approaches TAF.

The crew will transition to EOI C-I and perform Emergency

Depressurization to enable level restoration using low pressure

systems.

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

Appendix D Scenario Outline Form ES-D-l

Facility: BFN Scenario Number: HLTS-3-3 Op-Test Number: HLT0610

Examiners: - - - - - - - - - - Operators: _

Initial Conditions:

The HPCI system is tagged out for 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months to repair the Auxiliary Oil Pump. It is expected back in 3

hours. Flow indicator 3-78B is out of service. Instrument Mechanics are looking for a new transmitter.

The Main Generator voltage regulator has been placed in Manual for PMs on the Automatic voltage

regulator. The spare RBCCW pump in service to Unit 2.

Turnover:

Reduce power to 95% using recirculation flow due to low system load requirements. PMs on the

voltage regulator are complete. Return the Main Generator voltage regulator to Automatic operation.

Event Malfunction Event Event

Number Number Type* Description

N/A R-ATC The ATC operator will reduce reactor power to 95% using

R-SRO recirc flow using 3-01-68.

N/A N-BOP The BOP operator will return the Main Generator voltage

N-SRO regulator to Automatic using 3-01-47.

2 ior zdihs7542a C-BOP The crew will recognize and respond to an inadvertent start of

start C-SRO the 3D Core Spray pump.

TS-SRO The SRO will address Tech Specs.

3 imfrd0718-35 R-ATC The crew will recognize and respond to a control rod drifting

C-SRO into the core using 3-AOI-85-5.

TS-SRO The SRO will address Tech Specs.

4 imf ed12a C The crew will recognize and respond to a loss of 3A 480V

All RMOVboard.

TS-SRO The SRO will address Tech Specs.

5 bat NRC/ M The crew will recognize and respond to a recirc pump trip,

HLTS10-1 All power oscillations, scram and ATWS.

6 Timed out from M The crew will recognize and respond to a fuel failure during

batch file All the ATWS recovery actions.

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

Appendix D Scenario Outline Form ES-D-l

Facility: BFN Scenario Number: HLTS-3-4 Op-Test Number: HLT0610

Examiners:- - - - - - - - - Operators: _

Initial Conditions:

The unit is starting up following a refuel outage. Reactor power is at 1%. "C" RFP is uncoupled for

r-.J

performance of turbine overspeed testing. Currently at step 5.76.8 of3-GOI-IOO-IA.

Turnover: '

The 3C RFP is uncoupled and the suction and discharge valves are tagged for performance of turbine

overspeed. Currently at step 5.76.8 of3-GOI-100-1A and at step 5.6.13 of3-01-3 for warming 3B RFP.

Event Malfunction Event Event

Number Number Type* Description

none R-ATC Crew will continue to pull rods to increase power and start

N-BOP warming up 2B RFP

R-SRO

2 imfrd14a I-ATC Crew will respond to a RWM failure.

I-SRO SRO references Tech Specs.

TS-SRO

3 imf sw02a trip C-BOP Crew will respond to a RBCCW pump trip

7048FTC C-SRO Crew manually closes 70-48 after fails to auto close

4 ior zdihs468a C-BOP Crew will respond to feedwater controller malfunction which

imfth235 C-SRO results in cold water injection

5 imfth235 M Crew responds to fuel failure after cold water injection

All

6 imf cu0425 M Crew responds to a RWCU line break and scrams reactor

ior zdihs691 All before any area reaches max safe value.

null

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

HlTS-3--1

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SIMULATOR EXERCISE GUIDE

TITLE SLOW lOSS OF HP FEEDWATER HEATING ON B STRING, 2A FPC PUMP TRIP,RFP

TRIP, lOSS OF ALL FEEDWATER, UNISOlABlE RCIC STEAM LINE BREAK,

2 OR MORE AREA RAD lEVELS ABOVE MAX SAFE.

REVISION o

DATE January 2, 2008

PROGRAM BFN Hot License Training

PREPARED BY:

erations Instructor)

REVIEWED BY: rJlA-

. Date

REVIEWED BY: I/o);f

~ DJate

(Operations Traini g Manager or Designee)

CONCURRED:

Date

VALIDATION ----~~~~~::....---~~~-=----==---_\ ' J/9IvK

BY: Date

lOGGED-IN:

(librarian) Date

TASKS liST

UPDATED: Date

HLTS-3-1

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NUCLEAR TRAINING

REVISION/USAGE LOG

REVISION DESCRIPTION DATE PAGES REVIEWED

NUMBER I OF CHANGES I I AFFECTED I BY

0 Initial 01/02/2008 All RM Spadoni

1.

HLTS-3-1

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I. PROGRAM: BFN Licensed Operator Requalification Training

II. COURSE: License Requalification Training (Simulator Exercise Guide)

III. TITLE: SLOW LOSS OF HP FEEDWATER HEATING ON B STRING, FPC PUMP TRIP, RFP

TRIP, LOSS OF ALL FEEDWATER, UNISOLABLE RCIC STEAM LINE BREAK,

2 OR MORE AREA RAD LEVELS ABOVE MAX SAFE.

IV. LENGHT OF LESSON: 1 % to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

V. Training Objectives

A. Terminal Objectives

1. Perform routine shift turnover, plant assessment and routine shift operation in

accordance with BFN procedures.

2. Given uncertain or degrading conditions, the operating crew will use team skills

to conduct proper diagnostics and make conservative operational decisions to

remove equipment/unit from operation. (SOER 94-1)

3. Given abnormal conditions, the operating crew will place the unit in a stabilized condition

per normal, annunciator, abnormal, and emergency procedures.

4. Use step text procedural compliance.

B. Enabling Objectives

1. The operating crew will recognize and respond to a high pressure heater string

isolation as directed by 3-ARP-9-6A and 3-AOI-6-1A.

2. The operating crew will recognize and respond to a spurious FPC system trip and

will place the 3B pump I/S in accordance with 3-ARP-94 win 1 and 3-AOI-78-1.

3. The operating crew will recognize and respond to a RFP Trip with 3-AOI-3-1.

4. The operating crew will recognize and respond to a loss of feedwater event and

Rx SCRAM.

5. The operating crew will recognize and respond to unisolable RCIC steam line break, 2 or

more area rad levels above max safe requiring Emergency Depressurization.

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VI. References: The procedures used in the simulator are controlled copies and are used in development

and performance of simulator scenarios. Scenarios are validated prior to use, and any

procedure differences will be corrected using the procedure revision level present in the

simulator. Any procedure differences noted during presentation will be corrected in the

same manner. As such, it is expected that the references listed in this section need only

contain the reference material which is not available in the simulator.

A. SOER 94-01

B. SOER 96-01

VII. Training Materials:

A. Calculator (If required)

B. Control Rod Insertion Sheet (If required)

C. Stopwatch (If required)

D. Hold Order / Caution tags (If required)

E. Annunciator window covers (If required)

F. Steam tables (If required)

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VIII. Console Operator Instructions

A. Scenario File Summary

1. File: bat HLTS3-1

MF/RF/IOR# Description

a) ior zlofcv712[2] on Fails 71-2 and 71-3 open

b) ior zlofcv713[2] on

c) ior zlohs712a[2] on

d) ior zlohs713a[2] on

e) ior ypovfcv712 fail_now

f) ior ypovfcv713 fail_now

g) imf rm10h (e1 :25) 30 HCU-East rad ~ 30 mr/hr

h) imf rm10j (e1 :25) 25 HCU-West rad ~ 25 mr/hr

i) Imf rm10p (e1 2:00) 50 CS/RCIC area rad 50mr/hr

j) imf DG01D D DIG Fails to Start

k) imf DG02D D DIG Trip Protective Relay Operation

I) ior zi00hS211 Od20a[1] OFF 1816 Green Light Off

m) mrf DG01 D open Opens logic breaker

n) ior zdihs718a null Fails 71-8 valve closed

2. File: bat HLTS3-1-1

MF/RF/IOR# Description

a) mmf rm1 Op 1000 6:00 RCIC rad to max in 6 mins.

b) mmf rm10h 1000 13:00 HCU-West to Max in 13 mins.

c) mmf rm10j 1000 14:00 HCU- East to Max in 14 mins.

d) imf rc09 100 7:00 RCIC steam leak

e) imf ad01 b 0 MSRV 1-19 fails closed

f) imf ad01f 0 MSRV 1-34 fails closed

g) Imfad03b MSRV 1-19 Stuck Closed

h) Imf ad03f MSRV 1-34 Stuck Closed

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IX. Console Operator Instructions

B. Console Operators Manipulations

ELAP TIME PFK DESCRIPTION/ACTION

Sim. Setup rst 28 100 0h power MOC

Sim. Setup restorepref Establishes Preference Keys

HLTS3-1

Sim. Setup setup Verify Preference Keys

Sim. Setup esc Clears Popup Window

Sim. Setup F3 trg e1 MODESW Assigns trigger

Sim. Setup F4 bat HLTS3-1 see file summary

Sim. Setup manual Tag D D/G with Hold notices

ROLE PLAY: (After Stator coolant pumps alternated) As AUO, report 3-FIS-035-0065 reading 610 gpm, 3-HS-035-

0040 selected for "A" Stator Coolant pump on panel 25-114. If asked, inlet pressure is 10 psig on 3-PI-35-90.

When requested to reset local Stator F5 mrf an01 b reset Allows resetting MCR alarm

Coolant panel alarm then:

ROLE PLAY: As an 1M report that HPCI rupture diaphragm pressure switch PS-73-20B has failed low.

When directed from the Floor then: F6 imffw05b 1008:00 IB' HP heater string isolation

ROLE PLAY: If sent to investigate which valve is open, wait 2 minutes and report 3-LCV-22B light is out(B2 high

level dump)

ROLE PLAY: At ~ 79% power, as the Reactor Engineer, recommend inserting the first group of Emergency Insert

Control Rods.

If asked to reset local Cond Demin F7 mrf an01d reset allows reset of control room alarm

alarm

After conditions stabilized or as directed F8 imf sw10a Trips 3A FPC pump

by Floor Instr.

ROLE PLAY: (If asked) As AUO, report 3-78-506, 511, & crosstie 507 are ~pen & 3-78-510 (B hx outlet) is closed

ROLE PLAY: (If asked) As RW UO, 3-FRC-78-24 is in manual & set to OOk

If asked to throttle 3-FCV-78-66 F9 lor zlohs7866a[2] on

If asked to close 3-FCV-78-66 F10 dor zlohs7866a[2]

ROLE PLAY: (If asked) As Rx Bldg AUO, report 3B pump discharge pressure is 140 psig (PI-78-16 on 9-25-16)

ROLE PLAY: If sent to inspect breaker on 3A FPC pump, report bkr was found tripped and will not test

~~~MORE FOLLOWS~~~

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IX. Console Operator Instructions

G. Console Operators Manipulations (continued)

ELAP. TIME DESCRIPTION/ACTION

After FPC restored and as directed by F11 Imf fw13b Trips 3B RFP on thrust brng wear

Floor Instr.

ROLE PLAY: If sent to check 38 RFP report that there is no apparent cause but you will continue to check

When directed by Lead Examiner F12 bat rfpactrip trip a&c RFP's

If doesn't start on low level <shift>F1 imf rc02 Start of RCIC

After HPCI is in manual control and <shift>F2 imf hp07 HPCI 120V failure

injecting up to -50" or directed by Lead

Examiner then:

After 10 minutes of RCIC operations or <shift>F4 bat HLTS3-1-1 Max. Rad (2 areas in 13 mins.)

directed by Lead Examiner then:

ROLE PLAY: If directed to close RCIC valves 71-2 & 3 locally, respond that you are waiting on RadCon to enter

the Reactor building.

If decided to attempt to close valves mrf rc05k emer 71-2 to emerg

locally: mrf rc05s emer 71-3 to emerg

To return transfer switch to normal mrf rc05k norm 71-2 to norm

mrf rc05s norm 71-3 to norm

ROLE PLAY: Outside US reports that it appears to be a generic problem with RFP control oil system.

After RCIC is injecting and recovering <shift>F3 dmf fw13b Allows Crew to inject with "B" RFP

level then:

ROLE PLAY: Call the MCR and report that "8" RFP is repaired and ready for use

Terminates the scenario when the following conditions are satisfied or upon request of the floor instructor:

1. All rods fully inserted

2. Reactor Water level normal

3. Emergency Depressurization

HLTS-3-1

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x. Scenario Summary

With the unit operating at 1OO°A> , the operating crew will experience a slow loss of FW HTR level control

on the 8 high pressure heater string. Once the heater is isolated and power reduced, a trip of 3A FPC

pump will require the operator to start 38 FPC pump per 3-01-78. When plant conditions are stable the 38

Reactor Feedwater Pump will trip, the crew will respond per 3-AOI-1-3. After conditions stabilize, The

crew will experience a loss of the remaining RFPs which will cause the crew to scram and utilize RCIC for

level control. When RCIC is initiated it develops a steam leak which cannot be isolated forcing the crew to

emergency depressurize based on 2 Area Rad Monitors above maximum safe. If HPCI is used for water

level control the crew will experience a problem with the flow controller to respond in automatic.

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x. Information to Floor Instructors:

A. Ensure recorders are inking and recording and ICS is active and updating.

B. Assign Crew Positions based on the required rotation.

1. SRO: Unit Supervisor

2. ATC: Board Unit Operator

3. BOP: Desk Unit Operator

C. Terminate the scenario when the following conditions are satisfied or at the direction of the

Lead Examiner:

1. All rods fully inserted

2. Reactor Water level normal

3. Emergency Depressurize on 2 RADS above max safe

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XI. Simulator Event Guide

Event 1: NOR. OPS. & HPCI PRESSURE SWITCH FAILURE

POSITION EXPECTED ACTIONS SAT/UNSAT/NOTES

ATC/BOP Alternate Stator Cooling Pumps lAW OI-35A,

sect. 6.3.

-start standby pumps

-stops running pump

-coordinates local verification of system flow and

pressure

-coordinates local positioning of selector switch

Responds to Report by IMs of HPCI rupture

diaphragm pressure switch failure

(3-PS-73-20B), by relaying information to SRO.

SRO Consults Tech Spec 3.3.6.1 determines only

three pressure switches required.

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XI. Simulator Event Guide

Event 2: Slow Loss of HP Feedwater Heating on B string

POSITION EXPECTED ACTIONS SAT/UNSAT/NOTES

Crew Announces "BYPASS VALVE TO CONDENSER

NOT CLOSED"

ATC/BOP Dispatch AUO to JB 32-42 to determine which

bypass valve is open per ARP

Selects ICS screen FWHL

Announces "HEATER B2 LEVEL HIGH"

Dispatches personnel to Heater Level Controls

Verifies 3-FCV-6-95 open

Checks B2 heater shell pressure, drain flow

Announces B1 and B2 HP htr. Extraction isolation

SRO Enters 3-AOI-6-1:

Contacts Reactor Engineer

ATC/BOP Reduces power to 91 °lb rated with recirc flow (if

above)

Verifies 3B1 & 3B2 extraction valves closed

Verifies 3B1 & 3B2 MS Dr. Pump suction valves

closed

Identifies heater level still rising

SRO Directs isolating FW to B HP heater string

Directs power reduction to < 790/0 power (Mid-power

runback)

Enters 3-GOI-100-12, Power Maneuvering

Notifies Rx Eng. of Feedwater Heater isolation and

power reduction

HLTS-3-1

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XI. Simulator Event Guide

Event 2: Slow Loss of HP Feedwater Heating on B string (Continued)

POSITION EXPECTED ACTIONS SAT/UNSAT/NOTES

ATC/BOP Isolates FW to B HP heater string by closing 3-FCV-

3-31 and 76

Reduces Power to < 79°A> with Recirc. Flow

Monitors MT thrust bearing temps. (3-AOI-6-1A)

Closes 3-FCV-6-95

SRO Notifies ODS of reason for power reduction

ATC/BOP Notifies Chemistry & RACON

Crew Recognizes HTR level lowers as a result of isolating

the Condensate side of 3B HP HTR string (i.e. tube

leak)

HLTS-3-1

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XI. Simulator Event Guide

Event 3: Trip of 3A FPC pump

POSITION EXPECTED ACTIONS SAT/UNSAT/NOTES

Crew Recognizes 3A FPC pump trip, responds per the

ARP

ATC/BOP Performs the following:

Responds to alarm FPC system abnormal 3-

ARP-9-4C win 1

Enters 3-AOI-78-1 for start of a FPC pump

Coordinates with Rx Bldg AUO and Radwaste

UO to start 3B FPC pump

Starts 3B FPC pump

Verifies discharge pressure >120 psig with

AUO

Directs RW UO and Rx Bldg AUO place

demin in service

SRO/BOP Dispatch AUO/EMs to check breaker for 3A

FPC pump

SRO -Directs restoration of system after cause is

determined

SRO Evaluate Tech. Spec. (TRM 3.9.2/3.9.3)

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XI. Simulator Event Guide

Event 4: 3B RFPT Trip

POSITION EXPECTED ACTIONS SAT/UNSAT/NOTES

ATC/BOP Announces "RFPT B Abnormal" alarm and trip

of RFPT 'B'.

Refers to ARP, 3-AOI-3-1 and 3-01-3 and take

required action

SRO Dispatches AUO to RFP to determine cause of

trip

ATC/BOP Verifies that unit stable

Verifies Rx Thermal limits

SRO Contacts maintenance to check reason for

RFPT trip

NOTE: LEAD EXAMINER notify Console Instructor when ready to trip the next RFP (i.e. next event)

HLTS-3-1

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XI. Simulator Event Guide

Event 5: 3A and 3C RFP Trip

POSITION EXPECTED ACTION SAT/UNSAT/NOTES

ATC/BOP Recognizes 3A RFP trip and need for

reactor scram

SRO Directs Reactor scram

ATC Manually scrams the reactor

-mode switch in SID

-checks power lowering

-reports all rods in

-recognizes trip of 3C RFP and informs

SRO all RFP's are tripped

SRO Enters 3-EOI-1 on low reactor water level

Directs level be controlled by:

-RCIC

-CRD

-HPCI

-Enter AOI-1 00-1

ATC/BOP Utilizes RCIC for reactor water level control

Crew Recognizes radiation alarms associated

with RCIC operation

ATC/BOP Evacuates Reactor. Bldg.

SRO Enters EOI-3

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XI. Simulator Event Guide

Event 6: RCIC STEAM LEAK

POSITION EXPECTED ACTIONS SAT/UNSAT/NOTES

ATC/BOP If HPCI is used, recognizes auto control failure and

places HPCI controller in manual

ATC/BOP Places 3B RFP I/S after notified able to reset

Crew Monitors area radiation levels

ATC/BOP Recognizes and reports area radiation alarm for RCIC

room

Recognizes and reports high area temperature for

RCIC room

Recognizes RCIC failure to isolate and attempts to

manually isolate it

SRO Directs RCIC be isolated locally

Determines has two area radiation levels above max

safe lAW EOI-3 and directs emergency

depressurization by opening 6 ADS valves (C2)

ATC/BOP Opens 6 ADS valves and recognizes 2 valves failed to

open and opens 2 additional valves

Verifies RFP discharge valves closed

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XI. Simulator Event Guide

Event 6: RCIC STEAM LEAK

POSITION EXPECTED ACTIONS SAT/UNSAT/NOTES

SRO Directs RPV level be maintained

between +2" and +51"

with one or more of

the following: (After emergency

depress.)

-LPCI

-Core Spray

-Condensate

ATC/BOP After emergency

depressurization maintains RPV

water level TAF and restores

level +2" to +51" with one

or more of the following:

-LPCI

-Core Spray

-Condensate

SRO After EOI-2 entered on high

SP water level or temperature

directs the following:

- H2 0 2 analyzers placed in

service

ATC/BOP Places H2 0 2 analyzers

in service

SRO Directs all available Suppression Pool cooling be placed

into service due to Suppression Pool water temperature

ATC/BOP Places all available Suppression Pool cooling into service

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XII. Crew Critical Tasks (If an evaluated scenario)

SAT/UNSAT

1. Maintains reactor water level above TAF

2. Anticipates Emergency depressurize and rapidly

depressurizes using BPV's to main condenser

and/or Emergency depressurize based on 2 areas

radiation above maximum safe with a primary

system discharging to secondary containment (within

5 minutes)

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XIII. Scenario Verification Data

EVENT TASK# RO SR CONTROL MANIPULATION

o

1. Loss of HP Feedwater Heating 295014 3.7 3.9 817

U-068-NO-10

U-006-A8-01

S-006-A8-0 1

T-000-AD-17

2. 3A FPC pump trip

3. RFPs Trip 295001A2.01 3.7 3.7 83

U-003-A8-01

S-003-A8-01

U-003-NO-08 295001A4.02 3.9 3.7

295009G12 3.8 4.4

T-000-AD-17

4. HPCI Pressure Switch Failure U-073-AL-19 206000A2.09 3.5 3.7 85

S-000-AD-27 2.1.12 2.9 4.0

5. RCIC Leak/MSL Leak U-000-EM-10 295033 3.6 3.9 A7,814,A12,

U-000-EM-11 815,14,120

S-000-EM-10 295032 3.5 3.6

S-000-EM-12

U-000-EM-01 3.8 4.4

U-000-EM-02

U-000-EM-03 3.6 4.2

S-000-EM-01 3.5 4.1

S-000-EM-02 3.9 4.5

S-000-EM-03 3.9 4.5

U-000-EM-14 2.4.38 2.2 4.0

S-000-EM-15 295026 3.6 3.8

S-000-EM-24

T-000-AD-04

T-000-EM-09

T-000-EM-11

T-000-EM-16

HLTS-3-1

Revision 0

Page 20 of 21

SCENARIO REVIEW CHECKLIST

SCENARIO NUMBER HLTS 3-1

9- Total Malfunctions Inserted; List: (4-8)

1) High Pressure Heater Isolation

2) 3A FPC pump trip

3) RFPT trip

4) RCIC steam leak,

5) RCIC failure to isolate (auto or manual),

L Malfunctions That Occur After EOI Entry; List: (1-4)

1) RCIC steam leak

2) RCIC isolation failure (auto or manual)

..L Abnormal Events; List (1-3)

1) HP Heater Isol. (ARPs)

2) 3A FPC pump trip. (AOI & ARP)

3) RFPTs trip (ARP, AOI)

_1_ Major Transients; List: (1-2)

1) RCIC Line Break

..L EOls used; List: (1-3)

1) EOI-1

2) EOI-2

3) EOI-3

_1_ EOI Contingencies Used; List: (0-3)

1) C2

90 Run Time (minutes)

29 EOI Run Time (minutes); 30 0,lc, of Scenario EOI Run Time

L. Crew Critical Tasks

yes Technical Specifications Exercised (yes/no)

Page 21 of 21

XIV. SHIFT TURNOVER INFORMATION

Equipment out of service/LCOs: Unit 3 has been operating for 193 days. Unit 2 has been operating

for 56 days. Unit 1 has been operating for 290 days.

3ED Diesel Generator tagged for water jacket leakage repair Day 2 of LCO. will be returned to service

this shift.

Operation/Maintenance for the Shift: Support scheduled maintenance and testing activities

Alternate Stator Cooling Water Pumps per 3-01-35A. Sect 6.3 per scheduled OPA.

Unusual Conditions/Problem Areas: Fuel leakers on U3 are currently @ RFI 60.000.

Storms passing through the region. No Watches in effect fo r the immediate area.

3C RFW Pump was oscillating approximating 30 RPM during last shift. but currently working properly and

being monitored. Pump is operating in automatic to collect data for next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Trouble shooting

plan being developed.

HLTS-3-2

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PAGE 1 OF 21

SIMULATOR EVALUATION GUIDE

TITLE HPCI STEAMLINE BREAK, SRV FAILURE, RECIRC PUMP TRIP, DRYWELL LEAK,

EMERGENCY DEPRESSURIZATION ON LEVEL (C1)

REVISION o

DATE January 2, 2008

PROGRAM BFN Operator Training - Hot License

PREPARED BY: _~~~-+------.;;~=--_~ \ ) IZ !c8

~

REVIEWED BY:

(LOR Lead Instructor or Designee) Date

REVIEWED BY: 7~£~* (Operations Tlfaining Manager or Designee)

CONCURRED:

(Operations Superintendent or Designee'f(Required for Exam Scenarios only) Date

""-~--"'--'"" .'/ "'"

VALIDATION BY:

-;;-r: {;;~j~/ .c: /Lq /t)/{"

Date

LOGGED-IN:

(Librarian) Date

TASKS LIST

UPDATED: Date

HLTS-3-2

REVO

PAGE 2 OF 21

NUCLEAR TRAINING

REVISION/USAGE LOG

REVISION DESCRIPTION OF DATE PAGES REVIEWED BY

NUMBER REVISION AFFECTED

0 INITIAL 4/6/07 All

HLTS-3-2

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PAGE 3 OF 21

I. Program: BFN Operator Training

II. Course: Hot License Training

III. Title: HPCI STEAMLINE BREAK, SRV FAILURE, RECIRC PUMP TRIP, DRYWELL

LEAK, EMERGENCY DEPRESSURIZATION ON LEVEL (C1)

IV. Length of Scenario: ~1 to 1 % hours

V. Examination Objectives:

A. Terminal Objective

1. Perform routine shift turnover, plant assessment and routine shift operation in accordance

with BFN procedures.

2. Given abnormal conditions, the operating crew will place the unit in a stabilized condition

per normal, abnormal, annunciator and emergency procedures.

B. Enabling Objectives:

1. The operating crew will alternate EHC pumps.

2. The operating crew will continue power ascension from ~ 79% power.

3. The operating crew will experience a HPCI steam line break during performance of 3-SR-

3.5.1.7 , HPCI Flow Rate, with a failure of HPCI to auto isolate.

4. The operating crew will recognize and respond to a safety-relief valve failed open.

5. The operating crew will recognize and respond to a high vibration and trip of 3A Recirc

pump.

6. The operating crew will recognize and respond to reactor power oscillations by scramming

the reactor.

7. The operating crew will recognize and respond to a high drywell pressure condition.

8. The operating crew will Emergency De-pressurize when in C1 before reactor water level

reaches -190".

HLTS-3-2

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PAGE 4 OF 21

VI. References: The procedures used in the simulator are controlled copies and are used in development and

performance of simulator scenarios. Scenarios are validated prior to use, and any procedure differences will

be corrected using the procedure revision level present in the simulator. Any procedure differences noted

during presentation will be corrected in the same manner. As such, it is expected that the references listed

in this section need only contain the reference material which is not available in the simulator.

VII. Training Materials:

A. Calculator

B. Control Rod Insertion Sheet

C. Stopwatch

D. Hold Order/Caution tags

E. Annunciator window covers

F. Steam tables

HLTS-3-2

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PAGE 5 OF 21

VIII. Console Operators Instructions

A. Scenario File Summary

1. File: bat HLTS3-2

MF/RF/10R# Description

a) trg e1 MODESW Sets trigger

b) trg e2 adssrv1-22 Sets trigger

c) ior zlohs7416a[1] off Tag Out 3C RHR

d) imf rh01c

e) ior zdihs7416a null

f) mrf hw01 fast Advances all charts

g) imf th33b (e1 0) 1 2:00 B MSL break in DW

h) imf th21 (e1 5:00) 1 10:00 Recirc. line break

i) imf rd01 a (e1 10:00) 3A CRDP trip

j) imf rd01 b 3B CRDP trip

k) imf hp09 Failure of HPCI to auto isolate

I) ior zdihs718a close Fails RCIC

m) ior ypovfcv718 fail_power Keeps the 8 valve closed

n) imf rp11 (e1 1:00) MSIV logic fuse failure

0) ior zdihs261a null Prevents Fire pump A from starting

p) ior zdihs262a null B

q) ior zdihs263a null C

2. File: bat torhrc

MF/RF/10R# Description

1) ior zlohs7416a[1] off RHR C Tagout

2) imf rh01c

3) ior zdihs7416a null

3. File: bat RRPAVIB

MF/RF/10R# Description

1) imfth12a Inserts Vibration Alarm

2) imf th1 Oa (none 1: ) Fails Recirc Pump A Inboard Seal

3) imf th11 a (none 2: ) Fails Recirc Pump A Outboard Seal)

4) ior zdihs681 open Prevents Recirc Pump A Suction Valve Closure

HLTS-3-2

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PAGE 6 OF 21

B. Console Operators Manipulations

ELAP. TIME PFK# DESCRIPTION/ACTION

Simulator setup rst 28 ~ 78 %Power MOC, use mid-power runback push button

'Simulator setup restorepref Establishes Function Keys

HLTS3-2

Simulator setup setup Verify Function Keys

Simulator setup esc Clears Function Key Popup

Simulator setup F3 bat HLTS3-2 See Scenario File Summary

Simulator setup manual Place suppression pool cooling in service (Loop II)

Simulator setup manual Place HO tags on '3C' RHR pump

Simulator setup manual Place TESTING/MAINT frames on Panel 9-3F, Windows 5, 11,

26 for HPCI 3-SR-3.5.1.7 complete up to step 7.11

ROLE PLAY: If asked, state that the anti-rotation collar markings are aligned.

When HPCI is at rated pressure and F4 imf hp08 Steam leak into HPCI room

flow

ROLE PLAY: AUO at HPCI quad. Reports a large steam leak on HPCI and present location is elev. 565 Rx.Bldg.

When requested, wait 2 min. then: F5 imf rd01a trips 2A CRDP

When directed by Lead Instructor F6 imf ad01g 40 Fails SRV-1-4 open

When RO cycles SRV then: F7 dmf ad01g SRV-1-4 closes

When directed by Lead Instructor F8 bat RRPAVIB Recirc Pump A high vibration, seal failure,

suction valve fails to close and power oscillations.

When dispatched to check 2A Recirc Vibration, wait 2 minutes and report back swinging 10 to 14 mils

When 'A' Recirc trips F9 dmfth12a Deletes vibration high alarm

4 min. after 2A recirc. pump trip F10 imf cr02a 75 3:00 Core power oscillations

then: and

F11 imf th22 (none 1:30) 100 Bottom head leak

When requested, wait 3 minutes F12 bat app16fg Defeats RHR injection valve timers

Terminate the scenario when the following conditions are satisfied are at the direction of the Lead Examiner.

1. RPV water level +2" to +51"

2. Drywell sprayed

3. Emergency Depressurizarion completed

HLTS-3-2

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PAGE 7 OF 21

IX. Scenario Summary

Given Unit 2 at 79°/b power, the crew will alternate EHC pumps and resume power ascension to 100%. As 3-

SR-3.5.1.7, HPCI Flow Rate, is continued the crew will experience a ruptured HPCI steam line with a failure of

HPCI to automatically isolate. Manual HPCI isolation will be possible. As power ascension is continued, an

SRV fails open but can be closed as steps of 3-AOI-1-1 are performed. The crew experiences high vibration

with a subsequent trip and seal leakage on the 3A Recirc Pump resulting in high drywell pressure. When the

diesel generators automatically start the 3ED diesel generator fails to auto start but can be manually started.

Finally, the crew will Emergency Depressurizes before reactor water level reaches -190".

HLTS-3-2

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PAGE 8 OF 21

Information to Floor Instructors:

A. Ensure recorders are inking and recording and ICS is active and updating.

B. Assign Crew Positions based on the required rotation.

1. SRO: Unit Supervisor

2. ATC: Board Unit Operator

3. BOP: Desk Unit Operator

C. Conduct a shift turnover with the Unit Supervisor.

D. Direct the shift crew to review the control board and take note of present conditions, alarms, etc.

E. Terminate the scenario when the following conditions are satisfied are at the request of the floor/lead

instructor/evaluator.

1. RPV water level +2" to +51"

2. Emergency Depressurizarion completed

HLTS-3-2

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PAGE 9 OF 21

XI. Simulator Event Guide

Event 1: Alternate EHC Pumps

POSITION EXPECTED ACTION(S) SAT/UNSAT/NOTES

ATC/BOP Receive crew briefing and walk boards down

SRO Directs BOP to alternate EHC pumps

BOP Alternates EHC Pumps in accordance with 3-01-

47A

Starts 3B EHC Pump

Verifies EHC header pressure 1550 to 1650

psig

Verifies 3B EHC motor amps <140

Stops 3A EHC Pump

I

HLTS-3-2

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PAGE 10 OF 21

XI. Simulator Event Guide (Continued)

Event 2: Power Ascension continued

POSITION EXPECTED ACTION(S) SAT/UNSAT/NOTES

SRO Directs power ascension per 3-GOI-100-12 and 3-

01-68

ATC Raises reactor power at 8 Mwe/minute in

accordance with 3-GOI-100-12 and 3-01-68

BOP Performs as peer checker for recirc flow changes

HLTS-3-2

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PAGE 11 OF 21

XI. Simulator Event Guide (Continued)

Event 3: HPCI Steam Line Break

POSITION EXPECTED ACTION(S) SAT/UNSAT/NOTES

SRO Directs BOP to continue with 3-SR-3.5.1.7 at step

7.11

BOP Makes plant announcement HPCI is to be started

Responds to Reactor Bldg Hi Rad alarm per the

ARP

SRO Enters EOI-3 on High Rad. I High Temp.

BOP Determines HPCI area source of hi rad

Responds to HPCI Leak Detection Temp Hi alarm

per the ARP

Recognizes HPCI not isolated when isolation lights

are illuminated

SRO Directs HPCI manually isolated

BOP Manually isolates HPCI steam supply

Evacuates HPCI area

SRO Receives EOI-3 entry on flood level in HPCI room

BOP Notifies Rad Con and Fire Protection

Monitors for lowering temperature and radiation

levels in HPCI area

SRO Directs entry into 3-AOI-64-2B

Directs FCV-1-55 and FCV-1-56 Open

BOP Opens FCV-1-55 and FCV-1-56 Open

SRO Sends personnel to investigate

Determines unit in 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months LCO (TS 3.5.1.D - HPCI

and C RHR Inop)

Tech. Specs. 3.6.1.3, on FCV 73-2 or 73-3 when

tagged (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months )

HLTS-3-2

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PAGE 12 OF 21

XI. Simulator Event Guide (Continued)

Event 4: SRV-1-22 Fails Open

POSITION EXPECTED ACTION(S) SAT/UNSAT/NOTES

CREW Recognizes SRV open

Main Steam Relief Valve Open alarm

lowering generator output

SRO Directs response per AOI-1-1

BOP Determines SRV-1-22 from acoustic monitor

BOP Places SRV-1-22 control switch from close to open

to close several times

BOP Cycles relief valve and reports SRV closed

SRO Evaluates Tech Spec operability of ADS valve.

Determines valve operable, but requests Eng.

evaluation (Functional evaluation)

HLTS-3-2

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PAGE 13 OF 21

XI. Simulator Event Guide (Continued)

Event 5: Recirc Vibration, Seal Leakage, Power Oscillations and Scram

POSITION EXPECTED ACTIONS SAT/UNSAT/NOTES

ATC Announces Recirc "3A" high vibration alarm

Consults ARP for Panel 9-4

Directs AUO to Local Panel to check vibration

Monitors Recirc Pump Temperatures

SRO Contacts Reactor Engineer

Directs BUO to reduce speed of 3A RRP to

reduce vibration

ATC Reduces 3A RRP speed with peer check to clear

vibration alarm

Announces Recirc A Seal Leakage Alarm

Identifies Seal Failure via Instrumentation

Recognizes lowering pressure on Recirc Pump A

1 seal

SRO Directs crew to watch for signs of increased

leakage

ATC Acknowledges Recirc Pump A seal leakoff high

alarm; informs SRO; consults ARP

Recognizes lowering pressure on Recirc Pump A

outboard seal; informs SRO

Monitors drywell parameters; notes pressure and

temperature increasing; informs SRO

SRO When vibration report received or dual seal failure

is reported, directs 'A' Recirc Pump tripped

ATC Trips Recirc A and closes the discharge valve

SRO Directs actions per 3-AOI-68-1

I

HLTS-3-2

REVO

PAGE 14 OF 21

XI. Simulator Event Guide (Continued)

Event 5: Recirc Vibration, Seal Leakage, Power Oscillations and Scram (Continued)

POSITION EXPECTED ACTIONS SAT/UNSAT/NOTES

ATC Directs AUO to Recirc MG Set to monitor oil temp.

SRO Directs 'A' Recirc Isolated

ATC Notes that Recirc Pump A suction isolation valve will

not close; informs SRO

Directs AUO to close Recirc Pump suction valve

locally at Board.

Checks Power to flow map to verify in region 1

Checks APRMs and LPRMs for indication of power

oscillations

Informs SRO of Power Oscillations

SRO Directs inserting emergency shove sheet control rods

BOP Keeps SRO informed as drywell pressure

approaches 2.45 psig

SRO Directs venting per 3-01-64-1

BOP Vents per 3-01-64-1

Directs Logs person to monitor release rates

SRO Directs manual reactor scram prior to reaching

2.45psig DW pressure

ATC Scrams the reactor

SRO Directs 3-AOI-1 00-1

ATC Carry out actions of 3-AOI-1 00-1

SRO Enters EOI- 1 & 2 at 2.45 psig drywell pressure

SRO Directs venting per Appendix 12

HLTS-3-2

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PAGE 15 OF 21

XI. Simulator Event Guide (Continued)

EVENT 6: MSIV CLOSURE/LOCA

POSITION EXPECTED ACTION(S) SAT/UNSAT/NOTES

SRO - RPV pressure controlled 800 to 1000 psig

with one or more of the following:

- MSRV's (App 11A)

- RCIC (App 11B)

- RPV level be maintained between +2" to +51"

with one or more of the following:

-RCIC

-CRD

BOP Controls pressure 800 to 1000 psig with one or

more of the following:

- MSRV's (App 11A)

- RCIC (App 11B)

Recognizes MSIV closures

and reports to SRO.

SRO Directs determining the cause of the isolation

Directs App 8G, App 12, and H2 0 2 Analyzers in

service

BOP Performs App 8G, App 12, and Places H2 0 2

Analyzers in service

HLTS-3-2

REVO

PAGE 16 OF 21

EVENT 6: MSIV CLOSURE/LOCA (continued)

POSITION EXPECTED ACTION(S) SAT/UNSAT/NOTES

ATC/BOP Attempts to maintain RPV water level +2" to

+51" with one or more of the following:

-RCIC (App 5C)

-CRD (App 58) 3-8YV-85-551

-SLC (App78)

SRO Directs SP cooling be placed in service

BOP Places SP cooling in service

SRO Directs App 8G be performed

BOP Performs App 8G

Monitors containment parameters

SRO Enters EOI-2 on DW pressure and re-enters

EOI-1 and directs the following:

- Verify all available DW coolers in service

- Venting per App 12

- H202 analyzers placed in service

SRO Directs cooldown

ATC/BOP Verify all available DW coolers in service

ATC/BOP Commences a cooldown as directed

SRO Determines cannot maintain SC pressure less

than 12 psig and directs SC sprayed

BOP Sprays suppression chamber per App 17C

HLTS-3-2

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PAGE 17 OF 21

EVENT 6: MSIV CLOSURE/LOCA (continued)

POSITION EXPECTED ACTION(S) SAT/UNSAT/NOTES

SRO When SC pressure exceeds 12 psig or if

SRO determines cannot maintain DW temp.

<280 then directs the following:

- Ensures Recirc. pumps shutdown

- DW blowers secured

- DW sprayed per App 17B

ATC Trips Recirc. pumps

Secures DW blowers

Requests 16F & 16G be performed

Sprays the DW using RHR

SRO Directs DW sprays/SC sprays be

stopped when that area reaches

o psig

BOP Stops DW/SC sprays when that area reaches

o psig

SRO Directs CRD inject per App 5B

ATC Performs App 5B

Reports 3B CRDP tripped

Monitors containment parameters

SRO Monitors RPV water level, determines level is

lowering. Re-enters EOI-1 at +2" RPV level

- Directs performance of App 5B (CRD)

- Directs performance of App 7B (SLC)

Crew Monitors Drywell I PSC I and RPV water level

SRO Enters C1 at ~ -100" to - 122"

Directs ADS inhibited

ATC Closes RFP discharge valves

Reports 3A CRDP tripped

HLTS-3-2

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PAGE 18 OF 21

EVENT 6: MSIV CLOSURE/LOCA (continued)

POSITION EXPECTED ACTION(S) SAT/UNSAT/NOTES

BOP Inhibits ADS

SRO After entering C1 align all available injection

systems for injection.

-Containment sprays terminated

When water level reaches TAF

(-162") and before -190 directs the following:

Enters C2

- Six ADS valves opened

- RPV level returned +2" to +51"

BOP When directed by US terminates

Containment Sprays and lines up RHR for

LPCI

BOP Opens and verifies open 6 ADS

valves

ATC/BOP Restores RPV water level +2"

to +51" using:

-RHR

-Core Spray

-Condensate

SRO Classifies event as Site Area Emergency

(1.1-S 1)

HLTS-3-2

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PAGE 19 OF 21

XII. Crew Critical Tasks

SAT/UNSAT

1) Manually isolate HPCI before 2 areas exceed

Maximum Safe Radiation or Temperature

levels.

2) Prevents ADS actuation when Rx level reaches

-120".

3) Emergency depressurizes RPV based upon not

being able to maintain reactor water level above

-162, but before reaching -190"

4) Restores I maintains water level above TAF

I

HLTS-3-2

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PAGE 20 OF 21

XIII. SCENARIO REVIEW CHECKLIST

SCENARIO NUMBER HLTS-13

e Total Malfunctions Inserted; List: (4-8)

1) HPCI steam line break

2) RBCCW 3A pump trips

3) 3A Recirc. high vibration

4) 3A Recirc pump suction valve fails open and will not close

5) Failure of ADS/SRV 1-22

6) Drywell Leak

~ Malfunctions That Occur After EOI Entry; List: (1-4)

1) CRD pump 38 fails to start

2) CRD pump 3A trips

3) RCIC 71-8 fails to open

L Abnormal Events; List: (1-3)

1) SRV fails open

..L Major Transients; List: (1-2)

1) Loss of all high pressure makeup

2) Drywell Leak

~ EOls used; List: (1-3)

1) EOI-1

2) EOI-2

3) EOI-3

..L EOI Contingencies Used; List: (0-3)

1) C1

2) C2

90 Run Time (minutes)

45 EOI Run Time (minutes); 50 % of Scenario EOI Run Time

-L Crew Critical Tasks (2-5)

Yes Technical Specifications Exercised (yes/no) - Technical Requirements Manual

REVO

PAGE 21 OF 21

XIV. Shift Turnover Information

Equipment out of servtce/t.Cos: 3C RHR Pump -...._is____.;.._

out of_ ~

service. T.S ____.;..~_

3.5.1.A.1.___l_ _

3.6.2.3, 3.6.2.4, 3.6.2.5 have been entered. Unit 2 is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months into a seven day LCO.

Appendix R LCO addressed and in LCO tracking.

Operation/Maintenance for the Shift: Unit 3 is at 79°A> power. Alternate EHC Pumps per section

6.3 of 01 47A. Increase reactor power to 90% using Recirc flow (GOI-100-12.step 5.132) at 8 Mwe.

per minute. Continue with 3-SR-3.5.1. 7 which is in progress and is complete up to Step 7.11

(HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at Rated Reactor

Pressure). Loop II RHR has been vented within the hour in preparation for placing Torus cooling

in service.

Unusual Conditions/Problem Areas: 3-FCV-73-36 seal-in circuit has been disabled per step 7.6 of

3-SR-3.5.1.7

(

Browns Ferry Nuclear Plant

Unit3

Surveillance Procedure

3-SR-3.5.1.7

HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at

Rated Reactor Pressure

Revision 0044

Quality Related

Level of Use: .Continuous Use

Effective Date: 12-17-2007

Responsible Organization: OPS, Operations

Prepared By: MICHAEL S. RICE x6934

Approved By: John T. Kulisek

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 2 of 104

Current Revision Description

Pages Affected: 18,26,33,34,35,36,71,72,73,75,96,104

Type of Change: Revision Tracking Number: 048

PCRs: 07003919

Revised the stroke time criteria for the 3-FCV-73-18 valve to have a normal stroke time

range of 0.8 to 2.2 seconds and a maximum allowable stroke time range of 3.0 seconds.

This change in accepance criteria was evaluated and approved for use per 0-TI-383

Evaluation 07-1-IST-073-337.

Added instruction to restroke 3-FCV-73-18 if initial stroke time is less than maximum

allowable but outside the normal range. (PCR 07003919)

Added instruction to contact Duty Maintenance Manager if 3-FCV-73-18 was restroked and

to record the time. Per the OM Code, the restroked valve has to be evaluated within 96

hours.

Added new Illustration 1, Process for Stroke Timing Valves Per the ASME OM Code.

Added SR key number to Attachment 1 for scheduling.

Added instruction in Attachment 5 to contact OPS immediately if any evaluation results are

found to be NOT Acceptable.

Added new Attachment 10, ASME OM Code Restroke Time Record Form. (PCR

07003919)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 3 of 104

Table of Contents

1.0 INTRODUCTION 5

1.1 Purpose 5

1.2 Scope 6

1.3 Frequency 8

1.4 Applicability 8

2.0 REFERENCES 9

2.1 Technical Specifications 9

2.2 Updated Final Safety Analysis Report 9

2.3 Plant Instructions 9

2.4 Plant Drawings 9

2.5 Vendor Manuals 10

2.6 Other Documents 10

3.0 PRECAUTIONS AND LIMITATIONS 11

4.0 PREREQUiSiTES 19

5.0 SPECIAL TOOLS AND EQUIPMENT RECOMMENDED 23

5.1 Recommended Tools 23

5.2 Recommended Measuring and Test Equipment (M&TE) 24

6.0 ACCEPTANCE CRITERIA 25

7.0 PROCEDURE STEPS 27

8.0 ILLUSTRATIONS/ATTACHMENTS 72

Illustration 1: Process for Stroke Timing Valves Per ASME OM Code 73

Attachment 1: Surveillance Procedure Review Form 74

Attachment 2: HPCI Venting 76

Attachment 3: HPCI Cold Quick Start 83

Attachment 4: 3-FCV-73-18 Time Delay Adjustment 93

Attachment 5: ASME OM Code Inservice Testing Review Form 96

Attachment 6: HPCI Lube Oil Skid and Booster Pump Oil Level Settings 97

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 4 of 104

Table of Contents (continued)

Attachment 7: HFA Relay Contact Layout 98

Attachment 8: Installation and Removal of Yokogawa Recorders for 3-

FCV-73-18 99

Attachment 9: Annunciators Affected by Surveillance Procedure

Performance 103

Attachment 10: ASME OM Code Restroke Time Record Form 104

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 5 of 104

1.0 INTRODUCTION

1.1 Purpose

A. This procedure verifies the following High Pressure Coolant Injection (HPCI)

System Technical Specification (Tech Spec) surveillance requirements (SR):

1. The HPCI main and booster pump set must be capable of pumping

5,000 gpm against a simulated system head corresponding to reactor

pressure in order to satisfy SR 3.5.1.7.

2. HPCI discharge piping must be vented to meet SR 3.5.1.1 in lieu of

performing 3-SR-3.5.1.1 (HPCI) for the HPCI System if deemed necessary

by the Unit Supervisor (US).

3. This surveillance performs ASME OM Code Inservice Test (1ST) Program

testing of HPCI pumps and valves in order to satisfy Tech Spec 5.5.6

program requirements.

4. This surveillance provides overlap testing of the HPCI minimum flow valve

open and close functions to demonstrate compliance with SR 3.3.5.1.2 for

Table 3.3.5.1-1 Function 3f and SR 3.3.5.1.6.

B. This procedure also verifies the following additional licensing, INPO, and Fire

Protection Report (FPR) testing requirements:

1. Time-to-rated-flow testing is performed once an operating cycle or

whenever HPCI governor control system (GCS) corrective maintenance is

performed in order to satisfy a unit startup licensing commitment. This

testing is NOT specifically required by TS or 1ST Program requirements.

2. This procedure also accomplishes overspeed trip tappet trip valve

assembly testing recommended by INPO to ensure that the trip mechanism

is NOT binding. This testing is accomplished when the HPCI turbine is

cold and then when it is warm.

3. This surveillance is utilized to verify BFN FPR testing requirements which

demonstrate HPCI function operability.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 6 of 104

1.2 Scope

A. This surveillance verifies the HPCI turbine, main and booster pump set, and

supporting equipment (e.g., gland seal condenser) are capable of delivering

5,000 gpm against a simulated system head corresponding to reactor pressure.

1. This testing is accomplished by using startup test data which

conservatively approximates the required discharge head needed to

overcome system piping resistance and produce 5,000 gpm flow. This

pressure is added to the reactor steam dome pressure and is used as the

minimum discharge head required to satisfactorily meet SR 3.5.1.7.

2. The HPCI turbine is started and system flow is throttled back to a

condensate storage tank until a 5,000 gpm flow rate is attained while

verifying that the minimum, required discharge pressure can be obtained.

B. The same venting methodology utilized in 3-SR-3.5.1.1 (HPCI) is also used in

this surveillance to provide an alternate means of venting HPCI discharge

piping to comply with SR-3.5.1.1 for the HPCI System. This venting is

performed at the discretion of the US in lieu of performing 3-SR-3.5.1.1 (HPCI).

C. This surveillance in conjunction with SRs/Sls listed as being ASME type in

Surveillance Program Matrix fully implements the ASME OM Code 1ST Program

required by Tech Spec 5.5.6.

Satisfactory completion of this surveillance verifies Tech Spec 5.5.6 compliance

for the following valves:

Valve Test Description

ISV-73-23 HPCI turbine discharge pressure monitored to ensure that

valve is sufficiently open to perform its intended function.

CKV-73-603 HPCI turbine discharge pressure monitored to ensure that

valve is sufficiently open to perform its intended function.

CKV-73-559 The minimum flow valve is opened and the discharge

pressure drop is monitored to verify that this check valve is

opening to pass bypass flow.

FCV-73-18 This fast-acting valve's closure time is monitored.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 7 of 104

1.2 Scope (continued)

D. When Tech Spec and 1ST Program testing is accomplished, the turbine is

shutdown with the system throttled to simulate a system head corresponding to

reactor pressure.

1. If time-to-rated-flow testing is required, the turbine lube oil system and

turbine casing are given sufficient time to drain and cool to ambient

temperature, respectively. Once these two conditions are met, the system

is configured for a cold, quick turbine start and manual HPCI initiation

performed.

2. The time to reach 5,000 gpm flow against a simulated system head

corresponding to reactor pressure is verified to be ~ 30 seconds.

E. This surveillance in conjunction with 3-SR-3.3.5.1.6 performs the following

BFN FPR, Volume 1, Appendix R Safe Shutdown Program (Section V - Testing

and Monitoring) testing to verify that:

1. FCV-73-18 automatically opens and remains open during turbine startup

and operation,

2. FCV-73-18 closes when a manual turbine trip is initiated from the main

control room,

3. FCV-73-30 automatically closes when HPCI flow is greater than

approximately 1250 gpm,

4. FCV-73-30 automatically opens when HPCI flow is less than approximately

700 gpm, and

5. HPCI turbine and pump set operate per design during manual turbine

startup and operation.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 8 of 104

1.2 Scope (continued)

F. This surveillance satisfies SR 3.3.5.1.5 for Table 3.3.5.1-1 Function 3f by

functionally verifying that HPCI MIN FLOW VALVE FCV-73-30 closes when

HPCI pump set is operating above approximately 1250 gpm flow.

This testing is accomplished during turbine startup. FCV-73-30 is initially

opened to provide a minimum bypass flow path. When the turbine is started,

HPCI flow rises to the recirculation flow path back to the condensate storage

tank. The rising flow closes FCV-73-30 when an approximately 1250 gpm flow

rate is reached.

This surveillance also functionally verifies that FCV-73-30 will open when HPCI

flow is reduced below approximately 700 gpm flow.

This testing is accomplished when the ASME OM Code 1ST Program testing is

almost completed. A jumper is installed to allow FCV-73-30 to open when a low

flow signal is present. HPCI turbine speed is reduced with the flow indicating

controller in manual and HPCI pump set flow is reduced by throttling FCV-73-35

in the close direction until HPCI flow drops below approximately 700 gpm.

1.3 Frequency

A. This surveillance shall be performed once every 92 days when required by plant

conditions or whenever GCS corrective maintenance is performed which could

affect the GCS function. This SR shall be performed as required to satisfy BFN

GL 89-10 Program requirements.

B. [NRC/C] [NERlC] This surveillance is to be used for post-maintenance testing to

verify HPCI operability if the Governor Control System components require

corrective maintenance. [LER 296/85003] [INPO SOER 81-013]

1.4 Applicability

The surveillance requirements of this procedure are applicable in Mode 1. Modes 2

and 3 are also applicable except when RPV steam dome pressure s 150 psig.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 9 of 104

2.0 REFERENCES

2.1 Technical Specifications

Section 3.5.1, ECCS - Operating

2.2 Updated Final Safety Analysis Report

Section 6.3, Summary Description - Core Standby Cooling Systems

Section 6.4.1, High Pressure Coolant Injection System Description

Section 6.6, Inspection and Testing

Section 7.4, Core Standby Cooling System and Instrumentation

2.3 Plant Instructions

0-01-65, Standby Gas Treatment System

3-01-73, High Pressure Coolant Injection System

3-SI-3.1.5, HPCI Pump Performance

3-SI-3.1.12, HPCI System Pump Baseline Data Evaluation

3-SI-3.2.1, ASME Section XI Valve Performance

3-SR-3.3.5.1.5(F), High Pressure Coolant Injection System Pump Minimum Bypass

Flow Indicating Switch Calibration

3-SR-3.6.2.1.1, Suppression Chamber Water Temperature Check

3-SR-3.5.1.1 (HPCI), Maintenance of Filled HPCI Discharge Piping

0-TI-230, Predictive Monitoring Program.

0-TI-280, Calculations of Flow Transmitter Output for Use With ASME Section XI

SPP-8.1, Conduct of Testing

SPP-1 0.3, Verification Program

2.4 Plant Drawings

3-47E812-1 and -2, HPCI System Flow Diagram

3-47E610-73-1 and -2, HPCI System Mechanical Control Diagram

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 10 of 104

2.4 Plant Drawings (continued)

3-45E714-1 through -4, 250V DC RMOV Bd Schematic Diagram

3-45N3675-2, Panel 9-39 Wiring Diagram

3-45N3635-19, Local Instrument Panels Connection Diagram

3-730E928-1 through 5, -7 and -8, HPCI System Elementary Diagram

2.5 Vendor Manuals

BFN-VTM-G080-9270, BFN Unit 3 - Terry Model CCS HPCI Turbine Operation and

Maintenance Manual

BFN-VTM-B580-0010, Byron Jackson Technical Instructions High Pressure Coolant

Injection Pumps

2.6 Other Documents

NRC Inspection Report 82-13

Licensee Event Report 296/85003, Inoperability of HPCI System

Licensee Event Report 259/8232, Operator Notification

STI-15, HPCI Startup Test Instruction

Browns Ferry Nuclear Plant Fire Protection Report, Volume 1, Appendix R Safe

Shutdown Program

INPO SOER 89-001, Testing of Steam Turbine/Pump Overspeed Trip

GE SIL No. 336 R1, Surveillance Testing Recommendations for HPCI and RCIC

Systems

TVA Program Plan Implementation of NRC Generic Letter 89-10

Memorandum from D. Baker, GENE Power Ascension Operations Manager, to

M. Bajestani, BFN Technical Support Manager, dated June 25, 1991 (RIMS R40

910716805)

PGC-007-073-0, HPCI Operation Time Required to Raise Suppression Pool

Temperature 1 deg F (R40 910629 984)

INPO SOER 81-013, Concurrent Loss of High Pressure Core Cooling Systems

NRC Information Notice 91-50, A Review of Water Hammer Events After 1985

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 11 of 104

2.6 Other Documents (continued)

GE SIL No.1 06 R2, Suppression Pool Temperature Monitoring and Control

GE SIL No. 392 R1, Improved HPCI Turbine Mechanical-Hydraulic Trip Design

NRC Information Notice 93-67, Bursting of High Pressure Coolant Injection Steam

Line Rupture Discs Injures Plant Personnel

SEOPR 96-0-073-2, HPCI Turbine Administrative Vibration Limits

NEDC-32751 P, Power Uprate Safety Analysis for the Browns Ferry Nuclear Plant

Units 2 and 3 (RIMS R08-980316-888)

TVA-BFN-TS-384, Technical Specification (TS) Change TS-384 Request for License

Amendment for Power Uprate Operation (RIMS R08-980316-888)

GE-NE-B13-01866-39, Summary of System Evaluations and Proposed Changes to

Design Criteria Documents (RIMS W79-980427-005)

3.0 PRECAUTIONS AND LIMITATIONS

A. [NRC/C] LCO 3.5.1 requires the HPCI System to be OPERABLE in Mode 1 and

Modes 2 and 3 except when RPV steam dome pressure ~ 150 psig.

1. Entry into associated LCO 3.5.1 CONDITIONS AND REQUIRED

ACTIONS is NOT initially required provided the HPCI function is

demonstrated operable no later than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam dome

pressure reaches rated pressure from startup.

2. However, this surveillance removes the HPCI function from service (e.g.,

trip turbine using 3-HS-73-18A) for short duration's while performing

surveillance testing.

3. Consequently, entry into LCO 3.5.1 is administratively controlled within this

surveillance by declaring the HPCI function temporarily inoperable during

testing and verifying that LCO 3.5.1 CONDITIONS AND REQUIRED

ACTIONS have been met including tracking HPCI function inoperability in

Narrative Logs. [NCO 89-0216-002]

B. If maintenance other than what is provided in this surveillance procedure

becomes necessary, a work order should be generated.

C. Consult Attachment 9 for Panel 3-9-3 annunciators which will alarm during

performance of this surveillance.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 12 of 104

3.0 PRECAUTIONS AND LIMITATIONS (continued)

D. HPCI turbine operation below 2,400 rpm for extended periods except during

turbine startup and shutdown can result in inadequate oil pressure from the

turbine driven oil pump, higher system vibration, excessive exhaust line check

valve wear, or overheating of turbine driven oil pump when operating at low rpm

with auxiliary oil pump (AOP) running.

E. Suppression pool temperature will rise approximately 1°F every three minutes

during testing of HPCI System.

The temperature must NOT be allowed to exceed 105°F and must be returned

to ~ 95°F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after securing the HPCI Turbine as required by

LCO 3.6.2.1. The suppression pool temperature shall be monitored every

5 minutes and recorded in accordance with 3-SR-3.6.2.1.1, Suppression

Chamber Water Temperature Check.

F. Pressure suppression chamber (PSC) water shall NOT be used as the HPCI

water supply to perform this test because of its lower quality and the potential

water hammer risk when PSC water level is NOT high enough to swap HPCI

suction to the PSC.

G. The suppression pool shall be maintained at -5.5 to -2 inches as indicated by

3-LI-64-54A or 3-LI-64-66 on Panel 3-9-3.

H. Personnel stay time in HPCI Room during HPCI System operation should be

minimized if excessive exposure to noise, heat, or radiation is anticipated.

I. HPCI AOP operation should be minimized when HPCI System is in standby

readiness or following HPCI System shutdown. When AOP is operating,

turbine stop valve is held full open. If HPCI System is then manually or

automatically initiated, a HPCI turbine overspeed trip or high steam line flow

isolation may occur.

J. A radiation work permit (RWP) may be required for all personnel located in the

HPCI Room participating in the performance of this SR. RADCON shall be

consulted prior to turbine roll in order to determine the appropriate RWP

requirements.

K. Corrective Action shall be dispositioned in accordance with SPP-8.1, Conduct

of Testing.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 13 of 104

3.0 PRECAUTIONS AND LIMITATIONS (continued)

L. HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33 must be verified in AUTO mode

of operation with flow setpoint at 5,000 gpm if an automatic HPCI initiation

occurs during performance of this surveillance.

The HPCI flow controller 3-FIC-73-33 is a "FLOW X10" controller, 5000 gpm on

the controller digital display will read 500. The steps in this procedure which list

a flow value will be displayed as follows" flow as read on the digital display

followed by the actual flow in gpm" i.e. a flow of 1250 gpm is shown as "125

(1250 gpm)" a flow of 5000gpm is shown as "500 (5000 gpm).

M. The risk of steam emission to the surrounding area rises if a rupture disk breaks

during initial startup of turbine. Therefore, the number of personnel in HPCI

Room should be minimized until stable operation is achieved.

N. The identification number and calibration date for new test equipment, along

with step numbers for which it was used, shall be noted in the remarks

Section of Surveillance Procedure Review Form if during performance of this

surveillance it becomes necessary to change test equipment.

O. The HPCI PUMP MIN FLOW VALVE 3-FCV-73-30 will NOT open automatically

when low system flow is sensed unless a HPCI initiation signal is present.

P. HPCI pump and bearing temperatures should be monitored periodically using

HPCI/RCIC/RFWTEMPERATURES 3-TR-73-54 on Panel 3-9-47 or ICS to

ensure that temperatures are stable or NOT rising rapidly. Turbine shutdown

should be initiated if any oil temperature reading exceeds 155°F or any other

unsatisfactory oil condition is observed by personnel located in the HPCI Room.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 14 of 104

3.0 PRECAUTIONS AND LIMITATIONS (continued)

Q. [NRC/C] The HPCI System will be placed in configurations that make it susceptible

to overspeed tripping and motive steam loss should an initiation signal occur

when one of the following conditions is present:

1. When HPCI TURBINE STOP VALVE 3-FCV-73-18 is cycled open, the

GCS ramp generator will time out in approximately 12-13 seconds. If HPCI

System receives an automatic initiation signal after the ramp generator

times out and is reset by closure of turbine stop valve, a turbine overspeed

or a high steam line flow isolation may occur.

2. Manipulations of mechanical overspeed trip assembly (e.g., verifying

freedom of movement) may result in closure of turbine stop valve at a time

when it is required to be open for turbine operation.

3. Placing HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33 in manual mode or

changing its flow setpoint may NOT permit HPCI System to automatically

achieve design flow in the event of an automatic initiation.

4. Manual initiation of a turbine trip using 3-HS-73-18A will prevent turbine

stop valve from opening while trip push-button is depressed.

Since the above conditions may lead to a HPCI overspeed trip or loss of steam

supply if a HPCI initiation should occur during surveillance testing, the HPCI

System will be administratively removed from operable service to ensure that

RPV injection capability is maintained at all times when surveillance testing

could result in an overspeed condition of the HPCI turbine. [NCO 89-0216-002]

R. HPCI TURBINE STOP VALVE 3-FCV-73-18 operation should be observed for

visual and/or audible signs of a fast opening/closing transient during turbine

startup. Site Engineering and/or Mechanical Maintenance must be notified if

this type of transient occurs in order to evaluate the need for balance chamber

adjustments.

S. [II/F] Prior to initiating HPCI System and adding heat energy to suppression

chamber, the Unit Supervisor will evaluate need of placing Residual Heat

Removal System in suppression pool cooling mode to avoid the possibility of

thermal stagnation during sustained heat additions. [11-8-91-129]

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 15 of 104

3.0 PRECAUTIONS AND LIMITATIONS (continued)

T. The BFN ASME OM Code Ten Year Program for monitoring pump flow and

total developed head requires the use of measuring instruments capable of

+/- 2°Jb accuracy at full scale (FS) and having a maximum range which does NOT

exceed three times the maximum, expected process value. This accuracy

requirement is implemented for HPCI flow measurements by directly measuring

output of HPCI flow transmitter using the Integrated Computer System (ICS).

Existing local HPCI pump set suction and discharge pressure gages satisfy

ASME OM Code accuracy and range requirements and do NOT require

substitution with more accurate instrumentation.

Turbine speed indication (SI-73-51) on Panel 9-3 exceeds the 20/0 FS accuracy

requirement based on a review of two, as-found calibration checks performed

over a three year period. However, HPCI tachometer drift problems have made

it necessary to utilize local, hand held M&TE instrumentation (e.g., stroboscope)

to ensure that accurate turbine speed settings are established for ASME OM

Code purposes.

U. The ASME OM Code data recorded by this surveillance should be reviewed

and recorded in accordance with 3-SI-3.1.5 within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months of completion of this

surveillance.

v. ASME OM Code data collection requires that HPCI pump set be operated at a

predetermined flow rate and speed when discharge pressure readings are

taken. While the flow rate may be adjusted anywhere within the allowable

range specified (e.g., 4950 to 5050 gpm), UO must attempt to maintain the flow

rate as close as possible to midrange. This ensures that discharge pressure

readings do NOT vary significantly due to operating point changes from

performance to performance of this surveillance unless an actual deficiency

exists. UO must also ensure that turbine speed is adjusted as close as possible

to ASME OM Code test value of 3,800 rpm within the range 3790 to 3810 rpm.

Averaging techniques are acceptable.

W. Any control room ICS console may be utilized for collecting ICS data specified

by this surveillance. If ICS console originally selected fails to operate properly

during surveillance performance, another ICS console(s) may be used for

completion of test activities provided failure is isolated to console in use. If an

alternate ICS console(s) is used, then change(s) shall be noted in post-test

remarks Section of Attachment 1.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 16 of 104

3.0 PRECAUTIONS AND LIMITATIONS (continued)

X. HPCI System ICS startup transient data may be displayed and printed as

follows:

1. PRESS CANC key on ICS console keyboard.

2. SELECT GROUP.

3. SELECT MODIFY NON SYSTEMS GROUP.

4. SELECT test group to be modified (e.g., Test 15).

5. ENTER the following HPCI data points using F6 key to select fields:

FIELD POINT ID

03 73-31

06 73-33

09 73-51

12 DIG027

6. DELETE remaining data points from group by selecting Field 15 and

repeatedly pressing ENTER key until remaining data points are removed.

7. PRESS F3 to save group redefinition.

8. PRESS F1 to display group.

9. SELECT OT'HER GROUP FNCTS.

10. SELECT GROUP GRAPH 4 PTS ON 1 PLOT (Selection #4).

11. PRESS F2 to continue.

12. SETUP desired start data and time usinq F3 key.

13. SETUP time axis resolution of 1 minute per graticule.

14. PRESS Print Screen key to print plot of transient startup data.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 17 of 104

3.0 PRECAUTIONS AND LIMITATIONS (continued)

Y. Pressure switch 3-PS-73-47A controls AOP operation and resets at

approximately 35 psig lowering oil pressure.

AOP Operation causes 3-CKV-73-708 to close and the main, gear-driven oil

pump discharge pressure rises to its dead head value. This action causes

3-PS-73-47A to initiate at approximately 92 psig rising oil pressure which in turn

stops the AOP. The AOP will continue cycling on and off in this manner until the

main, gear-driven oil pump slows sufficiently to prevent initiation of

3-PS-73-47A.

Z. 3-FCV-73-6A and 3-FCV-73-6B close during HPCI turbine operation disabling

the drain path for inlet condensing pot (3-MCP-73-5).

Isolation of the drain path will eventually result in filling the inlet condensing pot

and HPCI TURBINE INLET DRAIN POT LEVEL HIGH 3-LA-73-5 (3-XA-55-3F,

window 26) will alarm. This is an expected condition and will NOT result in

turbine damage because steam flow into the turbine will prevent any excessive

accumulation of condensate in the inlet piping.

AA. 3-FCV-73-18 should be monitored for one continuous smooth action from full

closed to full open position.

The monitoring may be performed by either local visual line of sight, video

camera or video recorder, to ensure that once the 3-FCV-73-16 valve is opened

and Auxiliary Oil pump starts, the valve does NOT behave erratically (i.e.,

suddenly opening then closing and finally ramping open). (BFNPER 99-04221)

BB. During Starting, shutdown and tripping of the HPCI Turbine a second operator

should be utilized to assist in monitoring alarms and parameters for abnormal

conditions.

CC. Local vibration readings of the HPCI turbine and pump bearings (using portable

M&TE) may be obtained during each performance of this SR.

DO. Caution tags are available as prerequisites and are placed in Attachment 3 to

ensure that plant personnel do NOT operate these components prior to

completion of time-to-rated-flow testing.

EE. The Critical Steps warning represents a step or series of steps for an activity

which requires additional focus, attention, and increased awareness. The

Operator performing these steps for the activity needs to ensure the Unit

Supervisor and other Control Room staff are aware of the evolution. PEER

checks are required for this activity and short briefs need to be made prior to

performing the evolution. Included in the briefs are worst case scenario and

contingencies.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 18 of 104

3.0 PRECAUTIONS AND LIMITATIONS (continued)

FF. Step 7.0[21] serves to verify that the mechanical overspeed trip tappet

assembly is functioning properly and NOT binding when the HPCI turbine is at

its nominal, design operating temperature, and the overspeed trip automatic

reset time is approximately 4-6 seconds based on available turbine rpm

coastdown data and GE SIL 392 R1 recommendations.

GG. The discharge flow verification can be affected by how much air has been

introduced into the system and the fact that the discharge line is vented

for 1 minute through a closed drain prior to the discharge flow verification. The

most opportune time for this check is when the vent valve is opened when the

initial flow can be seen due to the turbulence initially created with the sightglass

empty.

Sight glass flow indication can be verified by any of the following: (Flashlight

should be used to assist in determination.)

1. Initial turbulence or bubbles seen through the sightglass when the

3-HS-73-63 push-button is depressed, followed by the sight glass filling

and the bubbles dissipating.

2. This occurs very fast therefore the operator must be monitoring prior to

depressing 3-HS-73-63.

3. Flowing water seen in sightglass

4. Lowering temperature gradient over the Ten minute period as seen by the

performance of Attachment 2 Step 1.0[23], if 3-FCV-73-45 is determined to

be seated.

5. Rising temperature over the Ten minute period as seen by the

performance of Attachment 2 Step 1.0[23], if 3-FCV-73-45 is determined to

have leakage.

HH. When timing the 3-FCV-73-18 valve, the 3.0 second requirement is such a tight

tolerance that using a stopwatch does NOT leave room for any errors. The use

of a Yokogawa recorder may be used as desired by System Engineering.

The Yokogawa Recorder can be connected in Panel 3-9-3 or Panel 3-9-39.

Only one location is required for testing, but may be connected to both for

additional data as required. System Engineering should determine the location

to be used. The Preferred location is Panel 3-9-3 for consistency and

communication. System Engineering will determine the location to be used.

II. When using the Yokogawa Recorder for measuring 3-FCV-73-18,

communications and countdown methods are to be established to ensure

recorders are on prior to operating the valve.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 19 of 104

Date

NOTES

1) Section 4.0 through Step 7.0[7], sets up the HPCI Surveillance for the Dynamic Run.

These steps may be performed up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the actual HPCI turbine

operation. Care should be given to all LCO entries. The latest revision of this

surveillance should be re-verified.

2) If this test continues for more than one shift, a Pre-job brief will be required for all new

personnel involved.

4.0 PREREQUISITES

[1] VERI,FY this copy of 3-SR-3.5.1. 7 is the most current revision.

[2] VERIFY the HPCI System is in a standby readiness

configuration in accordance with 3-01-73, High Pressure

Coolant Injection System.

[3] VERIFY the Reactor steam dome pressure is ~ 950 psig and

~ 1040 psig.

[4] VERIFY at least 2 turbine bypass valves full open (N/A if Main

Turbine is on-line).

[5] IF ICS will be utilized to collect HPCI flow rate data, THEN

CHECK that no gross instrument channel failure is present by

noting 'that HPCI flow rate on the ICS-displayed (single value

display (SVD 73-33) or the HPCI System mimic.), is within

100 gpm of flow rate indicated on HPCI SYSTEM

FLOW/CONTROL 3-FIC-73-33.

[6] VERIFY the following Operations personnel as a minimum are

available to perform this procedure. (This does NOT include

IV's or multiple shift performance or Peer Checking

requirements.)

UO: 2

AUO: 4

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure .Page 20 of 104

Date

4.0 PREREQUISITES (continued)

[7] NOTIFY each organization listed below.

REQUEST the number of Qualified personnel from each

organization to be available to support and perform their

associated activity: (If possible give a possible time reference

when personnel will be required.)

A. RADCON (1) will be available to

Determine RWP requirements

Will be available to monitor for airborne

contamination and radiation levels in the HPCI Room

during the startup and operation 'of the HPCI Turbine.

B. Electrical Maintenance (3 EM's) will be available to

Turbine/main pump speed readings using hand held

instrumentation locally in the HPCI Room during the

Startup and Operation of the HPCI Turbine.

Install jumper for Min Flow Valve.

Installation of Yokogawa Recorder for 3-FCV-73-18 if

used.

C. Mechanical Maintenance (2 MM's) will be available to:

Obtain a HPCI lube oil sample after turbine shutdown

prior to securing the Aux Oil Pump

Adjust 3-PCV-073-0501

Perform MPI-O-073-TRB001 if required.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 21 of 104

Date

4.0 PREREQUISITES (continued)

[8] CONTACT System Engineer or Duty Maintenance Manager to

determine if the following are to be performed:

[8.1] RECORD below if local vibration readings of HPCI pump

bearings using portable M&TE (Step 7.0[15.10]) are

required:

YES / NO (Circle one)

[8.2] RECORD below if verification of Time To Achieve Rated

Flow And Pressure is required: (Normally performed

once every two years after the refueling outage.)

YES / NO (Circle one)

[9] IF local vibration readings of HPCI pump bearings using

portable M&TE are NOT required, THEN

N/A Step 7.0[15.10]; (Otherwise N/A this step).

[10] IF verification of time to achieve rated flow is required, THEN

VERIFY that a Caution Order and associated tags have been

prepared to control operation of HPCI AUXILIARY OIL

PUMP 3-PMP-73-47 and HPCI PUMP CST TEST VLV

3-FCV-73-35. (Otherwise N/A)

[11] CHECK that a control room ICS console display is available to

monitor HPCI discharge pressure, flow, and manual initiation

status as a function of time; (REFER TO Step 3.0W and 3.0X).

(N/A if ICS is NOT available)

[12] PLACE "TESTING/MAINTENANCE" alarm window frame(s)

around the alarm windows listed in Attachment 9.

[13] CONTACT System Engineering to determine the following:

[13.1] CHECK the Timing Method to be used for 3-FCV-73-18.

STOPWATCH 0 YOKOGAWA RECORDER 0

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 22 of 104

Date

4.0 PREREQUISITES (continued)

[14] CHECK the Location(s) were the Yokogawa Recorder are to

be installed. (N/A if Yokogawa Recorders are NOT used.)

Panel 3-9-3 D Panel 3-9-39 D

[15] IF Yokogawa Recorders are to be used, THEN

NOTIFY Electrical Maintenance to install the Recorders per

Section 1.0 of Attachment 8 in the location(s) determined by

System Engineering. (Otherwise N/A)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 23 of 104

5.0 SPECIAL TOOLS AND EQUIPMENT RECOMMENDED

5.1 Recommended Tools

(1) Banana jack jumper

Carpenter's ruler or similar tool for measuring HPCI lube oil tank and booster

pump oil levels

Tape

Screwdriver for lifting leads.

Crescent wrench for adjusting lube oil pressures.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 24 of 104

Date

5.2 Recommended Measuring and Test Equipment (M&TE)

NOTE

The equipment data listed below may NOT be available at the time this procedure starts.

The data listed below may be collected at a later time but prior to using the equipment to

ensure calibrations requirements are met.

[1] ENTER information where required. Vibration M&TE accuracy

and frequency response range are controlled by the BFN

Vibration Program and have been verified to meet the listed

requirements. VERIFY required range and accuracy for

remaining M&TE by reviewing calibration sheets.

Recommended Frequency

Parameter Instrument (or Required Required Response Calibration

Measured equivalent instrument) Range Accuracy Range Due Date M&TE 10

MC Instruments Digital

Probe Tachometer

(Model 112) +/- 2% of

5428.6 rpm

Speed calibrated N/A

OR minimum

range

CSI Stroboscope

(Model No. 444)

CSI Model 2100 series +/-5% of

21.11-1000 Hz

Vibration vibration meter or N/A calibrated

minimum

equal range

(Local) digital or

N/A +/- 1 second N/A N/A

analog stopwatch

(MCR) digital or analog

N/A +/- 1 second N/A N/A

stopwatch

Time Yokogawa Recorder

(Panel 3-9-3 if used

for 3-FCV-73-18)

Yokogawa Recorder

(Panel 3-9-39 if used

for 3-FCV-73-18)

Omega Model HH22

digital thermometer 50°F to

Temperature 300°F +/- 1 of N/A

(surface and area air

minimum

temps are required)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 25 of 104

6.0 ACCEPTANCE CRITERIA

A. Responses which fail to meet the acceptance criteria stated below shall

constitute unsatisfactory surveillance procedure results and require immediate

notification of Unit Supervisor (US) at time of failure.

The following acceptance criteria shall be demonstrated as required by this

surveillance:

1. HPCI System is vented from high POINT VENT by observing continuous

water flow from vent when venting is performed at US discretion in lieu of

3-SR-3.5.1.1 (HPCI).

2. HPCI pump set delivers 5,000 gpm flow at a minimum discharge pressure

110 psi above reactor pressure.

3. [NRC/C] The HPCI System achieves 5,000 gpm flow at a minimum discharge

pressure 110 psi above reactor pressure within 30 seconds from a cold,

non-oil-primed, turbine quick start. (Only required following a refueling

outage or anytime maintenance affects Governor Control System

operation.) [LER 296/85003]

4. The differential pressure developed by the HPCI pump set shall be

~ 1034 psid and ~ 1201 psid when HPCI pump set is operating at

4950-5050 gpm flow and 3790-3810 rpm main pump speed.

5. HPCI PUMP MIN FLOW VALVE FCV-73-30 shall open when HPCI flow

rate lowers.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 26 of 104

6.0 ACCEPTANCE CRITERIA (continued)

6. The following valves shall comply with ASME OM Code Inservice Test

(1ST) acceptance criteria stipulated below:

Valve Acceptance Criteria

3-FCV-73-18 Valve shall close within 3.0 seconds or less when a

close signal is present.

3-ISV-73-23 Valve shall open sufficiently to perform its intended

function by noting that turbine exhaust pressure does

NOT exceed 40 psig when turbine is operating at or

near rated conditions.

3-CKV-73-559 Valve shall open sufficiently to perform its intended

function by noting at least a 70 psi drop in the HPCI

pump set discharge pressure when HPCI PUMP MIN

FLOW VALVE 3-FCV-73-30 is opened while HPCI

pump set is operating at or near rated conditions.

3-CKV-73-603 Valve shall open sufficiently to perform its intended

function by noting that turbine exhaust pressure does

NOT exceed 40 psig when turbine is operating at or

near rated conditions.

B. Steps which determine the above criteria are designated by (AC) next to initial

blank.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 27 of 104

Date

7.0 PROCEDURE STEPS

[1] CHECK that the following initial conditions are satisfied:

A. Precautions and limitations in Section 3.0 have been

reviewed.

B. Prerequisites listed in Section 4.0 are met.

C. The following annunciators are RESET:

HPCI TURBINE INLET DRAIN POT LEVEL HIGH

3-LA-73-5 (3-XA-55-3F, window 26)

HPCI TURBINE TRIPPED 3-ZA-73-18 (3-XA-55-3F,

window 11)

HPCI PUMP DISCH FLOW LOW 3-FA-73-33

(3-XA-55-3F, window 5)

HPCI TURB EXH DRAIN POT LEVEL HIGH

3-LA-73-8A (3-XA-55-3F, window 33)

D. The following indicating lights are EXTINGUISHED:

HPCI AUTO INIT 3-IL-73-59

HPCI AUTO ISOL LOGIC A 3-IL-73-58A

HPCI AUTO ISOL LOGIC B 3-IL-73-58B

HPCI TURBINE TRIP RX LVL HIGH 3-IL-73-18B

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 28 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

This surveillance will make HPCI INOP.

[2] PERFORM the following:

[2.1] The US and Unit Operator (UO) have been provided with

copies of this SR.

[2.2] UO has reviewed surveillance test scope including wire

lifts and jumper placements.

[2.3] OBTAIN permission from US to perform this

surveillance.

US

[2.4] [NRC/C] NOTIFY UO that this surveillance is commencing.

[RPT 82-16, LER 259/8232]

[3] RECORD date and time started, reason for test, and plant

conditions on Attachment 1, Surveillance Procedure Review

Form.

[4] VERIFY that suitable means of communication (e.g., hand

held radios, plant telephone system) will be available between

Main Control Room, HPCI Room, and HPCI vent station.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 29 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

Step 7.0[7.5] may be performed in parallel with remaining surveillance steps up to

Step 7.0[11] at the discretion and direction of US.

[5] VENT the HPCI discharge piping.

VERIFY the HPCI discharge piping has been vented within the

last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by one of the following:

[5.1] VERIFY 3-SR-3.5.1.1 (HPCI) has been performed within

the last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . (N/A if 3-SR-3.5.1.1 (HPCI) has NOT

been performed.)

[5.2] VENT the HPCI discharge piping by performing

Attachment 2. (N/A if 3-SR-3.5.1.1 (HPCI) has been

performed.)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 30 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTES

1) Three AUOs will be required in HPCI Room for performance of lube oil checks and

one Mechanical Maintenance available to adjust 3-PCV-073-501.

2) A crescent wrench may be required to position lube oil system stopcock valves.

[6] PERFORM the following lube oil system and control/stop valve

checks and adjustments:

[6.1] CHECK the following oil levels locally:

HPCI turbine lube oil reservoir level is per

Attachment 6.

Oil level in HPCI booster pump inboard and

outboard bearing oil sight glasses is per

Attachment 6.

[6.2] PERFORM the following:

[NRC/C] REVIEW Step 3.0Q for additional background

information regarding HPCI System removal from'

operable service. [NCO 89-0216-002]

IF Yokogawa Recorder(s) will be used to time

3-FCV-73-18, THEN

VERIFY installation of recorders in the location(s)

specified in Step 4.0[13] per Attachment 8.

(Otherwise N/A)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 31 of 104

Date

7.0 PROCEDURE STEPS (continued)

[6.3] IF HPCI System is Operable, THEN

PERFORM the following: (Otherwise N/A)

[6.3.1] VERIFY HPCI System may be removed from

operable service.

US

[6.3.2] DECLARE HPCI System inoperable.

US

[6.3.3] ENTER appropriate LCO information into Narrative

log.

US

NOTES

1) The TEST push-button 3-HS-73-47B is located in the HPCI Room at a local control

station on the south wall near the AOP.

2) Initial timing of 3-FCV-73-18 must be performed during the FIRST start of the Aux Oil

Pump with the oil system cold and de-pressurized.

3) Coordination between the operator starting the aux oil pump and the operator timing

the 3-FCV-73-18 valve must be performed to ensure proper timing.

4) Step 7.0[6.4] and Step 7.0[6.5] should be reviewed prior to starting the HPCI Aux Oil

Pump. These steps may be signed off after completion of Step 7.0[6.5].

[6.4] SIMULTANEOUSLY PERFORM the following:

DEPRESS and HOLD the HPCI AUX OIL

PUMP 3-HS-073-0047B TEST push-button until

Step 7.0[6.15].

AND

START the local STOPWATCH.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 32 of 104

Date

7.0 PROCEDURE STEPS (continued)

[6.5] WHEN Initial movement of local stem is observed on

HPCI TURBINE STOP VALVE 3-FCV-73-18, THEN

STOP the stop watch, and RECORD the time below:

seconds

NOTE

A time exceeding 13 seconds for turbine stop valve to begin opening may indicate a

problem with function of stop valve or lube oil system.

[6.6] VERIFY initial movement for turbine stop valve to begin

opening is less than 13 seconds.

IF recorded time is greater than 13 seconds, THEN

CONTACT Systems Engineering to determine if

diagnostic maintenance activities are required prior to

proceeding with testing.

[6.7] CHECK that HPCI TURBINE STOP VALVE

3-FCV-73-18 indicates OPEN by observing 3-ZI-73-18

position indicating lights.

[6.8] CHECK that HPCI TURBINE CONTROL VALVE

3-FCV-73-19 indicates OPEN by observing 3-ZI-73-19

position indicating lights.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 33 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

[NER/C] Step 7.0[6.9] verifies the turbine stop valve closing function using a Yokogawa

Recorder. Starting the recorder should be prior to operating the valve. Therefore a

countdown method or other method should be established between the Recorder Operator

and the Operator at Panel 3-9-3. must be ready to measure closure time of 3-FCV-73-18

since this is a fast acting valve. These steps may be signed off after completion of

Step 7.0[6.9.4]. [INPO SOER 89-001]

[6.9] IF a Yokogawa Recorder is to be used to measure

3-FCV-73-18, THEN

MEASURE closure time of HPCI TURBINE STOP

VALVE 3-FCV-73-18 by performing following:

(Otherwise N/A)

[6.9.1] NOTIFY the Recorder Operator to start the

Yokogawa Recorder on the desired point of the

countdown.

[6.9.2] DEPRESS and HOLD HPCI TURBINE TRIP

3-HS-73-18A until Step 7.0[6.9.4].

[6.9.3] [NRC/C] WHEN HPCI TURBINE STOP VALVE

3-FCV-73-18 is CLOSED as indicated by 3-ZI-73-18

position indicating lights, THEN

STOP the Recorder and RECORD closure time

below: [Appendix R]

3-FCV-73-18 CLOSURE TIME (SEC)

NORMAL MEASURED MAXIMUM

0.8 - 2.2 3.0

A. VERIFY the stroke time recorded is less than

or equal to the maximum value listed. _ _(AC)

[6.9.4] RELEASE HPCI TURBINE TRIP, 3-HS-73-18A.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 34 of 104

Date

7.0 PROCEDURE STEPS (continued)

[6.9.5] CHECK HPCI TURBINE STOP VALVE

3-FCV-73-18 indicates OPEN after a short time

delay by observing 3-ZI-73-18 position indicating

lights.

[6.9.6] IF the stroke time measured in step 7.0[6.9.3] is

less than or equal to the maximum listed but outside

the normal range, THEN

PERFORM the following: (Otherwise N/A)

A. RECORD on Attachment 10 the initial

measured stroke time from step 7.0[6.9.3]

above.

B. RESTROKE and TIME 3-FCV-073-0018 and

RECORD the restroke time on Attachment 10.

C. VERIFY the restroke time recorded on

Attachment 10 is less than or equal to the

maximum value listed. _ _(AC)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 35 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

[NER/C] Step 7.0[6.10] verifies the turbine stop valve closing function using a stopwatch. The

stopwatch must be ready to measure closure time of 3-FCV-73-18 since this is a fast acting

valve. These steps may be signed off after completion of Step 7.0[6.10.3]. [INPO SOER 89-001]

[6.10] IF a stopwatch is to be used to measure 3-FCV-73-18,

THEN

MEASURE closure time of HPCI TURBINE STOP

VALVE 3-FCV-73-18 by performing the following:

(Otherwise N/A)

[6.10.1] MEASURE closure time of HPCI TURBINE STOP

VALVE 3-FCV-73-18 by performing the following

substeps simultaneously:

DEPRESS and HOLD HPCI TURBINE TRIP

3-HS-73-18A until Step 7.0[6.10.3].

AND

START stopwatch at same time trip

push-button is depressed.

[6.10.2] [NRC/C] WHEN HPCI TURBINE STOP VALVE

3-FCV-73-18 is CLOSED as indicated by 3-ZI-73-18

position indicating lights, THEN

STOP stopwatch and RECORD closure time below:

[Appendix R]

3-FCV-73-18 CLOSURE TIME (SEC)

NORMAL MEASURED MAXIMUM

0.8 - 2.2 3.0

A. VERIFY the stroke time recorded is less than

or equal to the maximum value listed. _ _(AC)

[6.10.3] RELEASE HPCI TURBINE TRIP 3-HS-73-18A.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 36 of 104

Date

7.0 PROCEDURE STEPS (continued)

[6.10.4] CHECK HPCI TURBINE STOP VALVE

3-FCV-73-18 indicates OPEN after a short time

delay by observing 3-ZI-73-18 position indicating

lights.

[6.10.5] IF the stroke time measured in step 7.0[6.10.2] is

less than or equal to the maximum listed but outside

the normal range, THEN

PERFORM the following: (Otherwise N/A)

A. RECORD on Attachment 10 the initial

measured stroke time from step 7.0[6.10.2]

above.

B. RESTROKE and TIME 3-FCV-073-0018 and

RECORD the restroke time on Attachment 10.

C. VERIFY the restroke time recorded on

Attachment 10 is less than or equal to the

maximum value listed. _ _(AC)

NOTE

The removal of the Yokogawa Recorders may be performed in conjunction with the

remainder of the procedure.

[6.11] IF a Yokogawa Recorder was used to measure

3-FCV-73-18, THEN

NOTIFY Electrical Maintenance to REMOVE the

Yokogawa Recorders per Section 2.0 of Attachment 8:

(Otherwise N/A)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 37 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

The next section verifies free movement of turbine overspeed trip tappet trip valve

assembly prior to turbine operation. The trip knob reset action occurs automatically after a

variable time delay with no operator action required.

[6.12] VERIFY free movement of turbine overspeed trip tappet

trip valve assembly as follows:

[6.12.1] LIFT and HOLD HPCI TURBINE MECH TRIP VLV

3-XCV-073-0018 trip knob until Step 7.0[6.12.3].

[6.12.2] CHECK HPCI TURBINE STOP VALVE

3-FCV-73-18 closes by observing 3-ZI-73-18

position indicating lights.

[6.12.3] RELEASE HPCI TURBINE MECH TRIP VLV

3-XCV-073-0018 trip knob.

[6.12.4] CHECK HPCI TURBINE MECH TRIP VLV

3-XCV-073-0018 is reset by observing 3-ZI-73-18

position indicating lights and noting that HPCI

TURBINE STOP VALVE 3-FCV-73-18 reopens.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 38 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

The Target values and ranges in the table below are for information only. If the target is

NOT met, then surveillance testing may proceed with concurrence from Systems

Engineering and Unit Supervisor. 3-PCV-073~0501 may be adjusted by Mechanical

Maintenance to ensure target values are met. Adjustments should be documented in the

Post-Test remarks.

[6.13] CHECK that HPCliube oil pressures listed below are

within the desired range:

Parameter/Indicator Indicated Target

Value

HPCI TURB THRUST BRG

& EGR PRESS INDR psig ~ 15 psig

3-PI-073-0506

HPCI TURB OUTBD

JOURNAL BRG SUPPLY psig ~ 10 psig

3-PI-073-0508

HPCI TURB INBD BRG

SUPPLY PRESS INDR psig ~ 10 psig

3-PI-073-0510

HPCI MAIN PUMP BRG &

SPEED REDUCER SPLY psig ~ 20 psig

3-PI-073-0509

OIL SUPPLY PRESSURE

psig 36-40 psig

3-PI-073-0501A

HPCI OIL FILTER INLET

PRESS IND. psig 85-90 psig

3-PI-073-0053A

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 39 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

The trip knob reset action occurs automatically after a variable time delay with no operator

action required.

[6.14] TEST the HPCI Turbine Mechanical Trip Valve as

follows:

[6.14.1] VERIFY HPCI GOVERNOR CONTROL VALVE

CLOSURE BOOSTER VALVE 3-SHV-73-0707

one-half (1/2) turn open.

1st

2nd

[6.14.2] LIFT and HOLD HPCI TURBINE MECH TRIP VLV

3-XCV-073-0018 trip knob until Step 7.0[6.14.4].

[6.14.3] ADJUST the HPCI TURB OIL INLET THR VLV

FOR 3-PCV-073-0018C, 3-THV-73-714 as required

to obtain:

18-20 psig as indicated on HPCI MECH TRIP VLV

INLET PRESS 3-PI-073-0018B.

[6.14.4] RELEASE HPCI TURBINE MECH TRIP VLV

3-XCV-073-0018 trip knob.

[6.14.5] CHECK that HPCI TURBINE STOP VALVE

3-FCV-73-18 is OPEN by observing 3-ZI-73-18

position indicating lights.

[6.14.6] CHECK that HPCI TURBINE CONTROL VALVE

3-FCV-73-19 is OPEN by observing 3-ZI-73-19

position indicating lights.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 40 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

After releasing the AUX OIL PUMP 3-HS-073-0047B TEST push-button in Step 7.0[6.15],

ALLOW at least one minute for oil to drain back to oil tank before performing

Step 7.0[6.18].

[6.15] RELEASE the HPCI AUX OIL PUMP 3-HS-073-0047B

TEST push-button.

[6.16] CHECK HPCI TURBINE STOP VALVE 3-FCV-73-18

closes by observing 3-ZI-73-18 position indicating lights.

[6.17] CHECK HPCI TURBINE CONTROL VALVE

3-FCV-73-19 closes by observing 3-ZI-73-19 position

indicating lights.

NOTE

During the performance of Step 7.0[6.18], close coordination will be required. REVIEW

Step 7.0[6.18] though Step 7.0[6.21] for clear understanding of the operation of

3-FCV-73-18 and 3-FCV-73-19 upon Aux Oil Pump start.

[6.18] AFTER at least one minute from performing

Step 7.0[6.15], DEPRESS and HOLD the HPCI AUX OIL

PUMP 3-HS-073-0047B TEST push-button until

Step 7.0[6.21].

[6.19] VISUALLY CHECK that turbine control valve

3-FCV-73-19 approaches or reaches the full open

position while the turbine stop valve 3-FCV-73-18 is

closed.

[6.20] VISUALLY CHECK that when the turbine stop valve

3-FCV-73-18 begins to open, the turbine control valve

3-FCV-73-19 is initially driven in the closed direction,

then reverses and proceeds to the full open position

again.

[6.21] RELEASE the HPCI AUX OIL PUMP 3-HS-073-0047B

TEST push-button.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 41 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTES

1) Suppression pool temperature shall be monitored every 5 minutes in accordance with

3-SR-3.6.2.1.1, Suppression Chamber Water Temperature Check, when heat is being

added to suppression pool.

2) Step 7.0[7] thru Step 7.0[12] is performed in preparation of the HPCI Turbine start.

A pre-job brief should be considered at this time.

[7] CALCULATE the Minimum HPCI Main Pump Discharge

Pressure as follows:

[7.1] RECORD pretest suction pressure and thrust bearing

temperature below:

Indicated Acceptable

Parameter/Indicator Value Range

HPCI PUMP SUCT PRESS

psig 2 10 psig

3-PI-073-0028B (3-25-50)

HPCI TURB THR BRG

TEMP 3-TE-73-54F (ICS) of N/A

OR 3-TE-73-54F (3-9-47)

[7.2] RECORD reactor pressure indicated by REACTOR

WIDE RANGE PRESS A 3-PI-3-54 on Panel 3-9-5

below and in Step 7.0[7.3]:

Indicated Value Acceptance Criteria

2,950 psig

psig ~ 1040 psig

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 42 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

The basis for determining the minimum, HPCI main pump discharge pressure is derived

from startup testing performed by STI-15. Specifically, discharge pressure was measured

at 100 psig above reactor pressure for successful injection at rated flow. A 10 psig margin

has been added to this measured value based on engineering judgment to arrive at

110 psig value utilized by this SR.

[7.3] CALCULATE minimum HPCI main pump discharge

pressure required as indicated below:

Reactor Pressure = ---------

psig

(Step 7.0[7.2])

+ 110

(See Note)

Min Disch Press = -------------

psig

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 43 of 104

Date

7.0 PROCEDURE STEPS (continued)

[7.4] RECORD minimum HPCI main pump discharge

pressure calculated in Step 7.0[7.3] in the following

steps:

Step 7.0[13.1]

Step 7.0[14.2]

Step 7.0[17.1]

Attachment 3 (if required)

Step 1.0[17]

Step 1.0[19.2]

[7.5] VERIFY calculation performed in Step 7.0[7.3] is correct

AND pressure value obtained has been correctly

recorded in steps specified by Step 7.0[7.4].

IV

NOTE

Starting the HPCI turbine with HWC in service and flow is NOT at a reduced rate may result

in a higher than Normal Radiation Levels.

[8] VERIFY HWC Flow is at the Desired Setpoint or removed from

service as required by Radcon.

[9] PERFORM the following

VERIFY the M&TE equipment is available and ready to

support HPCI operation.

VERIFY 3-SR-3.6.2.1.1, Suppression Chamber Water

Temperature Check has been commenced.

VERIFY RHR is in Suppression Pool Cooling per 3-01-74

as determined by the Unit Supervisor.

[10] START Standby Gas Treatment System (SGTS) in

accordance with 0-01-65, Standby Gas Treatment System.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 44 of 104

Date

7.0 PROCEDURE STEPS (continued)

[11] ALIGN HPCI System for a manual start by performing the

following steps:

[11.1] CHECK HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33

is in AUTO and SET to control at 500 (5,000 gpm).

IF required, THEN

DEPRESS AUTO operation mode transfer switch and

ADJUST setpoint using Setpoint up/down keys.

[11.2] PLACE HPCI STEAM PACKING EXHAUSTER by

placing 3-HS-73-10A to START.

[11.3] VERIFY 3-FCV-73-36, using HPCI/RCIC CST TEST

VLV 3-HS-73-36A, is OPEN.

[11.4] OPEN 3-FCV-73-35, using HPCI PUMP CST TEST

VLV, 3-HS-73-35A.

WARNING

[NER] Failure of both HPCI steam exhaust piping rupture discs during turbine startup and

operation will result in a process steam release into HPCI Room. This release raises the

risk of personnel injury until steam line isolation occurs. Therefore, personnel in HPCI

Room should minimize stay time in close proximity to rupture disc cage assembly. [IE 93-67]

11**Startcd** CriticaIStep($)

[12] START the HPCI turbine by performing the following:

[12.1] [NER] VERIFY communication is established with

Operations personnel in HPCI Room. [IE 93-67]

[12.2] [NER] REQUEST Operations personnel in HPCI Room, to

ensure that all unnecessary personnel have exited HPCI

Room. [IE 93-67]

[12.3] [NER] ANNOUNCE HPCI turbine startup over plant public

address system. [IE 93-67]

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 45 of 104

Date

7.0 PROCEDURE STEPS (continued)

[12.4] PLACE HPCI AUXILIARY OIL PUMP 3-HS-73-47A to

START.

[12.5] OPEN 3-FCV-73-30, using HPCI PUMP MIN FLOW

VALVE, 3-HS-73-30A.

NOTES

1) Personnel Monitoring the 3-FCV-73-18 valve for smooth operation must pay close

attention to valve travel from the time 3-FCV-73-16 is opened until 3-FCV-73-18 is full

open and stable.

2) Smooth operation for 3-FCV-73-18 is a continuous operation from full close to full

open without erratic movement. Sound can be used to assist in determining operation

of valve. (i.e., The Valve slams open suddenly and then closed and then ramps open

is NOT smooth operation.)

[12.6] ENSURE personnel are ready to monitor 3-FCV-73-18

for smooth operation.

NOTIFY the personnel monitoring that the next step will

open 3-FCV-73-18.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 46 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTES

1) During the startup of the HPCI Turbine a second operator should be utilized to monitor

for abnormal conditions and alarms.

2) The HPCI Turbine parameters should be monitored during HPCI startup. This will

ensure proper response of the control systems. If HPCI pumps suction pressure

causes an auto swap of the HPCI suction valves from CST to the torus, then the HPCI

Turbine should be tripped.

3) REVIEW Step 7.0[12.8] to ensure actions occur when 3-FCV-73-16 opens.

CAUTIONS

1) If HPCI TURBINE STEAM SUPPLY VLV 3-FCV-73-16 fails to fully open, then the

governor control system ramp generator will time out and HPCI turbine speed,

discharge pressure, or flow will be lower than expected.

DO NOT RE-ATTEMPT to open HPCI TURBINE STEAM SUPPLY VLV 3-FCV-73-16

unless HPCI TURBINE STOP VALVE 3-FCV-73-18 is closed using HPCI TURBINE

TRIP 3-HS-73-18A. Failure to observe this caution will result in a turbine overspeed

trip if 3-FCV-73-16 is opened with the ramp generator timed out.

2) During the startup of the HPCI Turbine, the flow indication will remain high during the

transient until the Governor Control System stabilizes the HPCI Flow to the desired

setpoint.

The response time of the Governor Control System is slow. Therefore flow

should NOT be adjusted until the system has stabilized. During this time the

operator should monitor the speed indication for proper operation of the Governor

Control.

The Ramp Generator will cause the Turbine Speed to rise at a steady rate until

the Signal Converter circuit takes control and lowers the speed to stabilize the

flow at the desired setpoint.

3) Starting the HPCI turbine with HWC in service and without the flow being at a reduced

rate may result in higher than Normal Radiation Levels.

[12.7] OPEN 3-FCV-73-16, using HPCI TURBINE STEAM

SUPPLY VLV, 3-HS-73-1.6A.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 47 of 104

Date

7.0 PROCEDURE STEPS (continued)

[12.8] OBSERVE that the following actions occurs:

HPCI AUXILIARY OIL PUMP starts.

[NRC/C] HPCI TURBINE STOP VALVE 3-FCV-73-18

opens by observing 3-ZI-73-18 position indicating

lights. [Appendix R]

HPCI TURBINE CONTROL VALVE 3-FCV-73-19

partially or fully opens by observing 3-ZI-73-19

position indicating lights.

[NRC/C] HPCI PUMP MIN FLOW VALVE

3-FCV-73-30 closes when HPCI SYSTEM

FLOW/CONTROL 3-FIC-73-33 indicates

approximately ~ 125 (~ 1250 gpm) flow. [Appendix R]

HPCI turbine speed rises to greater than 2400 rpm

as indicated on HPCI TURBINE SPEED 3-SI-73-51.

HPCI STM LINE CNDS INBD/OUTBD DR VLVS

3-FCV-73-6A and 3-FCV-73-6B close by observing

3-ZI-73-6A and 3-ZI-73-6B position indicating lights.

HPCI AUXILIARY OIL PUMP stops as turbine

speed rises.

[12.9] VERIFY Smooth operation of 3-FCV-73-18 and mark

results below.

Yes 0 No 0

IF the Answer above is "NO", THEN

NOTIFY System Engineer to initiate a WO and

proceed with test. (Otherwise N/A)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 48 of 104

Date

7.0 PROCEDURE STEPS (continued)

[12.10] VERIFY RESET the following annunciators:

HPCI PUMP DISCH FLOW LOW 3-FA-73-33

(3-XA-55-3F, window 5)

HPCI TURBINE TRIPPED 3-ZA-73-18 (

3-XA-55-3F, window 11)

HPCI TURBINE GLAND SEAL DRAIN PRESSURE

HIGH 3-PA-73-46 (3-XA-55-3F, window 14)

HPCI TURBINE BEARING OIL PRESSURE LOW

3-PA-73-47 (3-XA-55-3F, window 19)

[12.11] VERIFY system flow, discharge pressure, and turbine

speed are stable prior to performing the next step.

II End of Critical Step($)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 49 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTES

1) PAUSE periodically as HPCI discharge pressure approaches the desired test pressure

to allow HPCI system flow, discharge pressure, and turbine speed to stabilize.

BFPER 00-003572-000

2) Due to discharge pressure indicator failures, speed should be monitored NOT to

exceed 4200 rpm to minimize exceeding design discharge pressure.

[13] WHILE maintaining HPCI Turbine Speed less than 4200 rpm,

ADJUST HPCI Pump Discharge Pressure as follows:

[13.1] [NRC/C] SLOWLY THROTTLE 3-FCV-73-35, using HPCI

PUMP CST TEST VLV, 3-HS-73-35A, as necessary,

until the followinq are achieved:

HPCI PUMP DISCH PRESS as indicated on

3-PI-73-31A is

psig

(Step 7.0[7.3])

Discharge flow steadies at or above 500

(5,000 gpm) as indicated by HPCI SYSTEM

FLOW/CONTROL 3-FIC-73-33. [Appendix R]

[13.2] [NRC/C] CHECK HPCI Room for evidence of steam, oil,

and gland seal condenser leaks.

[13.3] REQUEST RADCON to monitor radiation and

contamination levels to ensure either has NOT risen

significantly. [RPT-82-13]

[13.4] VERIFY system flow, discharge pressure, and turbine

speed are stable prior to performing the next step.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 50 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

Steps 7.0[14.1],7.0[14.2] and 7.0[14.3] may be performed concurrently.

CAUTION

HPCI main/booster pump bearing temperatures shall NOT be allowed to exceed 155°F.

[14] MONITOR and OBTAIN the following data:

[14.1] MONITOR the following HPCI turbine and pump set

temperatures using HPCI/RCIC/RFW TEMPERATURES

3-TR-73-54 on Panel 3-9-47 or ICS to verify

temperatures are NOT rising rapidly.

CHECK that no temperature exceeds 155°F:

PARAMETER INST CHANNEL

HPCI OIL COOLER DISCH 3-TE-73-54A

HPCI TURB HP BRG OIL 3-TE-73-54D (Gov End)

HPCI TURB LP BRG OIL 3-TE-73-54E (Cplg End)

HPCI TURB THRUST BRG 3-TE-73-54F

HPCI PUMP INBOARD BRG 3-TE-73-54G

HPCI PUMP OUTBOARD BRG 3-TE-73-54H

~ HPCI SPEED INCREASER 3-TE-73-54J

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 51 of 104

Date

7.0 PROCEDURE STEPS (continued)

[14.2] RECORD following data:

Indicated Acceptable

Parameter/I nd icator Value Range

HPCI SYSTEM FLOW

gpm ~ 5,000 gpm

3-FIC-73-33 or ICS

HPCI PUMP DISCH PRESS ~

psig

3-PI-73-31A (Step 7.0[7.3])

HPCI TURBINE SPEED

rpm ~ 2,400 rpm

3-SI-73-51

HPCI TURB EXH PRESS

3-PI-73-21A

psig < 40 psig

REACTOR WIDE RANGE ~ 950 psig

psig

PRESS A 3-PI-3-54 ~ 1040 psig

_ _(AC)

[14.3] RECORD following data:

Indicated Acceptable

Parameter/Indicator Value Range

HPCI PUMP SUeT Press

psig ~ 10 psig

3-PI-73-28A

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 52 of 104

Date

7.0 PROCEDURE STEPS (continued)

[15] OBTAIN ASME OM Code data for HPCI main and booster

pump set as follows:

[15.1] PLACE HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33

in MANUAL as follows:

DEPRESS the MANUAL operation mode transfer switch

on 3-FIC-73-33.

[15.2] ADJUST HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33

MANUAL operation lever, until approximately 3800 rpm

on HPCI TURBINE SPEED 3-SI-73-51.

[15.3] ADJUST HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33

MANUAL operation lever, to achieve 3790 to 3810 rpm

turbine speed, using hand held tachometer.

RECORD final turbine speed below:

HPCI Turbine Speed (M&TE) rpm

[15.4] VERIFYHPCI test condition flow rate as follows:

[15.4.1] IF ICS is utilized to obtain HPCI flow rate data,

THEN

CHECK that no gross instrument channel failures

have occurred by noting that ICS-displayed HPCI

flow rate is within 100 gpm of flow rate indicated on

HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33.

(Otherwise N/A)

[15.4.2] THROTTLE 3-FCV-73-35, using HPCI PUMP CST

TEST VLV, 3-HS-73-35A to obtain either of the

following:

An ICS display reading of 4950 to 5050 gpm.

495 to 505 (4950 to 5050 gpm) as indicated on

HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 53 of 104

Date

7.0 PROCEDURE STEPS (continued)

[15.5] ALLOW HPCI pump set to operate until steady-state

conditions are achieved, THEN

VERIFY conditions of Steps 7.0[15.3] and 7.0[15.4] are

met.

NOTE

The indicator on 3-PI-73-31 B may oscillate due to pump generated pressure pulses.

Should this condition exist, an average between the predominate high and low readings

should be recorded as the indicated value.

[15.6] OBTAIN the HPCI pump data as follows:

[15.6.1] On Panel 3-LPNL-25-0050

PERFORM the following:

A. OBSERVE 3-PI-73-31 B, while performing the

following to verify unobstructed

instrumentation.

CLOSE and OPEN PANEL ISOL VLV TO

3-PI~73-31 B, 3-PISV-73-9013 several times.

B. IF required to stabilize 3-PI-73-31 B indicator,

THEN

THROTTLE PANEL ISOL VLV TO

3-PI-73-31 B, 3-PISV-73-9013, as required to

stabilize 3-PI-73-31 B. (Otherwise N/A)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 54 of 104

Date

7.0 PROCEDURE STEPS (continued)

CAUTIONS

1) The HPCI pump set differential pressure is very sensitive to minor turbine speed and

pump set flow adjustments. Therefore, it is anticipated that the UO will be required to

make minor speed and flow rate adjustments in order to properly establish the ASME

OM Code operating point.

2) HPCI pump discharge pressure has no required range because it is a function of pump

speed and flow.

[15.6.2] COMPLETE following table entries stipulated

below:

Indicated Required

Parameter/Indicator Value Value

HPCI SYSTEM FLOW gpm 4,950-

3-FIC-73-33 or ICS 5,050 gpm

HPCI MAIN PUMP DISCH psig SEE

PRESS 3-PI-073-0031 B CAUTION

(HPCI RM) ABOVE

HPCI TURBINE SPEED rpm 3790-3810

HAND-HELD rpm

TACHOMETER

HPCI TURB EXH PRESS psig ~ 40 psig

3-PI-73-21A

REACTOR WIDE RANGE psig ~ 950 psig

PRESS A 3-PI-3-54

~ 1040 psig

HPCI PUMP SUCTION psig ~ 10 psig

PRESS 3-PI-073-0028B

(HPCI RM)

_ _(AC)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 55 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

Calculation independent verification (IV) consists of verifying arithmetic for accuracy and

arithmetic inputs have been properly transferred between the steps within the surveillance.

IV is NOT required to verify pressure data recorded at local instrument rack is correct.

[15.7] CALCULATE HPCI pump set differential pressure as

follows:

[15.7.1] Using applicable data recorded in Step 7.0[15.6.2],

CALCULATE HPCI pump set differential pressure:

Discharge Pressure (3-PI-73-31 B) psig

Suction Pressure (3-PI-73-28B) psig

Differential Pressure = psid

[15.7.2] VERIFY that the differential pressure calculated is

~ 1034 and ~ 1201 psid. _ _(AC)

[15.7.3] INDEPENDENTLY VERIFY HPCI pump set

differential pressure calculation is correct.

IV

[15.7.4] IF acceptance criteria is NOT met at

Step 7.0[15.7.2], THEN

NOTIFY the Unit Supervisor that the Unit 3 HPCI

pump is INOPERABLE due to low or high

differential pressure (N/A otherwise).

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 56 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTES

1) The HPCliube oil system oil filter inlet pressure minus outlet pressure shall NOT be

greater than 12 psi.

2) The target values and ranges in the table below are for information only. If the target

is NOT met, the out of range readings should be observed periodically to ensure that

the readings are NOT changing at a rate that could result in loss of oil pressure.

3) The data gathered in the following steps may be obtained concurrently with the

vibration data at Step 7.0[15.10].

[15.8] RECORD the following process data values obtained

locally at HPCI turbine:

Parameter/Indicator Indicated Target Value

Value

HPCI TURB THRUST BRG

& EGR SUPPLY PRESS psig ~ 13 psig

INDR 3-PI-73-506

HPCI TURB OUTBD

JOURNAL BRG SUPPLY psig ~ 8 psig

3-PI-73-508

HPCI TURB INBD BRG

SUPPLY PRESS INDR psig ~ 8 psig

3-PI-73-510

HPCI MAIN PUMP BRG &

SPEED'REDUCER SPLY psig ~ 18 psig

3-PI-73-50'9

HPCI MAIN OIL PUMP

DISCH PRESS INDR psig 105-110 psig

3-PI-73-505

HPCI OIL FILTER INLET

psig SEE NOTE 1

PRESS INDR 3-PI-73-53A

HPCI OIL FILTER OUTLET

psig SEE NOTE 1

PRESS INDR 3-PI-73-53B

HPCI OIL SUPPLY PRESS

psig 36-40 psig

INDICATOR 3-PI-73-501A

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 57 of 104

Date

7.0 PROCEDURE STEPS (continued)

[15.9] At the HPCI turbine skid:, CHECK Local HPCI Oil

Temperature 3-TI-073-1152 indication does NOT exceed

155°F.

RECORD the HPCI Oil Temperature in the table below.

Parameter/lnst Channel Indicated Value

HPCI LUBE OIL TEMP of

3-TI-073-1152

[15.10] OBTAIN HPCI turbine and pump set vibration levels and

RECORD data in table below: (N/A if NOT required)

VIBS Point Measured Value

CH in/sec

CV in/sec

CA in/sec

DH in/sec

DV in/sec

DA in/sec

EH in/sec

EV in/sec

FV in/sec

GH in/sec

GV in/sec

HH in/sec

HV in/sec

HA in/sec

EM

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 58 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTES

1) Steps 7.0[16.1] through 7.0[16.6] may be repeated to obtain accurate pressure drop

readings.

2) Two personnel that will be used to install jumpers on Panel 3-9-39 may be dispatched

at this time.

3) Opening 3-FCV-73-30 while HPCI pump set is operating at design flow will result in a

considerable rise in HPCI Room noise as flow in minimum flow line rises. This is an

expected condition which should be noted.

[16] PERFORM the following minimum flow function testing:

[16.1] RECORD below HPCI pump discharge pressure

measured in HPCI Room by 3-PI-73-31 B on Instrument

Rack 3-25-50:

HPCI Pump Disch Press psig

[16.2] NOTIFY Operations personnel in HPCI Room to monitor

HPCI pump discharge pressure measured by

3-PI-73-31 B on Instrument Rack 3-25-50 when HPCI

MIN FLOW VALVE 3-FCV-73-30 reaches open position.

[16.3] OPEN 3-FCV-73-30 as follows:

MOMENTARILY PLACE HPCI PUMP MIN FLOW

VALVE, 3-HS-73-30A in the OPEN position.

[16.4] RECORD below the lowest HPCI pump discharge

pressure measured in HPCI Room by 3-PI-73-31 B on

Instrument Rack 3-25-50.

HPCI Pump Disch Press psig

[16.5] CHECK HPCI PUMP MIN FLOW VALVE 3-FCV-73-30

has re-closed after stroking full open.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 59 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

Calculation independent verification (IV) consists of verifying arithmetic for accuracy and

arithmetic inputs have been properly transferred between the steps within the surveillance.

IV is NOT required to verify pressure data recorded at local instrument rack is correct.

[16.6] PERFORM the following to calculate the change in HPCI

pump set discharge pressure

[16.6.1] CALCULATE change in HPCI pump set discharge

pressure as stipulated below:

Initial Discharge Pressure psig

(Step 7.0[16.1])

Lowest Discharge Pressure psig

(Step 7.0[16.4])

Discharge Pressure Change = psig

[16.6.2] INDEPENDENTLY VERIFY pressure drop

calculation performed in above is correct.

IV

NOTE

Verification that discharge pressure change meets the acceptance criteria stipulated in

following step provides positive confirmation that HPCI PUMP MIN FLOW CHECK VALVE

3-CKV-73-559 has opened sufficiently to perform its intended desiqn function.

[16.7] CHECK that discharge pressure change recorded in

Step 7.0[16.6] is ~ 70 psig. _ _(AC)

[16.8] ADJUST HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33

MANUAL operation lever, until a turbine speed of

approximately 3000 rpm is indicated by HPCI TURBINE

SPEED 3-SI-73-51.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 60 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

Installation of the following jumper simulates the presence of a HPCI initiation signal which

allows the minimum flow valve to open on low flow.

[16.9] PLACE jumper across 3-RLY-073-23A-K24

Contacts 11-12 in Panel 3-9-39.

REFER TO Attachment 7.

1st

2nd

CAUTION

Throttling HPCI flow in the following step will result in the minimum flow valve opening.

This will cause rapid filling of the torus. Therefore, UO should ensure that jumper is

removed and minimum flow valve closed as quickly as possible to minimize torus filling.

[16.10] THROTTLE 3-FCV-73-35, using HPCI PUMP CST

TEST VLV, 3-HS-73-35A until HPCI SYSTEM

FLOW/CONTROL 3-FIC-73-33 indicates approximately

70 (700 gpm).

[16.11] [NRC/C] CHECK that HPCI PUMP MIN FLOW VALVE

3-FCV-73-30 is OPEN. [Appendix R] _ _(AC)

[16.12] REMOVE jumper placed across 3-RLY-073-23A-K24

Contacts 11-12 in Panel 3-9-39.

1st

2nd

[16.13] THROTTLE 3-FCV-73-35, using HPCI PUMP CST

TEST VLV, 3-HS-73-35A until HPCI SYSTEM

FLOW/CONTROL 3-FIC-73-33, indicates between 400

and 500 (4000 and 5000 gpm).

[16.14] VERIFY CLOSED 3-FCV-73~30 using HPCI PUMP MIN

FLOW VALVE, 3-HS-73-30A.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 61 of 104

Date

7.0 PROCEDURE STEPS (continued)

[16.15] VERIFY HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33

setpoint flow is within 50 gpm of indicated flow utilizing

setpoint up/down key adjustments.

[16.16] DEPRESS AUTO operation mode transfer switch on

HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33.

AND

ADJUST HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33

setpoint using Setpoint up/down keys to 500

(5,000 gpm).

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 62 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTES

1) Time To Achieve Rated Flow And Pressure will be performed by Attachment 3

following the HPCI Turbine Trip if required. The following steps will adjust the

3-FCV-73-35 to the required position and will have a Caution Order placed to control

the desired position.

2) Adjustments made in Step 7.0[17] should allow time for the system to stabilize prior to

making further adjustments. This may require several attempts to ensure both

conditions in Step 7.0[17] are met.

3) Due to discharge pressure indicator failures, speed should be monitored NOT to

exceed 4200 rpm to minimize exceeding design discharge pressure.

[17] IF Time To Achieve Rated Flow And Pressure is to be

performed (REFER TO Step 4.0[8]), THEN

PERFORM the following: (Otherwise N/A)

[17.1] WHILE maintaining HPCI Turbine Speed less than

4200 rpm, THROTTLE 3-FCV-73-35, using HPCI PUMP

CST TEST VLV, 3-HS-73-35A, as necessary, until the

following conditions are met:

HPCI PUMP DISCH PRESS 3-PI-73-31A reads

psig

(Step 7.0[7.3])

HPCI discharge flow steadies at or above 500

(5,000 gpm) as indicated by HPCI SYSTEM

FLOW/CONTROL 3-FIC-73-33.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 63 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTES

1) [NRC/C] CONSULT Step 3.0Q for additional background information regarding HPCI

System removal from operable service. [NCO 89-0216-002]

2) The intent of Step 7.0[18] is to depress and hold the trip push-button for thirty seconds,

verify the alarms, close 3-FCV-73-16, observe the aux. oil pump starts, then release

the push-button.

3) During the HPCI Turbine trip a second operator should be utilized to monitor for

abnormal conditions and alarms.

4) HPCI PUMP DISCH FLOW LOW 3-FA-73-33 (3-XA-55-3F, window 5) needs to be

verified prior to 3-FCV-73-16 becoming full close.

[18] PERFORM the following steps to shutdown HPCI turbine:

[18.1] VERIFY HPCI System has been declared inoperable

and ENTER appropriate LCO information into Narrative

log as required.

US

[18.2] DEPRESS and HOLD HPCI TURBINE TRIP

3-HS-73-18A until Step 7.0[18.8] is performed.

[18.3] 'WAIT 30 seconds and OBSERVE following

annunciators are in ALARM:

[NRC/C] HPCI TURBINE TRIP'PED 3-ZA-73-18

(3-XA-55-3F, window 11). [Appendix R]

HPCI PUMP DISCH FLOW LOW 3-FA-73-33

(3-XA-55-3F, window 5).

[18.4] CLOSE 3-FCV-73-16, using HPCI TURBINE STEAM

SUPPLYVLV,3-HS-73-16A.

[18.5] OBSERVE HPCI AUXILIARY OIL PUMP starts as

turbine slows.

[18.6] OBSERVE HPCI TURBINE SPEED 3-SI-73-51, reading

lowers to approximately zero.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 64 of 104

Date

7.0 PROCEDURE STEPS (continued)

[18.7] VERIFY HPCI TURBINE STEAM SUPPLY VLV

3-FCV-73-16 is closed.

[18.8] RELEASE HPCI TURBINE TRIP, 3-HS-73-18A.

[18.9] RECORD below and on Attachment 3 (if required) the

time of HPCI Turbine shutdown:

Time

[18.10] On Panel 3-LPNL-25-0050, OPEN or VERIFY OPEN

PANEL ISOL VLV TO 3-PI-73-31 B, 3-PISV-73-9013.

[19] VERIFY RESET the following annunciators:

HPCI TURBINE TRIPPED (3-XA-55-3F, window 11)

HPCI PUMP DISCH FLOW LOW (3-XA-55-3F, window 5)

[20] VERIFY HPCI STM LINE CNDS INBD/OUTBD DR VLVS

3-FCV-73-6A and 3-FCV-73-6B are OPEN by observing

3-ZI-73-6A and 3-ZI-73-6B position indicating lights.

NOTES

1) Step 7.0[21] should be reviewed prior to performance to ensure proper operation of

system.

2) Two people are needed to perform the 3-FCV-73-18 time delay test.

[21] PERFORM the following at HPCI turbine:

[21.1] Using the 3-FCV-73-18 valve stem position, CHECK

HPCI TURBINE STOP VALVE 3-FCV-73-18 is OPEN.

[21.2] LIFT and IMMEDIATELY RELEASE HPCI TURBINE

MECH TRIP VLV 3-XCV-73-18 trip knob.

[21.3] Using the 3-FCV-73-18 valve stem position, CHECK

HPCI TURBINE STOP VALVE 3-FCV-73-18 closes.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 65 of 104

Date

7.0 PROCEDURE STEPS (continued)

[21.4] START the stop watch when HPCI TURBINE STOP

VALVE, 3-FCV-73-18 is full closed.

[21.5] STOP the stop watch when the HPCI TURBINE STOP

VALVE 3-FCV-73-18 begins to open

RECORD time delay.

Time seconds

[21.6] IF the 3-FCV-73-18 time delay in Step 7.0[21.5] is NOT

within 4-6 seconds, THEN

PERFORM Attachment 4, 3-FCV-73-18 TIME DELAY

ADJUSTMENT. (Otherwise N/A.)

[22] IF time to achieve rated flow and pressure is to be verified

(REFER TO Step 4.0[8]), THEN

PERFORM the following: (Otherwise N/A this section)

[22.1] PERFORM Attachment 3.

[22.2] WHEN Attachment 3 is completed, THEN

CONTINUE in this procedure.

[23] CLOSE 3-FCV-73-35 using HPCI PUMP CST TEST VLV,

3-HS-73-35A.

[24] CLOSE 3-FCV-73-36, using HPCI/RCIC CST TEST VLV,

3-HS-73-36A.

[25] CHECK HPCI PUMP MIN F,LOWVALVE 3-FCV-73-30 is

CLOSED.

[26] CHECK HPCI PUMP INJECTION VLV 3-FCV-73-44 is

CLOSED.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 66 of 104

Date

7.0 PROCEDURE STEPS (continued)

[27] VERIFY HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33 is in

AUTO position.

IF required, THEN

DEPRESS AUTO operation mode transfer switch.

[28] VERIFY HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33 is set

to control at 500 (5,000 gpm).

IF required, THEN

ADJUST setpoint using Setpoint up/down keys.

NOTES

1) Care must be exercised to ensure that HPCI OIL TANK DRAIN 3-DRV-073-0703 is

cleaned with a clean rag and solvent to remove any impurities/contaminants that could

make their way into oil sample.

2) Pipe dope/sealant shall NOT be utilized for reinstallation of HPCI OIL TANK DRAIN

3-DRV-073-0703 pipe plug. This material is NOT required and serves only to,

contaminate oil samples.

3) Site Engineering will review and evaluate oil sample analysis as required per CI-130 to

determine if a Work Order is required to correct an oil quality deficiency.

[29] OBTAIN an Oil Sample with the Aux Oil Pump still running to

ensure thorough mixing as follows.

[29.1] OBTAIN two, one liter sample bottles from Chemistry

Lab for obtaining HPCI lube oil sample.

MM

[29.2] REMOVE pipe plug from HPCI OIL TANK DRAIN

3-DRV-073-0703.

MM

[29.3] OPEN HPCI OIL TANK DRAIN 3-DRV-073-0703 and

REMOVE two, one liter HPCliube oil samples.

MM

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 67 of 104

Date

7.0 PROCEDURE STEPS (continued)

[29.4] LABEL the first bottle of lube oil as FLUSH and LABEL

the second bottle of lube oil as SAMPLE.

MM

NOTES

1) Care must be exercised to ensure that HPCI OIL TANK DRAIN 3-DRV-073-0703 is

cleaned with a clean rag and solvent to remove any impurities/contaminants that could

make their way into oil sample.

2) Pipe dope/sealant shall NOT be utilized for reinstallation of HPCI OIL TANK DRAIN

3-DRV-073-0703 pipe plug. This material is NOT required and serves only to

contaminate oil samples.

3) Site Engineering will review and evaluate oil sample analysis as required per CI-130 to

determine if a Work Order is required to correct an oil quality deficiency.

[29.5] CLOSE HPCI OIL TANK DRAIN 3-DRV-073-0703 and

REINSTALL pipe plug in end of valve housing.

1st MM

2nd MM

[29.6] DELIVER HPCI lube oil bottle labeled SAMPLE to

Chemistry Lab for analysis and HPCI lube oil bottle

labeled FLUSH for disposal and RECORD delivery time

below:

Date Time

MM

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 68 of 104

Date

7.0 PROCEDURE STEPS (continued)

CAUTION

HPCI TURBINE SPEED 3-SI-73-51 could indicate zero rpm while turbine shaft is still

rotating. The auxiliary oil pump should NOT be stopped until visual confirmation is made

by personnel that turbine speed is zero.

[30] PERFORM the following after allowing approximately

15 minutes to pass after turbine shutdown:

[30.1] STOP HPCI AUXILIARY OIL PUMP and RETURN

3-HS-73-47A to AUTO position.

[30.2] STOP HPCI STEAM PACKING EXHAUSTER and

RETURN 3-HS-73-10A to AUTO position.

[30.3] CHECK HPCI TURBINE STOP VALVE 3-FCV-73-18 is

CLOSED by observing 3-ZI-73-18 position indicating

lights.

[30.4] CHECK HPCI TURBINE CONTROL VALVE

3-FCV-73-19 is CLOSED by observing 3-ZI-73-19

position indicating lights.

[31] EXIT HPCI System LCO by updating Narrative log.

US

[32] IF SGTS is no longer required, THEN

SHUT DOWN SGTS. REFER TO 0-01-65, Standby Gas

Treatment System. (Otherwise N/A)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 69 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTES

1) The following independent verifications are performed to ensure compliance with

SPP-10.3. First party verifications have already been performed previous to this step

and, therefore, have NOT been duplicated.

2) The independent verifications of the following step may be performed in any order.

3) If a deficiency(s) is identified during performance of the independent verifications in the

following step, the independent verifier shall stop and notify the Unit Supervisor

immediately for further instructions prior to correcting the deficient condition(s).

4) Successful completion of the following IVs returns HPCI System to its standby

readiness configuration.

[33] INDEPENDENTLY VERIFY on Panel 3-9-3:

[33.1] VERIFY HPCI TURBINE STEAM SUPPLY VLV

3-FCV-73-16 is CLOSED.

IV

[33.2] VERIFY HPCI PUMP CST TEST VLV 3-FCV-73-35 is

CLOSED.

IV

[33.3] VERIFY HPCI/RCIC CST TEST VLV 3-FCV-73-36 is

CLOSED.

IV

[33.4] VERIFY HPCI PUMP MIN FLOW VALVE 3-FCV-73-30

is CLOSED.

IV

[33.5] VERIFY HPCI PUMP INJECTION VALVE 3-FCV-73-44

is CLOSED.

IV

[33.6] VERIFY HPCI STEAM LINE INBD DRAIN VLV

3-FCV-73-6A is OPEN by observing 3-ZI-73-6A position

indicating lights.

IV

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 70 of 104

Date

7.0 PROCEDURE STEPS (continued)

[33.7] VERIFY HPCI STEAM LINE OUTBD DRAIN VLV

3-FCV-73-6B is OPEN by observing 3-ZI-73-6B position

indicating lights.

IV

[33.8] VERIFY HPCI TURBINE STOP VALVE 3-FCV-73-18 is

CLOSED by observing 3-ZI-73-18 position indicating

lights.

IV

[33.9] VERIFY HPCI TURBINE CONTROL VALVE

3-FCV-73-19 is CLOSED by observing 3-ZI-73-19

position indicating lights.

IV

[33.10] VERIFY HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33

is in AUTO and set to control at 500 (5,000 gpm).

IV

[33.11 ] VERIFY HPCI STEAM PACKING EXHAUSTER

3-HS-73-10A is in AUTO.

IV

[33.12] VERIFY HPCI AUXILIAR-Y OIL PUMP 3-HS-73-47A is in

AUTO.

IV

[34] INDEPENDENTLY VERIFY in the Auxiliary Instrument Room:

[34.1] IF Attachment 3 was performed, THEN

VERIFY no jumper is installed across

3-RLY-073-23A-K47 Contacts 1-2 in Panel 3-9-39.

(Otherwise N/A.)

IV

[34.2] VERIFY no jumper is installed across

3-RLY-073-23A-K24 Contacts 11-12 in Panel 3-9-39.

IV

[35] At Panel 3-LPNL-25-50 in the HPCI room,

INDEPENDENTLY VERIFY PANEL ISOL VLV TO

3-PI-73-31 B, 3-PISV-73-9013, is OPEN.

IV

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 71 of 104

Date

7.0 PROCEDURE STEPS (continued)

NOTE

ALLOW approximately one hour to pass before checking HPCI lube oil skid reservoir level

to ensure that oil in system has drained sufficiently to provide an accurate level reading.

[36] AFTER approximately one hour from HPCI Turbine shutdown:

PERFORM the following inspections:

VERIFY HPCliube oil skid reservoir level is per

Attachment 6.

CHECK oil level in HPCI booster pump inboard and

outboard bearing oil sight glasses is per Attachment 6.

[37] IF a Yokogawa Recorder was used to measure 3-FCV-73-18,

THEN

PERFORM the following: (Otherwise N/A)

VERIFY Attachment 8 is completed.

ATTACH the Chart Paper used for Timing the

3-FCV-73-18 to this procedure.

[38] IF restroking of 3-FCV-073-0018 was required during this

surveillance performance, (i.e., restroke time was recorded on

Attachment 10), THEN

PERFORM the following: (Otherwise N/A this step)

[38.1] NOTIFY Duty Maintenance Manager to:

OBTAIN a copy of Attachment 10 for delivery to the

ASME 1ST Program owner.

AND

CONTACT the Duty System Engineer to notify

ASME 1ST Program owner for evaluation of test

results.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 72 of 104

Date

7.0 PROCEDURE STEPS (continued)

[38.2] RECORD time below

Time: - - - - - " - -

[39] COMPLETE Attachment 1, Surveillance Procedure Review

Form, up to Unit Supervisor Review.

[40] NOTIFY UO that this surveillance is complete.

[41] NOTIFY US that this surveillance is complete.

8.0 ILLUSTRATIONS/ATTACHMENTS

Illustration 1 - Process for Stroke Timing Valves Per ASME OM Code

Attachment 1: Surveillance Procedure Review Form

Attachment 2: HPCI Venting

Attachment 3: HPCI Cold Quick Start

Attachment 4: 3-FCV-73-18 Time Delay Adjustment

Attachment 5: ASME OM Code Inservice Testing Review Form

Attachment 6: HPCI Lube Oil Skid and Booster Pump Oil Level Settings

Attachment 7: HFA Relay Contact Layout

Attachment 8: Installation and Removal of Yokogawa Recorders For 3-FCV-73-18

Attachment 9 - Annunciators Affected By Surveillance Procedure Performance

Attachment 10 - ASME OM Code Restroke Time Record Form

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 73 of 104

Illustration 1

(Page 1 of 1)

Process for Stroke Timing Valves Per ASME OM Code

CORRECTIVE STROKE TIME

DECLARE VALVE .-----..1

ACTION COMPLETE t------II VALVE PER SR

INOPERABLE AND

INITIATE

CORRECTIVE

ACTION

ENGINEERING

REVISE REFERENCE

STROKE TIME AND

INITIATE

PROCEDURE

CHANGE' RESTROKE TIME WITHIN

MAXIMUM LIMIT?

ENGINEERING

RESTROKE TIME DOCUMENTS

WITHIN NORMAL CAUSE OF STROKE

RANGE?

TIME VARIANCE

TEST STROKE TI ME

ACCEPTABLE?

ENGINEERING

INITIATES ANY

PROVIDE RESTROKE

ENGINEERING

TIME DATA TO

REQUIRED

EVALUATES STROKE CORRECTIVE

ENGINEERING

TIME WITHIN 96 HOURS ACTIONS

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 74 of 104

Attachment 1

(Page 1 of 2)

Surveillance Procedure Review Form

REASON FOR TEST: DATEITIME STARTED

D Scheduled Surveillance DATEITIME COMPLETED

D System Inoperable (Explain in Remarks) PLANT CONDITIONS

D Maintenance (WO No. - - - - - - -

D Other (Explain in Remarks)

PRE-TEST REMARKS:

PERFORMED BY:

Initials Name (Print) Name (Signature)

(Test Dir/Lead Perf)

Delays or Problems (If yes, explain in POST-TEST REMARKS)? DYes DNo

Acceptance Criteria Satisfied? DYes DNo

If the above answer is no, the Unit Supervisor shall

determine if an LCO exists. LCO DYes DNa

UNIT SUPERVISOR

- - - - - - - - - - - - - - - - - Date- - - - -

INDEPENDENT REVIEWER (OPS) Date

------

SCHEDULING COORDINATOR Date

POST-TEST REMARKS:

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 75 of 104

Attachment 1

(Page 2 of 2)

Surveillance Procedure Review Form

Continuation Page

PERFORMED BY:

Initials Name (Print) . Name (Signature)

POST-TEST REMARKS (Continued):

The SR Key number is a Cross Reference only and is not part of the procedure. Key # 3352A I

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 76 of 104

Attachment 2

(Page 1 of 7)

HPCI Venting

Date

1.0 HPCI VENTING INSTRUCTIONS

NOTES

1) The HPCI vent station is located on Elevation 565' of the Reactor Building near column

lines at R16-N..

2) This attachment requires IV of valves at the vent station.

3) A digital thermometer or equivalent device may be obtained from Hot Tool Room.

WARNING

The HPCI vent line piping may contain hot feedwater. Care shall be taken when working

around this potentially hot piping due to the possibility of a burn hazard existing.

[1] VERIFY the following valve positions from 3-PNL-9-3:

HPCI PUMP INJECTION VALVE 3-FCV-73-44 is

CLOSED.

HPCI SUPPR POOL OUTBD SUCT VLV 3-FCV-73-27 is

CLOSED.

HPCI SUPPR POOL INBD SUCT VLV 3-FCV-73-26 is

CLOSED.

HPCI CST SUCTION VALVE 3-FCV-73-40 is OPEN~

HPCI PUMP DISCHARGE VALVE 3-FCV-73-34 is OPEN.

[2] VERIFY CNDS SPLY TO SAFETY SYSTEMS 3-SHV-002-705

is LOCKED OPEN, locally at Elevation 541.5' of the

NEquadrant of the Reactor Building.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 77 of 104

Attachment 2

(Page 2 of 7)

HPCI Venting

Date

1.0 HPCI VENTING INSTRUCTIONS (continued)

[3] VERIFY the following valve positions from 3-PNL-9-6:

CNDS DISCHARGE NORMAL HDR VLV 3-FCV-2-167 is

OPEN by noting 3-ZI-2-167 position indicating lights on

Panel 3-9-6.

CNDS DISCHARGE EMERGENCY HDR VLV

3-FCV-2-166 is OPEN by noting 3-ZI-2-166 position

indicating lights on Panel 3-9-6.

[4] THROTTLE HPCI HIGH POINT TELL-TALE VENT SOV,

3-SHV-073-0552, approximately four turns open.

[5] PLACE thermometer probe on unpainted portion of vent line

piping near HPCI HIGH POINT TELL-TALE VENT,

3-FSV-073-0062.

[6] DEPRESS and HOLD HPCI HIGH POINT VENT PUMP

DISCH, 3-HS-073-0062, until Step 1.0[8].

[7] AFTER 60 seconds, THEN

MONITOR surface temperature of vent line piping near

3-FSV-73-62 and RECORD temperature below.

Temperature of

[8] RELEASE HPCI HIGH POINT VENT PUMP DISCH,

3-HS-073-0062.

[9] REMOVE thermometer probe.

[10] CHECK the surface temperature recorded in Step 1.0[7] of this

attachment is less than 255°F.

[11] CLOSE HPCI HIGH POINT TELL-TALE VENT SOV,

3-SHV-073-0552.

[12] OPEN 3-FCV-73-36, using HPCI/RCIC CST TEST VLV,

3-HS-73-36A.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 78 of 104

Attachment 2

(Page 3 of 7)

HPCI Venting

Date

1.0 HPCI VENTING INSTRUCTIONS (continued)

[13] OPEN 3-FCV-73-35, using HPCI PUMP CST TEST VLV,

3-HS-73-35A.

CAUTION

While opening HPCI PUMP INJECTION VALVE 3-FCV-73-44, HPCI discharge piping

pressure must be monitored using 3-PI-73-31A on Panel 3-9-3. If discharge pressure

readings equal or exceed a nominal value of 55 psig, HPCI PUMP INJECTION VALVE

3-FCV-73-44 shall be promptly closed and the Unit Supervisor contacted for additional

instructions prior to proceeding with venting since this condition may indicate a gross failure

of HPCI TESTABLE CHECK VLV 3-FCV-73-45.

[14] OPEN 3-FCV-73-44, using HPCI PUMP INJECTION VALVE,

3-HS-73-44A.

[15] MONITOR HPCI PUMP DISCH PRESS, 3-PI-73-31A on

Panel 3-9-3.

[16] IF HPCI PUMP DISCH PRESS, 3-PI-73-31A, exceeds 55 psig,

THEN

PERFORM the following: (N/A this section if 55 psig is NOT

exceeded.)

[16.1] CLOSE the HPCI PUMP INJECTION VALVE,

3-FCV-73-44.

[16.2] NOTIFY the Unit Supervisor contacted for additional

instructions prior to proceeding with venting.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 79 of 104

Attachment 2

(Page 4 of 7)

HPCI Venting

Date

1.0 HPCI VENTING INSTRUCTIONS (continued)

CAUTION

A sight glass blowout potential exists while performing the next steps. Stand clear of the

flow sight glass when first depressing 3-HS-73-63. If sight glass blows out or minimal flow

cannot be observed, this may indicate HPCI TESTABLE CHECK VLV 3-FCV-73-45

leakage.

[17] OPEN HPCI HIGH POINT TELL-TALE VENT SOV,

3-SHV-073-0551.

[18] PLACE thermometer probe on unpainted portion of vent line

piping near HPCI HIGH POINT TELL-TALE VENT,

3-FSV-073-0063.

[19] STATION personnel near the HIGH POINT VENT TELL-TALE

SIGHT GLASS, 3-FG-073-0513. (REFER TO Step 3.0GG and

the caution above.)

[20] VERIFY personnel involved in the venting have reviewed and

understands the indications and response that can be used by

Step 3.0GG, during the performance of Step 1.0[22].

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044.

Rated Reactor Pressure Page 80 of 104

Attachment 2

(Page 5 of 7)

HPCI Venting

Date

1.0 HPCI VENTING INSTRUCTIONS (continued)

NOTES

1) Steps 1.0[21] thru Step 1.0[23] should be performed concurrently to ensure flow is

observed in the sightglass.

2) If a steady flow of water cannot be observed from tell-tale sight flow indicator, HPCI

System must be declared inoperable.

CAUTIONS

1) [NRC/C] If vent line surface temperature is > 240°F, STOP and CONTACT the Unit

Supervisor for additional instructions prior to proceeding in the procedure.

2) A high surface temperature of > 240°F may indicate excessive feedwater leakage past

HPCI TESTABLE CHECK VLV 3-FCV-73-45. [NRC Information Notice 89~080]

[21] DEPRESS and HOLD HPCI HIGH POINT VENT TELL TALE,

3-HS-073-0063, until Step 1.0[24] of this attachment.

[22] CHECK that HPCI System is properly vented by observing a

steady flow of water in the HIGH POINT VENT TELL-TALE

SIGHT GLASS, 3-FG-073-0513. _ _(AC)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 81 of 104

Attachment 2

(Page 6 of 7)

HPCI Venting

Date

1.0 HPCI VENTING INSTRUCTIONS (continued)

[23] MEASURE vent line surface temperature near HPCI HIGH

POINT TELL-TALE VENT, 3-FSV-073-0063 and RECORD

below surface temperature at the time intervals indicated:

Time Temp (OF) Time Temp (OF)

1 min 6 min

2 min 7 min

3 min 8 min

4 min 9 min

5 min 10 min

[24] RELEASE HPCI HIGH POINT VENT TELL TALE,

3-HS-073-0063.

[25] REMOVE thermometer probe.

[26] CHECK that Step 1.0[23] of this attachment, peak vent line

surface temperature is less than 240°F.

[27] CLOSE HPCI HIGH POINT TELL-TALE VENT SOV,

3-SHV-073-0551.

[28] CLOSE 3-FCV-73-44, using HPCI PUMP INJECTION VALVE,

3-HS-73-44A.

1st

2nd

[29] CLOSE 3-FCV-73-35, using HPCI PUMP CST TEST VLV,

3-HS-73-35A.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 82 of 104

Attachment 2

(Page 7 of 7)

HPCI Venting

Date

1.0 HPCI VENTING INSTRUCTIONS (continued)

[30] INDEPENDENTLY VERIFY at Reactor Building Elevation 565':

VERIFY HPCI HIGH POINT TELL-TALE VENT SOV

3-SHV-073-0552 is CLOSED.

IV

VERIFY HPCI HIGH POINT TELL-TALE VENT SOV

3-SHV-073-0551 is CLOSED.

IV

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 83 of 104

Attachment 3

(Page 1 of 10)

HPCI Cold Quick Start

Date

1.0 HPCI COLD QUICK START INSTRUCTIONS

CAUTIONS

1) HPCI TURBINE SPEED, 3-SI-73-51, may indicate zero rpm while turbine shaft is still

rotating. The auxiliary oil pump should NOT be stopped until visual confirmation is

made locally by personnel that turbine speed is zero.

2) ALLOW approximately 15 minutes to elapse following HPCI turbine shutdown before

stopping HPCI PACKING EXHAUSTER to ensure removal of noncondensibles from

the HPCI turbine.

[1] RECORD time and date of turbine shutdown from

Step 7.0[18.9]:

Date Time

[2] PERFORM the following after allowing approximately

15 minutes to pass after turbine shutdown:

[2.1] VERIFY turbine speed is zero.

[2.2] PLACE HPCI AUXILIARY OIL PUMP, 3-HS-73-47A, to

STOP and RETURN TO AUTO position.

[2.3] STOP HPCI STEAM PACKING EXHAUSTER by placing

3-HS-73-10A to STOP and RETURN TO AUTO position.

[3] EXIT HPCI System LCO by updating Narrative log.

us

[4] IF SGTS is no longer required, THEN

STOP SGTS in accordance with 0-01-65, Standby Gas

Treatment System. (Otherwise N/A)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 84 of 104

Attachment 3

(Page 2 of 10)

HPCI Cold Quick Start

Date

1.0 HPCI COLD QUICK START INSTRUCTIONS (continued)

NOTES

1) The purpose of placing a caution tag on 3-FCV-73-35 and AOP is to alert the operator

that if AOP is run after turbine shutdown period has begun, turbine shutdown period

will have to begin again after AOP is stopped. This is to ensure that a non-oil primed,

cold, quick start time-to-rated flow test is performed.

2) If 3-FCV-73-35 is moved from position obtained during Step 7.0[17], this surveillance

may have to be reperformed. The position of 3-FCV-73-35 simulates reactor pressure

during the non-oil-primed, cold, quick start time-to-rated flow test.

3) The automatic and manual functions of 3-FCV-73-35 and AOP are NOT affected by

placement of caution tags.

[5] CLOSE 3-FCV-73-36, using HPCI/RCIC CST TEST VLV

3-HS-73-36A.

[6] PLACE caution tags on HPCI PUMP CST TEST VLV,

3-FCV-73-35 and HPCI AUXILIARY OIL PUMP control room

and local hand-switches to restrict manual operation of these

components.

RECORD Caution Order number below:

Caution Order No:

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 85 of 104

Attachment 3

(Page 3 of 10)

HPCI Cold Quick Start

Date

1.0 HPCI COLD QUICK START INSTRUCTIONS (continued)

NOTES

1) The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months wait period is based upon empirical data obtained by GE-San Jose and

provides sufficient time for the HPCI lube oil system to completely drain back to the

lube oil skid sump.

2) Empirical data obtained by GE - San Jose has demonstrated that a HPCI turbine

temperature which is within 25°F of ambient will show no observable variation in its

.start time from a completely cold turbine and may be considered cold.

3) Ambient, HPCI Room temperature may be obtained using either an analog or digital

temperature gage. HPCI TURB THRUST BRG temperature is recorded by

HPCI/RCIC/FW MISC TEMPERATURE 3-TR-73-54 on Panel 3-9-47 as Point

TE-73-54F.

4) Based upon temperature data from previous surveillance performances, time for HPCI

turbine to reach a cold condition is approximately 36-48 hours.

[7] VERIFY that at a minimum of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months have elapsed since

time and date recorded in Step 1.0[1].

[8] CHECK that HPCI TURB THRUST BRG temperature has

returned to within 25°F of the ambient, HPCI Room

temperature.

[9] RECORD below time, date and temperatures present when

performance of this surveillance was resumed.

Date Time

HPCI Turb Thrust HPCI Room Ambient

Brg Temperature Temperature

OF -----

OF

(3-TE-73-54F) (Portable M&TE)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 86 of 104

Attachment 3

(Page 4 of 10)

HPCI Cold Quick Start

Date

1.0 HPCI COLD QUICK START INSTRUCTIONS (continued)

[10] REMOVE caution tags placed on HPCI PUMP CST TESTVLV

3-FCV-73-35 and HPCI AUXILIARY OIL PUMP hand-switches

in Step 1.0[6].

[11] OPEN 3-FCV-73-36 using HPCI/RCIC CST TEST VLV

3-HS-73-36A.

NOTES

1) Placing a jumper across Contacts 1-2 of 3-RLY-073-23A-K47 allows for immediate

start of HPCI AUXILIARY OIL PUMP when 3-HS-73-47A is placed in START. This is

necessary to simulate an immediate start of the HPCI AUXILIARY OIL PUMP that

occurs during an actual HPCI initiation on high drywell pressure or low-low RPV water

level.

2) 3-RLY-073-23A-K47 is located on Panel 3-9-39. Opening back of Panel and facing

backs of relays, this relay is located on third row of relays from bottom and is third

relay from right.

[12] PLACE jumper across 3-RLY-073-23A-K47 Contacts 1-2 in

Panel 3-9-39. REFER TO Attachment 7.

1st

2nd

[13] START or VERIFY started SGTS in accordance with 0-01-65,

Standby Gas Treatment System.

[14] START HPCI STEAM PACKING EXHAUSTER by placing

3-HS-73-10A to START.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 87 of 104

Attachment 3

(Page 5 of 10)

HPCI Cold Quick Start

Date

1.0 HPCI COLD QUICK START INSTRUCTIONS (continued)

NOTES

1) During the startup of the HPCI Turbine a second operator should be utilized to monitor

for abnormal conditions and alarms.

2) The HPCI Turbine parameters should be monitored during HPCI startup. This will

ensure proper response of the control systems. If HPCI pumps suction pressure

causes an auto swap of the HPCI suction valves from CST to the torus, then the HPCI

Turbine should be tripped.

WARNING

[NER] Failure of both HPCI steam exhaust piping rupture discs during turbine startup and

operation will result in a process steam release into the HPCI Room. This release raises

the risk of personnel injury until steam line isolation occurs. Therefore, personnel in the

HPCI Room should minimize stay time in close proximity to the rupture disc cage

assembly. [IE 93-67]

[15] PERFORM the following prior to HPCI turbine startup:

[15.1] [NER] VERIFY communication is established with

Operations personnel in HPCI Room. [IE 93-67]

[15.2] [NER] REQUEST Operations personnel in HPCI Room

ensure that all unnecessary personnel have exited HPCI

Room. [IE 93-67]

[15.3] [NER] ANNOUNCE HPCI turbine startup over plant public

address system. [IE 93-67]

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 88 of 104

Attachment 3

(Page 6 of 10)

HPCI Cold Quick Start

Date

1.0 HPCI COLD QUICK START INSTRUCTIONS (continued)

NOTE

Step 1.0[16] may be signed off after the completion of Step 1.0[17].

CAUTIONS

1) If HPCI TURBINE STEAM SUPPLY VLV 3-FCV-73-16 fails to fully open, then the

governor control system ramp generator will time out and HPCI turbine speed,

discharge pressure, or flow will be lower than expected.

2) DO NOT REATTEMPT to open HPCI TURBINE STEAM SUPPLY VLV 3-FCV-73-16

unless HPCI TURBINE STOP VALVE 3-FCV-73-18 is closed using HPCI TURBINE

TRIP push-button 3-HS-73-18A. Failure to observe this caution will result in a turbine

overspeed trip if 3-FCV-73-16 is opened with the ramp generator timed out.

11**Start()fCritical.*$tep($)

[16] [NER/C] SIMULTANEOUSLY PERFORM the following sub-steps

in order to accomplish a cold, non-oil-primed, quick start of the

HPCI turbine: [INPO SOER 81-013] [GE SIL 336 R1]:

[16.1] PLACE HPCI AUXILIARY OIL PUMP, 3-HS-73-47A to

START.

[16.2] OPEN 3-FCV-73-16, using HPCI TURBINE STEAM

SUPPLY VLV, 3-HS-73-16A.

[16.3] START stopwatch.

11**iEnct** OfYrifiCalSfep(S)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 89 of 104

Attachment 3

(Page 7 of 10)

HPCI Cold Quick Start

Date

1.0 HPCI COLD QUICK START INSTRUCTIONS (continued)

[17] WHEN HPCI SYSTEM FLOW/CONTROL 3-FIC-73-33

indicates 2 500 (2 5,000 gpm) discharge flow and HPCI PUMP

DISCH PRESS 3-PI-73-31A indicates a pump discharge

pressure

psig, THEN

(Step 7.0[7~3])

STOP the stopwatch.

NOTE

Steps 1.0[18] thru 1.0[20] should be performed in parallel with remaining surveillance steps

to allow for turbine shutdown in order to limit heat addition to the suppression pool.

[18] RECORD below time taken to reach rated flow and pressure

measured in Step 1.0[17]:

seconds

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 90 of 104

Attachment 3

(Page 8 of 10)

HPCI Cold Quick Start

Date

1.0 HPCI COLD QUICK START INSTRUCTIONS (continued)

[19] IF ICS transient startup data is available, THEN

PERFORM the following. (Otherwise N/A the following

substeps)

[19.1] REVIEW ICS group tabular trend display data obtained

for HPCI discharge pressure, flow, and manual initiation

status.

[19.2] RECORD below time span from HPCI manual initiation

to when HPCI flow was 5,000 gpm with a discharge

pressure:

psig

(Step 7.0[7.3])

seconds

[19.3] CHECK that time recorded is less than or equal to

30 seconds. _ _(AC)

[20] IF ICS transient startup data is NOT available, THEN

CHECK that time recorded in Step 1.0[18] is less than or equal

to 30 seconds. (Otherwise N/A) (AC)

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 91 of 104

Attachment 3

(Page 9 of 10)

HPCI Cold Quick Start

Date

1.0 HPCI COLD QUICK START INSTRUCTIONS (continued)

NOTES

1) [NRC/C) Consult Step 3.0Q for additional background information regarding HPCI

System removal from operable service. [NCO 89-0216-002]

2) The intent of Steps 1.0[21] through 1.0[28] is to depress and hold the trip push-button

for thirty seconds, verify the alarms, close 3-FCV-73-16, observe the aux. oil pump

starts, then release-the push-button.

3) During the HPCI Turbine trip a second operator should be utilized to monitor for

abnormal conditions and alarms.

4) HPCI PUMP DISCH FLOW LOW 3-FA-73-33 (3-XA-55-3F, window 5) needs to be

verified prior to 3-FCV-73-16 becoming full close.

[21] VERIFY HPCI System has been declared inoperable and

ENTER appropriate LCO information into Narrative log as

required.

US

11.$tartC:>f** pritical$t~p(§)

[22] DEPRESS and HOLD HPCI TURBINE TRIP 3-HS-73-18A

until Step 1.0[28].

[23] WAIT 30 seconds and OBSERVE the following annunciators

are in ALARM:

HPCI TURBINE TRIPPED 3-ZA-73-18 (3-XA-55-3F,

window 11)

HPCI PUMP DISCH FLOW LOW 3-FA-73-33

(3-XA-55-3F, window 5)

[24] CLOSE 3-FCV-73-16, using HPCI TURBINE STEAM SUPPLY

VLV,3-HS-73-16A.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 92 of 104

Attachment 3

(Page to of to)

HPCI Cold Quick Start

Date

1.0 HPCI COLD QUICK START INSTRUCTIONS (continued)

[25] OBSERVE HPCI AUXILIARY OIL PUMP starts as turbine

slows.

[26] OBSERVE HPCI TURBINE SPEED 3-SI-73-51 reading lowers

to approximately zero.

[27] VERIFY HPCI TURBINE STEAM SUPPLY VLV 3-FCV-73-16

is closed.

[28] RELEASE HPCI TURBINE TRIP 3-HS-73-18A.

[29] RESET the following annunciators:

HPCI TURBINE TRIPPED 3-ZA-73-18 (3-XA-55-3F,

window 11)

HPCI PUMP DISCH FLOW LOW 3-FA-73-33

(3-XA-55-3F, window 5)

[30] VERIFY HPCI STM LINE CNDS INBD/OUTBD DR VLVS

3-FCV-73-6A and 3-FCV-73-6B are OPEN by observing

3-ZI-73-6A and 3-ZI-73-6B position indicating lights.

[31] REMOVE jumper across 3-RLY-073-23A-K47 Contacts 1-2 in

Panel 3-9-39. (Otherwise N/A)

1st

2nd

[32] RETURN TO Step 7.0[22.2] in the procedure.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 93 of 104

Attachment 4

(Page 1 of 3)

3-FCV-73-18 Time Delay Adjustment

Date

1.0 3-FCV-73-18 TIME DELAY ADJUSTMENT INSTRUCTIONS

NOTES

1) The following steps record the initial and final position of 3-SHV-73-712 to track the

adjustments performed.

2) Turning the valve closed causes a slower reset time and opening the valve causes a

faster reset time.

3) Step 1.0[3] may be performed multiple times to achieve a 4-6 second reset time for

HPCI TURBINE STOP VALVE 3-FCV-73-18.

4) Two people are needed to perform the 3-FCV-73-18 time delay test.

[1] DETERMINE the as-found position of 3-SHV-73-712 as

follows:

CLOSE 3-SHV-73-712 and RECORD the number of turns

valve was opened.

As-Found Turns Open

[2] RETURN 3-SHV-73-712 to its original position as follows:

OPEN 3-SHV-73-712 the number of turns open recorded in

Step 1.0[1] of this attachment.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 94 of 104

Attachment 4

(Page 2 of 3)

3-FCV-73-18 Time Delay Adjustment

Date

1.0 3-FCV-73-18 TIME DELAY ADJUSTMENT INSTRUCTIONS

(continued)

NOTE

The following steps should be reviewed prior to performance of Step 1.0[3.3] to ensure

proper operation of system.

[3] PERFORM the following until a 4-6 second reset time is

achieved for HPCI TURBINE STOP VALVE 3-FCV-73-18.

[3.1] ADJUST 3-SHV-73-712 to try and achieve a 4-6 second

reset time. (REFER TO Note above.)

[3.2] Using the 3-FCV-73-18 valve stem position, CHECK that

the HPCI TURBINE STOP VALVE 3-FCV-73-18 is

OPEN.

[3.3] LIFT and IMMEDIATELY RELEASE HPCI TURBINE

MECH TRIP VLV 3-XCV-73-18 trip knob.

[3.4] OBSERVE the 3-FCV-73-18 valve stem position to

CHECK that the HPCI TURBINE STOP VALVE

3-FCV-73-18 closes.

[3.5] START the stop watch when the HPCI TURBINE STOP

VALVE 3-FCV-73-18 is full closed.

[3.6] STOP the stop watch when the HPCI TURBINE STOP

VALVE 3-FCV-73-18 begins to open.

[4] VERIFY 3-FCV-73-18 time delay is within 4-6 seconds.

RE-PERFORM Step 1.0[3] of this attachment.

[5] RECORD the final 3-FCV-73-18 time delay.

Time seconds

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 95 of 104

Attachment 4

(Page 3 of 3)

3-FCV-73-18 Time Delay Adjustment

Date

1.0 3-FCV-73-18 TIME DELAY ADJUSTMENT INSTRUCTIONS

(continued)

[6] RECORD the Final number of turns open for 3-SHV-73-712,

by adding or subtracting the adjustments made in Step 1.0[3]

to the initial position recorded in Step 1.0[1] of this attachment.

Number of turns open

[7] RETURN TO Step 7.0[22] in the procedure.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 96 of 104

Attachment 5

(Page 1 of 1)

ASME OM Code Inservice Testing Review Form

N/Aor

Valves Tested Acceptable Not Acceptable Not Tested

3-FCV-73-18 o o D

(Step 7.0[6.9.3] or Step 7.0[6.10.2])

3-ISV-73-23 (Step 7.0[14.2]) D D D

3-CKV-73-559 (Step 7.0[16.7]) D D D

3-CKV-73-603 (Step 7.0[14.2]) D D D

N/A or

HPCI Pump Acceptable Not Acceptable Not Tested

Differential Pressure D D D,

(Step 7.0[15.7.2])

Date Received:

ASME OM Code Reviewer Date

IF any evaluation results are found to be NOT Acceptable, THEN Date

CONTACT OPS immediately. (Otherwise, N/A)

ASME OM Code data enter in 3-SI-3.1.5 and 3-SI-3.2.1

ANII Reviewer Date

REMARKS:

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 97 of 104

Attachment 6

(Page 1 of 1)

HPCI Lube Oil Skid and Booster Pump Oil Level Settings

NIPPLE

t----,.....-----r----......-----(TYP 2 PLCS)

LUBE OIL

SKID -----SiGHT GLASS

I----------i - ** - . . ** - OIL LEVEL

....----BEARING HOUSING NIPPLE

rSIGHT GLASS

I I

141/2"

TO

15 1/2"

- -. -- -j- -OIL LEVEL

1 1/2"

TO

2"

HPCI Lube Oil Tank Pedestal

HPCI LUBE OIL TANK HPCI BOOSTER PUMP

SIGHT GLASS SIDE VIEW SIGHT GLASS SIDE VIEW

(TYP 2 PLCSl

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 98 of 104

Attachment 7

(Page 1 of 1)

HFA Relay Contact Layout

FRONT

1

e

8 6 2

GENERAL ELECTRIC

RELAY

TYPE HFA

REAR

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 99 of 104

Attachment 8

(Page 1 of 4)

Installation and Removal of Yokogawa Recorders for 3-FCV-73-18

Date

1.0 INSTALLATION OF THE YOKAGAWA RECORDER.

[1] IF the Yokogawa will be connected in Panel 3-9-3, THEN

CONNECT the YOKOGAWA as follows: (Otherwise N/A)

For relay 3-RLY-23A-K31

CONNECT 1 Channel across Terminals 88-90 and

88-87.

For 3-ZS-73-18A

CONNECT 1 Channel across Terminals 88-90 and

88-91.

For 3-HS-73-18A

CONNECT 1 Channel across Terminals AA-69 and

AA-70.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 100 of 104

Attachment 8

(Page 2 of 4)

Installation and Removal of Yokogawa Recorders for 3-FCV-73-18

Date

1.0 INSTALLATION OF THE YOKAGAWA RECORDER. (continued)

[2] IF the Yokogawa will be connected in Panel 3-9-39, THEN

CONNECT the YOKOGAWA as follows: (Otherwise N/A)

For relay 3-RLY-23A-K31

CONNECT 1 Channel across Terminals 88-85 and

88-86.

For 3-ZS-73-188

CONNECT 1 Channel across Terminals CC-25 and

CC-26.

For 3-HS-73-18A

CONNECT 1 Channel across Terminals 88-11 and

88-12.

For 3-PCV-73-188

CONNECT 1 Channel across Terminals CC-31 and

CC-32.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 101 of 104

Attachment 8

(Page 3 of 4)

Installation and Removal of Yokogawa Recorders for 3-FCV-73-18

Date

2.0 REMOVING OF THE YOKAGAWA RECORDER.

[1] IF the Yokogawa was installed in Panel 3-9-3, THEN

REMOVE the Yokogawa Channels from the following

terminals: (Otherwise N/A )

Channel across Terminals 88-90 and 88-87

1st

2nd

Channel across Terminals 88-90 and 88-91

1st

2nd

Channel across Terminals AA-69 and AA-70

1st

2nd

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 102 of 104

Attachment 8

(Page 4 of 4)

Installation and Removal of Yokogawa Recorders for 3-FCV-73-18

Date

2.0 REMOVING OF THE YOKAGAWA RECORDER. (continued)

[2] IF the Yokogawa was installed in Panel 3-9-39, THEN

REMOVE the Yokogawa Channels from the following

terminals: (Otherwise N/A )

Channel across Terminals 88-85 and 88-86

1st

2nd

Channel across Terminals CC-25 and CC-26

1st

2nd

Channel across Terminals 88-11 and 88-12

1st

2nd

Channel across Terminals CC-31 and CC-32

1st

2nd

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 103 of 104

Attachment 9

(Page 1 of 1)

Annunciators Affected by Surveillance Procedure Performance

Panel Location Description Location

3-9-3 HPCI PUMP DISCH FLOW LOW 3-FA-73-33 3-XA-55-3F

Window 5

3-9-3 HPCI TURBINE TRIPPED 3-ZA-73-18 3-XA-55-3F

Window 11

3-9-3 HPCI TURBINE INLET DRAIN POT LEVEL HIGH 3-XA-55-3F

3-LA-73-5 Window 26

This Attachment provides the UO with a listing of Main Control Room alarms that will be

affected by performance of this SR. This Attachment is for information only.

BFN HPCI Main and Booster Pump Set 3-SR-3.5.1.7

Unit 3 Developed Head and Flow Rate Test at Rev. 0044

Rated Reactor Pressure Page 104 of 104

Attachment 10

(Page 1 of 1)

ASME OM Code Restroke Time Record Form

VALVE UNID NORMAL MEASURED MEASURED MAXIMUM

STROKE INITIAL STROKE RE-STROKE ALLOWED STROKE

TIME (SEC) TIME (SEC) TIME (SEC) TIME (SEC)

3-FCV-073-0018

0.8 - 2.2 3.0

(OPEN)

(

Browns Ferry Nuclear Plant

Unit 3

General Operating Instruction

3-GOI-100-12

Power Maneuvering

Revision 0031

Quality Related

Level of Use: Reference Use

Effective Date: 12-01-2007

Responsible Organization: OPS, Operations

Prepared By: William Fuller

Approved By: John Kulisek

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 2 of 29

Current Revision Description

Type of Change: Corrective Action Tracking Number: 034

PERs 961778,126211,116666,132198

PCRs 07004255, 07004297

The primary reason for this revision is to help minimize unplanned radiological exposures to

plant personnel during normal pJant operations. Because the performance of this procedure

does carry risks for such events, the following changes are made.

The procedure is revised to identify points in the procedure requiring Radiation

Protection notification to ensure any needed radiological controls are implemented to

prevent unintended radiological exposure during a reactor startup. The previous

revision contained steps requiring logging of Radiation Protection technician's name

and also contained a signature line for the Radiation Protection supervisor, if needed.

These logging and signature lines are removed from the procedure and their function

replaced by the new Appendix A, added to the procedure.

The function of Appendix A is to ensure proper communication between Operations

and Radiation Protection, and that Radiation Protection is allowed sufficient

opportunity to implement any needed radiological controls. A set of instructions is

included with Appendix A to insure proper data entry and control of any applicable

radiological protection hold points. The appendix is designed to encompass Radiation

Protection notifications from this GOI and also those initiated in any support procedure

implemented by this GOL

P&L Step 3.7, Radiation Protection Notifications and Radiological Protection Hold

Points (RPHPs), is added to provide information regarding how Radiation Protection

Notifications and Radiological Protection Hold Points are to be controlled. The P&L

also addresses the function of Appendix A. ,

The above changes are all primarily administrative in nature. Other changes to the

procedure are as follows:

Illustration 1 is changed to remove reference to specific reactor thermal limit values.

The thermal limits values frequently change. The change for this revision is to

reference the 0-TI-248 section by title for the limits.

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 3 of 29

Table of Contents

1.0 PURPOSE 4

2*.0 REFERENCES 4

2.1 Technical Specifications 4

2.2 Technical Requirements Manual-TRM 5

2.3 Final Safety Analysis Report 5

2.4 Plant Instructions 5

2.5 Miscellaneous Documents 6

3.0 PRECAUTIONS AND LIMITATIONS 8

3.1 General 8

3.2 Reactivity 8

3.3 Technical Specifications 9

3.4 Condensate System Limits at Normal Steady-StateOperations 9

3.5 Reactor Feedwater Pumps limits at Normal Steady-State Operations 9

3.6 Downpowering Of Nuclear Units Under Low System Load Conditions 10

3.7 Radiation Protection Notifications and Radiological Protection Hold

Points (RPHPs) [SOER 01-1, BFN PER 126211, PER 961778, PER

116666] 11

4.0 PREREQUiSiTES 14

5.0 INSTRUCTION STEPS 15

Illustration 1: Reactor Thermal Limits 26

Illustration 2: Reactor Thermal Power Versus Ultimate Heat Sink

Temperature Limit 27

Appendix A: Radiation Protection Notifications 28

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 4 of 29

1.0 PURPOSE

This instruction provides precautions and limitations, prerequisites and procedural

steps for power maneuvering between approximately 50% and 1Ooo~ power.

The following are examples of conditions that may require use of this procedure:

Load Following, as requested by TVA Operations Duty Specialist (ODS)

Control rod pattern adjustment

Control rod testing

Removing and/or returning a Recirc pump to service

Maintenance of plant equipment, such as Reactor Feed Pumps, Condensate or

Condensate Booster Pumps, Circulating Water Pump, Condenser Waterbox,

etc., that are required to support full power operations.

2.0 REFERENCES

2.1 Technical Specifications

Section 3.1, Reactivity Control Systems.

Section 3.1.3, Control Rod Operability.

Section 3.1.6, Rod Pattern Control.

Section 3.2.1 , Average Planar Linear Heat Generation Rate (APLHGR).

Section 3.2.2, Minimum Critical Power Ratio (MCPR).

Section 3.2.3, Linear Heat Generation Rate (LHGR).

Section 3.3.1.1, Reactor Protection System (RPS) Instrumentation.

Section 3.3.2.1, Control Rod Block Instrumentation.

Section 3.3.8.2, Reactor Protection System (RPS) Electric Power Monitoring.

Section 3.4.1, Recirculation Loops Operating.

Section 3.4.2, Jet Pumps.

Section 3.4.6, RCS Specific Activity.

Section 3.7.5, Main Turbine Bypass System.

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 5 of 29

2.1 Technical Specifications (continued)

Section 5.2.2, Unit Staff.

Section 5.4, Procedures.

Section 5.5, Programs and Manuals.

2.2 Technical Requirements Manual-TRM

TRM Section 3.1, Reactivity Control.

TRM Section, 3.3.1, Reactor Protection System (RPS) Instrumentation.

TRM Section 3.3.4, Control Rod Block instrumentation.

TRM Section 3.3.5, Surveillance Instrumentation.

TRM Section 3.4.1, Coolant Chemistry.

2.3 Final Safety Analysis Report

Chapter 3.0, Reactor.

Chapter 4.0, Reactor Coolant System.

Chapter 7.0, Control And Instrumentation.

Chapter 10.0, Auxiliary Systems.

Chapter 13.0, Condu,ct of Operations.

2.4 Plant Instructions

3-AOI-1 00-1, Reactor Scram.

3-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and

Reductions in Power During Power Operations.

3-01-2, Condensate System.

3-01-2A, Condensate Demineralizers System.

3-01-3, Reactor Feedwater System.

3-01-68, Reactor Recirculation System.

3-01-85, Control Rod Drive System.

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 6 of 29

2.4 Plant Instructions (continued)

3-01-92B, Average Power Range Monitoring System.

3-01-92C, Rod Block Monitoring System.

3-SR-3.1.3.5(A), Control Rod Coupling Integrity Check.

3-SR-3.3.1.1, Core Thermal Hydraulic Stability.

3-SR-3.4.1 (SLO), Reactor Recirculation System Single Loop Operation.

3-SR-3.4.1 (DLO), ReactorRecirculatlon System Dual Loop Operation.

3-SR-3.3.2.1.4(A), Rod Block Monitor (RBM) Calibration and Functional Test.

3-SR-3.3.2.1.4(B), Rod Block Monitor (RBM) Calibration and Functional Test.

OPDP-1, Conduct of Operations.

SPP-2.2, Administration of Site Technical Procedures.

SPP-10.3, Verification Program.

SPP-10.4, Reactivity Management Program.

0-TI-248, Station Reactor Engineer.

2.5 Miscellaneous Documents

BWROG-94078, BWR Owner's Group Guidelines for Stability Interim Corrective

Action.

GE SIL 380, BWR Core Thermal Hydraulic Stability.

INPO SER 89-006, Withdrawal of Safety Rod Group Out of Sequence.

INPO SER 91-024, Inadequate Control of Reactivity Changes During a Plant

Shutdown Results in an Unplanned Plant Transient.

INPO SER 92-008, Reactivity Management Expectations During Plant Shutdowns.

INPO SER 92-19, Power Oscillations at Boiling Water Reactors.

NRC Bulletin 88-07, Supplement 1, Power Oscillations in Boiling Water Reactors.

NRC Generic Letter 94-02, Long-Term Solutions and Upgrade of Interim Operating

Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors.

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 7 of 29

2.5 Miscellaneous Documents (continued)

NRC Information Notice 92-74, Power Oscillation at Washington Nuclear Power

Unit 2.

NRC Notice of Violation 94-24.

NSRB Item A258-4, Review procedures to preclude an event similar to SER 24-91,

inadequate control of reactivity changes during plant shutdown results in unwanted

transient.

Scram Frequency Reduction Committee Item SFRC-17, G-20-1 and 2.

T.A. Keys Memorandum to K.L. Welch, Use of Increased Core Flow (ICF) at Browns

Ferry Nuclear Plant (L32 920709 801).

Letter from O. D. Kingsley to W. J. Museler, DOWNPOWERING OF NUCLEAR

UNITS UNDER LOW SYSTEM LOAD CONDITIONS, March 1, 1996 (AOO 960226

150).

TVA-BFN-TS-384, Technical Specification (TS) Change TS-384 Request for License

Amendment for Power Uprate Operation (R08-980316-888).

NEDC-32751 P, Power Uprate Safety Analysis for the Browns Ferry Nuclear Plant

Units 2 and 3 (R08-980316-888)

GE-NE-B13-01866-39, Task Report 39 Summary of System Evaluations and

Proposed Changes to Design Criteria Documents (W79-980427-005).

LETTER TVAPUR- PROC-98003, Turbine Stop Valve and Turbine Control Valve

Surveillance Test Procedures (W79 980622-001).

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 8 of 29

3.0 PRECAUTIONS AND LIMITATIONS

3.1 General

A. While performing this procedure, plant conditions/status changes such that the

Unit Supervisor determines they are outside the scope of this procedure, he/she

may transition to 3-GOI-1 00-12A or 3-GOI-1 00-1A, as appropriate.

3.2 Reactivity

A. [INPO/C] Activities that can directly affect core reactivity are of a critical nature and

require strict procedural compliance, along with conservative actions. [INPO SER

89-006]

B. [NSRB/C] Reactivity can be added without moving control rods due to changing

plant conditions (such as lowering moderator temperature, lowering Xenon

concentration, rising reactor pressure, and rising feedwater flow) especially at

low power. Awareness of these conditions and monitoring core instrumentation

for these changes is required. [A258-4]

C. Reactor Engineering should be contacted to monitor flux shaping prior to all

power reductions.

D. [QAlC] SPP-10.4 requires approval of the Plant Manager or his designee prior to

any planned operation with the following reactivity control equipment bypassed

unless bypassing of this equipment is specifically allowed within approved

procedures:

1. Rod Worth Minimizer

2. Rod Block Monitor

3. Average Power Range Monitors

4. Integrated Computer System [ISE-NPS-92-R01]

5. OPRM Trip Function

E. Power Maneuvering Recommendations will be made by Reactor Engineering

(REFER TO 0-TI-248 for more detailed information.)

F. Refer to 0-TI-248, Station Reactor Engineer, for Feedwater Temperature Graph

and Power To Flow Map.

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 9 of 29

3.3 Technical Specifications

A. When the Reactor Recirculation System is operating in single loop operation,

3-SR-3.4.1 (SLO) is required to be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of entering single

loop operations to ensure the requirements of Tech Specs 3.4.1 are met.

B. [NRC/C] Core Thermal-Hydraulic Stability, is required by 3-SR-3.3.1.1.1 to be

verified outside Regions I, II & III. when OPRM's are INOP. [NCO 940245010]

C. Whenever Forebay Temperature is >92.5°F, as indicated on 2-TS-27-144,

Unit 3 power is required to be derated to within the limits shown in Illustration 2,

per Tech Specs 3.7.1.2.

3.4 Condensate System Limits at Normal Steady-StateOperations

A. [IIIC] Condensate flow should always be maintained within the following limits,

using 3-FC-2-29 in BAL if possible, to prevent Condensate Pump damage:

1. One Condensate Pump operation, greater than 1.5 X 106 Ibm/hr but less

than 6.25 x 106 Ibm/hr.

2. Two Condensate Pump operation, greater than 3.0 X 106 Ibm/hr but less

than 12.5 x 106 Ibm/hr.

3. Three Condensate Pump operation, greater than 4.5 X 106 Ibm/hr but less

than 15.0 x 106 Ibm/hr. [11-8-91-158]

B. Normal maximum line current to Condensate Pump Motors should not exceed

118 amps steady-state operations.

C. Normal maximum line current to' Condensate Booster Pump Motors should not

exceed 225 amps steady-state operations.

D. Changes in condensate system flow may require adjustment to SPE CNDS

BYPASS, 3-FCV-002-0190, either in the Control Room or locally. Personnel

adjusting this valve locally are required to be in direct communication with the

Control Room. Evolutions resulting in changes in condensate/feedwater flow

(condensate/booster pump start, feedwater pump start, changes in reactor

power, feedwater flow, steam flow, etc.) will affect flowrates through

3-FCV-002-0190, steam-jet air-ejector condenser(s), steam packing exhauster

condenser, and off-gas condenser. 3-FI-2-42, on Panel 3-9-6 should be

maintained between 2 X 106 Ibm/hr and 3 X 106 Ibm/hr.

3.5 Reactor Feedwater Pumps limits at Normal Steady-State

Operations

A. Individual Reactor Feedpump speed should be less than 5050 RPM.

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 10 of 29

3.6 Downpowering Of Nuclear Units Under Low System Load

Conditions

A. Due to having five nuclear units in an operating status, the frequency of

downpowering units under low system load conditions is expected to rise. The

following communications process will be used to coordinate downpowering a

unit at BFN under low load conditions:

1. The Electrical System Operator (ESO) will anticipate the potential need to

downpower nuclear units as far in advance as reasonable, normally one to

two days. The ESO will inform the Operations Duty Specialist (ODS) of

this potential need.

2. The ODS will notify the Browns Ferry Shift Manager that a potential need

to downpower exists.

3. The Shift Manager will notify the Operations Superintendent who will notify

the Operations Manager and Duty Plant Manager.

4. BFN will initiate a telecon with other operating nuclear units and senior

nuclear corporate management (normally, Senior Vice President, Nuclear

Operations, or, President, TVA Nuclear and Chief Nuclear Officer) to

formulate a contingency plan. The plan will address which units are to be

downpowered based on existing plant conditions, the reduction capability

of each unit, time to reach reduced power as well as return to full power,

and the preferred order for downpowering.

5. The contingency plan will be communicated to the appropriate site

management and Shift Manager for the impacted units as well as the

transmission/power supply organization.

6. The ESO will notify the designated Shift Managers approximately two to

four hours before the need to actually downpower. The Shift Manager will

notify the Operations Superintendent of any actual downpower.

7. Any change to unit status that would impact the agreed upon contingency

plan will cause the telecon to be reconvened with all affected parties and a

revised contingency plan developed. This will be initiated by the site

management who identifies the need to revise the plan.

BFN Power Maneuvering 3-GOI-100-12

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Page 11 of 29

3.7 Radiation Protection Notifications and Radiological Protection

Hold Points (RPHPs) [SOER 01-1, BFN PER 126211, PER 961778, PER 116666]

A. This General Operating Instruction initiates processes that cause a change in

area radiation levels in the plant. Generally, the procedure is used to reduce

power to some predetermined level, and then, after the purpose of the power

reduction is satisfied, the unit is returned to full power operation. The impact on

radiation levels in the plant is somewhat dependent of the purpose of the power

reduction, but generally plant radiation levels follow reactor power down and

then rise as the unit is returned to full power operation. The performance of this

procedure, in addition to the various other procedures used, carries the risk of

unintended radiological exposures and also radiation levels that warrant

changes in High Radiation Area or Locked High Radiation Area Controls.

Depending upon the extent of the power reduction, this GOI relies on System

Operating Instructions (support procedures) for system alignments required for

the various process systems. Many of these alignments can and do result in

changing the radiological impacts for the areas affected by the alignments.

Therefore, an increase in area monitoring may be required to determine

expected dose rates for areas that might require plant personnel to be present.

As the Unit is returned to full power operation, the risk of unintended radiation

exposures is increased if plant personnel remain in affected areas.

B. To reduce the probability of unintended radiation exposures, the following

controls are established by this procedure:

1. Radiological Protection Hold Points (RPHPs) are pre-established at

appropriate locations in this GOI and in the support procedures. The

function of RPHPs is to allow Radiation Protection to help ensure no

unintended radiological exposures occur as the result of a system

configuration change or raising reactor power. This may require holding at

the point identified in the procedure until verifying personnel are not in an

area before continuing in the procedure. These RPHPs also allow a

determination as to whether actions are required to relax or implement

RCI-17, Control of High Radiation Areas and Very High Radiation Areas,

controls.

2. The Radiation Protection notification steps have an (R) placed in the step

initial line, which means these steps can NOT be omitted 'unless the action

associated with the step is not performed, or the Radiation Protection

notification requirements are currently satisfied for the action, or the step

allows the notification to be N/A'd as determined by the Unit Supervisor.

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 12 of 29

3.7 Radiation Protection Notifications and Radiological Protection

Hold Points (RPHPs) [SOER 01-1, BFN PER 126211, PER 961778, PER 116666]

(continued)

3. An Appendix (Appendix A, Radiation Protection Notifications) is provided to

record Radiation Protection notifications, RPHPs, and release of RPHPs,

as necessary. The instructions for Appendix A are used to identify the

appropriate required logging of Radiological Protection entries. The primary

function of the appendix is to ensure proper communication with Radiation

Protection personnel and that they are allowed sufficient opportunity to

implement needed radiological controls.

4. Radiation Protection notification steps that require a RPHP are clearly

worded that an RPHP is in effect. For these steps, it should be made clear

to Radiation Protection that an RPHP is in effect so that they understand

that a signature on Appendix A will be necessary.

Radiation Protection notification steps that are not identified as RPHP

steps are considered courtesy notification steps to Radiation Protection.

These steps serve the purpose of informing Radiation Protection of

evolutions that are about to be implemented that may impact plant

radiological conditions and allow them to respond or "get their ducks in a

row". None of these steps imply that a hold in the procedure is necessary

unless Radiation Protection identifies one may be necessary at some point

after the notification is made. In many cases, the courtesy notifications are

related to an RPHP notification that will be reached later in the procedure.

These courtesy steps may also inform Radiation Protection that a system

has been returned to normal, has been shutdown, or a pump that was

previously started, is now shutdown. This information may be useful to

Radiation Protection for determining if area surveys should be performed

due to changing radiological conditions in an area. The courtesy

notification steps generally require an entry of the notification in the NOMS

narrative log, but mayor may not require Appendix A entry by operations,

depending upon expected radiological impact of the associated

evolution(s).

C. Because this procedure may be implemented to recover from system operation

problems and/or allow maintenance on plant equipment that may not be

operating correctly, there are a multitude of scenarios that can occur while the

procedure is in effect. If, at any time while performing this procedure, or while

performing a support procedure, Radiation Protection personnel, or Unit

Operator, Unit Supervisor, or other knowledgeable shift member identifies the

need for a RPHP, then the following is performed:

BFN Power Maneuvering 3-GOI-100-12

Unit 3 " Rev. 0031

Page 13 of 29

3.7 Radiation Protection Notifications and Radiological Protection

Hold Points (RPHPs) [SOER 01-1, BFN PER 126211, PER 961778, PER 116666]

(continued)

1. "RPHP" is written to the left of the affected procedure step number (this

Gal or the support procedure). If the RPHP is identified for a support

procedure, then RPHP is placed to the left of the step in this GOI that

initiates the support procedure.

2. Appropriate notifications made to Radiation Protection personnel, as

necessary.

3. The instructions for Appendix A are to be used to identify the appropriate

required logging of Radiological Protection entries.

D. Removal of any Radiation Protection Notification from this procedure requires

Operations Management and Radiation Protection Management approval

unless the action(s) related to the notification is also removed.

Removal or addition of any procedure actions that require Radiation Protection

notification requires that Radiation Protection be notified.

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 14 of 29

4.0 PREREQUISITES

[1] Reactor in MODE 1 with power greater than SODA>.

Initials Date Time

Performed by:

Name (Print) Initials

Reviewed by:

Shift Manager Signature Date

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 15 of 29

5.0 INSTRUCTION STEPS

NOTES

1) [NRC/C) Sequential completion is preferred in Section 4.0[1] unless unit conditions

dictate otherwise and the Unit Supervisor approves. [IR 84-45]

2) [NRC/C) Those steps preceded by an (R) are required for all power maneuvers and can

not be omitted unless provided for in the procedure. [IR 84-45]

3) [NRC/C) Those steps not preceded by an (R) may be signed off as NA for all power

maneuvers and initialed by the Unit Supervisor as appropriate. [IR 84-45]

4) Initials are NOT required after the step is reached where power reduction is terminated

up to the step where power ascension is commenced. These steps may be marked

N/A.

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

(R)

Initials Date Time

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

(R)

Initials Date Time

[3] NOTIFY Operations Duty Specialist (ODS) and/or Chattanooga Load

Coordinator of impending power reduction.

Initials Date Time

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 16 of 29

5.0 INSTRUCTION STEPS (continued)

NOTE

During the power reduction, Radiation Protection should be kept informed of systems or

equipment removed from service, significant power changes, and other actions or

conditions that may impact radiological control areas.

[4] NOTIFY Radiation Protection of purpose for power reduction, the target

power level (see above note), and RECORD time Radiation Protection

notified in NOMS Narrative Log.

(R)

Initials Date Time

[4.1] VERIFY appropriate data recorded on Appendix A in accordance with

Appendix A instructions.

(R)

Initials Date Time

[5] IF this instruction was entered due to a Recirc Pump startup or shutdown,

THEN

PERFORM the following:

N/A Step 5.0[6] through Step 5.0[11].

ENTER 3-GOI-1 00-12 at Step 5.0[12].

Initials Date Time

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 17 of 29

5.0 INSTRUCTION STEPS (continued)

[6] IF power is being reduced(less than 100/0) for any of the following reasons:

(N/A if entering 3-GOI-100-12 to recover from a Recirc Pump Trip or power

reduction of >10 01b)

Weekly Control Rod Exercise

Main Turbine Valve Testing

Ultimate Heat Sink temperature >92.5°F

PERFORM the following:

[6.1] REDUCE Recirculation flow. REFER TO 3-01-68.

Initials Date Time

[6.2] MAINTAIN Reactor thermal power within the limits shown on

Illustrations 1, 2, ICS, and 0-TI-248 as appropriate.

Initials Date Time

[6.3] WHEN desired to raise power after testing is complete, THEN

PERFORM the following as directed by Unit Supervisor. (N/A

Steps 5.0[7] through 5.0[19].

RAISE Recirculation flow. REFER TO 3-01-68.

MAINTAIN thermal power within limits shown on Illustrations 1, 2,

ICS, and 0-TI-248, Station Reactor Engineer.

Initials Date Time

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 18 of 29

5.0 INSTRUCTION STEPS (continued)

[7] IF required for power maneuvering, THEN

PERFORM the following as directed by Reactor Engineer using

3-SR-3.'1.3.5(A). (N/A if entering 3-GOI-100-12 to recover from a Recirc

Pump Trip)

A. OBTAIN the Control Rod Movement Data Sheet.

B. ALIGN control rods.

(R)

Initials Date Time

Reactor Engineer

NOTE

Refer to Illustration 1, ICS and/or 0-TI-248 for Reactor Thermal Limits.

[8] REDUCE reactor power by a combination of control rod insertions and core

flow changes, as recommended by Reactor Engineer.

REFER TO 3-SR-3.1.3.5(A) and 3-01-68. (N/A if entering 3-GOI-1 00-12 to

recover from Recirc Pump Trip)

(R)

Initials Date Time

[9] PERFORM the following while reducing Reactor power:

(N/A if entering 3-GOI-100-12 to recover from a Recirc Pump Trip)

[9.1] MONITOR Core thermal limits using Illustration 1, ICS, and/or 0-TI-248.

Initials Date Time

[9.2] MONITOR Power reduction on Nuclear Instrumentation.

(R)

Initials Date Time

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 19 of 29

5.0 INSTRUCTION STEPS (continued)

CAUTION

When operating with less than the full complement of condensate pumps, condensate

booster pumps, and/or reactor feedpumps, careful monitoring of motor amp limitations,

feedpump speed limitations, and reactor vessel makeup capacity should be performed.

This should include discussion between shift operating crews for contingency actions (e.g.

tripping one of the remaining Recirc Pumps) should any remaining Condensate/Feedwater

pumps trip.

NOTE

A condensate pump, condensate booster pump, and/or a reactor feedpump may be

removed from service at less than 85% power to support maintenance activities as directed

by the Shift Manager/Unit Supervisor.

[10] WHEN Reactor power is less than 85°A> , THEN

PERFORM the following: (N/A if entering 3-GOI-1 00-12 to recover from a

Recire Pump Trip).

[10.1] SHUT DOWN one of three Reactor Feedpumps, as directed by the

Shift Manager or Unit Supervisor. REFER TO 3-01-3. (N/A if NOT

performed.)

Initials Date Time

[10.2] REMOVE Condensate Demineralizers as desired. REFER TO 3-01-2A.

(N/A if NOT performed.)

Initials Date Time

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 20 of 29

5.0 INSTRUCTION STEPS (continued)

CAUTIONS

1) Failure to monitor SJAE/OG CNDR CNDS FLOW, 3-FI-2-42, on Panel 3-9-6 for proper

flow may result in SJAE isolation.

2) Changes in condensate system flow may require adjustment to SPE CNDS BYPASS"

3-FCV-002-0190, either in the Control Room or locally. Personnel adjusting this valve

locally is required to be in direct communication with the Control Room.

[10.3] MAINTAIN flow between 2 x 1061bm/hr and 3 x 106 Ibm/hr on SJAE/OG

CNDR CNDS FLOW, 3-FI-2-42 using COND SPE BYPASS FLOW

CONTROL, 3-HS-2-190A. REFER TO 3-01-2.

Initials Date Time

[10.4] SHUT DOWN one of three Condensate Booster Pumps, as directed by

Shift Manager or Unit Supervisor. REFER TO 3-01-2. (N/A if NOT

performed.)

Initials Date Time

[10.5] SHUT DOWN one of three Condensate Pumps, as directed by Shift

Manager or Unit Supervisor. REFER TO 3-01-2. (N{A if NOT

performed.) ,

Initials Date Time

NOTE

Duration of out of service time for remaining equipment should be taken into consideration

in the Step 5.0[10.6] evaluation.

[10.6] REQUEST Reactor Engineering to evaluate the need to lower Control

Rod Line (as a contingency) should the remaining Condensate or

Feedwater Pumps trip during time period pumps will be out of service.

Initials Date Time

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 21 of 29

5.0 INSTRUCTION STEPS (continued)

[11] IF necessary to continue power reduction to approximately 50%, THEN

REDUCE Reactor power by combination of control rod insertions per

3-SR-3.1.3.5(A) and core flow changes per 3-01-68, as recommended by

Reactor Engineer and directed by Unit Supervisor. (N/A if NOT performed or

if entering 3-GOI-1 00-12 to recover from a Recirc Pump Trip)

Initials Date Time

[12] IF Reactor Power is required to be lowered for Recirc Pump start-up or

shut down, THEN

LOWER Reactor Power to desired range required by 3-01-68.

(Otherwise N/A).

Initials Date Time

[13] IF continued power reduction is necessary (typically below approximately

50% power) and the Unit Supervisor determines reduction is outside scope of

this procedure, THEN

EXIT this procedure and PERFORM 3-GOI-100-12A. (Otherwise N/A)

Initials Date Time

NOTE

Illustration 1 provides Reactor thermal limits.

[14] REVIEW Precaution & Limitations. REFER TO Section 3.0.

(R) _

Initials Date Time

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 22 of 29

5.0 INSTRUCTION STEPS (continued)

[15] WHEN desired to restore Recirc System to dual loop operation, THEN

PERFORM the following: (N/A if Recirc System is already in dual loop

operation)

[15.1] RESTORE Recirc System to dual loop operation. REFER TO 3-01-68,

Recirc Pump Startup.

Initials Date Time

[15.2] NOTIFY Reactor Engineering to perform 3-SR-3.4.1 (DLO)

and O-TI-248, as necessary.

Initials Date Time

[16] BEFORE raising reactor power, NOTIFY Radiation Protection that an RPHP

is in effect for intentions to raise reactor power level, and RECORD time

Radiation Protection notified in NOMS Narrative Log.

(R)

Initials Date Time

[16.1] VERIFY appropriate data and signatures recorded on Appendix A in

accordance with Appendix A instructions.

(R)

Initials Date Time

[17] WHEN desired to restore Reactor power, THEN

PERFORM the following:

[17.1] RESTORE Reactor power using control rod withdrawals in combination

with core flow changes, as recommended by Reactor Engineer and

directed by Unit Supervisor. REFER TO 3-SR-3.1.3.5(A) and 3-01-68.

Initials Date Time

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 23 of 29

5.0 INSTRUCTION STEPS (continued)

CAUTIONS

1) Failure to monitor SJAE/OG CNDR CNDS FLOW, 3-FI-2-42, on Panel 3-9-6 for proper

flow may result in SJAE isolation.

2) Changes in Condensate system flow may require adjustment to SPE CNDS BYPASS,

3-FCV-002-0190, either in the Control Room or locally. Personnel adjusting this valve

locally is required to be in direct communication with the Control Room.

[18] WHEN Reactor power is approximately 70 0/0, THEN

PERFORM the following as directed by Unit Supervisor:

[18.1] PLACE additional condensate demineralizers in service to support

starting third Condensate Pump, Condensate Booster Pump, and

Reactor Feedpump. REFER TO 3-01-2A. (N/A if NOT performed)

Initials Date Time

CAUTION

When operating with less than the full complement of condensate pumps, condensate

booster pumps, and/or reactor feedpumps, careful monitoring of motor amp limitations,

feedpump speed limitations, and reactor vessel makeup capacity should be observed.

NOTE

A Condensate pump, Condensate booster pump, and/or a Reactor feedpump may be

returned to service between 70 0h and 85°h power following maintenance activities as

directed by the Shift Manager or Unit Supervisor.

[18.2] START third Condensate Pump. REFER TO 3-01-2.

(N/A if NOT performed.)

Initials Date Time

[18.3] START third Condensate Booster Pump. REFER TO 3-01-2.

(N/A if not performed.)

Initials Date Time

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 24 of 29

5.0 INSTRUCTION STEPS (continued)

[18.4] MAINTAIN flow between 2 x 1061bm/hr and 3 x 1061bm/hr on SJAE/OG

CNDR CNDS FLOW, 3-FI-2-42 using COND SPE BYPASS FLOW

CONTROL, 3-HS-2-190A. REFER TO 3-01-2.

Initials Date Time

[18.5] PLACE third Reactor Feedpump in service. REFER TO 3-01-3. (N/A if

NOT performed.)

Initials Date Time

[19] IF desired to raise power with only two(2) Reactor feedpumps in service,

THEN

RAISE Reactor power, as desired, maintaining each Reactor feedpump less

than 5050 RPM. (N/A if NOT performed)

Initials Date Time

[20] WHEN desired to restore Reactor power to 100 % , THEN

PERFORM the following as directed by Unit Supervisor and recommended by

the Reactor Engineer:

RAISE power using control rods or core flow changes.

REFER TO 3-SR-3.3.5(A) and 3-01-68.

MONITOR core thermal limits (Illustration 1).

Initials Date Time

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 25 of 29

5.0 INSTRUCTION STEPS (continued)

NAME (print) INITIALS

Performed by:

Reviewed by:

Shift Manager Date

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 26 of 29

Illustration 1

(Page 1 of 1)

Reactor Thermal Limits

Administrative Reactor Thermal Limits for MFLPD, MFLCPR, MAPRAT, and CTP (MWt) are

listed in 0-TI-248, Appendix for Administrative Limits. These limits should be reviewed with

Reactor Engineer.

Monitoring of core thermal limits at the following frequencies is recommended:

A. Following completion of planned power rises with control rods or recirc flow.

B. Following any unexpected power change.

C. Once every two hours during steady state operation.

If core mqnitoring software becomes unavailable, the Shift Manager and Reactor Engineer

are required to determine the appropriate frequency for monitoring core thermal limits using

the backup core monitoring computer taking into consideration current core conditions and

margin to thermal limits. Power changes should not normally be made without the core

monitoring software being available.

Maximum steady-state power averaged over 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is 3458 MWt. However, the reactor

should not be operated such that the steady state power (as indicated by 30 min avg, 1 hr

avg, or 2 hr avg) is above 3458 MWt.

Minor variations in process parameter inputs to the process computer may result in individual

edits or indications above 3458 MWt while true steady state core thermal power is

~3458 MWt. Normal variation is within 5 MWt of steady-state core thermal power. Running

averages (from core thermal power summary on the nuclear heat balance display) are not as

sensitive. The following g+uidance is provided:

RESULT (MWt) GUIDANCE

> 3463 REDUCE power.

3458 to 3463 ALLOW time for recent perturbations to

settle. Evaluate trend. IF the trend indicates

steady state core thermal power will be

above 3458, THEN

REDUCE power.

> 3458(any running average) REDUCE power.

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 27 of 29

Illustration 2

(Page 1 of 1)

Reactor Thermal Power Versus Ultimate Heat Sink Temperature Limit

101 :

...

100

100 :

\ Unacceptable ....

..-..

0~

""-'

Ir-

ev

~

99

\99 01

~

c;

0

0..

E I!:i- __

\

Ir- ._.L_II-

ev * ~L:L:~J- LClUle

..c:

I-

Ir-

0

......, 98

u

cu

ev

0:::

97

-. )) 96.3

96

91 92 93 94 95

RHRSW Forebay Inlet Temperature (degrees F)

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 28 of 29

Appendix A

(Page 1 of 2)

Radiation Protection Notifications

INSTRUCTIONS FOR APPENDIX A DATA ENTRY

This appendix provides record of Radiation Protection notifications, RPHPs, and required

signatures made during the performance of this Gal. Each notification step in this procedure,

or in any referenced support procedure, that requires Appendix A be entered requires the

following instructions to be used to complete the appropriate parts of the data entry page.

Copies are made as needed to support this data entry.

A. Ops ENTER name of the Radiation Protection Representative notified with date and time

of notification. Time of notification is also required in NOMS narrative log.

B. Ops ENTER step number (including Section number) associated with notification

requirement. If the notification is directed from a support procedure, then enter the

procedure number and current revision number

C. For all RPHP notifications, Radiation Protection DETERMINE if the RPHP is required to

prevent unintended exposures and/or to implement RCI-17, Control of High Radiation

Areas and Very High Radiation Area controls. IF RPHP is identified in a support

procedure to this GOI, THEN DETERMINE if an RPHP is also necessary for the Gal.

CONFER with Operations, as necessary.

D. For each identified procedure RPHP, Radiation Protection Supervisor's signature is

required to release the RPHP for the action associated with affected step. This signature

signifies one of two conditions: [SOER 01-1, Tech Spec 5.7, BFN PER 126211]

1. Radiation Protection actions are completed to prevent unintended exposures

and/or RCI-17 requirements have been met and any personnel working within

affected areas are on an appropriate RWP for the anticipated radiological

conditions.

OR

2. No actions were necessary because appropriate controls were already in place.

E. WHEN the use of this procedure is completed, FORWARD copies of the completed

appendix pages to the Radiation Protection Supervisor.

If, while performing this procedure, or while performing a support procedure, Radiation

Protection personnel, Unit Operator, Unit Supervisor, or other knowledgeable shift member

identifies the need for a RPHP, then "RPHP" is written to the left of the affected procedure

step number (this GOI or the support procedure. If the RPHP is identified for a support

procedure, then RPHP is also placed to the left of the step in this GOI that initiates the

support procedure and then A through E above is performed, as applicable.

BFN Power Maneuvering 3-GOI-100-12

Unit 3 Rev. 0031

Page 29 of 29

Appendix A

(Page 2 of 2)

Name Of Radiation Protection Person Notified: _

Date: / / - - - Time: - - - - - - -

Step# Procedure: (if not this procedure) Rev: _

RPHP Required by Ol? _ _(V) _ _(N) RPHP Required For Gal? _ _(V) __(N)

RCI-17 Controls Necessary? _ _(V) _ _(N)

Radiation Protection Supervisor Signature for Release

_ _ _ _ _ _ _ _ _ _ _ _ _ Date: / _ Time: - - - - - - -

Comments:

Name Of Radiation Protection Person Notified: _

Date: / / - - - Time: - - - - - - -

Step# Procedure: (if not this procedure) Rev: _

RPHP Required by Ol? _ _(V) _ _(N) RPHP Required For Gal? _ _(V) __(N)

RCI-17 Controls Necessary? _ _(V) _ _(N)

Radiation Protection Supervisor Signature for Release

- - - - - - - - - - - - - Date: - - - - / - - - Time: _

Comments:

FORWARD copies of the completed appendix pages to the Radiation Protection

Supervisor.

HLTS-3-3

Revision 0

Page 1 of 22

SIMULATOR EXERCISE GUIDE

TITLE POWER REDUCTION, CORE SPRAY 2D PUMP INADVERTANT START,

RECIRCULATION PUMP TRIP, REACTOR POWER OSCILLATIONS, ATWS WITH

MSIVS OPEN

REVISION o

DATE January 2, 2008

PROGRAM BFN Operator Training - HLT

NOTE: Provide examiners with copy of 2-01-99 section 8.3

Also provide copy of 0-01-578 Illustration 3

.

PREPARED BY:

erations Instructor)

REVIEWED BY: N!.,4

OR Lead Instr ctor or Designee)

REVIEWED BY:

VALIDATION

BY:- Date

LOGGED-IN:

(Librarian) Date

TASKS LIST

UPDATED: Date

HLTS-3-3

Revision 0

Page 2 of 22

NUCLEAR TRAINING

REVISIONISROAGE LOG

REVISION DESCRIPTION OF DATE PAGES REVIEWED BY

NUMBER REVISION AFFECTED

0 INITIAL 04/01/07 All

HLTS-3-3

Revision 0

Page 3 of 22

I. PROGRAM: BFN Operator Training

II. COURSE: Examination Guide

III. TITLE: POWER REDUCTION,CORE SPRAY SR FAILURE, RECIRCULATION PUMP

TRIP, REACTOR POWER OSCILLATIONS, ATWS WITH MSIVS OPEN

IV. LENGTH OF LESSON: 1 to 1 % hours

V. Training Objectives

A. Terminal Objective

1. Perform routine shift turnover, plant assessment and routine shift operation in

accordance with BFN procedures.

2. Given uncertain or degrading conditions, the operating crew will use team skills to

.conduct proper diagnostics and make conservative operational decisions to remove

equipment/unit from operation. (SOER 94-1 and SOER 96-01)

3. Given abnormal conditions, the operating crew will place the unit in a stabilized

condition per normal, abnormal, annunciator, and emergency procedures.

B. Enabling Objectives

1. Theoperatinq crew will recognize and respond to an inadvertent start of a Core

Spray pump and determine required actions per Technical Specifications.

2. The operating crew will recognize and respond to a recirculation pump trip with

reactor power oscillations in accordance with 3-AOI-68-1.

3. The operating crew will recognize and respond to CRD pump 3A trip per 3-AOI-

85-3.

4. The operating crew will recognize and respond to an ATWS in accordance with

EOI-1 and C-5.

5. The operating crew will recognize and respond to loss of 3A 480v RMOV board

and determine required actions per Technical Specifications

6. The operating crew will recognize and respond to high radiation in accordance

with EOI-3.

HLT8-3-3

Revision 0

Page 4 of 22

VI. References: The procedures used in the simulator are controlled copies and are used in development

and performance of simulator scenarios. Scenarios are validated prior to use, and any

procedure differences will be corrected using the procedure revision level present in the

simulator. Any procedure differences noted during presentation will be corrected in the

same manner. As such, it is expected that the references listed in this section need only

contain the reference material which is not available in the simulator.

VII. Training Materials: (If needed, otherwise disregard)

A. Calculator

B. Control Rod Insertion Sheet

C. Stopwatch

D. Hold Order / Caution tags

E. Annunciator window covers

F. Steam tables

HLTS-3-3

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Page 5 of 22

IX. Console Operation Instructions

A. Scenario File Summary

1. File: bat HLTS3-3

MF/RF/IOR DESCRIPTION

1) imffw26b 0 IB' FW flow failure

2) bat tohpci Tags out HPCI

3) imfth23 (e3 0) 2.5 15:00 Fuel failure

4) imf rp08a RPS A1 scram failure

5) imf rp08b RPS A2 scram failure

6) trg e1 CSD Set trigger e1 to file CSD

7) trg e2 CSDHS Set trigger e2 to file CSDHS

8) trg e3 MODESW Set trigger e3 to file MODESW

9) trg e1= dor zdihs7542a Trigger e1 initiates command

10) trg e2= ior zloil7542 on Trigger e2 initiates command

11) imf tc02 0 Fails Bypass valves closed

2. File: bat HLTS3-3-1

MF/RF/IOR DESCRIPTION

1) imfth03a (none 10:00) Trips 2A Recirc. pump

2) imf th03b Trips 2B Recirc. pump

3) imf cr02a 65 10:00 Power Oscillations

4) batatws90 90°A> Hydraulic ATWS

5) Imf tc01 (e35:00) Fail bypass valves closed 5 min after

mode sw

3. File: bat HLTS3-3-2

MF/RF/IOR DESCRIPTION

1 mrf ed27 rackin Rackin alternate feeder for 2A RMOV bd

2 mrf ed09 alt Transfer 2A RMOV bd to alternate

3 bat NRC/HLTS1 0-3 (none 1:00) Initiate file NRC/HLTS10-3

HLTS-3-3

Revision 0

Page 6 of 22

IX. Console Operation Instructions

A. Scenario File Summary

3. File: bat HLTS3-3-3

MF/RF/IOR DESCRIPTION

1 mrf rpO 1 reset Reset RPS A

2 mrf rp09 reset Reset RPS A Gross Failure alarm

4. File bat tohpci

MF/RF/IOR DESCRIPTION

1) ior ypomtrglesh fail_cn_po Tag gland seal exhauster

ior ypovfcv733a close 73-3 close

2) ior ypovfcv733 fail_now Tag FCV 73-3

ior ypovfcv7316 fail_now Tag FCV 73-16

3) ior ypovfcv7381 fail_now Tag FCV 73-81

4) ior zdihs7347a ptl Tag HPCI Aux oil pump

5) ior zohs7347a[1] off

6) imf hp05 HPCI trip

5. File bat app01f

MF/RF/IOR DESCRIPTION

1) mrf rp13a byp Bypasses automatic scrams

2) mrf rp13b byp (Appendix 1F)

6. File bat xferrmov2a

MF/RF/IOR DESCRIPTION

1 mrf ed27 rackin Rackin alternate feeder for 2A RMOV bd

2 mrf ed09 alt Transfer 2A RMOV bd to alternate

HLTS-3-3

Revision 0

Page 7 of 22

IX. Console Operation Instructions

A. Scenario File Summary

7. File batapp02

MF/RF/IOR DESCRIPTION

1) mrf rp12a test Bypasses ARI

2) mrf rp12b test (Appendix 2)

8. File batapp08ae

MF/RF/IOR DESCRIPTION

1) mrf rp06a byp Bypasses MSIV isolation on low

2) mrf rp06b byp RPV water level (Appendix 8A)

3) mrf rp06c byp

4) mrf rp06d byp

5) mrf rp14a byp Bypasses Rx Bldg ventilation

6) mrf rp14b byp isolation on low RPV level

9. File batatws90

MF/RF/IOR DESCRIPTION

1) imf rd17a SDV level switch failure

2) Imfrd17b

3) imf rd09a 90 90% hydraulic ATWS

4) Imf rd09b 90

10. File bat sdvtd

MF/RF/IOR DESCRIPTION

a) dmf rd17a Deletes SDV level switch failure

b) dmfrd17b

c) imf rd17a (none 7:00) Inserts level switch failure after 7

minutes

d) imf rd17a (none 7:00)

HLTS-3-3

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Page 8 of 22

IX. Console Operation Instructions

B. Console Operator Manipulations

ELAP TIME DESCRIPTION/ACTION

Sim.Setup reset 28 100% MOC

Sim. Setup restorepref HLTS3-3 Establishes Function Keys

Sim. Setup setup Verify Function Keys

Sim. Setup esc Clears Function Key Popup

Sim.Setup F3 See scenario file summary (bat HLTS3-3)

Sim.Setup manual Tag out HPCI. Hang out of service cover

on "B" FW Flow Indicator

Sim.Setup manual HPCI AOP and SPE pumps in PTL. Place

Main Generator Voltage Regulator in

Manual

ELAP TIME PFK DESCRIPTION/ACTION

2 minutes after Unit at 950/0 power F4 Core Spray pump 3D start

ior zdihs7542a start

If lockout light does not illuminate when CS F5 Illuminates lockout light

pump stopped, then lor zloil7542 on

2 minutes after Tech Specs addressed for F6 Control Rod xx-xx drifting in.

CS pump imf rd07 xx-xx

3 minutes after 3B CRD pump in service F7 Trip 3A 480v RMOV board normal feeder

imf ed12a

ROLE PLAY: When sent to investigate board loss, Report the breaker on the 480v SID bd feeding the normal feeder

is tripped and will not re-close

When requested to transfer board F8 Transfer RMOV board 3A to alternate

bat xferrmov2a

If requested to place A RPS normal F9 Resets A RPS and gross failures

bat HLTS3-3-3

If requested to place A RPS on xfmr F10 Place A RPS on xfmr

mrf rp04 a

~MORE FOLLOWS~

HLTS-3-3

Revision 0

Page 9 of 22

IX. Console Operation Instructions

B. Console Operator Manipulations

ELAP TIME PFK DESCRIPTION/ACTION

If requested to reset gross failures F11 Resets gross failures

mrf rp09 reset

If requested to secure A and B SBGT F12 Stops SBGT trains A and B

bat sgt_stop

Reset local RWCU panel <shift>F1 mrf an01e reset

3 minutes after Tech Specs for RMOV <shift>F2 Initiates Recirc pump trips and power

board addressed bat HLTS3-3-1 oscillations. ATWS

After scram inserted <shift>F3 SDV switches enabled

batsdv

When appendix 2 requested, wait 3 minutes <shift>F4 Bypass ATWS/ARI circuit

batapp02

When requested to perform appendix 1F, wait 5 <shift>F5 Jumper out scram logic

minutes bat app01f

When requested to perform Appendix 8A and <shift> F6 Allows restart of Reactor/Refuel zone

8E, wait 5 minutes batapp08ae ventilation

When scram is reset <shift> F7 SDV switches enabled

bat sdvtd

If requested to close 3-FCV-85-586, wait 5 <shift> F8 Provides drive water pressure for rod

minutes mrf rd06 close insertion

If requested to open 3-FCV-85-586, wait 1 <shift> F9 Pressurizes charging water header

minute mrf rd06 open

nd

When Reactor is manually scrammed (2 time) <shift>F3 SDV switches enabled

batsdv

nd

When Reactor is scram reset (2 time) <shift> F7 SDV switches enabled

bat sdvtd

Removes power oscillations

<shift> F10

dmf cr02a Deletes ATWS

<shift> F11

bat atws-1

Terminate the scenario when the following conditions are satisfied or upon request of the Chief Examiner:

1. All rods fully inserted

2. RPV water level +2" to +51"

3. Reporting requirements made

HLTS-3-3

Revision 0

Page 10 of 22

IX. SCENARIO SUMMARY:

The unit is operating at 100% power with a 5°A, power reduction scheduled. HPCI is tagged out

for maintenance on the Auxiliary Oil Pump and is expected to be returned to service within the

next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . It has been out of service for 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months .

The Main Generator voltage regulator was placed in Manual to allow PMs on the Automatic

regulator. PMs are complete and the voltage regulator can be returned to Automatic.

Core Spray 3D pump inadvertent start is received and the Crew must consult Tech Specs to

determine required actions.

Control Rod 18-35 drifts inadvertently into the core.

3A 480v RMOV board is lost due to breaker failure, the board will be transferred to alternate

supply. RPS half scram and PCIS isolations must be reset. SRO will refer to Tech Specs.

38 Recirc pump trips, resulting in power oscillations with some fuel failure. While responding to

the power oscillations per AOI-68-1, 3A Recirc pump trips and a manual scram must be inserted.

The crew will experience a hydraulic ATWS and respond per 3-EOI-1. The SDV will fail to drain

totally, thus requiring multiple reactor scrams to insert control rods.

HLTS-3-3

Revision 0

Page 11 of 22

X. Information to Evaluators:

A. Ensure recorders are inking and recording and ICS is active and updating.

B. Assign Crew Positions based on the required rotation.

1. SRO: Unit Supervisor

2. ATC: Board Unit Operator

3. BOP: Desk Unit Operator

C. Conduct a shift turnover with the Shift Manager and provide the Shift Manager with a copy of the

Shift Turnover.

D. Direct the shift crew to review the control board and take note of present conditions, alarms, etc.

E. Terminate the scenario when the following conditions are satisfied are at the request of the

floor/lead instructor/evaluator.

1. All rods inserted

2. Water level +2" to +51"

3. Reporting requirements have been made

HLTS-3-3

Revision a

Page 12 of 22

XI. Simulator Event Guide

Event 1: POWER REDUCTION AND VOLTAGE REGULATOR TO AUTOMATIC

POSITION EXPECTED ACTIONS

ATC Reduces power with recirculation flow

BOP Peer checks during power reduction

SRO Directs BOP to return voltage regulator to automatic per 3-01-47

section 8.14.

BOP VERIFY VOLTAGE REGULATOR MAN/AUTO SEL, 3-HS-57-27, is

in MAN.

PLACE GENERATOR FIELD VOLTAGE AUTO ADJUST (gOP), 3-

HS-57-26, to RAISE UNTIL the upper limit is reached (red light

illuminated).

PLACE GENERATOR FIELD VOLTAGE AUTO ADJUST (gOP), 32-

HS-57 -26, to LOWER UNTIL the lower limit is reached (green light

illuminated).

ADJUST GENERATOR FIELD VOLTAGE AUTO ADJUST (gOP), 3-

HS-57 -26, UNTIL GEN TRANSFER VOLTS, 2-EI-57-41, indicates

zero.

PLACE VOLTAGE REGULATOR MAN/AUTO SEL, 3-HS-57-27, in

AUTO.

HLTS-3-3

Revision 0

Page 13 of 22

XI. Simulator Event Guide

Event 2: SPURIOUS START OF 3D CORE SPRAY PUMP

POSITION EXPECTED ACTIONS

BOP Reports start of 3D Core Spray pump and verifies no valid

automatic start signal.

SRO Directs trip of 3D Core Spray pump

BOP Trips 3D Core Spray pump. Informs SRO that Lockout indicator for

3D Core Spray Pump is illuminated.

SRO Consults Tech Spec 3.5.1 determines that a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months LCO is in

effect with HPCI and one (1) Low Pressure ECCS system

inoperable.

HLTS-3-3

Revision a

Page 14 of 22

XI. Simulator Event Guide

Event 3: CONTROL ROD DRIFT IN

POSITION EXPECTED ACTIONS

ATC Announces "Rod Drift" alarm

Identifies rod 18-35 as drifting in

SRO Directs actions per 3-AOI-85-5

SRO Directs rod be continuously inserted to 00

ATC Continuously inserts rod to 00

ATC Informs Reactor Engineer

ATC Checks Thermal Limits

Verifies CRD operating parameters within limits

ATC Directs AUO to check the following per 3-AOI-85-5:

- scram pilot air header aligned

- check scram outlet valve for leakage

- check scram inlet valve for leakage

ATC Directs charging water to 18-35 be closed

SRO Declares accumulator inoperable per Tech Spec 3.1.5 and addresses

actions (when charging water is isolated)

ATC Directs scramming of affected rod from panel 9-16 in Aux. Inst. Room

ATC Directs operator to Aux. Inst. Room for rod scram

ATC

Establishes communication with operator and AUO

Directs operator to scram rod by taking scram switch to "down"

position

Verifies rod Full In overtravel

Direct operator to return scram switch to 'up' position

ATC Reports rod settles to 00 position

ATC Directs reopening Charging water isolation valve

ATC Resets drift and accum. Lights/alarms

SRO Initiates actions to determine CR operability and suggests actions

including maintenance and inspection

HLTS-3-3

Revision 0

Page 15 of 22

XI. Simulator Event Guide

Event 4: LOSS OF 3A 480V RMOV 8D

POSITION EXPECTED ACTIONS

CREW Announces loss of 3A 480v RMOV board, RPS half scram and

PCIS isolations.

Refers to 0-01-578 p&1 3.0 Z, Illustration 3 and 3-47E751-1

require removing 80 kva accident load to prevent overload on D

DIG (3D Core Spray would be a good choice)

SRO Directs Outside US to transfer RMOV board to alternate and

restore RPS A.

Directs ATC to reset half scram after RPS is restored, 80P to

reset PCIS and recover from isolations per 3-01-99 section 8.3.

(attached)

Refers to ITS and determines Unit is in 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months LCO from 3.8.7.8.

ATC Resets A Channel half scram after RPS A is restored.

80P Resets PCIS and restores:

Reactor and Refuel zone ventilation

S8GT

ECCS Keep Fill system

Drywell DP air compressor

Drywell Floor and Equipment drain pumps

Radiation Monitoring system

TIP system

HLTS-3-3

Revision 0

Page 16 of 22

XI. Simulator Event Guide

Event 5: RECIRCULATION PUMP TRIP/POWER OSCILLATIONS/ATWS

POSITION EXPECTED ACTIONS

ATC Recognizes B Recirculation pump trip

SRO Directs 3-AOI-68-1 entry

Contacts Rx Engr to place APRM's in SLO mode (If time permits)

ATC Recognizes power oscillations

Inserts rods on emergency shove sheet

BOP Verifies flow on A Recirc pump < 46,600 gpm and jet pump flow>

41,100Ibm/hr

ATC/BOP Notices failure to Scram on OPRM Trip or anticipates trip and

inserts a manual scram

SRO Directs manual reactor scram

ATC Inserts manual scram

Places Mode switch in shutdown

Recognizes hydraulic ATWS

Provides scram report

SRO Enters EOI-1/C-5 and verifies

Verifies Mode switch to shutdown

ARI initiated

Directs ADS inhibited

BOP Inhibits ADS

SRO Directs Appendix 1F and Appendix 2

Directs Appendix 1D

HLTS-3-3

Revision 0

Page 17 of22

XI. Simulator Event Guide

Event 6: ATWS WITH FUEL FAILURE

POSITION EXPECTED ACTIONS

BOP Recognizes 3A Recirc pump trip, if not tripped by ATC Operator

due to ATWS

Crew Recognize and reports

"OG Pretreatment Radiation High"

"OG Annual Release Limit Exceeded"

"Turbine Building High Radiation"

Notifies RadCon, and Chemistry

Evacuates appropriate area of Turb. Bldg.

SRO Directs RPV pressure be maintained 800-1000 psig

BOP Controls RPV pressure between 800-1000 psig with SRVs

SRO Directs water level be lowered to control power per Appendix 4

Directs Appendix 8A and 8E be performed

Reports SAE 1.2-S

Crew Monitors suppression pool temperature

ATC After Appendix 1F and 2 complete resets scram and drains SDV

ATC Inserts control rods per Appendix1 D

BOP Maintains water level as directed with RFP / RCIC

BOP Maintains Pressure control using the following appendices:

11D Main Steam line Drains

11F RFPT

11A SRV's

HLTS-3-3

Revision 0

Page 18 of 22

XI. Simulator Event Guide

Event 6: ATWS WITH FUEL FAILURE (continued)

POSITION EXPECTED ACTIONS

Crew Recognizes some control rod movement, but all control rods not in

SRO Directs reactor reset, drain SDV, and re-scram

Directs SLC injection if Torus temperature approaches 110 deg.

Enters EOI-2 on Torus water level

Directs Venting per Appendix 12

Directs placing H202 monitors in service

BOP Performs Appendix 12 to vent Torus

Places H20 2 monitors in service

ATC Maintains water level as directed to control power

ATC After SDV drained directs 3-85-586 re-opened

After accumulators recharged, scrams reactor and verifies all rods in

ATC

SRO Directs level be restored +2" to +51"

SRO Directs SLC stopped (if injected)

Crew Recognize RM-90-29A Rx Bldg High Radiation

(conditional)

BOP Evacuates Reactor Building

SRO Enters EOI-3 on high Rx Bldg radiation and directs ventilation

restored per Appx 8F.

ATC/BOP Closes MSIV's as directed by ARP for MSL HiHi Rad

(If received)

HLTS-3-3

Revision 0

Page 19 of 22

XII. Crew Critical Tasks (CCT)

SAT/UNSAT

1. Manual scram due to Scram failure of OPRM Trip

2. Prevent ADS actuation

3. Controls power by :

Inserting control rods per RC/Q-21

Lowering water level

4. Maintains RPV water level above -180" with rods out

5. When all rods are inserted restores and maintains RPV water level

above TAF.

HLTS-3-3

Revision 0

Page 20 of 22

XIII. Scenario Verification Data

EVENT KIA Number RO SRO

1. Control Rod Drift 201002A2.02 3.2 3.3

2. Loss of 3A 480V RMOV Board 262001 K3.06 3.8 4.1

3. APRM Failure 215005A2.03 3.6 3.7

2.1.12 2.9 4.0

4. Recirculation Pump Trip/Power Oscillations/ATWS - 295001 3.3 3.4

295025 3.8 3.9

295037 3.9 4.0

HLTS-3-3

Revision 0

Page 21 of 22

SCENARIO REVIEW CHECKLIST

SCENARIO NUMBER HLTS-3-3

7 Total malfunctions Inserted; List: (4-8)

1. Recirc Trip

2. Power Oscillations

3. ATWS

4. Fuel Failure

5. Core Spray pump 3D trip

6. Loss of 480v RMOV bd 3A

7. Control Rod Drift

2 Malfunctions That Occur After EOI Entry; List: (1-4)

1) ATWS

2) Fuel Failure

3 Abnormal Events; List: (1-3)

1) Control Rod Drift

2) Recirc Trip

3) 480v RMOV bd trip

2 Major Transients; List: (1-2)

1) ATWS

2) Fuel Failure

3 EOls used; List: (1-3)

1) EOI-1

2) EOI-2

3) EOI-3

EOI Contingencies Used; List: (0-3)

1) C5

75 Run Time (minutes)

35 EOI Run Time (minutes); _~ °h of Scenario EOI Run Time

5 Crew Critical Tasks (2-5)

Yes Technical Specifications Exercised (yes/no)

Page 22 of 22

XV. SHIFT TURNOVER INFORMATION

Equipment out of service/LCOs: HPCI tagged out for 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months to repair Auxiliary Oil Pump. Expected

back in 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . Flow indicator 3-78B out of service. 1M's are looking for a new transmitter. Main

Generator voltage regulator in manual for PMs on Automatic voltage regulator. Spare RBCCW pump in

service to Unit 2.

Operation/Maintenance for the Shift: Reduce power to 95°~ with recirculation flow (due to system load not

required). PMs on voltage regulator complete. return voltage regulator to automatic.

Unusual Conditions/Problem Areas: ~~N~o~n~e~~~~~~~~~~~~~~~~~~~~~~

Browns Ferry Nuclear Plant

Unit 2

Operating Instruction

2-01-99

Reactor Protection System

Revision 0073

Quality Related

Level of Use: Continuous Use

Effective Date: 04-02-2007

Responsible Organization: OPS, Operations

Prepared By: Terry Kenneth Boyer

Approved By: James A. McCrary

BFN Reactor Protection System 2-01-99

Unit 2 Rev. 0073

Page 2 of 77

Current Revision Description

Type of Change DCN 60717-03, Editorial Tracking Number: 084

PCR: 06002112,06003343

DCN 60717-03 removed Containment Isolation valves 2-FCV-32-0062 and

2-FCV-32-0063, associated DCA compressor suction piping, associated electrical and

pneumatic controls, and PCIS trip signal circuitry in the Control Room.

Page 45 - Deleted Step 8.3[8]. 2-HS-32-62A and 2-HS-32-63A were removed by DeN

60717-03. (PCR 06002112)

Page 62 - Deleted FCV-32-62 and FCV-32-63 which were removed by DCN 60717-03.

(PCR 06002112)

Page 64 - Changed the FUNCTION/SYSTEM of FCV-75-57 to PSC Pump Suction

Inboard Isolation Valve. Editorial change to reflect plant conditions. (PCR 06003343)

Page 65 - Changed the FUNCTION/SYSTEM of FCV-75-58 to PSC Pump Suction

Outboard Isolation Valve. Editorial change to reflect plant conditions. (PCR 06003343)

THIS REVISION DOES NOT AFFECT SYSTEM STATUS

BFN Reactor Protection System 2-01-99

Unit 2 Rev. 0073

Page 3 of 77

Table of Contents

1.0 PURPOSE 5

2.0 REFERENCES 5

2.1 Technical Specifications 5

2.2 Final Safety Analysis Report 5

2.3 Plant Instructions 5

2.4 Plant Drawings 6

2.5 Vendor Manuals 7

2.6 Miscellaneous Documents 7

3.0 PRECAUTIONS AND LIMITATIONS 8

4.0 PRESTARTUP/STANDBY READINESS REQUiREMENTS 12

5.0 STARTUP 13

5.1 Startup and Loading of RPS MG Sets 2A(2B) 13

5.2 Reset of Both RPS Trip Logic Channels with Mode Switch in

SHUTDOWN or in REFUEL 17

6.0 SYSTEM OPERATIONS 20

6.1 Reset of One RPS Trip Logic Channel 20

7.0 SHUTDOWN 22

7.1 De-energization of RPS Buses 22

7.2 Shutdown of RPS MG Sets !* * * * * * * * * * * * * * * 24

8.0 INFREQUENT OPERATIONS 25

8.1 RPS Bus A(B) Power Transfer from MG Set to Alternate 25

8.2 RPS Bus A(B) Power Transfer from Alternate to MG Set 32

8.3 Restoration to Normal Following RPS Bus Power Loss or Transfer 42

8.4 Restoring Power to RPS Bus A(B) Using Alternate Power Supply 48

8.5 Immediate Restoration of Power to RPS Bus A(B) Using Alternate

Power Supply 50

8.6 Preventing RWCU Isolations When Transferring RPS Power Supplies

In Cold Shutdown Condition 52

8.7 Preventing Shutdown Cooling Isolations When Transferring RPS 55

8.7.1 Preventing Shutdown Cooling Isolations Initial Lineup 55

BFN Reactor Protection System 2-01-99

Unit 2 Rev. 0073

Page 4 of 77

Table of Contents (continued)

8.7.2 Preventing Shutdown Cooling Isolations When Transferring

RPS A Power Supplies In Cold Shutdown Condition 57

8.7.3 Preventing Shutdown Cooling Isolations When Transferring

RPS B Power Supplies In Cold Shutdown Condition 59

Illustration 1: RPS Bus A or B Power Transfer 61

Illustration 2: Unit 2 Reactor Scram Initiation Signals 65

Illustration 3: Actions to Place RPS Instruments in Tripped Conditions

(TS Table 3.3.1.1-1) 67

ATTACHMENTS

Attachment 1: None

Attachment 2: None

Attachment 3: Reactor Protection System Electrical Lineup Checklist, Unit 2.

Attachment 4: None

Attachment 5: None

BFN Reactor Protection System 3-01-99

Unit 3 Rev. 0042

Page 42 of 71

8.6 Preventing RWCU Isolations When Transferring RPS Power

Supplies In Cold Shutdown Condition (continued)

[3.5] IF operation of the RWCU System isolation valves is

required while performing this section, THEN

REQUEST the operator at the breaker to PERFORM the

following:

[3.5.1] PLACE IN ON, 3-BKR-069-0002, RWCU SYSTEM

ISOL FCV-69-2 at 250V RMOV BO 3B, EI 593,

Compt 30. D

[3.5.2] PLACE IN ON, 3-BKR-069-0001, RWCU SYSTEM

ISOL FCV-69-1 at 480V RMOV BO 3A, EI 621,

Compt 16E. D

[3.5.3] PLACE IN ON, 3-BKR-069-0012, RWCU

ISOLATION VALVE FCV-69-12 at 480V RMOV

BO 3B, EI 593, Compt 17B. D

[4] TRANSFER RPS A power supply. REFER TO Section 8.1

or 8.2. D

[5] WHEN RPS power supplies are transferred and RPS

Restoration is completed, THEN

VERIFY PCIS RESET. D

[6] REQUEST the operator at the breaker to PERFORM the

following:

[6.1] PLACE IN ON, 3-BKR-069-0002, RWCG SYSTEM ISOL

FCV-69-2 at 250V RMOV BO 3B EI 593, Compt 3D. D

[6.2] PLACE IN ON, 3-BKR-069-0001, RWCU SYSTEM ISOL

FCV-69-1 at 480V RMOV BO 3A EI 621, Compt 16E. D

[6.3] PLACE IN ON, 3-BKR-069-0012, RWCU ISOLATION

VALVE FCV-69-12, at 480V RMOV BO 3B EI 593,

Compt 17B. D

BFN Reactor Protection System 3-01-99

Unit 3 Rev. 0042

Page 43 of 71

CAUTIONS

1) This section shall only be used when the Reactor is in the cold shutdown condition

(Mode 4 or Mode 5).

2) The amount of time in which Shutdown Cooling valves are prevented from operating

by the performance of this section should be minimized. This will minimize the

potential of preventing or delaying a real isolation requirement to operate the required

components. The safety features of these isolation functions are important to safe

operation of the plant, even with Primary Containment not required.

NOTES

1) The performance of this section will require one or two personnel qualified to operate

electrical breakers with radios in direct communication with the Unit Operator in the

Control Room.

2) Shift Manager/Unit Supervisor permission is required to perform this Section.

8.7 Preventing Shutdown Cooling Isolations When Transferring RPS

8.7.1 Preventing Shutdown Cooling Isolation Initial Lineup

[1] VERIFY the following:

Unit 3 is in Cold Shutdown Condition (Mode 4 or Mode 5). D

Reactor Mode Switch is in SHUTDOWN or REFUEL. D

Tech Spec Section 3.5.2 required actions are met if

appncable. D

Shutdown Cooling integrity is maintained.

(REFER TO Tech Spec Section 3.6.1.1 and

Table 3.3.6.1-1, function 6.b.) D

[2] The Operators are aware that during the performance of

Sections 8.7.2 (RPS 3A) or Sections 8.7.3 (RPS 38), that the

following should be performed if Containment System Isolation

is required.

CLOSE the associated Shutdown Cooling Isolation valve

breakers.

VERIFY the associated valves closed. D

BFN Reactor Protection System 3-01-99

Unit 3 Rev. 0042

Page 44 of 71

8.7.1 Preventing Shutdown Cooling Isolation Initial Lineup (continued)

[3] IF Shutdown Cooling Loop I is in service and Transferring

RPS A, THEN

PERFORM the following: (Otherwise N/A this Section)

[3.1] As Directed by the Unit Supervisor,

ALIGN RHR Loop II for Shutdown Cooling Loop.

(REFER TO 3-01-74) (N/A if RHR Loop II will not be

aligned.) D

[3.2] IF Shutdown Cooling Loop I will remain in service, THEN

PERFORM the following: (Otherwise N/A)

A. At 480 RMOV 3D - Compartment 2C

OPEN 3-BKR-074-0053, RHR SYS IINBD

INJECTION VLV FCV--74--53 (M010-25A). D

B. NOTIFY the Unit Supervisor breaker is open and

ENTER any applicable LCO's. D

[4] IF Shutdown Cooling Loop II is in service and Transferring

RPS B, THEN

PERFORM the following: (Otherwise N/A this Section)

[4.1] As Directed by the Unit Supervisor,

ALIGN RHR Loop I for Shutdown Cooling Loop ..

(REFER TO 3-01-74) (N/A if RHR Loop I will not be

aligned.) D

[4.2] IF Shutdown Cooling Loop II will remain in service,

THEN

PERFORM the following: (Otherwise N/A)

A. At 480V RMOV Bd 3E - Compartment 2C

OPEN 3-BKR-074-0067, RHR SYS IIINBD

INJECTION VLV FCV-74-67 (M010-25B). D

B. NOTIFY the Unit Supervisor breaker is open, and

ENTER any applicable LCO's. D

BFN Reactor Protection System 3-01-99

Unit 3 Rev. 0042

Page 45 of 71

8.7.2 Preventing Shutdown Cooling Isolations When Transferring

RPS A Power Supplies In Cold Shutdown Condition

[1 ] VERIFY Section 8.7.1 has been performed. D

[2] VERIFY Personnel are ready to transfer RPS A Power Supply. D

[3] ESTABLISH radio communication with the Control Room Unit

Operator. D

[4] At 480V RMOV BD 3A, EI 621, Compt 8C.

A. PLACE 3-BKR-074-0048 RHR SHUTDOWN COOLING

SUCT ISOL VLE FCV-74-48 (M010-18), in the OFF

position D

B. STATION an operator by the breaker for closure at the

request of the Control Room Unit Operator. D

C. IF operation of the RHR SHUTDOWN COOLING SUCT

INBD ISOL VLV is required while performing this section,

THEN

PLACE 3-BKR-074-0048 RHR SHUTDOWN COOLING

SUCT ISOL VLV FCV-74-48 (M010~18), in ON the

position. D

[5] TRANSFER RPS A power supply. (REFER TO Section 8.1

or 8.2.) D

[6] WHEN RPS A power supply has been transferred and RPS

Restoration has been completed, THEN

CONTINUE with this section of the procedure. D

BFN Reactor Protection System 3-01-99

Unit 3 Rev. 0042

Page 46 of 71

8.7.2 Preventing Shutdown Cooling Isolations When Transferring

RPS A Power Supplies In Cold Shutdown Condition (continued)

[7] On Panel 3-9-4

VERIFY PCIS will reset. D

[8] On Panel 3-9-3,

RESET RHR Loop I Logic as follows:

A. MOMENTARILY DEPRESS RHR SYS I SO CLG INBD

INJECT ISOL RESET, 3-XS-74-126. D

B. VERIFY 3-IL-74-126 extinguished. D

[9] On Panel 3-9-3,

RESET RHR Loop II Logic as follows:

MOMENTARILY DEPRESS RHR SYS II SO CLG INBD

INJECT ISOL RESET, 3-XS-74-132. D

VERIFY "3-IL-74-132 extinguished. D

[10] In Unit 3 Aux. Instrument Room,

VERIFY the following relays are energized:

Relay 16A-K29 on Panel 3-9-42 D

Relay 16A-K30 on Panel 3-9-43 D

[11] At 480V RMOV BD 3A, Compt 8C

PLACE 3-BKR-074-0048, RHR SHUTDOWN COOLING

SUCT ISOL VLV FCV-74-48 (M010-18) in the ON position. D

[12] At 480 RMOV 3D - Compartment 2C

PLACE 3-BKR-074-0053, RHR SYS I INBD INJECTION VLV

FCV-74-53 (M010-25A), in the ON position. (N/A if breaker

was not operated in step 8.7.1 [3.2]. D

[13] ALIGN Shutdown Cooling as required for current plant

conditions. (Refer To 3-01-74) D

BFN Reactor Protection System 3-01-99

Unit 3 Rev. 0042

Page 47 of 71

8.7.3 Preventing Shutdown Cooling Isolations When Transferring

RPS B Power Supplies In Cold Shutdown Condition

[1 ] VERIFY Section 8.7.1 has been performed. D

[2] VERIFY Personnel are ready to transfer RPS B Power Supply. D

[3] ESTABLISH radio communication with the Control Room Unit

Operator. D

[4] At 250V RMOV BD 3A, Compt R1A.

A. PLACE 3-BKR-074-0047 RHR SHUTDOWN COOLING

SUCT OUTBD ISOL VLV FCV-74-47 (M010-17), in the

OFF position. D

B. STATION an operator by the breaker for closure at the

request of the Control Room Unit Operator. D

C. IF operation of the Shutdown Cooling Loop II isolation

valves becomes necessary while performing this section,

THEN

PLACE 3-BKR-074-0047, RHR SHUTDOWN COOLING

SUCT OUTBD ISOL VLV FCV-74-47 (M010-17), in the

ON position. D

[5] TRANSFER RPS B power supply. (Refer To Section 8.1

or 8.2.) D

[6] WHEN RPS B power supply has been transferred and RPS

Restoration has been completed, THEN

CONTINUE with this section of the procedure. D

I

BFN 480V/240V AC Electrical System 0-01-578

Unit 0 Rev. 0171

Page 109 of 111

Illustration 3

(Page 1 of 3)

Board Restriction Verification Form

NOTES

1) This form is completed and verified to confirm compliance with Precaution and

Limitation 3.0N and 3.0Y, and associated board and board restrictions as given by

reference drawings.

2) Board restriction verification is performed twice per shift during normal plant conditions

while associated alignment is in place.

3) Board restriction verification is performed during normal plant conditions when manual

or automatic loads are added to boards affected by the associated alignment.

4) Independent verification of compliance with affected board restrictions is performed by

a qualified STA or SRO following first-party completion of this form.

5) This verification is to be completed for each individual board affected by the

manipulation/alignment.

1.0 BOARD RESTRICTION VERIFICATION

[1] RECORD board affected by the manipulation/alignment.

1st

IV

[2] RECORD board restriction from the associated drawing:

1st

IV

BFN 480V/240V AC Electrical System 0-01-57B

Unit 0 Rev. 0171

Page 110 of 111

Illustration 3

(Page 2 of 3)

Board Restriction Verification Form

1.0 BOARD RESTRICTION VERIFICATION (continued)

[3] RECORD actual board load after manipulation/alignment is

completed. (Use space below, as necessary, for calculations.

Actual Board kVA determination can be done by multiplying

board voltage, amps, square root of 3 and .001.

(kVA =.001 X 1.732 X (V) X (I))

The applicable Board Amp Limit can be determined by dividing

kVA limit above by board voltage(V) (.480kv or 4.16kv) times

the square root of 3

. 1 732) . e.g., AMp**

(I.e.,. Llmlt KVA

= -Limit

(V x .001) x 1.732

Some print notes require reducing Unit 2 auto starting loads

under accident conditions by some kVA value. To determine

appropriate load reduction, use 1 hp = 1 kVA. Affected loads,

which are to be prevented from starting, can be 4 kV load or a

480 V load which is powered from the 4 kV Shutdown Board.

kVA determination can be done by multiplying board

voltage, amps, square root of 3 and .001.

(kVA =.001 X 1.732 X (V) X (I).)

1st

IV

BFN 480V/240V AC Electrical System 0-01-57B

Unit 0 Rev. 0171

Page 111 of 111

Illustration 3

(Page 3 of 3)

Board Restriction Verification Form

1.0 BOARD RESTRICTION VERIFICATION (continued)

[4] VERIFY actual board load after manipulation/alignment will be

BELOW board restriction from the associated drawing.

1st

IV

[5] LOG results of this verification in Narrative Log.

1st

IV

HLTS-3-4

Revision 0

Page 1 of 21

SIMULATOR EVALUATION GUIDE

TITLE RWM FAILURE, RBCCW PUMP TRIP, FEED PUMP CONTROL

FAILURE, FUEL FAILURE, , RWCU LINE BREAK WITH FAILURE

TO ISOLATE, RAPIDLY DEPRESSURIZE WITH 2 AREAS

APPROACHING MAXIMUM SAFE RADIATION LEVELS

CONCURRED . /- - - - -

(Operations Superintendent or Designee) Date

VALIDATION: ~ _ 4/j yi / ,., /_~~~_

BY (Operations SRO: Req.ui.Oed~-for Exam Scenarios Only)

LOGGED-IN: ~ / _

(Librarian) Date

TASKS LIST

.UPDATED: / _

Date

HLTS-3-4

Revision 0

Page 2 of 21

NUCLEAR TRAINING

REVISION/USAGE LOG

REVISION DESCRIPTION DATE PAGES REVIEWED BY

NUMBER OF AFFECTED

REVISION

0 INITIAL 1/2/2008 All

HLTS-3-4

Revision 0

Page 3 of 21

I. Program: BFN Operator Training

II. Course: Examination Guide

III. Title: RWM FAILURE, RBCCW PUMP TRIP, FEEDPUMP CONTROL

FAILURE, FUEL FAILURE, RWCU LINE BREAK WITH FAI LURE

TO ISOLATE, RAPIDLY DEPRESSURIZE WITH 2 AREAS

APPROACHING MAXIMUM SAFE RADIATION LEVELS AND

EMERGENCY DEPRESSURIZE AFTER 2 AREAS REACH

MAXIMUM SAFE RADIATION LEVELS

IV. Length of Scenario: 1 to 1 % hours

V. Examination Objectives:

A. Terminal Objective

1. Perform routine shift turnover, plant assessment and routine shift

operation in accordance with BFN procedures.

2. Given uncertain or degrading conditions, the operating crew will

use team skills to conduct proper diagnostics and make

conservative operational decisions to remove equipment/unit from

operation. (SOER 94-1 and SOER 96-01)

3. Given abnormal conditions, the operating crew will place the unit in

a stabilized condition per normal, annunciator, abnormal, and

emergency procedures.

B. Enabling Objectives:

1. The operating crew will start and warm-up "B" RFP In accordance

with 01-6 section 5.7.

2. The operating crew will recognize and respond to a failure of RWM

in accordance with 3-01-85-5 and Tech. Specs.

3. The operating crew will recognize and respond to a RBCCW pump

trip in accordance with 3-AOI-70-1.

HLTS-3-4

Revision 0

Page 4 of 21

4. The operating crew will recognize and respond to Feedpump

control failure in accordance with 3-AOI-3-1.

5. The operating crew will recognize and respond to a fuel failure in

accordance with ARPs and EOI-3.

6. The operating crew will recognize and respond to a break in the

RWCU system and rapidly depressurize the RPV.

HLTS-3-4

Revision 0

Page 5 of 21

VI. References: The procedures used in the simulator are controlled copies and are

used in development and performance of simulator scenarios.

Scenarios are validated prior to use, and any procedure differences

will be corrected using the procedure revision level present in the

simulator. Any procedure differences noted during presentation will

be corrected in the same manner. As such, it is expected that the

references listed in this section need only contain the reference

material which is not available in the simulator.

VII. Training Materials: (If needed, otherwise disregard)

A. Calculator

B. Control Rod Insertion Sheet

C. Stopwatch

D. Hold Order / Caution tags

E. Annunciator window covers

F. Steam tables

HLTS-3-4

Revision 0

Page 6 of 21

VIII. Console Operator Instructions

A. Scenario File Summary

1. File: bat HLTS3-4

MF/RF/IOR# Description

a.) ior zdihs691 null Fails 69-1 to close

b.) imf cu04 25 RWCU suction line break

c.) imf cu06

d.) bat 7048FTC fail 70-48 to not auto close

2. File: bat HLTS3-4-1

MF/RF/IOR# Description

a.) imf rm1 Og 1000 5:00 Fails rm14 upscale

b.) imf rm1 Oe 1000 10:00 Fails rm09 upscale

3. File: bat7048ftc

MF/RF/IOR# Description

a.) ior zlohs704Ba[2] on Override red light on

b.) ior ypovfcv7048 fail_power_now Fails power to valve

c.) trg e1=bat 7048-1 Set trigger to 70-48 HS

d.) Imffw10a Fail 3A RFP auto trips

4. File: bat 7048-1

MF/RF/IOR# Description

a.) dor zlohs 7048a[2] Delete Override on red

light

b.) dor ypovfcv7048 Restore power to valve

5. File: bat HLTS3-4-4

MF/RF/IOR# Description

a.) imf fw30a (none 0) 60 30 50 Run up and stop 3A RFP

controller in manual

HLTS-3-4

Revision 0

Page 7 of 21

VIII. Console Operator Instructions

B. Console Operator Manipulations

UNSECURE file NRC - PW maryanne

ELAP TIME PFK DESCRIPTION/ACTION

Sim. Setup Pwrst 120 -2% power, MOC

(cst)

Sim. Setup restorepref Establishes Function Keys

HLTS3-4

Sim. Setup setup Verify Function Keys

Sim. Setup esc Clears Function Key Popup

Sim. Setup I Manual Place Hold Order Tags on C RFP

suction and discharge valves

Sim. Setup manual Ensure RWM is latched with no Insert or

Withdrawal blocks and comp/prog lights

reset, rod group 39 - 06-47 selected

Sim. Setup Manual Verify 3C RFP suct & disch valve lights

extinguished. If not, bat 2crfptag

Sim Setup <Shift F1 > Set 70-48 to not close on low pressure

bat 7048ftc and fail 3A RFP auto trips

After RFP warmed and F3 Fails RWM ( imf rd14a )

When requested by

Examiner

ROLE PLAY: If asked, have not performed a startup with RWM

bypassed within last calendar quarter

ROLE PLAY: If requested to verify open 3-1-155 and 3-1-156, report that they

are open

WATCH CAREFULLY TO SEE IF NEED TO MANUALLY FIRE TRIGGER

After Tech Specs F8 Trips A RBCCW pump

addressed for RWM (imf sw02a)

If trigger does not fire <Shift F2> Manually fire trigger

when 70-48 taken to

close

HLTS-3-4

Revision 0

Page 8 of 21

VIII. Console Operator Instructions (continued)

B. Console Operator Manipulations

If requested to align spare

RBCCW pump to Unit 3 F9 Aligns spare RBCCW pump to Unit 3

Wait 3 minutes (mrf sw02 align)

If requested to reset local F12 mrf an01 e reset

RWCU panel alarms

After spare RBCCW pump <Shift F3> Fails RFP governor in raise direction in

aligned and RWCU Bat HLT83-4-4 manual for 30 sec

returned to service

Two (2) Minutes after F6 fuel failure

Feedpump governor (imf th23 5 15:00)

problem

When directed by examiner F7 RWCU line break with failure to isolate

(bat HLTS3-4)

ROLE PLAY: If requested to attempt to close 69-1 locally at the breaker, wait

5 minutes and report it will not close

ROLE PLAY: If requested to check Aux Inst rm, report 835 A&C and 835

B&D reading 90 deg F and fairly steady

After attempts to close 69-1 F10 Causes Rad monitors to reach max

are made (bat HLTS3-4-1)

Terminate the scenario when the following conditions are satisfied or when requested

by Chief Examiner:

1. Reactor Water level restored between +2 to +51"

2. RPV rapidly depressurized

3. RPV emergency depressurized

SECURE file NRC - PW maryanne

HLTS-3-4

Revision 0

Page 9 of 21

IX. Scenario Summary

The plant is at approximately 2% power withdrawing control rods to open

sufficient bypass valves to roll the main turbine. "B" RFP needs to be started

and warmed in preparation for water level control.

During the control rod withdrawal, the RWM will experience a program fault

which will block rod movement. Tech. Specs will be addressed and control rod

withdrawal will contin-ue when a second licensed operator is present to ensure

withdrawal is in accordance with the BPWS.

An RBCCW pump will trip causing RWCU to be secured and the spare RBCCW

pump aligned to Unit 3 and the RWCU system returned to service.

The In-service RFP will experience a governor fault causing it to inject cold water

into the RPV causing a power spike and some fuel failure. Later the RWCU

system develops and leak and fails to isolate requiring entry into EOI-3 and

subsequent rapid depressurization due to 2 areas approaching max safe

radiation levels.

HLT8-3-4

Revision 0

Page 10 of 21

x. Information to Evaluators:

A. Ensure recorders are inking and recording and ICS is active and updating.

B. Assign Crew Positions based on the required rotation.

1. SRO

2. ATC

3. BOP

c. Conduct a shift turnover with the Shift Manager and provide the Shift

Manager with a copy of the Shift Turnover.

D. Direct the shift crew to review the control board and take note of present

conditions, alarms, etc.

E. Terminate the scenario when the following conditions are satisfied are at

the request of the floor/lead instructor/evaluator.

1. Reactor water level restored at +2" to +51"

2. RPV rapidly depressurized

3. RPV emergency depressurized when 2 areas are above max safe

values.

HLTS-3-4

Revision 0

Page 11 of 21

XI. Simulator Event Guide

EVENT 1: Warming up second RFP

POSITI,ON TIME EXPECTED ACTIONS

SRO Directs warming up B RFP in accordance with 3-01-3

BOP Warms up "B" RFP utilizing section 5.6 of 3-01-3.

Place in auto and verify open RFP min flow valve 3-FCV-3-13

Place 28 start/local enable 3-HS-46-138A in start and

observe RFP accelerates to 600 rpm

Verify no abnormal rubbing or vibration is observed

Raise speed to -1000 rpm using 3-HS-46-9A

Place TG motor 3-HS-3-127A in Auto

Depress 3B trip 3-HS-3-127A and verify HP and LP stop

valves close

Verify TG auto engages or RFP rolling on min flow

Depress 3B trip reset 3-HS-3-150A and verify blue light

extinguishes and HP and LP stop valves open

Place 3B start/local enable 3-HS-46-138A in start and

observe RFPT speed increases to - 600 rpm

HLTS-3-4

Revision 0

Page 12 of 21

XI: Simulator Event Guide

EVENT 2: RWM FAILURE

POSITION TIME EXPECTED ACTIONS

ATC Announces "RWM ROD BLOCK" 3-XA-55-5B window 35

alarm and refers to ARP.

Verifies Control Rod positions

SRO Directs ATC to bypass RWM per 01-85

Refers to T.S. 3.1, 3.3, table 3.3.2.1-1

Contacts Rx Engineer

ATC Refers to section 8.17 of 01-85 and places 3-XS-85-9025 in

Bypass.

ATC Checks manual bypass light lit and all others out._

SRO Determines T.S> 3.3.2.1 condition C applies. Greater than

nd

12 rods withdrawn and 2 person to verify compliance with

BPWS.

HLTS-3-4

Revision 0

Page 13 of 21

XI. Simulator Event Guide

EVENT 3: LOSS OF 3A RBCCW pump

POSITION TIME EXPECTED ACTIONS

BOP Responds to loss of RBCCW pump 3A trip and attempts to

restart 3A RBCCW pump and reports it failed to start.

SRO Directs securing RWCU pumps per 3-AOI-70-1

BOP Secures RWCU pumps and reports that the 3-FCV-70-48

sectionalizing valve failed to close.

US Directs closure of the 3-FCV-70-48

Directs placing Spare RBCCW pump in service.

Dispatches personnel to investigate pump loss

May contact Rx Engineer about heat balance

BOP Closes 3-FCV-70-48

After Spare RBCCW pump placed in service, re-opens 70-

48 and returns RWCU to service per 01-69. (conditional,

SRO may not direct valve to be opened after failure to auto

close.)

BOP Opens 69-8

Starts A(B) RWCU Pump

Coordinates with AUO to roll demins in service

Starts second RWCU Pump

HLTS-3-4

Revision 0

Page 14 of 21

XI. Simulator Event Guide

EVENT 4: FEEDWATER CONTROLLER FAILURE

POSITION TIME EXPECTED ACTIONS

Observes period rise by meter or annunciator and checks

for cause of reactivity addition

ATC Ranges IRMs as necessary to prevent a reactor scram

BOP Attempts to take control of A RFP by adjusting 3-HS-46-8A

and reports that "A" RFP cannot be controlled

SRO Directs tripping "A" RFP and using "8" RFP for RPV level

control

BOP Trips "A" RFP by depressing 3-HS3-125A and raises "B~'

RFP speed by using 3-HS-46-9A

BOP Opens "B" RFP discharge valve 3-HS-3-12A when "B" RFP

discharge pressure is within 250 Ibs of reactor pressure.

HLTS-3-4

Revision 0

Page 15 of 21

XI. Simulator Event Guide

EVENT 5: FUEL FAILURE DUE TO COLD WATER INJECTION

POSITION TIME EXPECTED ACTIONS

BOP Announces "TURBINE BUILDING HIGH RADIATION" and

determines which area and evacuates that area.

Announces "OFF-GAS ANNUAL RELEASE LIMIT

EXCEEDED" and responds per ARP

BOP/SRO Notifies Chemistry to perform analysis and Radcon

SRO

Declares a NOUE on a valid OG pretreatment rad alarm or

Main Steam Line rad Hi Hi.

HLTS-3-4

Revision 0

Page 16 of 21

XI. Simulator Event Guide

EVENT 6: RWCU LINE SUCTION BREAK

POSITION TIME EXPECTED ACTIONS

BOP Announces "RX BLDG HIGH RADIATION" and determines

which area and evacuates that area.

North and South RWCU area.

BOP Reports on RWCU leak detection alarms

SRO Enters EOI-3 on either high temp or high radiation

ATC Recognizes that 69-1 failed to isolate and attempts to

manually close

Crew Directs outside personnel to attempt to close 69-1 locally at

the breaker. .

SRO Directs Rx Scram before any area temp is above the

maximum safe operating temperature.

ATC Scrams reactor and provides scram report

SRO Directs ATC to perform actions of 3-AOI-1 00-1

HLTS-3-4

Revision 0

Page 17 of 21

XI. Simulator Event Guide

EVENT 6: RWCU LINE SUCTION BREAK

POSITION TIME EXPECTED ACTIONS

SRO/BOP Continue to monitor and trend secondary area temps and

radiation levels

BOP Reports that 2 areas are approaching maximum safe

radiation levels

SRO Directs rapid depressurization ot the RPV using BPVs

BOP Opens all BPVs using the Jack

BOP/ATC Coordinate level control during depressurization to prevent

flooding the RPV

BOP Determines that 2 areas are above max safe

radiation values

HLTS-3-4

Revision 0

Page 18of21

XI. Simulator Event Guide

EVENT 6: RWCU LINE SUCTION BREAK

POSITION TIME EXPECTED ACTIONS

SRO Determines that Emergency Depressurization is

Required and enters C2

Directs BOP to open all ADS valves

BOP Opens all ADS valves

SRO When the shutdown cooling pressure interlock

clears, directs BOP to place shutdown cooling in

service per Appx. 17D

BOP Places Shutdown Cooling in service per Appendix

17D

HLTS-3-4

Revision 0

Page 19 of 21

XII. Crew Critical Tasks

TASKS SAT/UNSAT

1. Trips "A" RFP prior to reaching Main

Steam Lines

2. Emergency Depressurize when 2 areas

reach maximum safe values.

HLTS-3-4

Revision 0

Page 20 of 21

XIII. Scenario Verification Data

EVENT TASK NUMBER KIA RO SRO CONTROL

MANIPULATION

1. Warm up RFP 259001 3.9 3.7

A4.02

2. RWM Failure 201006 3.2 3.4

A4.01

3. RBCCW Pump 295018

Trip AK3.03 3.1 3.4

AA1.01 3.3 3.4

AK3.04 3.3 3.3

4. RFP Governor 259001 A2.07 3.7 3.8

failure 295008 AA 1.08 3.5 3.5

5. Fuel Failure 295014 AA 1.05 3.9 3.9

AA1.07 4.0 4.1

6. RWCU Line Break 295033 EA1.05 3.9 4.0

EK3.01 3.3 3.5

HLTS-3-4

Revision 0

Page 21 of 21

SCENARIO REVIEW CHECKLIST

SCENARIO NUMBER HLTS-11

7 Total Malfunctions Inserted; List: (5-8)

1). RWM Failure

2) RFP controller failure

3) Fuel Damage

4) RWCU line break

5) "A" RBCCW pump trip

6) Failure of 69-1 to close

7) Failure of RFPs to trip on Hi level

1 Malfunctions That Occur After EOI Entry; List: (1-2)

1) RWCU line Break

2 Abnormal Events; List: (2-4)

1) RFP control failure

2) RBCCW pump trip

1 Major Transients; List: (1-2)

1) RWCU line break (small LOCA)

2 EOls used; List: (1-2)

1) EOI-1

2) EOI-3

1 EOI Contingencies Used; List: (0-2)

1) C2

63 Run Time (minutes)

52 EOI Run Time (minutes); 83 % of Scenario EOI Run Time

2 Crew Critical Tasks (2-3)

Yes Technical Specifications Exercised (yes/no)

Revision 0

Page 21 of 21

SHIFT TURNOVER SHEET

Equipment Out of Service/LCOs C RFP is u,ncoupled and awaiting overspeed testing.

Suction and Discharge valves are tagged.

Operations/Maintenance For the Shift: Continue with reactor startup at step 5.66.8 of

3-801-100-1A. Continue with warm-up of "8" RFP per 01-3 at step 5.6.2.2.16. Thrust

bearing! Overspeed/ Stop Valve and Control Valve tests are complete for "8" RFP.

Unusual Conditions/Problem Areas: Power System Alert in effect for the next 36

Hours.

Browns Ferry Nuclear Plant

Unit 3

General Operating Instruction

3-GOI-100-1A

Unit Startup

Revision 0074

Quality Related

Level of Use: Continuous Use

Effective Date: 12-05-2007

Responsible Organization: OPS, Operations

Prepared By: C. E. Heitzenrater

Approved By: John T. Kulisek

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 2 of 167

Current Revision Description

Type of Change: Enhancement Tracking Number: 081

Pages 81

PERs None

PCRs 07004789

Page 81 Step 5.0[30.2] - Modify the 6th bullet on Reactor Water temp to allow

obtaining data from either Recirc Discharge OR Suction temperature and the ability to

use ICS for either reading. This closes PCR 07004789.

BFN Unit Startup 3-GOI-100-1A

Unit 3 Rev. 0074

Page 3 of 167

Table of Contents

1.0 PURPOSE 5

2.0 REFERENCES 5

2.1 Technical Specifications 5

2.2 Technical Requirements Manual 6

2.3 Final Safety Analysis Report 7

2.4 Plant Instructions 7

2.4.1 Operating Instructions 7

2.4.2 Surveillance Instructions 11

2.4.3 Other Instructions 13

2.4.4 Administration Procedures 13

2.5 Miscellaneous Documents 13

3.0 PRECAUTIONS AND LIMITATIONS 16

3.1 General _ _ _..__ _ 16

3.2 Coolant and Metal Temperatures 17

3.3 Primary Containment 18

3.4 Control Rods, Reactivity Control and Relative Instrumentation 18

3.5 Thermal Limits 20

3.6 EHC and Main Turbine 20

3.7 Electrical Alignments and Load Considerations 21

3.8 Condensate and Feedwater 23

3.9 Radiation Protection Notifications and Radiological Protection Hold

Points (RPHPs) [SOER 01-1, BFN PER 126211, PER 961778, PER

116666] 24

4.0 PREREQUiSiTES 26

4.1 Prestartup Checklist 26

5.0 INSTRUCTION STEPS 64

Illustration 1: Reactor Thermal Limits 135

lllustratlon 2: Core Quadrants/Octants 136

Illustration 3: Reactor Vessel Heatup Graph 137

Illustration 4: Percent Power vs. Time (To obtain 4000 MWt-minutes) 138

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 4 of 167

Table of Contents (continued)

Attachment 1: Environmentally Qualified Barrier Doors 140

Attachment 2: Temperature Verifications from Cold Shutdown to 212°F 141

Attachment 3: Startup with MSIV's Closed 145

Appendix A: Radiation Protection Notifications 166

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 5 of 167

1.0 PURPOSE

This instruction provides precautions and limitations, prerequisites, and procedural

steps to take the unit from either MODE 4 or MODE 3 to full power operation. It also

provides an integrated plant approach to raising power back to full load after a power

reduction. Provided the Reactor remains in the MODE 1, the Shift .Manager will

enter this procedure at the appropriate step in Section 5.0 depending on the present

power level.

2.0 REFERENCES

2.1 Technical Specifications

Section 3.1.3, Control Rod Operability.

Section 3.1.7, Standby Liquid Control (SLC) System.

Section 3.1.8, Scram Discharge Volume (SDV) Vent and Drain Valves.

Section 3.2, Power Distribution Limits.

Section 3.2.1 , Average Planar Linear Heat Generation Rate (APLHGR).

Section 3.2.2, Minimum Critical Power Ratio (MCPR).

Section 3.2.3, Linear Heat Generation Rate (LHGR).

Section 3.3.1.1, Reactor Protection System (RPS) Instrumentation.

Section 3.3.2.1, Control Rod Block Instrumentation.

Section 3.3.3.1, Post Accident Monitoring (PAM) Instrumentation.

Section 3.4.1, Recirculation Loops Operating.

Section 3.4.2, Jet Pumps.

Section 3.4.3, Safety/Relief Valves (S/RVs).

Section 3.4.6, RCS Specific Activity.

Section 3.4.9, RCS Pressure and Temperature (PIT) Limits.

Section 3.5.1, ECCS Operating.

Section 3.5.3, RCIC System.

Section 3.6.1.1, Primary Containment.

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 6 of 167

2.1 Technical Specifications (continued)

Section 3.6.4.1, Secondary Containment.

Section 3.6.4.3, Standby gas Treatment (SGT) System.

Section 3.7.1, Residual Heat Removal Service Water (RHRSW) System.

Section 3.7.2, Emergency Equipment Cooling Water (EECW) System and Ultimate

Heat Sink (UHS).

Section 3.8, Electrical Power Systems.

2.2 Technical Requirements Manual

Section 3.1, Reactivity Control.

Section 3.3.1, Reactor Protection System (RPS) Instrumentation.

Section 3.3.3, ECCS Instrumentation.

Section 3.3.5, Surveillance Instrumentation.

Section 3.4.1, Coolant Chemistry.

Section 3.5, Emergency Core Cooling Systems.

Section 3.6, Containment Systems.

Section 3~6.1, Primary Containment Purge System.

Section 3.8, Auxiliary Electrical System.

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 7 of 167

2.3 Final Safety Analysis Report

Chapter 3.0, Reactor.

Chapter 4.0, Reactor Coolant System.

Chapter 5.0, Containment.

Chapter 6.0, Core Standby Cooling Systems.

Chapter 7.0, Control And Instrumentation.

Chapter 8.0, Electrical Power Systems.

Chapter 10.0, Auxiliary Systems.

Chapter 11.0, Power Conversion Systems.

Chapter 13.0, Conduct of Operations.

2.4 Plant Instructions

2.4.1 Operating Instructions.

3-AOI-100-1, Reactor Scram.

3-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and

Reductions in Power During Power Operations.

0-GOI-300-4, Switchyard Manual.

3-01-1, Main Steam System.

3-01-2, Condensate System.

3-01-2A, Condensate Demineralizers System.

0-01-2B, Condensate Storage and Transfer System.

0-01-2C, Demineralized Water System.

3-01-3, Reactor Feedwater System.

3-01-4, Hydrogen Water Chemistry System.

3-01-6, Feedwater Heating and Misc Drains System.

0-01-12, Auxiliary Boilers.

0-01-18, Fuel Oil System.

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 8 of 167

2.4.1 Operating Instructions. (continued)

0-01-20, Central Lubricating Oil System.

0-01-23, Residual Heat Removal Service Water System.

3-01-24, Raw Cooling Water System.

0-01-25, Raw Service Water System.

0-01-26, High Pressure Fire Protection System.

3-01-27, Condenser Circulating Water System.

3-01-27A, Screen Wash System.

3-01-27B, Amertap Condenser Tube Cleaning System.

0-01-27C, Cooling Tower System.

0-01-29, Potable Water System.

3-01-30A, Refueling Zone Ventilation System.

3-01-30B, Reactor Zone Ventilation System.

3-01-30C, Turbine Building Ventilation System.

0-01-300, Radwaste Building Ventilation System.

0-01-30E, Service and Office Building Ventilation System.

0-01-30F, Common and DG Building Ventilation.

0-01-31, Control Bay and Off-Gas Treatment Building Air Conditioning System.

0-01-32, Control Air System.

3-01-32A, Orywell Control Air System.

0-01-33, Service Air System.

0-01-34, Vacuum Priming System.

3-01-35, Generator Hydrogen Cooling System.

3-01-35A, Stator Cooling System.

3-01-35B, Generator Hydrogen Seal Oil System.

3-01-35C, Generator Circuit Breakers.

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 9 of 167

2.4.1 Operating Instructions. (continued)

0-01-39, CO 2 System.

0-01-40, Station Drainage System.

0-01-44, Building Heating System.

3-01-47, Turbine-Generator System.

3-01-47A, EHC System.

3-01-47B, Main Turbine Lube Oil System.

3-91-47C, Seal Steam System.

0-01-48, Integrated Computer System.

0-01-53, Demineralizer Backwash Air System.

0-01-57A, Switchyard and 4160V Electrical System.

0-01-57B, 480V/240V AC Electrical System.

0-01-57C, 208V/120V AC Electrical System.

0-01-57D, DC Electrical System.

3-01-63, Standby Liquid Control System.

3-01-64, Primary Containment System.

0-01-65, Standby Gas Treatment System.

3-01-66, Off-Gas System.

0-01-67, Emergency Equipment Cooling Water System.

3-01-68, Reactor Recirculation System.

3-01-69, Reactor Water Cleanup.

3-01-70, Reactor Building Closed Cooling Water System.

3-01-71, Reactor Core Isolation Cooling System.

3-01-73, High Pressure Coolant Injection System.

3-01-74, Residual Heat Removal System.

2-01-74, Residual Heat Removal System.

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 10 of 167

2.4.1 Operating Instructions. (continued)

3-01-75, Core Spray System.

3-01-76, Containment Inerting System.

0-01-77A, Waste Collector/Surge System Processing.

0-01-77B, Floor Drain Collector System Processing.

0-01-77C, Radwaste Filter and Demineralizer System.

0-01-770, Backwash Receivers and Phase Separators System.

3-01-78, Fuel Pool Cooling and Cleanup System.

0-01-82, Standby Diesel Generator System.

3-01-82, Standby Diesel Generator System.

3-01-84, Containment Atmosphere Dilution System.

3-01-85, Control Rod Drive System.

3-01-90, Radiation Monitoring System.

3-01-92, Source Range Monitors.

3-01-92A, Intermediate Range Monitors.

3-01-92B, Average Power Range Monitoring.

3-01-92C, Rod Block Monitor.

3-01-94, Traversing Incore Probe System.

3-01-99, Reactor Protection System.

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 11 of 167

2.4.2 Surveillance Instructions

3-SR-3.1.1.1, Reactivity Margin Test.

3-SR-3.1.2.1, Reactivity Anomaly and Exposure.

3-SR-3.1.3.5(A), Control Rod Coupling Integrity Check.

3-SR-3.1.3.5(B), CRD Coupling Integrity Check After Refueling or Maintenance.

3-SR-3.1.4.1, Scram Insertion Times.

3-SR-3.1.6.1, BPWS Compliance Verification.

3-SR-3.1.8.1, Scram Discharge Volume Valve Open.

3-SR-3.1.8.2, Scram Discharge Volume Valve Operability.

3-SR-3.3.1.1.2, APRM Output Signal Adjustment.

3-SR-3.3.1.1.3(IRMs), Intermediate Range Monitor Functional Test With Reactor

Mode Switch NOT In Run Position.

3-SR-3.3.1.1.5, Source Range Monitors (SRMs) and Intermediate Range Monitors

(IRMs) Overlap Verification.

3-SR-3.3.1.1.6(IRMs), IRM Gain Adjustment and IRM/APRM Overlap Verification.

3-SR-3.3.1.1.9(IRM A-H), Intermediate Range Monitor (IRM) Channel A-H

Calibration.

3-SR-3.3.1.1.13(APRM-1-4), Average Power Range Monitor Calibration-APRM-1-4.

3-SR-3.3.1.1.14(2e), Average Power Range Monitor (APRM) 2-0UT-OF-4 Voter

Logic Functional Test.

3-SR-3.3.1.1.16(APRM-1-4), Average Power Range Monitor Functional

Test-APRM-1-4.

3-SR-3.3.1.2.5&6, Instrumentation That Initiates Rod Block/Scrams Source Range

Monitor (SRM) Functional Test With Reactor Mode Switch NOT in RUN Position.

3-SR-3.3.1.2.7(SRM A-D), Source Range Monitor (SRM) Calibration and Functional

Test.

3-SR-3.3.2.1.1, Rod Block Monitor(RBM) Functional Test.

3-SR-3.3.2.1.2, RWM Functional Test For Startup.

3-SR-3.3.2.1.4(A), Rod Block Monitor (RBM) Calibration and Functional Test.

BFN Unit Startup 3-GOI-100-1A

Unit 3 Rev. 0074

Page 12 of 167

2.4.2 Surveillance Instructions (continued)

3-SR-3.3.2.1.4(B), Rod Block Monitor (RBM) Calibration and Functional Test.

3-SR-3.3.2.1.5, Verification Of RWM Automatic Bypass Setpoint.

3-SR-3.3.2.1.7, RWM Program Verification.

3-SR-3.3.5.1.6(ADS A), ADS Logic System Functional Test - Bus A, Time Delay

Relay Calibration, and Bus Power Monitor Test.

3-SR-3.3.5.1.6(ADS B), ADS Logic System Functional Test - Bus B, Time Delay

Relay Calibration, and Bus Power Monitor Test.

3-SR-3.4.1 (SLO), Reactor Recirculation System Single Loop Operation

3-SR-3.4.1 (DLO), Reactor Recirculation System Dual Loop Operation

3-SR-3.4.3.2, Main Steam Relief Valves Manual Cycle Test.

3-SR-3.4.9.1 (1), Reactor Heatup or Cooldown Rate Monitoring.

3-SR-3.4.9.5-7, RPV Temperature Monitoring with Head Tensioned.

3-SR-3.5.1.5, Reactor Recirculation Pump Discharge Valves Cycling.

3-SR-3.5.1.8, HPCI Main and Booster Pump Set Developed Head and Flow Rate

Test at 150 psig Reactor Pressure.

3-SR-3.5.1.10, Automatic Depressurization System Simulated Automatic Actuation

Test.

3-SR-3.5.3.4, RCIC System Rated Flow at 150 psig.

3-SR-3.6.1.2.1, Primary Containment Airlock Local Leak Rate Test.

3-SR-3.6.1.3.3, Primary Containment Isolation Manual Valves and Blind Flanges

Inside Containment Position Verification.

3-SR-3.6.1.3.5(SD), Valves Cycled During Cold Shutdown.

3-SR-3.3.1.1.1, Core Thermal Hydraulic Stability.

3-SI-3.3.1.A, ASME Section XI System Leakage Test of the Reactor Pressure

Vessel and Associated Piping (ASME Section III, Class 1).

3-SI-4.7.A.5.c, Control Air/Drywell Control Air Isolation Verification.

3-SI-4.6.B.1-4, Reactor Coolant Startup Chemistry.

BFN Unit Startup 3-GOI-100-1A

Unit 3 Rev. 0074

Page 13 of 167

2.4.3 Other Instructions

0-TI-248, Station Reactor Engineer.

MCI-0-064-HLT001, Drywell Personnel Airlock Doors.

MSI-0-001-VSL001, Reactor Vessel and Cavity Disassembly and Reassembly.

MSI-0-064-HLT002, Opening and Closing of Primary Containment Hatches.

RCI-17, Surveillance and Door Control of Prohibitive High Radiation Areas.

2.4.4 Administration Procedures

SPP-5.3, Chemistry Control.

SPP-10.3, Verification Program.

SPP-10.4, Reactivity Management Program

OPDP-1, Conduct of Operations

SPP-7.2, Outage Management

2.5 Miscellaneous Documents

BP-250, Corrective Action Program Handbook.

BWROG-94078, BWR Owner's Group Guidelines for Stability Interim Corrective

Action.

GE SIL 316, Reduced Notch Worth Procedure.

GE SIL 380, BWR Core Thermal Hydraulic Stability.

GE SIL 498, SB-1 and SB-9 Switch Lockup.

Incident Investigation II-B-91-129, Unit 2 manual scram due to high torus

temperature - caused by temperature stratification from extended RCIC run exhaust

steam discharge.

INPO SER 89-006, Withdrawal of Safety Rod Group Out of Sequence.

INPO SER 89-022, Intermittent Failure of Westinghouse Type DS and DSL Breakers

to Close.

INPO SOER 88-002, Premature Criticality Events During Reactor Startup.

INPO SOER 90-003, Nuclear Instrumentation Miscalibration.

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 14 of 167

2.5 Miscellaneous Documents (continued)

INPO SER 92-19, Power Oscillations at Boiling Water Reactors.

SOER 01-1, Unplanned Radiation Exposures

Licensee Event Report 260/94009, Missed Technical Specification (Tech Specs)

Surveillance before Reactor Startup as a Result of a Misunderstanding of

Tech Specs.

NRC Generic Letter 94-02, Long-Term Solutions and Upgrade of Interim Operating

Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors.

NRC IE Bulletin 79-12, Short Period Scrams at BWR Facilities.

NRC IE Bulletin 88-07, Supplement 1, Power Oscillations in Boiling Water Reactors.

NRC Inspection Report 84-45 response, RIMS L44850329806, Identify required

steps in startup procedures.

NRC Inspection Report 85-15, item 4.a, NOT Curves, Technical Specification

Figure 3.6-1, Out of Date.

NRC IE Notice 89-030, High Temperature Environments at Nuclear Power Plants.

NRC Information Notice 92-740, Power Oscillations at Washington Nuclear Power

Unit 2.

NSRB Item A258-4, Review procedures to preclude an event similar to SER 24-91,

inadequate control of reactivity changes during plant shutdown results in unwanted

transient.

Q13958, issued 09/05/90.

Q16997B - Doors, Hatches, and Penetrations Required to be Closed to Maintain EQ

Boundaries.

S17557B, Combined Zone Secondary Containment.

Scram Frequency Reduction Committee (SFRC) Recommendations 17, G-20-1 and

G-20-2 concerning additional SRO assisting during startup or shutdown.

Technical Specifications Assessment Report, Item 089, Clearly specify the

expectations for satisfying Tech Specs LCOs prior to changing operation conditions.

TVA-BFE-052, Extended Load Line Limit Analysis.

Jerry Robertson Memorandum to G.C. Campbell, Use of Increased Core Flow (L32

890302 901).

BFN Unit Startup 3-GOI-100-1A

Unit 3 Rev. 0074

Page 15 of 167

2.5 Miscellaneous Documents (continued)

W.N. Hannum (NSRB) Memorandum to R.R. Calabro and G.G. Campbell (L42

890927 800).

Letter from O. D. Kingsley to W. J. Museler, DOWNPOWERING OF NUCLEAR

UNITS UNDER LOW SYSTEM LOAD CONDITIONS, March 1, 1996 (AOO 9602,26

150)

T.A. Keys Memorandum to K.L. Welch, Use of Increased Core Flow (ICF) at Browns

Ferry Nuclear Plant (L32 920709 801).

TOE 0-97-064-0823 Parallel purging of the Torus and Drywell.

Drawing 0-48N954, R003 - Miscellaneous Steel Refueling Facilities General Plan &

Elevation.

TVA-BFN-TS-384, Technical Specification (Tech Specs) Change TS-384 Request

for License Amendment for Power Uprate Operation and NEDC-32751 P, Power

Uprate Safety Analysis for the Browns Ferry Nuclear Plant Units 2 and 3

(R08-980316-888) .

GE-NE-B13-01866-39, Task Report 39 Summary of System Evaluations and

Proposed Changes to Design Criteria Documents (W79-980427-005).

BFN-IPIP-TASK35, Computer Process Alarm Limits (W79-980319-002).

GE-NE-B13-01866-2, Task Report 2 Power Uprate Evaluation Report for

Power/Flow Operating Map (RIMS W79-971 023-002).

ED-N0003-980030, BFN Setpoint and Scaling Calculation (R14-980422-104)

ND-Q0068-980011, Power/Flow Map (RIMS R14-980423-102).

GE-NE-B13-01866-05, Power Uprate Evaluation Task Report for Browns Ferry

Units 1, 2 and 3 Transient Analysis (RIMS W79-971 004-005).

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 16 of 167

3.0 PRECAUTIONS AND LIMITATIONS

3.1 General

A. The Critical Steps warning represents a step, or series of steps, for an activity

that requires additional focus, attention, and increased awareness. The

Operator performing these steps for the activity needs to verify the Unit

Supervisor and other Control Room staff are aware of the evolution. PEER

checks are required for this activity and short briefs need to be made prior to

performing the evolution. Included in the briefs are worst case scenario and

contingencies.

B. [SFRC/C] Scram Reduction Recommendations, G-20-1 and G-20-2, require an

additional SRO licensed operator to assist with BOP operations when

performing power maneuvers during unit startup or shut down with Reactor

power less than 60%. This individual is not limited to the Control Room.

[SFRC-17, G-20-1 & -2]

C. Unit Supervisor's permission is required to reject water to main condenser from

the Reactor Water Cleanup (RWCU) System without a RWCU filter in service.

D. [TSAR/C] Coolant Leakage Detection Systems is required to be in service prior to

reaching 212°F. [Item 045]

E. The bottom layer of Reactor Well shield blocks are required to be in place prior

to exiting Mode 4 (cold shutdown). The top layers of Reactor Well Shield

Blocks are required to be installed BEFORE exceeding 10 days of power

operation. This requirement is established based on evaluation of Source Term

effects to Operations Personnel. [BFN PER 02-005145-000]

F. [TSAR/C] Any time a unit, system, or plant mode or operational condition change is

required, the Unit Supervisor, Unit Operators, Shift Manager, and STA if

manned, are required to review all applicable LCOs prior to the mode change

(and as soon as practical during an emergency) to ensure compliance with

Tech Specs. [Item 089]

G. To ensure that all 3-SR-2 instrumentation meets the Instrument Checks for the

required modes. The 3-SR-2 readings will be taken prior to the Mode or

Condition changes. The STA will verify that the readings will allow the Reactor

Mode or Condition Changes.

H. The maximum rate of temperature change (rise or fall) in Wheeler Reservoir is

limited to 10°F per hour as measured at the downstream temperature control

point.

I. Reactor Core Isolation Cooling or High Pressure Coolant Injection Systems are

not normally used for level control during Reactor startup.

BFN Unit Startup 3-GOI-1 00-1A

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Page 17 of 167

3.1 General (continued)

J. Chemistry parameters are specified and corrective actions for any out-of-limit

chemistry parameter are delineated in SPP-5.3. Special attention is required

for chemistry parameters in Tech Specs 3.4.6 and Technical Requirements

Manual 3.4.1, Coolant Chemistry.

K. Noble Metal Injection will cause a higher radiation level than Normal throughout

the plant. Following a startup from Noble Metals injection this condition will

diminish over time up to 6 weeks.

Therefore, if this startup is being performed following Noble Metal Injection or

during the time period where radiation levels are still higher than normal from a

Noble Metal Injection Shutdown, then, to minimize radiation levels, the

Hydrogen Water Chemistry System should not be aligned during the startup.

The Duty Engineer will make recommendations in determining if the Hydrogen

Water Chemistry System should be placed in service.

L. All MSIVs should be OPEN prior to 25°A, Reactor power.

3.2 Coolant and Metal Temperatures

A. Lowering Reactor head flange and/or head temperature below the temperature

of fully tensioning Reactor head bolts may result in bolt relaxation and potential

leakage when Reactor vessel is pressurized during startup.

B. [TSAR/C] Monitoring coolant temperature when in MO,DE 4 with the vessel head

tensioned is performed using 3-SR-3.4.9.5-7. [Item 041]

C. The following limitations apply to Reactor heatup and/or cooldown:

1. When Reactor coolant temperature is less than 215°F, a maximum heatup

rate limit of 50°F/hr will reduce the O2 and Hydrogen Peroxide content of

the coolant.

2. During Reactor Heatup with Reactor coolant temperature greater than or

equal to 215°F, and during Reactor Cooldown, the optimum rate of

temperature change is 20°F every 15 minutes. This will ensure the

administrative limit of 90°F/HR is not exceeded. Do not Attempt to

"makeup" for time intervals which fall short of 20°F. If the 20°F is

exceeded in any 15 minute period, subtract the amount of

heatup/cooldown rate over 20°F from the 20°F for the next 15 minute

period. These guidelines will assist in achieving a target heatup/cooldown

rate of 80°F/Hr and ensure the administrative limit of 90°F/Hr is not

exceeded.

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 18 of 167

3.2 Coolant and Metal Temperatures (continued)

3. During Reactor heatup, operators should use metal temperatures as a

reminder that as metal heats up, the moderator HEATUP RATE will rise

with the same amount of heat input.

D. Minimizing operation with low feedwater flow and temperature or cold feedwater

flow cycling limits thermal duty on feedwater nozzles (REFER TO 3-01-3).

3.3 Primary Containment

A. [II/F] Prior to initiating any event which adds, or has the potential to add, heat

energy to the suppression chamber, the Unit Supervisor or Shift Manager will

evaluate the necessity of placing suppression pool cooling in service. This is

due to the potential of developing thermal stagnation during sustained heat

additions. [1I-B-91-129]

B. When containment integrity is required, airlock door seals should be tested

within seven days after each containment access per 0-TI-360 App A.

3.4 Control Rods, Reactivity Control and Relative Instrumentation

A. [NRC/C] Startups are performed using 3-SR-3.1.3.5(A) to incorporate Reduced

Notch Worth Procedure (RNWP) and Banked Position Withdrawal Sequence

(BPWS) recommended by G.E. [IE Bulletin 79-12, LER 260/84004]

B. [NER/C] Periodic pauses during control rod withdrawal are necessary to allow for

stabilization of neutron level and collection of data for estimating proximity to

critically. [SER 89-006, SOER 88-002]

c. [INPO/C] Adjustment of Nuclear Instrumentation readings downward to

match other indications without a full investigation and comparison with all

available methods to measure power level may result in non-conservative

power readings and protective setpoints. [SOER 90-003, SOER-88-002]

D. [NER/C] If SRMs or IRMs exhibit noise spikes during startup, control rod

withdrawal should be suspended and an assessment of SRM or IRM operability

performed in accordance with 3-01-92 or 3-01-92A, as applicable. [SOER 88-002]

E. [NER/C] Activities that can directly affect core reactivity are of a critical nature and

require strict procedural compliance, along with conservative actions. [INPO SER

89-006, SOER 88-002]

F. [NSRB/C] Reactivity can be added without moving control rods due to changing

plant conditions (such as lowering moderator temperature, lowering xenon

concentration, rising Reactor pressure, and rising feedwater flow) especially at

low power. Awareness of these conditions and monitoring core instrumentation

for these changes is required. [A258-4]

BFN Unit Startup 3-GOI-100-1A

Unit 3 Rev. 0074

Page 19 of 167

3.4 Control Rods, Reactivity Control and Relative Instrumentation

(continued)

G. In the event of an unexplained change in reactivity during an approach to

criticality, the approach to criticality must cease and the reactor core be made

sufficiently subcritical to prevent an inadvertent criticality. Approval of the Plant

Manager or his designee is required to resume the approach to criticality.

H. Reactor Engineering is required to be contacted to monitor flux shaping prior to

all power ascensions.

I. During the initial startup from MODE 4 following a refueling outage,

3-SR-3.1.1.1, Reactivity Margin Test, is required to be performed in conjunction

with the performance of 3-GOI-100-1A.

J. [NRC/C] Core Thermal-Hydraulic Stability, the reactor is required to be verified

outside Regions I, II & III. When OPRM's are INOP, REFER TO 3-SR-3.3.1.1.1.

[NCO 940245010]]

K. For Unit 3 Middle of Core Life to End of Core Life, the moderator temperature

coefficient of reactivity becomes positive as control rods are withdrawn for

startup when moderator temperature is below 350°F. The resulting effect will

be for Reactor power to rise until the moderator begins boiling and inserting

negative void reactivity. Exercise additional caution when withdrawing control

rods under this condition.

L. [OAlC] SPP-10.4 requires approval of the Plant Manager or his designee prior to

any planned operation with the following reactivity control equipment bypassed

unless bypassing of this equipment is specifically allowed within approved

procedures:

1. Rod Worth Minimizer

2. Rod Block Monitor

3. Source Range Monitors

4. Intermediate Range Monitors

5. Average Power Range Monitors

6. Refueling Interlocks

7. Integrated Computer System [ISE-NPS-92-R01]

8. OPRM Trip Function

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 20 of 167

3.5 Thermal Limits

A. Expected changes for Core Power and CPR during transients.

Raise In Variables CP CPR Reason I Transient

Subcooling r .1-* Loss of Feed water heating.

Cold Water Injection .

Core Flow r .1-* Runaway Recirc Pump.

Pressure .1- .1- Turbine Trip W/O Bypass.

MSIV Closure

Local Power Factor .1- .1- Control Rod Drop.

Xenon Shift

Axial Flux Shape .1- .1- Core Age

Bundle Power Raise is greater than critical power.

B. Operating in Single Loop Operation requires Safety Limit adjustments by

performing 3-SR-3.4.1 (SLO). Per Tech Specs 3.4.1, a completion time of

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed from the time of the Recirc Pump trip. However, these

actions should be performed as soon as possible.

C. Per Unit 3 TRM COLR the Thermal limits and off-rated corrections are provided

for Recirculation Pump Trip out-of-service and/or Turbine Bypass out-of-service

conditions. These events are analyzed for separate and/or concurrent

inoperability. The Shift Manager is required to make determination if startup

with the EOC-RPT will be disabled for startup.

D. [TSAR/C] Steady-state power operation is not permitted at Reactor vessel

pressure of greater than or equal to 1055 psia. MCPR analyses are not valid

above 1055 psia Reactor pressure. [Item 094]

3.6 EHC and Main Turbine

A. If hotwell pressure drops below -7"Hg with EHC pressure set less than 50 psi

above Reactor pressure and bypass jack above zero, bypass valve operation

could result.

B. Hotwell pressure above -25 inches Hg could result in low pressure turbine last

stage bucket failure.

C. Abnormal vibration in the main turbine during startup could result in turbine

damage.

BFN Unit Startup 3-GOI-1 00-1 A

Unit 3 Rev. 0074

Page 21 of 167

3.6 EHC and Main Turbine (continued)

D. If Turbine seals have been in service with the Turbine Turning Gear secured

and the unit is to be returned to operation, the following restrictions apply:

1. Turbine placed on the turning gear for 10 times as long as the period it was

stopped, up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , then check the eccentricity,

AND

2. If the eccentricity is higher than normal, the turbine is required to be left on

the turning gear until the eccentricity indication has reached and

maintained its normal minimum width for at least one hour. REFER TO

3-01-47.

E. During a Reactor startup, with an initial pressure greater than 150 psig and

EHC being unavailable prior to the startup, the EHC system should be placed in

service when it becomes available. The Main Turbine Shell and Chest warming

may begin when the conditions are met. Due to main turbine shell and chest

warming requirements, the EHC System should be placed in service prior to

950 psig.

F. The EHC Control System can be used in either Reactor Pressure control or

Header Pressure control. While in Header Pressure control, a single header

pressure input failing high could cause the bypass valves to open. While in

Reactor Pressure control, a single Reactor Pressure input failing high will not

affect the bypass valves. For this reason, Reactor Pressure control is the

preferred mode of operation for the EHC Control System.

G. Swapping pressure control sources ("HEADER PRESSURE CONTROL" to

"REACTOR PRESSURE CONTROL" or "REACTOR PRESSURE CONTROL"

to "HEADER PRESSURE CONTROL") may cause the turbine bypass valves to

open, depending on actual plant conditions.

H. When the pressure control swaps from "HEADER PRESSURE CONTROL" to

"REACTOR PRESSURE CONTROL" the pressure set will be actual Reactor

pressure at the time the swap is done, regardless of any previously raised

Reactor pressure set done during a Reactor startup.

3.7 Electrical Alignments and Load Considerations

A. Downpowering of Nuclear Units Under Low System Load Conditions:

Due to having five nuclear units in an operating status, the frequency of

downpowering units under low system load conditions is expected to rise. The

following communications process will be used to coordinate downpowering a

unit at BFN under low load conditions:

BFN Unit Startup 3-GOI-1 00-1 A

Unit 3 Rev. 0074

Page 22 of 167

3.7 Electrical Alignments and Load Considerations (continued)

1. The Electrical System Operator (ESO) will anticipate the potential need to

downpower nuclear units as far in advance as reasonable, normally one to

two days. The ESO will inform the Operations Duty Specialist (ODS) of

this potential need.

2. The ODS will notify the Browns Ferry Shift Manager that a potential need

to downpower exists.

3. The Shift Manager will notify the Operations Superintendent who will notify

the Operations Manager and Duty Plant Manager.

4. BFN will initiate a telecon with other operating nuclear units and senior

nuclear corporate management (normally, Senior Vice President, Nuclear

Operations, or, President, TVA Nuclear and Chief Nuclear Officer) to

formulate a contingency plan. The plan will address which units are to be

downpowered based on existing plant conditions, the reduction capability

of each unit, time to reach reduced power as well as return to full power,

and the preferred order for downpowerinq.rsrscci Scram Reduction

Recommendations G-20-1 and G-20-2 require an additional SRO licensed

operator to assist with BOP operations when performing power maneuvers

during unit startup or shut down with Reactor power less than 60%. This

individual is not limited to the Control Room. [SFRC-17, G-20-1 & -2]

5. The contingency plan will be communicated to the appropriate site

management and Shift Manager for the impacted units as well as the

transmission/power supply organization.

6. The ESO will notify the designated Shift Managers approximately two to

four hours before the need to actually downpower. The Shift Manager will

notify the Operations Superintendent of any actual downpower.

7. Any change to unit status that would impact the agreed upon contingency

plan will cause the telecon to be reconvened with all affected parties and a

revised contingency plan developed. This will be initiated by the site

management who identifies the need to revise the plan.

B. Electrical alignments and Bus loading are to be made in accordance with

0-01-57A, Switchyard and 41,60V Electrical System and 0-01-57B, 480V/240V

AC Electrical System.

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 23 of 167

3.8 Condensate and Feedwater

A. Changes in Condensate System flow may require adjustment to SPE CNDS

BYPASS, 3-FCV-002-0190, either in the Control Room or locally. Personnel

adjusting this valve locally shall be in direct communication with the Control

Room. Evolutions resulting in changes in condensate/feedwater flow

(condensate/booster pump start, feedwater pump start, changes in Reactor

power, feedwater flow, steam flow, etc) will affect flow rates

through 3-FCV-002-0190, steam-jet air-ejector condenser(s), steam packing

exhauster condenser, and Off-Gas condenser. SJAE/OG CNDR CNDS FLOW,

3-FI-2-42, on Panel 3-9-6 should be maintained between 2 X 106 Ibm/hr

and 3 X 106 Ibm/hr.

B. The following are the limitations on the Condensate system and the Reactor

Feedwater Pumps during normal, steady-state operations:

1. Condensate System:

a. [II/C] Condensate flow should always be maintained within the following

limits, using 3-FC-2-29 in BAL if possible, to prevent Condensate

Pump damage:

(1) One Condensate Pump operation, greater than 1.5 X 106 Ibm/hr

but less than 6.25 x 106 Ibm/hr.

(2) Two Condensate Pump operation, greater than 3.0 X 106lbm/hr

but less than 12.5 x 106 Ibm/hr.

(3) Three Condensate Pump operation, greater than 4.5 X 106 Ibm/hr

but less than 15.0 x 106 Ibm/hr. [11-8-91-158]

b. Normal maximum line current to Condensate Pump Motors should not

exceed 118 amps steady-state operations.

c. Normal maximum line current to Condensate Booster Pump Motors

should not exceed 225 amps steady-state operations.

2. Reactor Feedwater Pumps:

a. Individual Reactor Feedpump speed should be less than 5050 RPM.

C. RFW START-UP LCV, 3-LCV-3-53, does not have a hole in the disc allowing

flow at low pressures. The valve does have relief ports that may allow a small

amount of water to pass. The flow should not be of significant amount, but

3-FCV-3-53 may be isolated at the Unit Supervisors discretion.

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 24 of 167

3.9 Radiation Protection Notifications and Radiological Protection

Hold Points (RPHPs) [SOER 01-1, BFN PER 126211, PER 961778, PER 116666]

A. This General Operating Instruction initiates processes that are likely to cause

increased radiation levels that raise the risk of unintended radiological

exposures and also radiation levels that warrant High Radiation Area or Locked

High Radiation Area Controls.

This GOI relies heavily on a multitude of System Operating Instructions (support

procedures) for system alignments required for the various process systems.

Many of these alignments can and do result in raising the radiological impacts

for the areas affected by the alignments. Therefore, there are increased

probabilities of unintended radiation exposures to personnel that may be

occupying these areas when alignments take place, and when reactor power

increases occur.

B. To reduce the probability of unintended radiation exposures, the following

controls are established by this procedure:

1. Radiological Protection Hold Points (RPHPs) are pre-established at

appropriate locations in this GOI and in the support procedures. The

function of RPHPs is to allow Radiation Protection to help ensure no

unintended radiological exposures occur as the result of a system

configuration change or raising reactor power. This may require holding

actions for a step (actions typically identified with a BEFORE conditional

step) until verifying personnel are not in the area before continuing in the

procedure. These RPHPs also allow a determination as to whether actions

are required to implement RCI-'17, Control of High Radiation Areas and

Very High Radiation Areas, controls.

2. The Radiation Protection notification steps have an (R) placed in the step

initial line, which means these steps can NOT be omitted unless the action

associated with the step is not performed, or the step allows the notification

to be N/A'd as determined by the Unit Supervisor.

3. An Appendix (Appendix A, Radiation Protection Notification Record) is

provided to record Radiation Protection notifications, RPHPs, and release

of RPHPs, as necessary. The instructions for Appendix A is used to identify

the appropriate required logging of Radiological Protection entries. The

primary function of the appendix is to ensure proper communication with

Radiation Protection personnel and that they are allowed sufficient

opportunity to implement needed radiological controls.

4. Radiation Protection notification steps that require a RPHP are clearly

worded that an RPHP is in effect. For these steps, it should be made clear

to Radiation Protection that an RPHP is in effect so that they understand

that a signature on Appendix A will be necessary.

BFN Unit Startup 3-GOI-1 00-1A

Unit 3 Rev. 0074

Page 25 of 167

3.9 Radiation Protection Notifications and Radiological Protection

Hold Points (RPHPs) [SOER 01-1, BFN PER 126211, PER 961778, PER 116666]

(continued)

Radiation Protection notification steps that are not identified as RPHP

steps are considered courtesy notification steps to Radiation Protection.

These steps serve the purpose of informing Radiation Protection of

evolutions that are about to be implemented that may impact plant

radiological 'conditions and allow them to respond or "get their ducks in a