Text

Draft - Section a Operating
	ML031710473
Person / Time
Site:	Crane
Issue date:	05/14/2003
From:	Gumbert R; AmerGen Energy Co
To:	Conte R; NRC/RGN-I/DRS/OSB
	Conte R
References
	50-289/03-301 50-289/03-301
	Download: ML031710473 (117)
	v • d • e

TMI SRO License Exam 0511 2/03 TMI-I OPERATOR TRAINING JOB PERFORMANCE MEASURE A.l-1 PERFORM ESTIMATED CRITICAL BORON CONCENTRATION CALCULATION

\\

Page 1 of 7

A.1-I 11.2.05.122 Revision 3 0511 2/2003 TASK TITLE:

PERFORM ESTIMATED CRITICAL BORON CONCENTRATION CALCULATIONS L

TASK NUMBER:

00 I C030 1 0 1 TIF: 3.05 KIA

REFERENCE:

System:

NA KIA:

2.1.25 Rating(RO/SRO):

2.8/3.1 POSITION:

SRO@RO@ N L O n EVALUATION METHOD:

PERFORM [x1 SIMULATE 0 EVALUATION LOCATION:

SIMULATOR 0 IN-PLANT 0 CONTROL ROOM 0 OTHER [XI TASK STANDARDS:

Examinee calculates RCS boron concentration required to achieve criticality at the desired critical rod position, within the tolerances described within this JPM.

APPROXIMATE COMPLETION TIME: 30 minutes u

TIME-CRITICAL TASK COMPLETION TIME:

NA minutes REQUIRED TOOLS OR MATERIALS:

Calculator.

Straight Edge.

OP 1103-15B, Estimated Critical Conditions, Revision 34.

REFERENCES:

OP 1 103-1 5B, Estimated Critical Conditions, Revision 34.

HU-AA-104-101, Procedure Use and Adherence, Revision 0.

ALTERNATE PATH JPM? NO SIMULATOR SETUP: NA INITIALIZATION: NA EVENT TRIGGERS: N/A MALFUNCTIONS: NIA REMOTE FUNCTIONS: N/A OVERRIDES: N/A MONITOR: N/A Page 2 of 7

A.1-1 11.2.05.122 Revision 3 0511 212003 READ TO STUDENT L

When I tell you to begin, you are to CALCULATE THE RCS BORON CONCENTRATION REQUIRED TO ACHIEVE CRITICALITY AT THE DESIRED CRITICAL ROD POSITION. Before you start, 1 will describe the general plant conditions, state the initiating cues, and answer any questions. Perform procedure steps as if you were actually performing the task.

INITIAL CONDITIONS:

The reactor was manually tripped from full power, equilibrium conditions, 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ago.

Prior to the reactor trip, power was constant at 100% for the past 3 months.

Current conditions:

Reactor is at Hot Shutdown.

Preparations are in progress to go critical Reactor trip is reset.

CRD Groups 1-4 are fully withdrawn.

CRD Groups 5, 6, and 7 are fully inserted.

0 Group 8 APSR positions: 30% withdrawn.

RCS T-ave is 534°F.

RCS Pressure is 2155 psig.

Core Burnup is 600 EFPD.

Reactor startup (achieving criticality) is scheduled to occur 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from this time.

RCS Boron Concentration is 190 ppm.

Mixed Boron Depletion Correction Factor = 0.99.

PPC and Nuclear Engineering are unavailable to provide value for Xenon reactivity.

INITIATING CUE:

The Shift Manager directs you to calculate the RCS boron concentration required to achieve criticality at the desired control rod positions described in Enclosure 1 of OP 1103-1 5B, Estimated Critical Conditions (provided).

\\d ARE THERE ANY QUESTIONS?

TIME CRITICAL: NO Page 3 of 7

JPM INSTRUCTION SHEET DIRECTIONS TO STUDENT:

When I tell you to begin, you are to CALCULATE THE RCS BORON CONCENTRATION REQUIRED TO ACHIEVE CRITICALITY AT THE DESIRED CRITICAL ROD POSITION described in enclosure 1 of OP 11 03-15B, Estimated Critical Conditions (provided). Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions. Perform procedure steps as if you were actually performing the task.

INITIAL CONDITIONS:

The reactor was manually tripped from full power, equilibrium conditions, 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ago.

Prior to the reactor trip, power was constant at 100% for the past 3 months.

Current conditions:

Reactor is at Hot Shutdown.

Preparations are in progress to go critical Reactor trip is reset.

0 CRD Groups 1-4 are fully withdrawn.

0 CRD Groups 5, 6, and 7 are fully inserted.

Group 8 APSR positions: 30% withdrawn.

RCS T-ave is 534°F.

RCS Pressure is 21 55 psig.

Core Burnup is 600 EFPD.

Reactor startup (achieving criticality) is scheduled to occur 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from this time.

RCS Boron Concentration is 190 ppm.

Mixed Boron Depletion Correction Factor = 0.99.

PPC and Nuclear Engineering are unavailable to provide value for Xenon reactivity.

INITIATING CUE:

The Shift Manager directs you to calculate the RCS boron concentration required to achieve criticality at the desired control rod positions described in Enclosure 1 of OP 1103-15B, Estimated Critical Conditions (provided).

d TIME CRITICAL: NO Page 4 of 7

Denotes Critical Elements

Denc L -

1 2

v L

3 10 11 1s Sequential Step STEP Examinee obtains a copy of OP 1103-15B, Estimated Critical Conditions, to calculate the ECB in accordance with guidance provided in Section 3.1, using Enclosure 1.

Examinee reviews procedure Section 2.0, Limits And Precautions.

Examinee begins implementation of Section 3.1, Estimated Critical Boron Concentration.

Examinee verifies data included on is correct, in accordance with Initial Conditions.

Using Figure 1, the examinee determines fuel excess reactivity.

Using Figure 6, the examinee determines reactivity worth of Control Rod Groups 5-7 at the desired critical position.

Using Figure 2, the examinee determines reactivity worth of Control Rod Group 8 at the desired critical position.

Using Figure 4, the examinee determines Xenon reactivity at the time of startup.

Using Figure 5, the examinee determines reactivity associated with samarium and plutonium buildup after shutdown.

Examinee calculates boron reactivity worth required for criticality at the desired critical rod position.

~~

Using Figure 3, the examinee determines Hot Zero Power inverse boron reactivity worth.

A.1-I 11.2.05.122 Revision 3 05/12/2003 STANDARD OP I 1 03-1 58, Estimated Critical Conditions, is a Level 2 procedure.

Adherence, Section 2.1.2 defines Level 2 as Reference Use: referring to a procedure periodically during the performance of an activity to confirm that all procedure segments of an activity have been performed, performing each step in the sequence specified, and where required, signing appropriate blocks to certify that all segments have been completed.

The procedure should be at the work location.

Adherence, Section 4.1.6 requires the Procedure User to observe all Precautions, Limitations and applicable Prerequisites.

HU-AA-101-101, Procedure Use and HU-AA-101-101, Procedure Use and Expected value = +6.9% AWK.

Interpolation between curves for 400 and 693 EFPD is required in order to determine value for 600 EFPD.

Expected value for 600 EFPD = -1.5% AWK.

Interpolation between curves for 400 and 672 EFPD is required in order to determine value for 600 EFPD.

Expected value = -0.1 36% AWK.

Correct time (hours after shutdown) is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Expected value = -3.6% AWK.

Interpolation between curves for 400 and 693 EFPD is required in order to determine value for 600 EFPD.

Expected value = -0.033% AWK.

Expected value = -1.631 % AWK.

Expected value = 127.5 ppm B/% AWK.

Page 5 of 7

12

I 3 STEP

~

~~

~

__ ~

Using the inverse boron reactivity worth, and the required boron reactivity, the examinee calculates the corrected Critical Boron Concentration Using the mixed boron depletion correction Factor (provided in the Initial Conditions),

the examinee calculates the Estimated measured Critical Boron Concentration.

A. 1-1 11.2.05.122 Revision 3 051' STANDARD Expected value = 208 ppm.

Lower Acceptance Limit =

107 ppm.

Expected Value =

210 ppm.

Upper Acceptance Limit =

313 ppm.

Basis for acceptance range:

Reference OP 1 1 03-1 5B, Estimated Critical Conditions, Enclosure 2 Section 4.3, for Critical Rod Position Tolerance Band:

Transient Xenon Startup (if Xenon is MORE negative than -0.5% AKIK):

0 Critical Rod Position Tolerance Band is k 0.8% AKIK from desired critical rod position.

(+0.8)(127.5/0.99) = 210 +I03 ppm.

Lower Acceptance Limit = 107 ppm.

Upper Acceptance Limit = 313 ppm.

0 0

0 Xenon Free Startup (if Xenon is LESS negative than -0.5% AUK):

0 Critical Rod Position Tolerance Band is k 0.5% AUK from desired critical rod position.

12003 SIU -

END TASK Page 6 of 7

REVISION 3

DATE 05/12/2003 A. 1-1 11.2.05.122 Revision 3 031 2/2003 JPM CHANGE HISTORY PAGE REFERENCE TITLE OP 11 03-1 58 Rev. 34, dated 10/28/02.

DESCRIPTION (Include AI # if Appropriate)

Modified Bank JPM 11.2.05.122.

Page 7 of 7

I 1 03-1 5 8 Revision 34 Page 1 of 1 W

4.l 4.2 4.3 ENCLOSURE 1 Estimated Critical Boron Concentration (3.1)

IndividuA dala eiitries may be coiiipleted in any sequence Sign-oil of the Enclosure signifies the coinpletion of the Enclosure calculation CALCULATION IS FOR AN ECBP AT 532 k 2°F ON TAVE ?j3L/

DATEiTIbIE -

CYCLE BURNUP PRESENT klEASURED EORON CONCENTRATION

/ 9 ppm6 DESIRED CRITICAL ROD POSl [ION I

h.

00

-.E F p D 0

CRG 1-4 100

"'a WD CKG 5

/o '% WD CRti 6 7

WD CKG 7 c'

Yo W D CRG 8 3 (3 'YO W D FUEL EXCESS REACTIVITY (FIG 1)

DESIRED CRITICAL CRG 5-7 REACTIVITY WORTH (FIG 6 )

DESIRED CRITICAL CRG 8 REACTIVITY WORTH (FIG 2)

XENON REACTIVITY WORTH (PPC, NUCLEAR ENGR FIG 4)

SAILIARIlIRd AND PLUTONIUM BUILDUP (FIG 5)

TIME SINCE SHUTDOVVN HKS WAC1 IVITY DUE TO BUILDUP "10 \\kik EORON REACTIVITY WORTH REQUIRED FOR CRITICALITY 1

b x (-1) -

(FUEL)

(CRG 5 - 7)

(CRG 8) (Xenon)

[SM) 3 a 3 b 3 c 3 d 3 e

+

?'IO

\\kik INVERSE BORON WORTH (FIG 3) ppi>iB/% Ik/k CRIl ICAL. 60KON CONCENTRATION

-1 3 1 CORRECTED CRITICAL BORON CONCENTRATION i [Inverse Boron)

(Boron Reacl)

I ppmE I =

sy x ( 1) x 4,1 4.2 L

4 3 2 FINAL MIXED BORON DEPLETION CORRECTION FACTOR (NAS Display 10, Control Room Log, Nucleai Engineenng) 4 3 3 ESTIMATED MEASURED CRITICAL BORON CONCENTRATION (4 3 1) / ( I 3 2 ) =

-[>kJiljf<

Sane <opy ot this Enclosure to Nuclear Engineering Szna original to Oparations tor filing 12

I Number I

TMI - Unit 1 Operating Procedure AmerGen Title L

11 03-1 5 8 Revision No Estimated Critical Conditions Applicabili ty/Scope USAGE LEVEL TMI Division 2

List of Effective Pages 34 Effective Date 10128/02 Paqe This document is within QA plan scope X

50.59 Applicabie X

1 2

3 4

5 6

7 8

9 IO 11 12 13 14 15 Yes No Yes No u 16 17 18 19 20 Revision Paqe Revision Revision Paqe Revision 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 1

TMI - Unit 1 Operating Procedure Title Number 1 103-1 5B Revision No.

Section

?.O REFERENCES Estimated Critical Conditions 3

34 2.0 LIMITS AND PRECAUTIONS 3

3.0 OPERATING PROCEDURE 4

3.1 Estimated Critical Boron Concentration 4

3.2 Estimated Critical Rod Position 7

3.3 Actual Subcriticality Calculation 10 4.0 ENCLOSURES

.., Estimated Critical Boron Concentration-, Estimated Critical Rod Position-, Actual Subcriticality Calculation Figure q, Fuel Excess Reactivity Figure 2, Group 8 Integral Rod Worth Figure 3, Inverse Boron Worth Figure 4.

Transient Xenon Reactivity Worth Figure 5, Samarium and Plutonium Buildup following Reactor Shutdown Figure 6, Integral Control Rod Worth 12 13 14 15 16 17 18 19 20 2

TMI - Unit 1 Operating Procedure Title Estimated Critical Conditions 1.O REFERENCES 1.I TMI-1 FSAR Section 3.2.2 Number 1 1 03-1 5 6 Revision No.

34 1.2 FRA-ANP Doc. No.61-501 4400-01, Physics Manual Three Mile Island Unit 1, Cycle 14 I

Reactor=

Engineering may maintain6 a computer program which can calculate shutdown margin and reactivity balance for INFORMATION ii ONLY and trending purposes. All official record calculations shall use the 1 =sed figures and data sheets.

1.3 TMI-I Technical Specifications Sections 3.5.2 and 3.1.3 1.4 Plant Process Computer (1 105-IOA)

I.5 Approach to Criticality (1 103-8) 1.6 Soluble Poison Concentration Control (1 103-4) 1.7 Plant Process Computer Xenon Program 1.8 Transient Monitor System (1 105-1 5) 1.9 BWFC Operating Guidelines, 64-1234740-00, Reactivity Balance Guidelines 1.10 TMI Reactivity Management Program (1085) 2.0 LIMITS AND PRECAUTIONS NOTE 1 When a conditional steo does not aodv in the oerformance of this I

I procedure, the performer shall continue with the next numbered step.

I 2.1 All figures used to determine core reactivities shall be verified to be for the current fuel cycle as identified by title.

Rod withdrawal sequence shall be in accordance with 1103-8, Approach to Criticality.

2.2 3

TMI - Unit 1 Operating Procedure

- Title Estimated Critical Conditions 3.0 OPERATING PROCEDURE Number 1 103-1 5B Revision No.

34 NOTE When using the procedure Enclosures to perform the calculations, the entry of the data signifies the completion of the applicable procedure step.

No additional steD sian-off is reauired.

3.1 Estimated Critical Boron Concentration NOTE The purpose of this section is to estimate the critical boron concentration for the desired rod position prior to startup. This portion of the procedure need only be performed to obtain a critical boron concentration. The RCS is then deborated (prior to withdrawing CRG 5-7) to this concentration, and CRG 5-7 are withdrawn in accordance with 1103-8 Approach to Criticalitv.

3.1.I Prerequisites It is desired to perform an Estimated Critical Boron Concentration 3.1.2 Procedure

1.

Obtain Enclosure 1 and use the following to complete.

NOTE The reference conditions for an ECB are 532"F, 0 %FP, 21 55 psig, no xenon, and CRG 1-8 at 100% WD.

2.

Obtain the following data:

a.

Obtain the average reactor coolant temperature, TAVE, from the PPC or from the digital display window on the Control Room Center Console.

b.

Obtain the cycle burn-up from NAS Display 1 or the hourly log.

C.

Obtain the latest measured boron concentration from the Reactor Coolant Chemistry Analysis, and check the Control Room log to verify that no major boron concentration changes have been made since the analysis.

4

TMI - Unit 1 Operating Procedure

~--

Title Estimated Critical Conditions If major boron concentration changes have been made since the latest sample, request a new RCS boron concentration measurement.

Number 11 03-1 5B Revision No.

34 Until the new boron concentration is available, use 1103-4, Soluble Poison Concentration Control to estimate the current boron concentration to calculate a preliminary ECB.

d.

Record the desired critical rod positions.

d NOTE Normally for a xenon free core, the desired critical rod position should be between 75% WD on CRG-6 and 25% WD on CRG-7. This should allow a power increase to 90% F.P. using CRG-6 and CRG-7. With CRG-7 at approximately 90% WD, it may be necessary to feed and bleed for control of equilibrium xenon buildup.

3.

Determine the following:

a.

Determine the fuel excess reactivity per Figure 1.

b.

Determine the CRG 5-7 reactivity worth of Step 2.d. per Figure 6.

C.

Determine the CRG 8 reactivity worth of Step 2.d. per Figure 2.

d.

Obtain the xenon worth from NAS Display 22.

If this program is not available. obtain a value from the program XENC# (where # is the current cycle number). Figure 4 may be used if Plant Process Computer Program and Nuclear Engineering are unavailable provided that power was constant (within f 2% FP) for at least the last 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months prior to shutdown.

e.

Determine the reactivity associated with samarium and plutonium buildup after shutdown by using Figure 5.

4.

Calculate the following:

I.

Determine the boron reactivity worth required for criticality.

5

Number I

TMI - Unit 1 Operating Procedure

~--

Title I

103-1 5B Revision No.

Estimated Critical Conditions

2.

Determine the HZP Inverse Boron Worth associated with the cycle burn-up per Figure 3.

34

3.

Calculate the estimated measured critical boron concentration as follows:

3.1 Determine the corrected Critical Boron Concentration by multiplying the required boron worth (Step 4.1.) by the HZP Inverse Boron Worth (Step 4.2).

NOTE Step 4.3.1 is the Critical Boron Concentration based on undepleted boron.

3.2 Determine the Final Mixed Boron depletion correction factor based on current correction factor and accounting for predicted boron additions.

NOTE Only boron additions will affect the depletion correction factor.

4.

5.

3.3 Determine the Estimated Measured Critical Boron Concentration by dividing the corrected CBC (Step 4.3.1) by the depletion correction factor (Step 4.3.2).

an anomaly exists after checking the calculations, notify Shift Manager and evaluate the discrepancy.

IF -

THEN Approve and save as follows:

1.

Sign the appropriate location on Enclosure 1

2.

Have an independent licensed SRO review and approve the calculation.

3.

Send a copy of the completed Enclosure 1 to Nuclear Engineering.

4.

Forward original to the appropriate operations person for filing.

6

Number TMI - Unit 1 Operating Procedure

-u-Title Estimated Critical Conditions 11 03-1 5 6 34 Revision No.

3.2 Estimated Critical Rod Position 3.2.1 Prerequisites It is desired to perform an Estimated Critical Rod Position.

3.2.2 Procedure The purpose of this section is to estimate the critical rod position prior to startup. It may be necessary to perform this section multiple times if plant conditions change before actual startup to account for changes in xenon NOTE The reference conditions for an ECP are 532"F, 0 %FP, 2155 psig, no xenon. and CRG 1-8 at 100% WD (Ref. 1.2).

1.

Obtain Enclosure 2 and use the following complete:

2.

Obtain the following data:

a.

Obtain the average reactor coolant temperature, TAVE, from the PPC or from the digital display window on the Control Room Center Console.

b.

Obtain the cycle burn-up from NAS Display 1 or the hourly log.

C.

Obtain the latest measured boron concentration from the Reactor Coolant Chemistry Analysis, and check the Control Room log to verify that no major boron concentration changes have been made since the analysis.

If major boron concentration changes have been made since the latest sample, do not proceed.

Request a new RCS boron concentration measurement.

0 Before continuing, ensure the RCS is in boron concentration equilibrium. If necessary, request additional samples until two consecutive measurements are within 10 ppm of each other and no additional changes have been made to RCS boron.

Adjust the measured boron concentration to account for B-10 depletion using the boron correction factor (NAS Display 10, Control Room Log, or Nuclear Engineering).

7

TMI - Unit 1 Operating Procedure L-' Title Number I

103-1 5B Revision No.

d.

Record the expected CRG 8 position at criticality.

Estimated Critical Conditions

3.

Determine the following:

34

a.

Determine the fuel excess reactivity per Figure 1.

b.

Determine the CRG 8 reactivity worth of Step 2.d. per Figure 2.

C.

Determine the Hot Zero Power (HZP) Inverse Boron Worth associated with the cycle burn-up per Figure 3.

c. 1 Determine the reactivity associated with the boron concentration by dividing the current concentration (Step 2.c.3) by the HZP Inverse Boron Worth (Step 3.c).

d.

Obtain the xenon worth from NAS Display 22.

If this program is not available, obtain a value from the program XENC# (where "#" is the cycle number).

Figure 4 may be used if Plant Process Computer Program and Nuclear Engineering are unavailable provided that power was constant (within ? 2% FP) for at least the last 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months prior to shutdown.

e.

Determine the reactivity associated with samarium and plutonium buildup after shutdown by using Figure 5.

4.

Calculate the following:

1.

Determine the inserted CRG 5-7 worth required to be critical NOTE Inserted rod worth is used because the reference condition is for an ARO configuration.

2.

Determine the estimated critical rod position per Step 4.1. and Figure 6.

3.

Determine the critical rod position tolerance band for either a xenon transient or steady state xenon situation.

NOTE For the purposes of this procedure, xenon is defined as being in steady state when the xenon worth is between zero and -0.5% Ak/k; transient xenon is when xenon worth is more negative than

-0.5% Aklk.

8

I Number TMI - Unit 1 Operating Procedure Title Estimated Critical Conditions 11 03-1 5B Revision No.

34

4.

IF an anomaly exists after checking the calculations, THEN notify Shift Manager and evaluate the discrepancy.

5.

Approve and save as follows:

1.

Sign the appropriate location on Enclosure 2.

2.

Have an independent licensed SRO review and approve the calculation.

3.

Send a copy of the completed Enclosure 2 to Nuclear Engineering.

4.

Forward original to the appropriate operations person for filing.

9

Number TMI - Unit 1 Operating Procedure

'-- Title 1 103-1 5 8 Revision No.

Estimated Critical Conditions I

34 3.3 Actual Subcriticality Calculation (ASC)

NOTE I

I The purpose of this section is to determine the actual subcriticality at current plant conditions (Ref. 1103-8).

3.3.1 Prerequisites e

It is desired to perform an Actual Subcriticality Calculation (ASC).

3.3.2 Procedure

1.

Obtain Enclosure 3 and use the following to complete:

NOTE Reference conditions for this calculation are 532" F and 0% FP conditions.

2.

Obtain the following data:

3.

a.

Obtain the average reactor coolant temperature, TAVE. from the PPC or from the digital display window on the Control Room Center Console.

b.

Record the present control rod positions.

C.

Determine the Integral Control Rod Worth for the present control rod position using Figure 6.

Calculate the following:

1.

Perform an Estimated Critical Rod Position per Section 3.2.

2.

Record the "Inserted CRG 5-7 Worth Required for Criticality" (Step 4.1 of Enclosure 2).

3.

Calculate the Actual Subcriticality.

4.

IF an anomaly exists after checking the calculations, THEN notify Shift Manager and evaluate the discrepancy.

10

Number Operating Procedure

.- Title I

TMI - Unit 1 I

11 03-1 5B Revision No.

Estimated Critical Conditions 34

4.

Approve and save as follows:

1.

Sign t h e appropriate location on Enclosure 3.

2.

Send a copy of completed Enclosure 3 to Nuclear Engineering

3.

Forward to the Operations Department for filing 11

2.a 2.b 2.C 2.d 3.a 3.b 3.c 3.d 3.e 4.1 4.2 4.3 ENCLOSURE 1 Estimated Critical Boron Concentration (3.1)

NOTE Individual data entries may be completed in any sequence. Sign-off of the Enclosure sicrnifies the comdetion of the Enclosure calculation.

I 1103-15B Revision 34 Page 1 of 1 I

CALCULATION IS FOR AN ECBP AT 532

2°F ON TAVE DATElTl ME CYCLE BURNUP EFPD PRESENT MEASURED BORON CONCENTRATION PPmB DESIRED CRITICAL ROD POSITION CRG 1-4 100

%WD CRG 5

% WD CRG 6

% WD CRG 7 Yo WD CRG 8

% WD FUEL EXCESS REACTIVITY (FIG 1)

DESIRED CRITICAL CRG 5-7 REACTIVITY WORTH (FIG 6)

DESIRED CRITICAL CRG 8 REACTIVITY WORTH (FIG 2)

XENON REACTIVITY WORTH (PPC, NUCLEAR ENGR., FIG 4)

Yo Aklk Yo Aklk

'/o Aklk Yo Aklk SAMARIUM AND PLUTONIUM BUILDUP (FIG 5 )

0 TIME SINCE SHUTDOWN HRS e

REACTIVITY DUE TO BUILDUP

% ilklk BORON REACTIVITY WORTH REQUIRED FOR CRITICALITY x (-1) =

1 (FUEL)

(CRG 5 - 7 )

(CRG 8)

(Xenon)

(SM)

+

[ 3.a

+

3.b 3.c 3.d 3.e INVERSE BORON WORTH (FIG 3)

CRITICAL BORON CONCENTRATION ppmBI% Aklk 4.3.1 CORRECTED CRITICAL BORON CONCENTRATION (Boron React) 4.1 I =

[

(1 nvery;oron) x (-1) x Yo Aklk 4.3.2 FINAL MIXED BORON DEPLETION CORRECTION FACTOR (NAS Display 10, Control Room Log, Nuclear Engineering)

ESTIMATED MEASURED CRITICAL BORON CONCENTRATION (4.3.1) / (4.3.2) =

ppmB 4.3.3 CALCULATED BY:

DATEITIME APPROVED BY (SRO):

DATEITIME Send copy of this Enclosure to Nuclear Engineering Send original to Operations for filing 12

ENCLOSURE 2 Estimated Critical Rod Position (3.2) 1103-1 5B Revision 34 Page 1 of 1 2.a 2.b 2.c.l 2.c.2 2.c.3 2.d 3.a 3.b 3.c 3.c.l NOTE Individual data entries may be completed in any sequence. Signoff of the Enclosure signifies the completion of the Enclosure calculation.

CALCULATION IS FOR AN ECP AT 532 k 2°F ON TAVE CYCLE BURNUP EFPD PPmB FINAL MEASURED BORON CONCENTRATION BORON DEPLETION CORRECTION FACTOR (NAS Display 10, Control Room Log, Nuclear Engineering)

DATE/T I ME FINAL CORRECTED BORON CONCENTRATION (2.c.l) X (2.c.2) =

PPmB Yo WD CRG 8 POSITION AT CRITICALITY FUEL EXCESS REACTIVITY (FIG 1)

% Aklk CRG 8 REACTIVITY WORTH (FIG 2)

INVERSE BORON WORTH (FIG 3)

BORON REACTIVITY WORTH

% Sklk ppmBI% Aklk x (-1) =

1

[ p2p:; /

InverTcBoran

% Aklk 3.d 3.e XENON REACTIVITY WORTH (PPC, NUCLEAR ENGR., FIG 4)

SAMARIUM AND PLUTONIUM BUILDUP (FIG 5)

% Aklk

\\

0 TIME SINCE LAST SHUTDOWN v

REACTIVITY DUE TO BUILDUP 4.1 INSERTED CRG 5-7 WORTH REQUIRED FOR CRITICALITY x (-1) =

(FUEL)

(CRG 8)

(BORON)

(XENON)

(SM)

+

3.d 3.e

+

[ 3.2

+

3.b 3.c.l H RS

% Aklk

% Ak/k 4.2 4.3 ESTIMATED CRITICAL ROD POSITION (FIG 6)

CRITICAL ROD POSITION TOLERANCE BAND (FIG 6 )

0 CHECK ONE:

% WD on CRG Steady State conditions if xenon (3.d) is 0.0 to -0.5% Ak/k 0 5% Aklk 0

Transient conditions if xenon (3.d) is more negative than -0.5% Aklk 0.8% Akl k MINIMUM ROD WITHDRAWAL LIMIT (4.1-4.3)

% WD on CRG MAXIMUM ROD WITHDRAWAL LIMIT (4.1+4.3)

% WD on CRG DATE/TI M E CALCULATED BY:

APPROVED BY (SRO):

DATElTlME Send copy of this Enclosure to Nuclear Engineering Send original to Operations for filing 13

ENCLOSURE 3 Actual Subcriticality Calculation (3.3) 11 03-1 5B Revision 34 Page 1 of 1 DATEITI ME 2.a CALCULATION IS FOR ASC AT TAVE 2.b PRESENT ROD POSITIONS CRG 1-4 100

%WD CRG 5

% WD 0

CRG 6 Yo WD 0

CRG 7

%WD 2.c INTEGRAL CONTROL ROD WORTH (Fig. 6) 3.2 INSERTED CRG 5-7 WORTH REQUIRED FOR CRITICALITY (ENCLOSURE 2, Step 4.1) 3.3 ACTUAL SUBCRITICALITY (Step 2.c - Step 3.1)

O/o A kl k Yo Aklk

% aklk CALCULATED BY:

DATElTlME APPROVED BY (SRO):

DATElT I M E Send copy of this Enclosure to Nuclear Engineering Send original to Operations for filing 14

U J T

FIGURE 2 11 03-1 5B Revision 34 Page 1 of 1 Cycle 14 Group 8 Integral Rod Worth HZP, No Xenon 0

10 20 30 40 50 60 70 80 90 100 Group 8 Position Each Div. = 2% WD 0 EFPD 200 EFPD 400 EFPD 612 EFPD NOTE Linearly interpolate between EFPD Curves.

16

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FIGURE 4 Cycle 14 Transient Xenon Reactivity Worth

-4.0,

-3.5

-3.0

-2.5

-2.0

-1..5

-1.0

-0.5 0.0 I

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11 03-1 5B Revision 34 Page 1 of 1 10 20 30 40 50 60 70 80 90 100 Fach IXv. = 1 Hour Hours After Shuldown 18

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I 1 Number Estimated Critical Conditions ApplicabilitylScope USAGE LEVEL 34 Effective Date TMI Division This document is within QA plan scope X

Yes 50.59 Applicable X

Yes List of Effective Pages 2

10128102 No No 1

2 3

4 5

6 7

8 9

10 1 1 12 13 14 15 16 17 18 19 20 u

Revision 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 34 Page Revision Pase Revision Paqe Revision 1

3. a 3 b 3 i:

3.d 3.2 4.1 42 1.3 ENCLOSURE 1 Estimated Critical Bo ro n C o n ce n tration (3. I )

Iildividual data entries niay be completed in any sequence. Slyii-off of tlie Enclosure sisnifies the comoletion of the Enclosure calculation.

11 03-1 56 Revision 34 Page I of I CALCULATION IS FOR AN ECBP AT 532 t 2°F ON TAVE -5 5 9 DATE/TIME I

?r;E 'Q-EFPD CYCLE BURNUP PKESENT [LIEASURED BORON CONCEN7 RATION

~-

1 8 ppnIB DESIRED CRITICAL ROD POSITION 100

% WD CRG 1-4 CKG 5 I!W 74 WD CRG 6 75 YoWD CRG 7 0 '1, W D CRG 8 ZdT

% W D FUEL EXCESS REACTIVITY (FIG I )

DESIRED CRITICAL CRG 5-7 REACTIVITY WORTH (FIG 6)

L~ESIRED CRITICAL CRG 8 REACTIVITY WORTH (FIG 2)

XENON REACTIVITY WORTH (PPC, NUCLEAR ENGR, FIG 4)

SAMARIUM AND PLUTONIUM BUILDUP {FIG 5)

TIME SINCE SHUTDOWN 1 z. HRS REACTIVITY DUE TO BUILDUP (3,c\\. 3.7

'/u Lk/k

-' 7 60RON REACTIVITY WORTH REQUIRED FOR CRITICALITY I

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(SM) 1 3.a 3.b 3.c 3.d 3.e INVERSE BORON WORTH (FIG 3)

/J77,s ppmB/% lkik C KIT I CAL H OR0 N CON C ENTRATI 0 N 13 1 CORRECTED CRITICAL BORON CONCENTRATION

[(I iiverse Boron)

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4.2 4.1 (Boron React) '

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ESTIMATED MEASURED CRITICAL BORON CONCENTRATION (4 3 1) / (4 3 2) = L/c7 1)~>11it3 a -9$

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00 0.00 FIGURE 2 Cycle 14 Group 8 Integral Rod Worth HZP, No Xenon 20 40 50 60 70 Group 8 Position 80 90 100 Each Div. = 2% W D 1103-1 5B Revision 34 Page 1 of 1 NOTE Linearly interpolate between EFPD Curves.

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Y TMI SRO License Exam OW1 2/03 TMI-I OPERATOR TRAINING JOB PERFORMANCE MEASURE A.l-2 MAINTAIN MINIMUM SHIFT STAFFING, CONTROL OVERTIME Page 1 of 7

A. 1-2 Revision 0 0511 2/2003 TASK TITLE:

MAINTAIN MINIMUM SHIFT STAFFING, CONTROL OVERTIME TASK NUMBER:

343006060303 TIF: 2.6 KIA

REFERENCE:

System:

NA KIA:

2.1.4 Rating(RO/SRO):

2.3/3.4 Knowledge of shift staffing requirements.

POSITION:

S R O B R O O N L O O EVALUATION METHOD:

PERFORM SIMULATE 0 EVALUATION LOCATION:

SIMULATOR IN-PLANT CONTROL ROOM OTHER TASK STANDARDS:

Examinee identifies required actions to restore minimum staffing, and selects personnel in accordance with requirements to control overtime.

APPROXIMATE COMPLETION TIME: 30 minutes TIME-CRITICAL TASK COMPLETION TIME:

NA minutes w REQUIRED TOOLS OR MATERIALS:

OP-TM-101-111-1001, Shift Manning Requirements, Rev. 2.

Tech Spec 6.2.2 and Table 6.2-1, Amendment 21 9.

LS-AA-119, Overtime Controls, Rev. 1

REFERENCES:

OP-TM-101-111-1001, Shift Manning Requirements, Rev. 2.

Tech Spec 6.2.2 and Table 6.2-1, Amendment 219.

OP-AA-101-111, Roles And Responsibilities Of On-Shift Personnel, Rev. 0.

LS-AA-I 19, Overtime Controls, Rev. 1 ALTERNATE PATH JPM? NO SIMULATOR SETUP: NA INITIALIZATION: NA EVENT TRIGGERS: N/A MALFUNCTIONS: N/A REMOTE FUNCTIONS: N/A OVERRIDES: N/A MONITOR: N/A Page 2 of 7

A. 1-2 Revision 0 0511 2/2003 READ TO STUDENT v

When I tell you to begin, as the Unit Supervisor, you are to PERFORM THE STEPS NECESSARY TO ENSURE THAT YOUR SHIFT IS APPROPRIATELY STAFFED. Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions. Perform procedure steps as if you were actually performing the task.

INITIAL CONDITIONS:

Reactor power is loo%, with ICs in automatic.

The time is 2355.

The shift is staffed as indicated on the provided Shift Manning Log.

Third CRO John Doe is not licensed. He is in the Auxiliary Building performing an Independent Verification of a valve lineup change.

INITIATING CUE:

When I tell you to begin, as the Unit Supervisor, you are to PERFORM THE STEPS NECESSARY TO ENSURE THAT YOUR SHIFT IS APPROPRIATELY STAFFED.

ARE THERE ANY QUESTIONS?

TIME CRITICAL NO W

Page 3 of 7

JPM INSTRUCTION SHEET DIRECTIONS TO STUDENT:

When I tell you to begin, as the Unit Supervisor, you are to PERFORM THE STEPS NECESSARY TO ENSURE THAT YOUR SHIFT IS APPROPRIATELY STAFFED. Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions. Perform procedure steps as if you were actually performing the task.

INITIAL CONDITIONS:

Reactor power is loo%, with ICs in automatic.

The time is 2355.

The shift is staffed as indicated on the provided Shift Manning Log.

Third CRO John Doe is not licensed. He is in the Auxiliary Building performing an Independent Verification of a valve lineup change.

INITIATING CUE:

When I tell you to begin, as the Unit Supervisor, you are to PERFORM THE STEPS NECESSARY TO ENSURE THAT YOUR SHIFT IS APPROPRIATELY STAFFED.

TIME CRITICAL: NO Page 4 of 7

A. 1-2 Revision 0.

OW1 212003 STEP STANDARD s/u CUE: Unit reactor Operator (URO) Barnes reports that his contact lenses have just popped out and are lost. He reminds you that he has a license restriction that requires him to wear corrective lenses.

His backup eyeglasses are missing and cannot be located.

CUE: If asked, report that there are no other licensed CROs on site.

CUE: If requested, provide Examinee with copy of:

0 0

OP-TM-101-111-1001, Shift Manning Requirements.

0 LS-AA-I 19, Overtime Controls.

0 Overtime Callout List.

Technical Specifications pages 6-1 and 6-2.

1 Examinee references Shift Manning Log to determine current shift manning status.

"2 Examinee references Shift Manning Log and determines that current staffing is acceptable.

"3

~~

Examinee references Technical Specifications and/or OP-TM-101-111-1001, Shift Manning Requirements, to determine minimum shift manning requirements for current conditions.

Technical Specifications:

TS 6.2.2.2.a requirement:

0 Each on-duty shift shall be composed of at least the minimum shift crew composition shown on Table 6.201.

At least one licensed Reactor Operator shall be present in the control room when fuel is in the core.

Two licensed CROs must be on shift when Tave >2OO0F.

Two licensed CROs are required to be in the Control Room during scheduled reactor shutdown.

TS 6.2.2.2.b requirement:

0 Table 6.2-1 requirement:

0 TS 6.2.2.2.c requirement:

0 OP-TM-101-111-1001, Shift Manning Requirement:

0 Three CROs (at least 2 RO) must be on shift when RCS Temperature >2OO0F.

Examinee initiates action to comply with Tech Spec requirement for two licensed ROs.

Examinee determines three CROs are required, two of which must be RO licensed. One licensed CRO is required to be in the Control Room.

Examinee initiates action to comply with Tech Spec requirement for two licensed ROs.

Page 5 of 7

(I 4

Examinee references LS-PA-119, Overtime Controls, to evaluate callout restrictions.

5 The Examinee cancels the callout to CRO X, even though he is low man on the cumulative overtime list.

6 NOTE: CRO is limited to working 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in 7 consecutive days.

Examinee identifies next individual to be called to replace the CRO.

STEP Examinee selects CRO Y for callout due to overtime restrictions on CRO X.

Examinee seeks a replacement for the second licensed CRO position left vacant by Barnes situation.

Examinee calls the Operations Scheduler or directly references the Overtime Callout List to identify a replacement CRO to be called.

CUE: Provide Examinee with Overtime Callout List that identifies CROs X, Y, and 2 as personnel to be contacted for callout.

A. 1-2 Revision 0 05/12/2003 STANDARD Examinee selects (CRO X) low man on overtime list for callout.

SIU I

END TASK Page 6 of 7

DATE 0511 2/2003 A. 1-2 Revision 0 05/12/2003 JPM CHANGE HISTORY PAGE REFERENCE TITLE OP-TM-101-111-1001, Shift Manning Requirements.

Tech Spec 6.2.2.2.

Tech Spec Table 6.2-1.

LS-AA-119, Overtime Controls.

DESCRIPTION (Include AI # if Appropriate)

~~

Original issue.

Page 7 of 7

Title I

CRO -SWITCHING AND TAGGING LOG Night Shift Wynne cough1 i n

w i 1 son Barnes Masters CRO Trainee Revision No.

33 I

Day Shift SHIFT MANNING LOG Complete this log by indicating by name those aualified individuals fulfilling the listed Shift Manning Duties. The CRO signature on the cover page indicates that the qualification requirements of the minimum shift manning has been verified and that personnel listed on the Fire Brigade are on the Fire Brigade Qualified Listing.

1 Shift Manager (SM) required at all times 1

In-Plant SupervisodSTA 1

Control Room Supervisor - May be waived <200°F by Director, Operations 3 Control Room Operators >2OO0F At least 2 must have RO license if >200°F Only 2 CROs with minimum of 1 RO license required if <200°F Designate resp. person for E-Plan notifications & Fire Brigade (3rd CRO or I&C), if 3rd RO is not available, then designate and notify I&C or other qualified individual.

URO ARO FIRE BRIGADE/

E-PIAN 4 Auxiliary Operators required at all times.

u (When >200 4 AOs required to satisfy AP 1029) t

Emergency response assigned IAW OS-24 Attachment E.

Fire Brigade - Minimum of 5 members per AP 1029 (suggested manning) 1 Maintenance Foreman 2

2 A Auxiliary Operators 1

RadTech (Fire Brigade must have 2 qualified AOs as a minimum. AOs if available (must be extra) may fill Maint. slots if not also assigned as safe shutdown AO)

Maintenance (Electrical or Mechanical Discipline)

Sec Rdgs SeclPri Floater (Fire Brigade)

Pri Rdgs OB - EOP20 - Minimum Manning (Fire Brigade)

Extra (Fire Brigade)

Extra SCBA Ansul SCBA Ansul Rad Tech Fire Brigade Leader John Doe t A1 thouse Kohl Flowers Lutz Murray I Fel denzer Murray Kohl Esworthy cobaugh Barth Baumbach

W Rank Lowest Middle Hia hest Overtime Callout List Overtime Charged This Quarter Name Overtime Hours Charged Phone CRO X 9

948-1 234 CRO Y 12 948-5678 CRO Z 14.5 948-6543

7 6 0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Vice President-TMI Unit 1 shall be responsible for TMI-1 operations and may, at any time, delegate his responsibilities in writing to the Plant Manager. He shall delegate the succession of his responsibilities in writing dyring his absence.

The Shift Manager (or during his absence from the Control Room, a designated individual), shall be responsible for the Control Room command function. A management directive to this effect signed by the Chief Nuclear Officer shall be reissued to all unit personnel on an annual basis.

I

\\--

6.1.2 I

6.2 ORGANIZATION 6.2.1 CORPORATE 6.2.1.I An onsite and offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

6.2.1.2 Lines of authority, responsibility and communication shall be established and defined from the highest management levels through intermediate levels to and including operating organization positions. These relationships shall be documented and updated as appropriate, in the form of organizational charts. These organizational charts will be documented in the Updated FSAR and updated in accordance with IO CFR 50.71e.

6.2.1.3 The Chief Nuclear Officer shall have corporate responsibility for overall plant nuclear safety and shall take measures to ensure acceptable performance of the staff in operating, maintaining, and providing technical support so that continued nuclear safety is assured.

\\,

6.2.2 UNIT STAFF 6.2.2.1 The Vice President-TMI Unit 1 shall be responsible for overall site safe operation and shall have control over those on site activities necessary for safe operation and maintenance of the site.

6.2.2.2 The unit staff organization shall meet the following:

a.

Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b.

At least one licensed Reactor Operator shall be present in the control room when fuel is in the reactor.

6-1 Amendment No. 7 1, 22, 77, !%, 179,242.7,278-,219

c TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION(iii) u LICENSE CATEGORY QUALIFICATIONS Tave 2 200" Tave 5 200" RO (iv) 1 N on-L icwszd Au si linn*

o p L1,ltor Shift Tcchnicnl Advisor Rcquirzd 2

l(ii) 1 None (i)

Does not include the Licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising (a) irradiated fuel handling and transfer activities onsite, and (b) all unirradiated fuel handling and transfer activities to and from the Reactor Vessel.

(ii)

May be on a different shift rotation than licensed personnel.

(iii)

Except for the Shift Manager, shift crew composition may be one less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an incoming shift crewman being late or absent.

L-,

(iv)

Pursuant to the requirements of 10 CFR 50.54(m).

6-2 Amendment No. -

,219

AmerGen TMI Cover Page I

Title Overtime Controls This document is within QA plan scope Yes No USAGE LEVEL 50.59 Applicable Yes 0 No 3

AmerGen 1 Revision No.

I 0 913 0102 Effective Date TMI - Unit 1 Number LS-AA-119

Exelan,.

1.

Nuclear OVERTIME CONTROLS (This revision is a major rewrite)

PURPOSE LS-AA-1 I 9 Revision 1 Page I of 9 Level 3 - Information Use 1.1.

Implements the requirements for controlling the use of overtime as applied to personnel within the scope of the Technical Specifications. In addition, this procedure implements the requirements for an overtime guideline deviation approval process to prevent situations where fatigue could reduce the ability of operating personnel to maintain the reactor in a safe condition. The requirements specified in this procedure only pertain to safety-related work.

1.2.

Ensures compliance with Technical Specifications which require that overtime worked by plant staff members performing safety related functions be limited and controlled in accordance with the NRC Policy Statement on working hours (Le., Generic Letter (GL) 82-1 2, Nuclear Power Plant Staff Working Hours).

I

.3.

Applicable Bargaining Unit agreements and arbitration awards will also be considered in addition to the GL 82-12 limitations when assigning overtime to personnel. These agreements and awards are addressed in a site-specific administrative document.

I

.4.

Scope I

.4.1.

The overtime controls shall apply to all plant I corporate personnel who perform safety-related work activities including the following but not limited to:

I. Senior reactor operators, reactor operators, and non-licensed operators working on shift actively manipulating and operating plant equipment; or personnel responsible for the immediate supervision of the performance of such activities.

2.

Health physics / Chemistry personnel, including both Health Physicists and Radiation Protection Technicians performing work that could directly impact dose to the public, assigned emergency response duties including in plant rescue teams, environmental monitoring and dose calculations or who handle, process or provide data and input to emergency response decision makers.

3.

Key maintenance personnel (including contractor, health physics, and chemistry personnel), who are responsible for the correct performance of maintenance, repair, modification, quality verification, or calibration of safety-related structures, systems or components; or supervisory personnel responsible for the immediate supervision of the performance of such activities.

4.

Engineering or other personnel who are responsible for the correct performance of surveillance tests, maintenance, repair, modification or calibration of safety-related structures, systems or components; or personnel responsible for the immediate supervision of the performance of such activities. In addition, these

LS-AA-I I 9 Revision 1 Page 2 of 9 functions include reactor engineers supporting reactivity manipulations during critical reactor operations.

1.4.2.

All hours, except as noted in Step 1.4.4, worked at any nuclear station (i.e., both safety-related and non-safety-related work), must be counted as hours accumulated toward the overtime guideline hour limitations specified in GL 82-1 2. Work activities on both safety-related and non-safety-related structures, systems, and components will contribute toward personnel fatigue; therefore, all work hours are considered in the accumulated time related to GL 82-12 overtime guideline limitations. Note that only safety-related work requires an overtime guideline deviation authorization when exceeding GL 82-1 2 limitations.

1.4.3.

Non-safety-related work activities do not require an overtime guideline deviation authorization.

1.4.4.

Activities whose time may be subtracted (Le., exempted) from the hours accumulated toward the GL 82-1 2 overtime guideline limitations.

1.

All shift turnover time for the on-coming and off-going shift. There is no specific limitation regarding the duration of turnover time that may be exempted.

2.

Casual personal time before and after work, (Le., time spent having a cup of coffee before the work period begins; or stopping in the cafeteria to have a soft drink after the work period ends prior to leaving the station).

2.

TERMS AND DEFINITIONS 2.1.

Safetv Related: The term "safety-related" as used in this procedure shall have the same meaning as the term "safety-related" as defined in 10 CFR 50.49(b)(I) which states, "This equipment is that relied upon to remain functional during and following design basis events to ensure the integrity of the reactor coolant pressure boundary; the capability to shut down the reactor and maintain it in a safe shutdown condition; or the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures.....I' Therefore, any work associated with "safety-related" structures, systems, or components, to be performed after the GL 82-12 overtime guidelines will be exceeded, requires an Overtime Guideline Deviation Authorization.

2.2.

Work Hours Countina Toward GL 82-12 limitations: As defined in Step 1.4.2.

2.3.

Work hours that may be exempted from GL 82-12 limitations: Activities listed in Step 1.4.4 above may be subtracted (Le., exempted) from the hours accumulated toward the GL 82-12 overtime guideline limitations.

LS-AA-I 19 Revision 1 Page 3 of 9 2.5.

2.5.1.

2.5.2.

2.5.3.

2.5.4.

2.5.5.

2.6.

Work Hours Requiring Overtime Guideline Deviation Authorization: Hours worked in excess of the GL 82-12 overtime guidelines shall be evaluated to determine if an overtime guideline deviation authorization is required prior to the work being performed. Overtime that is planned to be worked in excess of the GL 82-1 2 guidelines will require an overtime guideline deviation authorization only if the work is safety-related. Overtime that is planned to be worked in excess of the GL 82-1 2 guidelines does not require an overtime guideline deviation authorization provided the work is non-safety-related.

EXAMPLE:

If a maintenance person has worked 68 hours7.87037e-4 days 0.0189 hours 1.124339e-4 weeks 2.5874e-5 months within a six-day period and plans to work an eight-hour shift on the seventh day; and the first three hours of that eight hour shift will be spent working on safety-related equipment, followed by five hours of non-safety-related work, then an overtime guideline deviation authorization will goJ be necessary for any of this work. It should be noted that it does not matter if the 68 hours7.87037e-4 days 0.0189 hours 1.124339e-4 weeks 2.5874e-5 months was safety-related or non-safety-related work, since all work is counted toward the GL 82-1 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limitations.

GL 82-12 Overtime Guidelines An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any sevenday (168 hr) period.

A break of at least eight hours should be allowed between work periods.

Except during extended shutdown periods such as refueling outages, forced outages, major maintenance or major plant modifications, the use of overtime should be considered on an individual basis, (i.e., one person at a time, as necessary), and not for an entire work group or the entire staff on shift.

During extended shutdown periods such as refueling outages, forced outages, major maintenance or major plant modifications, the use of overtime, within the GL 82-12 guidelines, may be applied to an entire work group or the entire staff on shift.

GL 82-12 Overtime Guideline Deviation: Recognizing that very unusual circumstances may arise (e.g., emergency situations to protect the public safety, critical emergent work requiring specialized skills or actions needed to avoid an unnecessary shutdown), requiring deviation from the above guidelines, such deviation shall be authorized by the Plant Manager or designated manager. Authorized deviations to the working hour guidelines shall be documented on Attachment 1 and available for NRC review.

LS-AA-I 19 Revision 1 Page 4 of 9

3.

RESPONSIBILITIES 3.1.

Plant Management will ensure that enough plant personnel will be employed to maintain adequate shift coverage without routine heavy use of overtime. It should be recognized that due to the nature of some job functions, a normal work week may entail a nominal amount of time greater than 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months to accommodate shift turnover time and shift rotation schedules.

3.2.

Each individual is responsible for:

Monitoring their own overtime hours by using time keeping records or personal tracking methods.

Informing their supervisor prior to exceeding any GL 82-12 overtime guideline.

Informing their supervisor when fatigue or reduced mental alertness could negatively affect the individuals job performance.

3.3.

The coqnizant supervisor is responsible for:

Ensuring that no personnel exceeds GL 82-12 overtime guidelines without the appropriate prior authorization.

Monitoring personnel who have come from a different nuclear station to ensure that they will not exceed GL 82-12 overtime guidelines without the appropriate authorization.

Initiating the Overtime Guideline Deviation Authorization form (i.e., Attachment 1 ) as necessary.

Assessing each employee under their supervision for fatigue and mental alertness prior to an employee exceeding the GL 82-12 overtime guidelines and performing safety-related work.

Assessing each employee under their supervision to reconfirm their mental alertness during the course of the overtime deviation period. This assessment is not documented.

3.4.

The Plant Manager or designated manager is responsible for:

Oyster Creek The Department Managers, or higher levels of management shall authorize Overtime Guideline Deviations.

- 0 Approving and signing the Overtime Guideline Deviation Authorization form prior to an individual exceeding the GL 82-12 overtime guidelines and performing safety-related work. The Plant Managers approval may be received verbally by phone and noted as such on Attachment 1.

Reviewing overtime usage on a monthly basis to ensure excessive work hours have not been assigned and to assess the use of Overtime Guideline Deviation Authorizations for potential abuse.

LS-AA-I 19 Revision I Page 5 of 9 3.5.

..j 3.6.

4.

4.1.

4.1.1.

4.1.2.

4.1.3.

The Human Resources Department or other designated der>artment is responsible for generating a monthly overtime monitoring report. This report shall be submitted to the Plant Manager for review.

The Nuclear Oversight (NOS) Department will assess overtime control as required based on station performance at the direction of the NOS Manager consistent with the guidance in the NOS Continuous Assessment procedure. It should be noted that an audit of security gate times or.payroll times may not be an appropriate measure of GL 82-1 2 compliance since these measures may not be representative of safety-related work activities.

MAIN BODY Use of Overtime In the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, forced outages, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight.

I. Activities listed in Step 1.4.4 may be excluded from time counted toward this time limitation.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any seven-day (I 68 hr) period.

1.

Activities listed in Step 1.4.4 may be excluded from time counted toward these time limitations.

A break of at least eight hours should be allowed between work periods.

I. An eight hour break between work periods only becomes applicable if one of the limitations in Steps 4.1.I and 4.1.2 has been met or exceeded.

EXAMPLE:

Assume a worker has been working normal 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months days from 7:OO a.m. to 3:30 p.m. At 1O:OO p.m. one night, this worker gets called into work for an emergent issue and works until 1 :00 a.m. The worker goes home and subsequently returns to work for his normally scheduled shift at 7:OO a.m.

Even though this individual did not get eight hours off before he was called into work at 1O:OO p.m. nor did he get eight hours off before he returned to work for his normally scheduled shift, an Overtime Guideline Deviation Authorization is not rewired since no GL 82-12 limitation listed in Stem I 4.1.1 and 4.1.2 was exceeded.

LS-AA-I 19 Revision I Page 6 of 9 4.1.4.

4.1.5.

4.2.

4.2.1.

4.2.2.

4.2.3.

4.3.

4.3.1.

2.

Shift turnover time and casual personal time noted in Step I

.4.4 may be included as part of the eight hour break between work periods.

Except during extended shutdown periods such as refueling outages, forced outages, major maintenance or major plant modifications, the use of overtime should be considered on an individual basis, (Le., one person at a time, as necessary), and not for an entire work group or the entire staff on shift.

During extended shutdown periods such as refueling outages, forced outages, major maintenance or major plant modifications, the use of overtime, within the GL 82-1 2 guidelines, may be applied to an entire work group or the entire staff on shift.

Overtime Guideline Deviation Recognizing that very unusual circumstances may arise (e.g., emergency situations to protect the public safety, critical emergent work requiring specialized skills or actions needed to avoid an unnecessary shutdown), requiring deviation from the above guidelines, such deviation shall be authorized by the Plant Manager or designated manager. Authorized deviations to the working hour guidelines shall be documented on Attachment 1.

Overtime hours worked for activities such as vacation coverage, normal shift coverage, do not constitute very unusual circumstances. Overtime worked for these types of activities should not exceed the GL 82-12 limitations.

During very unusual circumstances as noted in Step 4.2, overtime guideline deviations may be approved for an entire work group or the entire staff on shift. The cognizant supervisor(s) will assess each individual for fatigue and mental alertness prior to an individual exceeding the GL 82-1 2 overtime guidelines and performing safety-related work. The cognizant supervisor(s) will assess each individual over the duration an individual is working in excess of the GL 82-1 2 guidelines. These assessments are not documented.

I. If an individual is observed to be fatigued or exhibits unsatisfactory mental alertness, then the individual will be removed from performing safety-related work.

A single Overlime Guideline Deviation Authorization will be completed and remain active for the entire time period an individual exceeds a GL 82-12 overtime guideline.

When it has been determined that a deviation from the GL 82-1 2 overtime guidelines is necessary, then prior to an individual exceeding the GL 82-12 overtime guidelines and performing safety-related work, initiate Attachment 1, "Overtime Guideline Deviation Authorization." (cognizanf supervisor)

COMPLETE columns one through four of Attachment 1.

LS-AA-I 19 Revision 1 Page 7 of 9 b

4.3.2.

4.3.3.

4.3.4.

4.4.

4.4.1.

4.4.2.

4.5.

4.5.1.

- 4.5.2.

4.6.

4.7.

4.8.

4.9.

PROVIDE a description of the work to be performed verifying that the work is safety-related.

PROVIDE a justification describing why the safety-related work needs to be performed on overtime.

FORWARD the form to the Plant Manager or designated manager for approval.

REVIEW the Overtime Guideline Deviation Authorization form prior to an individual exceeding the GL 82-1 2 overtime guidelines and performing safety-related work.

(Piant Manager or designated manager)

SIGN the form as the approver. The Plant Managers or designees approval may be received verbaliy by telephone.

FORWARD the form back to the cognizant supervisor.

ASSESS each individual for fatigue and mental alertness prior to an individual exceeding the GL 82-1 2 overtime guidelines and performing safety-related work.

(cognizant supervisor)

ASSESS each individual for fatigue and mental alertness throughout the duration or the overtime deviation period.

If the supervisor observes the individual to be fatigued or not mentally alert, then that individual will not be assigned to perform safety-related work.

COMPLETE column 5 of Attachment I after an individual receives a rest period of sufficient length to be removed from the overtime deviation. (cognizant supervisor)

FORWARD original Attachment 1 to Document Services for retention and a copy to the Human Resources Department or designated department for overtime monitoring.

(cognizant supervisor)

ASSESS the use of Overtime Guideline Deviation Authorizations on a monthly basis to ensure excessive work hours have not been assigned and to assess the use of Overtime Guideline Deviation Authorizations for potential abuse. (Plant Manager or designated manager)

ASSESS overtime control as required based on station performance consistent with the guidance in the NOS Continuous Assessment procedure. It should be noted that an audit of security gate times or payroll times may not be an appropriate measure of GL 82-1 2 compliance since these measures may not be representative of safety-related work activities. (NOS Manager)

5.

5.1.

6.

6.1.

6.2.

6.3.

7.

7.1.

LS-AA-119 Revision I Page 8 of 9 DOCUMENTATION LS-AA-119, Attachment I, Overtime Guideline Deviation Authorization, forms shall be retained and be available for NRC review.

REFERENCES Technical Specifications for each Station:

Braidwood TS 5.2.2.d Byron TS 5.2.2.d Clinton TS 5.2.2.e Dresden TS 5.2.2.d LaSalle TS 5.2.2.d Limerick TS 6.2.2.f Oyster Creek TS 6.2.2.2.i Peach Bottom TS 5.2.2.d Quad Cities TS 5.2.2.d TM I TS 6.8.1.j NRC Generic Letter (GL) 82-12, Nuclear Power Plant Staff Working Hours NRC Generic Letter (GL) 83-1 4, Definition of Key Maintenance Personnel (Clarification of Generic Letter 82-12)

ATTACHMENTS LS-AA-119, Attachment I, Overtime Guideline Deviation Authorization

LS-AA-1 I 9 Revision I Page 9 of 9

~

I ATTACHMENT I Overtime Guideline Deviation Authorization Page 4 of 1 I

I I

The following individuals are needed to perform safety-related work and will exceed the overtime guidelines of Generic Letter (GL) 82-1 2.

I I

Note I The cognizant supervisor will complete the information in columns 1 - 4 above. This form is then forwarded to the Plant Manager (or designed Manager) for review.

Note 2:

The Plant Manager's (or designated manager's) approval shall be completed prior to an individual exceeding the overtime guidelines and performing safety related work. The Plant Manager's approval may be received verbally by telephone. This form is then forwarded back to the cognizant supervisor.

The cognizant supervisor will perform an assessment of each individual's fatigue and mental alertness prior to the start of performing safety-related work; and throughout the duration of the overtime deviation period. If the supervisor observes the individual to be fatigued or not mentally alert, that individual will not be assigned to perform safety-related work.

The cognizant supervisor will complete column 5 after the overtime deviation period is complete.

Note 3:

Note 4:

Note 5:

GL 82-12 Overtime Guideline to be exceeded:

A-more than 16 consecutive hours B - more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period C - more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period D - more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a seven day (168 hr) period E - less than an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months break between work periods Description of safety-related work to be accomplished:

Justification for needing overtime:

Submitted by:

I Cognizant Supervisor Time Date Approved by:

I Plant Manager (or designated manager)

Time Date

Exelan,,

1.

1.1.

2.

2.1.

2.2.

3.

3.1.

4.

Nuclear OP-TM-I 01 -1 I 1 -1 001 Revision 2 Page 1 of 3 TMI Training and Reference Material SHIFT MANNING REQUIREMENTS PURPOSE To provide a consolidated listing of all shift manning requirements for on shift personnel.

TERMS AND DEFINITIONS SRO - Senior Reactor Operator holding an Active NRC license as defined in 10 CFR 55.

CRO - Reactor Operator holding an Active I NRC license as defined in 1 OCFR55.

RESPONSIBILITIES It is the responsibility of the Shift Manager on duty to ensure these manning requirements are met at all times.

MAIN BODY NOTE Specific permission must be granted by the Plant Manager and the Site Vice President to deviate from the below listed requirements.

4.1.

Listed below are the minimum shift operations manning requirements:

Plant >2OO0F RCS TemPerature 1 Shift Manager (SM) 1 Control Room Supervisor (SRO)

Plant ~200°F RCS Temperature 1 Shift Manger (SM) 1 Control Room Supervisor*

3 Control Room Operators***

(at least 2 RO) 2 Control Room Operators (at least 1 RO) 4 Auxiliary Operators**

4 Auxiliary Operators Shift Technical Advisor N/A

OP-TM-1 01 -1 11 -1001 Revision 2 Page 2 of 3 4.2.

4.3.

4.4.

4.4.1.

4.4.2.

4.4.3.

4.4.4.

4.4.5.

4.4.6.

5.

5.1.

L.-/

May be waived by the Senior Manager, Operations. Either a qualified SRO, Shift Manager, or Control Room Supervisor must be on shift at all times when below 200°F.

- The minimum shift crew composition of 4 Auxiliary Operators assumes 4 qualified operators.

a.

Two of the Auxiliary Operators must meet the requirements to be designated as Fire Brigade members.

b.

Two of the Auxiliary Operators must meet the requirements to be assigned as personnel designated to facilitate the safe shutdown of the Unit.

- - Except for the Shift Manager, shift crew composition may be one less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an incoming shift crewman being late or absent.

A minimum of 1 SRO and 1 RO must be in the Control Room at all times when RCS temperature is >2OO0F.

At least 1 SRO temperature is <2OO"F.

1 RO must be in the Control Room at all times when RCS A Fire Brigade of at least 5 members shall be maintained on-site at all times.

Only those personnel who have satisfactorily completed the required fire fighting training shall be assigned as Fire Brigade members.

The Fire Brigade shall not include those personnel required in the Control Room as noted in 4.2 and 4.3 above.

The Fire Brigade shall not include those personnel necessary for the safe shutdown of the Unit as specified in the Technical Specifications The duty STA shall not satisfy any Fire Brigade position.

The members making up minimum shift manning shall be listed on the Control Room Log Sheet each shift.

The members of the Fire Brigade shall be listed on the Control Room Log Sheet each shift.

DOCUMENTATION None

OP-TM-101-1 i 1-1 001 Revision 2 Page 3 of 3 REFERENCES TMI Unit 1 Technical Specifications Section 6.2.2.2 1038, "Administrative Controls - Fire Protection Plan"

-d 6.

6.1.

6.2.

7.

ATTACHMENTS 7.1.

None

TMI SRO License Exam 05/12/03 TMI-I OPERATOR TRAINING JOB PERFORMANCE MEASURE A. 2 USE STATION DRAWINGS TO PREDICT IMPACT OF INSTRUMENT FAILURE Page 1 of 6

A.2 Revision 0 O W 1 212003 TASK TITLE:

USE STATION DRAWINGS TO PREDICT IMPACT OF INSTRUMENT FAILURE TASK NUMBER:

3420070303 TIF: 2.2 KIA

REFERENCE:

System:

NA KIA:

2.1.24 Rating(RO/SRO):

2.813.1 POSITION:

SRO[XIROO N L O O EVALUATION METHOD:

PERFORM SIMULATE 0 EVALUATION LOCATION:

SIMULATOR 0 IN-PLANT 0 CONTROL ROOM 0 OTHER TASK STANDARDS:

Examinee identifies impact of isolating MU-42-FS on RC-P-I D.

RC-P-1 D cannot be started from the Control Room.

APPROXIMATE COMPLETION TIME: 30 minutes TIME-CRITICAL TASK COMPLETION TIME:

NA minutes REQUIRED TOOLS OR MATERIALS:

u 208-1 13 Rev. 15.

SS-209-065, Rev. 9.

Marker to identify electrical contacts affected.

REFERENCES:

302-661, Rev.52.

208-1 13 Rev. 15.

SS-209-065, Rev. 9.

ALTERNATE PATH JPM? NO SIMULATOR SETUP: NA INITIALIZATION: NA EVENT TRIGGERS: N/A MALFUNCTIONS: N/A REMOTE FUNCTIONS: NIA OVERRIDES: N/A MONITOR: N/A Page 2 of 6

A.2 Revision 0 0511 2/2003 READ TO STUDENT When I tell you to begin, you are to USE ELECTRICAL PRINTS TO PREDICT THE IMPACT THE CLEARANCE REQUEST (TO MECHANICALLY ISOLATE MU-42-FS) ON OPERATION OF RC-P-I D. As you reference the electrical prints, use the marker provided to circle the effected flow switch contacts that produce the operational impact you predict.

INITIAL CONDITIONS:

RCS heatup is in progress.

RC-P-1A and RC-P-1B are operating.

RC-P-IC and RC-P-1 D are not operating.

Intermediate Closed Cooling Pump IC-P-?A is operating.

Operators are prepared to start RC-P-I D.

A piping leak has identified at MU-42-FS.

Maintenance has submitted a Clearance Request that MECHANICALLY isolates MU-42-FS in order to terminate and repair the leak.

INITIATING CUE:

When I tell you to begin, you are to use electrical prints (208-1 13, Rev. 15, and 209-065, Rev. 9) to predict the impact of the clearance request (to mechanically isolate MU-42-FS) on operation of RC-P-1 D. Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions. As you reference the electrical prints, use the marker provided to circle the effected flow switch contacts that produce the operational impact you predict.

u ARE THERE ANY QUESTIONS?

TIME CRITICAL: NO Page 3 of 6

JPM INSTRUCTION SHEET v

DIRECTIONS TO STUDENT:

When I tell you to begin, you are to USE ELECTRICAL PRINTS TO PREDICT THE IMPACT THE CLEARANCE REQUEST (TO MECHANICALLY ISOLATE MU-42-FS) ON OPERATION OF RC-P-1 D. Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions. As you reference the electrical prints, use the marker provided to circle the effected flow switch contacts that produce the operational impact you predict.

INITIAL CONDITIONS:

RCS heatup is in progress.

RC-P-1A and RC-P-1 B are operating.

RC-P-IC and RC-P-I D are not operating.

Intermediate Closed Cooling Pump IC-P-1A is operating.

Operators are prepared to start RC-P-I D.

A piping leak has identified at MU-42-FS.

Maintenance has submitted a Clearance Request that MECHANICALLY isolates MU-42-FS in order to terminate and repair the leak.

INITIATING CUE:

When I tell you to begin, you are to use electrical prints (208-1 13, Rev. 15, and 209-065, Rev. 9) to predict the impact of the clearance request (to mechanically isolate MU-42-FS) on operation of RC-P-1 D. Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions. As you reference the electrical prints, use the marker provided to circle the effected flow switch contacts that produce the operational impact you predict.

TIME CRITICAL: NO Page 4 of 6

A. 2 Revision 0 05/12/2003 1

Denotes Critical Elements STEP STANDARD SIU NOTE: Steps may be performed in any sequence.

CUE: Provide Examinee with the following tools:

0 Print 209-065, Rev. 9.

0 Examinee obtains a copy of print 209-065 Auxiliary Relays of MU42-FS (LOW RC PP. TOTAL SEAL INJECTION FLOW).

Marker to identify electrical contacts i 2

3 4

5

6

--It-

~ ~ _ _ _

~

Examinee circles affected flow switch and auxiliary relay contacts on print 209-065, in accordance with the initiating cue.

Examinee determines impact of low flow condition following isolation of the flow switch.

Examinee obtains a copy of print 208-1 10, RC Pump ID.

Examinee circles contacts affected by low flow condition following isolation of MU24-FS, in accordance with the initiating cue.

Examinee determines impact of low flow condition following mechanical isolation of the flow switch on RC-P-ID operation.

fected.

Examinee obtains a copy of 209-065 Auxiliary Relays of MU42-FS (LOW RC PP. TOTAL SEAL INJECTION FLOW).

Examinee circles the following three contacts (associated with impact on RC-P-1 D) on print 209-065 affected by low flow condition at MU24-FS following isolation by Maintenance:

Examinee circles one contact operated directly from MU42-FS (labeled 86/1FS, low setpoint) in the center of the print.

Examinee circles two contacts operated by auxiliary relay 80WMU42-FS2 associated with RC-P-1 D (at bottom of the print).

NOTE: Examinee may circle additional contacts 0

0 associated with the other three RCPs.

Examinee may verbalize the following description of impact on print 209-065:

1.) MU42-FS low flow condition energizes auxiliary relays 80WMU42-FSl and 80WMU42-FS2.

2.) When auxiliary relay 80WMU42-FS2 energizes:

RC-P-ID start circuit contact (C31 to C3) opens.

RC-P-ID trip circuit contact (PT to TI) closes.

0 0

Examinee obtains a copy of print 208-1 13 RC Pump ID.

Examinee circles the following two contacts on print 208-1 13:

0 0

Examinee circles one MU42-FS contact in RC-P-1 D start circuit.

Examinee may circle one additional MU42-FS contact in RC-P-1 D trip circuit.

Examinee describes impact of low flow condition following mechanical isolation of the flow switch on RC-P-1 D operation:

RC-P-I D can not be started from the control room, due to failure of the low seal injection flow starting interlock.

0 END TASK Page 5 of 6

A.2 Revision 0 05/12/2003 JPM CHANGE HISTORY PAGE REV1 SI ON 0

DATE 05/12/2003 REFERENCE DESCRl PTlON TITLE (Include AI # if Appropriate) 208-1 13 Rev. 15.

SS-209-065, Rev. 9.

Page 6 of 6

1. ETROPOL I T A N E D I SON COt.IPA:!Y T H R E E E I L E I S L A N D NUCLEAR S T A T I O N U N I T I P

ECTR 1 C A L ELEMENTARY D I AGFIAt.1 D. C.

& MISCELLANEOUS REFERENCE DWGS.:

NOTES:

LEGEND:

SS-208-Gi31 INDEX: Ss-2)g-ooi 08041280H I. SEE D N G D8041276K,D804127~K, 2, RELAYS - CLARK

4LJ6-3

~ G U L A T O R Y REQUIRED GILBERT ASSOCIATES, INC.

M A O E /D$

ENGINEERS AND CONSULTANTS R u o w a. PENNA cnum.

D,I cF. DFN. J REV. cn. APP. DATE SQ. CF.D II/ 8 4192 jSS-209-065 3

I

, 1 2

WORK O R D E R S I Z E D R A W I N G R E V PN I

10-3-3-13 W T M i % 0-3-3-14

'I

+$:

80X/MU42-FSl AXCC FD92 4%

AXCC w

1 N

8OX/MU42-FS 1 C3 1 c 3 T

+ - -

RCP-1A RCP-1C

. SPARES s s-208-1 10 ss-208-112 I

uc(o91r AUXI IARY RELAYS OF: MU42-FS.

(LOW Rc PP. TOTAL SEAL INJECTION FLOW)

FD219 3 A 0-3 15 0-3-3-16 t

I 1

80X/MU42-FS2 AXCL A XCL N

r c3\\

c3 RCP-SS-208 REVISED TO INCORPORATE

'FCN-C085931,

1 L

cMA& k///Y/

I ATS 1% -

REV IDRAFT~CHECK~

APPROVED I D A

~

APP B

RCP-10 111 SS-208-113 SPARES

'I li!

4-I

I I

+&OM IO 0

A. 3 Revision 0 05/12/2003 TMI-I OPERATOR TRAINING JOB PERFORMANCE MEASURE A. 3 LIQUID RELEASE PERMIT APPROVAL Page 1 of 6

A.3 Revision 0 05/12/2003 TASK TITLE:

LIQUID RELEASE PERMIT APPROVAL TASK NUMBER:

068C010101 Initiate, make and complete a liquid waste release.

TIF: 2.5 KIA

REFERENCE:

System:

Generic WA:

2.3.6 Rating

2.1/3.1 POSITION:

SRO [XIRO 0 NLO EVALUATION METHOD: PERFORM [XI SIMULATE 0 EVALUATION LOCATION: SIMULATOR 0 IN-PLANT 0 CONTROL ROOM OTHER IxI TASK STANDARDS: Examinee does NOT approve the liquid release, due to exceeding ODCM Calendar Quarter cumulative organ (Liver) dose limit of 1.5 Mrem.

APPROXIMATE COMPLETION TIME: 30 minutes TIME-CRITICAL TASK COMPLETION TIME: NA REQUIRED TOOLS OR MATERIALS:

Releasing Radioactive Liquid Waste, 661 0-ADM-4250.01, Rev. 16.

Offsite Dose Calculation Manual, 6610-PLN4200.01, Rev 23.

REFERENCES:

Releasing Radioactive Liquid Waste, 6610-ADM4250.01, Rev. 16.

Offsite Dose Calculation Manual, 6610-PLN4200.01, Rev 23.

ALTERNATE PATH JPM? NO SIMULATOR SETUP:

INITIALIZATION: NA EVENT TRIGGERS: NA MALFUNCTIONS: NA REMOTE FUNCTIONS: NA OVERRIDES: NA MONITOR: NA Page 2 of 6

A. 3 Revision 0 0511 212003 READ TO STUDENT When I tell you to begin, you are to (AS SHIFT MANAGER) REVIEW AND APPROVE A LIQUID RELEASE PERMIT IN ACCORDANCE WITH 661 0-ADM4250.01, RELEASING RADIOACTIVE LIQUID WASTE. Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions. Perform procedure steps and make notifications as if you were actually performing the task.

INITIAL CONDITIONS:

The plant is stable at 100% power.

WDL-T-I IA, Waste Evaporator Condensate Storage Tank, is isolated and "DO NOT OPERATE' tagged.

A Liquid Release Permit has been initiated for WDL-TI IA.

WDL-TI 1A level is 5.8 feet.

The current time is 2000.

INITIATING CUE:

Review and approve the Liquid Release Permit for WDL-TI I A in accordance with 6610-ADM4250.01, Releasing Radioactive Liquid Waste.

ARE THERE ANY QUESTIONS?

TIME CRITICAL: NO Page 3 of 6

JPM INSTRUCTION SHEET u

DIRECTIONS TO STUDENT:

When I tell you to begin, you are to (AS SHIFT MANAGER) REVIEW AND APPROVE A LIQUID RELEASE PERMIT IN ACCORDANCE WITH 6610-ADM4250.01, RELEASING RADIOACTIVE LIQUID WASTE. Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions. Perform procedure steps and make notifications as if you were actually performing the task.

INITIAL CONDITIONS:

The plant is stable at 100% power.

WDL-T-1 IA, Waste Evaporator Condensate Storage Tank, is isolated and "DO NOT OPERATE" tagged.

A Liquid Release Permit has been initiated for WDL-TI IA.

WDL-TI 1A level is 5.8 feet.

The current time is 2000.

INITIATING CUE:

As Shift Manager, review and approve the Liquid Release Permit for WDL-TI IA, in accordance with 661 O-ADM-4250.01, releasing Radioactive Liquid Waste.

TIME CRITICAL: NO Page 4 of 6

A. 3 Revision 0 0511 212003 STEP STANDARD SIU Accompanying documents are:

Forms 1621-1,1621-2,1621-3,16214, and 1621-5.

Dose Summary Report - Year-to-Date.

Dose Summary Report - Calendar Quarter to-Date.

Dose Summary Report - Month-to-Date.

Dose Summary Report for this release.

"3

4 Examinee ascertains Shift Manager signature responsibilities and dose limits for liquid releases by referencing 6610-ADM-4250.01, Releasing Radioactive Liquid Waste.

NOTE: Offsite Dose Calculation Manual, 6610-PLN4200.01, may be referenced to ascertain dose limits for releasing radioactive liquid waste.

Examinee reviews Cumulative Dose Summary for the current year to verify projected doses for this release will not cause the station to exceed individual dose limits from radioactive materials in liquid effluents released to the site boundary.

Examinee reviews Cumulative Dose Summary for the current quarter to verify projected doses for this release will not cause the station to exceed individual dose limits from radioactive materials in liquid effluents released to the site boundary.

Examinee signs Form 1621-3 (line 30) to approve the release IAW 6610-ADM-4250.01 step 4.10.4.

661 0-ADM-4250.01, Releasing Radioactive Liquid Waste, section 4.10.4, describes Shift Manager responsibilities.

6610-ADM-4250.01, Releasing Radioactive Liquid Waste, section 3.2, describes quarterly and yearly dose limits for releasing radioactive liquid.

661 0-PLN4200.01, Offsite Dose Calculation Manual, section 2.2.1.I, describes quarterly and yearly dose limits for releasing radioactive liquid.

In accordance with 661 0-ADM-4250.01, Releasing Radioactive Waste, section 3.2.b, during any calendar year, dose will be:

13 mrem to the whole body.

Dose will be 11 0 mrem to any organ.

Examinee determines that the calendar year dose limits will not be exceeded.

In accordance with 661 0-ADM-4250.01, Releasing Radioactive Waste, section 3.2.a, during any Calendar Quarter, dose will be:

51.5 mrem to the whole body.

Dose will be 15 mrem to any organ.

Examinee determines that the projected whole body dose (1.67 mrem) WILL EXCEED the quarterly whole body dose limit (1.5 mrem).

The examinee DOES NOT SIGN Form 1621-3 to approve this release, due to exceeding total Offsite Dose Calculation Manual (ODCM) allowable dose for the calendar quarter.

END TASK Page 5 of 6

A. 3 Revision 0 091 2/2003 JPM CHANGE HISTORY PAGE DESCRIPTION (Include AI # if Appropriate)

Initial issue.

Page 6 of 6

Number TMI - Unit 1 Radiological Controls Procedure

le L'

Releasing Radioactive Liquid Waste 661 0-ADM-4250.01 Revision No.

16 EXHIBIT 4 Form 1621-1 (Example)

Page 1 of 5 OPERATIONS INPUT TO LIQUID RELEASE PERMIT (41 Release Number 1200XXX035

11) Datenime: 2 Davs ~ ~ 0 ~ 1 1 0 0 Requester (Signature):

Shift Manager I

(signed)

(Shift Mgr.)

12) x WDL-T-11A Tank A (31 Tank Put on Recirculation (min. of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )

Time: I000 Date:rl Davs ACIO Tank Isolated and "Do Not Operate" Tagged:

Time: I100 date:^ Davs A ~ O Tank Volume Ft. 5.8 gallons 4044 Recirc Time >40 Hours 3, SignedlDate Shift Mananer/(Signed)/ Dated Yesterday 132) (Record all start and stop date times)

Tank Level at Start of Release ft.

gal.2 Time Release Stopped Tank Level at End of Release ft.

g a l. 1 Time Date Tank Volume Released (Actual) gal.

Time Release Started MDCT Effluent Totalizer at Stop gal.

Time Date MDCT Effluent Totalizer at Start gal.

Total Time of Release Total Dilution Flow gal.

Minutes (33)

Actual Release Rate = Actual Gallons Released =

min Utes gPm

~

~~~

~

t34)

Cancelled or Partial Release State reason this release was cancelled or only partially released:

C35) Release data completed and chemistry notified of actual gallons released:

Signature DatelTime Shift Manager Date/Time Shift Manager Date/Time All data required on this form has been completed.

Completed Release Permit has been forwarded to Rad Eng.

1 Tank level IS read on WDL-4-132 for W D L - ~ - I l A or WDL-u-133 for WDL-1-11 B WDL-E-126 may be used rf primary instruments are not available.

u

Number TMI - Unit 1 Radiological Controls Procedure Title i-,'

661 0-ADM-4250.01 Revision No.

Releasing Radioactive Liquid Waste (4) RELEASE NUMBER I

~2ooxxx035 I

16 NOTE RESULT AFTER COMPLETION FORWARD THIS DATA SHEET ALONG WITH COPIES OF THE GAMMA AND TRITIUM ANALYSIS TO RADIOLOGICAL CONTROLS FOR INCLUSION IN THE RELEASE PERMIT.

SIGNATURE (PRINTISIGN)

LIMIT TANK RECIRCULATED 2 8 HRS.

BY: RM Punliese I (signed) DATE/TIME Today I 0005 RELEASE SAMPLE(S) COLLECTED BY: RM Pugliese I (signed) DATE/TlME Today I 0010 RELEASE GAMMA SCAN BY: RM Pugliese I (signed) DATEnlME Today I O 0 1 8 RELEASE TRITIUM ANALYSIS BY: RIM Pugliese I (siqned) DATE/TlME Today I0050 WEEKLY COMPOSITE SAMPLES BY: RM Pugliese I (signed) DATE/TlME Today I O 1 0 0 II PH I

5.39 I

4.5-9.5 (NOTE 1) 11 II BORON 26 PPM NOTE

1.

2.

The limit of 4.5 - 9.5 will ensure that the NPDES limit of 6 - 9 is not exceeded at the main station discharge to the Susquehanna River.

Must be 10 uMHO and pH is less than 6.0 or greater than 9.0, release must be approved by Chemistry Supervisor or his designee with a written evaluation attached to release form.

TMI - Unit 1 Radiological Controls Procedure Title u

Number 6 6 1 0 -AD M 4250.0 I Revision NO.

Form 1621-3 (Example)

Page 3 of 5 Releasing Radioactive Liquid Waste Radiological Controls Input to Liquid Release Permit 16 (4) Release Number (241 Minimum Estimated Time for Release 1

150 LZOOXXX035 Minutes Instrument Readings:

Expected Reading (23)FR-84 2.7E+01 gpm (301 Release Information Completed By S. Edelman I (signed)

Rad. Controls Tech. A (Sign)

Release Recommended By D. Viola I (signed)

Group Rad. Controls Supervisor (Sign)

Reading at Reading After Reading After Reading After Reading After Start 114 Complete 112 Complete 314 Complete Release Comp.

Release Approved By Shift Manager (Final Approval) 1311 Release Data:

u WDL-V-257/RM-L6 Hi Radiation Interlock Test Sat.

initials (1 9)

FR-146 Alarm Setpoint at 7.2E+03 gpm FR-146 (Low MDCT Flow) Test Sat.

(221 FR-84 Alarm Setpoint at 3.00E+01 gpm FR-84 (Hi Liq. Release Flow) Test Sat.

initials initials initials initials RM-L6 Operable per SP 1301-1 (Check Source) initials

Form 1621-4 (9)

Nuclides H-3 661 0-ADM4250.01 Revision 16 Page 4 of 5 L200XXX035 Release Number I

(IO)

(1 1)

(12)

(14)

(15)

Specific (10) x (14)

Controlling Activitv Sensitivity of Monitor Specific Activity Effluent Conc.

Controlling RM-L6 to each

Response

pCi/ml pCi/ml Conc.

Nuclide (CPm) 3.12E-04 2E-3.

1.56E+02

\\-

(13)

Required D.F.

(81 Radiological Analyses Review by:

(GRCS)

D. Viola I (siqned)

(1 6)

Monitored Response of RM-L6 Above Bkgd.

Date/Time Today I AM I

Ce-141 I

I 3E-5.

I I

5.8E7 I

I Ce-144 I

I 3E-6.

I I

1.I E7 I

I CO-58 I

I 2E-5.

I I

1.2E8 I

Cr-51 I

I 5E-4.

I I

1.3E7 I

(171 Dilution Factor Required Based on Boron =

26PPM =

3.71 E+01 Record Results in Step (20) 0.7 PPM

4 Releasing Radioactive Liquid Waste Form 1621-5 Page 5 of 5 16 Contact Control Room Operator for following data

a.

MDCT Flow 8.00E+03 gpm (must be >5000 gpm)

MINIMUM MDCT FLOW (0.9 x usable MDCT Flow from 11 8 ) a) =

Form 1621-3, (must be > 5000 GPM).

Radionuclide D.F. From ((1 3) 4.57E+02 Boron D.F. From (17) 3.71E+Ol Gorse D.

Name of CRO contacted 7.20E+03. Record on Step (1 91 on MAXIMUM = 1.57E+02 DF RR max = Min MDCT from (19) 7,20E+03

= 3.00E+Ol R. R. MAX (5.55 to 30 gpm)

Required D.F. from (20) 1.57E+02 FR-84 ALARM SETPOINT = RR Max. Record this value at (22) on Form 1621-3.

RR Actual = 0.9 x RR max (3,OE+01) = 2.7E+01 gpm Record this value at (5 to 27 gpm) (23) on Form 1621-3.

Estimated time of release = Est. Vol. to be released 4044 nal = 150 MIN RR Actual (from (23))2.7OE+Ol gpm Record this value at (24) on Form 1621-3.

Estimated Reading of RM-L6:

5.99E+01 cpm (From (16)) -c I

.00E+03 cpm (RM-L6 Background) = 1.06E+03 cpm Record this value on Step (25) on Form 1621-3.

Estimated Reading of RM-L7:

5.99E+01 cpm (From (16))

1.57E+02

+

1.00E+02 cpm (RM-L7 Background) = 1.00E+02 cpm Dilution Factor (From (20))

Record this value on Step (26) on Form 1621-3.

CUMULATIVE DOSE

SUMMARY

Month (This Month) to Date Cum. Dose (Bound.)

Cum. Dose (Cr. Rec.)

GENERATED Today Inhalation Meat Gr. Plane Cow/MWI Vegetation Tot. Dose 4.91 E-01 1.19E-01 O.OOE+OO 1.58E-03 2.68E-03 5.37E-03 9.24E-04 2.23E-04 00.00E+00 1.64E-03 2.77E-03 5.56E-03 Dose Due to Liquid Releases (MREM)

Tot. Body 1 Liver I

Bone 1

Thyroid I

Kidney 1

Lung I

GI - LLI I

Cum.Dose I 5.1E-01 I 1.36E+00 I 1.39E+00 I 5.38E-01 I I.lIE+OO I 7.59E-01 I 6.05E-01 Includes Batch Releases 029 thru 035 & Cont. Releases 540-542 Dose Due Part./lodine/Tritium Releases (MREM)

CUMULATIVE DOSE

SUMMARY

This Quarter - 2003 GENERATED Today Gamma Air Beta Air Dose Due to Liquid Releases (MREM)

Cum. Dose (Bound.)

Cum. Dose (Cr. Rec.)

1 Tot.Body I Liver I

Bone I

Thyroid I

Kidney 1

Lung I

GI - LLI I Cum.Dose I 1.67E+OO 1 1.63E+00 I 1.39E+00 I 5.38E-01 I I.lIE+OO I 7.59E-01 I 6.05E-01 Includes Batch Releases 029 thru 035 & Cont. Releases 540 thru 542 Inhalation Meat Gr. Plane CowlMWl Vegetation Tot. Dose 8.91E-04 2.16E-04 O.OOE+OO 1.58E-03 2.68E-03 5.37E-03 9.24E-04 2.23E-04 O.OOE+OO 1.64E-03 2.77E-03 5.56E-03 Dose Due Part./lodine/Tritium Releases (MREM)

CUMULATIVE DOSE

SUMMARY

2003 GENERATED Today Gamma Air Beta Air Dose Due to Liquid Releases (MREM)

Cum. Dose Tot. Body I Liver I

Bone I

Thyroid I

Kidney I

Lung I

GI - LLI I Cum. Dose I 2.91E+00 I 2.79E+00 I

2.65E+00 I 1.88E+00 I 1.24E+00 I

2.95E+00 I 2.15E+00 Includes Batch Releases 001 thru 035 & Cont. Releases 501 thru 624 5.03E-04 3.23E-04 Dose Due Part./lodine/Tritium Releases (MREM)

Cum. Dose (Bound.)

Cum. Dose (Cr. Rec.)

Inhalation Meat Gr. Plane Cow/MWI Vegetation Tot. Dose 1.88E+00 4.77E-01 4.43E-08 3.51E+00 5.94E+00 1.91 E+OO 1.91 E+OO 1.89E+00 4.82E-01 2.83E-08 3.54E+00 5.99E+00

DOSE

SUMMARY

REPORT 10CFR50 DOSE REPORT FOR ADULT (MREM)

Release # 2OOXXXO35 GENERATED: Today

i I Number TMI - Unit 1 Radiological Controls Procedure AmerGen Title 661 0-ADM-4250.01 Revision No.

Releasing Radioactive Liquid Waste Applicability/Scope USAGE LEVEL 16 Effective Date 1

16 2

16 3

16 4

16 5

16 6

16 7

16 8

16 9

16 10 16 I 1 16 12 16 13 16 14 16 15 16 16 16 TMI Division This document is within QA plan scope X

Yes 50.59 Applicable X

Yes 1

3 10123/02 No No

TMI - Unit 1 Radiological Controls Procedure L, Title Releasing Radioactive Liquid Waste 1.0 PURPOSE Number 661 0-ADM-4250.01 Revision No.

16 This procedure describes the requirements applicable to radioactive liquid discharges to unrestricted areas and the monitoring programs, computer codes designed to ensure compliance with these regulations.

2.0 APPLICABILITY This procedure applies to Operations,.Chemistry, and Radiological Controls personnel responsible for actions to release liquid effluent to the environment.

3.0 DEFINITIONS 3.1 Liquid Radioactive Waste ODCM Concentration Limits (ODCM Part 1 Control 2.2.1.1)

The concentrations of radioactive material released at any time from TMI-1 to the environment shall not exceed ten times the 10 CFR 20, Appendix B, Table 2, Column 2 Effluent Concentrations, except that total dissolved or entrained noble aases shall not exceed 3 X I O 3 Gilml.

Liquid Radioactive Waste ODCM Dose Limits (ODCM Part 1, Control 2.2.1.2)

The dose to an individual from radioactive materials in liquid effluents released from TMI-1 to the site boundary shall be limited to:

3.2

a.

During any calendar quarter to I 1.5 mrem to the whole body I 5 mrem to any organ.

b.

During any calendar year to i 3 mrem to the total body and < 10 mrem to any organ.

4.0 PROCEDURE 4.1 Prerequisites A.

Prior to the release of liquid to the station effluent from the Waste Evaporator Condensate Storage Tanks A or B, a Liquid Release Permit must be obtained and estimated dose calculations performed.

NOTE Exhibit 3 provides a chart illustrating the typical sequence of processing a release.

B.

These tanks shall be recirculated a minimum of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to being sampled. (The time the tank is on recirculation while filling the tank may be considered in the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months requirement.) This shall be done to insure proper mixing in the tank(s) and to assure that a representative sample of the tank can be drawn for release calculations. Tank shall be left on recirculation until release is initiated.

2

TMI - Unit 1 Radiological Controls Procedure L

Title Releasing Radioactive Liquid Waste W

Number 6610-ADM-4250.01 Revision No.

16 4.2 4.3 4.4 C.

Prior to sampling and until the release is complete, the input to the tank shall be isolated.

D.

The river flow is obtained from the York Haven Dam at 848-7277. (The value in CFS converts to gpm by multiplying by 449). If this source is unavailable:

a.

Additional sources of river flow are the National Weather Service recording at 1-888-881-7555, or the USGS website on the Internet.

b.

If the river flow cannot be obtained, an average 75 year monthly historical value may be used from Exhibit 2.

NOTE In these INSTRUCTIONS the item number (XI of Section 4.0 will refer to the identically numbered section of the Liquid Release Permit.

NOTE For any corrections made throughout the release permit, the information being replaced shall be changed in accordance with AP 1001G.

The Shift Manager or designee initiates the request for release by completing (1) and is responsible to complete 421 and 13) on Form 1621-1.

The tank level is recorded, and the volume in gallons determined by Operations as read by WDL-LI-132 (133). The form is then turned over to the Chemistry Department for sampling/analyses. WDL-LR-126 may be used if other indication is not available.

The next sequential release number is assigned by Chemistry on Forms 1621-1 and 1621-2 MI.

Appropriate information is recorded on Exhibit 5 of Reference 6.3. The release name must consist of 10 characters composed of:

Character 1:

Must be an "L", signifying a liquid release Character 2,3,4,5:

Represents the year of release, ex. 2000 Character 6,7:

Represents month of release, ex. 05 Character 8,9,10:

Represents Number of release, ex. 014 (must always use 3 numbers for release Number)

Example:

L200005008, would be the release for the 8th liquid release of the year, 2000, released in May.

3

Number TMI - Unit 1 Radiological Controls Procedure L

Title Releasing Radioactive Liquid Waste 661 0-ADM4250.01 Revision No.

16 4.5 4.6 4.7 4.8 The Chemistry Technician will sample the designated tank for the required volume needed to perform all analyses required to release tank, after verlfying that the tank has been on recirculation for a minimum of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (51. Sampling and analysis is in accordance with Reference 6.1, and shall include the following:

Gamma isotopic analysis Tritium analysis 0

pH, conductivity, and boron analysis The technician's name and the date and time of analyses are recorded on Form 1621-2 a7l. After all analyses are completed, copies of all radiological analyses are turned over to Rad. Controls along with Forms 1621-1 and 1621-2.

NOTE Gross Alpha Analysis - The gross alpha is not a limiting condition for release, therefore is not performed on each batch. Gross alpha will be evaluated in a proportional monthly composite per Reference 6.6.

If the LAGER Computerized Release Permit is available, the following steps will be performed:

4.7.1 The Radiological Controls Technician will enter the information required by the computerized release permit, print it out, and sign 130).

4.7.2 The GRCS will review all analyses and sign and date the permit 48).

4.7.3 Proceed to Section 4.10 of this procedure.

If the LAGER Computerized Release Permit is not available, but the backup spreadsheet WECST.XLT is available, the following steps will be performed.

4.8.1 The Radiological Controls Technician will enter the information required by the spreadsheet, print it out, and sign (30).

4.8.2 The GRCS will review all analyses and sign and date the permit (81.

4.8.3 Proceed to Section 4.10 of this procedure.

4

Number TMI - Unit 1 Radiological Controls Procedure Title Releasina Radioactive Liauid Waste 66 I O-AD M-4250.Ol Revision No.

16 4.9 If both the LAGER Computerized Release Permit and WECST.XLT are not available, the Radiological Controls Technician/GRCS/Radiological Engineer will complete appropriate information on permit forms 1621-3, 1621-4, and 1621-5 using the following steps.

4.9.1 4.9.2 4.9.3 4.9.4 4.9.5 4.9.6 4.9.7 4.9.8 4.9.9 4.9.10 4.9.1 1 4.9.12 Nuclides normally found are listed as well as space for any additional nuclides identified 191.

The specific activities (SA) of positively identified nuclides are recorded 1101. No calculations are required for any nuclides not positively identified.

10 CFR 20 Appendix B, Table 2, Column 2 Effluent Concentrations are tabulated except H-3 which is listed at 2 times the IOCFR20 concentration 11 11. If other nuclides are identified, enter the value from I O CFR 20, Appendix B, Table 2, Column 2. The 10CFR20 concentrations are 10% of the limits in the TMI ODCM. TMI-1 ODCM limits for dissolved and entrained gases, also require that the total dissolved and entrained gases not be in excess of 3E-3 pCi/ml. Therefore, each noble gas is listed at 10% of that limit.

Enter the ratio of < I O 1 to 11 11, in 1121.

Enter the sum of all items in column 1121 in 1131. This will estimate the Dilution Factor (DF) for releasing radionuclides the specified concentrations. Record the results at Step 1201 of Form 1621-5.

Monitor (RM-L6) gamma sensitivity for each nuclide is listed in (141. If RM-L6 sensitivity is not listed for a particular nuclide, determine the predominant gamma energy (MeV) of the particular nuclide and use Exhibit I to obtain the nuclide's RML-6 sensitivity, or contact Radiological Engineering.

Enter the product of 11 01 x 1141 for each positively identified nuclide listed in 191.

Sum the values listed in (151 in 1161. This is the expected value of the RM-L6 response (above background).

The Radiological Controls technician will use the boron result to calculate a dilution factor 1171, and then transfer that factor to 1201.

Obtain from the Control Room Operator (CRO) the present MDCT flow in gpm from FR-146. The MDCT flow shall be > 5000 gpm. The name of the CRO supplying this information is to be recorded.

The minimum MDCT flow (1 91 used to calculate the release rate (RR) shall be 90 percent of the usable MDCT flow. This is used as the FR-146 setpoint; however, the setpoint may not be less than 5000 gpm. Record this value on Form 1621-3.

The dilution factor determine in 1131 is compared with the dilution factor based on boron from 1171. The maximum is the actual dilution factor required for release.

5

TMI - Unit 1 Radiological Controls Procedure L.

Title Releasing Radioactive Liquid Waste 4.9.13 The RR Max is computed by dividing the minimum MDCT flow from (191 by the maximum required D.F. from (201. (If the RR Max calculated above is greater than 30 gpm the RR Max shall be 30 gpm). If the Max is < 5.55 gpm notify the Shift Manager and Group Radiological Controls Supervisor.

Number 661 0-ADM-4250.01 Revision No.

16 4.9.14 The FR-84 alarm setpoint (221 is RR Max. Record this value at 422) on Form 1621-3.

NOTE If FR-84 and/or FR-146 is not operable, refer to Table 2.1-1, 2a, and 2b in the TMI ODCM, Part 1.

4.9.15 The calculated release rate is 90 percent of RR Max. It must be greater than or equal to 5 gpm but less than or equal to 27 gpm. Record this value at (231 on Form 1621-3. If it is not, notify the Shift Manager.

The estimated time period to complete tank release is calculated by dividing the tank volume from (31 by the actual release rate from (23). Record this value at (24) on Form 1621-3.

4.9.16 4.9.1 7 The estimated reading of RM-L6 is calculated by adding (161 to the background reading of RM-L6. Record this value at (251 on Form 1621-3.

4.9.1 8 The estimated reading of RM-L7 is calculated by dividing (161 by the dilution factor required from (201, then adding that result to the background of RM-L7. Record this value at (261 on Form 1621-3.

4.9.1 9 Calculate estimated offsite dose from this release using Reference 6.8.

4.10 Release Signoffs 4.1 0.1 The Radiological Controls Technician A that completed the information required signs in the appropriate space (301, if not already signed.

4.1 0.2 The Group Radiological Controls Supervisor will evaluate the data, check the calculations and recommend the release be approved or disapproved 4301.

4.10.3 Radiological Controls shall then send the completed release permit with the input parameters (sample results) for this release, the associated dose (manually calculated or by computer), due to this release and updated copy of the monthly, quarterly, and yearly dose contributions (manually calculated or by computer), from 4 liquid effluent releases to the Shift Manager.

4.10.4 The Shift Manager confirms that the projected dose from this release, when added to the total integrated dose to date, will not exceed the total ODCM allowable dose. The Shift Manager then signs Form 1621-3, (301, to approve the release.

6

Number Radiological Controls Procedure

=-.-

Title I

TMI - Unit 1 I

66 1 O-AD M 4250.01 Revision No.

Releasina Radioactive Liauid Waste I

16 4.1 1 4.12 4.13 Operations is responsible for procuring sign offs in Section t31) on Form 1621-3. Operations with assistance from I and C shall complete the following in accordance with Reference 6.2.

a.

Test WDL-V257/RM-L6 Hi radiation interlock

b.

Set FR-146 Setpoint C.

Test WDL-V257/FR-146 interlock

d.

Set FR-84 Setpoint

e.

Test W D L-V257/FR-84 interlock NOTE Just prior to release of liquid effluent through WDL-V257, RM-L6 must be proven operable using the installed check source or equivalent (reference SP 1301-1).

f.

Operations shall source check RM-L6 (Reference SP 1301-1).

Operations may then commence the liquid release by performing Reference 6.2.

NOTE MDCT flow shall not be changed during the liquid release unless absolutely necessary.

When the release is complete, Operations will perform the following:

e If a release is cancelled or only partially released, Operations shall complete t34) on Form 1621-1.

e Time and date are filled in when release starts and ends 1321.

e Tank level and volume as determined by WDL-LI-132 (133) at start and completion of The total gallons of MDCT Flow are calculated by subtracting the MDCT Flow Totalizer release are filled in. WDL-LR-126 may be used if other indication is not available (32).

e reading at start from reading at end of release on Form 1621-1 t32).

e Operations shall complete t33) on Form 1621-1 with the actual release rate.

e The Operations personnel that completed release data signs his name. Shift Manager will verify that all data from t31) thru (35) are complete and accurate.

e Operations will notify Chemistry of the actual volume (gallons) released for this release permit.

7

TMI - Unit 1 Radiological Controls Procedure L-Title Number 661 0-ADM-4250.01 Revision No.

Re1 easing Radioactive Liquid Waste 16 0

The Chemistry personnel notified will log the value in the Chemistry Log Book. This value will be used to perform Surveillance Procedure 1301-6.3.

0 The permit is then forwarded to the Radiological Engineering Department for updating liquid release records in accordance with Reference 6.3.

5.0 RESPONSIBILITIES Responsibilities of each group involved {Chemistry, Operations and Radiological Controls) in this procedure have been outlined in the procedure section (Section 4.0) of this procedure.

6.0 REFERENCES

6.1 Chemistry Procedure N1857 6.2 OP 1104-29s 6.3 661 0-ADM-4250.03 - Liquid Radioactive Release Records 6.4 LAGER User Manual 6.5 NPDES Permit PA 0008820 6.6 SP 1301-6.3 6.7 661 0-PLN-4200.01 - (ODCM) - Offsite Dose Calculation Manual 6.8 661 0-ADM-4250.05 - Dose Calculations for Liquid and Gaseous Effluents 6.9 AP 1001G 6.10 Appropriate Chemistry Gamma Spectroscopy Procedures 7.0 EXHIBITS 7.1 Exhibit 1 - RM-L6 Sensitivities 7.2 Exhibit 2 - Historical Monthly River Flow 7.3 Exhibit 3 - Liquid Release Flow Chart Summary 7.4 Exhibit 4 - Forms 1621-1 through 1621-5 (Examples) a

i EXHIBIT I ENERGY RESPONSE CURVE 643-90 DETECTOR IN A #I-33 0 J OAM/DIS 1

CALCULATCO VAUIES ENERGY In keV

+

ACTUAL TEST VALUES To determine the sensitivity for an isotope:

6610-ADM-4250.01 Revision 16 Page 1 of 1

1.

2.

3.

4.

Determine the predominant gamma energy of the nuclide Find the sensitivity for that energy from the above graph Multiply that sensitivity by the gamma yield The product is a conservative estimate of the monitors sensitivity to that nuclide.

9

TMI - Unit 1 Radiological Controls Procedure Title Releasing Radioactive Liquid Waste AVERAGE MONTHLY RIVER FLOW (gpm)

Number 661 0-ADM-4250.01 Revision No.

16 40.0 30.0 20.0 10.0 0.0 MILLIONS ( x 1 E6 )

17.1 Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec I Historical I 16.5 20.2 35.0 34.3 22.5 13.1 7.8 5.6 5.6 8.3 13.4 17.1 I I +Historical 1 Based on the Past 75 Years Data from the Safe Harbor Dam Hydro-coordinators.

10

Releasing Radioactive Liquid Waste SHIFT SUPERVISOR 1 INITIATES 16 EXHIBIT 3 LIQUID RELEASE FLOW CHART

SUMMARY

4 A-

+

TANK RELEASE ACTUAL #'S RAD. ENGR.

UPDATE CURIES AND CHEMISTRY 0 REVIEWS SHIFT MANAGER FINAL APPROVAL Page 1 of 1 MANUAL CALC-PERMIT SAMPLESIANALYSES I

v I

I RAD. CON.

COMPUTER DOSE CALCS.

CODE-PERMIT MANUAL DOSE CALC.

PERFORMED FIELD I

GRCS APPROVAL 1

0 P ERAT IO N S NOTIFIES CHEM.

OF ACTUAL GALLONS RELEASED FOR PERFORM I NG SURV.

PROCEDURE 11

TMI - Unit 1 Radiological Controls Procedure L/ Title Number 661 O-ADM-4250.01 Revision No.

W (4) Release Number (1) Datemime:

Requester (Signature):

(Shift Mgr.)

(2)

WDL-T-11A Tank A (31 Tank Put on Recirculation (min. of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) Time:

Date:

Tank Isolated and "Do Not Operate" Tagged: Time:

Date:

0 WDL-T-11 B Tank B Tank Volume Ft.

gallons Recirc Time 0

Sig ned/Date (32) (Record all start and stop date times)

Tank Level at Start of Release ft.

gal. 0 Time Release Stopped Tank Level at End of Release ft.

gal. 0 Time Date Tank Volume Released (Actual) gal.

Time Release Started MDCT Effluent Totalizer at Stop gal.

Time Date MDCT Effluent Totalizer at Start gal.

Total Time of Release Total Dilution Flow gal.

Minutes (331 Actual Release Rate = Actual Gallons Released =

mi nu tes 9 Pm (34) Cancelled or Partial Release State reason this release was cancelled or only partially released:

(35) Release data completed and chemistry notified of actual gallons released:

Sig nature Da te/Time All data required on this form has been completed.

Completed Release Permit has been forwarded to Rad Eng.

Shift Manager Datemime Shift Manager Daterrime 0 Tank level is read on WDL-LI-132 for WDL-T-11A or WDL-LI -133 for WDL-T-11B. WDL-LR-126 may be used if primary instruments are not available.

12

I Number I

Releasing Radioactive Liquid Waste 16 Form 1621-2 (Example)

Chemistry Data Sheet for Releasing Radioactive Liquid Waste Page 2 of 5 (41 RELEASE NUMBER 7 1

RESULT PH CONDUCTIVITY BORON NOTE LIMIT 4.5 - 9.5 (NOTE 1)

10 uMHO and pH is less than 6.0 or greater than 9.0, release must be approved by Chemistry Supervisor or his designee with a written evaluation attached to release form.

13

TMI - Unit 1 Radiological Controls Procedure

~-

Title Releasina Radioactive Liauid Waste Form 1621 -3 (Example)

Page 3 of 5 Number 6610-ADM-4250.01 Revision No.

16 Radiological Controls Input to Liquid Release Permit Instrument Readings:

Expected Reading 4231 FR-84 4251 RM-L6 (4) Release Num ber 7 1

Reading at Reading After Reading After Reading After Reading After Start 1/4 Complete 1/2 Complete 3/4 Complete Release Comp.

(241 Minimum Estimated Time for Release 1

Minutes 430)

Release Information Completed By Rad. Controls Tech. A (Sign)

Release Recommended By Release Approved By Group Rad. Controls Supervisor (Sign)

Shift Manager (Final Approval)

(311 Release Data:

u WDL-V-257/RM-L6 Hi Radiation Interlock Test Sat.

initials 4191 FR-146 Alarm Setpoint at 9 Pm FR-146 (Low MDCT Flow) Test Sat.

4221 FR-84 Alarm Setpoint at 9 Pm FR-84 (Hi Liq. Release Flow) Test Sat.

initials initials initials initials RM-L6 Operable per SP 1301-1 (Check Source) initials 4191 FR-146 4261 RM-L7 14

Form 1621-4

19) 661 0-ADM-4250.01 Revision 16 Page 4 of 5 110) 11 1) 1121 1141 11 5)
Specific 1101 x I141 Activity Specific Activity Effluent Conc.
Controlling RM-L6 to each
Response
Sensitivity of Monitor Controlling Release Number I
I Ce-141 (8)
Radiological Analyses Review by:
(GRCS) 3E-5.
I I
5.8E7 1
Datemime i
1 Ce-144 I
3E-6.
1.1E7 CO-58 I
2E-5.
1 I
1.2E8 1
CO-60 3E-6.
1.9E8 cs-I 34 I
9E-7.
I I
2.5E8 1
cs-I 37 1 E-6.
1.05E8 Fe-59 1-131 Mn-54 4171 Dilution Factor Required Based on Boron = Result (in PPM) =
Record Results in Step 1201 0.7 PPM 1 E-5.
1.1E8 1 E-6.
1.1E8 3E-5.
l.lE8 15 Mo-99 Zn-65 Cr-51 2E-5.
1.5E7 5E-6.
5.4E7 5E-4.
1.3E7 Ag-I 1 OM I
6E-6.
I I
3.2E8 1
Sb-125 4
3E-5.
8.3E7 Xe-I 33 3E-4.
I I
3.7E6 1
Xe-l33m 3E-4.
1.3E7 Xe-I 35 3E-4.
I I
1.2E8 1
Kr-85m 3E-4.
1.2E8 1131 Required D.F.
11 61 Monitored Response of RM-L6 Above Bkgd.
Number TMI - Unit 1 u
Releasing Radioactive Liquid Waste Radiological Controls Procedure I 661 0-ADM-4250.01 Title I Revision No.
16 Form 1621-5 Page 5 of 5 Contact Control Room Operator for following data
a.
MDCT Flow gpm (must be >5000 gpm)
MINIMUM MDCT FLOW (0.9 x usable MDCT Flow from (181 a) =
Form 1621-3, (must be > 5000 GPM).
Name of CRO contacted
. Record on Step (19) on Radionuclide D.F. From (13)
Boron D.F. From (17)
MAXIMUM =
DF R. R. MAX (5.55 to 30 gpm)
RR max = Min MDCT from (191 Required D.F. from (20)
FR-84 ALARM SETPOINT = RR max Record this value at (221 on Form 1621-3.
RR Actual = (0.9 x RR max
) =
gpm Record this value at (5 to 27 gpm) 123) on Form 1621-3.
Estimated time of release = Est. Vol. to be released aal =
MIN RR Actual (from (23))
9 Pm Record this value at (241 on Form 1621-3.
Estimated Reading of RM-L6:
cpm (From (161) +
cpm (RM-L6 Background) =
cpm Record this value on Step (25) on Form 1621-3.
Estimated Reading of RM-L7:
cpm (From (16))
+
cpm (RM-L7 Background) =
cpm Dilution Factor (From 120))
Record this value on Step (26) on Form 1621-3.
16
TMI SRO License Exam 5/12/2003 TMI-I OPERATOR TRAINING JOB PERFORMANCE MEASURE A.4 DAY # I Page 1 of 6
A.4 Revision 0 0511 212003 TASK TITLE:
EMERGENCY ACTION LEVEL IDENTIFICATION AND EVENT DECLARATION.
TASK NUMBER:
5001 045001 TIF: 3.35 KIA
REFERENCE:
System:
Generic KIA:
2.4.41 Rating:
2.314.1 POSITION:
S R O H R O O N L O O EVALUATION METHOD:
PERFORM SIMULATE EVALUATION LOCATION: SIMULATOR IN-PLANT CONTROL ROOM OTHER TASK STANDARDS: Examinee classifies the event as an ALERT under FA1 or MA4 within 15 minutes of direction to classify the event, and then completes EP-MA-114-100 MAROG Notifications Attachment 1 (PNMD Notification Form) to support initial off-site notifications.
APPROXIMATE COMPLETION TIME: 20 minutes.
TIME-CRITICAL TASK COMPLETION TIME:
Classification: 15 minutes.
REQUIRED TOOLS OR MATERIALS:
0 EP-AA-111 Emergency Classification and Protective Action Recommendations, Rev. 5a.
Exelon Nuclear Radiological Emergency Plan Annex for Three Mile Island (TMI) Station.
Table TMI 3-1, Emergency Action Level (EAL) Matrix.
EP-MA-114-100 MAROG Notifications Attachment 1 (PNMD Notification Form)
REFERENCES:
0 0
0 EP-AA-111 Emergency Classification and Protective Action Recommendations, Rev. 5a.
Exelon Nuclear Radiological Emergency Plan Annex for Three Mile Island (TMI) Station.
Table TMI 3-1, Emergency Action Level (EAL) Matrix.
EP-MA-114-100 MAROG Notifications Attachment 1 (PNMD Notification Form)
d ALTERNATE PATH JPM? NO SIMULATOR SETUP:
INITIALIZATION: NA EVENT TRIGGERS: NA MALFUNCTIONS: NA REMOTE FUNCTIONS: NA OVERRIDES: NA MONITOR: NA READ TO STUDENT When I tell you to begin, you are to CLASSIFY THE EVENT, AND COMPLETE THE MAROG NOTIFICATIONS Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions.
Perform procedure steps to identify the EAL,classify the event, and complete the initial notifications form as if you were actually performing the task.
FORM, ATTACHMENT I (PNMD NOTIFICATION FORM) TO SUPPORT INITIAL OFF-SITE NOTIFICATIONS.
Page 2 of 6
A. 4 Revision 0 0511 2/2003 INITIAL CONDITIONS:
The plant was stable at 68% power with only one operating Feedwater Pump. ICs was in manual mode.
SEQUENCE OF EVENTS:
1. One (Group 7) control rod dropped, requiring the operators to manually reduce reactor power to less than 60%.
2.
3.
4.
5.
6.
7.
8.
9.
After the automatic controller failed, Pressurizer level was controlled manually.
The only operating Feedwater Pump tripped.
Trip of both FW Pumps caused a turbine trip.
EFW automatically started, and is injecting flow into the OTSGs.
RPS failed to automatically execute an automatic reactor trip, but manual reactor trip was successful.
RCS leak into the Containment Building resulted in automatic ES Actuation.
Loss of RCS Subcooled Margin occurred.
MU-P-1 C was started manually following failure to automatically.
Current Conditions:
Reactor is shutdown.
RCS is 35°F subcooled.
The current time is 1605.
Wind Speed 12 mph.
Wind direction is from 295.
The EOF is NOT activated.
INITIATING CUE:
Based on these conditions, classify this event and complete EP-MA-114-100 MAROG Notifications Attachment 1 (PNMD Notification Form) to support initial off-site notifications.
ARE THERE ANY QUESTIONS?
TIME CRITICAL: YES Page 3 of 6
JPM INSTRUCTION SHEET DIRECTIONS TO STUDENT:
v When I tell you to begin, you are to CLASSIFY THE EVENT, AND COMPLETE THE MAROG NOTIFICATIONS FORM, ATTACHMENT 1 (PAIMD NOTIFICATION FORM) TO SUPPORT INITIAL OFFSITE NOTIFICATIONS.
Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions.
Perform procedure steps to identify the EAL,classify the event, and complete the initial notifications form as if you were actually performing the task.
INITIAL CONDITIONS:
The plant was stable at 68% power with only one operating Feedwater Pump. ICs was in manual mode.
SEQUENCE OF EVENTS:
1. One (Group 7) control rod dropped, requiring the operators to manually reduce reactor power to less than 60%.
2. After the automatic controller failed, Pressurizer level was controlled manually.
3. The only operating Feedwater Pump tripped.
4. Trip of both FW Pumps caused a turbine trip.
5. EFW automatically started, and is injecting flow into the OTSGs.
All 3 Emergency Feedwater Pumps are operating.
6. RPS failed to automatically execute an automatic reactor trip, but manual reactor trip was successful.
7. RCS leak into the Containment Building resulted in automatic ES Actuation.
8. Loss of RCS Subcooled Margin occurred.
9. MU-P-1 C was started manually following failure to automatically.
w Current Conditions:
Reactor is shutdown.
RCS is 35°F subcooled.
The current time is 1605.
Wind Speed 12 mph.
Wind direction is from 295.
The EOF is NOT activated.
INITIATING CUE:
Based on these conditions, classify this event and c o m p t e EP-M.
(PNMD Notification Form) to support initial off-site notifications.
.I 00 n ROG Notifications,ttachmer 1 TIME CRITICAL: YES Page 4 of 6
A.4 Revision 0 0511 2/2003 STEP STANDARD SIU NOTE: Record time that direction to classify the event is given.
I
2 3
Examinee obtains a copy of Exelon Nuclear Radiological Emergency Plan Annex for Three Mile Island (TMI) Station and/or Table TMI 3-1, Emergency Action Level (EAL) Matrix, to identify the specific EAL(s) applicable to current conditions, and to classify the event.
The examinee declares the event and identifies himself as the ED. (Examinee may indicate to the examiner that they would make the announcement.)
Examinee completes MAROG Notifications form, Attachment 1 (PNMD Notification Form) to support initial off-site notifications.
The examinee determines that an ALERT condition exists, specifically:
0 ALERT FAI, due to loss of RC System barrier (~25°F SCM) or 0
ALERT MA4, due to failure of automatic RPS System trip, with successful Manual Trip.
The examinee declares the event and identifies himself as the ED. (Examinee may indicate to the examiner that they would make the announcement.)
NOTE: This declaration is required to be completed within 15 minutes from the time the examiner provides direction to classify the event.
Time of Declaration:
~
~~~
Examinee completes MAROG Notifications form, (PNMD Notification Form) to support initial off-site notifications.
Fields to be completed:
2 - Classification, Affected Unit, Initial Declaration 3 - Brief Non-Technical Description 4 - Non-Routine Radiological Release Status:
NO non-routine radiological release in progress.
5 - PAR, not applicable 6 - Meteorology:
0 Wind from 295 degrees at 12 miles per END TASK Page 5 of 6
c A.4 Revision 0 OW1 2/2003 JPM CHANGE HISTORY PAGE REFERENCE DESCRIPTION (Include AI # if Appropriate)
Initial issue.
Page 6 of 6
Three Mile Island Station Annex Exelon Nuclear Table 3-2: TMI EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA4 INITIATING CONDITION Auto SCRAM NOT Successful EAL THRESHOLD VALUE
1. Failure of RPS to automatically INITIATE AND COMPLETE a reactor trip when ANY RPS set point has been exceeded Manual trip from Control Room was successful AND MODE APPLICABILITY 0
Power Operations (PwrOps) 0 Hot Standby (HStby)
Hot Shutdown (HSD) 0 Heatup / Cooldown (HU/CD)
. Startup(SU)
BASIS (References)
Reactor Protection System (WS) trip setpoints are designed and set to maintain the plant inside (less than) the Core Safety Limits.
An Alert is warranted because conditions exist that lead to potential loss of fuel clad or RCS inventory.
Successful follow up to the ATWS means that the Control Personnel were able to de-energize the Control Rod drives from the control room. This may occur by depressing the main or backup trip pushbutton. Additionally, the electrical bus may be de-energized from the Control Room.
'The activation of the Emergency Response Organization is essential to evaluate and possibly mitigate the consequences of the event.
This EAL satisfies NESP-007 Alert SA2.
December 2002 TMI 3-73 Revision Ob
Three Mile Island Station Annex Exeion Nuclear Tabie 3-2: TMI EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FA1 INITIATING CONDITION ANY LOSS or ANY POTENTIAL LOSS of EITHER the Fuel Cladding or Reactor Coolant System EAL THRESHOLD VALUE Comparison of conditions / values with those listed in Fission Product Barrier Matrix indicates:
LOSS OR LOSS POTENTIAL LOSS of the Fuel Cladding Barrier POTENTIAL LOSS of the Reactor Coolant System Barrier MODE APPLICABILITY Power Operation (PwrOps) 0 Hot Standby (HStby)
Hot Shutdown (HSD)
Startup(SU)
Heatup / Cooldown (HU / CD)
BASIS: (References)
The Fuel Cladding and the Reactor Coolant System are weigded more heavily lLLan the Containment Barrier.
A LOSS or POTENTIAL LOSS of either the Fuel Cladding or the Reactor Coolant System would be a substantial degradation in the level of plant safety.
Guidance for development of this EAL was taken from Recognition Category F in the NUMARCINEI Methodology for Development of Emergency Action Levels.
December 2002 TMI 3-33 Revision Ob
EP-MA-114-100 Revision la Page 10 of 15 ATTACHMENT I PA / MD NOTIFICATION FORM Page 1 of 2 UTILITY MESSAGE NO.
EMERGENCY DIRECTOR APPROVAL:
STATUS:
[ ] This is a drill.
[ ] This is NOT a drill.
1. This is
at [ ] LIMERICK I [ 1 PEACH BOTTOM Station.
My phone number is
p Completed by Communicator at fime notification is performed.]
. The current time is *
[ I ALERT DECLARED AT:
i ~ E S C A ~ T I O N
[ ] SITE AREA EMERGENCY TIME:
[ ] NO CHANGE
[ ] GENERAL EMERGENCY DATE:
1 1
[ ] REDUCTION
[ ] TERMINATED EAL#:
- IN CLASSIFICATION STATUS r 1 RECOVERY
3. BRIEF NON-TECHNICAL DESCRIPTION:
[ ] EVACUATE 360 DEGREES FROM 0 MILES (SITE BOUNDARY) TO 5 MILES;
[ ] EVACUATE 360 DEGREES FROM 0 MILES (SITE BOUNDARY) TO 10 MILES I I d 6
6. METEOROLOGY: Wind Direction (FROM)
STATUS:
Wind Speed is:
(MILES PER HOUR)
(DEGREES)
[ J This is a drill.
[ ] This is U T
a drill.
TMI SRO License Exam 517 2/2003 TMI-I OPERATOR TRAINING JOB PERFORMANCE MEASURE A.4 DAY #2 Page 1 of 6
A.4 Revision 0 05/12/2003 TASK TITLE:
EMERGENCY ACTION LEVEL IDENTIFICATION AND EVENT DECLARATION.
. b, TASK NUMBER:
5001 045001 TIF: 3.35 KIA
REFERENCE:
System:
Generic WA:
2.4.41 Rating:
2.3/4.1 POSITION:
S R O R R O O N L O O EVALUATION METHOD:
PERFORM SIMULATE 0 EVALUATION LOCATION: SIMULATOR IN-PLANT 0 CONTROL ROOM OTHER TASK STANDARDS: Examinee classifies the event as an ALERT under HA4, due to Non-Bomb Explosion inside the Vital Area, within 15 minutes of direction to classify the event, and then completes EP-MA-114-100 MAROG Notifications Attachment 1 (PNMD Notification Form) to support initial off-site notifications.
APPROXIMATE COMPLETION TIME: 20 minutes.
TIME-CRITICAL TASK COMPLETION TIME:
Classification: 15 minutes.
REQUIRED TOOLS OR MATERIALS:
0 0
0 0
EP-AA-111 Emergency Classification and Protective Action Recommendations, Rev. 5a.
Exelon Nuclear Radiological Emergency Plan Annex for Three Mile Island (TMI) Station.
Table TMI 3-1, Emergency Action Level (EAL) Matrix.
EP-MA-114-100 MAROG Notifications Attachment 1 (PNMD Notification Form)
EP-AA-111 Emergency Classification and Protective Action Recommendations, Rev. 5a.
Exelon Nuclear Radiological Emergency Plan Annex for Three Mile Island (TMI) Station.
Table TMI 3-1, Emergency Action Level (EAL) Matrix.
EP-MA-114-100 MAROG Notifications Attachment 1 (PNMD Notification Form)
REFERENCES:
0 0
0 ALTERNATE PATH JPM? NO SIMULATOR SETUP:
INITIALIZATION: NA EVENT TRIGGERS: NA MALFUNCTIONS: NA REMOTE FUNCTIONS: NA OVERRIDES: NA MONITOR: NA When I tell you to begin, you are to CLASSIFY THE EVENT, AND COMPLETE THE MAROG NOTIFICATIONS Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions.
Perform procedure steps to identify the EAL,classify the event, and complete the initial notifications form as if you were actually performing the task.
~-
FORM, ATTACHMENT I (PNMD NOTIFICATION FORM) TO SUPPORT INITIAL OFF-SITE NOTIFICATIONS.
Page 2 of 6
A.4 Revision 0 0511 212003 INITIAL CONDITIONS:
The plant was stable at 100% power with only one operating Feedwater Pump. ICs was in manual mode.
SEQUENCE OF EVENTS:
4
1. Selected RCS T-Hot instrument failed high, affecting ICs T-Ave indication and control.
2. A hydrogen gas leak reduced Main Generator gas pressure, requiring a forced load reduction to protect the generator. Because of an Integrated Control System malfunction the load reduction had to be performed manually.
3.
Following the load reduction, protective relay operation transferred loads off the 1 A Auxiliary Transformer to 1 B Auxiliary Transformer and Emergency Generator EG-Y-I B.
4. A major steam line rupture inside the Containment Building caused the reactor to trip.
5. Excessive OTSG heat transfer results in a core overcooling event, and ESAS actuation.
6.
Following isolation of feedwater sources to the affected OTSG, crew members were required take actions to prevent RCS reheat and re-pressurization.
7. All Emergency Feedwater Pumps started.
8. Control and termination of HPI flow was complicated by a stuck open High Pressure Injection valve.
Current Conditions:
Reactor is shutdown.
RCS is 45°F subcooled.
OTSG 1A is isolated and depressurized.
HPI flow has been terminated.
RCS pressure and temperature have been stabilized.
All Emergency feedwater Pumps are operating (on recirculation).
The current time is 1605.
Wind Speed 8 mph.
Wind direction is from 270 degrees.
The EOF is NOT activated.
u INITIATING CUE:
Based on these conditions, classify this event and complete EP-MA-114-100 MAROG Notifications Attachment 1 (PNMD Notification Form) to support initial off-site notifications.
ARE THERE ANY QUESTIONS?
TIME CRITICAL: YES Page 3 of 6
JPM INSTRUCTION SHEET DIRECTIONS TO STUDENT:
When I tell you to begin, you are to CLASSIFY THE EVENT, AND COMPLETE THE MAROG NOTIFICATIONS Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions.
Perform procedure steps to identify the EAL,classify the event, and complete the initial notifications form as if you were actually performing the task.
FORM, ATTACHMENT I (PAIMD NOTIFICATION FORM) TO SUPPORT INITIAL OFF-SITE NOTIFICATIONS.
INITIAL CONDITIONS:
The plant was stable at 100% power with only one operating Feedwater Pump. ICs was in manual mode.
SEQUENCE OF EVENTS:
1. Selected RCS T-Hot instrument failed high, affecting ICs T-Ave indication and control.
2. A hydrogen gas leak reduced Main Generator gas pressure, requiring a forced load reduction to protect the generator. Because of an Integrated Control System malfunction the load reduction had to be performed manually.
3. Following the load reduction, protective relay operation transferred loads off the 1A Auxiliary Transformer to I B Auxiliary Transformer and Emergency Generator EG-Y-1 B.
4. A major steam line rupture inside the Containment Building caused the reactor to trip.
5. Excessive OTSG heat transfer results in a core overcooling event, and ESAS actuation.
6. Following isolation of feedwater sources to the affected OTSG, crew members were required take actions to prevent RCS reheat and re-pressurization.
7. All Emergency Feedwater Pumps started.
8. Control and termination of HPI flow was complicated by a stuck open High Pressure Injection valve.
u Current Conditions:
Reactor is shutdown.
RCS is 45°F subcooled.
OTSG 1A is isolated and depressurized.
HPI flow has been terminated.
RCS pressure and temperature have been stabilized.
All Emergency feedwater Pumps are operating (on recirculation).
The current time is 1605.
Wind Speed 8 mph.
Wind direction is from 270 degrees.
The EOF is NOT activated.
INITIATING CUE:
Based on these conditions, classify this event and complete EP-MA-114-100 MAROG Notifications Attachment 1 (PNMD Notification Form) to support initial off-site notifications.
ARE THERE ANY QUESTIONS?
TIME CRITICAL: YES Page 4 of 6
A.4 Revision 0 OW1 2/2003
Denotes Critical Elements STEP STANDARD SIU INITIATING CUE: Identify the EAL and classify the event based on current plant conditions as required.
I The examinee determines that an ALERT condition
2 3
Examinee obtains a copy of Exelon Nuclear Radiological Emergency Plan Annex for Three Mile Island (TMI) Station and/or Table TMI 3-1, Emergency Action Level (EAL) Matrix, to identify the specific EAL(s) applicable to current conditions, and to classifv the event.
The examinee declares the event and identifies himself as the ED. (Examinee may indicate to the examiner that they would make the announcement.)
Examinee completes MAROG Notifications form, Attachment 1 (PNMD Notification Form) to support initial off-site notifications.
exists, specifically:
ALERT HA4, due to Non-Bomb Explosion inside the Vital Area.
The examinee declares the event and identifies himself as the ED. (Examinee may indicate to the examiner that they would make the announcement.)
NOTE: This declaration is required to be completed within 15 minutes from the time the examiner provides direction to classify the event.
Time of Declaration:
Examinee completes MAROG Notifications form, (PA/MD Notification Form) to support initial off-site notifications.
Fields to be completed:
2 - Classification, Affected Unit, Initial Declaration 3 - Brief Non-Technical Description 4 - Non-Routine Radiological Release Status:
AIRBORNE non-routine radiological release in progress (EF-P-1 operation in conjunction with OTSG tube leakage).
5 - PAR, not applicable 6 - Meteorology:
Wind from 270 degrees at 8 miles per hour.
END TASK Page 5 of 6
A.4 Revision 0 0511 2/2003 JPM CHANGE HISTORY PAGE REVISION 0
DATE 1
REFERENCE TITLE 5/12/2003 NA DESCRIPTION (Include AI # if Appropriate)
Initial issue.
Page 6 of 6
Three Mile Island Station Annex Exelon Nuclear Table 3-2: TMI. EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HA4 INITIATING CONDITION Fire OR Explosion Affecting Operability of Safety Systems Required for Safe Shutdown EAL THRESHOLD VALUES
1. Fire that affects the operability of one safety system train OR
2. Fire inside the Protected Area requiring off-site assistance OR
3. Unanticipated NON-bomb explosion (violent combustion / pressurized equipment failure) inside Vital Area NOTE: At a minimum, the plant vital areas listed Table H-1 shall be evaluated for damage based on fire / explosion location.
Table H-1 Plant Vital Area Structures Reactor Building Intake Building Intermediate Building Control Tower Diesel Generator Buildings Auxiliary & Fuel Handling Buildings MODE APPLICABILITY:
ALL Plant Conditions BASIS: (References)
The purpose of this EAL is to identify when the level of safety of the plant is in question because of a fire or explosion. The fire or explosion may be impacting safety systems directly (in a Vital Area) or indirectly (in the Protected Area) but it is challenging a Vital Area (Area where vital equipment for Safe Shutdown is located). Damage to equipment or structures inside Vital Area that could impact on the ability ofthe plant to protect the health and safety of the public.
Evaluate, as a minimum, the following areas (TMI-1) for damage based on fire location: Reactor Building, Intake Building, Intermediate Building, Control Tower, Aux. and Fuel Handling Building, and Diesel Generator Building.
Threshold Value 1: Considered to be met if a single Emergency Diesel Generator or Engineered Safeguards system string is rendered inoperable AND it is required to be operable for present plant conditions for event mitigation. This EAL is meant to satisflr NESP-007 EAL HA2.
December 2002 TMI 3-100 Revision Ob
Three Mile Island Station Annex Eaelon Nuclear Table 3-2: TMI EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HA4 - Contd BASIS (References) - Contd Threshold Value 2: Considers that extensive damage to a structure inside the Protected Area may affect normal day-to-day operations. This is especially true for the TMI-2 buildings that do not have water and off-site assistance is required to extinguish a fire. This EAL is meant to satisfy NESP-007 EAL HA2.
Threshold Value 3: Addresses violent unconfined combustion or a catastrophic failure of pressurized equipment. This EAL is meant to satisfy NESP-007 EAL HA2.
December 2002 TMI 3-101 Revision Ob
EP-MA-114-100 Revision l a Page 10 of 15 ATTACHMENT 1 PA / MD NOTIFICATION' FORM Page I of 2 UTILITY MESSAGE NO.
EMERGENCY DIRECTOR APPROVAL:
PERFORM INITIAL ROLL CALL PRIOR TO TRANSMITTING - Refer to back of Form TIME INITIAL ROLL CALL COMPLETED:
I
1. This is
at [ ] LIMERICK / [ ] PEACH BOTTOM Station.
STATUS:
[ ] This is a drill.
[ ] This is U T
a drill.
2. CLASSIFICATION:
AFFECTED UNIT(S):
THIS REPRESENTS NAN:
[ ] INITIALDECLARATION My phone number is
p Completed by Communicator at time notification is performed.]
. The current time is *
[ ] UNUSUAL EVENT
[ 1 ALERT DECLARED AT:
[ ]ESCALATION
[ ] SITE AREA EMERGENCY TIME:
[ ] NO CHANGE
[ ] GENERAL EMERGENCY DATE:
I
/
[ ] REDUCTION
[ ] RECOVERY
[ ]ONE [ ]TWO [ ]THREE
[ ] TERMINATED EAL#:
- IN CLASSIFICATION STATUS
3. BRIEF NON-TECHNICAL DESCRIPTION:
6. METEOROLOGY: Wind Direction (FROM)
STATUS:
Wind Speed is:
(DEGREES)
(MILES PER HOUR).
[ ] This is a drill.
[ ] This is NOT a drill.

ML031710473

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ML031710473

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