IR 05000529/2003301

May 2003 Retake Exam 50-529-2003-301 Final JPMs & Outline

ML040560299
Person / Time
Site:	Browns Ferry, Palo Verde
Issue date:	05/08/2003
From:	NRC/RGN-II
To:
References
50-259/03-301, 50-260/03-301, 50-296/03-301 50-259/03-301, 50-260/03-301, 50-296/03-301
Download: ML040560299 (210)

Text

Final Submittal BROWNS FERRY RETAKE EXAM 50-25912003-301 ADMIN PART A MAY 8,2003

ES-301 Administrative Topics Outline Form ES-301-1 (R8,Sq)

xility: BFN Date of Examination: 05/08/2003 xamination Level (circle one): SR8 Operating Test Number: Remedial Administrative Describe method of evaluation:

Topic!Subject 1, ONE Administrative JPM, OR Description 2. TWO Administrative Questions p_s_

_i

,.4 MODECHANGE JPM A I .1R TECH SPEC 3.0.4 DETERMINATION PERFORM JPM A I .2R SPECIFIC AND DETERMINE REGULATORY REPORTING REQUIREMENT INTEGRATED PLANT PROCEDURE:

SURVEILLANCE JPM A2.1R TESTING CORE SPRAY SYSTEM VALVE TIMING

__ RADIATION JPM A3.lR EXPOSUR REVIEW AND INTERPRET A RADIOLGOICAL SURVEY MAP LIMITS AND RADIATION CONTROL

_ .

JPM A4.2 R EMERGENCY CLASSIFY EVENT PER REP PROCEDURE ACTION LEVEL ICBMMUNICATI ONS

F;NA E§-301 Oeeratino Test Qualitv Checklist Form ES-301-3 Initials 1. GENERAL CRITERIA The operating lest conforms with the previouslyapproved outline; chan@5 we amSiStent idth sampling rwuirements (e.0.. 10 CFR 55.45, operational importance. safety function distribution). There is IKI day-to-day repetition between this and other operating tests to be administered during this examiMtio The operating test shall not duplicate items hom the applicant$ audit test(s)(see S&ion 0.l.a). Overlap with the written examination 8mi between operating test cafegories is within acceptable limit . WALK-THROUGH (CAEGORY A B 8 ) CRITERIA Each JPM imludes the foliowing. as applicable:

. initial codifions

, initiating we9

. refermcesand tods.induding associated procedures

~ reasonable and vali&atedtime limits (average time allowed for mpletion) and specific desigmtlon if desmal to be time critical by the facilty licsn5ee

. 5pxific perfofmance criteria that include:

~ detailed expected actions with exad criteria and nomenclature

~ system re5ponse and other ernminer wes

- statements describing important observations to b@made Lv the applicant

. criteria for sumssful cwnpl@tionof the task

. identification of critical steps and their assdated pedoimanca standards

. restrictions on the sequence of steps. ifapplicakle The prescripted questions in Catqory A are predomhsntly open reference and meet the criteria in Attachment 1 of ESdO Printed Name / Sig I. Author I. Facility ReviewW

. NRC Chief Examiner (#)

I. NRCSu~~or IQTE: * The faciiity signature is no! appiicable for NRC-developedtest of 26 NUREG-1021, Revision 8, Supplement 1

JPM NO. A1.1R REV.NO 0 PAGE 1OF4 BROWNS FERRY NUCLEAR PLANT JOB PERFORMANCE MEASURE JPM NUMBER: Al.lR TITL TECH SPEC DETEKL'INATION - 3 TASK NUMBER: S-000-AD-27 PLANT CONCURRENCE: _ _ _ - ~ ~ .- DATE:-.

OPERATIONS t

Examination JPMs Require Operations Traioing Manager Or Designee Appro:.a!

and Plant Concurrence

. .

JPMNO. A1.1R REV.NO 0 PAGE 20F4 BROWNS FERRY NUCLEAR PLANT JOBP~RFORMANCEMEASURE TASK TITLE: TECH SPEC DETERMINATION - 3. WA NUMBER: 2.1.12 W A RATING: R O L 9 - SRO: f **+***t*d*************t**t*c**t******************~*~~~****~*~*******~*~**~***~*~*****~******~~+****

TASK STANDARD: B etermine that a mode change from Mode 2 to Mode 1 with inoperable. required equipment is not allowed in accordance with Technical Specification

.. . ~ .

3. LOCATION OF PERFORMANCE: SIMULATOR ,~x- PLANT -x- CONTROL ROOhl x-REFERENCESPWOCEDURES NEEDED: Unit 2 technical Specifications, 3.0.4, 3. VALIDATION TIME: CONTROL ROOM: NIA LOCAL: NiA ..

MAX.TIME ALLOWED: NIA . (Completed for Time Critical JPMs onlyj PERFORMANCE TIME: NIA CONTROL ROOM NIB LOCAL Jfi.~-

COMMENTS:__ .l__.l_____ ~

Additional comment sheets attached? YES NO RESULTS: SATISFACTORY UNSATISFACTORY __

EXAMINER SIGNATURE: - DATE: __

EXAM1 NER

JPMNO. A1.1R REV.NO 0 PAGE 3 0 F 4 BKOWNS FERRY NUCLEAR PIANT JOE PERFORMANCE MEASURE REFERENCES ALLOWED INITIAL CONDITION

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The current time is 1900 hrs, Unit 2 is starting up f r o r cold shutdown and is now hclding at 5% rated thermal power with the Mode Switch in Startiip. Ari oil !eak was discovered on 2A RHR pump wotor at 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> tc<!a A 7 day LCO was ente:eb. Mainterance was pei%rrned and !he oil leak repaired a! 1DO0 hours today The F Order has been released. Cleawp and post mr?ir:!cnance testing ,s.exxegt~e< be completed by 2'100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> toda INITIAT!NG .- CUE:

.-

Determine If reactor stamp !o 30% rated ?ne:rr,al power may contiwe while cleawp and uost imaintetxmr:

testirg to prove 2.4 WHR pump operaf:le is iii progres ....~. Indicate the documentation which defends vour answer EXPECTED RESPONSE k x a c t wording not recuiygd)

No, this would require a mode change from Mode 2 to Mode I with inoperable, required ECCS equipmen REFERENCES:

Knowledge item The operator will have to reach the conclusion that for the startup to continue to ?hedesirecj power level that entry into Mode 1 would be required. This will required applicath? of I. CO 3.0.4 and LCQ 3. Tech Spec LCO 3.0.4(Page 3.0-2)

When an LCO is net metL

__ ~ entry

.- -~

into

_- a hIO_DE_or other specified condition in th,e Applic:jbilit.;.

shall not be made -- ~ ~ ~ econt~!&

e. x c s t when ?hea F ; s o ~ a t e d ~ ~ C ~ I ( ~ _ N ~ t ~ - permit nlered m a t -. __

i o .n- in the MODE-. or other specified condition i r i the Applicabiiity for aEu>!lizj:gd period ofLm3. This Specification shall not Frwcnt chancjes in MODES or other s p e c i t ' d

..- +.

conditions in the Applicability that are required to comply with ACTONS or that are a part of a shutdown of the uni Tech Spec LCO 3.6.4 Basis (Page 33.0-6)

The basis fer LCQ 3.0.4 states, " K O 3.0.4 establishes limitation on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It precludes placing the unit in a different Mode or other specified condition stated in that Applicability (e.g.,

Applicability desired to be entered) when the folioiving exist:

a. Unit conditions are such that requirements of the LCB would not be met in the Applicability desired to be entered; and b. Continued noncompliance with the LCO requirements, if the Applicability were entered, would result in the unit being required to exit the Applicability desired to !)e entered to comply with the Required Actions."

Tech Spec LCO 3. ECCS The applicability of LCO 3.5.1 is Mode 1 and Mode 2 & 3 (with an exception for HPCl and ADS at el5Opsig). This LCQ requires ALL ECCS systems to be operable. Current Mode is Mode 2 and for startup to continue would require entry into Mode 1, LCO 3.5.1 Conditioc A requires the inoperable system to be restored within 7 days, condition B would be entered lf the subsystem were not restored and woulc! require the Unit be placed in Mode 3 then

~...~ ~

Mode Conclusion:

Although it is expected that RHR would be operable and LCO 3.5.1 met prior to exceeding the 7 day completion time of 3.5.1.A, b C 0 3.0.4 forces the operator to consider that the inoperability would be "continued noncompliance" and uitimately action statement 3.5.";.8 would be entered which would require exit from Mode 1. For the current plant conciitioris and application of TS 3.0.4 and TS 3.5.1, the startup can not continue because entr{ into Mode 1 would be required.

c

CANDIDATE'S HAN5OUT REFERENCES

_ _ _ ~ -ALLOW=

~

INITIAL CONDITION The current time is 190Qhours, Unit 2 is starting up from cold shutdown and is now holding at 8% rated thermai power with the Mode Switch in Startup. 4n oil leak was discovered on 2A WHR pump motor at 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> today. A 7 day LCO was entered. Maintenance was performed and the oil leak repaired at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> today. The Hold Order has been released. Cleanup and post maintenance testing is expected to be completed by 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br /> today.

INITIATING CUE:

Determine if reactor startup to 30% rated thermal power may continue while cleanup and post maintenance testing to prove 2A WHR pump operable is in progress.

Indicate the dscurnentz.lion which defend5 your answe LCO Applicability .0 LCO APPLICABILITY (continued)

LCO 3. When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued operation in the MQDE or other specified condition in the Applicability for an unlimited period of time. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are a part of a shutdown of the uni Exceptions to this Specification are stated in the individual Specification LCO 3.Q.4is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3 LCO 3. Equipment removed from service or declared inoperable to coinply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is

.. an exception to LCQ 3.0.2 for the system returned to service ,rider administrative control to perform the testing required to demonstrate OPERABILIT LCO 3. When a supported system LCO is not met solely due to a suppcrt system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. Phis is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.1Z , "Safety Function Determination Program (SFBP)." If a loss of safety function I S

ECCS - Operating 3. EMERGENCY COKE COOLING SYSTEMS (ECCS)AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3 5 1 ECCS - Operating LCO 3.5 1 Each ECCS injectionlspray subsystem and the Automatic Depressurization System (ADS) function of six safetykelief valves shall be OPERABL APPLICABILITY MODE 1, MODES 2 and 3 except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig ACTIONS .. .

CONDITION REQUIRED ACTION

.. , .

A. One low pressure ECCS Restore low pressure 7 days injectionlspray subsystem ECCS injectionlspray

._

-

inoperabl subsystem(s) to OPERABLE statu One low pressure coolant injection (LPCI) pump in both LPCI subsystems inoperabl B. Required Action and Be in MODE hours associated Completion Time of Condition A not yb&

me .2 Be in MODE 4 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> BFN-UNIT 2 3.5-1 Amendmerd No. 253, 2E9 March 12 23C1

ECCS - Operating 3 __

ACTIONS (continued)

._ CONDITION REQUIRED ACTION COMPLETION TIME

- ___.___

C. HPCl System inope C1 Verify by administrative Immediately means RClC System IS OPEBABLE 4ND C2 Restore HPCl System to 14 days OPERABLE status D. HPCl System inope D1 Restore HPCI System to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status ANB SR Condition A enterec D2 Restore low pressure ECCS injection/spray 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> I subsystetn to OPERABLE status E One ABS valve Restore ABS valve to I4 days inoperable OPERABLE statu ^

-

F. One ADS valve F1 Restore ADS valve to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperabl OPERABLE status AND OR F2 Restore low pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Condition A enterec ECCS iryectionispray I subsystem to OPERABLE status BFN-UNIT 2 3.5-2

ECCS - Operating 3. ACTIONS (continued) __ ~ _.___

. . CONDITION REQUIRED ACTION COMPLETION TIME

.____

G. TWQor more AD§ valves Be in MODE hours inoperabl AN D

-

OR Reduce reactor steam 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Required Action and dome pressure to associated Completion s 150 psi Time of Condition C, D, E, or F not me ~

H. Two or more low pressure H . l Enter LCO 3. Immediately ECCS injectionlspray subsystems inoperable for reasons other than Condition OR

-

HPCl System and one or more ABS valves inoperable BFN-UNIT 2 3.5-3 Amendment N3 253 260 March 12 ^:c'04

ECCS - Operating 3. SURVEILLANCE REQUIREMENTS

-.iaiY- . - . ~. ~ . -._

~. :

S URVEI LbANCE FREQUENCY

- . .~ .~ ~

SR 3.5. Verify, for each ECCS injectionkpray 31 days subsystem, the piping is filled with water from the pump discharge valve to the injection valv S R 3.5.1.2 --.-.-.----------------...N~TE-------------------------

Low pressure coolant injection (LPCI)

subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RtiR) tow pressure permissive pressure in MODE 3. if capable of being manually realigned and not otherwise inoperabl Vel ,fy each ECC§ injectionispray subsystem 31 days manual, power operated, and automatic valve in the flow path, that is riot locked, sealed, or otherwise secured in position, is in the correct positio SR 3.5. Verify ADS air supply header pressure is I? 81 31 days psi . .

SR 3.5. Verify the bPCl cross tie valve is closed and 31 days power is removed from the valve operato (continuedi

._ BFN-UNIT 2 4.5-4 Amendment No. 253

- -

ECCS - Operating 3. SURVEILLANCE

.- R E Q U l R E M E N T S m i n u e d )

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SURVEILLANCE FREQUENCY SR 3.5. s _ _ _ _ _ _ _ _ _ _ _I___-____

____ NOTES _s_____s s______________

1. Only required to be performed when in MODE 4 > 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> __ssl______s____.-__I_D______l_s________-~-~~~~~~..-----~-~.

Verify each recirculation pump discharge Once prior to valve cycles through one complete cycle of entering MODE 2 full trave!. from MODE 3 or

specified flow rate against a system head with the lcservice corresponding to the specified pressur Testing Program TO A VESSEL TO TORUS NO OF DIFFERENTIAL SYSTEM

_ l FLO'W RATE,

- u PSESSURE OF, Core Spray 2 6250 gpm 2 r105psid INDICATED NO OF SYSTEV SYSTEM FLOW R A l E pVME PFiESSURE

__-

LPCI LPCI

_ I _

2 12,000 gp z 9,CCOgpm

__

_.-

.. .I_

~

1

? 250 p i g 1' 125 3sig 1 .

(cor.t i zu ed j BFN-UNIT 2 3.5-5 Amendment No 252

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS(eontrnue3

._I ~

SURVEILLANCE FREQUENCY

. I - - . ~ . ...

SR 3.5. _s_s~s______________...___NO~~ _____ _ss________ ~ _____

__

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the tes _ s s _ _ _ _ _ _ _ _ _ _ s s _ _ . _ _ - - ~ ~ ~ ~ ~ - - - - ~ - - - ~ ~ . - - ~ - ~ " - ~ - - - - - - - - ~ ~ ~ - - ~

Verify, with reactor pressure c 1040 and 22 days 2 950 psig, the HPCI pump can develop a flow rate ;, 5000 gpm against a system head corresponding to reactor pressur SR 3.5. Verify, with reactor pressure c 165 psig, the 24 months HPCl pump can develop a flow rate r 5000 gpm against a system head corresponding to reactor pressur Verify each ECCS injectionlspray subsystem 24 monttx actuates on an actual or simulated automatic initiation signa ~ ~ _ _ -

(ccntinued)

BFN-UNIT 2 3 5-6 Amendment No 255 November 33 13%

ECCS - Operating 3. __

SURVEILLANCE B ~ Q U I B E M E N T S c o n t i n u..e d ~ ~

.. SURVEILLANCE FREQUENCY SR 3.5.1.10 _ _____s___.._s____....~--- NOTE _ _ _ _ _ _ _ _ _ _ _ _ _ _ s~ ssss_s

___

Valve actuation may be exclude Verify the ABS actuates on an actual or 24 months simulated automatic initiation signa SR 3.5.1.11 ___-___ ss___ _____.._ NOTE---- ____.___ _ ss__.-______

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the tes Verify each ADS valve opens when manually actuate months !I

- ~ _ I _ . . _ _ ~ ~ _ . _ _ - . . . _ ~ _ _ ~ . ~ _ _ _ _ _ ~ I _ . - ~

S R 3.5.1.12 Verify automatic transfer of the power supply 24 months from the normal source to the alternate source for each LPCl subsystem inboard injection valve and each recirculation pump discharge valv _:., -

.

I_ _~-~.-=i~~-~..~--.~.-l ~ _i

- BFN-UNIT 2 3.5-7

ECCS - Shutdown 3. .5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5 2 ECCS - Shutdown LCO 3. Two low pressure ECCS injectionlspray subsystems shall be OPERABL APPLICABILIVY: MODE 4, MODE 5, except with the spent fuel storage pool gates removed and water level 2 22 ft over the top of the reactor pressure vessel flange

- . 0. N~S ~~

ACT1 , ~ ~ ~: . ~~ . =~ ... ~ .-- . - _

~ j~-

- ~., ~

I .~ - - _ _ . - ~ = -..._i J

CONDITION REQUIRED ACTION COMPLETlON 1;

TIM%

________ ~ - .-

A. One required ECCS Restore required ECCS 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> injectionlspray subsystem injectionlspray subsystem inoperabl to OPERABLE status.

. .. .

~ ~

B. Required Action and Initiate action to suspend Immediately Time of Condition associated A not Completion Q P5 RV me ~ ~ _- --

C. Two required ECCS Initiate action to suspend Immediately injectionispray OP DRV subsystems inoperabl I C2 Restore one ECCS 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> iryectionlspray subsystem to OPERABLE status L

BFN-UNIT 2 3 5-8 Amendrrert No 253

ECCS - Shutdewn 3. ACTlONS

- - (continuek

.. ... .. CONDITION REQUIRED ACTION COMPLETION TIME

- _______I-D. Required Action C.2 and Initiate action to restore immediately associated Completion secondary containment to Time not me OPERABLE statu AND Initiate action to restore Immediately two standby gas treatment subsystems to OPERABLE statu Initiate action to restore immediately isolation capability in each required secondary containment penetration flow path net isolate EFN-UNIT 2 3.5-9 Amendmer? Ne 253

..

ECCS - Shutdown 3. REQUIREMENTL SURVEiLLANCE......

.... . ._. .-. ,-. . .. = . .: . -=

SURVEILLANCE FREQUENCY SR 3.5. Verify, for each required ECCS 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> injection/spray subsystem, the suppression poor water level is 2 -6.25 inches with or-7.25 inches without differential pressure, contro SI? 3.5. Verify, for each required ECCS 31 days injectionispray subsystem; the piping is filled with water from the pump discharge valve to the injection valv .~

SR 3.5. ...-----.------NOTE-----.--------.----~----.

One LPCi subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperabl Verify each required ECCS injectionlspray 31 days subsystem manual: power operated: and automatic valve in the flow path, that is not locked, sealed: or otherwise secured in position, is in the correct positio (cortiwed)

- BFN-UNIT 2 3.5-10 Amendment No 253

ECCS - Shutdown 3. SURVEILLANCE I-- ~..REQUIREMENTS (continued) .-

SURVEILLANCE FREQUENCY SR 3.5. Verify each required ECCS pump develops In accordance the specified flow rate against a system head with the Inservice corresponding to the specified pressur Testing Program TORUS NO OF DIFFERENTIAL SYSTEM FLOW R A E FLMS PRESS.U.RE.OE cs 26250 gpm 2 ;: 105 psid INDICATED NO OF SYSTEM SYSTEM FLOVVRRATE PUMPS PRESSURE LPCl 25.000 gpm 1 2 125 psig Verify each required ECCS injeetionkpray 24 months subsystem actuates on an actual or simulated automatic initiation signa .~,,~~li.~.i~~~

.~-.- __ ^. ., ~

BFN-UNIT 2 3.5-11

.,.

RClC System 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTQR CORE ISOLATkON COOLING (RCIC) SYSTEM 3 5 3 RCIC System LCO 3. The RClC System shall be OPERABLE APPLICABILITY: MODE. 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig ACTIONS __ -

.-, CONDITION

-_ . ._

A. RCIC System inoperabl A.l Verify by administrative lmmediateiy means High Pressure Coolant Injection System is OPERABL AND A2 Restore RCIC System to 14 days OPERABLE statu . Required Action and B.1 Be in MODE hours associated Completion Time not me BND i 8.2 Reduce reactor steam 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> dome pressure to 5; 150 psig.

BFN-II *IT 2 3.5-12 Amendment NO Z Y i

RClC System 3.5.3 SURVEILLANCE WE Q UIREMENTS .

SURVEILLANCE

_- --..-___

SR 3.5. Verify the RCIC System piping is filled with 31 days water from the pump discharge valve to the injection valv SR 3.5. Verify each RClC System manual, power 31 days operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct positio Verify, with reactor pressure z 1040 psig and 32 days 2 950 psig, the RClC pump can develop a flow rate z 600 gpm against a system head corresponding to reactor pressure Verify, with reactor pressure 5 165 psig, the 24 months RClC pump can develop a flow rate i 600 gpm against a system head corresponding to reactor pressure (cont i r uecl )

BFN-UNIT 2 3.5-13 Amendment Nc 255 November 33 15%

RClC System 3.5.3

....

SURVEILLANCE REQUIREMENTS (continued)

-. I SR 3.5. SURVEILLANCE

..-...

t NOYE.........................

Vessel injection may be exclude ______________s_----Isss_____sp_________-------~----------~.

Verify the RClC System actuates on an actual or simulated automatic initiation signa FREQUENCY 24 months V

BFN-UNIT 2 3.5-14 Amendment No 255

-.., November 31) '998

LCO Applicability B BASES LCO 3. Therefore, this LCQ can be applicable in any or all MODES. If the (continued) LCO and the Required Actions of LCO 3.7.6 are not met while in MODE 4 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LCQ 3.7.6 of "Suspend movement of irradiated fuel assemblies in the spent fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of bCO 3. These exceptions are addressed in the individual Specification LCQ 3. LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It precludes placing the unit in a different MODE or other specified condition stated in that Applicability (e.@,Applicability desired to be entered) when the following exist:

a. Unit conditions are such that requirements of the b C 0 would not be met in the Applicability desired to be entered; and b. Continued noncompliance with the LCO requirements, if the Applicability were entered, would result in the unit being required to exit the Applicability desired to be entered to comply with the Required Action Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of !he unit before or after the MODE change. Therefore, in such cases. entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions. The provisions of this Specification should not b e interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status befor-a m l i startu !contirim+

BFN-UNIT 2 B 3 0-6 Revislor? (3

_ d

LCO Applicability B BASES LCO 3. The provisions of LCO 3.0.4 shall not prevent changes in MORES (continued) or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3. shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdow Exceptions to bCO 3.0.4 are stated in the individual Specification The exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time. Exceptions may apply to all the ACTIONS or to a specific Required Action of a Specificatio LCO 3.0.4 Is only applicable when entering MODE 3 from MODE 4, MOBE 2 from MODE 3 or 4, or MODE 1 from MODE Furthermore, LCO 3. is applicable when entering any other specified condition in the Applicability only while operating in MORE 1, 2, or 3. The requirements of LCO 3.0.4 do not apply in MODES 4 and 5:or in other specified conditions of the Applicability (unless in MOBE 9 , 2, or 3) because the ACTIONS of individual specifications sufficiently define the remedial measures to be taken. [In some cases (e.g.,..) these ACTIONS provide a Note that states "While this bCO is not met. entry into a MODE or other specified condition in the Applicability is not permitted, unless required to comply with ACTIONS." This Note is a requirement explicitly precluding entry into a MODE or other specified condition of the Applicability.]

~. .- (continued)

BFN-UNIT 2 B 3 0-7 Revision 3

-.

LCO Applicability B 3.0 BASES LCO 3. Surveillances do not have to be performed on the associated (continued) inoperable equipment (or on variables outside the specified limits),

as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, either in compliance with LCO 3.0.4 or where an exception to LCO 3 0 . 4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs mtist be met to ensure OPERABlblTY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.

LCO 3. LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action(s)) to allow the performance of required testing to demonstrate:

a. The OPERABILITY of the equipment being returned to service:

or b. The OPERABILITY of other equipmen The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to pedorm the required testing to demonstrate O P E R A B I L I ~ .This Specification does not provide time to perform any other preventive or corrective maintenanc ECCS - Operating B 3. B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (KCIC) SYSTEM B 3.5.2 ECCS - Operating BASES

~ . ... : ~... ____

BACKGROUND The ECCS are designed, in conjunction with the primary and secondary containment, to limit the release of radioactive materials to the environment following a loss of coolant accident (LOCA). The ECCS uses two independent methods (flooding and spraying) to cool the core during a LOCA. The ECCS network consists of the High Pressure Coolant Injection (HPCI) System, the Core Spray (CS) System, the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System, and the Automatic Depressurization System (ADS). The suppression pool provides the required source of water for the ECCS. Although no credit is taken in the safety analyses for the condensate storage tank (CST), it is capable of providing a source of water for the HPCI, RHR and CS system The ECCS design requirements ensure that the criteria of Reference 12 are satisfie On receipt of an initiation signal, ECCS pumps automatically start; simultaneously, the system aligns and the pumps inject water, taken either from the CST or suppression pool, into the Reactor Coolant System (RCS) as RCS pressure is overcome by the discharge pressure of the ECCS pumps. Although the system is initiated, ADS action is delayed, allowing the operator to interrupt the timed sequence if the system is not neede The HPCI pump discharge pressure almost immediately exceeds that of the RCS, and the pump injects coolant into the vessel to cool the core. If the break is small: the HPCI System will maintain coolant inventory as well as vessel level while the

- - . (ContlnueQ BFN-UNIT 2 B 3 5-1 Reblslcn 0

-

ECCS - Operating B 3. BASES Bi. '.GROUND RCS is still pressurized. If HPCl fails, it is backed up by ADS in (continued) combination with LPCl and CS. In this event, either the vessel would be manually depressurized or the ABS timed sequence would be allowed to time out and open the selected safetykelief valves (SIRVs) depressurizing the RCS, thus allowing the LPCl and CS to overcome RCS pressure and inject coolant into the vessel. If the break is large, RCS pressure initially drops rapidly and the LPCI and CS cool the cor Water from the break returns to the suppression pool where it is used again and again. Water in the suppression pool may be circulated through a heat exchanger cooled by the RHR Service Water System. Depending on the location and size of the break, portions of the ECCS may be ineffective; however, the overall design is effective in cooling the core regardless of the size or location of the piping brea All ECCS subsystems are designed to ensure that no single active component failure will prevent automatic initiation and successful operation of the minimum required ECCS equipmen The CS System (Ref. 1) is composed of two independent subsystems. Each subsystem consists of two 50% capacity motor driven pumps, a spray sparger above the core, and piping and valves to transfer water from the suppression pool to the sparger. The LOCA analysis (Ref. 13) requires both pumps in a subsystem (loop) to be OPERABLE for the subsystem to be OPERABLE. Failure of one CS pump results in the loss of the associated CS loop for LOCA mitigation. The CS System is designed to provide cooling to the reactor core when reactor pressure is low. Upon receipt of an initiation signal. the CS pumps in both subsystems are automatically started (A pump

- I (continued)

BFN-UNIT 2 B 3.5-2 Revision 3

-

ECCS - Operating B 3.51 BASES BACKGROIINB immediately when offsite power is available and B, C, and D (continued) pumps approximately 7, 44, and 21 seconds afterwards and if offsite power is not available all pumps 7 seconds after AC power is available). When the RPV pressure drops sufficiently, CS System flow to the RPV begins. A full flow test line Is provided to route water from and to the suppression pool to allow testing of the CS System without spraying water in the RP LPCl is an independent operating mode of the RHR Syste There are two LPCl subsystems (Ref. 2), each consisting of two motor driven pumps and piping and valves to transfer water from the suppression pool to the RPV via the corresponding recirculation Coo The two LPCl pumps and associated motor operated valve? in each LPCI subsystem are powered from separate 4 kV shutdown boards. Both pumps in a LPCl subsystem inject water into the reactor vessel through a common inboard injection valve and depend on the closure of the recirculation pump discharge valve following a LPCl injection signa Therefore, each LPCI subsystem's common inboard injection valve and recirculation pump discharge valve are powered from one of the two 4 kV shutdown boards associated with that subsystem. The ability to provide power to the inboard injection valve and the recirculation pump discharge valve from two independent 4 kV shutdown boards ensures that a single f a h e of a diesel generator (DG)will not result in the failure of both LPCI pumps in one subsyste (continhedl BFN-UNIT 2 B 3.5-3

%_

ECCS - Operating B 3. BASES BACKGROUND The two LPCI subsystems can be interconnected via the LPCl (continued) cross tie valve; however, the cross tie valve is maintained closed with its power removed to prevent loss of both LPCk subsystems during a LOCA. The LPCl subsystems are designed to provide core cooling at low. KPV pressure. Upon receipt of an initiation signal, all four LPCl pumps are automatically started (A pump immediately when offsite power is available, and B, C: and D pumps approximately 7, 14, and 21 seconds afterwards; if offsite power is not available, all pumps immediately when AC power is available). RHR System valves in the LPCI flow path are automatically positioned to ensure the proper flow path for water from the suppression pool to inject into the recirculation loops. When the RPV pressure drops sufficiently, the LPCl flow to the RPV, via the corresponding recirculation loop, begins. The water then enters the reactor through the jet pumps. Full flow test lines are provided for the four LPCl pumps to route water from the suppression pool. to allow testing of the LPCI pumps without injecting water into the RPV. These test lines also provide suppression pool cooling capability, as described in LCQ 3.6.2.3, "RHR Suppression Pool Cooling."

The HPCl System (Ref. 4) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbin as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger. Suction piping for the system is provided from the CST and the suppression pool. Pump suction for HPCl is normally aligned to the CST source to minimize injection of suppression pool water into the RPV. However, if the CST

- - (continued)

BFN-UNIT 2 B 3 5-4 Revision 0

-

ECCS - Operating B 3. BASES BACKGROUND water supply is low, or if the suppression pool level is high, an (continued) automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the HPCI System. With HPCl taking suction from the condensate storage tank and injecting to the reactor vessel, there is sufficient inventory in the tank such that the high'suppression pool level suction transfer will occur before a low condensate header level would be created. The steam supply to the HPCl turbine is piped from a main steam line upstream of the associated inboard main steam isolation valv The HPCl System is designed to provide core cooling for a wide range of reactor pressures (I 50 psig to 1174 psig). Upon receipt of an initiation signal, the HPCl turbine stop valve and turbine control valve open and the turbine accelerates to a specified speed. As the HPCl flow increases, the turbine governor valve is automatically adjusted to maintain design flow. Exhaust steam from the WPCl turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the HPCI System during normal operation without injecting water into the RP The ECCS pumps are provided with minimum flow bypass iines, which discharge to the suppression pool. The valves in these lines automatically open (for CS and RHR they are already open) to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects. ali ECCS pump discharge lines are filled with water. The LPCl and CS System discharge lines are kept full of water using the pressure suppression chamber head tank or condensate head tank. The HPCl System is normally aligned to the CST The height of water in the CST is sufficient to maintain the piping full of water up to the first isolation valve. The relative heigi;t of !:e feedwater line connection for HPCl is such that the water in the feedwater lines keeps the remaining portion of the HPCl discharge line full of wate ~

____l.__l (continued?

BFN-UNIT 2

,.-,-

ECCS - Operating B 3. BASES BACKGROUND The ADS (Ref. 4) consists of 6 of the 13 S/RVs. bt is designed (continued) to provide depressurization of the RCS during a small break LQCA if HPCl fails or is unable to maintain required water level in the RPV. ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure ECCS subsystems (CS and LPCI), so that these subsystems can provide coolant inventory makeup. Each of the SlRVs used for automatic depressurization is equipped with one air accumulator and associated inlet check valves. The accumulator provides the pneumatic power to actuate the valve APPLICABLE The ECCS performance is evaluated for the entire spectrum of SAFETY ANALYSES break sizes for a postulated I B C A . The accidents fer which ECCS operation is required are presented in References 5 and 6. The required analyses and assumptions are defined in Reference 7 . The results of these analyses are described in Reference 8.

... This LCQ helps to ensure that the following acceptance criteria for the ECCS, established by 10 CFR 50.46 (Ref. 91,will be met following a LOCA, assuming the worst case single active component failure in the ECCS:

a. Maximum fuel element cladding temperature is i 2200° b. Maximum cladding oxidation is 5~ 0.17 times the total cladding thickness before oxidation; (continuedl BEN-UNIT 2  % 4 5-6 aevlslon 0

. I

ECCS Operating

~

B 3.5.1 BASES AF . ' .CABLE c. Maximum hydrogen generation from a zirconium water SAFETY ANALYSES reaction is s 0.01 times the hypothetical amount that would (continued) be generated if all of the metal in the cladding surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; d. The core is maintained in a coolable geometry; and e. Adequate long term cooling capability is maintaine The limiting single failures are discussed in Reference 13. For a large or small pipe break LOCA and events requiring ADS operation, selected battery failure is considered the most severe single failure. The remaining OPERABLE ECCS subsystems provide the capability to adequately cool the core and prevent excessive fuel damag The ECCS satisfy Criterion 3 of the NRC Policy Statement (Ref. 14).

LCO Each ECCS injectionkpray subsystem and six ADS valves are required to ae OPERABLE. The ECCS injection/spray subsystems are defined as the two CS subsystems, the two LPCl subsystems, and one HPCI System. The low pressure ECCS injectionlspray subsystems are defined as the two CS subsystems and the two LPCI subsystem With less than the required number of ECCS subsystems OBERBBLE, the potential exists that during a limiting design basis LOCA concurrent with the worst case single failure, the limits specified in Reference 9 could be exceeded. All ECCS subsystems and AD§ must therefore be OPERABLE to satisfy the single failure criterion required by Reference (continued)

BFN-UNIT 2 B 3.5-7 Revision 0

ECCS Operating

~

B 3. BAS%S LCO LPCI subsystems may be considered OPERABLE during (continued) alignment and operation for decay heat removal \ h e n below the actual RHR low pressure permissive pressure in MOBE 3, if capable of being manually realigned (remote or local) to the LPCl mode and not otherwise inoperable. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core coaling, thereby allowing operation of RHR shutdown cooling when necessar AFPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3,when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event uf a break in the primary system piping. In MODES 2 and 3:when reactor steam dome pressure is 5 150 psig, ABS and HPCl are not required to be OPERAsbE because the low pressure ECCS subsystems can provide sufficient flow below this pressure. ECCS requirements for MODES 4 and 5 are specified in LCO 3.5.2: "ECCS -

Shutdown."

,-

ACTIONS If any one low pressure ECCS injection/spray subsystem is inoperable, or if one LPCI pump in both LPCI subsystems is inoperable: the inoperable subsystem(s) must be restored to OPERABLE status within 7 days. In this condition, the I

remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced, because a single failure in one of the remaining OPERABLE subsystems, concurrent with a LOCA, may result in the ECCS not being able to perform its intended safety functio (continued)

BFN-UNIT 2 B 3.5-8 Amendment No 259 aevtston Q March 12 20C1

JPM NO. Al.2R REV.NO 0 PAGE 1 O F 5 BROWNS FERRY NUCLEAR PLANT JOBP'ERFORMANCEMEASURE JPM NUMBER: A I .2W TITLE: DETERMINE REGULATORY REPORTING REQUIREMENTS TASK NUMBER: S-000-AD89 VALIDATED BY: . ._ __ DATE:-. -

APPROVED: - - _~ I _ _-- DATE.-.

TRAINING PLANT CONCURRENCE: ~ - BATE:-.

OPE RAT10NS

• Examination JPMs Require Operatiom Training Manager or Desigree Appro\ial and Plant Concurrence

JPM NO. A I 2 R REV.NO 0 PAGE 2 0 F 5 BROWNS FERRY NUCLEAR PLANT JOB PERFORMANCE MEASURE RO __ SRO -XY DATE:

JPM NUMBER: AI2R TASK NUMBER: S-000-AD-89 TASK TITLE: DETERMINE REGULATORY REPORTING REUUIRMENTS K/A NUMBER: 2.4.30 KiA RATING: R6-2.?L SRO: - 3 2 -

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

TASK STANDARD: U sing Tech Spec and available references, determine that condition require entry into TS 3.0.3 and determine per SPP-3.5, Regulatory Reporting Requirements, that a 4-hour report and follow-cp 60 days report are required for a Tech

....- Spec required shutdow LOCATION OF PERFORMANCE: SIMULATOR x - PLANT % -. CONTROL ROOM x.-

REFERENCESiPROCEDURES NEEDED: SPP-3.5 - Regulatory Reporting Requirements VALIDATION TIME: CONTROLROQM: N!A - LOCAL: N!A MAX. TIME ALLOWED: NIA (Completed for Time Critical JPMs only)

PERFORMANCE TIME: NlA CONTROL. ROOM BLA-, LOCAL -I& -

COMMENTS:-

~ .~ .- -

-~ ~ .- . __I.__ ~ - -~

Additional comment sheets attached? YES ,

- NO RESULTS: SATISFACTORY _ _ UNSAT~SFAC'TORY EXAMlNE,R SIGNATURE.: ~ .- -. DATE:

E)(AM INEK

_J.'

.-

~

BROWNS FERRY NUCLEAR PLANT JOB PERFORMANCE MEASURE Unit 2 was operating at 100% rated thermal power when HPCI was declared inoperabie due to a failed area temperature surveillance requirement. Two hours later 2-PCV-1-19, an ADS valve, was declared inoperable due to a failed power suppl You are the Shift Manage INITIATIPJGCUE DeterMine the required action by Technical Specifications as a result of the above condition Examiners CUE (this is not on the student handout)

Once TS 3.0.3 entry is identified, then state "Per your direction, the Unit Supen;iso/

has commenced a shutdown per TS 3.0.3, Determine any Regulatory Repoding Requhments, if any."

Critical Step SAT- UNSAT Student should identify that the combination of HPCI and an ADS valve in Mode 1 is Tech Spec Condition 3.5.1.H which requires entry into TS LCO 3.8.4 Immediatel Reference Ucit 2 Technica! Specification Provide CUE to determine reporting requirements (above)

+ * * t * * i * * * * * * * * * * * * * * * * * ~ * ~ * , * * * * * * * ~ * * * * * ~ * * * * * * ~ ~ , ; * * * , , ~ ~ ~ ~ * * * * * * * * * * * * * + ~ * ~ ~ * ~ + * ~ + * . ~ * ~ * ~ ,

Critical Step SAT___ UNWT 3 Hour Notification to NRC once the shutdowr. is initiated as reqwed Cy I S (Continued on nex? page)

BROWNS FERRY NUCLEAR PLANT JOB PERFORMANCE MEASURE EMI\R[NER'SKEY

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

NON Critical Step SAT- UNSAT 68 Day LER is required upon completion of shutdown, which is defined as entry into :he first shutdown mode as required by the TS condition and required action.

Examiners Note: The immediate notif tion requiiement (4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report) is a responsibility of Operations and the 60 day LER is Site Licensing's responsibility.

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

wrencs?

SPP-3.5, Regulatorj Reporting Req. iremcnts Appensix A 3.1.C. The following criteria require Wur.notlficai:cn: ,72(b)(2)(i) -- The initiation of any rilic,'ear o h ! shutdown required by.the.gl&

Technical SpecificaJjon 3. Licewee Event Reports Awritten report shall be prepared ;n sccordawe wit+ 13 CFR 50.73(a)(i) for i!err:s in !he @&?y regod criteria or Technical Specification. The wpcrt shall be , . . . . . . . . .

Report Criteria ,73(a)(2)(i)fA)- The m m a of a r y nuciear snutclowl required by the pkant's Technical Specificatiofi CANDIDATE'S HANDOUT WEFERENCS

-. . ALLOWED -

INITIAL CONDITIONS

-. .-

Unit 2 was operating at 100% rated thermal power when HPCl was declared inoperable due to a failed area temperature surveillance requirement. Two hours later 2-PCV-1-19, an ADS valve, was declared inoperable due to a failed power supply.

You are the Shift Manager on duty.

__

INITIATING

-. -CUE Determine the required action (if any) by Technical Specifications as a result of the above condition Tennessee Valley Authority I

Page Iof 53 ___.___

N A N STANDARD REGW 4TORY RPORTING Quality Related 5 Yes El No PROGRAMS AND REQUIREMENTS PORC Required OYes El PROCESSES 10CFRSO 59 Review 0Yes 5 No I Effective Date - ~

11/19/2002

_ _ . ^ _ _

EESPONSIELE PEER TEAM: Licensing

.- -

Organization

-- -. Mark J. Burzynski

.- . .- -

10/2&/02 -

• Primary Sponsor Rate Ashok S. Bhatnagar 11113iO2 Peer Team Mentor Date

.- N/A - -

General Manager, KA Cafe Karl W. Sinaer 11/14/02

~ .- - -

Bate

• Senior V;ce President, Nucicnr Opertitiorrs

TQANSTANDARD SPP- PROGRAMS R N D REGULATOKY REPORTlNG REQUIREMENTS Fiev. 4i PROCESSES Page 2 of 53 REVISION LOG Page 1 of 2 Revision Effective Pages Description Number Date Affected of Revision 0 06-30-97 All Initial issue. Replaces STD-4.5, SSP-4.5 (SQN and BFN) and SSP-4.05 (WEN).

1 10-24-97 2, 17, 18 This revision adds: I) a new condition to the regulatory reporting matrix contained in Appendix A and 2) the actual locations of the cooling lower, off-gas stack. and nieteorological tower obstruction lights to Appendix , 4; 8,9, Revised to modify definition of "Safe Shutdown" and make other mhor 13-17, 20 editorial change . 11, 13, Revised Appendix A, "Site Event Notification Matrix" to inc!ude 17, 18,28 additional notification requirements. Revised Appendix A, Sectioz "Immediate Notification - NRC" and Section 3.2, "Twenty-Four Hour Notification NWC" IQ

~ incorporate minor clarifying NRC change (replaced "eye" with "lens') per BIN 3150-AF46 [Ref: FR: 7i23!08 (Volume 63, Number 141)]. Revised Appendix B, "Other Regulatory Reportiny" to add telephone number for the FAA Also corielet: typos on the "NRC Event Notification Worksheet."

4 72/21/98 2, 3, 28, Added Appendix E. "Reporting of Decommissioriing Funding' a d I 29, and 30 corrected typo /17/99 2,12,18, Revised Appendix A , "Reporting of Event Fiela!ed or Con5tiorral 28-30 Reports", to ciatify reporting requirements for follow-up notification Revised Append;x 8, 'Other ReguIatory Reporting", to revise telephone numbers for the FAA. Revised Appendix E, "Repotting of Decornmissioriing Funding", to address requirerner:ts for :iotifying NRC when shutting down the operation of a reactor, as :eqaired b y 10 ,: "i 50.§4(bb).

12ilO/99 2. 40 Added note to clarify requirements for making an ENS nc:ificatior, for events or conditions that are discovered which met the errwger'ry plan criteria but no emergency was declared and the basis for the emergency no longer exist /23/01 2-7, 10-36 Procedure updated to refled changes to 10 CFR 50.72 and 59.1'3 :?'IC a general update to the organization of the reporting gu:da.ic Appendix 8 to provide guidance regarding reporting criterit? f or Conditions Affecting Activities Involving By-product, Scar::?

Special Kuclear ?Ja:eriai Lkenses. Added Appendix C i*iiiiZ iriformation previously included it? Appendix A and fullhel gG regarding expectatinns for notifica?ionof Senior Managerrc:'tt regarding plant eve,.., TVANSTANDARD s PP-3.6 PROGRAMS AND REGULATORY REPORTING UEQLIIREMEMTS Rev. 1 3 PROCESSES Page ifof 53

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Defect (a) A deviation in a basic component delivered for use in a facility, installed, used, or operated if, on the basis of an evaluation, the deviation could create a substantial safety hazard; or (b) the installation, use. or operation of a basic component containing a defect as defined above: (c) A deviation in a portion of a facility subject to the construction permit provided the deviation could, on the basis of an evaluation, create a substantial safety hazard and the portian of the facility containing the deviation has been offered to the purchaser for acceptance; or (d) A conditiori or circumstance involving a basic component that could contribute to the exceedirrg of a safety limit as defined by the plant operating technical specification Department Level Manager - Any manager who functionally or administratively reports directly to the site vice president or plant manage Deviation - A departure from the technical or quality assurance requirements included in a procurement document. safety analysis report, coristrt~ctioripermit or other documents provided for basic component Evaluation - T h e process of determining whether a particular deviation could create a substantial safety hazard or deterriiinirg whether a failure to comply is associated with a substantial safety hazar .

Event Any cccurrence surrounding un:t operatio External Conditions -Events created by thicgs outside the desiyn features of the plan Government Agencies - A n acjency of the Federal governrrlent as defined in 10 CFR 5 Incident Investigation - Process conducted by the KRC for the purpose of accident preventio The process inCIlJdeSgat3eriny and ana!yzing information, determining findings and cc~wlusion including the cause(s) of a significant operatjoGal event; and the disseminating of the investigation results for NRC, industry, and public revie Initiation of Shutdown - Physical act of redecing power or temperature to change moces Invalid Actuation (Signal) -Signals that do not meet the criteria for being valid. invalid actuations include instances where instrument drift, spuriow signals, human error or other invalid signals that result in manual or alitornatic actuation of the systems listed in !O CFR 50,73(a)(Z!)(iv)(B).

Major Loss of Cornniunication - Consti!utes t 3 e loss of ccrnmunication capabilities Major Deficiency - A condition or circumstance which under normal operating conditions, an anticipated transient. or postulated design basis accident could contribute to exceeding a saf?ty limit or came an accident. "Major deficiency" also means a condition or circumstance which in the event of an accident due to other causes c.oulif. considering an independent single failure, result in a loss of safety function necessaiy to niitiyate the consequences of the acciden Natural Phenomenon - Act of nature (e g., f're. flood, tarnado)

Mews Release -Known items which may be distributed to the media (UPI, television, :ad: mwspaper, etc.) and those items identified to be going on TVA news tape distribl;ted by the TVA Public Affairs Staf TVANSTANDAWB SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 1t PROCESSES Page 12 of 53 Noncompliance (Failure To Comply) - A noncompliance for the purposes of this procedure means any failure to comply with the Atomic Energy Act of 1954, as amended, or with any applicable rule or regulation of the MRC relating to substantial safety hazards. A noncompliance may be in operations, engineering, or construction of the farjlity or basic component thereQ Organization Manager -This is the most senior rriariager available who is in the same organization as the individual who discovered the abnormal event. The senior manager is not normally interpreted to be the plant manager or site vice presiden Preplanned Sequence Part of an approved procedure, including warkplans, work request, work orders, surveillance instructions, general operating instructions and system operating instruction Prevented The Fulfillment Failure or possible failure of a safety system to properly c o ~ e l e t g a safety functio Principal Safety Barrier - Fuel cladding, RCS pressure boundary. or the containment Redundant Equipment -Equipment, systems, structures capable of perforrr;iny the same intended function within the same Technical Specification allowable values. (In most cases, this

.

means opposite train equipment.)

Safe Shutdown - Mode 3, as defined by the Techrrical Specific-t' d lolls Safety Function - A component or structure designed to actuate upon receiving the pr3per signal (ESF or RPS).

Significant Operational Event - Any radiological, safeguards, or other safe!y-related operational event at an NRC licensed facility :hat poses an actual or potential hazard to !!le public health and safety, property, or the environment. These everits or those that typically result in a 1OCFR50.72 immediate nctifica!ion (See Appendix A of this procedure) A significant operational event also may be referred to as "an incident". Examples of these events inclu":

operations that exceeded. or were riot included in the design basis of the facility, a major deficiency in design, construction, or operatior having potectial yeneric Safety implications, a significant loss of integrity of the fuel, the primary coolant boundary, or the p r i n a v containment boundar a 105s of safety function or multiple failures in systems used to mitigate an actual e:,t:nt significant unexpected system interactions, repetitive failures or events involving safety related equipment or deficiencies in operatio questions or concerns peeaining to licensee opera!ional perromanc Substantial Safety Hazard -Loss of safety function to the exlent that there is a major reilxt Lvl in the degree of protection provided to public health and safety for any facility or activity license Threat Physical hazard (e.g., fire. severe rad

~

W A N STANDARD SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 11 PROCESSES Page 13 of 53

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Unanalyzed Condition Plant Condition outside the bounds of the initial conditions as described in the FSAR accident aiialysi Valid Actuation (Signal) - Signals that are initiated in response to actual plant conditions or parameters satisfying the requirements for initiation,

.

TVANSTANDARD SPP- PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. $ 4 PROCESSES Page 44 of 53 APPENDIX a Page 1 of 10 REPORTING OF EVENTS OR CONDlTlONS AFFECTING LICENSED NUCLEAR POWEK PLANTS PURPOSE This Appendix identifies reporting requirements; and instructions for determining reportability, preparation, and transmittal of LERs; and notification to NRC for events occurring at TVA's licensed nuclear plant .0 ScCIpE TVA is required by 10 CFR 50.72 and 50.74 to promptiy report varicus types of conditions or events and provide written follow-up reports. as appropriate. This appendix provides repotting guidance applicable to licensed power reactor NOTE Appendix 6 provides addi!ioia! reporting criteria found in 10 CFR Parts 20, 30, 40, and 70 that may be applicable to events involving byproduct, source or special nuclear 1 material possessed by the licensed nuclear plant. Site Licensing and Site RadCon are responsible for making the repc:taSi!ty determinations for 10 CFR Part 20, 30, 40. or 70 events associated with their site. Corporate Licensing and Corporate RadChem are responsible for making the reportability determinations for 10 CFR Part 20, 30, 40, or 70 events associated with ail other TVA licensed activities. Licensing is responsible for developing (with input from affec!e3 or~i:ini7a!icns) and submitting the immediate notification and wi..:en reports lo N?C in accordance with 10 CFR Part 20, 30,40, or 70 requirements. Reporting require!nen:s for ?ersanrrel exposure required by :O CFK Fart .

20 are contairied in RCD?-4. "Personnel Iriprocessing and Dosimetry Administraiive

,. -- Processes." -

Appendix C contains the cri!eria fcr re9cli:lg if events or conditicns affecting ISFS TVA. as the general licensee of ttie ICFS!. is required by 10 CFR 72.216 !o make initial and written reports in accordance with 1 C CFR 72.74 and 10 CFR 72.75. Operations is responsible for making the repaitability determinations for :O CFR 72.74 an5 10 CFR 72.75 reports. Operations is respcnsible for rnakirig the immediate notification to NRC ir?

accordance with 10 CFR 72.74. Cperations is responsible for makicg the immediate, 4-hour, and 24-hOur notifications lo NRC in accordance with 10 CFR 72.75. Licensing is responsible for developicg (with kput from affected organizations) and submitting the written reports required by 10 CFR 72.7 NOTE Repofling requirements for events or condi:ions affecting the physical protection O f the licensed nuclear plant specified in 10 CFR 73.71 are contained in SPP-I .3 *Plant A-W P S S and Security." Responsibilities for reportability de!erminations and immediate notification requlrements are assigned to Site Nuclear Security and Corporate Nuciew Security. Licensing is responsible for developi;ly (with input from affected orgarliza and submitting the written reports required hy 10 CFF? Part 73.71 REQUIREMENTS

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NOTE Internal management notifica:ion requirerients f3r plant events are found in Apl?endix D Operations and the Plant Manager (or Duly Plant Ma:iager) are responsible for makitla these internal rnanngement notificatiorls.

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W A N STANDARD SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. I 1 PROCESSES Page I 5 of 53 APPENDIX A Page 2 of 10 NOTE NRC NUREG-1022, Supplements arid subsequent revisior,s should be used as guidance for determining reportability of plant events pursuant to 10CFR50.72 and 10CFR50.7 .j &mediate Notification - NRC TVA is required by 10 CFR 50.72 to notify NRC immediately if certain types of everits occur. This appendix contains !he types of events and the allotted time in which NRC must be notified. (Refer to Form SPP-3.5-1). Operations is responsible for making !he reportability determinations for 50.92 arid 50.73 reports. Operations is responsible for making the immediate notification to NRC in accordance with 10 CFR 50.7 Notification is via the Emergericy Notification System. If the Emergency Notification System is not operative, use either a telephone, telegraph, mailgram, or facsimil NOTE The NRC Event Notification Worksheet may be used in preparing for notifying

_ I -

the NR . The immediate Notification Criteria of 10 CFR 50.72 is divided into I-hour, 4-hour, and 8- hour phone calls. Notify the NRC Operations Center within !he applicable time limit for any item which is identified in the Immediate Notifica!ion Criteri The following criteria require l 4 o u r no!ific.ation: (Technical Specifications) - Safety Limits as defined by the Teckn'ca'

Specifications which have been violate . 50.72 (a)(l)(i) The dec!aration of any of the Emergency classes

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specified iri the licensee's approved En;e:gency Pla MOTE If it is discovered that a condition existed which rnet the Emergency P'a9 criteria but no emergency was declared and the basis for the emergency class no longer exists at the time cf discovery. an ENS notification (and notification of the Operations Duty Specialist), within one hour of discovery of the undeclared (or rnisclassified) event. shall be mad However. actual declaration of the emergency class is not necessaly iii these circumstance . -

50,72(b).(I)) Any deviation from !he plant's Technical Specificatiilrls authorized pursuant to 10 CFS 50.54(x The following criteria require 4-hour nctiflca!!cn: ,72(b)(Z)(i) - The initiation of any nuciear plant shutdown required 13)

the plant's Technical Specification W A N STRNBARD SPP-3.5 PROGRAFAS AND REGULATORY REPORTING REQUIREMENTS Rev. I f PROCESSES Page 16 Of 53 APPENDIX A Page 3 of 10

W A N STANDARD SPP-3.5 PROGRAMS A N D REGULATORY REPORTING REQUIREMENTS Rev. I 1 PROCESSES Page 17 of 53 APPENDIX A Page 4 of 10 (4) ECCS for boiling water reactors (BWRs) including: core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal syste (5) BWR reactor core isolation cooling system (6) PWR auxiliary or emergency feedwater system (7) Containment heat removal and depressurization systems, including containment spray and fan cooler system (8) Emergency ac electrical power systems, including: Emergency diesel generators (EDGs). .72(b)(3)(v) Any even! or condition that at the time of discovery could have prevented the fulfillment of the safety function cjf structures or systems that are reeded to:

(A) Shut down the reactor and maintain it in a safe shutdown'

condition; (B) Remove residual heat:

(6) Control the release of radioactive material; or (D' Mitigate the consequences of ari acciden NOTE According to 5G.72 (b)(3)(vi) events covered by 50.72(bj(3jjv)

may include one or more procedural errors, equipment faiigl-es and/or discbery of design. analysis, fabrication, construclior,.

andlor procedurai inadequacies. However, individual component failares need not be reported pursuafit this paragraph if redundant equipment in the same system was operable and available to perform the required safety func!;o . 50,72(b)(3)(xii) - Any event requiring the transport of a radioactively contaminated person to an offsite medical lacility for treatmen . 50,72(b)(3)(xiii) Any event that results in a major loss of emergency

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assessment capability, offsite response capability, or offsite communications capabiiity (e.g., significant portion c j f con:rol room indication, emergency notification system, or offsite notification system). Follow-up Notification (50.72(cjj With respect to the telept1or.e notifications i%de under paragraphs (aj and ct!i

[50.42 (a) and 58.72 (b), respectively] 3f this section [50.72]. in addition 13 rnakiiig the required initiai n3:ifization. during the course of the even!:

NANSTANDARD SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQULREMENTS Rev. 11 PROCESSES Page 18 of 53 APPENDIX A Page 5 of 10 Immediately report (i) any further degradation in the level of safety of the plant or other worsening plant conditions including those that require the declaration of the Emergency Classes, if such a declaration has not been previously made; or (ii) any change from one Emergency Class to another, or (iii) a termination of the Emergency Clas Immediately report (i) the results of ensuing evaluations or assessments of plant conditions, (ii) the effectiveness of response or protective measures taken, and (iii) information related to plant behavior that is not understoo Maintain an open, continuous communication char:nel with the NRC Qperetions Center upon request by the NR .2 -

Twe n k - -

F ou r KO- ur NotificEn-- 1 Any violation of the requirement contained in specific operating license conditions, shall be reported to NRC in accordance with the license conditio .3 Two-Bav Notification - NRC

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50.9(b) The NRC shall be rictified of incomplete.or inaccurate infomation wkich contains significant implications fnr the public health and safety or common defense a:ld security. Notificatiorl shill he p-ovided t o the administrator of the appropriate regional oftice within two working days of identifying the information. Licensing is responsible for determining reportahility (with input from affected organiza!ions) and notifying NRC in accordance with 10 CFR 5 .4 Sixty-Bay Verbal ReDort 50,73(a)(Z)(iv)(A) requires that any w e n t or co:idition that resulted in inanual 01 automatic actuation of the specified systems be reported as a Licensee Event Report (LER (Refer to Appendix A, Section 3.51). This CFR section also allows that ill the case of an invalid actuation, other than actuation of the reactor protection system when the reactor is critical, an optional teiephcne r.o!ification may be placed to the NRC Operations Center within 60 days after discovery of the event imtead of scbtnittiiiy a written LE Verbal Report Required Content:

If the verbal notification option is selected (NUREG 1022, Revisicn 2 , Sec.ti 32.6.. "System Actuation"), instead of a LER. the verbal report: Is not considered ar: LER Shoiild identify !hat the report is being made Lxder 50.73(a)(2)(iV)jk)

TVANSTANDARD SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. f 3 PROCESSES Page 19 of 53 APPENDtX A Page 6 of 10 Should provide the following information:

(a) The specific train@) and system(s) that were actuated (b) Whether each train actuation was complete or partial (c) Whether or not the system started and functioned successfully N(STE Licensing will ellsure that the information that is provided to NRC during the Sixty-Day Verbal Report is verified in accordance with EP-21 Verbal Report Development and Review Licensing will: Develop (with :npl;t from responsible organization) the response (i.e.,

report summary) to address the required inpu . Ensure that the reporting W a i l s are reviewed by MRC Telephone Reporl Tirneliness Licensing "till make the 60-day telephone report promptly after the PER for the invalid actuatio!? event is reviewed by MR .5 -

Written Report N&C_ A report on a Safety Limi?Violstion shall be submitted to the NRC, the NSRB, and the Site Vice Fresident if required by Technical Specification . Any violation of the requirements contained in the Operating license conditiorW in lieu of other reporting requirements requires a written follow-up report if specified in the licens Reporting Radiation Injuries 10 CFR 140.6(a) requires, as promptly as possible, submittal of a written notice

[e.g.. report] in the event of: Bodily injury or property damage arising out of or in connection with tt!.?

possession or use of !hc radioactive material at the licensee's faciiity

[location]; or In the course of transpcjrtation; or In the event any radiatior? exposure claim is made. (Refer to RCDP-9

• Radiological ar,d Chemistry Control Radiological Exposure InqtiirPS"?

The written notice shall contain prticala:'s sufficient to identify the licensee and reasonably obtainable inforrnatic~r:with respec! to time, place, and circamstar - .

thereof, or the nature of the tiai W A N STANDARD SPP-3.5 PROGRAMS AND REGULATORY REPOKTLNG REQUIREMENTS Rev. I$

PROCESSES Page 25 of 53 APPENDIX A Page 7 of 10 Licensee Event Reports A written report shall be prepared in accordance with 10 GFR 50.73(a)(i) for items in the GO-day report criteria or Technical Specifications. The report shall be complete and accurate in accordance with the methods outlined in this procedure. The completed forms shall be submitted to the USNRC, Document Control Desk, Washington, DC 20555. NUREG 1022, Revision 2 , contains the instructions for conipletion of the LER form. Licensing is responsible for deve!oping (with input from affected organizations) and submitting the written reports (or optional telephone reports (refer to Appendix A, Section 3.41) required by 10 CFR 50.7 NOTE Unless otherwise specified in ?he reportirig criteria below, an event shall be reported if it occurred within three years of the date of discovery regardless of the plant mode or power level. and regardless ofthe significance of the structure, system, or component that initiated the even Report Griteria .73(a)(2)(i)(A) - The completion of any nuclear plant shutdown required by the plar.t's Technical Specification . 50.73(a)(2!(i)(B) - A n y operation or condition which was prohibited by +he plant's Technical Specifications, except when: The T e c t i n i h Specification is administrative in nature; The event consisted solely of a case of a late surveiilance !c:st where the oversight was corrected, the test was performed. and the equipment was found to be capable of performing its specified safety functions; or The Technical Specification was revised prior to discovery ot !he event such that the operation or condition was no longer prohibited at the time of discovery of the even . 50,73(a)(Z)(i)(C) - Any deviation from the piant's Technical Specifications authorized pursuant to IO CFR 50.54(x). ,73(a)(2)(ii)(A) - Any event or condition that resulted in the conditic1r. c f the nuclear power plant, inclliding its principal safety barriers, being seriously degrade . 50,73(a)(Z)(iij(B) - Any event or condition that resiilted il! the IluAeJ:

power plant being in an unanalyzed condition that significantly i:egrai:ieG plant safet B V A R I S i , 4ARD SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 111 PROCESSES Page 21 of 53 APPENDIX A Page 8 of 10 .73(a)(Z!(iii) Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the sate operation of the rruelear power plan . 50,73(a)(2)(iv)(A) - Any event or condition that resulted in manual o i automatic actuation of any of the systems listed in paragraph (a)(a)(ivj(B) [see list in item no. 8 below], except when (1) the actuation resulted from and was part of a pre-planned sequence during testing or reactor operation: or (2) The actuation was invalid and (i) occurred while the system was properly removed froni service or (ii) occurred after the safety function ilad been already complete NOTE

. I t i the case of an iwalid actuation, other than actiiation

.

o f the reactor protection system (RPS) when the reactor is critical, a telephorie notification to the NRC 3perations Center within 60 days after discovery of the event may be provided instead of submitting a written E R (1 3 CFR 50.73(a)). [Refer to Appendix A, Section 3.41 .73(a)(2)(iv)(B) - The systems to which :he requirements to paragraph (a)(Z)(iv)(A) of this sxtiori apply are:

(1) Reac!rr protection system (RPSj inciuding: reactor scram cr ;eactor tri (2) General containment isolation signals affecting contain-nent isolation valvPS in more than one SyStEm 07 multiple main steam isolatidii valves (MSIVs).

(3) Emergency ccre cooling systems (ECCS)for pressur:zed water reactors (PWRs) including: high-head intermediate-head, and low-head injection systems and the low pressure injec!ion function of residual (decay)

heat removal system (4) ECCS !or boiling water reactors (BWRs) inciuding: core spray systems; high-pressure coolant injection s p t m low pressire injec:ion function of the residua! heat removal systeir (5) BWR :eactor core isoiaticn cooling systei?,

TVANSTANDARD SPP-3.5 PROGRAMS AND R E G U U T O R Y REPORTING REQUIREMENTS Rev. I 1 PROCESSES Page 22 of 53 APPENDIX A Page 9 of 10 (6) PWR auxiliary or emergency feedwater system (7) Containment heat removal and depressurization systems, including containment spray and fan coolei system (8) Emergency ac electrical power systems, including:

emergency diesel generators (EBGs).

(9) Emergency service water systems that do not normally run and that serve as ultimate heat sink . 50.73(a)(2)(~)- Any event or condition that could have prevented the fulfillment of the safely function of structures or systems that are nceded to:

(A) Shut down the reactor and maintain it in a safe shutdown corrditian; (B) Remove residC.i;l heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an acciden ___

NOTE Events reported above may include one or more procedural errors, equ-pment failures, and/or discovery of design. analysis fabrication, construction, andior procedural inadequacie However, individuai component failures need not he reported pursuant to this criterion if redundant equipment in the same system was operable and available to perform the required safety functioq (50.93(a)(Z)(vi!!

1 ,73(a)(2)(vii) - Any ever.1 where a single cause or condition Caused at least one independent train or channel to become inoperable in nlulticle systems or two independent trains or channels to become inoperable in a single system designed to:

(A) Shut down :he reactor and main!ain it in a safe shutdown condition; (3) Remove residual heat; (C) Control the release <Ifradioactive materid 01 (D) Mitigate the consequences of a n acciden NANSTAMDARB SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 11 PROCESSES Page 23 of 53 APPENDIX A Page 10 of 10 1 .73(a)(2)(viii)(A) - Afly airborne radioactivity release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in appendix B to part 20, table 2, colum1n i .73(a)(2)(viii)(B) -Any liquid effluent release that, when averaged over a time period of Ihour, exceeds 20 times the applicable concentrations specified in appendix E? to part 20, table 2, column 2, at the point of entry into the receiving waters (i.e.. unrestricted area) for all radionuclides except tritium and dissolved noble gase . 50.73(a)(Z)(ix)(A) - Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

(1) Shut down the reactor and maintain it in a safe shutdowt?

condi:ion; (2) Remove residual heat; (3) Control the reiease of radioactive material; 01 (4) Mitigate the cmseqmnces of an acciden Events covered above may include cases of procedural erro equipment ?ailiire, and/or discovery of a design, analysis, fabrication, constrtlctisn, andior prccedural inadequac However, licensees are not required to report an event pursuant to this criterion if the event results from a shared dependency among trains or channels that is a natural or expected consequence of the approved plant design or normai and expected wear or degradatiotl [50,73(a)(Z)(ix)(B)].

14 50.73(@(2)(x) -Any event that posed an actual threat to the safety 0 : :!le nuclear power plant or significantly hampered site personnel in the nerfnrrnance of .. duties necessaw for the safe operation of the nuClCar

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power plant including fires, toxid gas releases, or radioactive release W A N STANDAKD SPP- PROGRAMS AND 6(EGULATORY REPORTING REQUIREMENTS Rev. I f PROCESSES Page 24 of 53 APPEGDIX B Page 1 of 5

'EPORTING OF EVENTS OR CONDITIONS AFFECTING ACTIVITIES INVOLVING BYPRODUCT, SOURCE OR SPECIAL NUCLEAR MATERIAL LICENSES PURPOSE This Appendix identifies reporting requirements; arid instructions for determining reportability, preparation, and transmittal of written reports. and notification to NRC for events affecting TVA activities governed by NRC byproduct. source or special nuclear materia! license TVA is required by its various NRC licenses to report certain events or conditions. 10 CFR Part 20 contains reporting requirements for events involving licensed byproduct, source, or special nuclear material. 10 CFF? 30.50 contains reporting requirements for events involving licensed byproduct material. 10 CFR 40.60 contains reporting requirements for events involving licensed source material. 10 CFR Part 70 contains reporting requirements for events and conditions involving licensed special nilciear material. This procedure contains the reporting requirements for these activitie . REQUIREMENTS NOTE Internal manayement notification require men?^ for events reported to N X are fourid in Appendix D. Operations and the Plant Manager [or Duty Plant Manager) are responsible for making these i,.ternal managemerit notification .1 Immediate Notification NRC

.

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TVA is required by the various byproduc:, so:irce or special nuciear materia! licenses to notify NRC immediately if certain types of events or conditions occur. This appendix contains the types of events and the ailo!ted time in which NRC must be notifie Site Licensing and Site RadCon are responsible for rnakiny the reportabi!ity determinations for 10 CFR Part 20, 30, 40, or 70 events associated with their sit Corporate Licensing and Corporate RadChem are responsible for making the reportability determinations for 10 CFR Part 20, 30,40, or 90 events assoc,ia!ed with all other TVA licensed activities. Licensing is responsible for making the immediate notification and developing (with input from affected organizations) and submitting written reports to NRC in accordance with :O CFR Part 20, 30, 40, or 70 requirement NOTE- Reporting requirements for personnel exposure required by 10 CFR Part 20 are contained in RCDP-4, "Personcel Inprocessing and Dosimetry Administrative Processes."

Notification should be made to the NRC office identified in the specific repoiling regulation.

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W A N STANDARD SPP-3.5 PROGKAMS AND REGULATORY REPORTING REQUIREMENTS Rev. i1 PROCESSES Page 25 of 53 APPENDIX B Page 2 of 5 The following criteria require immediate notification: lOCFR20.1906(d)(l) Upon discovery, any removable radioactive surface contamination that exceeds the limits of 10CFR71.87(1). OCFR20.1906(d)(2) - Upon discovery, any external radiation leveis that exceed the limits of 10CFR71.4 . 20.220I(a)(l)(i) - Upon discovery, any lost stolen, or missing licensed material has occurred in an aggregate quantity eqiiai to or greater than 1,000 times the quantity specified in 10 CFR Part 20: Appendix C under circumstances that appears that exposure could result lo persons in unrestricted area . 20.2202(a)(1) - Any event involving byproduct, source, or speclai nuclear material possessed by the licewee that may have caused or threatens lo cause an individual to receive a total effective dose

.

equivalent of 25 rems or more; a lens dose equivalent of 75 rems or more; or a shallow d x e equivalent to the skit1 or ex',:emities of 250 rads or mor . 20.2202(a)(2) - Any event involving byproduct, source, ar special nuclear material possessed by the licensee that may have caused or threatens lo cause the release o f radioactive material, inside or oc!ts.de of a restricted area, so that. had an individua; been present for 24 ilwr '

the individual could have received an intake five times the occupatior?ai annual limit on intaf The following criteria require I-hour notification: .52(a) - A n y case of accidental criticality and any loss, other than normal operating loss, of special nuclear materia . 70.52(b) - A n y loss or theft or unlawful diversion of special nuclear material which the licensee is licensed to possess or any incident in which an attempt has been made or is believed to have been made !O

~ o m n i ial theft or unlawful diversion of such materia!. The fOlbwing criteria require 4-hOUr notification: .50ja) and 40.6G(a) and 70.50(a) - Upon discovery, any event involving licensed byproduct, source or special nuclear material !53t prevents immediate protective actions necessary to avoid exposures ;t radiation or radioactive material that could exceed regulatory iirrits 2 ' '

releases of material that could exceed regulatary limit W A N STANDARD SPP-3.5 PROGRAMS AND R,EGULBTORY REPORTING REQUiREMENTS Rev. il PROCESSES Page 26 of 53 APPENDIX B Page 3 of 5 13. The following criteria require 24-hour notification: .2202(b)(1) Upon discovery of the event, report any event involving

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loss of control of licensed material possessed by the licensee that may have caused, or threatens to cause, an individual to receive, in a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a total effective dose equivalent exceeding 5 rems, a lens dose equivalent exceeding 15 rems, or a shallow-dose equivalent to the skin or extremities exceeding 50 rems (0.5 Sv). .2202@)(2) Upon discovery of the eveat, repott any event involving loss of control of licensed material possessed by the licensee that may have caused, or threatens to cause the release of radioactive material, inside or outside of a restricted area, so that. had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the individual could have received an intake in excess of one occupational annual limit on intske (the provisions of this paragraph do not appIy to locations where personnel are not normaily stationed during routine operations, such as hot-cells or process

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enclosures). .50(b) and 40.60(bj and 70.50(b) - Upon discovery of ar.y of the following w e n t $ involving licensed material: An unplanned contamination event that requires access to the contaminated area. by workers or the public, to be restricted for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by imposing additional radiological controls or by prohibiting entry icto :tie area a d_involves a quantity of material greater than five !inies the lowest annllai limit on intake specified in Appendix E of $$20.lC01 - 20.2401 of 10 CFR part 20 for the materiai has access to the area restricted for a reason other than to allow isotopes with a half-life of Cess than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to decay prior to decontaminalio An event in which equipment is disabled or fails to function as designed when the equipment is required by regulation or license condition to prevent releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences c f an accident 301 the equipment is required :o be available and operable when it is disabled or fails to functierl R O redandant equipment is available and operable to perform the reqtiired safety functio An event that requires unplanned niedical treatment at a medical facility of :in individual with spreadable radicactive contamination or? the i:idividual's clothing 0: bod W A N STANDARD SPP-3.5 PUBGRAMS AND REGUUTQUY REPOUTING REQUIREMENT$ Rev. I 1 PROCESSES Page 27 af 53 APPENDIX B Page 4 of 5 An unplanned fire or explosion damaging any licensed material or any device. container, cr equipment containing licensed material when the quantity ~f material involved is greater than five times the lowest annual limit on intake specified in Appendix

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B of 9j20.1001 20.2401 of 10 CFR Part 20 for the material the damage affects the integrity of the licensed material or its containe .2 -

Verbal Report NRC The following criteria required 30-day notification: .2201 (a)(ii) - Within 30 days after the occurrence of any !os Stolen. or missinrJ licensed material becomes known, verbally report all licensed material in a quantity greater than I O times the quantity specified in Appendix C to Part 20 that is still missing at this tim .

3.3 Written Repa@2..N Week Repoil 10CFR20, Appendix G(III)(E) - k written report is required witbin 2 weeks of cornpleticn of the investigation of any shipment or part of a shipment for which ac,knowledgment is not received within the t i r e s set forth in 13CFR23 Appendix G.This investigaticn must: Be petformed by the shipper. if the shipper has not received notification of receipt within 23 days after transfer, Be traced and reported. The investigation shail include this shipment ard filing a report with the nearest Corrirrission Regional Office listed in 1QCFR20 Appendix Day Report A written report is required within 30 days for the following items: .2201(5) - Events reported in acccrdance with 10 CFR 20.2201 (a). .2203(a)(:) - Events repor:ed in accordance with 10 CFR 20.220 . 20.2203(a)(2) - Doses in excess of any of the following the occupational dose 1ir:iits for ad:tlts in s20.1201, or the occupational dosr lirllits for a minor in $20 1207. 01 the IitlliIS '(I1 a n embryok!us o t a deciared pregnant woman in $20.1208 2 C the limi!s for an irldivitlgal member of the public in S20.!3CI, 2'

any applicable lirnit in thc license or the A L A W cols::ain!s f 0 i air emissions es:ablist:r?d under $20.11Cll(d)

W A N STAMDARD SPP-3.6 PROGRAMS AND REGULATORY REP0RTING REQUIREMENTS Rev. I?

PROCESSES Page 28 of 53 APPENDIX B Page 5 of 5 .2203(aj(3) - Levels of radiation or concentrations of radioactive rrraterial in a restricted area in excess of any applicable limit in the license an unrestricted area in excess of 10 times any applicable limit set forth in this part or in the license (whether or not involving exposure of any individual in excess of the limits in §20.1301) .2203(a)(4) - For licensees subject to the provisions of EPA's generally applicable environmental radiation standards in 40 CFR part 190, levels of radiation or releases of radioactive material in excess of those standards. or of license conditions related to those standard . 20.2204 - Any planned special exposure conducted in accordance with §20.1206, informing the Commission that a planned special exposure was conducted and indicating tt;e date the plaoned special exposure occurred and the information

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required by $20.210 . 30.50(~)(2)- Events reported in accordance with 10 CFR 30.50(a) and 30.50(b). .60(c)(2) - Events reported in accordance with 10 CFR 40.60.a and 40.60(b). .50(c)(2) - Events reported in accordance wi!h 10 CFR 70.50.a and70.50)b).

The written report will contain the information specified by the appropriate regula?ion. The report shall be complete and acca:ate in accordance with the rriethoas outlined in this procedure. The cornpietcd forms shall be submitted to the address identified in the specific regulatio Day Report A written report is required within 90 days for the follcwing items: .34(b) and 40.46 and 70.36 Notification of intent to trc;nsfer

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ownership or control of licensed activities shalt be m2de 90 : 3 W prior to the proposed action. (Reference: NRC information Notice 89-25, Revision 1).

W A N STANDARD SPP- PROGUAMS AND REGULATORY REPORTING UEQLIIUEMENTS Rev. ii PROCESSES $age 29 Qf 53 APPENDIX C Page 1 O f 5 REPORTING OF EVENTS OR CONDITIONS AFFECTING INDEPENDENT SPENT FUEL STORaGE INSTALUTiON (ISFSI) PURPOSE This Appendix identifies NRC reportirig requirernerts; and provides instructions for determining reportability, preparation, and transmittal of writteii reports; and notification to NRC for events occurring at spent fuel storage installations at TVA's licensed nuclear plants. SCOQE TVA. as the general licensee of the ISFSI is required by 10 CFR 72.216 to make initial and written reports in accordance with 10 CFR 72.74 and 10 CFR 72.75. 10 CFR 72.74 and 10 CFR 72.75 requires the general licensee to promptly report various types of conditions or events and provide written follow-up reports, as appropriate. This appendix provides reporting guidance applicable to spent fuel storage installations at licensed power reactors. REQUIREMENTS

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This Section contains the types of events and the allotted time in which NRC must be notifie Operations is responsible for making the reportability deterrninations for 18 CFR 72.74 and 10 CFR 72.75 reports. Operations is respon!jible for rnaking the immediate notification to NRC in accordacce with 10 CFR 72.74 and 10 CFK 72.75. and is respoi?sible for making the four-hour and twenty-four hour 10 C F 7 72.75 report Notification is via the Emergency Notification System. If the Emergency Notification Syste:ri is inoperative, make the required notification "i,a commercial telephonic service or any other dedicated telephonic system or any other method that wiii ensure that a report is received by the NRC Operations Center within the required tirnefrarn .1 -

~ m r r t @ e l i a t e ~ ~ ~ f i c aNRC tion TVA is required by I C CFR 72.74 and 77 75 to notify NRC imniediately if ceitai:? types of events occu The Immediate Notification Criteria of 70 CFR 72.74(a) require the licensee to notify the NRC Operations Center within 1-hour. The following criteria require hour notification: Discovery of accidental criticality. or Any loss of special nuclear material The Immediate Notificatior! Critcria of 10 CFR 72.75(a) require the licensee ?n notify the NRC Operations Center within 1-hour. The following criteria require 1-hour notification: Emergency Notifications - The declaration of an emergency as specified in TVA's approved emergency Glan addressed in 10 GFK 72.3 TVANSTANDAHD SPP- PROGRAMS AND REGULATORY REPQRTING REQUIREMENTS Rev. i t PROCESSES Page 30 of 53

.. APPENDIX C Page 2 of 5 our Non-Ernerqencv EoIficatioo - NRC TVA is required by 10 CFR 72,75(b) to notify NRC as soon as possible, but no Cater than four hours after the discovely of any of the following events involving spent fuel, high level radioactive waste, or reactor-rela!ed Greater than Class C waste: CFR 72.75(b)(l) - An event that prevents immediate actions necessa:y to avoid exposures to radiation or radioactive materials that could exceed regulatory limits or releases of radioactive materials that could exceed regulatory limits (e.g., events such as fires, explosions, and toxic gas releases). U CFR 72.75(b)(2) - A defect in any storage structure, system, or component which is impoltant to safet CFR 72.75(b)(3) - A significant reduction in the effectiveness of any storage confinenient system during us . CFR 72.75(b)(4) - A n action taken in an emergency that departs from a condition or a technical specification contained in a license or certificate o f compliance issued under this pait when the action is immediately needed to protect the public heaIth and safety and no action consistent with license or certificate of cornpliarce coriuitions or technical specifications that can provide adequate or equivalent protection is irnrriediately apparen CFR 72.75(b)(5) - A n event that requires unplanned medical treatment at. an offsite medical facility of an itidividilal with radioactive con!amination on the individual's clothing or body which could cause further radioactive contaminatio CFR 72.75(b)(6) - A n unplanned fire or explosion damaging any spent filel, high level radioactive waste, andior reactor-related Greater than Class C wastc or any device, container, or equipment containing spent fuel, high level radioactive waste, andicr reactor-related Greater than Class C waste when the damage affects the integrity of the material or its containe .3 -

Twentv Four Hour N o n - E n a e r s e n c y _ ~ ~ ~ ~ a t iNoWnC TVA is required by 10 CFR 72.75(c) to notify NRC within twenty four hours after the discovery of any of the following events involving spent fuel, high level radioactive waste, or reactor-related Greater than Clsss C waste: CFR 72.75(~)(1)- Ar,y unplanned contamination event that requires access tL'

the contarriinated area by workers or the public to be restricted for mole than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by imposing additional radio!ogical controls or by prohibiting eritty intc the are W A N STANDARD SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 11 PROCESSES Page 3.1 Of 53 APPENDIX C Page 3 of 5 CFR 72.75(c)(2) - A n event in which safety equipment is disabled or Fails to function as designed when: CFR 72.75(c)(2)(i) - The equipment is required by regulation, license condition, or certificate of compliance to be available and operable to prevent releases that could exceed regulatory limits, to prevent exposures to radiation or radioactive materials that could exceed regulatory limits, or to mitigate the consequences of an accident; and CFR 72.75(c)(2)(ii) - No redundant equipment was available and operable to perform the required safety functio .4 Information R e w i r e d Duiiriglni&al NotificafLo~n- NRC 10 CFR 72.75(6)(1) requires that the nctifications made under 10 CFR 72.75(a), (b), or (c) [Section 3.1 .B, 3.2, or 3.3 of this Appendix] by telephone to the NRC Operations Center provide. to the ex?ent that the information is available at the time of notification, the following information: The caller's name and cail hack telephone number; A description of the event, including date and time; The exact location of the event; The quanlities, and chernic8 aiid physical forms of the spent fuel, high [eve; radioactive waste, or reactor-related areater than Class C waste invo!ved; and Any personnel radiation exposure dat .5 Thirty Dav Written Report NRC ~

10 CFR 72.75(d) requires submittal of a written report to NRC within thirty days following initial notification required by 10 CFR 72,75(a),(b), or (c) [Section 3.1.6, 3 . Z or 3.3 Of this Appendix]. These written reports must be sent to the NRC, in accordance with 70 CFR 72.4, "Communications."

NBTE Written reports prepared pursuant ?o other regulations may be submitted to f.ilfill this requirement if the reports coctain all the necessary inforination and the appropriate distribution is mad The written report rnmt include: A brief abstract describicy the ri cr occarrences during the event, including ail component or system failures that contributed to the event and sigt1ificai:t corrective action taken o r planaed to prevent recurrence;

W A N STANDARD -. .

SP PROGRAMS A N D REGULATORY REPORTING REQUIREMENTS Rev. Z I PROCESSES Base 32 of 53 APPENDIX C Page 4 of 5 A clear, specific, narrative description of the event that occurred so that knowledgeable readers conversant with the design of ISFSI or Monitored Retrievable Storage Installations (MRS), but not familiar with the details of a particular facility, can understand the complete event. The narrative description must include the following specific information as appropriate for the particular event: ISFSi and MRS operating conditions before the event; Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event: Dates arid approximate time of occurrences; The cause of each component or system failure or peixmnei ertor, if known;

. The failure mcide, rnechatism, and effect of each failed component, if known; A list of systems or secondary functions that were also affected for failures of components with muitipie functions; For .vet spent fuel storage systems only, after failure that rendered a ttairi of a safety system inoperable, an estimate of the elapsed !:me frcini the discovery of the failue until the train was returned t o service;

• The method of discovery of each component of system failure or procedural error; Operator actions thst affected the course cf the event, including operator errors, procedural deficiencies, or both, that contributed to the event; 1 For each personnel error !he licensee shall discuss:

a) Whether the error was a cognitive eiror (e.g., failure to recognize the actual facility condition, failure to realize which systems should be functioning, failure to recognize the trui:

nature of the event) or a procedural error:

b) Whether the error ivas coctrary to an approved procedure, vi&s :i direct result of an error in an approved ptocedure, or was associated with an activity or task that was not covered by a n approved procediire; C) Any unusual charac.teristics of the work location (e.g.. lieat, noise) that directly contributed to ::le error; a r d

W A N STANDARD SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 11 PROCESSES Page 33 of 53 APPENDIX C Page 5 of 5 d) The type of personnel involved (e.g., contractor personnel, utility-licensed operator, utility nonlicensed operator, other utility personnel);

1 Automatically and manually initiated safety system responses (wet spent fuel storage system only);

1 The manufacturer and model number (or other identification) of each component that failed during the event:

1 The quantities, and chemical arid physical foirns of the spent fuel, high level radioactive waste, or reactor-related greater than Class C waste; An assessment of the safety consequences and implications of the event. This assessment must include the availability of other systems or cornporrerits that could have performed the same function as the components and systems :hat failed during the event;

. A description of any c0rrectiv.e actions planned as a result of the event, including those to reduce the probabiiity of simiiar events occurring in the future; Reference to any previous similar events at the same facility that are known to the licensee; The name and telephone number of a person within the licensee's organizatic;,i who is knowledgeable about the event and can provide additional iEfGmiaticri concerning the event and the facility's characteristics; The extent of exposure of individuals to radiation or to radioactive materials without identification of individuals by nam .0 Records Retention

-

10 CFR 72.8O(c) Records that are required by 10 CFF! 72 or by the license conditions must be maintained for the period specified by the appropria!e regulation or license condition. If a retention period is not otherwise spec.ifird the above records must be maintained until NRC terminates the licens .2 License Termination and T g r ~ s f e r A. 10 CFR 72,80(e) - Prior to license termination, records required by IC CFR 20.2103(b)(4) and 10 CFR 72.38(d) shall be forwarded to the NRC Regl3n 1 1 offic . -

1 0 CFF? 72.80(f) if licensed activities are transferred or assigned in accordance with 70 CFR 72.44(b)(1). recards reqKired by 10 CFR 20.2103(b)(4) and I O CFR 72.30(d) shall be forwarded to the new !icen:m:

arid the new licensee will be rcsponsible for maintsining ?hese recor:li; !!n:il the IiccI'se is terminate ' JSTANQARD PROGRAMS AND PROCESSES REGULATORY REPORTING REQfflREMENTS SPP- Rev. 11 Page 34 of 53 I Notification Re uirements EvenUCondition Plant i i i : - i t y Plant Ops. Duty Spe Operations , Manager (ODS)

v o g e n e r a t o r trip, msdheduled unit power reduction, i or conscheduled unit shutdown; and when unit is restored to full sewic ' Unplanned entry;zo a Linliting Candition for Operation with time I

\

Yes" Yes*

i Only for Yes'

Yes" I

Yes OnlywhenTS

j d-ration of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or les I duration of 24 safety limits i ~

~ hours or !e55 I exceeded

N A N STANDARD SPP- PROGRAMS AND REGULATORY KEPORTLNG REQUIREMENTS Rev. II PROCESSES Page 35 of 53 APPENDIX E

..- Page 9 of 2 OTHER REGULATORY REPORTING PURPOSE This Appendix identifies other reports required by various regulatory agencies other than !he NR .0 NOTIFICATION .CRIPERIA/REPOWT INSTRUCTION Immediate Notification Federal Aviation Administration (FAA)

~

The Plant Operations Department is responsible for verifying the proper cperaticn of the cooling tower, off-gas stack and ineteoroloyical tower obstruction lights. This should be accomplished through visual obseruation at least once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Any observed 01 otherwise known extinguishmefit or improper functioning of a top obstruction high:

which will last more than 30 minutes should be immediately reported to the FAA at (8CO)

352-6751 (for SQN and WEN) and (800) 772-0547 (for BFN and BLN) by the shift v

manager (or designee). Information !he FAA will reqclire is latitude and longitude for the tower, height of the tower (to mean sea level), facility, name of person making the notification. the condition of the light. or fights. the circumstance which caused the fai!ilre (if known), and the probable date nornial operation will resume. This information is detailed in Notice to Airmen 7AA 79-30.2f. Further notification should be given upor resumption of normal operation of the obstruction light _Plant

_ 10-r ILatitud~g

- Lonqitudc m h A l m e a n sea l e v ~ l ;

. .

ELN Coolini; Tower 1 34" 42' 27.46"N 85" 55' 51.39"W 4C? ft (1109)

Cooling Tower 2 340 42' 2 2 . 8 3 : ~ 85" 55' 44.40"W 45 ::. (7109)

MET Tower 34" 43' 08.33"N 85" 54' 56.99"W 100 ft. (1056)

BFN Off-gas Stack 34O 42' 16.O"N 87" 07' 15.0"W 600 ft. (1 165)

MET Tower 34" 42' 03.19"N 87" 06' 29.28"W 300 tt. (865)

SQN Cooling Tower 1 35" 13' 21.72"N 85" 05' 22.22"W 459 ft. (1 159)

Cooling Tower 2 35" 13' 15.07'" 85' 05' 19.85"W 459R. (1159)

MET Tower 35" 13' 10.50N 85" 06' 04.3O"W 300 ft. (!056)

WBN Cooling Tower 1 35" 36' 05.95N 84" 47' 09.18"W 509 ft. (1240)

Cooling Tower 2 35" 36' 11.88"N 84" 47' 10.37"W 509 ft. (4 240)

MET Power 35" 36' 10"N 84" 47' 24.24W 200 :t. (1011)

NRC notification is not required when the FAA is notified of cooling towe meteorological tower, or Browns Feriy off-gas stack lighting deficiencie .2 m e d i a t e Notification Tennessee Ernerqencv M a n a s e r n g m q e s y (TEMA)iAlabama Emerqencv Manaqerner~tAqency ( A E W The following is a clarification cf what constitutes non-emergency events that reqiiirc t i l e Operatiorrs Duty Specialist lo imrnediate!y notify TEUA or AEM WANSTANDARD SPP-3.5 PROGRAMS AND REGULATORY REPBRTLNG REQUIREMENTS R@V. 14 PROCESSES Page 36 of 53 APPENDIX E Page 2 of 2

'

A. A confirmed fatality, whether or not it is related to nuclear operation Confirmation to he provided by TVA Medical Services when the death is the immediate result of an accident or iilness occurring on sit B. Strikes or honoring of picket iines affecting plant operation C. Accidental activation of the Prompt Notification Systern (PNS) sirens This will include confirmed inadvertent activation of the offsite PNS siren system or portions of the system. The shift manager rnay not he cognizant of ail inadvertent activations; howevcr, if the shift manager becomes aware of such an incident, the Operations Duty Specialist (ODs) shall be called. If the ODs becomes aware of such an incident, from sources other than the shift manager, then the ODS shall notify the shift manage . Undeclared emergency events reported to NRC If an event or condition is discovered which meets the emergency class criteria but no emergency was previously deciared and the basis for the emergency class no longer exists; a report to NRC shall be made. The Sta:e personnel should be notified that the condition was reported to NRC and could resirlt in potential media coverag E. Incidents which could attract a!tention from the immediate local residect Examples include: explosioris, fiies, release of steam or liquid ansite accountability sic-efl soucdirgs. and TVA or local emergency vehicles (fire, rescue, medical, law enforcement), with sirens sounding whiie entering 3r leaving the owner-controlled area (C3C.A) in response to a TVA situatior,. The above examples do not require notification if the State has been notified in advance. The testing or use of TVA vehicle sirens (fire, ambulance. security)

within the OCA only does not require notificatio NOTE Notify the NRC Resident when notifications are to he made to TEMA or AEMA. Notification of the NRC Operations Center is not required due t@

notification of TEMA or AEMA, unless required by other applicabie reporting requirement W A N STANDARD SPP- PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 1 4 PROCESSES Page 37 o f 53 APPENDIX F Page 1 of 2 EVALUATION AND REPORTING OF DEFECTS AND FAILURES 80 COfvlPLY ASSOCIATED WITH SUBSTANTIAL SAFETY HAZARDS PER '10CFR 50.55(e) REPORTING REQUIREMENTS I .Q PURPOSE The purpose of this Appendix is to identify the requirements for the evaluation and reporting of defects and failures to comply associated with substantial safety hazards to the NRC in accordance with 10 CFR 50.55(e) and defines the interface with 10 CFR 21 reporting procedures where appropriat .5 SCOPE s__

Reportability for defects and failures to comply associated with substantial safety hazards under 10 CFR 58.55(e) is applicable to PVA nuclear plants with construction permits and requires TVA to notify the NRC of these deficiencies. These deficiencies are initially identitied by the Administrative Control Programs defined in "Corrective Action Program", SPP- v Reportability of deficiencies which contain Safeguards information Pursuant to 18 CFR 73.71 will he processed in accordance with SPP-1.4: "Safeguards Information." REQUIREMENTS General

--- CFR 50.55(e) requires that holders of construction permits evaluate deviations and failures to copply associated with substantial safe?y hazards as soon as practicable and in all cases within 60 days of discovery, except as provided iri Subsection 3.1 ,D,, in crdet- to identify the reportable defect or fai'ure to comply that could create a substantial safety hazard were it to remain uncorrected If the deficiency is related to any deficiency which has a!ready been repo:ted to NRC and the description of the deficiency and corrective actions for the new deficiency are within the scope of the previously reported deficiency. a memorandum should be prepared documenting that no further repotiiily is require Ifthe deficiency is related to any deficiency which has already been reported i 3 NRC and the description of the deficiency and corrective actions for the deficiency are not within the scope of the previously reported deficienc . Initiate actions to develop a submittal to NRC exFanc!ing the Scope 0; the previously reported deficiency, OR Initiate actions to deveiop an initial report to NRC

N A N STANDARD SPP-3.5 PROGRAMS AND ~ E G U L A T O R YREPORTING REQULREMENTS Rev. i 1 PROCESSES Page 38 of 53 APPENDIX F Page 2 of 2 If an evaluation of an identified deviation or failure to comply potentially .

associated with a substantial safety hazard cannot be completed within 60 days of discovery, then an interim report must be submitted to NRC within 60 days of discovery of the deviation or failure to compl .2 Determininq Repoflability ofDeficiencies Line managenient determines whether the deficiency is potentially reportable or not repoitable under 10 CFR 50.55(e) in accordarice with the timeframe as specified in Section 3.2 above, in accordance with the corrective action program requirements specified in SPP-3.1, If the deficiency is determined to be potentially reportabl . Notify the Site Licensing Manager arid the Site Vice President, within five working days of the completion of the evaluation, that a report is required. Site Licensing will make the final determinatio , Initial NRC notification preferably by facsimile, to the NRC Operations Center. [telephone numbers specified in 10 CFR 50.55(e)(6)(i)] must then be made within two caiendar days following receipt of information by the Site Vice President or Site Licensing Manager. Verification t i a t the facsimile has been received stiol;ld be m a d e by calling the NRC Operations Cente Within 30 days following notification of the Site Vice President or Site Licensing Manager of a substantial safety hazard. a written report is required to be submitted to the NR If deviations are evaiuated under 50.55(e) and result in either a negative reportability determination or reportable defect; then this satisfies the requirements of Part 21

TVAi< STANDARD SPP- PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 4 4 PROCESSES Page 39 of 53 APPENDIX G Page 1 sf 6 DETERMINATION OF RPORTABILITY UNDER I 0 CFR PART 21 PURPOSE_

The purpose of this Appendix is to specify instructions for reporting requirements and for the evalua!ion of potential defects and rroncompliances pursuant to 10 CFR Part 21. Part 21 requires managers and responsible officers of certain firms and organizations which are building, operating. or owning NRC-licensed facilities OF conducting NRC-licensed activities to (1)

report any defects in basic components; (2) report any failures to comply with NRC requirements that could result in a substantial safety hazard; (3j post 10 CFR 21 regulations, Section 206 of the Energy Reorganization Act of 1974, and procedures adopted pursuant to Part 21 regulations; (4) specify in procurement documents Part 21 applicability: and (5) maintain evaluations of all deviations and failures to comply for a minimum of five years after the date of the evaluatio Posting will be in accordance with NADP- .

Reportability of deficiencies which contain Safeguards Information pursuant lo 10 CFR 73.71 will be processed in accordance with SPP- This Appendix establishes the methods for eva1uatir:g defects and failures to comply to determine if they are reportable in accordance with 10 CFR 21, This Appendix also implem1ePts the reqiiirements in the regulation for timing and content of report A flowchart illustrating the process for evaluating potential defects and failures to conlply is shown in Figure 1 of this Appendi .' REQUIREMENTS_ CFR 21 Evaluation and Reporting If a deviation or failure to comply as evaluated could create a Substantial Safety Hazar it must be reported to the NRC unless the Site Vice President has actual knowledge that it has been previously reported to the NRC in writin Reportina Criteria Operating plants with potential defects in instailed equipment must be evali:a!ed under 10 CFR 50.72, 50.73, or 73.71 as appropriate rather than 10 CFR 21 Only those poten!ial defects in basic components which have never been installed or used in the plant are required to be evaluated under 10 CFR 21 Plants with construction permits are required to report under 10 CFR 5 0 . W e j rather than under 10 CFR 21

W A N STANDARD SPP- PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 4 T PROCESSES Page 40 of 53 APPENDIX G Page 2 of 6 CFR 21 Evakgtjon Criteria bine management will perform an evaluation of any defect or failure associated with substantial safety hazards in accordance with the correc!ive action program requirements specified in SPP-3.1. If the deficiency is determined to be potentially repodable under 10 CFR 21, notify the Site Licensing Manager within five working days of the completion of the evaluatio .2 Written ReDort Content Each written report submitted to the NRC shall contain the following:

. The name and address ofthe individual or individuals informing the NR Identification of the facility, !he activity, or the basic component supplied for such facility or such activity within the United States which fails to comply or contains a defec . Identification of the firm constructing the facility or supplying the basic component whic3 fails to comply or contains a defec . Nature of the defect or failure to corriply and the safety hazard which is created or could be created by such defect or failure to compl . The date on which the information of siich defect or failure to comply was obtaine In the case of a basic component which contains a defect or fails to comply. the number and loca!ion of all such components in use at, supp!ied for, or being supplied for one or more facilities or activities subject to the regulations in this par . The corrective action which has been, is being. or will be taken; the name of the individual or organization responsible for the action; and the length of time tha!

has been or will be taken to complete the actio . Any advice related to the defect or failure to comply about the facility, activity, or basic component :hat has been, is being, or will be given to purchasers or licensees.

.

TVAN STANDARD SPP-3.5 PROGRAMS A N D VEGULATORY REPORTING REQUIREMENTS Rev. I d PROCESSES Page 41 of53 APPENDIX G Page 3 of 6 Identification of Deviations and Noncornpliances The principal means for identifying deviations and noncompliances in TVAN are: (a) the Problem Evaluation Report (PER) and the Administrative Control Programs (ACP), (b)

the Receiving Inspection Reporl, (c) notices received from vendors, ( d ) Preoperational, Post-Modification, Surveillance and Post-Maintenance testing, (e) Licensee Event Reports of other sites (LERs). and (f} 10 CFR 50.55(e) repofis at plants still under construction Time Requirements for Reportability Completed E v a l u a u

~ Site Licensing will review the completed Part 21 evaluation within 60 days of the discovery of the deviation or failure to compl Site Licensing will, within five working days, submit the completed evaluation to senior site rnanagerneo . The NRC shall be notified by facsimile, which is the preferred method, by contacting ?he NRC Operatiqns Center at the telephone numbers located in 10 CFR 21.21 (c)(3j(i) within two days of the irlforniation being provided to Senior site management. Verification ?hatthe facsimile has been received should be made by calling the KRC Operations Cente D, A written report will be submitted to the NRC Document Control desk and a copy will be sen? to the appropriate Regional Administrator within 30 days following .

.

receipt of information by seniar site managemen .5 Interim R e p o m g An interim report shall be made in writing within 60 days of discovery and shall containl as a minimum, available information about the deviation or failure to comply that is being evaluated, and shall also state when the evaluation will be completed, if ?he evaluation can not be completed within 60 days of discover W A N STANDARD SPP- PROGRAMS AND REGULATORY REPORTING REQUREMENTS Rev. t l PROCESSES Page 42 of 53 APPENDIX G Page 4 of 6 PART 21 EVALUATION SHEET Type of Document References - Does the deficiency involve a component, that is installed in the plan!?

YesU No0 If "yes", then reportability should be evaluated under 10 CFR 50.72, 50.33 (LER). or 73.71 (Safeguards Events). A separate Part 21 report is not required. If the event is determined to be not reportable under 50.72, 50.73, or 73.71,!hen the obligations of Part 21 are stili met by !he evaluation. If

"no", continue with 8. Does the cornponeiit or service meet the Part 21 definition of a "Basic Yes8 NoU Component"?

If "yes", go to C. If "no", go to E; the item is not reportable Does !he deficiency involve: A failure of the facility, activity. or basic component supplied to TVA, tu compiy with the Energy Reorganizatiorl Act of 1974, or any applicable NRC license requirenents arid regulations, or any rule or order issued by NRC to TVA? Yesn ~o[_1 It "yes", go t o D A loss of safety function to the extent that if the component was installed in the plant there would be a major reduction in the degree of protection provided to the public health and safety? Examples would include moderate exposure to or release of licensed material or major degradation of essential safety-reiated equipment or major deficiencies involving desigri, construction, inspection, test, or us Yesn Nom If "yes", go to D A departure from the technical requiremerits for a delivered component or service as set forth in a procurement document?

Delivery occurs upon acceptance by TVA (e.g., at receipt YesU Nom inspection). If "yes". go to D Answer all three questions under C above. If all "No's", this deficiency is not reportable. Go to "E". If any Yes's", continue as indicated by the applicable 'GO TO" statement. Could the deviation or noncompliance have caused a substantial safety hazard (By definition, item CZ constitutes a substantial safety hazard.)?

(If "No," this item is not reportabie - If "Yes." this iterri is reportable; undm Part 2 WAN STANDARD SPP- PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. $ 1 PROCESSES Page 43 of 53 APPEMBbX G Page 5 of 6 PART 21 EVALUATION SHEET Is this item potentially reportable by TVAN under Part 21?, Yesa NOD Prepared by: Date ~

Responsible Manager: Date

- . I _ _ _ - . ~ . _ ~ . , * ~ _ _ ~ . . . ~ . ~ - i__

-

SITE LICENSING MANAGER REVIEW (OR DESIGN,C& Has this item been reported by TVA, another licensee. or by a vendor? Yesm Nom If "yes," a separate 19 CFR 21 report is not require If "no," complete a Part 21 Report (App. F) and notify Site Licensing or

-__

Senior Site Managemen p . ~ - - ~ . ~ ~ .. i- . . - = .-

.

Tracking Data:

Date forwarded to Site Licensing -. .

Date received by Site Licensing -, -

Date Senior Site Management informed -

...

.,

~

Date NRC notified (initial notification) -. .- ~.~

Date NRC notified (written report) __.._I_.~

W A N STANDARD SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQUIKEMEN'IS Rev. i f PROCESSES Page 44 of 53 APPENDIX G Page 6 of 6 EVALUATION LOGIC FOR PART 21 Figure i

N A N STANDARD SPP- PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. $ 1 PROCESSES Page 45 of 53 APPENDIX H

-_ Page 4 of 3 REPORTING OF DECOMMISSIONING FUNDING PURPOSE The purpose of this Appendix is to identify the minimum requirements for submitting a decommissioning funding report to the NRC as required by 10 CFR 50.75, "Reporting and Recordkeeping for Decommissioning Planning." This Appendix also identifies requirements for notifying NRC when permanently shutting down !he operation of a reactor, as required by 40 CFR 50.54(bb). TVA shall report decommissioning funding plans for its nuclear plants as required by 10 CFR 50.33(k), "Contents of Applications; General information." 10 CFR 58.75 establishes requirements for: 1) indicating to NRC how a licensee will provide reasonable assurance !?at funds are available for the decommissioning process, and 2) reporting tirneframes. This Appendix provides both the minimum information that is required to be submitted in the deconimissioning funding report and the required reporting timeframes. Additionally, this

.

Appendix provides the minimum reporting requirements (e.g., reporting timeframes, description of the program that TVA intends to implement for managing all irradiated fuel until it is transferred to the Department of Energy) for notifying NRC when permanently shutting down the operation of a reactor, as required by 1 0 CFR 50.54(bb). REQUlREMEN.TA General 10 CFR 50.35 Repoflinq R e q y j r e m s 10 CFR 50.33(k) and 10 CFR 50.75(%) require nuclear plant license applicants and licensees to submit decommissioning funding repoits. License applicants are reqiiired t o submit information regarding how reasonable assurance will be provided that funds wil; be available for decommissioning pursuant to 10 CFR 50.33(k) and 10 CFR 50.7 License holders are required to report periodicaliy on the status of their deconrrT1issionir;g funding pursuant to IO CFR 50.75(fj(1) as described be!o Written Report Content The information in this report must include, at a minimum: The amount of decommissioning funds estimated to be required pursuant to 10 CFR 50.75(b) and (c): The amount accumulated to the end of the calendar year precedirg I3e date of the report: A schedule of the annual amounts rernainicg to be COlleCted;

N A N STANDARD SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 41 PROCESSES Page 46 of 53 APPENDIX H Page 2 Qf 3 The assumptions used regarding:

a) Rates of escalation in decommissioning costs; b) Rates of earnings on decommissioning funds; and C) Rates of other factors used in funding projections; Any modifications occurring to a licensee's current method of providing financial assurance since the last submitted report; and Any material changes to trust agreement Responsibilities The Treasurer's Office is responslble for providing input to Corporate Licensing for the decommissioning report for items 2, 3, 4.b, 5,and 6 prescribed in Section . Corporate Nuc.lear Engineering is responsible for providing input to Corporate Licensingfor the decommissioning report for items 1 4a, and 4c. prescribed in Section . Corporate Licensing is responsible for preparing and submitting the final decommissioning repor Required Reporting Pimeframes TVA shall report. oira calendar-year basis, to NRC by March 31, 1999, and at least once every 2 years thereafter on the status of its decommissioning funding for each licensed reactor that it own . TVA shall submit a decommissioning report annually, when:

a) A reactor power plant is wiihin 5 years of the projected end of its operation; b) Conditions have changed such that a reactor powe: plant will close within 5 years (before the end of its licensed life);

c) A reactor power plant has already closed (before the end of its licensed life); or d) A plant is involved in a merger or acquisi?ion

hVANSTANDhRB SPP-3.5 PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. I 1 PROCESSES Page 47 of 53 APPENDIX H Page 3 of 3 Pennaneot Reactor Shutdown Reportinca Requirements TVA shall, within 2 years following permanent cessation of operation of reactor or 5 years before expiration of the reactor operating license, which ever occurs first, submit written notification to the Commission in accordance with 10 CFR 50.54(bb). This written notification shall: Include a description of TVAs program for managing and funding the irradiated fuel at the reactor until title to and possession of the irradiated fuel is transferred to the Secretary of Energ . Demonstrate to NRC that the elected actions are consistent with NRC requirements for licensed possession of irradiated nuclear fuel and that the actions will be implemented in a timely basi . Verify, for actions requiring NRC prior approval. that submittals have been or will be made to NR . Be retained as a record until expiration of the reactor operating license TVA shall notify the NRC of any significant changes in the proposed waste management program as described in the initial notificatio W A N STANDARD SPP- PROGRAMS AND REGULBT'ORY REPORTING REQUIREMENTS Rev. Z 1 PROCESSES Page 48 of 53 APPENDIX I

.- Page 1 of 2 COMMUNICATION WITH THE NRC FOLLOWING A SIGNIFICANT OPEKATIONAL EVENT PURPOSE The purpose of this Appendix is to define the communications that needs to be established with the site resident inspectors and the regional adniinistrator staff within 24 to 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following a significant operational event that could result in an incident investigation by the NR .0 SCOPE This appendix briefly discusses the NRC decision making process for determining their inspection response following notification of a significant operational event and provides guidance on the need for and types of information required by the NRC during this decision making proces .

During the site investigation phase following a 10 CFR 50.72 notification for a plant trip or another significant equipment maifunction or failure, a clear communication path needs to be maintained with the resident inspecto Bacmnd Upon notification of a significant operational event, the regional administrator and his staff will .

perform the initial review to assess the safety significance of the event. The guidance provided in NRC's Management Directive MD-8.3 'NRC Incident Investigation Program" is used lo assess the level of response required. The criteria ?or determining between an Incident investigation TEarn (IIT), an Augmented Inspection Team ( A l n or a special inspection is based on a combination of deterministic criteria and an estimation of the Conditional Core Damage Frequency (CCDF) of the actual plant configuration at the time of the significant operational even in determining the risk significance of the everit the NRC is instructed to assess 1) the potential influence on risk of the dominant core damage sequences, 2) level of confidence in failure Or unavailability values assumed for these sequences, and 3) level of confidence of equipment tailure/recovery and their influence on the CCD An accurate estimated CCDF is crucial to the determination of the inspection response. With t l e high reliance of an estimated CCDF on the level of confidence of equipment recovery and satisfactory completion of operator actions, an accurate CCDF requires a thorotigh understanding of the status of equipment at the ?ime of the event and the causes of fai!ures f31 any equipment that did not perform as designe .0 REQUIUEMENTS A communication path could be accomplished through normal resident to licensee manager cornmunicat.ion or through the trip response team llcensing representative. This decision should be based on the complexity of the event and the event investigative team structure. The communication process must assess that sufficient and timely information is being provided to the NRC decision maker ~~NsTANDARD SPP-3.5 PROGRAMS AND .REGULATORY REPORTING REQUIREMENTS Rev. 1t PROCESSES Page 49 of 53 APPENDIX I Page 2 of 2 A s the site investigation evolves and the failure modes and causes of equipment and performance issues are identified, it is critical that the NRC be kept informed of the evolving analysis findings as they pertain to the following crucial estimated CCDF inputs:

Clear understanding of the event sequences, equipment failures, equipment not available foi mitigation, and operator performance problems;

~

Perspectives on equipment failures that address suspected causes, potential extent of condition, arid ability of operators to recover failed equipment:

~

Perspectives on unavailable equipment and ability to restore functions to support risk significant scenarios;

- Perspectives on operator performance problems and their ability to recover or restore critical function NOTE It is important to recognize that the MD-8.3 decision process not only assesses what happened, it also assesses what might have happened with respect to risk Significant

scenario W A N STANDARD SPP- PROGRAMS A N D PEGULATORY REPORTING REQUIREMENTS Rev. 13 PROCESSES Page 50 of 53 APPENDIX J

._ Page 1 of 2 INTERNAL NOTIFICATION O F EVENTS REQUIRING SERIOUS ACCIDENT INVESTIGATI'ONS PURPOSE The purpose of this appendix is to provide the infemai management notificafion requirements for serious accidents, as prescribed in TVA-SPP-I 8.010, "Conduct Serious Accident Investigation."

This appendix applies to all serious accidents that result in any of the following occurrence EXCE~~ION--Radiological control and nuclear operational safety incidents subject to other specific reporting and investigation requirements are investigated by TVA Nuclear (TVAN) as required by applicable procedure .0 REQUIREMENTS Serious Accidents include any of the following occurrences: A fatality or in-patient hospitalization of three or more TVA employees within 30 days of an acciden . Other events (including property damage only) which under slightly differerr!

circumstances would have met or may meet the following provisions: Falls (usually from elevation) causing head injury, broken bones, acdior other serious injur . Electric contact resulting in current flow through the body andior lcss of consciousnes . Electric are causing SeGor,d- and/or third-degree burns to the bod . Being caught in or by equipmenthachinery causing head injury, broken bones, andior other serious injur . Thermal burns causing second and/or third degree burns to the body Being struck by equiprnent'machinery or falling objects causing head injury, broken bones: and/or other serious injur . Overpressure resulting in component failure (gas, hydraulic. or air!

causing head injury, broken bones, and/or other significant traumatic injur . Release of latent or kinetic energy (tension [objects under cornpressior; or projectiles [objects being thrown]) causing head injury broken 'sarle andior other serious injur .2 Manager in Charge of Workplace, or designee, performs the following initial actiors:

NOTE Checklist items are listed in the general order of preference: however, many of these can occur simultaneousl WANSTANBARB SPP-3.5 PROGRAMS AND pEGULATQRY REPQRTING REQUIREMENTS Rev. I 9 PROCESSES Page 51 of 53 APPENDIX J Page 2 of 2 Notifies the TVA Police for assistanc Gathers the following notification information: Accident location: Time of the accident: Name(s) of the individual(s) involved: Extent of injuries: Contact person (name and telephone number): Brief description of the occurrence: Immediately notifies the following persons and provides the information listed in Section 2. NOTE If either of the individuals listed in Sections 2.2.C.1, or 2.2.6.2. is unavailable, do not delav in con_!g.ctina the T.yb 0peratqr.or the Prooram .Manaqer, C o w

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SafetaRefer to Section 2.2.C.31, Responsible executive vice president of the organization where tKe accident occurre . Safety manager of the organizatio . TVA Operator at (865)-632-2101. Inform the operator that this call is to report a serious accident to the Program Manager, Corporate Safet Alternate numbers for Corporate Safety are:

a) Office: (865)-632-7753 or 7756 (during business hours).

b) Cell phone:'(865)-414-881 c) Pager: (800)-201-813 .3 Licensing will evaluate those serious accidents that Operations determined were not reportable per 1OCFR§0,72(b)(2)(xi) [refer to Appendix A, Section 3.1 .C.4.] to determine if a "courtesy" phone call to NRC is appropriat W A N STANDARD s PP- PROGRAMS AND REGLLUTORY REPORTING REQUIREMENTS Date ~- 05-02-2001 PROCESSES Page 52 OP 53 NRC EVENT NOTIFICATION WORKSHEET Page 4 of 2 NRC EVENT NOTIFICATION WORKSHEET f kVtNI I IMC & L U N C 8 H r Non-Emergency 10 CFR 50.72(b)(3)

Page 1 of 2

W A N STANDARD SPP- PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Date 05-02-2001 PROCESSES Page 53 of 53 NRC EVENT NOTIFBCATION WORKSHEET Page 2 of 2 LnlK Rate:

......

Leak Start Date:

L I S of safety Related Equipment not BpelaUOOEl:

.

\.. ..

JPM NO.A3. REV. NO. 0 PAGE 1OF R BROWNS FERRY NUCLEAR PLANT JOB PERFORMANCE MEASURE JPM NUMBER: A3.1 R TITLE: REVIEW A RADIOLOGICAL SURVEY MAP TASK NUMBER: NIA SUBMITTED BY: ._ DATE:

VALIDATED BY: DATE, APPROVED: --- - DATE:

TMINING PLANT CONCURRENCE: ~ BATE:

OPERATIONS

Examination JPMs Require Operations Training Manager or Designee Approva!

and Plant Concurrence

--.

JPM KO. A>.1F REV. NO. c PAGE I @E' 1 C

J i ? M NO, A2.lR RE PAGE 2 OF 10 REVISION LOG

...

ZPM NO. A%.IF.

R E V . XO. 0 PAGE 2 OF 1 C

J E ' M N O . A2.11; REV. NC . 9 PAGE 5 GF 1 C Perfcrir.nrice - -~

Step: Criti.cai X Kotl C r i t i - a l -~.

7.24 FULLY OPEN 2 - F C V - . / 5 - 5 0 i i s i n g U O Z SPFAY SYS I1 TEST VALVE, 2 HS - '7 5 - 5. A .

S t -~

-. and-ard:

PLACES HANDSWITCH in open position and fully opens valve as inidicated by RED L\GWT onl MOTES :

REFER TO Illustration 1, Process f o r Stroke Timing Valves per the ASME OM Code.

• * * * * * * * * * * * * * * * * * * * * * * * f t * * * * * * * * * * * * t * * * * * * ~ * * ' k * * * * * * * * * * * * * * * *

?erforrnance

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Stc,p: __ Crit i ca!. X Mot. C r i z i cz1 7.25 CLOSE and TIME 2-PCV-95-50 using CORE SPRAY SYS PI TEST VALVE, 2 - H S - 9 5 - 5 0 8 and RECORD the stroke time below.

S~-t andar-d -:

Records closure time of valve in table for step 7.25. Ckosure time should be below the normal range and maximum valu ?eerfoi:mance St.eo: Criti.ca1 X ~ ~ ~ ~ Not C r i , t . i c a l--

9.2 VERIFY the time recorded is within the m a x i m u m value listed.

_

Stan -d a r-d :

Notes that time recorded in step 7.25 is within the maximum value listed as Acceptance Criteri * * * * r * * * * * * * * * * * * * * * t * * * n * f * * * * * * * * * * * * * ' * k * * * . ~ * * * * * * * * * ~ * * * ~ * * * * * * * ~ ' k * * ' ~ ~

Perf:)rmaric~ - -c S t.e= ~

IZ t~ical<: -

>Jot c:rif.;<:d 7.26 IF the stroke time measured in s t e p 7.25 is within the maximum value listed and outside the normal range, then P e r f o r m the following (otherwise NA this ection)!BFPER97l?861 7.2 OPEN the 2-FCV-75-50 using CORE SPRAY SYS I1 TEST VALVE, 2-HS-75-50A.

-j :t andar-cl :

DET'ERM1NE.S time outside the normal ranae and OPENS 2-FCV-75-50 to allow re-stroking and retiming the valv ?erfori-iancr -

~- 3 t e-gi: Crificai X .. N a t C r j . t i c d.

7.2 CLOSE and T I M E 2-FCV-75-50, using CORE SPRaY SYS I% TEST VALVE, 2-HS-75-5QA and RECORD the restroke time on Attachment _ .:

5;tan:iarii PLACES AND HOLDS handswitch in close and times valve closure from movement of HS to Green L ight Only. Notes time is again below normal expected range. Records Information on Attachment 7 of SW.

7.2 VERIFY the restroke time recorded on Attachment 7 is within the maximum values listed. [BFPER9713861.

.itLiri&rd-.

DETEMINES re-stroke time is below the maximum value listed. Acceptance criteria met

• * * * * * z * * * * * * * * * .k * * * * * * * .k * * * -k * * * * * * * * * i* * * * * * ; e * * * * i* * * i.t + * i*

). I* r _i EXAMINERS NOTE; IF STEP 7.27 IS STARTED TREN ASK STLTENT TO STOP SR PERFROMANCE, EVALUATE THE ITEMS FOLLOWING;

E n t e r s fl@w c h a r t at. STAKr Stroke T i m e Valve Per SR Strcke 'Time X i t h i n Maxi~n;urr!L i m i t ' ? -fES C t r::)ic.; T;.me w i t h i ri No riiia 1 Ea rize NO

._s_ ...

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TENNESSEE VALLEY AUTHORITY BROWNS FERKY NUCLEAR PLANT SURVEILLANCE PROCEDURE 2-SR-3.5.1.6(CS 61)

CORE SPRAY FLOW RATE LOOP S I REVISION 12 QUALITY RELATED PREPARED BY: Keith Smith

~~~ ~ ~.

RESPONSIBLE ORGANIZATION: OPERATIONS , .-

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APPROVED BY: -. PHILLIP CHABkVELh ____~~ __

EFFECTIVE DATE ___

-C3:05/2C!~3 LEVEL OF USE: CQPdVlNUOUS USE

.. PAGES AFFECTED 61 KEVlSlON DESCRIPTION IC-13 ENHANCEMENT Page 61, Replaced illustration 1 that referenced attachment 3 for the restroking of any valves, with a new illustration. The original iilustration could not be changed without special software program. The New illustration is a generic restroke illustration that may can be modified using the current Microsoft Word prograrn The actual illustration being deleted does not show as a revision and o d y the illustration title appears as a chang .~ ~ ~~

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CORE SPRAY FLOW RATE LOOP I 1 2-SR-3.5.1.G(CS II)

Rev GO1 2

Page I of 61

~. . ~ ~ .... ~ INTRODUCTION This surveillance procedure is performed to determine the operability of Core Spray (CS)

System LOOP II in conformance with requirements specified in Technical Specification (TS) Surveillance Requirements (SRs) 3.5.1.6and 3.5.2.4,and to functiona!ly test the CS LOOP II minimum flow valve as required by SR 3.3.5.1.2. This SR also tests CS Pumps 2 8 and 2D, and associated valves, to the ASME OM Code Program of TS 5. The equipment area coolers associated with Core Spray Pump 28 and 2D as specified in the Technical Requirements Manual (TRM -- Sections 3.3.3and 3.5.3)will also be verifie .2 Seeps 1. This procedure verifies CS Pur:ps 28 and 2D can be operated from the miair1 control room to deliver rated ilow and pressure against a simulated reactor pressure. This satisfies TS SRs 3.5.1.6 and 3.5.2.4for CS Loop I . This procedure also serves to dezonstrate compliance with the TS SRs. TRM Surveillance Requirements (TSRs), and ASME OM Code program requirements indicated below:

CORE SPRAY SYS II TEST VALVE 2-FCV-75-50 is exercised (i.e, open and

. closed) by this procedure to demcrsirate its sperability per the ASME OM Code Progra The CS pump room cooler located is the NE Corner Room at Eleva!ion 541 5 the Reactor Building is operated during performance of this procedure to demonstrate its cooling ability. Specifically, ti;e room cooler must start a:;d operate upon start of either CS Pump 2B or 2D. This procedure satisfies TSR 3.3.3.2.2 (for Table 3.3.3.2-1furctiori 4) and TSR 3.5.3.1for CS L o w I The CS Loop Minirnum Flow Valve, 2-FCV-75-33, is verified to open and close as system parameters require to satisfy TS SR 3.3.5.1.2for Table 3.3.5.1-1 Function 1 . The CS System LOOP II discharge p i p i q venting is verified by this procedzrt:

This satisfies TS SI3 3.5.1.Iand SR 3 5 2 2 testipg requirenents for CS Loop 1 . This procedure is a!so used to coiiec! appiicabie data as required by 2-Sl-3.1.7 c:

2-SI-3.1.8. This procedure in ccnjuncticn with SRslSls listed as being ASME tyce in Surveillance Program Matrix. fuily impiernents the ASME Ow1 Coce Prograni :I TS 5.5.6and 0-TI-36 . ~ -~ .~ ~_~ - -. - ___~ -

--~!

~ ~~ ~

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CORE SPRAY FLOW W Y E LOOP II 2-SR-3.5.1 6(CS 11)

a- UNIT 2 BFN -.

Scope (Continued)

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Rev 0012 Page 2 of 61 I

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1. Satisfactory completion of this procedure meets the ASME OM Code requirements as specified in 2-SI-3.2.1 for cyc!ing the following valves:

2-FCV-95-50 2-CW-75-5376 (cycle valve open and closed)

2-CKV-751-5375 (cycle valve open and closed)

2-CKV-95-570B (cycle valve open and closed)

2-CKV-75-57OD (cycle valve open and closed)

1. This procedure and 2-SR-3.5.1.6(CSI) fully satisfy SR 3.3.5.1.2for Table 3.3.5.1-1 Function . This procedure and 2-SR-3.5.l.G(CSI) fully satisfy SRs 3.5.1.6 and 3.5.24 for CS pump . This procedure may be performed in lieu of 2-SR-3.5.1.1(CSI I ) to satisfy SRs 3 5.1.Iand 3.5.2.2for CS Loop II. Note that !he freqtiency of SRs 3.5.1 1 anrl 3.5.2.2 is once per 32 day .3 Frequencv This procedure is to be performed once per 92 days. This satisfies the requiremer:ts :;i

.~ S R 3.3.5.1.2(once per 92 days), and SRs 3,5.1.6and 3.5 (in accordance with ' -

lnservice Testing Program). Appiicabilitv LCO 3.3.5.1 Functions: MODES 1, 2, 3. 4'"). and 5'"'.

LCQ 3.5.1 Functions: MODES 1, 2, and LCO 3.5.2 Functions: MODE 4, and U6BE 5 except with the spent fuel storage p ~ 3 l gates removed and water level 2 22 ft over the top of the reactor pressure vessel ilar:ge TRM LCO 3.3.3.2 Functions: MODES 1 , 2 . 3, 4'"', and 5")

TRM LCO 3.5.3 Functions: (a).

(a) When associated subsystemjs) are required to De operable

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2-SR-3.5.1.6(CS II)

Rev 001 2 Page 3 of 61 i 2.0 mRENCES

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2.1 BFN Unit 2 Technical Specific-Section 3.3.5.1,Emergency Core Cooling System (ECCS) Instrumentation Section 3.5.1, ECCS-Operating Section 3.5.2. ECCS-Shutdown Section 5.5.6,Inservice Testing Program 2.2 Technical Recluirerien&Manual Section TI? 3.3.3.2, Low Pressure Area Cooler Instrumentatio Section TR 3.5.3, Equipment Area Cooler .3 BFNUFSAR Section 6.4.3, Core Spray System Description Section 6 5.2.4, Core Spray System Section 7.4, Core Standby Cooling System Conti-oi and 1ns:rumen:ation

.~ .

2.4 Plant Instructions 0-GOI-300-2, Electrical 0-01-23, Residual Heat Removal Service Water System 0-01-67, Emergency Equipment Cooling Water Sys:em 2-01-75, Core Spray System 2-SI-3.1. I . Core Spray Pump Performance 2-SI-3.1 8, Core Spray System Baseline Data E v a ! u a t i n 2-Sl-3.2,1, ASME Section XI Valve Perfc,r -.: ance

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2-S1-4.2.B-60FT(Il) Core Spray Pump Area Cooler Fan Therrnostat Fuilctiow (2-TS-64-73).

2-SIMI-75B, Core Spray System Scaling and Setpoint Documents

- __ __ --

Rev 8012 Page 4 of 61 Plant Instruct& (Continued)

0-TI-23Q,Vibrating Monitoring and Diagnostics 0-TI-280, Calculations of Flow Transmitter Output for Use with ASME Section XI 0-TI-362, lnservice Testing of Pumps and Valves SPP-8.1, Conduct of Testing SPP-3.2, Electrical Equipment Environmental Qualification (EQ) Program SPP-10.3, Verification Program 2.5 Plant Drawin=

0-73OE930 Sheets Z and 13 through 20, C S System Elementary Diagrarls 0-731 E761 Sheets ? O and 1Z , Emergency Equipment Elementary Diagrams 2-45N2750-121480V Reactor MOV Board 28 Concection Diagram 2-45E779-8, 480V Shutdown Aux Power Schematic Diagram 2-45E779-IO, 480V Shutdown Aux Power Schematic Diagram 2-45E779-16, 480V Shutdown Aux Power Schematic Diagram 2-47E814-I, CS System Flow Diagrarn 2-4?E610-35-lI Mechanical Control Biasrams CS System 2-47E611-75-1 through -3, CS Systeli Mechanical Logic Diagrams 2.6 Vendor Mag-

~FN-0-CVM-GO80-210~-87, GEK-779A. Controls Sys Maruals Vol 7 [Unit 2 oflly'i BFN-2/3-VTM-B260-0010, Vendor Technical Manual for Bingham-Vdilliar?ette 12x1CZ, 14 1/2 Single Stage CVDS Pump Other Documentation 2.7 _____.-.__

GE Services Information Letter (SIL) No 93, Systern Protective Device MD-Q2075-893109; Core Spray Acce!:tancc: Criteria 'or Technical Specificat'on Operability Surveillance (B22 9001 18 174)

NESSD 2T-O64-0072-~0-01,Nuc!ear Engirwerlng St?!po~fltijnn Scaiirig OOcu ED-Q2075-883532 Se:point and Sccjling ana Caicii.ntlon for 2-F-75-21 arid

__

2-SR-3.5.1.6(CSII)

Rev 0012

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Page 5 of 61

! PRECAUTIONS AND LIMITATIONS

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31 This test manually starts CS Pumps 28 and 2D and measures system pressure with various pump configuration .2 If maintenance other than what is provided in this procedure becomes necessary, a work order should be generate .3 [NRC/C] The CS pump motor nameplate full load current is 80 amperes. The maximum allowable continuous running current is 92 amperes which is based on full load current multiplied by the motor service factor of 7 .15. [SLT 861087005] The stroke time for all motor-operated valves shall be measured to the nearest tenth of a second. Stroke time is defined as the period from initial switch movement to Panel indication of completed valve trave .5 ASME OM Code information should be reviewed and recorded in accordance ;Nith 2-SI-3.1. I , 2-SI-3.1.8, and 2-SI-3.2.1: as appropriate, within 4 days of the compIet!cn of this procedur .6 Annunciators which will alarm during perfornlance of this procedure are specifled on Attachment 5 and within the procedure text as note When starting pumps with :he injection valves and test valve closed, the CORE S F H A Y SYS I1 DISCI4 PRESS HIGH (2-xB-55-3F, wicdcw 30) an:iuncia!or may alarm momentaril This procedure lifts the power lead (Cable 2ES3308-ll, Wire No. 8EX) at Termirial 2 i:i JB 8790 for 2-TS-64-73 to enable testing of the CS pump room cooler fan due to CS pump operation regardless of area temperature. This operation disables the CS pump room cooler fan thermostat. 2-S1-4.2.B-60FT(lIj must be performed on the thermostat return it to operable statu All wire lifts shall be taped or covered in such a mancer as to prevent perscnnel cr equipment hazards. Wire lifts performed in conjunction ivith this procedure be restored to their as-fcund configuration (e.g.. bend radiiis).

3'3 If the CS pump room cooler fan thermostai is disribled for grester than 24 hoar-s a TRM LCO may resul .10 A radiation work permit (RWPj may be required to perform this procedure 3 11 The motor start limitations cf 0-GOI-SOO-2 !irnit CS p u n p starts to twa starts ir succ;ession ccasting to rest between starts with the mator initia:iy at ambient teiTip+:r:i::ar<:

or one start with the motor initially at ncrrnai o0ernt:ng ternperatijre

L-

I 2-SR-3 5 1 6(CS 11)

Rev 0012 I Page 6 of 61 PRECAUTIONS AND LIMITATIONS (Continued)

. .-.

3.12 DL * perforrnance of this test, RHRSW Pumps A I , 62. B3, and D3 may start if they are

~

v a l i d in to Emergency Equipment Cooling Water (EECW) System, operable, and NOT running. This constitutes a planned actuation of the EECW System, an Engineered Safeguards Feature (ESF) andi therefore, is NO1 reportabl .13 The BFN ASME OM Code Ten Year Program for monitoring pump flow and pressure requires that measuring instruments are accurate within two percent. This accuracy requirement is implemented in this procedure by using temporary pressure gauges with this accuracy to monitor pump suction pressure and by directly measuring flow modifier inputioutput using either a digital volt-ohm meter (BVOM) or the Integrated Computer System (ICs). Existing ,:ischarge pressure gauges satisfy the accuracy requirement and do N8T require substitution with temporary measuring instrument The ASME OM Code also provides the f d l scale (F§) range of compliance instrumentation be three times (3x1the expected process value OF less to ensure an accurate measurement since compliance instrument accuracy requirements are based on the full scale range of the instrument. The f d scale range of the temporary suctlor?

pressure gauges specified by step 5. account for this requirement. The existing discharge pressure gauges comply with the ASME OM Code range requiremert and C;o NOT require substitution ,vith temporary measuring instrument .14 Corrective Actions shall be dispositioned in accordaxe with SPP-8.1, Conduct of Testin .15 ASME OM Code data collection requires the CS pilrnps be operated at a repeatable reference value when discharge presstire readings are taken. While t5e flow rate may be an average of the required value (32CO gp:?,), the UO should attempt to main!ain the flow rate as close as possible to the stated value. This helps to ensure that discbarge pressure readings do not vary significantly due to operating point changes from performance to performance of this procedure urless an actual problem exis [BFPER98-004734-000]

3.16 Any ICs console in the main control rocm may be used for collecting ICS data ssecified by this procedure. If the ICs console originally selected fails i o operate preper!y durir'q performance of this procedure, another IC§ cor?sole(s) may be m e d for comple!im of test activities provided the failure is isolatec tc, the console in use If an alternate ICs console(s) is used, then the changejsj shall Se ncted in the post-test rer?.arkssectlo!: (cf Attachment 1

....

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30 PRECAUTIONS AND LIMITATIONS_ (Continued)

3 17 The suction pressure gauges used for ASME OM Code data measurements will be measuring a nominal pressure of approximately 5 to 8 psig. The gauges shall have a minimum range of 0 to 10 psig and a maximum range of 0 to 30 psig to ensure compliance with the ASME 01\61Code The gauges need have the same range. For example, Gauge A may have a range of 0 to 1 Q psig while Gauge B may have a range of 0 to 30 p i g .

The process connection fittings used for gauge installation shall have a minimurn withstand pressure rating of 450 psig. The pressure gauges are NOT reqtiired to meet this withstand requirement due to the sense line sine and root valve isohtior! capability.

3 18 Vibration readings (obtained by Electrical Maintenance using the portable vibratior, instrument) should be verified to not be in ALARM status. This is _NOT acceptance criteria, but is a means of verifying vibration data obtained is valid. Any vibraticn reading exceeding an ALARM limit should be remeasured to verify the reading is valid.

3 19 The observers who verify lights should have spares available to replace burned au bulbs.

a 20 REFER TO Illustration 1, Process for Stroke Timing Valves per the ASME Ob1 Code Valves that exceed the Maximuin allowed stroke time shall not be restroked a r d w!.!st r a declared inoperabi I

. ~.~ ~ ..~ . -

CQRE SPFLAY FLOW RATE LQQP It 2-SR-3.5.'.6,(GS II)

Rev 002 2 Page 8 of 61

-~___ __ .. ..

NOTE:

Prerequisite steps may be performed in any order at the discretion of the Unit Superviso .0 PKEREBUISITTES This copy of 2-SR-3.5.1.6(CSII) is verified the most current revisio ;.)q'.

. r - -

' i EECW I S in service to support CS N E pump reom cooler operation A : - . CS System LOOP II IS available for test:ng I

..

.--

1 Qualified personnel listed below are available to perform this procedure UO1 IM2 AUOZ EM2

...... Electrical Maintenance (EM) has been potified of this procedure performance and is ready to perform the zpolicabie pump vibration monitoring sections of this procedire and 3-TI-53 .6 Instrument Maintenance (IM) has been notified of this procedure performance and is ready to perform the applicable portions of :his procedure and 2-Sl-4.2.B-60FT(li _ *

INITIAL3 40 PR%RE-QUlSITE§ (Continued)

NOTE:

Separate equipment qualification work record packages are required for each maintenance discipline (i.e,, EM and IM) to facilitate review and approva .7 Computer-gmerated environmental qualification work record forms (SPP-9.2-26) have been obtained from Work Planning and will be completed by IM during performance of this procedure for the following terminal block and cable:

2-TB-64-8790 2-fQES-064-3308/11 Computer-generated environmental qualification work record forms (SPP-9.2-26) have been obtained from Work Planning and will be

.~ .

completed by EM to document CS pump Totor vibration levels during performance of this procedure:

2-MTR-075-0033 2-MTR-C' 5-0042 . PREWEQUISITES (Continued)

NOTE:

The CS LOOP II flow rate may be monitored from an ICs console by requesting either a single value display (SVD 75-43) or the CS mimic. IF the ICs is used to collect CS LOOP II flow rate data, THEN CHECK no gross instrument channel failures have occurred by noting t h e IC§ display to be used indicates within :GO gpm of CORE SPRAY SY§ II ;(

FLOW Indicator 2-Fl-55-49 on Panel 2-9-3; Otherwise NIA.

4.10 IF the ICs is used to collect CS LOOP I1 flow rate data, THEN PLACE NIA in t h e fol!owing steps;:

7.5.5,7.15.2, 7.32.2, 7.48, and 7.52.8 (Otherwise NIAthIs step.)

4 11 IF the ICs IS NOT used to collect CS LOOP II flow rate data, THEN PLACE NIA in the following steps:

7.5.1 and 7.32.1; (Otherwise NIA this s:e SPECIAL TOOLS AND EQUIPMENT R E C O M M E U

-.

51 -

RE ImendedTools 5. Lead seals and crimping tool 5. Slotted head screwdriver set 5. Crescent wrench 5.1.et Needle nose pliers Recommeqded Measurinq and Test Equipment IM&TE)

5. Enter information where reqtiired. Vibration M&TE accuracy and frequency response range are controlled by !he BFN Vibration Program and have been verified to meet the listed reqiiiremen!s. Verify required range and accuracy for remaining M&TE by reviewing calibra:ion sheets. NIA blanks for instruments not used. For example, only two suction pressure gauges are required and they are usually analog. in this case the digital gauge .ilanks would be NIA. Likewise. the DVQM is only required if the IC§ is no! available, and those blanks will usca!!y be Ni Reqkired Accuracy 7 9.89-1000 NIA

- ~.

.- .- . .-

Suction maximum

.- ~

N!A I .

.- ~..- 1

50 SPECIAL TOOLS AND EQUlPhldl~TRECOMMENQEQ (Continued)

NOTES:

(1) Accuracy ratios should be calculated in the space provided before leaving the IM Shop area (2) M&TE with an accuracy ratio less than 1 0 shall NOT be used for procedure test record purposes 5. Two analog or digital pressure gauges (as listed in Section 5.2.7) and associated process connection fittings (e.g.,Tyyon tubing) per Section 3.1 GauQe A Required Accuracy 0.6 psig -

Accuracy Ratio -- -

I  :, .

M8TE Accuracy (, ~,1 psig Gauqe 5 Accuracy Ratio =

Required Accuracy

- --

- 0 6 psig- -

>

M&TE Accuracy

~

( I ps1g - --

-_ ,~.. -. -

2-SR-3.5.1.6(CS 11)

Rev 007 2 Page 13 of 61

- . ~ ~ . . _____ ACCEPTANCE CRITERZ

. Responses which fail to meet the acceptance criteria stated in this section shall constitute unsatisfactory SR results and require immediate notification of the Unit Supervisor at the time of failur The following acceptance criteria shall be demonstrated as required by this procedure:

NOTE:

Step 6 1.1 is used to satisfy the Tech Spec requirement to provide 6258 gpm flow at 105 psid between the reactor pressure vessel (RPV) and primary containmen . CS LOOP II provides at least 6250 gpm flow at 228 psig or greater discharge pressur . The operatlng point for CS Pump 28 and 2D shall be within the acceptance criteria range determined by tke BFN ASME OM Code Inservice Test (6s')

Program. Specifically, CS Pump 2R differential pressure shall lie within ihe range 229 to 279.8 psid and CS Pump 2D differentia! pressure shall lie within the range 220.6 to 269.6 psid while either ptirno is operating within the ASME OM Code-specified discharge flow rate rang CQRE SPRAY SYS II PEST VALVE 2-FCV-75-50 closir?g stroke time does exceed 30 second . CS LOOP II discharge piping is ven!ed 6. The CORE SPRAY MINIMUM FLOW VALVE. 2-FCV-75-37, optjiis on lowering flew and closes on rising flo I

' Rev 0012 Page 14 of 61 I ACCEPTANCE CRITERIA (Continued)

..

6. The following valves shall comply with the ASME OM Code IST Program acceptance criteria stipulated below:

intended function by noting at least a 40 psi drop in CS Pump 28 discharge pressure when CORE SPRAY SYS II MIN FLOW VALVE 2-FCV-75-37 is opened while CS Pump 28 is operating at or near rated condition liverin~2FO gj I Valve shall open Sufficiently to perform its intended function by noting at least a 10 psi drop in 6s Pump 2 D discharge pressure when CORE SPRAY SYS II MIN FLOW VALVE 2-FCV-75-27 i s ii opened w!-ile CS Pump 2D is operatir,g at or near rated condition Valve shall close sufficiently to prevent backflow by noting CS Pump 28 alone is capable of

.. . -.

6.2 Responses which fail to meet the acceptance criteria stated in this section shall constitute unsatisfactory TSW results and require irnrnediate notificaticr7 of the Unit Supervisor at the time of failur . Core Spray NE Room Cooler autorra:ically starts when either Core Spray Pump 26 or 2D star .3 Steps which determine the above criteria are designated by (AC) zext to the irii!ia i blan i - -

i BFN UNIT 2

._

~-

INITIALS 70 PRQCEBURE STEPS NOTE:

1) All checks, verifications of manipulations by UO will be performed at Panel 2-9-3 unless otherwise note ) [NRCIC] The area fan cooler thermostat is disabled during this procedure and a TWM LCO may be entered VERIFY the following initial conditions are satisfied:

_-7 I : K-7. Precautions and limitations in Section 3.0 have been reviewe 'c ::.

7. Prerequisites listed in Section 4.0 2re me .2 OBTAIN permission from Unit Supervisor to perform this tes ..

- ,

. ,

.

.. [NCO 89-0216-0021 I_..._-.

us 73 [NRCIC] NOTIFY Unit 1 , Unit 2, and Un!t 3 Unit O~erators(UO)prior to the start of this procedure [RPT 82-16, LER 259/8232j _- RECORD the date and time started, reason for test, and plant conditions I

on Attachment I,Surveillance Procedure Review For I I

BFN UNIT2 1 CORE SPRAY FLOW RATE LOOP II

~~

2-SR-3.5.1.6(CS 11)

Rev 0012

.

.

Page 16 of 61

-

INITIALS 70 E, EBURE STEPS (Continued)

NOTE§:

(1) Lifting of the CS pump room therrnostat power lead will enable testing of the NE Corner Room CS pump rootn cooling fan operation due to CS pump operation regardless of room temperature, This operation disables the area fan cooler thermostat. 2-SI-4.2.-60FT(II)must be performed on the thermostat to return it to an operable statu (2) [NRC/C] When the GS pump roorn cooler thermostat is disabled, an LCO associated with the area cooler fan logic may be entered (TR Table 3.3.3.2-1). If the GS pump room cooler thermostat is disabled for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a CS System LCO i r a y result if CS Loop II is required to be operable. Step 7.51 verifies 2-S-4.2.B-6OFT(II)was performed and records the time. 'The perfwner shall notify the Unit Supervisor prior to exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or if it becomes apparent the LCO trine limit cannot be met. [NCO 89-0216-002]

(3) JB 8790 is located in the NE Corner Room at Elevation 541.5'

'. ' (4) Steps 7.5, 7.6, and 7.7 may be performed concdrrently PERFORM the following:

751 LIFT, TAPE, and LABEL the power lead (Cable 2ES3308-1 Wire No 8BX) at Terrninal 2 in JB 8790 for 2-TS-64-73 7. [Nf?C/C] RECORD below !he time step 7.5 1 is completed:

[NCO 89-0216-002]

. ,

.! ' , ,

Time

.

._

70 PROCEDURE STEPS (Continued)

NOTES:

(1) The work described in the following step shall be performed using skill of craf (2) The pressure gauges installed in the following step are located at Instrument Rack 2-LPNL-25-60 in the Northeast Corner Room at Elevation 519 of the Reactor Buiidin (3) The centerline of Pressure Gauge A must be at the same elevation as 2-PI-75-32 centerline when readings are take (4) The centerline of Pressure Gauge E must be at the same elevation as 2-PI-75-41 centerline when readings are take . INSTALL a temporary pressure gauge and associated process connection hardware at each test tee connection for 2-PI-75-32 and 271-7541 per Attachment 5 . Section __ .. .

.. IM 7 54 TAG the gauge on 24'1-75-32 test tee as PRESSURE GAUGE A and the gauge on 2-PI-75-42 test tee as PRESSURE GAUGE B -

iM NOTE:

The BVOM test connections which may be insta!!ed in the foilowing step are loczted at Panel 2-9-19 in the Auxiliary Instrument Roo . IF the 16s is used to collect CS LOOP II f!cw rate data. THEN TEMPOWaWlhY INSTALL DVOM test concections at Fiow Modifier 2-FM-75-49 per Attachment 3; (Otnewise NiA.) __

I :VI

.

INlTlALs 70 PROCEDURE STEPS (Continued)

NOTE:

1) Section 7.6 for CS Loop l l venting is not required to be performed if venting was performed by this procedure, 2-01-75, or 2-SR-3.5.1.1(CS II)within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and no activities have occurred to invalidate the venting. If this is applicable: Section 7.6 can be NIA with performance of the applicable 2-SR-3.5.1.1(CS 11)) 2-01-75, or 2-SR-3.5.1.6(CS 11)

documented in the post-test remark ) 2-HS-75-72 and 2-SHV-75-92 are located on the north wall of Elevation 593 of the Unit 2 Reactor Building near column lines RI I- .6 VERIFY CS LOOP I1 discharge piping is vented as follows:

7. OPEN CS SYSTEM I1 HP VENT TELL-TALE FLOW SOLENOID VALVE 2-FSV-75-72 by depressing ar.d holding push-button 2-HS-75-92.

.. .

7. SLOWLY OPEN CS SYSTEM I I TELL-TALE VENT SHUTOFF VALVE %-SHV-75-7 I

___..-

7. VISUALLY VERIFY a steady stream of water exitipg the drain side Of CS SYSTEM / I HP VENT TELL-TALE FLOW SOLENOIQ VALVE 2-FSV-75-72 after venting four minute !,AC)

764 CLOSE CS SYSTEM II HP VENT TELL-TALE FLOW SOLENOID VALVE 2-FSV-75-72 by releasing pushmutton 2-HS-75-72 _. - -

765 CLOSE CS SYSTEM I1 TELL-TALE VENT SHUTQFF VALVE 2-SF 75-72 t I -

9. VERIFY CS LOOP I1 discharge pipirig static pressure is greater than 39 psig as indicated by CORE SPRAY SYS II DISCH PRES>

Indicator 2-PI-75-49 on Panel 2-9- IAC!

-

70 PROCEDURE STEPS (Continued)

NOTE:

The RHR SEVlCE WATER PUMP RUNNING annunciators (Reference Attachment 5) will alarm in steps 7 7 1 through 7 7 4 for those pumps not already running and wilt collectively be verified in step 7 7 5 PLACE RHRSW pumps in operatim as fellows to prevent an RHRSW automatic start:

7. IF RHRSW Pump 83 is capable of being started AND is N8T operating, THEN START WHRSW PUMPI33 using kand-switch C-HS-23-88N2 (Othenvise MIA.)

772 IF RHRSW Pump B3 is capable of being started AND is NOT operating. THEN START RHRSW Pump D3 using hand-switch O-HS-23-94N2 (Otherwise NIA ) _-

INITIALS PROCEDURE STEPS (Continued)

7. IF RHRSW Pump A I :

(1) 1s capable of being starte &

(2) 1s lined up to supply EEC AND I _

(3) !s operating, THEN START RHRSW Pump A4 using hand-switch O-HS-23-1W2:

(Otherwise NIA.).

774 IF RHRSW Pump 6 1 (1) Is capable of being starte ANB (2) Is lined up to supply EECW (3) Is operating. THEN START RHRSW Pump C l using hand-switch O-HS-23-8N2:

(,Otherwise NIA.).

775 VERIFY afd ACKNOWLEDGE alarnis recelvcd an Panel 0-9-23 ___ -

INITIALS 70 EDURESPEPS (Continued)

____

CAUTION The static pretest suction pressure must be 1 psig a CS pump can be operated NOTES:

Pressure indicators are located on instrclrnent Rack 2-LPNL-25-60 which is in the NE Corner Room at Elevation 51 9'.

Ensure Pressure Gauges A and 5 are mounted at the same elevation as 2-PI-75-32 and 2-PI-75-47 when readings are take VERIFY CS Pump 28 and 2 8 pretest static suction pressures are adequate and RECORD belo i Pressure

. Gauge A Pressure Gauge -. B

~~

'i i

INITIALS 70 PROCEDURE STEPS (Continued) VERIFY Core Spray NE Room Cooler Fan is OFF by observing fan motor power on light at 480V RMOV Board 2 8 , Compartment 8B, Electrical I

,-.-..-

Board Room 2B (El. 593') is extinguishe ,, ,

__.I__ .

7.10 CHECK Core Spray NE Room Cooler Fan is U T operating by noting no air flow from the duct louvers above CS Pump 2 8 and 2D can be felt while standing next to each pum ,, .~

-

-.-....-

. -

7 II START CS Pump 2 8 using hand-switch 2-HS-75-33A ~ _--

7.12 CHECK the following annunciators on Panels 2-9-3 and 0-9-23-8 are in alarm e CORE SPRAY SYS II PUMP B START (2-XA-55-3F, window 1) _... .

e RHR OR CS PUMPS RbiNNlNG ADS BLOWDOWN PERMISSIVE (2-xA-55-3C, WlndOW IO) ~ -.

e CORE SPRAY PUMP 26 RUNNING (O-XA-55-23C, window 37) -. --

NOTE:

Indication of the required flow in the following step demonstrates 2-CKV-75-537B is open and 2-CKV-75-5378 and 2-CKV-75-57GD are closed sufficiently to allow CS Pump 2B to operate at rated conditions.

7.13 THROTTLE 2-FCV-75-50 using CORE SPRAY SYS I/ TEST VALV HS-75-5QA to obiair: an average CS LOOP ll flow of 3200 g@mas indicated by 2-FI-75-49. [BFPER 98-0004734-0GQj -. .- . At>)

, ' 7 Date ? /i 7

- INITIALS 70 PROCEDURE STEPS (Continued)

NOTES:

(I) Satisfactory completion of step 7.14 verifies 2-CKV-75-570B has opened sufficiently to perforrn its intended functio (2) Step 7.14 may be repeated as required if data acquisition steps are incomplete due to timinglcoordination problems while observing gauges and indicating light .14 PERFORM the following to verify 2-CW-75-578B operation:

7.1 THROTTLE 2-FCV-75-59 using CORE SPRAY SYS II TEST VALVE, 2-HS-75-50A to obtain a CS LOOP II flow of ' i.,-

approximately 1800 gprn as ind cated by 2-FI-75-49 ____.

.~

7.1 VERIFY COKE SPRAY SYS II MIN FLOW VALVE 2-FCV-75-37 opens (4C)

7.1 THROTTLE 2-FCV-95-50 using CORE SPRAY SYS I I TEST VALVE 2-HS-75-508 to obtain a CS LOOP I I flow of approximately 2800 gpm as indicated by 2 - f 1-75-49 7.1 VERIFY CORE SPRAY SYS II MlN FLOW VALVE, 2-FCV-75-37 IS closed ~

I AC)

7.1 THROTTLE 2-FCV-75-58 using CORE SPRAY SYS il TEST VALVE. 2-HS-95-508 to obtain a CS LOOP II flow of approximately 3200 gpm 3s indicated uy 2-Fl-75-49.

._ -

~'

INITIALS PKQCEQURE STEPS (Continued)

7.1 RECORD below CS Pump 28 discharge pressure measured locally by 2-P1-75-35 on 2-LPNL-25-60:

!

/[-,: 1...-

CS Pump 2 8 Diseh Press i ,~; ~ PSi!3 I 9.1 NOTIFY Operations personnel to monitor CS Pump 2 8 discharge pressure measured locally by 2-PI-75-35 on 2-LPNL-25-60 for minimum reading when CORE SPRAY SYS II MIN FLOW VALVE 2-FCV-75-37 reaches open position ( ,L 7.1 CONTBNLBQUSLY HOLD the CORE SPRAY SYS II MIN FLOW VALVE 2-FCV-75-37, 2-HS-75-37A in the OPEN position until ,' _.!,/_,

step 7.14.1 _ _ _ ~

7.1 RECORD below the lowest CS Pump 23discharge pr,assure measured locally by 2-PI-75-35 on 2-LPNL-25-60 CS Pump 28 Bisch Press -_ Fs:Q 7.14.1Q RELEASE hand-switch 2-HS-75-37A to the AUTO positior:

70 PROCEDURE STEPS (Continued)

7.14.11 GALCULATPE the change in CS Pump 2B discharge pressure as stipulated below:

Initial Discharge Pressure 3 0 0 -Ps'g (Step 7.14.6)

Lowest Discharge Pressure - I

Pig (Step 7.14.9)

!

!I i..:

Discharge Pressure Change -

-

-__~- > _I psid NOTE:

Verification the discharge pressure change meets the acceptance criteria stipulated in the following step provides positive confirmation 2-CKV75-570B ias opened sufficiently to perform its intended design functio .14.12 VERIFY the discharge presscre change recorded in step 7 1 4 . 1 1 is greater than or equal to 10 psi !AC)

9.14.13 CHECK CORE SPRAY SYS l i MIN FLQW VALVE 2-FGV-75-39 is closed by noting valve position indicating !igl?ts above hand-switch c

2-H S-75-37 __

- ..

.~ -. ~~. ~-

CQRE SPRAY FLOW RATE LOOP II 2-SR-3.5.1.6(CS II)

Rev 0012 Page 26 of 61 I Date . / L ; / c 5

-

INITIALS 70 EF EDURE STEPS (Continued)

NOTES:

(2) Pressure indicators 2-PI-75-32 and 2-PI-75-35 are located in the NE Corner Room at Elevation 519' on local instrument Rack 2-LPNL-25-6G (2) The centerline of Pressure Gauge A must be at the same elevation as 2-PI-75-32! centerline when readings are taker,.

(3) CS LOOP II flow rate data may be obtained by using the ICs (e.g., ICs console display, Panel 2-9-5 ICs digital displzy) or by manual means using a DVOM to locally monitor Flow Modifier 2-FM-75-49 input. If the ICs is used, the average flow rate shall be determired by a visual estimate of the average of the ICs display. A nbmerical calculation is not required.

(4) ASME OM Code data collection requires the CS pumps be operated at a stable reference flow rate when discharge pressure readings are taken. Whi!e the flow rate may be an average value, the Unit Operator (UO)should maintain the flow rate as close as possible to the required 3200 gum (349 mV;) flow rate.

7 15 PERFORM the following ASME OM Code pump flow and pressure measurements for CS Pump 2 8 operation 7.1 IF the ICs is available to obtain CS LOOP 11 flow rate data, THEN PERFORM the following, ( O t h e w s e NIA )

7.15. CHECK no gross instrument channel failures have occurred by noting the ICs-displayed flow rate is within 100 gpm of the flow rate shown on CORE SPRAY SYS I1 FLOW Indicator 2-Fl-75-4 ~

7 15 1 2 THROTTLE CORE SPRAY SYS tl PEST VALVE 2-FCV-75-50 using hand-swtch 2-tiS-75-50A to obtain an average ICs display reading I>f 3203 gpm

[BFPER98-004734-000] __ .- -~ ~--

ilj 70 PROCEDURE STEPS (Continued)

7.1 IF the ICs is available to obtain CS LOOP li flow rate data, THEN THROTTLE 2-FCV-75-50 using CORE SPRAY SYS II TEST VALVE, 2-HS-75-50A to obtain an average reading of 349 mV at the BVOM installed at Panel 2-9-1 9. (Otherwise N/ [BFPER98-004734-000]

7.1 AFTER stable conditions are obtained, THEN RECORD CS Pump 28 suctior: pressure from Pressure Gauge A:

.I'

CS Pump 2B suction pressure (M&l"E) psig 7.1 RECORD the pressure reading at 2-PI-75-35 below:

CS Pump 2i3 discharge pressure ' , i

'/

, psig 7.1 CALCULATE CS Pump 28 differential pressclre as follows and VERIFY the differential pressure meets the acceptance criteria:

~. ~ .-

Discharge Pressure psig (Step 7.15.4)

-.

.4 Suction Pressure I .- psig (Step 7.15.3) j

.

Differential Pressure - . psid

. ~

~~ - .-

229 to 279 8 psid

- . .- - - __ i

INITIALS 70 PROCEDURE STEP'S (Continued)

7.1 RECORD the following data for CS Pump 28: .~ .

Pararneterilndicator

~

__ Measured Acceptance

. Criteria

-

Core Spray Sys I I Flow, or -7,i I . gpm 3200 gpm

.s.a v e m

____I ICS. D i s w .

. .. -

Core SpraySys II Disch Pressure 2-PI-75-48 NIA LPanel 2 - 9 - x ... ~ : . ~

!.

~.~~

7 ~ __

Core Spray Sys II Flow '. .- mV 349 mV (! 0 5 mV)

2-FM-75-49 (Panel 2-9-19) j (.averaye) (average)

ow Transm Core Spray pump is Motor Current 2-1-75-33

- (Panel

~ ~ ~ 2-9-31

~ . _ _ . ~.

4kV Shutdown Bd C i. 1 .,

t r L .

VAC

.. ~~~ VolW&anel.~~ 9-23-8). -

NOTE:

j) NIA reading for 2-FM-75-49, CS SYS ll FLOW or7 step 7 15 6 if BVBM or, step 7 5 5 was installed

,

Date i./J/O INITIALS 70 PROCEDURE STEPS (Continued)

NOTE:

Verify vibration readings obtained in the next step are not in ALARM status on the portable M&TE before shutting down the CS Pump. This will preclude having to restart the pump in case of questionable data. The ALARM h i t s are NOT acceptance criteria, but are used to flag questionable data.

7.16 [QMDS] NOTIFY EM to perform 0-TI-230 vibration measurements as I L.-e~.

indicated on Attachment 4 for CS Pump 28.

7.17 RECORD CS Pump 2 8 vibration readings below:

r . ~ B ~ ~ - ~ . ~

MEASURED ---1 I* POINT

.~ I VALUE-I I AA I 1 inisec i I

-

EM 7 ? 8 VERIFY the Core Spray NE Room Cooler Fan is ON by OBSERVING fan motor power on light at 480V RMBV Board 28, Compartment 88, Electrical Board Room 2E (El 593) I S illuminated _ _

7.19 CHECK the Core Spray NE Room Cooler Fan is cperaticg by:

NOTING air flow from the duct louvers above C S Pwnp 28 and 2D can be felt while standicg next to each pum , ({c:

~ .~ ~

CORE. SPRAY FLOW RATE LOOP B I 2-SR-3.5.1.6(CS II)

Rev 0012

- .. ~ , __ .. ..

90 PRBCBURE STEPS (Continued)

,!LLL 7 20 START CS Pump 2D using hand-switch 2-HS-75-42A.

7 21 CHECK the following annunciators on Panels 2-9-3 and 0-9-23-8 are in alarm:

>,

CORE SPKAY SYS I1 PUMP D START (2-XA-55-3F, window 2). L ac-.

e CORE SPRAY PUMP 2D RUNNING (0-XA-55-23D, window 37) ~-

.,

7 22 THROTTLE 2-FCV-75-50 using CORE SPRAY SYS II TEST VALVE, 2-HS-75-50A as necessary to obtain a CS LOOP II flow of 6250 to C 6350 gpm as indicated by 2-Fl-75-49 or ICS display - -

7 23 RECORD the following CS LOOP II parameters and VERIFY parameter values are within acceptance bavd

__~

Parameterhdieator -

Care Spray Sys II Flow 2-Fl-75-49 anel 2-9-3) or ICs Core Spray Sys II Disch Pressure 2-PI-75-48Eanel

__ ., 2-3-3)

~ ~ _ _ ~ ~

~

' " '

-

iACj 7 24 FULLY OPEN 2-FCV-75-50 using CORE SPRAY SYS I I TEST VALVE 2-HS-75-5C Date I INITIALS 70 &URE STEPS (Continued)

NOTES:

REFER TO Illustration 1 Process for Stroke Timing Valves per the ASME OM Code 7 25 CLOSE and TIME 2-FCV-95-50 using CORE SPRAY SYS II TEST VALVE, 2-HS-75-50A and RECORD the stroke time below

___ ~.~~~

2-FCV-75-50 Closure Time (Seconds) 7 Normal Maximum 23.0 to 3 .~~

. ~~

7.2 VERIFY the time recorded is within the maximun! value listed !AC)

7.26 IF the stroke time measured in step 7 25 is within the maximum value

....

listed and outside the normal range, THEN PERFORM the following; (Othenvise NIA this section) [BFPER971386]

7 26 1 OPEN the 2-FCV-95-50 using CORE SPRAY SYS II TEST VALVE, 2-HS-75-50A ~ __

7 26 2 CLOSE and TIME 2-FCV-75-50 using CORE SPRAY SYS I I TEST VALVE 2-HS-75-50A and RECQRD the restroke time on Attachment 7 [BFPER971385] ~-

7.2 VERIFY the restroke time recorded on A:tachment 7 is within the maximum values iisted. [BFPER971386] -. . :AC!

7.27 CHECK CORE SPRAY SYS II MIN FLOW VALVE 2-FCV-75-37 is open by noting valve position indicating lights above hand-switch 2-HS-75-37A. s _ _ ~. .~.

INITIALS 70 PROCEDURE STEPS (Continued)

NOTE:

The following steps requires the operator to hold until conditions have stabilized as determined by Engineering ur Components Organization for obtaining Temperature profiling and Thermograph .28 [F Temperature profilingflherrnography is being performed on the 28 Core Spray Pump. THEN PERFORM the following: (Othenuise N/A this section.)

7.2 THROTTLE 2-FCV-75-50 using CORE SPRAY SYS II TEST VALVE, 2-HS-75-50A as necessary to obtain a CS LOOP II flow of 6250 to 6350 gpin as indicated by 2-Fl-75-49 or 16s display 7.28.2 HOLD until Engineering/ Components group obtains steady state Temp profiIe/Therrnography for the 28 Core Spray pump

..., .~

bearingslcomponent ~.

7.2 WHEN Notified b y Engineering/ Components group that steady state Temp profilenhermography data is obtained for tPe 28 Core Spray pump, THEN CONTINUE with this procedure 7.2 CLOSE -FCV-75-50 using COKE SPRAY SYS ll TEST VALV HS-75-50A -

7.2 CHECK CORE SPRAY SYS I/ MIN FLOW VALVE 2-FCV-75-37 is open by noting valve position indicating ligrits above hand-switch 2-HS-75-37 _ - -

INITIALS 70 PROCE-DURE STEPS (Continued)

7 29 STOP CS Pump 28 using hand-switch 2-HS-75-33 I 7.30 RESET and CHECK the following annunciators on Panels 2-9-3 and 0-9-23-8 are reset:

CORE SPRAY SYS II PUMP 8 START (2-xA-55-3F, window 1).

CORE SPRAY PUMP 2B RUNNING (0-XA-55-236, window 37).

NOTE:

Indication of required flow in the following step demonstrates 2-CW-75-537B is open and 2-CKV-75-5378 and 2-6KV-75-570B are closed sufficiently to allow CS Pump 2D to operate at rated conditions 7.31 THROTTLE 2-FCV-75-53 using CORE SPRAY SYS II TEST VALVE, 2-HS-7.5-5BA to obtain a 6s LOOP II flow of 3280 gpm as indicated by 2-Fl-75-4 (ACj

[

~~ ~ .- ~

BFN UNIT2 I! CORE SPRAY FLOW RATE LOOP b l 2-SR-3.5.1.6(CS 11)

Rev 0012

~..~.. . .~ . . .- - .

Date L

____

INITIALS 70 PROCQURE STEPS (Continued)

NOTES:

(1) Pressure indicators 2-PI-75-41 and 2-PI-75-44 are located in the NE Corner Room at elevation 5 1 9 on local instrument Rack 2-LPNL-25-60 (2) The centerline of Pressure Gauge B must be at the same elevation as 2-PI-75-41 centerline when readings are take (3) ASME OM Code data collection requires the CS pumps be operated at a stable reference flow rate when discharge pressure readings are take While the flow rate may be an average value, the Unit Operator (UO)

should maintain the flow rate as close as possible to the required 3200 gpm (349 mV) flow rat .32 PERFORM the following ASME OM Code pump flow and pressure measurements for 6s Pump 2D operation:

7 32 1 IF the ICs is available to obtain CS LOOP I1 flow rate data, THEN c-THROTTLE: 2-FCV-75-50 using CORE SPRAY SYS I1 TEST VALVE, 2-HS-75-50A to obtam an average IC§ display reading cf 3200 gpm. (Otherwise N/A ) [BFPER98-004734-000] s_ .3 IF the IC§ is NOT available to obtain CS LOOP I1 flow rate data, THEN THROTTLE 2-FCV-75-50 using CORE SPRAY SYS II PEST VALVE 2-HS-75-50A to obtain an average reading of 349 mV (LO 5 mV) at the DVOM installed at Panel 2-9-19 (Otherwse NIA ) [BFPR98-004734-000]

--_ -

1 BFN I CORE SPRAY FLOW RATE LOOP Ii 2-SR-3 5 1 6(CS 11)

l... UNIT2 ! Rev 0012 Page 35 of 61 Date INITIALS 70 PKOCEDURE STEPS (Continued)

7.32.3 AFTER stable conditions are obtained, THEN RECORD CS Pump 213 suction pressure from Pressure Gauge B:

CS Pump 2D suction pressure (M&TE) psig 7.32.4 RECORD the pressure reading at 2-Pi-75-44 below:

CS Pump 2B discharge pressure lasig 7 32 5 CALCULATE CS Pump 2D differential pressure as follows and VERIFY the differential pressure mee!s the acceptance criteria

_ I _ psig (Step 7.32.4)

_

plscharge Pressure . .~

- psig (Step 7.32.3)

~ .~ ~ ~ ~ ~~

I

Bate

-.

INITIALS 70 &p . m R E STEPS (Continued)

7 3 RECORD the following data for CS Pump 2D:

. .-

Pararneter/lndicator ... .-

Core Spray ~ y s ' FIOW, or 16s D i s m..~..~ ...

l~

-__-gpm 3200 gpm (average) !

Core Spray Sys II Disch Pressure 2-PI-75-48 P.a?e!Es?l .~

Core Spray Sys I I Flow 2-FM-75-49 (Panel 2-9-19)

--

(average)

Note ___

1 349mV(+0,5mV)

(average)

jElow Transmitter Input)-

~

~

Core Spray Pump 2D Mator Current 2-El-75-42 (Panel 2 - 9 3 ~~ ~

I NIA i 4kV Shutdown Ed D NOTE:

1) N/A reading for 2-FM-75-49, CS SYS II FLOW on step 7 32 6 if DVOM on step 7 5 5 was NOT installed

~ . - ~ ~

L-. .-

CORE S P W Y FLOW RATE LOOP b

___ .- -

I 2-SR-3.5.1.6(CS I I )

Rev 0012 Page 37 of S I

~ ~ ..~ ~ .. i Date PWOCERURE STEPS (Continued)

NOTE:

Verify vibration readings obtained in the next step are not in ALARM status on the portable M&TE before shutting down the CS Pump. This will preclude having to restart the pump in case of questionable data. The ALARM limits are NOT acceptance criteria, but are used to flag questionable data.

___

7.33 [QMDS] NOTIFY EM to perfor,,, 0-TIL230 vibration data measurements as indicated on Attachinent 4 for CS Pump 2D 7.34 RECORD CS Pump 2 D vibration reading bebw

- .-

VlBS MEASURED POINT VALUE AH I inisec inlsec inlsec CH2 inisec

.~ .-

~ ~~ -- ..

2-SR-3.5.1.6(6S II)

Rev 0012 !

Page 38 of 61 i Bate INITIALS PROCEDURE STEPS (Continued)

NOTES:

(1) Satisfactory completion of step 7.35 verifies 2-CKv-75-570D has opened sufficiently to perform its intended function.

(2) Step 7.35 may be repeated as required if data acquisition steps are incomplete due to timingicoordination problems while observing gauges and indicating lights.

7.35 PERFORM the following to verify 2-CW-75-570D operation:

7.3 RECORD below the CS Pump 2 0 discharge pressure rneasured locally using 2-PI-75-44 on 2-LPNL-25-60 CS Pump 2D Disch Press - - psig 7 35 2 NOTIFY Operations personnel to monitor CS Pump 2D discharge pressure measured locally rising 2-PI-75-44 on 2-LPNL-25-60 for minimum reading when CORE SPRAY SYS II MIN FLOW VALVE 2-FCV-75-37 reaches open positio .3 CONTINUOUSLY HOLD the CORE SPRAY SYS II MIN FLOW VALVE, 2-US-75-37A in the OPEN position until step 7 35 5 7.3 RECORD below the lowest CS P u r p 2D discharge pressure measured locally by 2-PI-75-44 on 2-LPNL-25-60, CS Pump 2D Disch Press psig 7.3 RELEASE hand-switch 2-HS-75-37A to the AUTO position

~~ - - ~ ~- -

CORE S P W I FLOW RATE LOOP II 2-SR-3.5.1.6(CS II)

Rev 0012 Date INiTlALS

_ _ I _

70 PROCBURE STEPS (Continued)

9.3 CALCULATE the change in CS Pump 2 5 discharge pressure as stipulated below:

Initial Discharge Pressure Psig (Step 7.35.1)

Lowest Discharge Pressure - Pslg (Step 7 35 4)

Discharge Pressure Change -

- psid NQTE:

Verification the discharge pressure change meets ti-e acceptance criteria stipulated in the following step provides positive confirmation 2-CW-75-57GD

. ... ~. has opened sufficiently to perform its intended design function.

~.

7 35 7 VERIFY the discharge pressure change recorded IS greater than or equal to 10 psid __ (ACj 7.3 CHECK CORE SPRAY SYS li MIN FLOW VALVE 2-FCV-75-37 is closed by noting valve position indicating lights above hand-switch 2-HS-75-37 Date INITIALS PROCEDURE STEPS (Continued)

NOTE:

The following steps requires the operator to hold until conditions have stabilized as determined by Engineering or Components Organization for obtaining Temperature profiling and Thermography.

7.36 IF Temperature profiling/Therrnography is being performed on the 2D Core Spray Pump, THEN PERFORM the following: (Otherwise N/A this section.)

7 36 1 THROTTLE 2-FCV-75-50 using CORE SPRAY SYS II PEST VALVE, 2-HS-75-50A as necessary to obtain a CS LOOP II flow of 3200 gprn as indicated by 2-FI-75-49 or I C s dlsplay 7 36 2 HOLD until Engineering/ Compone:its group obtains steady state Temp profilelvherrnography for the 213 Core Spray pump bearingslconiponent .

7 36 3 WHEN Notified by Engineering/ Coir,ponents group that steady state Temp profilenhermography data is cbtained for the 28 Core Spray pump, THEN CONTINUE with this procedure

- - . - __

I BFN CORE S P W Y FLOW RBTE LOOP I1 2-SR-3 5 9 6(CS I I )

UNIT 2 Rev 0012 i 1 Page 41 of E;I i

Date i-...'

INITIAL_S_ PR:

_ I QURE STEPS (Continued)

7.37 THROTTLE CLOSE 2-FCV-75-50 using CORE SPRAY SYS Ii TEST VALVE, 2-HS-75-50 .38 CHECK the Core Spray NE Room Cooler is operating by noting air flow from the duct louvers above CS Pump 26 and 2 8 can be felt while standing next to each pump (86)

7.39 STOP CS Pump 2D using hand-switch 2-HS-75-42 .40 CHECK CORE SPRAY SYS 11 MlN FLOW VALVE 2-FCV-75-37 is open by noting valve position indicating lights above 2-HS-75-37A -.

7.41 VERIFY the Core Spray NE Room Cooler Fan is OFF by observing fan motor on light at 48OV R K 3 V Board 28, Comoartment 8 8 , EIectrical Board Room 2B (El 593') is extinguished 7.42 CHECK the Core Spray NE Rooin Cooler has stopped operating by noting no air flow from the duct louvers above CS Pump 28 and 2D can be felt while standing next to each pump "

- I_

7.43 RESET and CHECK the following annunciators on Panel 2-9-3 and 0-9-23-8 are reset:

CQRE SPRAY SYS II PUMP B START (2-xA-55-3F, window 2). ~

0 RHR OR CS PUMPS RUNNING ADS BLOWDOWN PERMISSIVE (2-xA-55-3C, window IO) -

o CORE SPRAY PUMP 20 RUNNING (C-XA-55-23D. window 37) ~. ~ -

.. __

CORE SPRAY FLOW RATE LOOP II

.... -...~..-~... ~~

I

!

2-SR-3.5.1.6(CSIlj Rev Q012 Page 42 of 61 Date

.~~. ... __

INITIALS PROCEDURE STEFS (Continued)

NOTE:

The NE Corner Room cooling fan will automatically start if area temperature IS greater than 90 degrees Fahrenheit 7 44 RETERMINATE the power lead (Cable 2ES3308-11, Wire No 8BX) at Terminal 2 in JB 8790 for 2-TS-64-73 and REIN§TALL J E cover __ --

1st 7.45 COMPLETE applicable environrnentai qualification work record entries for terminal block and cabl iM 7.46 COMPLETE aDDlicable

, , environmental atiahfication work record entries for CS pump motor vibration measurements EM NOTE:

The following step shall be performed using skill of craf .47 REMOVE Pressure Gauges A and B installed at test tees on Instrument Rack 2-LPNL-25-60 and RETURN instrunient sense lines to their as-found condition per Attachment 6, Section B IM 7.48 IF a DVOM was used to collect CS Loop II flow rate data, THEN REMOVE the DVOM and associated test probe leads installed a!

Panel 2-9-19 on Flow Modifier 2-FM-75-49 and RETURN the flow modifier to its as-found condition; Othemiise M A .

IM 7.49 RETURN RHRSW Pumps A I , C1, E3 and D3 to their pretest al!gnment in

~..-

accordance with 0-01-69 where applicab;e (Reference step 7 7 ) ~ --

Rev 0012 Page 43 of 61 Date

-

INITIALS 70 PROCEDURE STEPS (Continued)

NOTES:

(1) Lifting of thermostat power lead enabled testing of NE Corner Room cooling fan operation due to CS pump operation regardless of room temperature. This operation disabled the area fan cooler thermostat 2-Sl-4.2.5-6OFT(II) must be run on the thermostat to return it to an operable statu (2) [NRC/C] When the CS pump Room cooler thermostat was disabled, an LCO on the thermostat may have been entered fTR Table 3.3.3.2-1). If the CS pump room cooler thermostat was disabled for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a CS System LCO may have resulted. Step 7.5 lifted the lead and recorded the time. If the time the thermostat was disabled approaches 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the performer shall notify the Unit Supervisor immediately. [NCO 89-0216-6021 7.50 REQUEST IM to perform portions of 2-SI-4.2 B-60FT(ll) whck are

. .

applicable to 2-TS-64-73 functional testing 7.51 [NRUC] VERIFY portions of 2-SI-4.2.B-60FT(ll) which 2re applicable to 2-TS-64-73 functional testing are successfully complete AND RECORD the time of completion below: [NCO 89-0216-0021 Time IM

....

~ .~ ~. ~. . ..~

~ ~

. ~ .~~

-~ .

CORE SPaRY FLOW RAPE LOOP I! !2-SR-3.%.1.6(CSil)

Rev 0012 Page 44 of 61 I Date -

-

INITIALS PROCEDURE STEPS (Continued)

NOTES:

(1) Section 7.52 substeps may be performed in any order.

(2) 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).

7.52 PERFORM the following independent verifications to ensure CS LOOP II has been returned to its pretest configuratioc:

7.5 VERIFY CS Pump 28 is in a standby configuration by checking the green indicating light is illuminated ,white, and red indicating lights above 2-HS-75-33A, and the amber CS PUMP AUTO-INIT LQCKQUT LIGHT (PUMP 2B 2-lL-75-33), are extinguished on Panel 2-9- ~-

IV 7.5 VERIFY CS Pump 2 8 is in a standby configuration by checking the green indicating light is illuminated white. and red indicating lights above 2-HS-75-428, and the amber CS PUMP AUTO-INIT LOCKOUT LIGHT (PUMP 2 8 2-IL-75-42j, are extinguished on Panel 2-9- ~

IV 7.5 VERIFY CORE SPRAY SYS II TEST VALVE 2-FCV-75-50 IS closed using valve position indicating lights above 2-HS-75-508 on Panel 2 - 9 3 -.

'J 7.5 VERIFY CORE SPRAY SYS II MIN FLOW VALVE 2-FCV-75-37 is open using valve position indicating

iyh;s above 2-HS-75-37A on Panel 2-9- IV

Bate

-,

INITIALS PROCEDURE STEP'S (Continued)

NOTE:

Calculation IV consists of verifying arithmc.. 3 for accuracy and arithmetic inputs have been properly transferred between steps within this procedure. IV is required to verify local pressure readengs were correctly recorde .5 PERFORM the following:

7.52. VERIFY calculation performed in step 7.14.11 is correc IV 7.52. VERIFY calculation performed in step 7.15.5 is correc IV 7.52. VERIFY calculation performed in step 7.32.5 is correc IV 7.52. VERIFY calculation performed in step 7.35.6 is correc IV NOTE:

The following step is performed outside the main control room at the CS LOOP I1 vent station which is located ~n the north wall of Elevation 593 of the Reactor Building approximately 8 feet east of column lines R11 - VERIFY CS SYSTEM li TELL-TALE VENT SHUTOFF VALVE 2-SHV-75-72 I S closed Iv

Date

_-

1NlTlAI-S 70 p i:D4JRE STEPS (Continuea)

NOTE:

The following step is performed by IMs in the NE Corner Room at Elevation 519 7.5 INDEPENDENTLY VERIFY the followmg 7.52. VERIFY OPEN 2-PISV-075-0032 PANEL ISOh VALVE TO 2-PI-075-0032 iv 7.52. VERIFY installed, test tee cap for 2-PI-075-0832 IV 7.52. VERIFY OPEN, 2-PISV-075-GO41, PANEL ISOL VALVE TO 2-PI-075-004 :.

IV 75274 VERIFY installed, test tee cap for 2-PI-075-0041 -

IV NOTE:

The following step is performed by IMs in the Unit 2 Auxiliary Instrument Roo .5 IF a DVOM was used to co!led CS LOOP I! flow rate data, THEN VERIFY DVOM and associated test probe !eads installed at Panel 2-9-19 on Flow Modifier 2-FM-75-43 have been removed and flow modifier has been returned to its inservice condition; Othem/ise NI IV

Bate r m INiTlALs 70 PROCEDURE STEPS (Continued)

NOTES:

(1) Steps 7.52.9 through 7.52.15 are NOT independent verifications but are performed by the independent verifier to facilitate procedure performance efficiency.

(2) Steps 7.52.9 through 7.52.14 verify the closing spring for each CS and RHRSW pump operated by this procedure is properly charged upon termination of test activities.

(3) The VO should be contacted to determine which RHRSW pumps were operated. If there exists a question as to whether an RHRSW pump has been operated during performance of the procedure then the breaker charging spring status shall be verified to be conservativ .5 At 416OV Shutdown Board C, Compartment 7, in Electriea! Board Room 2 VISUALLY VERIFY CORE SPRAY PUMP 28 BKR CHARGED amber light is illuminated and closrng spring target indcates CHARGED [11-6-92-068]

7 52 10 At 4160V Shutdown Board D Compartment 8, in Electrical Board Room 2 6 VISUALLY VERBFY CORE SPRAY PUMP 2D BKR CHARGED amber light is illuminated and closing spring target ir?dica!es CHARGED [11-6-92-068]

7.52.11 IF RHRSW Pump A I was operated by this procedure. THEN At 4160V Shutdown Board A, Compartn:ent IO. in Electrical Board Room 1 VISUALLY VERIFY RHR SERVICE WATER PUMP A I BKR CHARGED amber light is illuminated and closing spring target indicates CHARGED; Othenvise NIA. [Il-B-92-068] __ .

Date

- INITIALS 70 PROCEDURE STEPS (Continued)

7.52.12 IF RHRSW Pump C4 was operated by this procedure, THEN At 4160V Shutdown Board B,Compartment 10, in Electrical Board Room 1 VISUALLY VERIFY RHR SERVICE WATR PUMP C1 BKR CHARGED amber light is illuminated and closing spring target indicates CHARGED, Otherwise N/A (Il-B-92-068] ___.._I 7.52.14 SF RHRSW Pump W3 was operated by this procedure, THN At 41EOV Shutdown Board C, Compartment 5, in Electrical Board Room 2 VISUALLY VERIFY RHR SERVICE WATER PUMP 83 BKR CHARGED amber light is illuminated and closing spring target indicates CHARGED; Qthewise N!A. [11-B-52-068] --

i -. .

7.52.14 IF RHRSW Pump D3 was operated by tbis procedur THEN At 4?60V Shutdown Board D:Compartment 19. in Electrical Board Room 2 VISUALLY VERIFY RHR SERVICE WATER PUMP 0 3 BKR CHARGED amber light is illuminated anel closing spring target indicates CHARGED; Otherwise N/A. [11-8-92-068] ~~ .

.

7 52.15 RSET and VERIFY annunciators CORE SPRAY PUMP 2B RUNNING (O-XA-55-4!C, window 37';

CORE §PRAY PUMP 2D RUNNING (C-XA-55-41D, windcw 3 7 )

e Any applicable RHR SERVICE WATER PUMP RUNNING at Central Diesel Information Center Panei 25-41 are reset --

'-

Bate

%__

INITIALS 70 PROCEDURE STEPS (Continued)

7.53 COMPLETE Attachment 1, Surveillance Procedure Review Form, up to Unit Supervisor revie .54 IF any valves required restroking (i.e. stroke times were recorded on Attachment 71, THEN PROVIDE a copy of Attachment 7 to the Duty Maintenance Manager to deliver to the ASME IST Program owner, (Otherwise NlA.).

[BFPER999 3861 7.55 NOTIFY Unit 1, Unit 2, and Unit 3 Unit Operators this procedure is complet .56 NOTIFY Unit Supervisor this procedure is complete Attachment 1, Surveillance Procedure Review For Attachment 2, ASME OM Code lnservice Testing Review For Attachment 3, Flow Modifier 2-FM-75-49 Froct Vie Attachment 4, Vibration Data Point Locations Attachment 5, Annunciators Affected by Surveillance Procedure Performance Attachment 6 : Temporary ASME OM Code Pressure Indicators Installation and Rer:oval Attachment 7 . ASME OM Code Res:roktt T/i?e Record Form Illustration 1, Process for Stroke Timing Va:ves Per the ASME OM Cade END OF TEXT

~~ ~...

CORE §&RAY FLOW RATE LOOP II 2-SR-3.5.1.6(CS 11)

Rev 0012 Page 50 of 61

.

ATTACHMENT I (Page 1 of 2)

SURVEILLANCE PROCEDURE REVIEW FORM REASON FOR TEST: DATETTIME STARTEB_~//~~L, 2 ~'6312 Scheduled Surveillance DATE/TlME COMPLETED J -System inoperable (Explain in Remarks) PLANT CONDITIQNS ,VI~ ~ I T I P-" Maintenance (WO No.03- L b j J - u / )

Other (Explain in Remarks)

PBE-TESTBEMARKS: 7 . , r i i : , - c d PIV\T c r > - C C V ' ' ' ~ ~ - Y C A6-1&,2---

.~

,b\,+<V k"&, 'A Jp-. r l i 1-7 I.-.-__

AT-I-ACHM%NT1

.. (Page 2 of 2)

Surveillance Procedure Review Form (Continuation Page)

PERFORMED BY:

-_

Name (Print) Name (Signature)

ATTACHMENT 2 (Page 1 of 2 )

ASME OM CODE INSERVICE TESTING REVIEW FORM COMPONENT ACCEPTABLE NOT NIA OR TESTED STEP RANGE

. ..- ACCEPTAM NOT TESTED CS Pump 2B

~1 .~.

Differential step 7.1 i' 1 ~.

Pressure CS Pump 2D Differential step 7.3 Pressure

ATTACHMENT 2 (Page 2 of 2)

ASME OM CODE INSERVICE TESTING REVIEW FORM COMPONENT - :LY NOT NIA OR TESTED

-. - --

STEP &+._I('iABLE ACCEPTABLE NOT TESTED 2-FCV-95-50 step 7 25

-

L ,.

d 2-FKV-75-537B step 7.13 ri and step 7.31 2-CKV-75-537D step 7.13 ~

and step 7 31 2-CKV-75-57CB step 7.14.12 and step 7 31 2-CKV-75-5700 step 7 13 and steu 7 35.7 NOTE:

Any valve with a stroke time exceeding it5 maximum value shah! be declared inoperable. Any valve having a stroke time outside of its normal stroke time range shall be res:roked immediately. If the restroke time is also outside the r,ormal range, then the recorded stroke times shall be evaluated by engineering within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> of the completion of this pror,ed::re for acceptability.

Date Received: ___

ASME OM CODE REVIEWER (Components) - DATE ASME OM Code Data entered in Sl(s)2-SI-3 1 1 and 2-SI-3 2 1 ANI1 REVIEWER - - __ BATE .

BFN CORE S P W Y FLOW RATE LOOP I1 2-SR-3.5.1.6(CS II)

UNIT 2 Rev 0012 Page 54 of 61

. ~~

~ ~ ~ ~ ~~~~ ~ -~ ~. . . -. ~-

&E

~~~-~ SPRAY FLOW RATE LOOP 81 2-SR-3.5.1.6(6S II)

Rev 0012 Page 55 of 61

.... . ~.

ATTACHMENT 3 (Page 1 of 1)

FLOW MODIFIER 2-FM-75-49 FRONT VIEW NOTES:

(1) Connect (+) DVOM probe lead to IIP (+) flow modifier test jack and (-)

DVOM test probe lead to IIP (-) flow modifier test jack per the diagram shown' below.

(2) Set DVOM to 2 V DC voltage range and observe voltage readings for any signs of anomaly. Voltage readings should be 248 mV plus or minus process noise voltage variations which should NOT exceed +I- I' mV when a no flow condition is presen ASME OM CODE DVOM TEST POINTS FOR MEASURING FLOW RATE

. - - -__.- - .-. ___ -

i~BFN

~- - CORE .§PRAY FLOW RATE LOOP S I 2-SR-3.5.1.6(CSII)

i - UN'T2 .~ .. ...... .

Rev 0012 Page 56 of 64 DATA POINT AA DATA POINT A (AH1, AH2)

DATA POINT B (W

~- ~ .~ ~ ~ ~ ~. ~~~ -~ ~- ~ . ..

CORE SPRAY FLOW RATE LOOP I1 ' 2-SR-3.5.1.6(CS11)

Rev 8012 Page 57 of 61

..... - . .

ATTACHMENT 5 (Page 1 of 1)

ANNUNCIATORS AFFECTED BY SURVEILLANCE PROCEDURE PERFORMANCE

.. ..

1 UNlD

...

DESCRIPTION

....

LOCAT ION

..... ......

RHR OR cs PUMPS~UNNINGABS I-- 28-3

........... ~

-. -PERMISSIVE

.

CORE SFRAY SYS. . .I1. . .PUMP

- ,.

.

....

B START .....

.....

I Window 4 -.

...........

.

t I 2-xpI-55-3F 2-9-3 2-9- .... CORE SPRAY SYS I1 PUMP CORE SPRAY SYS II DISCH

......

....

B START......

PRESS ..... HIGH

~-

1 Window 2

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

2-XA-55-3F Window 30

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

~~ ~

~~

0-9-23-7

...... RHR SERVICE WATER PUMP A I RUNNING

............. ..... _ _ _ _ . ~ ~

I Window34 t 0-XA-55-23B 0-9-23-7-. . RHR SERVICE

. _ _ _ RHR

. .~. SERVICE ...

.........WATER

.

WATER PUMP 83 PUMP C1 .... RUNNING ~.

~. RUNNING ..___...

t

. ~ ~ . Window ~~~.~ 34 0-XA-55-23C

...... .____

~~~~~

t . 0-9-23-8.. .- CORE SPRAY . . . . . .25

. . . . PUMP . . .RUNNING

........ ~

!

I Window 37 0-XA-55-23D 1 I

0-9-23-8 0-9-23-8

.~..___ .

.~

RHR SERVICE ....

CORE SPRAY PUMP.2 0 .RUNNING WATER PUMP D3 ..RUNNING

....

~

1 Window 34 ~ ~

Q-W-55-23D I

... . _ _ _

. .

RHR SERVICE WATER . PUMP ........ A I RUNNING ..... I Window34 0-XA-55-4A~n3

_RHR _ ~ _ SERVICE

. ~ . _ _ WATER _ I . ~ . PUMP Cl RUNNING Window 34 1 0-25-41 CORE SPRAY PUMP 25 RUNNING

__..________..~ .... ___

0-XA-55-41 B

...

Window37 0-XA-55-41C I j

RHR. SERVICE . WATER PUMP B 3 RUNNING ..... ~.

Window 34~- .

Q-XA-55-41D I

1.-

0-25-41

-__-

RHR SERVICE WATER PUMP

.....__

..........

CORE SPRAY

-~ ,.__~

~

PUMP 20 RUNNING

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

. D3 RUNNING

~

...

4 . . -.Window 34 ~ - - ~ _

0-XA-55-41 D Window 37 -4

.

~ ~ ~ ~

~

.__ 1 .___ . 1 This attachment provides the UO with a listing of main control room and lecal alarms th3t ~VJIII1-;e affected by performance of this procedure, This attachment is for information o-1).

~ ~

2-SR-3 5 1.6(CS II)

Rev 0012 Page 58 of 61 ATTACHMENT 6 (Page 1 of 2)

TEMPORARY ASME OM CODE PRESSURE INDICATORS INSTALLATION AND REMOVAL Bate PERFORM the following steps to install the temporary pressure gauges described by step 5.2.2 at Instrument Rack 2-LPNL-25-0068:

A. 1 CLOSE 2-PISV-095-0032 PANEL ISOL VALVE TO 2-PI-75-32 REMOVE 2-PI-075-0032 test tee cap below gauge A. 3 CONNECT temporary pressure gauge at 2-Pi-075-0032 test tee OPEN 2-PISV-095-0032 PANEL ISOL VALVE TQ 2-PI-75-32 ana VENT temporary gauge connection as required

..-. CLOSE 2-PISV-075-0041 PANEL ISQL VALVE TO 2-PI-75-41 REMOVE 2-PI-095-0041 test tee cap located below gauge CONNECT temporary pressure gauge at 2-PI-075-0841 test tee OPEN 2-PIS-075-0041 PANEL ISOL VALVE TO 2-PI-75-47 and VENT temporary gauge connection as required

... ~ ~

~

CORE S P r n Y FLOW RATE LOOP 81 ATTACHMENT 6 I 2-SR-3.5.1.6(CS II)

Rev0012 Page 59___

of 61 ]

(Page 2 of 2)

TEMPORARY ASME OM CODE PRESSURE INDICATORS INSTALLATION AND REMOVAL Date INIT1ALS PERFORM the following steps to remove the temporary pressure gauges installed at Instrument Rack 2-LPNL-25-0060:

CLOSE 2-PISV-0754032 PANEL lSOb VALVE TO 2-PI-75-32 DISCONNECT temporary pressure gauge at 2-PI-075-0032 test tee INSTALL 2-PI-075-0032 test lee cap OPEN 2-PISV-075-0032 PANEL ISOb VALVE TO 2-PI-75-32 CLOSE 2-PISV-095-0041 PANEL ISOL VALVE TO 2-PI-75-41 -

DISCONNECT temporary pressure gauge at 2-PI-075-0041 test tee - -

INSTALL 2-PI-075-0041 test tee cap OPEN 2-PISV-075-0041 PANEL ISOL VALVE TO 2-PI-75-47

r - - -- - ~ -- __ ~ _ _ ~ ~ ~__

T

~~ ~ ~

BEN CORE SPRAY FLOW RATE LOOP I! 2-SR-3.5.1.6JCS II)

UNiT 2 Rev 0012 Page 60 Qf 64 ATTACHMENT 7

. (Page 1 of I)

ASME OM CODE RESTROKE TIME RECORD FORM F - A F V E

~.

uN'DFzL 2-FCV-95-50 (CLOSE')

STROKE TIME 23.0 to 3 'FI IMEASURED .--MUM---

STROKE TIME STROKE

.~-

(SECTIME 30.0'

t VALVE PER SR TIME II

~~

1 -

MAXIMUM LIMIT?

L-f ENGINEERING

--

E i

TIME NORMAL RANGE?

WIT HI^

I REVISE REFERENCE STROKE TIME AND .-

INITIATE J

I II I r

PROCEDURE _c...~_.~

~

CHANGE RESTWOKE TIME WITHIN d

MAXIMUM LIMIT?

t r ---A TEST STROKE TIME-ACCEPTABLE? I ,

.

NO3 I ~ ~  !

ENGINEERING

-~

RESTROKE TIME DATA TO

EVALUATES STROKE ENGINEERING

~, -~96 HOURS, WITHIN

JPM NO.A3.1.f?

REV. NQ. 0 PAGE 2 0 F 8 BROWNS FERRY NUCLEAR PLANT JOB PERFORMANCE MEASURE REVISION LOG Revision Effective Pages Description Number Date Affected of Revision 0 9/9/02 ALL NEW

JPM NO.A3. REV. NO. 0 PAGE 3 0 F 8 BROWNS FERRY NUCLEAR PLANT JOB PERFORMANCE MEASURE RQ I_

SRO __ BATE:

JPM NUMBER: A3.1R TASK NUMBER: ADMlN TASK TITLE: NiA WA NUMBER: 2.3.10 K'A RATING: R O L SRO: 3.3

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

TASK STANDARD: R EVlEW A RADIOLOGICAL SURVEY MAP TO DETERMINE IF A TASK CAN BE COMPLETED WITHOUT EXCEEDING AN INDIVIDUAL'S E EXPOSURE LIMITS.

LOCATION OF PERFORMANCE: SIMULATOR - PLANT - CONTROL ROOM REFER%NCE§1~ROCEDUKES NEEQED: NIA VALIDATION TIME: CONTROL ROOM: LOCAL:

MAX. TIME ALLOWED: __ (Completed for Time Critical JPMs only)

PERFORMANCE TIME: I I_ CONTROL ROOM - LOCAL COMMENTS: - I

_ ____.-- __

s_____-- _---

Additional comment sheets attached? YES -_ NO RESULT§: SATISFACTORY ~ UNSATISFACTORY EXAMINER SIGNATURE: BAT EXAMIN 2

JPM NOA3.1 .R REV. N PAGE 4QF a

-____

EXAMINERS KEY (Page 1 of 2)

INITIAL CONDITIONS: You a r e the Scpervisor of a Browns F e r r y eniployee w h o h a s o b t . a i n e d a n accuniu1at:ive y e a r l y r a d i a t i o n dose sf 630 mrem.

INITIATING CUES: Given t h e a t c a c h e d r a d i o l o g i c a l . s u r v e y map of > A R e a c t s r Water Cleanup Pump room, QEsEgM;l?Thl ~~o!~y~~wo~kfr__car~

corn@ e t e t he-aL?&Le ~2~55 rr.- !Li~t&qgt . - e : ~ = ~ ~ . G rrij

~:</

T V R a~- chj~ ~ ~ ~ ~ t ~ ~ t ~ ~ , ~ , ~ ~ ~ ~ r ~ . ~ . , . ~

The j o b r e q u i r e s r e p l a z i n g tne c l e a n u p pLIRiF v e n t . T h e worker will have t o trar:spcrt. weldinn l e a u s and a cut.tijng torck. along t h ~ i d e s i 2nat.ea wal.kway- (marked a s EQLX FMEKT MCVSXEKT P A W ) , cut tile e x i s t i n g v e n t v a l v e off, weld i.n ::he n e w vent villve, and t r a n s p o r t :.he equipment back t 3 t.he C-zoce s t e p c f f p a d .

T r a n s p o r t . i n g t h e r e q u i r e d equipment ~ i l . 1 require approximatel.:,.. 3 miniltes, e a c h way. Once c h e pumg ,<rea or. t h e survey map i s r e a c h e d (marked a s P ~ m pVeiit) , it w i i l r e q i l r e 32-r.&r~gJg?t..c ri9place che punip vent . Tke a t t a c i i e d Rad?:o:.ot:ic-..? S c r v e y Map

, c o c t a i n s t.he i n f o r m a t i o n y0.u must i.riterpretr t o s u c c e s s f u l l l r complete t . h i s J P M .

JPM NQ.83.1 .R REV. NO. 0 PAGE 5 0 F 8 I

I : i ClOOO c~oclo

JPM NO.A3. REV. NO. 0 PAGE 6 0 F A E

. -M I N E R ' S KEY

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

These questions may be asked to clarify or expand on the required knowledge items at the discretion of the examiner.

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

What is the Yearly TVA Administrative Dose Limit?

1000 Mr

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

According to information on the survey map, Would the worker be required to wear protective clothing in this area?

Yes, this is a C-Zone

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

Where would the worker find the specified protective clothing requirements?

On the RWP

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

According to information on the survey map and radiological posting requirements, Is this area a Radiation-or t&h Radiation Area?

High Radiation Area

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

When the worker is working on the Pump Vent, what is the general area dose rate?

408 rnr/hr

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

Where in the room would the worker stay to avoid unnecessary exposure whi!e waiting in the room. E, Where is the Low Dose Waiting Area Near the Step Off Pa BROWNS FERRY NUCLEAR PLANT JOB PERFORMANCE MEASURE CANDID=§ HANDOUT (Page 1 of 2)

bf$&a INITIAL CONDITIONS: You a S u p e r v i s o r of a Rrcwns F e r r y employee wna h a s obt.ained a n a c c u m u l a t i v e y e a r l y r a d i a t i o n d o s e of 8 O C mrcr INITIATING CUES: G i v e n t h e att.ac:her! r a d i c l o g i c a l s u r v e y map of 3~

R e a c t o r Water Cleanup Pump room, EZTERVINE i f the worker c a n complete t h e a s s i g n e d t a s k i n t h e rcom witjlout exceedi:ig h i s / h e r TVA a d m i n i s t r a t i v e y e a r l y d o s e limi The j o b r e q u i r e s rep1acir.g t h e c l e a n u p pump v e n t . The worker xi11 nave t o t r a n s p o r t w e l d i n g l e a d s and a c u t t i n g t o r c h a l o n s ?I?F d e s i g n a t e d walkway ( n a r k e d a s EQUIPMENT MOVEMENT PATH), cut t l i o e x i s t i n g v e n t valve off, weld i n t h e new v e n t v a l v a , a x 3 t r a n s p o r t t h e equipment back t c t h e C-zone s t e p o f f p a d .

T r a n s p o r t i n g t h e r e q u i r e d equipment w i l l r e q u i r e a p p r o x i n a t e l y 8 m i n u t e s , e a c h way. Once t n e 2um.p a r e a on t h e s c r v e y map i s r e a c h e d (marked a s Pump Vent) ~ i t will r e q u i r e -.3 0 .-m i n u t e s t o r e p l a c e t h e pcmp v e n t . The a t t a c h e d Radlologica: Survey Map con t a i n s t h e i rif o rnia t ion you m us t i n t e r p re t t o succ e s s f u 11y complete t h i s ,SP ~1000 e1000 5.000 e1OCO 5 el000 BA-IGenerd Area

l ~1000

[

si-Hot Spot Q-a 10,000

WANSTANDARB SPP- PROGRAMS AND RADIOLOGICAL CONTROLS Rev. 4 PROCESSES Page 40 of 33 RaBCON (or RSO) shall investigate TLD and secondary

.. dosimeter reading discrepancie Individuals shall inform RADCON (or RSO) if they lose their TLB or their secondary dosimeter is lost, damaged, exhibits an unexpected response or is off-scal NVLAP accredited TLD results are normally used as the official record of radiation exposur .4.1.6 Administrative Dose Lev&.*

Occupational radiation dose limits at W A N facilities are consistent with the limits given in 10 CFR 2 Occupational radiation dose limits at TVAN facilities are consistent with the limits given in 10 CFR 2 Administrative dose levels (ADLs) to be used as guidelines for maintaining doses below regulatory limits have been established within TVAN and shall be observed for routine work. This program is not applicable to minors or declared pregnant women. The TVAN Administrative Dose Level Program is surnrnarized in Table 1 below:

TABLE 1 ADMINISTRATIVE DOSE LEVEL PROGRaM-Requirement _ I_

=ation

-

to excee~~~signalures)

Statement of current year dose and previous l o t appllnbk years dose signed by individual Up tu i.0 TERE NRC FORM-4 or equivalent to document Not applicable (or 3 0 LDE or 10 SDE) all current year and prsvious years dose sources

-. -

equivalent -

To exceed 1.0 ?ERE Same as above Sie Radiologicala i d Chernisii'-

(or 3.0 LDE or 10 SDE) all Control Manageriffs0 sources

___.__

TO exceed ~ . ~ - T E D Eall

' ~orrn-4information must be ver:fied and a Sie Radiologicaland Chemistry sources Planned Special Exposure initiated Control ManageriRSQ, Plant Manager', and Site VP'or SED as

- appropriate Site Radiologicaland Chemistry

_-

Form4 must b@ verified Contrui Manager:RSO. Plant Manaoer'. and SRe VP' or SED as I, At non-nuclear plant sites, this will be the RSQ's immediate superviso . At non-nuclear plant sites, this will be the applicable TVA V . Authorizations for a planned special exposure will only be considered in an exceptional situation when alternatives that might avoid the dose estimated to result from the planned special exposure are unavailable 01 impractica . Total effective dose equivalent should not exceed I N rem, where N equals the individual's age in years at last birthday. without the

-- authorization signatures delineate BROWNS FERRY NUCLEAR PLANT JOB PERFORMANCE MEASURE JPM NUMBER: A4.1R TITLE: CLASSIFY THE EVENT PER THE REP (LOSS OF SHUTDOWN COOLING)

TASK NUMBER: S-000-EM-21 SUBMITTED BY: BATE:-

VALIDATED BY: . DATE:__

APPROVED: DATE:-

TRAINING PLANT CONCURRENCE: DATE:-

OPERATIONS 0 ExaminationJPMs Require Operations Training Manager or Designee Approval 2nd Plant Concurrence

JPM NO. A4.1R REV. NO. 0 PAGE 2 OF 2 2 BROWNS FERRY NUCLEAK PLANT JOB PERFORMANCE MEASURE REVISION LOG Revision Effective Pages Description Number Bate Affected of Revision 0 New

.

JPM NO. A4.1 R REV. NO. 0 PAGE 3 OF 2 2 BRQWNS FERRY NUCLEAR. PLANT JOB PEKFORMANCE MEASURE OPERATOR:

RO __ SRO X DATE:

JPM NUMBER: A4.1R TASK NUMBER: S-000-EM-21 (SROONLY)

.

I TASKT1TLE:CLASSIFY THE EVENT PER THE REP (LOSS OF SHUTDOWN COOLING)

WA NUMBER: 2.4.38 WA RATING: R O 2 . 2 SRO: 4 . 0

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

TASK STANDARD: C lassify the event as an ALERT and. perform associated SRO actions.

LOCATION OF PERFORMANCE: SIMULATOR X PLANT - CONTROL ROOM I 3a **..v*s REFEWENCESlPKOCEDURES NEEDEB: EPlP 1, R E V X EPIP 3. REVZfgq VALIDATION TIME: CONTROL ROOM: 25:OO LOCAL: N/A MAX. TIME ALLOWED: __ (Completed for Time Critical JPMs only)

PERFORMANCE TIME: __ CONTKOL ROOM - LOCAL NIA COMMENTS:

Additional comment sheets attached? YES - NO RESULTS: SATISFACTORY UNSATISFACTORY I__

SIGNATURE: DATE:

EXAMINER

-

BROWNS FERRY NUCLEAR PLANT JOB PERFQRMANCE MEASURE CANDIDATE HANDOUT INITIAL CONDITIONS: You are the SHIFT MANAGER. Following a 400 day run, Unit 2 was in Mode 4 with the first pass of head bolt de-tensioning complete and was placed in Shutdown Cooling at 0800. At 1030 shutdown cooling isolated due to a failed pressure switch. 2- AOI-74-1 has been implemented. Unit 2 conditions are as foilows:

Reactor Power Shutdown, All rods inserted Reactor Level +80 inches on Shutdown Vessel Flood Range Reactor Pressure 3 psig Moderator Temperature 215 Degrees Fahrenheit BW Pressure 0 psig DW Temperature 105 degrees F

.. BW Radiation 88-90-256 Normal DW Leakage Rate Torus Temperature Torus Level None 88 degrees F-3 inches

-

Torus Pressure 0 psig NOTE: No release in progress Wind speed is 5 rnph from the SW INITIATING CUES: Determine the EVENT CLASSIFICATION and PERFORM THE REQUIRED actions until the emergency centers are staffed and you are relieve JPM NO. A4.1W

.- REV. NO. 0 PAGE 5 OF 2 2 BROWNS FEKKY NUCLEAR PLANT JOB PERFORMANCE MEASURE EXAMINERS COPY INITIAL. CONDITIONS: You are the SHIFT MANAGER. Following a 400 day run, Unit 2 is in Mode 4 with the first pass of head bolt de-tensioning complete and was placed in Shutdown Cooling at 0800. At 1030 shutdown cooling isolated due to a failed pressure switch. 2- AOI-74-1 has been implemented. Unit 2 conditions are as follows:

Reactor Power Shutdown, All rods inserted Reactor Level +80 inches OR Shutdown Vessel Flood Range Reactor Pressure 3 psig Moderator Temperature 21 5 Degrees Fahrenheit DW Pressure 0 psig DW Temperature 105 degrees F DW Radiation RR-90-256 Normal BW Leakage Rate None Torus Temperature 88 degrees F Torus Level -3 inches

..

Torus Pressure 0 psig NOTE: No release in progress Wind speed is 5 mph from the SW lNlTlATlNG CUES: Determine the EVENT CLASSIFICATION and PERFORM THE REQUIRED actions until the emergency centers are staffed and you are relieve JPM NO. A4.lR REV. NO. 0 PAGE 6 OF 2 2 START TIME:

Performance Step : C r i t i c a l X Not Critical-REFERS TO EPIP 1 to classify emergency event Standard:

Candidate REFERS TO EPIP-1 Section l i s 1.5A, LOSS OF DECAY HEAT REMOVAL and deciares an ALERT based on reactor moderator temperature CANNOT be maintained below 212' F whenever Technical Specifications require Mode 4 conditions or during operation in Mode v SAT-UNSAT-NIA- COMMENTS:

Performance Step : C r i t i c a l X Not Critical-

.

IMPLEMENTS EPIP-3 ALERT Standard:

SHFT MANAGEWSEI3 RECOGNlZESilMPLEMENTS an ALERT per EPIP-3.

SAT-UNSAT-N/A- COMMENTS:

JPM NO. A4.1 R REV. NO. 0 PAGE 7 OF 22 WSTRUGTiONS

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

Petformance Step : Critical-Not Critical& If all Emergency centers ARE STAFFED Then notify the following that an ALERT Emergency Classification has been issued and EPIP 3 is being implemented, and continue in this procedure at Step 3.4. If ail Emergency Centers ARE NOT STAFED, Then NIA this step and continue in this procedur CECC, T§C, O X , CONTROL ROOMS, PLANT PA ANNOUNCEMENT CUE: EMERGENCY CENTERS ARE NOT STAFFE >

Standard:

NA §TEP 3.1 & continue in procedure

-

SAT-UNSAT-N/A- COMMENTS:

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

JPM NO. A4.1 W REV. NO. 0

..

PAGE 8 OF 2 2 3.2 Notification of the Operations Duty Specialist (ODs) & Ernerqency Responders Note: The OD§ should be notified within 5 minutes after the emergency event is declare ......................................................................................................

Performance Step : CriticalXNot Critical-3. Complete Attachment A (Initial Notification Form).

Standard:

ATTACHMENT A is completed as shown in Examiners Copy, Page 7 of 1 .

(INFORMATION GIVEN IN INITIAL CONDITIONS & INITIATING CUES EXCEPT EAL DESIGNATOR) NOTE: THIS IS GENERIC INFORMATION FOR DESCRIPTION OF EVENT--ALL THE EXACT INFORMATION IS NOT REQUIRED FOR ACCEPTANCE UNDER BRIEF DESCRIPTION OF EVEN SAT-UNSAT-NIA- COMMENTS:

. .

JPM NO. A4.1R REV. NO. 0 PAGE 9 OF 2 2 Performance Step : Critical& Not Critical-3.2.2 Activating Emergency Response Organization (ERO)

3.2. if ongoing/anticipated on-site security events may present a d<nger to the emergency responders, Then consult with Nuclear Securit .2. E ongoinglanticipated events present a danger to emergency responders, Then direct the Unit 1 Unit Operator to make notifications per Attachment B and select "Staging Area" as the option for the Emergency Paging Syste .2. E there are no ongoing/anticipated danger to emergency *

responders, Then direct the Unit I Unit Operator to make notifications per Attachment 3 and select as applicable, "Brill" or "Emergency" as the option for the Emergency Paging System.

Standard:

DIRECTS Unit 1 Operator to make notifications per Attachment B (3.2.2.3)1 SAT-U NSAT-N/A- COM MENTS:

.

JPM NO. A4.1R

. . ~ ~ i REV. NO. 0 PAGE 10 OF 2 2

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

Performance Step : Critical& Not Critical-3. the ODS and Provide the information from Attachment Note: Utilize the direct ring-down OBS phone when making this notification or as applicable dial direc OBS Telephone Numbers 5-1-751-1700,2495-If the ODS cannot be reached within I 0 minutes, Then contact the State of Alabama directiy by requesting the Rad

.

Health Duty Officer at:

Day Shift 8 a.m.-5 Holidays-Weekends-Offshifts 9-1-334-206-5391 9-1-334-242-4378 Standard:

-

Attempts Notification of the OB ..

CUE: The ODS Cannot be contacted by phone due to communications problems, no estimate on repair tim Notifies the State of Alabama directly by one of the number listed abov SAT-UNSAT-NIA- COMMENTS:

JPM NO. A4.1 R REV. NO. 0 PAGE 11 OF 2 2 Performance Step : Critical- Not Critical&

3. Fax a copy of Attachment A to the ODS for confirmation of information or if the state is contacted directl ODS Fax AL Rad Health Fax 5-752-8620 9-1-334-206-5387 CUE: FAXING TO THE STATE WILL BE SIMULATE >

Standard:

SIMULATES faxing a copy of Attachment A to the stat SAT-U NSAT-NiA- COMMENTS:

..

3.2.5 Receive confirmation call from the ODS (to verify notification of the State of Alabama)(NA this step, if the state was contacted directly).

I1 CUE: FHREE MINUTES AFTER THE PERFORMER FAXES ATTACHMENT STATE RAD HEALTH OFFICER CALLS AND CONFIRMS RECEIPT OF FAX.

.

JPM NO. A4.IR REV. NO. 0 PAGE 12 OF 2 2 NOTIFICATION OF SITE PERSONNEL Performance Step : Critical Not C r i t i c a l 2 3. Make the following P.A. announcement:

THIS IS (NAME), SHIFT MANAGE AN ALERT HAS BEEN DECLARED ON UNIT HAVE ASSUMED THE DUTIES OF SITE EMERGENCY DIRECTO REPORT TO YOUR ASSIGNED EMERGENCY RE§PONSE FACILITY AT THiS TIME!

Standard:

MAKES P. A. announcement as above.

SAT-UNSAT-N/A- COMMENTS:

JPM NQ. A4.1R REV. NO. 0 PAGE 13 OF 2 2 CAUTION: Bo not initiate Assembly and Accountability if:

1 A severe weather condition exists or projected'on-site, such as a Tumad ~

2. An on-site security risk condition exists that may present a danger to site personnel during the assembly/accountability process.(Consult with Nuclear Security) ACCOUNTABILITY Performance Step: Critical Not Critical X 3. E the emergency situation warrants an Assembly, Accountability, Phen -

implement EPIP-8, Appendix C, concurrently with this procedure. (N/A STEP IF NOT APPLICABLE)

3.4.2 If the emergency situation does not warrant an Assembly, Accountability at this time, Continue to assess the situation, implementing EPIP-8 when necessar Standard: .

ADDRESSES Accountability and at candidate's discretion may or may not implement the accountability section EPIP- SAT-UNSAT-N/A- COMMENTS:

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

EXAMINERS NOTE: If candidate chooses NOT to initiate Assembly and Accountability then continue at Step 3.5, OFFSITE ,. DOSE ASSESSMENT,

.~ page 20 (of this JPM): JPM NO. A4.1 R REV. NO. 0 PAGE 14 OF 2 2 EPIP 8 APPENDIX C Page 1 of 3 SHIFT MANAGERBITE EMERGENCY DIRECTOR -ASSEMBLY AND ACCOUNTABILITY ACTIONS The following appendix shall be utilized by the Shift ManagedSite Emergency Director (SMISED) or designee for the purpose of conducting site assembly and accountability action ....................................................................................................

Performance Step: Critical Not Critical X The SMlSED has determined that conditions require the activation of the assembly and accountability siren system and proces Standard:

ENTERS initials and tim SAT-UNSAT-N/A- COMMENTS.

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

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

Performance Step: Critical Not C r i t i c a l 3 NOTIFY ... Nuclear Security (NS) at extension 3238 or 2219 that:

A. The assembly and accountability sirens will be activated hmediately AND B. NS should implement EPIP-8, Appendix Standard:

NOTIFIES NS assembly and accountability sirens will be activated and DIRECTS NS to implement EPIP 8, Appendix SAT-UNSAT-N/A- COMMENTS:

__ .

I --

JPM NO. A4.1R

. ~. REV. NO. 0 PAGE 15 OF 2 2

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

Performance Step: Critical Not Critical X NOTIFY.. . Radiological Control (RADCON) at extension 7865 that:

A.The assembly and accountability sirens will be activated immediatel AND B. RADCON should implement EPIP-8, Appendix Standard:

NOTIFIES PADCON assembly and accountability sirens will be activated and DIRECTS

.

WDCON to implement EPIP 8, Appendix SAT-UNSAT-N/A- COMMENTS:

-_

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

+

Performance Step: Critical Not Critical X MAKE... a public address announcement similar to:

"Attention all plant personnel, the site assembly and accountability process has been initiated. All personnel report immediately to your assigned assembly areas."

(REPEAT)

Standard:

MAKES P.A. announcement SAT-UNSAT-N/A- COMMENTS:

JPM NO. A4.1 R REV. NO. 0 PAGE 16 OF 2 2

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

Performance Step: Critical Not Critical X ACTIVATE ... the assembly and accountability siren Standard:

ACTIVATES the assembly and accountability sirens by DEPRESSING red button on 0-CNTL-244-6398 on Shift Manager's desk. (Critical) ENTERS initials and time. (Not Critical)

SAT-U NSAT-N/A- COMMENTS :

v Performance Step: Critical Not Critical X

-- cycle and silence .

WHEN ... the Assembly and Accountability Sirens have completed the 3-minute MAKE.. . a PA announcement similar to:

"Attention all plant personnel, the site assembiy and accountability process has been initiated. All personnel report immediately to your assigned assembly areas."

(REPEAT)

Standard:

MAKES P.A. announcement SAT-IINSAT-N/A- COMMENTS:

JPM NO. A4.ZR REV. N PAGE 17 OF 2 2 If at any time during the assembly and a bility process RADCON determines that radiation guidelines for an assembly area($) have been exceeded, request NS to re-locate affected personnel to another assembly area or evacuate affected personnel off-

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

Performance Step: Critical Not Critical X NOTIFY... Central Emergency Control Center (CECC) Director either by the direct ring-down telephone in the TSC or at extension 751-161 w OR If the CECC Director can not be reached, notify the Operations Duty Specialist (ODS) at extension 751-1700 that:

. .

A. The assembly and accountability sirens have been activate AND

.... B. BFN EPIP-8 is currently being implemented for assembly and accountabilgy, Standard:

NOTIFIES the Operations Duty Specialist (ODS) at extension 751-1700 that:

A. The assembly and accountability sirens have been activate AND B. BFN PIP-8 is currently being implemented for assembly and accountabilit SAT-UNSAT-N/A- COMMENTS:

)CUE: CECC HAS NOT YET BEEN STAFFED. Communication problems with

JPM NO. A4.1 R REV.NO 0 PAGE 18 OF 22 CCOUNTABILITY ARE SUCCESSFULLY COMPLET Performance Step: Critical Not Critical X WHEN... Notified by NS that the assembly and accountability process has been complete THEN .... MAKE a public address announcement similar to:

Attention all plant personnel, the site assembly and accountability process has x been completed. All personnel remain in your assigned assembly area (REPEAT)

Standard MAKES P A. announcement

._

SAT-U NSAT-N/A- CQ MME NTS: .

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

Performance Step: Critical Not Critical X VERIFY with the SMISEB, that conditions at this time require an order to evacuate all non-emergency response personnel from the Owner Controlled Are Standard:

ENTERS NIA SAT-U NSAT-N/A- COMMENTS:

JPM NO. 84.1 R

-.... REV. NO. 0 PAGE 19 OF 2 2

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

Performance SteE Critical Not Critical X I IF ... Conditions at this time, DO require an order to evacuate all non-emergency response personnel from the Owner Controlled Are THEN ... Initiate Appendix F of this procedure (EPIP-8).

Standard:

Evacuation not required for an ALERT, candidate does not require evacuation of non-emergency response personne SAT-UNSAT-N/A- COMMENTS:

-

1I. IF... Conditions at this time, DO NOT require an order to evacuate all non- .

.

emergency response personnel from the Owner Controlled Are THEN..Exit this procedure. Re-enter this procedure at Appendix F when it has been determined by the SM/SED that conditions require an order tQ evacuate all non-emergency response personne JPM NO.A4.1 R REV. NO. 0 PAGE 20 OF 22

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

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

3.5OFFSITE DOSE ASSESSMENT Performance Step : Critical- Not Critical&

3.5.1 Evaluate the need for offsite dose assessment (NIA STEP IF NOT APPLICABLE)

CUE: CECC IS NOT OPERATIONA .5. When offsite dose assessment is required obtain the

.

information from the CECC when operationa .5. If the CECC is not operationai, contact the TSC, when staffed or the M D C O N Shift Supervisor and request the implementation of EPlP 14, for dose assessment.

Standard:

SHIFT MANAGEWEED addresses the OFFSITE DOSE ASSESSMENT, may request activation of RABCON VANS in accordance with EPIP-14.

SAT-UNSAT-N/A- COMMENTS:

JPM NO. A4.1R

--. REV. N PAGE 21 OF 2 2

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

Performance Step : C r i t i c a l X Not Critical- NOTIFICATION OF THE NRC 3. Notifv the NRC immediately or within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and if requested by the NRC maintain an open and continuous communications channe . Note: Utili+e the Emergency Notification System (ENS) when making this notification. Dial the first number listed on the sticker affixed to the ENS telephone, by dialing 9-l-"The Ten Digit Number listed on the ENS E

telephones". the number is busy, THEM select in order, the alternate numbers until a connection is achieved. No access codes require Standard:

NOTIFIES NRC on simulator red phon SAT-U NSAT-N/A- COMMENTS:

I

. -_ . -

FkAMINER'S CUE: NRC does not require an open-line at this tim JPM NO.A4.1R REV. NO. 0 PAGE 22 OF 2 2 PERIODIC EVALUATION OF THE EVENT I CUE: THE EMERGENCY CENTERS ARE STAFFED AND THE PLANT MANAGER ( S I T E EMERGENCY DIRECTOR) I S HERE TO RELIEVE YO END OF TASK STOP TIME:

EMERGENCY EPIP-1 CLASSIFICATION SECTION I1 EVENT CLASSIFICATION MATRKX 1.0 REACTOR PROCEDURE P

-

DESCRIPTION Valid MAIN STEAM LINE RADIATION HGII-HIGE alatni, KA-90-135C OR Valid 00 PKKIT?EATMDJ"' RADIAlTON HIGH alaml, NA-90-1 F A IPER4TING CONDITION:

.Model - M& 3

.M&2 Reactor moderator tempmature CANNOT be maintained below 212" I: whenever Tehnical Specificationsreqnire Mode 4 conditions or dlmng operations in Mode 5. '

OPERATING CONDITION:

-M&4-M d e 5 Suppression Pool temperatwe, level and W V pressure CANNOT be maintained inthe safe areaofCnrve 1.5- OPERATLVG CONNu~l?ON:

-Model -Male3 I/lodc2 1.0 REACTOR PAGE 19 OF 207 REVISION 30

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT EMERGENCY PLAN r m L E M E m r x G PRQCEDLRE EPIP-3 ALERT REVISION 29 PREPAREDBY T W CORNELIUS PHoh%: 2038 RESPONSE3I.E ORGANIZAlTOK EMERGENCY PREPAREDNESS APPROVED BY JEW LEWIS DATE: 04/02/21303 EFFECTIVE DATE 04/07/2001 LEVEL OF USE: REFERENCE USE QIJAKITY-RELATED

REVISION 3OG Procedure Number: EPIP-3 R e v i s i o n Nurrker: 2 9 Pages A f f e c t e d : 3,5,6,8,11 D e s c r i p t i o n o f Change:

IC-30 This change is being conducted to incorporate the management of NRC Commitment changes as prescribed in the correspondence from site licensing RIMS R08000214113, and to human factor the notification and fullow-up notification form Page 2 change to step 3.2.1 involves human factoring the Notification Form Titl ~

Page 6 - changes involves removing the NRC Comniitment Brackets to step requiring the review of PORC and the human factoring of applicable step Page 7 - change involves human factoring attachment title and modifying information to ensure consistenq with NRC guidanc Page 8 change involves adding information regarding the support of the Unit 1 Operator in staffing the ER Page 9 - IJpdated information for the Unit Operator to use during the ERO s w i n g proces Page IO - change involved adding a clarify statenlent concerning the appropridte use of the Follow-up Notification For I C 31 EPIP-3, revision 26 is being issued to incarporate changes regarding assemhly and account~bility actions. All actions to initiate the accountability and evawatian processes are nuw located in EPIP-8. The revision additionally standardizes telephone numbers, and POKC reviews. This revision also adds clarification for the aL7ions taken by the Unit 1 Unit Operator during their stifling of the ERO proces Page 3 added a statement to the caution information regarding securiq threat. Clarifitd steps 3.4.1

-- ~

and 3.4.2 to implement EPIP-8 regarding actions to be taken for asSeMbly~acCOUntabll1~

and evacuatio Page 6 standardize Alert procedure ciosnre inforinatio Page 8,9 - Clarify actions taken by the Unit 1 Unit Operator during tlc notification attachmen IC-32 EPP-3, revision 21 is being conducted to incorporate changes regarding actions to be taken when dangerous conditions exist on site that mould require the assembly of the ERO at the staging are Additionally page 3 and 5 were revised to update telephone information regarding the Ofice of Radiation Contro Page 2 - change instruct the SED when to direct the Unit 1 Unit Operator to assembly the ERO at the staging are Page 4 revision adds clarification to the caution note regarding on-site security conditions for

~

asseiiirly/accoiintabiIi Page 8 revision adds option for staging area IC-13 EPIP-3, revision is being conduct lo change the procedure reference for Dase Assessment from EPIP-14 to EPIP-13. Page 4 of this procedure is be revise IC-34 EPIP-3: rev. 29 is being condnct to standardize record Wention @age 6) and revise the notification f o m to include NRC Terminology from RIS 2002-16 for normal and ahnormal rclcases @age 8 &

11). Additionally the revision will provide a place to doannent the time and EM, Designition when centers are staffed (page 2). Attachment section was renumbered (page 5 ) .

ALERT BROWNS FERRY EPIP-3 NUCLEAR PLANT RRPOSE Provide for timely notification o f appropriate individuals or organizations when the Shift Manager/Site Emergency Director (SED) has determined by EPIP-I that an incident has occurred which is classified as an ALER .2 Provide for periodic evaluation ofthe current situation by the Shift ManagerlSED to determine whether the ALERT should be terminated, continued, or upgraded to a more serious classificatio .0 SCOPE This procedure appiies to emergency events that are classified as Alert by EPP-1. Emergency Classification Procedure.

...

PAGE I OF 11 KEVISIOW 0029 I

ALERT BROWNS FERRY EPIP-3 NUCLEAR PLANT

._~ :

3.0 INSTRUCTIONS Date: --- I /

3.1 Hall Emergency Centers -4RESTAFFED, Notify the following that a ALERT Emergency Classification was declared at Time: -9 EAL Designator ,and EPIP 3 is being implemente Then continue in this procedure at Step CECC Control Rooms 0 IMTIAlS TBlE TSC Plant PA Announcement 0 OSC This is WE,Site Ernerpncy Diroder. an Mort ha3 been Jcclarcd at BFN.we are cuirently implementing EPIP-3. Smdby fur further updat If all Emergency Centers ARE NOT STAFFED, Then N/A this step and continue in this proc.edur .2 Notification of the Operations Duty Srxxialist (ODs) & Emervencv Responders Note: The ODs be notified within 5 minutes afier the emergency event is

.~ declare .2.1 Complete Attachment A (Initial Notification Form).

s1 TBle 3.2.2 Activating Emergency Response Organization (ERO)

3.2. If ongoinglanticipated on-site security events mg present a danger to the emergency responders, IrnIALS TIM&

~ Then consult with Nuclear Securit .2. zf ongoinglanticipated events present a danger to BNITLUS TlME emergency responders, Then direct the Unit 1 Unit Operator to make notifications per Attachment B and select Staging Area as the option for the Emergency Paging Syste .2. zf there are no ongoinganticipated danger to IMTW TIME emergency responders, Then direct the tJnit 1 Unit Operator to make notifications per Attachment B and select as applicable, Drill or Emergency as the option for the Emergency Paging Syste PAGE 2 OF 11 RE\IISION 0029 I

ALERT BROWNS FERRY EPTP-3 NUCLEAR PLANT 3.0 INSTRUCTIONS (CO~TJNUED)

3.2.3 Nctifv the ODS and Provide the information from Attachment Note: Utilize the direct ring-down ODS phone when making this notification or a5 applicable dial direc ODS Telephone Numbers - 5-751-1700,or 2495-If the ODS cannot be reached within 10 minutes, Then contact the State of Alabama directIy by requesting the Ofice of Radiation Control at:

Dav Shifl8 a.m. - 5 p.m.(CentraJ) aolidavs-Weekends-Off-Shi~s Primary: 9-1-334-206-5391 Montgomery State Trooper Post Backup: 9-1-800-582-1866 9-1-334-242-4378 3.2.4 Far a copy ofhttachment A to the ODS for confirmation of information or state if the state was contacted directly rMTI T m ODs Fax Ofice of Radiation Control Fax 5-751-8620 9- 1-334-206-5387 3.2.5 Receive confirmation call from the ODs (to verify notification of the State of Alabama)(NA this step, if the 'rim's 'rnm state was contacted directly).

3.3 NOTIFICATION OF SITE PERS0Nh .3.1 Make the following plant P.A. announcement:

1.Wl TIME THIS IS (NAME), SHIFT MANAGER. A ALERT HAS BEEN DECLARED ON UNIT -. HHAW ASSUMED THE: DUTIES OF SITE EMERGENCY DIRECTOR. REPORT TO YOUR ASSIGNED EMERGENCY- RESPONSE FACILITY AT THIS TIM PAGE 3 OF 1I REVISION OOZY 1

ALERT BROWNS FERRY EPIP-3 NJCLEAR PLANT 3.0 INSTRUCTIONS(CONTIMJED)

CAUTION: Do not initiate Assembly and Accountability if:

1. A severe weather condition exist/projected on-site, such as a Toniad . An on-site security risk condition exists that may present a danger to site personnel during the assemblp/accountability process (Consult with Nuclear Security).

3.4 ACCOKJNTABILATY 3. If the emergency situation warrants an Assembly, Accountability, Then implement EPIP-8, Appendix C, IhrrJALS TIhlE concurrently with this procedur (N/A STEP IF NOT APPLICABLE)

3. zfthe emergency situation does not warrant an Assenibly, Accountability at this time, Continue to assess the situation, implementing EPIP-8 when necessar .5 OFFSITE DOSE ASSESSMENT 3.5.1 Evaluate the need for offsitedose assessmen INlllALS TIMI?

@/A STEP IF NOT AFKICABLEI)

3.5.1.1 When offsite dose assessment is required obtain the information from the CECC when operationa .5.1.2 Hfthe CECC is not operational, contact the TSC, when staffed or the RADCON Shift Supemisor and request the implementation of E P P 13, for dose assessmen PAGE4OF11 REVISION 0029 I

ALERT BROWNS FERRY EPIP-3 NUCLEAR PLANT k-.

3.0 INSTRIJCTIONS (CONTMUED)

3.6 NOTIFICATION OF THE NRC 3.6.1 -the NRC immediately or within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and if requested by the NWC maintain an open and continuous INITIALS TIME communications c.hanne Note: the Emergency Notification System (ENS) when making this notification. Dial the first number listed on the sticker affixed to the ENS telephone, by dialing 9-1-The Ten Digit Number Listed on the ENS Teleohone If the number is busy, Then select in order, the alternate numbers until a connection is achieved. No access codes are require .7 PERIODIC EVALUATION OF THE EVENT 3.7.1 Continue to Evaluate the event using EPIP-I as conditions warrant, 3.7.2 zfplant conditions warrant the need for follow up information, Comalete the Follow Up Notification Form, Attachment Note: Conditions that warrant this evaluation are as a minimum when other EAL conditions exist indicating the current emergency classification or significant changes in plant conditions have occurre . rf the CECC is not staffed, Then notify the ODS and provide follow up information from the completed Attaclinient C form. Utilize the direct ring-down ODS phone when making this notification or as applicable dial direc ODS - 5-751-2495, 1700 Note: zf the ODS cannot be reached, contact the State of Alabama directly by requesting the Office of Radiation Control at:

=Shift 8 a.m. - 5 p.m. (Central Time) Holidavs-Weekends-Off-shifts Primary: 9-1-334-206-539I Montgomery State Trooper Post Backup: 9-1-800-582-1866 9-1-334-242-4378 3. Ifthe conditions warrant upgrading to a higher classification, Then initiate the appropriate EPH PAGE 5 OF 11 REVISION 0029 I

ALERT BROWNS FERRY EPIP-3 NUCLEAR PLANT

.

I 3.8 INSTRIJCTIONS (CONTINUED)

3. If the conditions warrants termination of the classifications, Then enter EPIP-16, Termination and Recovery Procedur . After the evaluation has been completed, .if stafed, Notify the following of the status:

(!ECC OSC NRCPNS) CONTROI.ROOMS TSC PI.mr PA ANNOUNCEMF,NT 3.7.7 Re-enter this procedural section as conditions warrant at step 3.7.1 or until directed to exit this procedure by steps 3.7.4 or 3. .8 CLOSIJREOF THE ALERT 3.8.1 Upon termination of the Notification of Alert, the Shift -. .-

Manager shall send the completed EPIP-3 and all IMTTA TIRlE attachments to Emergency Preparedness (El).

-

3.8.2 Upon receipt of completed EPIP-3 and all attachments, 1 m . s TiME Emergency Preparedness shall forward documents for the purpose of documentation storag .8 RECORD RETENTION 4.1 RECORDS OF CLASSIFIED EMERGENCIES The materiaIs generated in support of key actions during an actual emergency classified as NOL% or higher are considered Lifetime retention Non-QA records. Materials shall be forwarded to the El Manager who shall submit any records deemed necessary to demonstrate performance to the Corporate EP Manager for storag .2 DRILL AM) EXERCISE RECORDS The materials deemed necessary to demonstrate performance of key actions during drills are considered Non-()A records. These records shall be forwarded to the EP Manager who shall retain records deemed necessaly to demonstrate six-year plan performance for six years. The EP Manager shall retain other records in this category for three year PAGE 6 OF 1 1 KEVISION 0029 I

ALERT BROWNS FERRY EPIP-3 NUCLEAR Pl,A?JT ATTACHMENTS I Attachment A - Initial Notification Fonn Alert

-

Attachment B Unit 1, Unit Operator Notifications Attachment C Follow Up Information Form Alert PAGE 7 OF I 1 WVlISION0029 I

ALERT BROWNS FERRY EPIP-3 NUCLEAR PLANT ATTACHMENT A (Page 1 of 1)

INITIAL NOTIFICATION FORM This is a Drill ALERT This is an Actual Event - Repeat - This is an Actual Event I

2. This is , Browns Ferry has declared a ALERT affecting: eS Unit 1 Unit 2 0 Unit 3 JI Common AL Designator(8):

4. Brief Description of the Event:

5. Radiological Conditions: (Check one under both Airborne and Liquid column.)

Airborne Releases Offsite Liquid Releases Offsite Minor releases within federally approved 0Minor releases within federally approved limits limits Releases above federally approved limits Releases above federally approved 5 Release information not known limits ( Tech Specs) Release information not known ( Tech Specs)

6. Event Declared: Time: Date:

7. Provide Protective Action Recommendation: None 8. Please repeat the information you have received to ensure accuracy.

9. Time and Bate this information was provided I Action: When completed, telecopy this informatio I PAGE 8 OF 11 REVISION 0029 I

ALERT BROWNS FERRY EPIB-3 NUCLEAR PLANT

_. ~~~~ ~

ATTACHMENT B (Page 1 of2) I UNIT 1, UNIT OPERATOR NOTIFICATIONS Date: 1 1 NOTES: (1) The Emergency Paging System F P S ) consists of a dedicated touch screen CR Activation of any screen feature requires the user place their fingertip within the boundary ofthe select button and leave it there for at least I second. The CRT Screen will normally display a large rectangle that indicates that the paging system is available but currently inactiv (2) If the EPS fails to operate, contact the SNVSED immediately. Request that the ODS be c.ontacted to initiate the system from his location. If the system fails to operate from the ODS area, then utilize the Weekly Duty List and Call-Out List to manually staff the Emergency Responders, implementing this attachment at step . Activation ofthe Emergency Paging System (EPS). PRESS the EPS CRT Screen once to activatc the paging INITIALS TIME option l PRESS the appropriate option as instructed by the SED

• PAGERTEST LYITIALS TIME 0 DRILL

• EMERGENCY

• STAGMGAREA ABORT PRESS the START Button to initiate the option or ABORT to deny t.he option reques LTTIAES TIME MONITOR the Paging System Terminal Display INITIALS TIME F... A NO response is observed OR The position being paged has not responded within approximately 20 minutes THEN... Utilize the Weekly Duty List and attempt to contact the position representative with available information. (No Fitness for Duty Question Required) IF... The individual cannot be reached utilizing the Weekly Duty List THEN... Utilize the Call-Out List and attempt to contact an alternate position representative. (Fitness for Duty Question Required)

PAGE 9 OF 1I REVISION 0029 I

ALERT BROWNS FERRY EBIP-3 NUCLEAR PLANT L--

ATTACHMENT B (Page 2 of 2)

UNIT 1, UNIT OPERATOR NOTIFICATIONS Date: I I E). Manual Call-Out (N/A step ifEPS operates normally)

INITIALS TI^ Utilize the current Weelkv Dutv List and contact positions as liste . I f a position can not be reached from the current Weekly Duty list, then refer to the Call-out List as applicable to fill all vacant position CONTINUE until all positions have been fille INIT~ALS TIME 2. Notifv the Unit Supervisors on shif INITIALS TIME 3. Notifv Nuclear Security Shitt Supervisor and state AN ALERT HAS BEEN DECLARED and direct to activate EPIP-11, INITIALS TIME Security and Access Contro e Plant Extension 3150 or 2219 4. Notifv the Chemistry Lab Supervisor and state AN ALERT HAS BEEN DECLARED and direct to implement 2/3-TI-33 1, INITIALS TIME Post Accident Sampling Procedure and CI-900 series, Analysis Procedure e Plant Extension 2364 or 2368 5. Notifv the RADCON Shift Supervisor and state AN ALERT HAS - ~ --

BEEN DECLARED and direct to activate EPIP-14, INITIALS TIME Radiological Control Procedure e Plant Extension 7865 or 3 104 6. Notifv the On-Call NRC Resident and state AN ALERT HAS BEEN DECLARED. per BFN-EPIP-03 INITIALS TIME W Plant Extension 2572 [Secretary] or from weekly duty list PAGE 10 OF 11

ALERT BROWNS FERRY 13~13-3 NUCL.EAR PLANT ATTACHMEAT C (Page 1 of 1)

FOLLOW-UP INFORRIATION FORM ALERT THIS IS A REAL E\ENT THIS IS A DRILL Note: This form is for conducting Follow-up Information only.

rhis is _____.____- at Browns Ferr Name There has been a Alert declared at Browns Ferry affecting:

0 IJnit 1 Unit 2 119 Unit 3 Common The Reactor is Shutdown At Power Plant Conditions are Stable aDeteriorating

.Follow-Up Information (e.g., Key Events, Status Changes)

Current Radiological Conditions are: (Check one under both Airborne and Liquid c.oluma.)

Airborne Release Offsite Liquid Releases QfYsite 2 Minor releases within federally approved 6] Minor releases within federally approvec limits limits 0 Releases above federally approved limits Releases above federally approved limit:

Release Information not known 61 Release Information not known ( Tech Specs) ( Tech Specs)

Additional Rad information: (e.g., release duration)

______ _ I _ _ _

63 There is no Protective Action Recommendation at this time.

Please repeat the information YOU have received to ensure accuracy.

The time for this follow up is: Time: Date:

SIGNATURE:

LAST PAGE PAGE 11 OF 11 REVISION 0029 I

TL, dTAlUDARD SPP-3.5 P R Q G M M S AND REGULATORY REPORTING REQUIREMENTS Rev. 11 PROCESSES Page 3 of 53 REVISION LOG Page 2 of 2 Revision Effective Pages Description Number Date Affected of Revision 8 5/2/01 2, 37 Update Form "NRC Event Notification Worksheet" to reflect changes to NRC Form 361. (Minor/editorial changes.)

9 6/29/01 2-4, 6,8-10, Added guidance on key information to be communicated to the 38,39 NRC Resident Inspectors following a significant operating event.

10 2/6/02 3-9,11-27,32, Revised procedure to clearly identify organizational 34 responsibilities for making reportability determinations. immediate notifications, and follow-up written reports. Also added section lo Appendix A, pertaining to the optional verbal notifica!ion that is allowed under 10 CFR 50.73(a)(2)(iv)(A).

11 14/19/02 3,4, 6-8,14, 18-20, 23-51 Revised procedure to add reporting requireinents (App. C) for Independent Spent Fuel Storage Facilities prescribed in 10 CFH

.

72. Also added flexibility to Appendix A, Section 3.5 pertzining to written reporting requirements. Also added Apperidix'J to provide guidance pertaining to serious Accident Internal Notificatiors Added 30-day verbal NRC notification that is required by 10 CFR 20.2281 (a)@) to address PER 02-000344-30 W A N STANDARD SPP- PROGRAMS AND REGULATORY BEPORTING REQUIREMENTS Rev. li PROCESSES Page 4 of 53 TABLE OF CONTENTS Sectior Title Page Revision bog............................................... .......................................................................... 2 Table of Contents ........ .......................................................................... 4 .0 SCOPE .......................... ................................ 5 IN~YRUCTIOPJS........................................................................................................... 5 Periodic Reports ..... Event OF Condition Reporting ............ RECORDS .................................................................................... ......................... 10 B~FBNITIQNS .........................................................................................................

APPENDIXES Appendix A Reporting of Everit or Cmditions Affecting Nuclear Power Plants ......14 Appendix E3 Reporting of Events or Conditions Affecting Activities Invol..'ng ByProduzt, Source or Special Nuclear Material Licenses Appendix C Reporting of Events or Conditioris Affecting Independent Spent File1 Storage installation (ISFSI) ........................................................ 29 Appendix D Site Everit Notification bl a !rix ...............................................

Appendix E Other Regulatory Repor!irig ...............

Appendix F Evaluation and Reporting of Defects a Associated with Subs!ari!lal Safety Hazards Per 10CFR50,55(e) Reporting Requirements Appendix 6 Determination of Reportabiiity Under 10 Appendix H Reporting of Decommissioning Funding ................................

Appendix I Communication with the NRC FO Operational Event ...................... 4%

Appendix J Internal Notificatio investigations ....... ....................................

FORMS SPP-3.5-1 NRC Event Notification Worksheet .................................................... 52

N A N STANDARD SPP- PROGRAMS AND REGUMTORY REPORTING REQUIREMENTS Rev. 11 PROCESSES Page 5 of53 PURPOSE This Standard Program and Process (SPP) specifies the requirements for various reports to the Nuclear Regulatory Commission (NRC) and other regulatory agencies to ensure compliance with the reporting requirements. _SCOPE This SPP identifies the reporting requirements specified in the following: Title 10, Code of Federal Regulations Technical Specifications Final Safety Analysis Reports Correspondence to various regulatory agencie ?he appendixes provide guidelines for reporting of conditions andlor events INSTRUCTIONS Periodic Reports Each site licensing organizatizn shall emure that a rnalrix o: their site's periodic reporting requirements is maintaine Each report matrix should contain the report topic or titie?regulatory basis for the reocrl time requirements for submitting the report, and the groupjs) responsible for preparirig, coordinating review, obtaining approval and issuing the repor Report Preparation a The organizatioii or individual respcnsibk for preparing the report shall ensure that the report is initiated in a timely manner (for proper rev;t?:v and approval), addresses all necessary items, is technically accurate. is reviewed and coordinated, and meets the reqiirernents for submittal. I f submitted to NRC, the iepcrt must be processed in accordance with Business Practice (BP) 213, "Managing TVA's Interface with NRC." For reports to other reguletoty agencies. the responsible 0rganii'a:io;r shall ensure applicabic: reporting guide:ines are satisfie . Reports with commitments shall meet the requirements of SPP-3.3,

'NRC Commitment Management." Review and Approval of Reports The organization or individual responsit~lefor review shiill ensure that the !-t.pLil'

is consistent with TVA policy an:! shall resolve comments. The orgaiiizatior ~>:

individual responsible for approval of the report shall perform a final revie NANSTANDARD SPP-3.5 PROGRAMS AND REGULATORY REPORTING REBUIKEMERITS Rev. 12 PROCESSES Page 6 of 53 Distribution of Reporls The organization responsible for preparing the report shall ensure that adequate copies are made in a timely manner to supporl the transmittal of the report to the regulatory agency. Internal copies of the report shall be distributed in accordance with organizational procedures or instructions on preparation of the repor Complete and accurate information must be provided to NRC at all time Information can be in violation of this requirement even if it is not in writing, supplied under oath, or supplied wit.hoLit knowledge of its falsity. lnforrnation can be considered incomplete or inaccurate due to any one of the following reasons: An affirmative statement which is false An omissio . Inadequate revie . Failure to review Careless disregard or deliberateness Negligence not amoiirting to careless disregard Inadvertent clerical or similar error involving information which, ha6 it

'

been available to NRS and accurate at the time the inforrmtion shculd have been submitted, waul:! probably have resalted in regulatory or NRC seekiriy abditional informatio Failure to correc: material informa::on tha: has significant implication for public health and safety which was correct when submitted but Oeccrnes incorrect due to subsequen! changes or event Identification of incorrect statements or misrepresentations having significartt implication for public health and safety or c o m r n w defense and security made in previously submitted information, including reports, must be rcported to NRC's Regional Office within two working days of identifying the information. Phis requirement is not applicable to informatiort which is already required to be provided to NRC by other repofiing or updating requirements. Licensir!g is responsible for making the deteirnination of reportability and notifying NRC i n accordance with 10 CFR 5 .2 Event O F Condi!ion Repo3. The specific event or condi!lon repoeing requirements applicable tc different licenses or permits TVA possesses are identified in the followirly Appendices: Appendix A. "Repotting cf Events 3: Conditions AffectinQ Licensed Nuclear Power Plants" contains the criteria for reporting of events c:

conditions affecting licensed nuclear power plants. Operations is responsible for making the reporlabi!ity de!erminations for 50.72 and 50.73 repoits. Operatioris is it.sporisihle for making the immediate

W A N STANDARD SPP- PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 11 PROCESSES Page 7 of 53

. . notification to NRC in accordance with 10 CFR 50.72. Licensing is responsible for developing (with input from affected o!ganizations) and submitting the written reports (or optional telephone reports) required by I O CFR 50.7 N6)TE Reporting requirements for personnel exposure required by 10 CFR Part 20 are contained in RCRP-4, "Personnel lnprocessing and Bosimetty Administrative Processes."

Appendix A also contains criteria for reporting "radiation injuries" in accordance with 10 CFR 140.6. Site WDCON is responsible for reporting "radiation injuries" to Licensing. Licensing is responsible for developing and submitting the written report to NRC (with input frorn Site BADCON). Appendix B, "Reporting of Everits or Conditions Affecting Activities involving Byproducts, Source, or Speciai Nuclear Material Licenses" contains the cri?eriafor reporting of events or conditions affecting activities involvirig byproduct. source or special nuclear material licenses. Site Licensing and Site RadCon are responsible for making the reportability determinations for 10 CFR Part 20, 30, 40, or 70 events associated with their site. Corporate Licensing and Corporate R a d C k n are responsible for making the reportability determinations for 10 C F 3 Part 20, 30. 40:or 70 events associated with ail other TVA licensed activities. Licensing is responsible fcr making the immediate notification an< developiog (with ir.put from affected organizations) and submitting written reports to NRC in a c c o r d a r k with 10 CFR Part 20. 30, 48, or 7 0 requirement . Appendix C, "Repohng of Events or Conditions Affecting i7dependent Spent Fuel Storage Instaliatioxs (ISFSI)" contains the criteria for reporting even!s or ccriditions affecting ISFSi. TVA, as the general licensee of the ISFSI, is required by 18 CFR 72.216 to make inltial and written reports in accorc?ancewith 10 CFR 72.74 and 10 CFR 12.7 Operatioris is responsible fcr making the reportability deterrnhatiorls for 10 CFR 72.74 and 70 CFR 72.75 reports. Operations is tespor?sible for-making the immediate r:ctiiication to NRC in ascordance with 10 C'R 72.74. Operations is responsibie for making the imrnediaie, 4-'lour. and 24-hour notifications in accordance with 10 CFR 72.75. LiceXing is responsible for developing (with input from affected c:ganiza!ions) atid submitting the written reports required by 10 C F R 72.7 . Appendix D, "Site Evect Notificatioo Matrix" contains :>e in!emI management notjfication req2iremerits for plant evevts. If designatcj TVA Manager is unavaiiahie for- no!ification due to temporsry assignment. ( e . & INFO loanee) notificaticn should be made :ii GeSi!il or next higher manage;. Cperations and the Fiani Manager (c: DC;Iy Plant Manager) are rcspcmlble for makirrg these internal rrianarJen notifications.

. .

WANSTANBARD SPP-3.5 PROGRAMS AND REGULATORY REPOFITWG REQUIREMENTS Rev. 71 PROCESSES Page El of 53 Appendix E, "Other Regulatory Reporting" contains the criteria for reporting of events or conditions to Federal and State regulator)l, agencies other than the NRC. Operations is responsible for making the repQltability determinations and notifications for these non-NRC Federal and State regulatory agency reporting requirement Additional reporting guidance for defects is contained in SPP-3.1, "Corrective Action Program." Appendix F. "Evaluatien and Reporting of Defects and Failures to Comply Associated with Substantial Safety Hazards Per 10 CFR 50.55(e) Reporting Requirernents" contains the criteria tor reporting of deficiencies for nuclear plants with a construction permit in accordance with 10 CFR 55.55(e). Licensing is responsible for making the final reportability determination and written report to NRC in accordance with 10 CFR 50.55(e).

Additional reporting guidance for deficiencies is contained iri SPP- . Appendix G, "Determination of Reportability Under 10 CFR Part 21" contains the criteria for reporting defects in basic components in accordance with 10 CFR Part 21. Licensing is responsible for making the final repoitability determination and written report to NfiC in accordance with 10 CFR Part 2 . Appendix H, '"Reporting of Decomrnissionirlg Funding" contains the criteria for notifying NRC when permaneo?ly shtitting down thcoperatiori of a reactor, as reqJired by 10 CFR 50.54(bb). Licensing is respopsitla for making the written r,ctification to NRC in accordance with 10 CPR 50.54(bb). Appendix l, "Communication with the NRC Following a Significant Operational Event" contciins guidance on communica!ions t9at needs to be established with the NRC within 24 to 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following a significant operational event that could result in an incident investigatio:i by the NRC. Site Licensing coordinates this commurricatioo with NR . Appendix J. "lnternal Notification of Events Requiring S e r i w s AcZid-?ct Investigations" provides internal management notification reqciirements for serious accidents, as prescribed in TVA-SPP-18.ClO. "Conduc:

Serious Accident Investigation."

1 fieporting requirements :or fitness for duty events recuired by 10 C F Q Part 26 are contained in SPP-1.2 "Fitness For Duty." Respo:lsibillties :ill repoitahility determitations ard irnrnediate notification reqdi:ernen!s d i e assigned to Site Nuc!ear Security and Corporate Nuclear Securit Licensing is respcnsible for making the written reports required b y 1 C CFR Part 2 . Reporting requiremerits fsr events o i conditions affecting the ohysicdl protection of the licensed nuclear plant specified in , J CFR 73,71 a??

contained in SPP-1.3 'Pla!jt Access and Security." Respnsihiiities is1

W A N STANDARD SPP- PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. 31 PROCESSES Page 9 of 53 reportability determiriations and immediate notification requirements are assigned to Site Nuclear Security and Corporate Nuclear Security. In accordance with SPP-1.3, if NRC notification is required (e.g.. o w or twenty-four hour phone call), the Site Security Manager will request the Plant Shifl Manager to call the NRC 0perat.ions Center. Licensing is responsible for making the written reports required by 10 CFR Part 73.7 Report Preparation The organization or individual responsible for preparing the report shall ensure that the report is initiated in a timely manner (for proper review and approval), addresses all necessary items, is technica!ly accurate, is reviewed and coordinated, and meets the requirements for submittal. If subrnitted to NRG, the report must be processed in accordance with Business Practice (BP) 213, "Managing TVA's Interface with NRC." For reports to cther regulatory agencies, the responsible organization shall ensure applicable reporting guidelines are satisfie m Reports with cornrnitrnents shall meet the requirements of SPP-3.3,

'NRC Conlrriitrrient Management. Review and Approval of Reports The oryar:-ration or individual responsible for review shall enskire that the report is consistent with TVA policy and shall resolve comments. ?he organization cr individual responsible for approval of the report shall perform a final review.

.. . Distribution of Reports The organization responsible for preparicg the report shall ensure that adequate copies are made in a tirnely manner to support the transmittal of the repott to the regulatory agency. lnterrial ccpies of the report shall be distributed in accordance with organizational procedures or irlstrilctions on preparation of the repor Complete and accurate informatiori must be provided to &RC at RII tirne Information can be in violation c f this requirernerrt ever1 if it is not in writin supplied under oath, or supplied without knowledge of its falsity. Information can be considered incomplete or inaccurate dce to any one of the followiXl reasons: An affirmative statement which is fals . An omission Inadequate review Failure to review Careless disregard or i!elibe:atenes . Negligence not arnoi:ntirig to CareleSS disregard

W A N STANDARD SPP- PROGRAMS AND REGULATORY REPORTING REQUIREMENTS Rev. <I PROCESSES Page 10 of 53 Inadvertent clerical or similar error involving information which, had it been available to NRC and accurate at the time the information shocld have been submitted, would probably have resulted in regulatory action or NRC seeking additional informatio Failure to correct material irifomation that has significant implication tor public health and safety which was correct when submitted but becoiries incorrect due to subsequent changes or events. RECORD2 Records of the reports and their transmittals shall be maintained in RIMS as non-QA records unless separate procedures require the reports to be maintained as QA records. DEFINITIONS Actuation - The minimum number cf ttipped channels required to complete the logic of a function. (Example: 214 logic - at !east tw3 (2) channe!s n!ust trip to be considered an actuation Administrative Control Program (ACP) ),I? approved, proceduralized method

~ 0: docume;l!ivic adverse conditions and iolplerrlei~tingcoriec.tive actio Basic Component when applied to n x l e a r power reactcrs: means a plant structure, system component, or part thereof necessaty to ersure: The integrity of the reactor coolant pressure boundav, The capability lo shut dowil the rea,i?orand maintain it in a safe shutdown c o n d i t k cr The capability to prevent or mitigate the consequences of accidents whicn could resu't. it-,

potential offsite exposures comparable to those referred to in 10 CFR 100.11 In all cases, "Basic Component" includes safety-related design, analysis, inspection, testicg, fabrication. replacement parts, or consulticy services that are associated with the coniponerit hardware whether these sewices are performed by the component suppEier or other (10 CFR 21.3 and 10 CFR 50.2).

Completion Of Any Nuclear Plant Shutdown -This is when reactor is taken subcritica Construction - T h e analysis, design, manufacture, fabrication, quality assurance, placemert, erection, installation, modification, inspection, or testing of a facility or activity and ccnscltinil services related to the facility or activity that are important to safet Control Of The Items TVA is subject to Part 21 requirernents regarding procured materia'.

components, or parts after TVA h a s taken control of the items. This cccu?s after conductirig the required receipt inspection. Evaluation of defects famd during receipt inspectiorl is the responsibility of the supplier if the iterns are relurned to the supplier. If TVA accepts ownrrsh!p of the itern, any defect must be evaiuated and. if applicahle, reported under Palt 21