Text

Initial Exam 2008-301 Final Administrative JPMs
	ML091480143
Person / Time
Site:	Farley
Issue date:	05/28/2009
From:	; NRC/RGN-II
To:	; Southern Nuclear Operating Co
References
	Download: ML091480143 (120)
	v • d • e

HLT-32 ADMIN exam A.1.1 RO Page 1 of 6 A.I.IRO Conduct of Operations ADMIN 006AI.02 - RO TITLE: Determine required quantity of Boric Acid solution and Reactor Makeup water and integrator settings for makeup to the RWST.

TASK STANDARD: Determine the required quantity of Boric Acid and Reactor Makeup water to restore RWST level at the current Boric Acid Concentration, and correctly determine the setting of the Reactor Makeup system integrators and the potentiometer setting for the Boric Acid Flow controller.

PROGRAM APPLICABLE: SOT SOCT OLT --.X..... LOCT __

ACCEPTABLE EVALUATION METHOD: ~PERFORM SIMULATE DISCUSS EVALUATION LOCATION:

SIMULATOR CONTROL ROOM ----X-CLASSROOM PROJECTED TIME:

20 MIN SIMULATOR IC NUMBER:

N/A ALTERNATE PATH TIME CRITICAL PRA'----_

Examinee:

Overall JPM Performance:

Satisfactory 0

Unsatisfactory 0 Evaluator Comments (attach additional sheets if necessary)

EXANUNER: ________________ _

HLT-32 ADMIN exam A.1.1RO Page 2 of 6 CONDITIONS When I tell you to begin, you are to determine RWST Makeup quantity, Boric acid concentration, and integrator setting for make up to the RWST per FNP-l-S0P-2.3, Chemical And Volume Control System Reactor Makeup Control System, starting at Step 4.2.3.2 Makeup to Refueling Water Storage Tank (RWST). The conditions under which this task is to be performed are:

a. Unit 1 is at 100% power and stable.

b. RWST Level is at 37.7 feet.

c. RWST Boron concentration is at 2400 ppm.

d. On Service BAT concentration is 7001 ppm.

e. RWST Purification (Recirc) is NOT on-service.

f.

The Reactivity Spreadsheet is not available.

g. You are the extra plant operator and have been directed by the Shift Supervisor to perform SOP-2.3 steps 4.2.3.2 - 4.2.3.6 to:

1. Determine the quantity of blended flow required to raise level in the RWST from 37.7 feet to 39.5 feet while maintaining the current RWST Boron Concentration.

2. Determine the integrator settings for:

FIS 113, BORIC ACID BATCH INTEG FIS-168, TOTAL FLOW BATCH INTEG

3. Determine the potentiometer setting to makeup to the RWST at a reduced flow of 60 gpm total flow for:

FK-113, BORIC ACID MKUP FLOW EVALUATION CHECKLIST ELEMENTS:

STANDARDS:

START TIME I NOTE:

This is a classroom setting ADMIN JPM task.

1.

Determines Gallons needed per RWST Tank

Determines RWST Volume at curve 31B is 22,378 gallons.

39.5 feet=491064 gals Determines RWST Volume at 491064 - 468686 = 22378 37.7 feet=468686 gals Calculates total volume RESULTS:

(CIRCLE) addition= 22378 gals S / U 491064 - 468686 = 22378 gals

[no tolerance: whole numbers from a table]

HLT-32 ADMIN exam A.1.1 RO Page 3 of 6 EVALUATION CHECKLIST RESULTS:

ELEMENTS:

STANDARDS:

(CIRCLE)

2.

3.

NOTE:

Determines Boric Acid amount from Figure Determines from Figure 1, SOP-1, SOP-2.3, for the current concentration of 2.3 ratio of Boric Acid amount to 2400 ppm.

total amount from the ratio of Boric Acid flow to Total Flow.

Then calculates total Boric Acid in gallons to obtain total 22,378 gals of blended solution at 2400 ppm:

7664-7672 gals Boric Acid Solution S I U PER FIG. 1 PAGE 1:

22378( 41.1) = 7664.465gals 120 PER FIG 1 PAGE 3:

2237s( 41.14) = 7671.924 gals 120

[tolerance: 7664.0-7672 based on using either page 1 or page 3 numbers and rounding to nearest whole number of 7664 or rounding up for conservative 7672 gals.]

Determines the totalizer settings for Total Determines totalizer settings are:

flowFIS-168, TOTAL FLOW BATCH

FIS-168=22378 gals S I U INTEG and FIS-113, BORIC ACID

FIS-113=7664-7672 gals S/U BATCH INTEG: Based on Figure 1.

In element 4, FK-113 pot setting* is critical, but the manual position ofFIS-168 demand which corresponds to 60 gpm is not critical, since this controller would need to be adjusted while flow was present. There is no corresponding demand that will ensure 60 gpm flow prior to initiating flow and adjusting as necessary.

Examiner NOTE:

In element 4, IF applicant desires to raise the setpoint above the minimum required, ask them what setpoint they are going to use and ensure it is greater than the minimum required.

HL T -32 ADMIN exam A.1.1 RO EVALUATION CHECKLIST ELEMENTS:

4.

Determines the Minimum Flow controller potentiometer setting for FK-113, BORIC ACID MKDP FLOW for the 60 gpm Total flow directed by the Shift Supervisor at 2400 ppm Pot setting of 10.29 according to Figure 1, SOP-2.3 would correspond to 120 gpm total flow at 2400. Since the Boric Acid Flow Controller pot only goes to 10.0, the total flow must be reduced to less than 120 gpm within the capacity of the system. The Shift Supervisor has directed 60 gpm total flow (12012=60) in the initial conditions, and the boric acid flow for 120 gpm total flow must be divided by 2, and the pot setting must be divided by

2. FK-168 will need to be adjusted in manual to obtain 60 gpm.

Per Fig 1:

10.2912=5.145 Boric Acid FK-113 pot setting STOP TIME STANDARDS:

States that FK -168 will need to be adjusted in manual to obtain 60 gpm.

Based on Figure 1, states that MINIMUM setting for FK -113 is 5.145 pot setting*.

[Tolerance of 5.14 to 5.15 based on the accuracy of the pot indication which is Y2 the smallest increment of 0.02. The procedure directs initiating more boric acid flow than necessary to ensure the boric acid flow totalizer reaches the setpoint and stops Boric Acid Flow prior to the Total Flow totalizer reaching its setpoint. This flushes the Boric Acid from the lines and delivers it all to the RWST. There is no procedural requirement to limit the boric acid flow to a specified maximum amount.]

Terminate when all elements of the task have been completed.

Page 4 of 6 RESULTS:

(CIRCLE)

SID CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.

HL T-32 ADMIN exam A.1.1 RO GENERAL

REFERENCES:

1.

FNP-l-SOP-2.3 Version 48.0

2.

FNP-I-ARP-1.5 EG4, Version 50.0

3.

KIA:

G2.006A1.02 RO 3.0 SRO 3.6 GENERAL TOOLS AND EQUIPMENT:

Provide:

1. FNP-l-SOP-2.3, Version 48.0

2. Curves 31A & 31B

3. Calculator (or applicant may supply their own calculator)

COMMENTS:

Page 5 of 6

A.1.1RO (1 Page)

HANDOUT CONDITIONS When I tell you to begin, you are to determine RWST Makeup quantity, Boric acid concentration, and integrator setting for make up to the RWST per FNP-I-S0P-2.3, Chemical And Volume Control System Reactor Makeup Control System, starting at Step 4.2.3.2 Makeup to Refueling Water Storage Tank (RWST). The conditions under which this task is to be performed are:

a. Unit 1 is at 100% power and stable.

b. RWST Level is at 37.7 feet.

c. RWST Boron concentration is at 2400 ppm.

d. On Service BAT concentration is 7001 ppm.

e. RWST Purification (Recirc) is NOT on-service.

f.

The Reactivity Spreadsheet is not available.

g. You are the extra plant operator and have been directed by the Shift Supervisor to perform SOP-2.3 steps 4.2.3.2 - 4.2.3.6 to:

1. Determine the quantity of blended flow required to raise level in the RWST from 37.7 feet to 39.5 feet while maintaining the current RWST Boron Concentration.

2. Determine the integrator settings for:

FIS 113, BORIC ACID BATCH INTEG FIS-168, TOTAL FLOW BATCH INTEG

3. Determine the potentiometer setting to makeup to the RWST at a reduced flow of 60 gpm total flow for:

FK-l13, BORIC ACID MKUP FLOW

(

09/02/08 10:49: 10 T 1 FARLEY NUCLEAR PLANT SYSTEM OPERATING PROCEDURE FNP-I-SOP-2.3 CHEMICAL AND VOLUME CONTROL SYSTEM REACTOR MAKEUP CONTROL SYSTEM FNP-I-SOP-2.3 July 29,2008 Version 48.0 PROCEDURE USAGE REQUIREMENTS PER FNP-0-AP-6 SECTIONS Continuous Use ALL Reference Use Information Use Approved:

J. L. Hunter (for)

Operations Manager Date Issued 08/1112008 S

A F

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~ 200000 100000 Refueling Water Storage Tank Capacity Q1F16T501 Capacity (Gal) vs. Level (Ft)

Revision 3.0 February 4, 2005 JSJ Approved:

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HLT-32 ADMIN exam A.1.1SRO Page 1 of 6 A.l.lSRO Conduct of Operations ADMIN 006Al.02 - SRO TITLE: Determine required quantity of Boric Acid solution and Reactor Makeup water and integrator settings for makeup to the RWST, and determine which TS ACTIONS are required, if any.

TASK STANDARD: Determine the required quantity of Boric Acid and Reactor Makeup water to restore RWST level at the current Boric Acid Concentration, and correctly determine the setting of the Reactor Makeup system integrators and the potentiometer setting for the Boric Acid Flow controller.

Determines that LCO 3.5.4 CONDITION B is in effect, and restoring RWST to greater than 37.9 feet in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is required.

PROGRAM APPLICABLE: SOT SOCT OLT --L LOCT __

ACCEPTABLE EVALUATION METHOD: --...L. PERFORM SIMULATE DISCUSS BV ALUATION LOCATION:

SIMULATOR CONTROL ROOM ~

CLASSROOM PROJECTED TIME:

20 MIN SIMULATOR IC NUMBER:

N/A ALTERNATE PATH TIME CRITICAL PRA. __

Examinee:

Overall JPM Performance:

Satisfactory 0

UnsatisfactQIT 0 Evaluator Comments (attach additional sheets if necessary)

EXANUNER: ________________ _

HLT-32 ADMIN exam A.1.1SRO Page 2 of 6 CONDITIONS When I tell you to begin, you are to determine RWST Makeup quantity, Boric acid concentration, and integrator setting for make up to the RWST per FNP-1-S0P-2.3, Chemical And Volume Control System Reactor Makeup Control System, starting at Step 4.2.3.2 Makeup to Refueling Water Storage Tank (RWST). The conditions under which this task is to be performed are:

a. Unit 1 is at 100% power and stable.

b. RWST Level is at 37.7 feet.

c. RWST Boron concentration is at 2400 ppm.

d. On Service BAT concentration is 7001 ppm.

e. RWST Purification (Recirc) is NOT on-service.

f.

The Reactivity Spreadsheet is not available.

g. You are the extra Shift Support Supervisor and have been directed by the Shift Supervisor to perform SOP-2.3 steps 4.2.3.2 - 4.2.3.6 to:

1. Determine the quantity of blended flow required to raise level in the RWST from 37.7 feet to 39.5 feet while maintaining the current RWST Boron Concentration.

2. Determine the integrator settings for:

PIS 113, BORIC ACID BATCH INTEG FIS-168, TOTAL FLOW BATCH INTEG

3. Determine the potentiometer setting to makeup to the RWST at a reduced flow of 60 gpm total flow for:

FK-113, BORIC ACID MKUP FLOW

(

4. Determine which TS ACTION(S) is(are) required, if any.

(

EVALUATION CHECKLIST ELEMENTS:

STANDARDS:

START TIME I NOTE:

This is a classroom setting ADMIN JPM task.

1.

Determines Gallons needed per RWST Tank

Determines RWST Volume at curve 31B is 22,378 gallons.

39.5 feet=491064 gals Determines RWST Volume at 491064-468686 = 22378 37.7 feet=468686 gals Calculates total volume RESULTS:

(CIRCLE) addition=22378 gals S / U 491064 - 468686 = 22378 gals

[no tolerance: whole numbers from a table]

(

HLT-32 ADMIN exam A.1.1SRO Page 3 of 6 EVALUATION CHECKLIST RESULTS:

ELEMENTS:

STANDARDS:

(CIRCLE)

2.

3.

NOTE:

Determines Boric Acid amount from Figure Determines from Figure 1, SOP-1, SOP-2.3, for the current concentration of 2.3 ratio of Boric Acid amount to 2400 ppm.

total amount from the ratio of Boric Acid flow to Total Flow.

Then calculates total Boric Acid in gallons to obtain total 22,378 gals of blended solution at 2400 ppm:

7664-7672 gals Boric Acid Solution S I U PER FIG. 1 PAGE 1:

22378( 41.1) = 7664.465ga18 120 PER FIG 1 PAGE 3:

22378( 41.14) = 7671.924 gals 120

[tolerance: 7664.0-7672 based on using either page 1 or page 3 numbers and rounding to nearest whole number of 7664 or rounding up for conservative 7672 gals.]

Determines the totalizer settings for Total Determines totalizer settings are:

flow FIS-168, TOTAL FLOW BATCH

FIS-168=22378 gals S/U INTEG and FIS-113, BORIC ACID

FIS-I13=7664-7672 gals S I U BATCH INTEG: Based on Figure 1.

In element 4, FK-113 pot setting* is critical, but the manual position of FIS-168 demand which corresponds to 60 gpm is not critical, since this controller would need to be adjusted while flow was present. There is no corresponding demand that will ensure 60 gpm flow prior to initiating flow and adjusting as necessary.

Examiner NOTE:

In element 4, IF applicant desires to raise the setpoint above the minimum required, ask them what setpoint they are going to use and ensure it is 2reater than the minimum required.

HL T-32 ADMIN exam A.1.1 SRO EVALUATION CHECKLIST ELEMENTS:

STANDARDS:

4.

Determines the Minimum Flow controller States that FK-168 will need to be potentiometer setting for FK-113, BORIC adjusted in manual to obtain 60 ACID MKUP FLOW for the 60 gpm Total gpm.

flow directed by the Shift Supervisor at 2400 ppm Based on Figure 1, states that MINIMUM setting for FK-113 is Page 4 of 6 RESULTS:

(CIRCLE)

Pot setting of 10.29 according to Figure 1, SOP-2.3 5.145 pot setting*.

S / U would correspond to 120 gpm total flow at 2400. Since the Boric Acid Flow Controller

[Tolerance of 5.14 to 5.15 based on pot only goes to 10.0, the total flow must be the accuracy of the pot indication reduced to less than 120 gpm within the which is Y2 the smallest increment capacity of the system. The Shift Supervisor of 0.02. The procedure directs has directed 60 gpm total flow (12012=60) initiating more boric acid flow than in the initial conditions, and the boric acid necessary to ensure the boric acid flow for 120 gpm total flow must be divided flow totalizer reaches the setpoint by 2, and the pot setting must be divided by and stops Boric Acid Flow prior to

2. FK-168 will need to be adjusted in the Total Flow totalizer reaching manual to obtain 60 gpm.

its setpoint. This flushes the Boric Acid from the lines and delivers it Per Fig 1:

all to the RWST. There is no 10.2912=5.145 Boric Acid FK-113 pot setting procedural requirement to limit the boric acid flow to a specified maximum amount.]

5.*

Determines that LCO 3.5.4 CONDITION B Determines LCO 3.4.5 is in effect.

CONDITION B is in effect, and RWST level must be raised to 37.9 feet level in one hour or less.

STOP TIME

[Applicant may add that if CONDITION B is not met, CONDITION C is entered which requires MODE 3 entry in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND MODE 5 entry in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

This is correct, but not required for this ADMIN JPM]

Terminate when all elements of the task have been completed.

S / U

HLT-32 ADMIN exam A.1.1 SRO Page 5 of 6 CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.

GENERAL

REFERENCES:

1.

FNP-I-SOP-2.3 Version 48.0

2.

FNP-I-ARP-I.5 EG4, Version 50.0

3.

Tech Specs & Basis Amendment No. 146 (Unit 1), Amendment No. 137 (Unit 2)

4.

KIA:

G2.006A1.02 RO 3.0 SRO 3.6 GENERAL TOOLS AND EQUIPMENT:

Provide:

1. FNP-I-SOP-2.3, Version 48.0.

2. Curves 3IA & 3IB

3. Calculator (or applicant may supply their own calculator)

4. LCO 3.5.4 & Basis.

COMMENTS:

(

A.1.1SRO (1 Page)

HANDOUT CONDITIONS When I tell you to begin, you are to determine RWST Makeup quantity, Boric acid concentration, and integrator setting for make up to the RWST per FNP-I-SOP-2.3, Chemical And Volume Control System Reactor Makeup Control System, starting at Step 4.2.3.2 Makeup to Refueling Water Storage Tank (RWST). The conditions under which this task is to be performed are:

a. Unit 1 is at 100% power and stable.

b. RWST Level is at 37.7 feet.

c. RWST Boron concentration is at 2400 ppm.

d. On Service BAT concentration is 7001 ppm.

e. RWST Purification (Recirc) is NOT on-service.

f.

The Reactivity Spreadsheet is not available.

g. You are the extra Shift Support Supervisor and have been directed by the Shift Supervisor to perform SOP-2.3 steps 4.2.3.2 - 4.2.3.6 to:

1. Determine the quantity of blended flow required to raise level in the RWST from 37.7 feet to 39.5 feet while maintaining the current RWST Boron Concentration.

2. Determine the integrator settings for:

FIS 113, BORIC ACID BATCH INTEG FIS-168, TOTAL FLOW BATCH INTEG

3. Determine the potentiometer setting to makeup to the RWST at a reduced flow of 60 gpm total flow for:

FK-l13, BORIC ACID MKUP FLOW

4. Determine which TS ACTION(S) is(are) required, if any.

(

09/02/08 10:49: 1 0 FARLEY NUCLEAR PLANT SYSTEM OPERATING PROCEDURE FNP-I-S0P-2.3 CHEMICAL AND VOLUME CONTROL SYSTEM REACTOR MAKEUP CONTROL SYSTEM FNP-I-S0P-2.3 July 29, 2008 Version 48.0 PROCEDURE USAGE REQUIREMENTS PER FNP-0-AP-6 SECTIONS Continuous Use ALL Reference Use Information Use Approved:

J. L. Hunter (for)

Operations Manager Date Issued 08/1112008 S

A F

E T

Y R

E L

A T

E D

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600000 --'~1V;;.;;;uCR\\(31A---

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RevIsion 3.0 February 4, 200S JSJ Approved-I '.,. i/.n,,,: (-I. ~

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3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.4 Refueling Water Storage Tank (RWST)

LCO 3.5.4 The RWST shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION A.

RWSTboron A.1 Restore RWST to concentration not within OPERABLE status.

limits.

OB RWST borated water temperature not within limits.

B.

RWST inoperable for B.1 Restore RWST to reasons other than OPERABLE status.

Condition A.

C.

Required Action and C.1 Be in MODE 3.

associated Completion Time not met.

~

C.2 Be in MODES.

Farley Units 1 and 2 3.5.4-1 RWST 3.5.4 COMPLETION TIME 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1 hour 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

RWST 3.5.4 SURVEILLANCE REQUIREMENTS SR 3.5.4.1 SR 3.5.4.2 SR 3.5.4.3 SURVEILLANCE

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

Only required to be performed when ambient air temperature is < 35°F.

Verify RWST borated water temperature is ~ 35°F.

FREQUENCY 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Verify RWST borated water volume is ~ 471,000 7 days gallons.

Verify RWST boron concentration is ~ 2300 ppm and 7 days

S 2500 ppm.

Farley Units 1 and 2 3.5.4-2 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

RWST B 3.5.4 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.4 Refueling Water Storage Tank (RWST)

BASES BACKGROUND Farley Units 1 and 2 The RWST supplies borated water to the Chemical and Volume Control System (CVCS) during abnormal operating conditions, to the refueling pool during refueling, and to the ECCS and the Containment Spray System during accident conditions.

The RWST supplies both trains of the ECCS and the Containment Spray System through separate, redundant supply headers during the injection phase of a loss of coolant accident (LOCA) recovery. A motor operated isolation valve is provided in each header to isolate the RWST from the ECCS once the system has been transferred to the recirculation mode. The recirculation mode is entered when pump suction is manually transferred to the containment sump following receipt of the RWST -

Low alarm. Use of a single RWST to supply both trains of the ECCS and Containment Spray System is acceptable since the RWST is a passive component, and passive failures are not required to be assumed to occur coincidentally with Design Basis Events.

The switchover from normal operation to the injection phase of ECCS operation requires changing centrifugal charging pump suction from the CVCS volume control tank (VCT) to the RWST through the use of isolation valves. Each set of isolation valves is interlocked so that the VCT isolation valves will begin to close once the RWST isolation valves are fully open. Since the VCT is under pressure, the preferred pump suction will be from the VCT until the tank is isolated. This will result in a delay in obtaining the RWST borated water. The effects of this delay are discussed in the Applicable Safety Analyses section of these Bases.

During normal operation in MODES 1, 2, and 3, the residual heat removal (RHR) pumps are aligned to take suction from the RWST.

The ECCS and Containment Spray System pumps are provided with recirculation lines that ensure each pump can maintain minimum flow requirements when operating at or near shutoff head conditions.

When the suction for the ECCS and Containment Spray System pumps is transferred to the containment sump, the RWST flow paths must be isolated to prevent a release of the containment sump (continued)

B 3.5.4-1 Revision 0

BASES BACKGROUND (continued)

APPLICABLE SAFETY ANALYSES Farley Units 1 and 2 RWST B 3.5.4 contents to the RWST, which could result in a release of contaminants to the atmosphere and the eventual loss of suction head for the ECCS pumps.

This LCO ensures that:

a. The RWST contains sufficient borated water to support the ECCS during the injection phase;

b. Sufficient water volume exists in the containment sump to support continued operation of the ECCS and Containment Spray System pumps at the time of transfer to the recirculation mode of cooling; and

c. The reactor remains subcritical following a LOCA.

Insufficient water in the RWST could result in insufficient cooling capacity when the transfer to the recirculation mode occurs. Improper boron concentrations could result in a reduction of SDM or excessive boric acid precipitation in the core following the LOCA, as well as excessive caustic stress corrosion of mechanical components and systems inside the containment.

During accident conditions, the RWST provides a source of borated water to the ECCS and Containment Spray System pumps. As such, it provides containment cooling and depressurization, core cooling, and replacement inventory and is a source of negative reactivity for reactor shutdown (Ref. 1). The design basis transients and applicable safety analyses concerning each of these systems are discussed in the Applicable Safety Analyses section of B 3.5.2, "ECCS-Operating"; B 3.5.3, "ECCS-Shutdown"; and B 3.6.6, "Containment Spray and Cooling Systems." These analyses are used to assess changes to the RWST in order to evaluate their effects in relation to the acceptance limits in the analyses.

The RWST must also meet volume, boron concentration, and temperature requirements for non-LOCA events. The volume is not an explicit assumption in non-LOCA events since the required volume is a small fraction of the available volume. The deliverable volume limit is set by the LOCA and containment analyses. For the RWST, the deliverable volume is different from the total volume contained (continued)

B 3.5.4-2 Revision 0

BASES APPLICABLE SAFETY ANALYSES (continued)

Farley Units 1 and 2 RWST B 3.5.4 since, due to the design of the tank, more water can be contained than can be delivered. The minimum boron concentration is an explicit assumption in the main steam line break (MSLB) analysis to ensure the required shutdown capability. The minimum boron concentration limit is an important assumption in ensuring the required shutdown capability. The maximum boron concentration is an explicit assumption in the inadvertent ECCS actuation analysis, although the results are very insensitive to small changes in boron concentrations. The minimum temperature is an assumption in both the MSLB and inadvertent ECCS actuation analyses.

The MSLB analysis has considered a delay associated with the interlock between the VCT and RWST isolation valves, and the results show that the departure from nucleate boiling design basis is met.

The delay has been established as 27 seconds, with offsite power available, or 42 seconds without offsite power. This response time includes 2 seconds for electronics delay, a 10 second stroke time for the RWST valves, and a 15 second stroke time for the VCT valves.

For a large break LOCA analysis, the minimum water volume limit of 321,000 gallons and the lower boron concentration limit of 2300 ppm are used to compute the post LOCA sump boron concentration necessary to assure subcriticality. The large break LOCA is the limiting case since the safety analysis assumes that all control rods are out of the core.

A water volume of 506,600 gallons and the upper limit on boron concentration of 2500 ppm are used to determine the maximum allowable time to switch to hot leg recirculation following a LOCA.

The purpose of switching from cold leg to hot leg injection is to avoid boron precipitation in the core following the accident.

In the ECCS analysis, the containment spray temperature is assumed to be equal to the RWST lower temperature limit of 35°F. If the lower temperature limit is violated, the containment spray further reduces containment pressure, which decreases the rate at which steam can be vented out the break and increases peak clad temperature. An upper temperature assumption of 120°F is used in the small break LOCA analysis and containment OPERABILITY analysis. Exceeding this temperature would result in a higher peak clad temperature, because there would be less heat transfer from the core to the

( continued)

B 3.5.4-3 Revision 0

(

BASES APPLICABLE SAFETY ANALYSES (continued)

LCO APPLICABILITY ACTIONS Farley Units 1 and 2 RWST B 3.5.4 injected water for the small break LOCA and higher containment pressures due to reduced containment spray cooling capacity. For the containment response following an MSLB, the lower limit on boron concentration and the upper assumption on RWST water temperature are used to maximize the total energy release to containment.

The RWST satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

The RWST ensures that an adequate supply of borated water is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA), to cool and cover the core in the event of a LOCA, to maintain the reactor subcritical following a DBA, and to ensure adequate level in the containment sump to support ECCS and Containment Spray System pump operation in the recirculation mode.

To be considered OPERABLE, the RWST must meet the water volume, boron concentration, and temperature limits established in the SRs.

In MODES 1, 2, 3, and 4, RWST OPERABILITY requirements are dictated by ECCS and Containment Spray System OPERABILITY requirements. Since both the ECCS and the Containment Spray System must be OPERABLE in MODES 1, 2, 3, and 4, the RWST must also be OPERABLE to support their operation. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops -

MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level," and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level.

With RWST boron concentration or borated water temperature not within limits, they must be returned to within limits within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Under these conditions neither the ECCS nor the Containment Spray

( continued)

B 3.5.4-4 Revision 0

(

BASES ACTIONS SURVEILLANCE REQUIREMENTS Farley Units 1 and 2 A.1 (continued)

RWST B 3.5.4 System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE condition. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months limit to restore the RWST temperature or boron concentration to within limits was developed considering the time required to change either the boron concentration or temperature and the fact that the contents of the tank are still available for injection.

With the RWST inoperable for reasons other than Condition A (e.g.,

water volume), it must be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In this Condition, neither the ECCS nor the Containment Spray System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE status or to place the plant in a MODE in which the RWST is not required. The short time limit of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore the RWST to OPERABLE status is based on this condition simultaneously affecting redundant trains.

C.1 and C.2 If the RWST cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SR 3.5.4.1 The RWST borated water temperature should be verified every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to be above the minimum limit assumed in the accident analyses. This Frequency is sufficient to identify a temperature change that would approach the limit and has been shown to be acceptable through operating experience.

(continued)

B 3.5.4-5 Revision 0

BASES SURVEILLANCE REQUIREMENTS REFERENCES Farley Units 1 and 2 SR 3.5.4.1 (continued)

RWST B 3.5.4 The SR is modified by a Note that eliminates the requirement to perform this Surveillance when ambient air temperature is within the operating limit of the RWST. With ambient air temperature within the limit, the RWST temperature should not exceed the limit.

SR 3.5.4.2 The RWST water volume should be verified every 7 days to be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. Since the RWST volume is normally stable and is protected by an alarm, a 7 day Frequency is appropriate and has been shown to be acceptable through operating experience.

SR 3.5.4.3 The boron concentration of the RWST should be verified every 7 days to be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA. Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized. Since the RWST volume is normally stable, a 7 day sampling Frequency to verify boron concentration is appropriate and has been shown to be acceptable through operating experience.

1. FSAR, Chapter 6 and Chapter 15.

B 3.5.4-6 Revision 0

HL T-32 ADMIN exam A.1.2 SRO&RO Page 1 of 6 A.l.2 SRO & RO Equipment Control ADMIN OlSAl.04 - SRO & RO TITLE: Perform A Quadrant Power Tilt Ratio Calculation TASK STANDARD: Perform a QPTR calculation per STP-7.0 and identify that the current value does not meet acceptance criteria.

PROGRAM APPLICABLE: SOT SOCT OLT ~

LOCT-.-X ACCEPTABLE EVALUATION METHOD: ~

PERFORM EV ALUATION LOCATION:~CLASSROOM SIMULATE PROJECTED TIME:

20 MIN SIMULATOR IC NUMBER:

N/A ALTERNATE PATH TIME CRITICAL PRA Examinee:

Overall JPM Performance:

Satisfactory 0

Unsatisfactory Evaluator Comments (attach additional sheets if necessary)

EXAMINER: ____________________ _

DISCUSS 0

(

HL T-32 ADMIN exam A.1.2 SRO&RO CONDITIONS When I tell you to begin, you are to PERFORM A QUADRANT POWER TILT RATIO CALCULATION. The conditions under which this task is to be performed are:

a.

Reactor power is 70%.

b. N-41, N-42, & N-43 PR NI detectors are operable.

c.

N-44 PR NI detector is inoperable.

Page 2 of 6

d. You are directed by Shift Supervisor to perform STP-7.0, using DATA SHEET 2 and curves 71A-D provided, and determine if the acceptance criteria is met.

e. The IPC and QPTR computer spreadsheet are not available.

f.

A pre-job brief is not required.

EVALUATION CHECKLIST ELEMENTS:

START TIME STANDARDS:

RESULTS:

(CIRCLE)

I NOTE: Critical to use the correct 0% AFD values from curves.

1.

2.

3.

Obtain normalized currents from curves 71A, 71B, & 71C.

Record data for power range detector A and detector B from Data sheet 2.

Calculate upper and lower quadrant power tilt ratios.

4.

Enter the greater of the upper or lower quadrant power tilt ratio.

5.

Records power level.

6.

Determines acceptance criteria NOT met.

Obtains normalized current values (Curve 71) and records them on ofSTP-7.0.

Values from Data sheet 2 for detector A and detector B NI-41, 42, & 43 displays recorded on of STP-7.0.

Upper ratio calculated at 1.00 to 1.00244.

Lower ratio calculated at 1.04 to 1.043.

Greater of the above two values Lower: between 1.04 to 1.043 entered.

S / U S / U S / U S / U Current avg power level recorded.

S / U Determination made that S / U acceptance criteria was NOT met due to Lower Detector higher than acceptable and between 1.04 to 1.043.

HLT-32 ADMIN exam A.1.2 SRO&RO EVALUATION CHECKLIST ELEMENTS:

7.

8.

Reports to Shift Supervisor that acceptance criteria is NOT met.

Fills out Surveillance Test Review sheet per attached key.

STOP TIME STANDARDS:

Page 3 of 6 RESULTS:

(CIRCLE)

Reports to Shift Supervisor that S / U acceptance criteria is NOT met due to lower detector QPTR. (CUE:

Shift Supervisor acknowledges).

Fills out Surveillance Test Review S / U sheet per attached key. (CUE IF STATED THAT CR WOULD BE WRITIEN: CR # 2008001010 has been written by the Unit Operator).

Terminate when assessment of acceptance criteria is performed.

CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) preceding the element number.

(

GENERAL REFERENCES

1.

FNP-1-STP-7.0, Version 17.0

2.

Core Physics curves 71A-D Rev. 16.0

2.

KlAs: 015A1.04 RO-3.5 SRO-3.7 GENERAL TOOLS AND EQUIPMENT Provide:

1.

Calculator

2.

STP-7.0

3.

Core Physics curves 71A-D

4.

DATA SHEET 2 COMMENTS

HLT-32 ADMIN exam A.1.2 SRO&RO ATTACHMENT 1 QUADRANT POWER TILT RATIO CALCULATION CALCULATION SHEET UPPER QUADRANT POWER TILT Channel Detector Detector A A

+ Detector = Calibrated Indicated A 100%

Output

_1 Current Current Total Average N41B 104.4 155

=

0.673548 Number Upper N42B 107 159.51

=

0.670804 Operable Detector X

N42B 114.1 169.95 = 0671374 Upper Calibrated Detectors Output N44B N/A N/A

= N/A 1

Total Detector A Calibrated

= 2.015727 3

=

0.671909 X

Output

Obtamed from Curve 71,0% AFD Current LOWER QUADRANT POWER TILT

"::hannel Detector

Detector Detector B B

+B 100%

= Calibrated Indicated Current Output Current 1

N41B 107 162.46

= 0.658624 Total Average N42B 110 163.29

= 0.673648 Number Lower Operable Detector N43B 125 176.06

= 0.709985 Lower Calibrated N44B N/A

+ N/A

= N/A Detectors Output Total Detector B Calibrated Output 2.042257 3

1

=

0.680752 Page 4 of 6 Maximu Upper m

Quadrant Detector

= Power A

Tilt Calibrated Ratio Output 1.00244 0.673548 = [1.00 to 1.00244]

Lower Maximum Quadrant X

Detector B = Power Calibrated Tilt Output Ratio 1.042942 0.709985

[1.04 to

= 1.044118]

Obtained from Curve 71, 0% AFD Current Power 70%

Record~~

Upper o~wer Quadra.n.!Jfilt Ratio 1.04 (rounded from all digit calc) to 1.044118 (not rounded: 2 significant digit calc.)

[note: Tolerance determined by using only 2 significant digits in one calculation and another calculation by using all digits in calculator, rounding to 2 significant digits was used and compared to non-rounded numbers. Used max tolerance of either roundin or not roundin.]

\\.CCEPT ANCE CRITERIA: Maximum of Upper or Lower Quadrant Power Tilt Ratio does not exceed 1.02

I FNP-1-STP-7.0 SURVEILLA TITLE FARLEY NUCLEAR PLANT SURVEILLANCE TEST REVIEW SHEET TECHNICAL SPECIFICATION REFERENCE SR 3.2.4.1 MODE(S) REQUIRING TEST:

QUADRANT POWER TILT RATIO CALCULATION 1 (>50% Rated Thermal Power)

TEST RESULTS (TO BE COMPLETED BY TEST PERFORMER)

PERFORMED BY dtl"; <e4 ni: "!t."~.t'c"4L~

~;_-_~

COMPONENT OR TRAIN TESTED (if applicable) _____________________ _

[~TIRE STP PERFORMED

[] FOR SURVEILLANCE CREDIT

[] PARTIAL STP PERFORMED:

[] NOT FOR SURVEILLANCE CREDIT REASON FORPARTIAL:, ____________________________ _

TEST;PMPLETED:

~

following deficiencies occurred:

[ ] Satisfactory

[~satisfactory L..OuJe.r (LlC1clf'd..ftt Pc>W& r t ~ I + r,1c. i 0

~orrective action taken or initiated:

<... jl ~ 2 Do8a c> I 0' 0 UJ r \\ ~ ~ c.,t\\

SHIFT SUPPORT SUPERVISOR REVIEW REVIEWED BY _________________ _

DATE ____________ __

[ ] Procedure properly completed and satisfactory

[]Commen~: __________________________________ __

ENGThffiERINGSUPPORT SCREENED BY _____________ DATE, _______ _

REACTOR ENG. REVIEW (If applicable)

REVIEWEDBY _____________

D.ATE, _______ _

[ ] Satisfactory and Approved

[] Comments: ________________________________ __

Version 17.0

08/11103 13:06:50 I 1 FNP-I-STP-7.0 FARLEY NUCLEAR PLANT UNIT 1 SURVEILLANCE TEST PROCEDURE STP-7.0 QUADRANT POWER TILT RATIO CALCULATION 1.0 Purpose To detennine the quadrant power tilt ratio using power range nuclear instrumentation.

2.0 Acceptance Criteria The quadrant power tilt ratio shall be :::;; 1.02.

3.0 Initial Conditions

~3.1 3.2 Nti3.3 The version of the procedure has been verified to be the current version and correct unit for the task. (OR 1-98-498).

This STP may be performed at less than 50% power for verification of power range instrument indications. In this case, the STP is not for surveillance credit.

Above 50% of rated thermal power.

IF DVM is used to collect data, THEN have I&C obtain a Fluke 45 or equivalent with shielded test leads with NO exposed metal connectors.

DVM Serial No. _____ _

Cal. due ________ _

4.0 Precautions and Limitations 4.1 Reactor power, rod position and reactor coolant temperature should be constant while taking data.

4.2 A QPTR calculation should be done prior to rescaling of Power Range Nuclear Instruments, and after completing the rescaling of ALL Power Ranges Nuclear Instruments. A QPTR calculation performed between individual Power Range rescaling may provide erroneous results 4.3 If one Power Range NI is inoperable and thermal power is ~ 75% RTP, the remaining power range channels can be use for calculating QPTR.(SR 3.2.4.1) 4.4 Above 75% RTP, with one Power Range NI inoperable, QPTR must be determined by SR 3.2.4.2 Version 17.0

HLT-31 ADMIN exam A.1.1 RO ATTACHMENT 1 QUADRANT POWER TILT RATIO CALCULATION CALCULATION SHEET UPPER QUADRANT POWER TILT Channel Detector Detector A A

-+ Detector =

Calibrated Indicate A 100%

Output

_1 d

Current Total Average Current Number Upper N41B 104.4 155

=

0.673548 Operable Detector N42B 107 159.51

=

0.670804 Upper Calibrate X

N42B 114.1 "7"

169.95 =

0671374 Detectors d

Output N44B N/A N/A

=

N/A 1

Total Detector A Calibrated

=

2.015727 3

=

0.671909 X

Output

Obtamed from Curve 71, 0% AFD Current LOWER QUADRANT POWER TILT Channel Detector

Detector DetectorB B

-+B 100%

= Calibrated Indicated Current Output Current 1

N41B 107 162.46

= 0.658624 Total Average N42B 110

-+ 163.29

= 0.673648 Number Lower Operable Detector N43B 125 176.06

= 0.709985 Lower Calibrated N44B N/A

-+ N/A

= N/A Detectors Output Total Detector B Calibrated Output 2.042257 3

1

=

0.680752 Page 4 af6 Maximum Upper Detector Quadran A

t Calibrate

= Power d

Tilt Output Ratio 1.00244 0.673548

= [1.00 to 1.00244]

Lower Maximum Quadrant X

Detector

= Power B

Tilt Calibrated Ratio Output 1.042942 0.709985

[1.04 to

=

1.044118]

Obtained from Curve 71, 0% AFD Current Power 70%

RecordMax~.

Upper or Lo~t Tilt Ratio 1.04 (rounded from all digit calc) to 1.044118 (not rounded: 2 significant digit calc.)

[note: Tolerance determined by using only 2 significant digits in one calculation and another calculation by using all digits in calculator, rounding to 2 significant digits was used and compared to non-rounded numbers. Used max tolerance of either roundinK or not rounding.]

ACCEPTANCE CRITERIA: Maximum of Upper or Lower Quadrant Power Tilt Ratio does not exceed I 1.02

(

A.1.2 SRO & RO (2 Pages)

HANDOUT CONDITIONS When I tell you to begin, you are to PERFORM A QUADRANT POWER TILT RATIO CALCULATION.

The conditions under which this task is to be performed are:

a.

Reactor power is 70%.

b. N-41, N-42, & N-43 PR NI detectors are operable.

c. N-44 PR NI detector is inoperable.

d. You are directed by Shift Supervisor to perform STP-7.0, using DATA SHEET 2 and curves 71A-D provided, and determine if the acceptance criteria is met.

e.

The IPC and QPTR computer spreadsheet are not available.

f.

A pre-job brief is not required.

A.1.2 SRO & RO (2 Pages)

HANDOUT DATA SHEET 2 DETECTOR A INDICATED CURRENT N41 104.4 N42 107.0 N43 114.1 N44 0.0 DETECTORB INDICATED CURRENT N41 107.0 N42 110.0 N43 125.0 N44 0.0

(

PCB-I-VOLI-CRV71A UNIT 1 VOLUME 1 CURVE 71A PRESENT NIS CHANNEL N41 CURRENT SETTINGS Rev. 16 04/06/06 WRM Approved: ~~-/;;r Engineering Support Manager 4#6 ate AFD % (Values at 100% Power) 30 I

0 I

-30 Computer Channel Detector Current K

Constant Gpot N41T N41B NOTES:

l)

2)

3)

4) 179.18 I 155.00 I 130.82 129.93 I 162.46 I 195.00 84.2149 20.206

.6063 Revised Detector Equations Channel IDet=

(M)

AO +

10 N41T lDet=

.8059

AO +

155.0014 N41B lDet=

-1.0845

AO +

162.4641 nns CURVE IS FOR CYCLE 21 STARTUP CURRENTS.

CALCULATED PER FNP-O-ETP-3605 At 100% Power AFD% = AO%

T refers to the Top or Upper Detector. and B refers to the Bottom or Lower Detector I&C Procedures for N-41 Calibration are FNP-I-IMP-228.8 & FNP-I-STP-228.5 This curve is exempted from 50.59 screening per AP-I. Attachment 1. Note 1 Curve Placed in Effect:

/

Shift Supervisor Date / Time (To be completed following scaling in rack)

(

\\

Channel N42T N42B NOTES:

PCB-l-VOLI-CRV71B UNIT 1 VOLUME 1 CURVE 71B PRESENT NIS CHANNEL N42 CURRENT SETTINGS Rev. 16 04/06106 WRM Approved: ~dL,-.. -¥' 6¥c:::

'EI1gil1eeIiIlg Support Manager

. ate AFD % (Values at 100% Power) 30 I

0 I

-30 Computer Detector Current K

Constant Gpot 185.26 I 159.51 I 133.76 129.24 J 163.29 I 197.33 81.0977 19.463 0.5839 Revised Detector Equations Channel lDet=

(M)

AO +

10 N42T lDet=

.8583

AO +

159.5140 N42B lDet=

-1.1348

AO +

163.2865 THIS ClJRVE IS FOR CYCLE 21 STARTUP CURRENTS, CALCULATED PER FNP-O-ETP-3605

1) At 100% Power AFD% = AO%

2) T refers to the Top or Upper Detector, and B refers to the Bottom or Lower Detector

3) I&C Procedures for N-42 calibration are FNP-I-IMP-228.9 & FNP-I-STP-228.6

4) This curve is exempted from 50.59 screening per Ap-l, Attachment 1, Note 1 Curve Placed in Effect:

I Shift Supervisor Date / Time (To be completed following scaling in rack)

Channel N43T N43B NOTES:

PCB-1-VOL1-CRV71C UNIT 1 VOLUME 1 CURVE 71C PRESENT NIS CHANNEL N43 CURRENT SETTINGS Rev. 16 04/06/06 WRM Approved: A~.L--~$~

Engineering Support Manager Date AFD % (Values at 100% Power) 30 I

0 I

30 Computer Detector Current K

Constant 196.47 I 169.95 I 143.43 140.38 I 176.06 I 211.74 83.6349 20.07 eVls elector ~quations R

edD E

Channel IDet=

(M)

AO +

10 N43T IDet=

.8840

AO +

169.9474 N43B IDet=

-1.1893

AO +

176.0636 nns CURVE IS FOR CYCLE 21 STARTUP CURRENTS, CALCULATED PER FNP-O-ETP-3605 Gpot 0.6022

1) At 100% Power AFD% = AO%

2) T refers to the Top or Upper Detector, and B refers to the Bottom or Lower Detector

3) I&C Procedures for N-43 calibration are FNP-I-IMP-228.10 and FNP-I-STP-228. 7

4) This curve is exempted from 50.59 screening per AP-I, Attachment I, Note 1 Curve Placed in Effect:

I Shift Supervisor Date I Time (To be completed following scaling in rack)

Channel N44T N44B NOTES:

PCB-I-VOLI-CRV71D UNIT 1 VOLUME 1 CURVE 71D PRESENT NIS CHANNEL N44 CURRENT SETTINGS Rev. 16 04/06/06 WRM Approved: L3/"YL~ J;;- tiM 6 Engineering Support Manager Date AFD % (Values at 100% Power) 30 I

0 I

-30 Computer Detector Current K

Constant Gpot 189.88 I 163.76 I 137.64 131.03 I 166.86 I 202.70 80.1607 19.235 NIA ev e

or ~qua IOns R ised D teet E

f Channel JDet=

(M)

AO +

10 N44T JDet=

.8707

AO +

163.7593 N44B JDet=

-1.1945

AO +

166.8637 TInS CURVE JS FOR CYCLE 21 STARTUP CURRENTS, CALCULATED PER FNP-O-ETP-3605

1) At 100% Power AFD% = AO%

2) T refers to the Top or Upper Detector, and B refers to the Bottom or Lower Detector

3) J&C Procedures for N-44 Calibration are FNP-I-JMP-228.11 & FNP-I-STP-228.8

4) This curve is exempted from 50.59 screening per AP-l, Attachment I, Note 1 Curve Placed in Effect:

I Shift Supervisor Date I Time (To be completed following scaling in rack)

(

08/11103 13:06:50 FARLEY NUCLEAR PLANT FNP-I-STP-7.0 August 2, 2003 Version 17.0 SURVEILLANCE TEST PROCEDURE FNP-I-STP-7.0 QUADRANT POWER TILT RATIO CALCULATION PROCEDURE USAGE REQUIREMENTS PER FNP-0-AP-6 SECTIONS Continuous Use Reference Use Information Use Approved:

TODD YOUNGBLOOD Operations Manager ALL Date Issued 8-4-03 S

A F

E T

Y R

E L

A T

E D

08/11103 13:06:50 FNP-1-STP-7.0 SURVEILLANCE TEST NO.

FNP-I-STP-7.0 TITLE FARLEY NUCLEAR PLANT SURVEILLANCE TEST REVIEW SHEET TECHNICAL SPECIFICATION REFERENCE SR 3.2.4.1 MODE(S) REQUIRING TEST:

QUADRANT POWER TILT RATIO CALCULATION 1 (>50% Rated Thennal Power)

TEST RESULTS (TO BE COMPLETED BY TEST PERFORMER)

PERFORMED BY DATE/TIME COMPONENT OR TRAIN TESTED (if applicable)

[ ] ENTIRE STP PERFORMED

[] FOR SURVEILLANCE CREDIT

[ ] PARTIAL STP PERFORMED:

[] NOT FOR SURVEILLANCE CREDIT REASON FOR PARTIAL:

TEST COMPLETED:

[ ] Satisfactory

[ ] Unsatisfactory

[ ] The following deficiencies occurred:

[ ] Corrective action taken or initiated:

SHIFT SUPPORT SUPERVISOR REVIEW REVIEWED BY DATE

[ ] Procedure properly completed and satisfactory

[ ] Comments:

ENGINEERING SUPPORT SCREENED BY DATE REACTOR ENG. REVIEW (If applicable)

REVIEWED BY DATE

[ ] Satisfactory and Approved

[ ] Comments:

Version 17.0

08/11103 13:06:50 FNP-l-STP-7.0 TABLE OF CONTENTS Procedure Contains Number of Pages Body.......................................................... 2............................................. 1............................................. 1 STRS......................................................... 1

(

Page 1 of 1

08/11103 13:06:50 I

1 FNP-1-STP-7.0 FARLEY NUCLEAR PLANT UNIT 1 SURVEILLANCE TEST PROCEDURE STP-7.0 QUADRANT POWER TILT RATIO CALCULATION 1.0 Purpose To determine the quadrant power tilt ratio using power range nuclear instrumentation.

2.0 Acceptance Criteria The quadrant power tilt ratio shall be S 1.02.

3.0 Initial Conditions

~

3.1 The version of the procedure has been verified to be the current version and correct unit for the task. (OR 1-98-498).

NOTE:

This STP may be performed at less than 50% power for verification of power range instrument indications. In this case, the STP is not for surveillance credit.

~3.2

~

3.3 Above 50% of rated thermal power.

IF DVM is used to collect data, THEN have I&C obtain a Fluke 45 or equivalent with shielded test leads with NO exposed metal connectors.

DVM Serial No. ______ _ Cal. due ________ _

4.0 Precautions and Limitations 4.1 Reactor power, rod position and reactor coolant temperature should be constant while taking data.

4.2 A QPTR calculation should be done prior to rescaling of Power Range Nuclear Instruments, and after completing the rescaling of ALL Power Ranges Nuclear Instruments. A QPTR calculation performed between individual Power Range rescaling may provide erroneous results 4.3 If one Power Range NI is inoperable and thermal power is ~ 75% RTP, the remaining power range channels can be use for calculating QPTR.(SR 3.2.4.1) 4.4 Above 75% RTP, with one Power Range NI inoperable, QPTR must be determined by SR 3.2.4.2 Version 17.0

08/11103 13:06:50 I 1 FNP-I-STP-7.0 5.0 Instructions 4NOTE: QPTR may be determined using detector current meter data with normalized currents from Curve 71A, 71B, 71C, AND 71D, or by using detector currents read by DVM with normalized currents from Curve 71A, 71B, 71C, AND 71D, DVM data is obtained using Attachment 2.

5.1 Obtain normalized currents from Curve 71, and enter on the Calculation Sheet.

NOTE:

With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER S 75% RTP, the remaining three power range channels can be used for calculating QPTR.

5.2 Read detector current meters in NI-41B, 42B, 43B, and 44B POWER RANGE B drawer DETECTOR A and DETECTOR B or have I&C obtain detector currents using Attachment 2 for the desired detectors.

5.3 Enter total number of operable detectors in space provided on the Calculation Sheet.

5.4 Calculate the upper and lower Quadrant Power Tilt Ratios.

5.5 Record the greater of the upper or lower Quadrant Power Tilt Ratio value in the space provided.

ACCEPTANCE CRITERIA:

Maximum value of upper or lower Quadrant Power Tilt Ratio shall be S 1.02.

5.6 Record the Power Level (Avg.) in the space provided.

6.0 References 6.1 FSAR - Chapter 4.1.

6.2 Unit 1 Technical Specification 3.2.4 Version 17.0

08/11103 13:06:50 1

T 1 FNP-I-STP-7.0 ATTACHMENT 1 QUADRANT POWER TILT RATIO CALCULATION CALCULATION SHEET UPPER QUADRANT POWER TILT Channel Detector A

Detector A Detector A Indicated 100%

=

Calibrated Current Current Output N41B

=

1 Total Average N42B

=

Number Upper Operable Detector X

N43B

=

Upper Calibrated Detectors Output N44B

=

Total Detector A Calibrated Output 1

X

=

Obtained from Curve 71, 0% AFD Current LOWER QUADRANT POWER TaT Channel DetectorB

DetectorB DetectorB Indicated 100%

= Calibrated Current Current Output N41B

=

1 Total Average N42B

=

Number Lower Operable Detector X

N43B

=

Lower Calibrated Detectors Output N44B

=

Total Detector B Calibrated Output 1

X

=

Obtained from Curve 71, 0% AFD Current Maximum Detector A Calibrated Output Maximum DetectorB Calibrated Output Upper Quadrant

= Power Tilt Ratio

=

Lower Quadrant

= Power Tilt Ratio

=

% Power ________ _

Record Maximum of Upper or Lower Quadrant Tilt Ratio _____ _

ACCEPTANCE CRITERIA:

Maximum of Upper or Lower Quadrant Power Tilt Ratio does not exceed 1.02 Page 1 of 1 Version 17.0

08/11103 13:06:50 FNP-1-STP-7.0 ATTACHMENT 2 USING A DVM TO OBTAIN DETECTOR CURRENT VALUES ACCEPTANCE CRITERIA:

Maximum of Upper or Lower Quadrant Power Tilt Ratio shall be S 1.02.USING A DVM TO OBTAIN DETECTOR CURRENT VALUES NOTE:

Detector current values may be obtained for as many drawers as required. Unused spaces in the Table should be marked NA.

1.0 Obtaining NI Detector Currents using a DVM.

1.1 Using a Fluke 45 or equivalent (Do Not use a Fluke 8600) and shielded test leads connect and obtain detector voltage readings as follows:

INOTE:

Voltage values should be in the 2 to 3 volt range.

Al/A I&C V/A I&C 1.1.1 1.1.2 For Upper Detector connect to TP301 (+) and TP305 (-) and record voltage in appropriate space of table below.

For Lower Detector connect to TP302 (+) and TP305 (-) and record voltage in appropriate space of table below.

NOTE:

To calculate detector currents use the following formula:

Measured Detector Voltage x Curve 71" 0% AFD, 100% Current" Value = Calculated Detector Current 2.083 V/A 1.2 Using the 0% AFD, 100% current value from Curve 71, calculate the detector current value and record in appropriate space of table below.

N41 N42 N43 N44 Upper Lower Upper Lower Upper Lower Upper Lower Detector Detector Detector Detector Detector Detector Detector Detector DVM DVM DVM DVM DVM DVM DVM DVM Voltage Voltage Voltage Voltage Voltage Voltage Voltage Voltage Step 1.1 Calculated Calculated Calculated Calculated Calculated Calculated Calculated Calculated Current Current Current Current Current Current Current Current Step 1.2 Page 1 of 1 Version 17.0 I

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HLT-32 ADMIN exam A.2R Page 1 of 7 A.2R Equipment Control ADMIN 004A2.22 - RO TITLE: Perform an RCS Leakage Test TASK STANDARD: Perform an RCS Leakage Test by performing the required Surveillance Calculation, STP-9.0, and identify that the RCS leak rate does NOT meet acceptance criteria.

PROGRAM APPLICABLE: SOT SOCT OLT --L LOCT -X ACCEPTABLE EVALUATION METHOD: --L PERFORM SIMULATE DISCUSS EVALUATION LOCATION:

SIMULATOR CONTROL ROOM ~

CLASSROOM PROJECTED TIME:

20 MIN SIMULATOR~NUMBER: _____ ~N~A~_

ALTERNATE PATH TIME CRITICAL PRA Examinee:

Overall JPM Performance:

Satisfactory 0

Unsatisfactory 0 Evaluator Comments (attach additional sheets if necessary)

EXAMINER: ________________ _

(

HLT-32 ADMIN exam A.2R Page 2 of 7 CONDITIONS When I tell you to begin, you are to perform an RCS Leakage Test. The conditions under which this task is to be performed are:

a.

The unit is in Mode 1 at 100% power.

b.

You are directed by the Shift Supervisor to determine RCS leakage per STP-9.0, STEPS 5.4 to 5.9

- WHEN step 5.9 is complete, THEN another operator will complete the STP starting at step 5.10 determine whether or not RCS leakage acceptance criteria is met EVALUATION CHECKLIST ELEMENTS:

STANDARDS:

RESULTS:

(CIRCLE)

START TIME NOTE:

A Key is provided at the end of this ADMIN JPM with expected values fmed in.

The applicant must obtanin some of the values from the Handout and some from curve 28B, RCDT Rallons vs. level.

1 *.

Step 5.4 Read and record initial readings on Records initial readings on data S I U 2*.

3*.

data sheet 1.

sheet 1.

Step 5.5 Records final values on data sheet

1.

Step 5.6 Records the R-ll, R-12, and Ctmt Sump Ivl. Readings.

Records final values on data sheet

1.

Records the R-ll, R-12, and Ctmt Sump lvl. Readings.

S I U S/U I CUE IF REQUESTED: "Other leakage is 0 gpm".

NOTE:

[IF applicant inquires about any "other leakage" it is 0 gpm. This is information that is obtainable in the plant from turnover, and since there is no "other known leakage" provided, it may be assumed to be 0 gpm OR it may be requested. Either is acceptable].

Step 5.7 has already been marked NA in the handout.

4*.

Step 5.8 Calculates identified and Calculates identified and S I U unidentified leakages using the formulas on unidentified leakages using the STP-9.0 data sheet 1.

formulas on STP-9.0 data sheet 1.

HL T -32 ADMIN exam A.2R EVALUATION CHECKLIST ELEMENTS:

5*.

6*.

Step 5.9 Marks NA on step: "IF unidentified leakage is more negative than

-0.2, THEN re-perform leak rate measurement. "

Compares actual Leak rates with the acceptance criteria, and determines that the Unidentified Leakage does not meet acceptance criteria.

__ STOP TIME STANDARDS:

Marks NA on step 5.9 due to leak rate being positive.

Determines that the Identified leakage MEETS acceptance

criteria, but the Unidentified Leakage does NOT meet acceptance criteria.

Page 3 of 7 RESULTS:

(CIRCLE)

SIU S I U Terminate JPM when determination of acceptance criteria is complete for leak rate.

CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.

GENERAL REFERENCES

1.

FNP-I-STP-9.0, Version 42.0

2.

Plant tank curves 27 A, 27B, 27C, 28A, & 28B

3.

KIA: 004 A2.22 RO-3.2SRO-3.1 GENERAL TOOLS AND EQUIPMENT Provide:

1.

FNP-I-STP-9.0, Version 42.0

2.

Plant tank curves 27A, 27B, 27C, 28A, & 28B

3.

Calculator (or the Applicant may supply a calculator)

4.

Plant Conditions at 1000 & 1200 COMMENTS

HL T -32 ADMIN exam A.2R Page 4 of 7 KEY: STP-9.0, DATA SHEET 1, RCS Leakage All tolerances based on differences in rounding. One calc was performed rounding to the least significant digits at each step of the calculation, and one was performed using all digits in the calculator until the end of each step of the calculations to round to the least significant digits.

INSTRUMENT NAME INITIAL FINAL FINAL - INITIAL Computer TIME 1000 1200 A=120 Minutes (MCB)

TE0453 LIQ PRZR TEMP 650.8 OF 650.8 OF No significant change (51 OF)

(TI0453)

PC0482, PT0455, PT0456 or PT0457 PRZRPRESS 2239.4 2239.4 No significant change (PI 455, PI 456 (Note 1) psig psig (55 psig) or PI 457)

TC0484 (preferred),

OR L1T =

0 OF TY0412K, OR RCSTAVG TY0422K,OR (Note 1) 571.9 OF 571.9 OF Maximum change of O.3°F TY0432K (Note 7) allowed ifTAVG is 545°F or greater, O.I°F if TAVG is less (Average ofTI 412D, than 545°F.

422D & 432D)

RCS Temperature CF (Note 5) 99.7 N/A B =L1T x CF = 0 Gal.

Correction Factor OR NA LC 1600 OR Average of LT0459,

PRZRLVL 47.8 %

47.8 % C=56.3 x 0 %= o Gal.

LT0460 & LT0461 (LI-459, 460, 461)

LT0115 VCTLVL 51.0 %

37.3 % D = 14.18 x (-)12.7 % =

(LI115)

(-)194 to 194.3 Gal.

LI 1003 Waste Pnl or RCDTLVL 36.4 38.1 % E = 6.01 to 6.02 Gal.

BOP LS261 Pos 6 127.69 *Gal 133.71 *Gal (Enter 0 if negative)

LT0470 PRTLVL 69.7 %

69.7 % F= 0 Gal.

(LI470)

(Note 2)

NA *Gal NA *Gal (Enter 0 if negative)

FIS 168 TOTAL FLOW 3489 Gal.

3489 Gal. G = 0 Gal. Dilution and BATCH INTEG NAGal.

NAGal.

Blended Makeup

From Tank Curve Book

HLT-32 ADMIN exam A.2R Page 5 of 7 KEY (continued): STP-9,O, DATA SHEET 1, Res Leakage Total Leakage B-C-D+G (0 )-( 0 )-( {-)194 to (-)194.3)+ ( 0)

=

A

=

(

120

)

=

{+)1.617 to (+)1.62 GPM

{Note 6)

Identified Leakage E+F (6.01t06.02)+(

A

(120) o ) +

0

= 0.050 no tolerance GPM Other Leakage:

Source o

Total Other Other leakage (Note 6)

Rate (GPM) o U 'd

'f' d Leak 1.617 to 1.62 0.050 to 0.052 1.567 to 1.57 GPM m ent11e age = ----- -------=-------

Total Leakage Identified Leakage (Notes 3, 4, & 6)

ACCEPTANCE CRITERIA:

Identified Leakage ~ 10 gpm Unidentified Leakage ~ 1 gpm

A.2R (2 pages)

HANDOUT CONDITIONS When I tell you to begin, you are to perform an ReS Leakage Test. The conditions under which this task is to be performed are:

a.

The unit is in Mode 1 at 100% power.

b.

You are directed by the Shift Supervisor to determine ReS leakage per STP-9.0, STEPS 5.4 to 5.9

- WHEN step 5.9 is complete, THEN another operator will complete the STP starting at step 5.10 determine whether or not ReS leakage acceptance criteria is met

A.2R (2 pages)

HANDOUT Plant Conditions at 1000:

INSTRUMENT NAME Computer Points N/A TIME 1000 TE0453 LIQ PRZR TEMP 650.8 of PC0482 PRZR PRESS 2239.4 psig TC0484 RCS TAVG 571.9 of LC 1600 PRZR LVL 47.8 LTOl15 VCT LVL 51.0 BOP LS261 Pos 6 RCDT LVL 36.4

Gal LT0470 PRT LVL 69.7

Gal FIS 168 TOTAL FLOW 3489 Gal.

BATCH INTEG Plant Conditions at 1200:

INSTRUMENT NAME Computer Points N/A TIME 1200 TE0453 LIQ PRZR TEMP 650.8 of PC0482 PRZR PRESS 2239.4 Psig TC0484 RCS TAVG 571.9 of LC 1600 PRZR LVL 47.8 LTOl15 VCT LVL 37.3 BOP LS261 Pos 6 RCDT LVL 38.1

Gal LT0470 PRT LVL 69.7

Gal FIS 168 TOTAL FLOW 3489 Gal.

BATCH INTEG CTMT R-11 Particulate 187 CPM Rad Monitor R-12 CTMT Gas Rad 75 CPM Monitor Q1G21LI3282A Ctmt. Sump lvl 14 Inches Q1G21LI3282B Ctmt. Sump lvl 14 Inches

08/08/08 09:39:04 T 1 FARLEY NUCLEAR PLANT SURVEILLANCE TEST PROCEDURE FNP-I-STP-9.0 RCS LEAKAGE TEST FNP-I-STP-9.0 May 30, 2008 Version 42.0 PROCEDURE USAGE REQUIREMENTS PER FNP-0-AP-6 SECTIONS Continuous Use ALL Reference Use Information Use Approved:

J. L. Hunter (for)

Operations Manager Date Issued:

05/30/2008 S

A F

E T

Y R

E L

A T

E D

08/08/0809:39:04 1

1 FNP-l-STP-9.0 FA EY CLE PLANT SURVEILLANCE TEST REVIEW SHEET SURVEILLANCE TEST NO.

TECHNICAL SPECIFICATION REFERENCE FNP-I-STP-9.0 SR3.4.13.1 TITLE MODE(S) REQUIRING TEST:

RCS LEAKAGE TEST 1,2,3,4 TEST RESULTS (TO BE COMPLETED BY TEST PERFORMER)

PERFORMED BY DATEITIME COMPONENT OR TRAIN TESTED (if applicable)

[] ENTIRE STPPERFORMED

[ ] FOR SURVEILLANCE CREDIT

[] PARTIAL STP PERFORMED:

[ ] NOT FOR SURVEILLANCE CREDIT REASON FOR PARTIAL:

TEST COMPLETED:

[ ] Satisfactory

[ ] Unsatisfactory

[ ] The following deficiencies occurred:

[ ] Corrective action taken or initiated:

SHIFT SUPERVISOR! SHIFT SUPPORT SUPERVISOR REVIEW REVIEWED BY DATE

[ ] Procedure properly completed and satisfactory

[] Comments:

ENGINEERING SUPPORT SCREENED BY DATE GROUP SCREENING REVIEWED BY DATE (IF APPLICABLE)

[ ] Satisfactory and Approved

[ ] Comments:

Version 42.0

08/08/08 09:39:04 1

T 1 FNP-I-STP-9.0 TABLE OF CONTENTS Procedure Contains Number of Pages STRS......................................................... 1 Body.......................................................... 6 Data Sheet 1.............................................. 3 Page 1 of 1 Version 42.0

08/08/08 09:39:04 T 1 FNP-1-STP-9.0 1.0 Purpose FARLEY NUCLEAR PLANT UNIT 1 SURVEILLANCE TEST PROCEDURE STP-9.0 RCS LEAKAGE TEST To determine identified and unidentified reactor coolant system leakage by performance of an RCS water inventory balance.

NOTE:

Asterisked steps (*) are those associated with Acceptance Criteria.

2.0 Acceptance Criteria 2.1 Unidentified leakage is $; 1 gpm.

2.2 Identified leakage $; 10 gpm.

NOTE:

FNP-I-STP-9.0 RCS Leakage Test (SR 3.4.13.1) is only required to be performed during steady state operation. AI 2004201338 3.0 Initial Conditions

(.. ~

3.1 The version of this procedure has been verified to be the current version.

(OR 1-98-498)

(.. ~

3.2 This procedure has been verified to be the correct unit for the task.

(OR 1-98-498)

(.. ~

3.3 Reactor power and reactor coolant temperature should be stabilized and held approximately constant for I hour prior to and during the test. (In Mode 3 or 4 not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation.)

(.. ~

3.4 The pressurizer level and pressure control systems are in automatic or are in manual control and are stable.

(.. ~

3.5 The level of the VCT is in the normal operating band high enough to prevent the occurrence of an Auto Makeup during the test. Version 42.0

08/08/08 09:39:04 FNP-I-STP-9.0 c...~3.6 c...~3.7 to: 3.8 The CVCS system is aligned per FNP-I-SOP-2.1A, CHEMICAL & VOLUME CONTROL SYSTEM.

Notify the Shift Chemist and Shift Radiochemist of the performance of the test to ensure that no sampling of the RCS or CVCS will be done during this test.

IF required for step 5.2, THEN ensure the following instrument is in calibration.

Calibrated Digital Voltmeter FNP I.D. #

Cal Due Date Version 42.0

(

08/08/08 09:39:04 1

I 1

FNP-I-STP-9.0 4.0 Precautions And Limitations 4.1 No sampling of the RCS or CVCS shall be done during this test.

4.2 Any of the following will render this test void:

4.2.1 Emergency boration 4.2.2 Diversion ofletdown to the recycle holdup tanks.

4.2.3 Make up from any source which does not go through the boric acid blender.

4.2.4 Boration ofless than 10 gpm, due to Batch Integrator counter inaccuracies.

4.3 To minimize the inaccuracy introduced into the calculation by RCS temperature changes, RCS temperature should be maintained as follows:

4.3.1 IF RCS temp is < 545 of, THEN the RCS temperature should not change by more than 0.1 of during the test.

4.3.2 IF RCS temp is 2: 545 of, THEN the RCS temperature should not change by more than 0.3 of during the test.

4.3.3 IF required to maintain RCS temperature, THEN control rods, turbine load or boron concentration should be adjusted as necessary.

4.4 The calculation assumes that changes in RCS volume due to PZR temperature 1 pressure fluctuations are negligible. Pressurizer parameters should be maintained stable to minimize inaccuracy.

4.5 The following guidelines should be followed to maximize precision:

IF available, THEN computer points should be used for obtaining data.

Otherwise, the available indications are to be read as accurately as possible.

For RCS Tavg, the computer point data should be entered to include three decimal places (i.e., 572.204 OF).

For other computer points and RCDT level, the data should be entered to include at least one decimal place (i.e., 50.1 %).

Identified and unidentified leakage rates are to be reported in two decimal places (e.g., 0.07 gpm).

IF possible, THEN normal makeup to the VCT should be avoided.

4.6 IF the RCDT or PRT level indication is invalid, THEN use 0 gpm for RCDT or PRT portion of identified leakage unless leakage into the RCDT or PRT is to be determined using another approved method.

4.7 To ensure that the STP-9.0 Computer Program remains current, the Engineering Support Group should be notified of any revision or TCN to the Data Sheet 1. Version 42.0

08/08/08 09:39:04 1

FNP-1-STP-9.0 5.0 Instructions 5.1 NOTE:

I&C N,f\\ 5.2 CV L~

5.3 NOTE:

5.4 NOTE:

5.5 5.6 The RCDT system is aligned as follows:

5.1.1 RCDT level is in the normal operating band.

5.1.2 Close RCDT PUMPS DISCH LINE ISO QIG21HV7136 The following step is only required if increased accuracy is necessary for determination of leak rate into PRT or the MCB PRT level indicator has a problem.

IF required, THEN have I&C connect a calibrated digital voltmeter across the output ofLQY-470, location C5-231.

Place VCT HI LVL DIVERT VLV, QIE21LCVl15A, in the VCT position.

Batch Integrator readings will be taken prior to and at the conclusion of each make up evolution.

Read and record initial readings on data sheet 1.

A time span of at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months should be used during normal steady state plant operations, however if plant conditions dictate, a shorter time span may be used. (30 minutes minimum).

After the desired time span (normally 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) record final values on data sheet 1.

Record the following readings (recorded for trending purposes):

Rad Monitor R-ll CPM Rad Monitor R-12 CPM Ctmt. Sump Iv!. Q1G21LI3282A Inches Ctmt. Sump Iv!. Q1G21LI3282B Inches Version 42.0

l

\\,

08/08/08 09:39:04 FNP-I-STP-9.0 NA 5.7 IF the RCS leakrate program is to be used, THEN verify that the program is revision 3.

NOTE:

If the RCS leakrate program is used, then the remainder of data sheet 1 may be left blank.

5.8 Calculate identified and unidentified leakages using the RCS leakrate program or formulas on data sheet 1.

ACCEPTANCE CRITERIA:

Identified Leakage ~ 10 gpm

_1-IV

_1-IV Unidentified Leakage ~ 1 gpm 5.9 IF unidentified leakage is more negative than -0.2, THEN re-perform leak rate measurement.

5.10 5.11 5.12 Open RCDT PUMPS DISCH LINE ISO QIG21HV7136.

Place VCT HI LVL DIVERT VLV, QIE2lLCV115A in the AUTO position.

IF computer point LC0500 is available, THEN review the RCS leakrate trend (last 30 days if possible) on IPC to determine if any abnormal trends exist.

5.13 IF unidentified leakage is >0.15 gpm, THEN re-perform leak rate measurement to confirm the results.

5.14 IF unidentified leakage is confirmed to be >0.15 gpm, THEN perform the following: (steps may be performed in any order) 5.14.1 Perform inspection to identify the leakage path(s) (AOP-l.O, attachments 2 through 5).

5.14.2 Perform evaluation including any recent maintenance, plant evolutions or filter alignments to locate source ofleakage, determine corrective actions and the effects of the leakage.

5.14.3 IF leakage is NOT known to be outside CTMT, THEN request chemistry sample CTMT via R-67 for iron analysis.

5.14.4 Submit CR to document the leakage and actions taken. Version 42.0

08/08/08 09:39:04 1

I FNP-1-STP-9.0 I&C

_1_ 5.15 IV IF applicable, THEN have I&C remove the calibrated digital voltmeter installed in step 5.2.

5.16 Update OPS home page (ULR Data spreadsheet) with unidentified leakage rate.

5.17 IF used for RCS leakrate calculation, THEN attach the computer generated Data Sheet 1 to this procedure.

6.0 References 6.1 P&ID D-175037 - RCS, sheet 2 6.2 P&ID D-175039 - CVCS, sheet 2 6.3 P&ID D-175042 - Waste Processing System, sheet 1 Version 42.0

08/08/08 09:39:04 INSTRUMENT NAME Computer TIME (MCB)

TE0453 LIQ PRZR TEMP (TI0453)

PC0482, PT0455, PT0456 or PT0457 PRZRPRESS (PI 455, PI 456 (Note 1) or PI 457)

TC0484 (preferred),

OR TY0412K, OR TY0422K, OR RCSTAVG TY0432K (Note 7)

(Note 1)

(Average ofTI 412D, 422D &

432D)

RCS Temperature CF (Note 5)

Correction Factor LC 1600 OR Average ofLT0459, PRZRLVL L T0460 & LT0461 (LI-459, 460, 461)

LT0115 VCTLVL (LI 115)

LI 1003 Waste Pnl or RCDTLVL BOP LS261 Pos 6 LT0470 PRTLVL (LI 470)

(Note 2)

FIS 168 TOTAL FLOW BATCH INTEG

From Tank Curve Book DATA SHEET 1 RCS Leakage INITIAL of psig of

Gal

Gal Gal.

Gal.

Page 1 of3 FINAL of psig of N/A

Gal

Gal Gal.

Gal.

A=

FNP-1-STP-9.0 DATA SHEET 1 FINAL - INITIAL Minutes No significant change (::: 1°F)

No significant change

(:::5 psig)

~T =

of Maximum change of O.3°F allowed ifTAVG is 545°F or greater, 0.1 of if TAVG is less than 545°F.

B =~T x CF =

Gal.

C = 56.3 x

%=

Gal.

D = 14.18 x

%=

Gal.

E=

Gal.

(Enter 0 if negative)

F=

Gal.

JEnter 0 if neAative)

G=

Gal. Dilution and Blended Makeup Version 42.0

(

08/08/08 09:39:04 Total Leakage 1 I FNP-I-STP-9.0 DATA SHEET 1

= B - C - D + G = -,-( _---'--)_-(-0..----;---'-)_--'-( ~----,,-)-+....o-( _---"-) - ---GPM A

()

(Note 6)

Identified Leakage

=_E+_F =~( __ ~)_+~(~ __ ~)+

=-------GPM A

(

)

Other leakage (Note 6)

Other Leakage:

Source Rate (GPM)

Total Other Unidentified Leakage =

=

GPM Total Leakage Identified Leakage (Notes 3, 4, & 6)

ACCEPTANCE CRITERIA:

Identified Leakage S 10 gpm Unidentified Leakage S 1 gpm Page 2 of3 Version 42.0

(

08/08/08 09:39:04 NOTES:

T 1 FNP-I-STP-9.0 DATA SHEET 1 1 IF TAVG < S30°F, THEN use: PI-402A (PT0402) and PI-403A (PT0403), lC and lA Loop RCS WR PRESS (Avg. of Readings)

AND TR-410 (TE04l0) and TR-413 (TE0413), RCS COLD AND HOT LEG TEMP (Avg. of Readings) 2 Calibrated fluke may be used for PRT level determination if deemed necessary.

3 For reporting purposes values between -0.2 and 0 gpm shall be reported as 0 gpm. Values more negative than -0.2 gpm indicate a potential problem and therefore shall be reported as is.

4 If unidentified leakage> 0.9 but < 1 gpm, test should be reperformed with ZAS secured. At maximum injection rate, ZAS can introduce -8.03 gpm error into calculation.

S Obtain CF from Table 1 using the nearest value of RCS temperature. N/A if RCS Leakrate program is used.

6 Leakage calculations are to be reported in two decimal places (e.g.,

0.07 gpm).

7 TC0484 is preferred for ReS Tavg, but an individual loop temperature may be used if desired due to instability in the average reading.

TABLE 1 RCSTemp Page 3 of3 Version 42.0

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VOLUME II CURVE 27A PRESSURIZER RELIEF TANK CAPACITY CAPACITY (GAL} VS LEVEL (FEET)

REV. 1 March 8,1978 GAF APPROVED:

R 0>

<:::,)

7t)(

"p

~

r-c

~

K-E'. 1; L';f""}',,NT!~rr~~:'" " eM t

471510 Pef>lvoLl eeV.116

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VOLUME II CURVE 27C PRESSURIZER RELIEF TANK CAPACITY TABLE N1B32T001 CAPACITY (GAL) VS % LEVEL REV. 0 May 7, 1980 GAF

('

APPROVED:

Ie/, VI) t do." (.J.V').? c

~~~

tA). t-4 c UM~~

~hdto TECHNICAL SUPERINTENDENT D./{TE

% LEVEL GALLONS

~,~ LEVEL GALLONS

-,,' LEVEL GALLONS

" LEVEL GALLm~S 0.0 222.12 26.0 2272.2r;i 52.0 5086.:38 7'8.0 7842.44 1.0 27'2.18 27'.0 237'2.49 53.8 5198.21 7'" I 7'936.8:3 2.0 325. :39 28.0 2473.7'3 54.8 5:309. '32 80.8 8029. '~.)

3.0 381. 81 29.0 2575.91 55.0 5421.45 81.0 8121.58 4.0 441.16 30.0 2678.98 56.0 5532.7',

82.0 8211.8(1 5.0 503.28 31.0 2782. ~~2 57.0 564:3.89 83.8

33121121.53 6.0 568.105

32. I 2887.710 58.0 5754.6:3 84.0 8:387.71 7'.0 635.34 33.0 2'3'33.24 59.8 5865.14 85.0

3473.2:3

8. (1 705.02 34.6

3099.49 60.0 5975.22 86.0 8557.106 9.0 776. '39 35.0 3206.4:3

61. 0 6084. 9(~

87.8 863-3" 11 10.0 851*.16 36.0 3314.00 62.0 61'34. 11 88.0 8719.2'3 11.0 927.43 37.0 3422.18 63.0 6:302.82 89.0

3797.57 12.0 1005.71 38.0 35:30.89 64.0 6411. 60 96.0 8873. :34 13.0 1085.89 39.0 3640.10 65.0 6518.57' 91.0 8948.01 14.0 1167.94 40.6

37'49.78 66.0 6625.51 92.6 9019.98 15.0 1251.77 41.0 3859.86

67. £1 6731. 7'6 9:3.0 9089.66 16.0 1337'.29 42.0

1970.: 32 68.0 6837'. :30 94.0 9156.95 17.0 1424.47' 43.0 4081.11 69.0 6',42.08 95.13 9221.72 18.0 1513.20 44.0 4192.21 70.0 7'046. (12 96.0 9283.84 19.(1 1603.42 45.6 4303.55 7' 1. ~J 714',.09 97.0 934:3.19 20.0 16',5.10 46.0 4415.0:3 72.0 7251_27

~38. £1 93'31'3~61 21.0 1788.17 47.0 4526.79 7:3.0 7':352.51

',9.13 9452.. 98

22. 1 1882.56 48.0 4638.62 74.0 7'452.71 1 (10. (1

'3502. ~:n3 23.0 197'8.22 4'~. 0 4750.54 75.0 7551. 86 24.0 2075.11 50.0 4862.56 76.0

?64 c3.8*3 25.0 217'3.14 51.0 4974.46 77'.0 7746.7:3

VOLUME II CURVE 28A REACTOR COOLANT DRAIN TANK CAPACITY

,+~,",' NIG21TOOl CAPACITY (GAL) VS LEVEL (FEET)

REV 2 June 11, 1986 JMR APPROVED:

L.().~

GALLONS rmIGHT 0.00 0.25 0.50 0.75 1.00 1.25 1.47 1.72 1.97 2.23 2.47 2.72 VOUJME ft' 0,0 1.8 4.8 8.6 12.9 17.4 21.4 26.0 30.4 34.6 38.3 41.3 GALLONS 0.0 13.7 38.2 69.0 103.9 141.5 175.1 213.9 251.2 285.7 315.8 339.0

Unit 1 Volume II Curve 28B Reactor Coolant Drain Tank Capacity pc8-I-vol..:"J.-C<<Y~813 NlG21TOOl Capacity (Gallons) vs % Level Rev.

2~ December 14, 1981, C.A.P.

Approved:

) -ft'2-a~

Technical Superintendent Date

~; LE"/EL 1" (1 4.U

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HLT-32 ADMIN exam A.2S Page 1 of 8 A.2S Equipment Control ADMIN 004A2.22 - SRO TITLE: Perform an RCS Leakage Test & determine Tech Spec requirements if applicable.

TASK STANDARD: Perform an RCS Leakage Test by performing the required Surveillance Calculation, STP-9.0, and identify that the RCS leak rate does NOT meet acceptance criteria, then determine proper Tech Spec TS 3.4.13 CONDITION B for Pressure boundary Leakage.

PROGRAM APPLICABLE: SOT SOCT OLT --.1L LOCT ---X ACCEPTABLE EVALUATION METHOD:

X PERFORM SIMULATE DISCUSS EVALUATION LOCATION:

SIMULATOR CONTROL ROOM ~

CLASSROOM PROJECTED TIME:

25 MIN SIMULATOR IC NUMBER:

NA ALTERNATE PATH TIME CRITICAL PRA Examinee:

Overall JPM Performance:

Satisfactory 0

Unsatisfactory 0 Evaluator Comments (attach additional sheets if necessary)

EXANUNER: ________________ _

HLT-32 ADMIN exam A.2S Page 2 of 8 CONDITIONS When I tell you to begin, you are to perform an RCS Leakage Test and identify if any Tech Spec ACTIONS are applicable, and if so, which ACTION(S). The conditions under which this task is to be performed are:

a.

The unit is in Mode 1 at 100% power.

b.

You are directed by the Shift Supervisor to determine RCS leakage per STP-9.0, STEPS 5.4 to 5.9

- WHEN step 5.9 is complete, THEN another operator will complete the STP starting at step 5.10 determine whether or not RCS leakage acceptance criteria is met IF acceptance criteria is NOT met, THEN determine proper Tech Spec ACTION(S).

EVALUATION CHECKLIST ELEMENTS:

STANDARDS:

RESULTS:

(CIRCLE)

__ START TIME NOTE:

A Key is provided at the end of this ADMIN JPM with expected values t1Iled in.

The applicant must obtanin some of the values from the Handout and some from curve 28B, RCDT gallons vs. level.

1 *.

Step 5.4 Read and record initial readings on Records initial readings on data S / U 2*.

3*.

data sheet 1.

sheet 1.

Step 5.5 Records fmal values on data sheet

1.

Step 5.6 Records the R-ll, R-12, and Ctmt Sump lvl. Readings.

Records fmal values on data sheet

1.

Records the R -11, R -12, and Ctmt Sump lvl. Readings.

S / U S / U I CUE IF REQUESTED: "Other leakage is 0 gpm".

NOTE:

[IF applicant inquires about any "other leakage" it is 0 gpm. This is information that is obtainable in the plant from turnover, and since there is no "other known leakage" provided, it may be assumed to be 0 gpm OR it may be requested. Either is acceptable].

Step 5.7 has already been marked NA in the handout.

4*.

Step 5.8 Calculates identified and Calculates identified and S / U unidentified leakages using the formulas on unidentified leakages using the STP-9.0 data sheet 1.

formulas on STP-9.0 data sheet 1.

HLT-32 ADMIN exam A.2S EVALUATION CHECKLIST ELEMENTS:

5*.

6*.

Step 5.9 Marks NA on step: "IF unidentified leakage is more negative than

-0.2, THEN re-perform leak rate measurement."

Compares actual Leak rates with the acceptance criteria, and determines that the Unidentified Leakage does not meet acceptance criteria.

STANDARDS:

Marks NA on step 5.9 due to leak rate being positive.

Determines that the Identified leakage MEETS acceptance

criteria, but the Unidentified Leakage does NOT meet acceptance criteria.

Page 3 of 8 RESULTS:

(CIRCLE)

S I U S I U 7*.

~~~~~~-~-+--~~--~+-~~;

Determines that Tech Spec 3.4.13 1, /,1Jt.

STOP TIME in NOT met, S I U IVI l' AND CONDITION B is in effect for "Pressure boundary LEAKAGE exists".

Be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND BE MODE 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is required.

S I U S/U Terminate JPM when determination of applicable Tech Spec ACTION is complete for leak rate.

CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) before the element number.

HL T-32 ADMIN exam A.2S GENERAL REFERENCES

1.

FNP-1-STP-9.0, Version 42.0

2.

Plant tank curves 27 A, 27B, 27C, 28A, & 28B

2.

KIA:

004 A2.22 RO-3.2SRO-3.1

3.

Tech Spec 3.4.13 AND Basis GENERAL TOOLS AND EOUIPMENT Provide:

1.

2.

3.

4.

5.

COMMENTS FNP-1-STP-9.0, Version 42.0 Plant tank curves 27A, 27B, 27C, 28A, & 28B Calculator (or the Applicant may supply a calculator)

Plant Conditions at 1000 & 1200 Tech Spec 3.4.13 AND Basis Page 4 of 8

(

HLT-32 ADMIN exam A.2S Page 5 of 8 KEY: STP-9.0, DATA SHEET 1, RCS Leakage All tolerances based on differences in rounding. One calc was performed rounding to the least significant digits at each step of the calculation, and one was performed using all digits in the calculator until the end of each step of the calculations to round to the least significant digits.

INSTRUMENT NAME INITIAL FINAL FINAL - INITIAL Computer TIME 1000 1200 A=120 Minutes (MCB)

TE0453 LIQ PRZR TEMP 650.8 OF 650.8 OF No significant change (:::; 1°F)

(TI0453)

PC0482, PT0455, PT0456 or PT0457 PRZRPRESS 2239.4 2239.4 No significant change (PI 455, PI 456 (Note 1) psig psig

(:::;5 psig) or PI 457)

TC0484 (preferred),

OR i1T =

0 OF TY0412K, OR RCSTAVG TY0422K, OR (Note 1) 571.9 OF 571.9 OF Maximum change of O.3°F TY0432K (Note 7) allowed ifTAVG is 545°F or greater, 0.1 OF if TAVG is less (Average ofTI 412D, than 545°F.

422D & 432D)

RCS Temperature CF (Note 5) 99.7 N/A B = i1T x CF = 0 Gal.

Correction Factor OR NA LC 1600 OR Average of LT0459, PRZRLVL 47.8 %

47.8 % C = 56.3 x 0 % = o Gal.

LT0460 & LT0461 (LI-459, 460, 461)

LT0115 VCTLVL 51.0 %

37.3 % D = 14.18 x (-)12.7 % =

(LI 115)

(-)194 to 194.3 Gal.

LI 1003 Waste Pnl or RCDTLVL 36.4 38.1 % E = 6.01 to 6.02 Gal.

BOP LS261 Pos 6 127.69 *Gal 133.71 *Gal (Enter 0 if negative)

LT0470 PRTLVL 69.7 %

69.7 % F= 0 Gal.

(LI470)

(Note 2)

NA *Gal NA *Gal (Enter 0 if negative)

FIS 168 TOTAL FLOW 3489 Gal.

3489 Gal. G = 0 Gal. Dilution and BATCH INTEG NA Gal.

NA Gal.

Blended Makeup

From Tank Curve Book

HLT-32 ADMIN exam A.2S Page 6 of 8 KEY (continued): STP-9,O, DATA SHEET 1, Res Leakage Total Leakage

_ 8-C-D+G _ ( 0 )-( 0 )-( (-)194 to (-)194.3 )+( 0 )_

A

(

120

)

(+)1.617 to (+)1.62 GPM (Note 6)

Identified Leakage E + F (6.01 to 6.02 ) + (

A

(120) o ) +

0

= 0.050 no tolerance GPM Other leakage (Note 6)

Other Leakage:

Source o

Total Other Rate (GPM) o U 'd if'ed Leak 1.617 to 1.62 0.050 to 0.0502 ill ent 1 age =

Total Leakage Identified Leakage 1.567 to 1.57

GPM (Notes 3, 4, & 6)

ACCEPTANCE CRITERIA:

Identified Leakage :s: 10 gpm Unidentified Leakage :s: 1 gpm

A.2S (2 pages)

HANDOUT CONDITIONS When I tell you to begin, you are to perform an RCS Leakage Test and identify if any Tech Spec ACTIONS are applicable, and if so, which ACTION(S). The conditions under which this task is to be performed are:

a.

The unit is in Mode 1 at 100% power.

b.

You are directed by the Shift Supervisor to determine RCS leakage per STP-9.0, STEPS 5.4 to 5.9

- WHEN step 5.9 is complete, THEN another operator will complete the STP starting at step 5.10 determine whether or not RCS leakage acceptance criteria is met IF acceptance criteria is NOT met, THEN determine proper Tech Spec ACTION(S).

A.2S (2 pages)

HANDOUT Plant Conditions at 1000:

INSTRUMENT NAME Computer Points N/A TIME 1000 TE0453 LIQ PRZR TEMP 650.8 of PC0482 PRZR PRESS 2239.4 psig TC0484 RCS TAVG 571.9 of LC 1600 PRZR LVL 47.8 LTOl15 VCT LVL 51.0 BOP LS261 Pos 6 RCDT LVL 36.4

Gal LT0470 PRT LVL 69.7

Gal FIS 168 TOTAL FLOW 3489 Gal.

BATCH INTEG Plant Conditions at 1200:

INSTRUMENT NAME Computer Points N/A TIME 1200 TE0453 LIQ PRZR TEMP 650.8 of PC0482 PRZR PRESS 2239.4 Psig

(

TC0484 RCS TAVG 571.9 of LC 1600 PRZR LVL 47.8 LTOl15 VCT LVL 37.3 BOP LS261 Pos 6 RCDT LVL 38.1

Gal LT0470 PRT LVL 69.7

Gal FIS 168 TOTAL FLOW 3489 Gal.

BATCH INTEG CTMT R-11 Particulate 187 CPM Rad Monitor R-12 CTMT Gas Rad 75 CPM Monitor Q1G21LI3282A Ctmt. Sump lvl 14 Inches Q1G21LI3282B Ctmt. Sump lvl 14 Inches

08/08/08 09:39:04 T 1 FARLEY NUCLEAR PLANT SURVE~LANCETESTPROCEDURE FNP-I-STP-9.0 RCS LEAKAGE TEST FNP-I-STP-9.0 May 30, 2008 Version 42.0 PROCEDURE USAGE REQUIREMENTS PER FNP-0-AP-6 SECTIONS Continuous Use ALL Reference Use Information Use Approved:

J. L. Hunter (for)

Operations Manager Date Issued:

05/30/2008

-......;..;.,;",,;;.~~--

S A

F E

T Y

R E

L A

T E

D

08/08/08 09:39:04 1

T 1 FNP-l-STP-9.0 FARL CLEAR PLANT SURVEILLANCE TEST REVIEW SHEET SURVEILLANCE TEST NO.

TECHNICAL SPECIFICA nON REFERENCE FNP-1-STP-9.0 SR 3.4.13.1 TITLE MODE(S) REQUIRING TEST:

RCS LEAKAGE TEST 1,2,3,4 TEST RESULTS (TO BE COMPLETED BY TEST PERFORMER)

PERFORMED BY DATEITIME COMPONENT OR TRAIN TESTED (if applicable)

[ ] ENTIRE STP PERFORMED

[ ] FOR SURVEILLANCE CREDIT

[] PARTIAL STP PERFORMED:

[ ] NOT FOR SURVEILLANCE CREDIT REASON FOR PARTIAL:

TEST COMPLETED:

[ ] Satisfactory

[ ] Unsatisfactory

[ ] The following deficiencies occurred:

[ ] Corrective action taken or initiated:

SHIFT SUPERVISOR! SHIFT SUPPORT SUPERVISOR REVIEW REVIEWED BY DATE

[ ] Procedure properly completed and satisfactory

[ ] Comments:

ENGINEERING SUPPORT SCREENED BY DATE GROUP SCREENING REVIEWED BY DATE (IF APPLICABLE)

[ ] Satisfactory and Approved

[ ] Comments:

Version 42.0

08/08/08 09:39:04 1

I 1

FNP-1-STP-9.0 TABLE OF CONTENTS Procedure Contains Number of Pages STRS......................................................... 1 Body.......................................................... 6 Data Sheet 1.............................................. 3 Page 10f1 Version 42.0

08/08/08 09:39:04 T 1 FNP-I-STP-9.0 1.0 Purpose FARLEY NUCLEAR PLANT UNIT 1 SURVEILLANCE TEST PROCEDURE STP-9.0 RCS LEAKAGE TEST To determine identified and unidentified reactor coolant system leakage by performance of an RCS water inventory balance.

NOTE:

Asterisked steps (*) are those associated with Acceptance Criteria.

2.0 Acceptance Criteria 2.1 Unidentified leakage is ~ 1 gpm.

2.2 Identified leakage ~ 10 gpm.

NOTE:

FNP-I-STP-9.0 RCS Leakage Test (SR 3.4.13.1) is only required to be performed during steady state operation. AI 2004201338 3.0 Initial Conditions C. ~

3.1 The version of this procedure has been verified to be the current version.

(OR 1-98-498)

C. ~

3.2 This procedure has been verified to be the correct unit for the task.

(OR 1-98-498)

C. ~

3.3 Reactor power and reactor coolant temperature should be stabilized and held approximately constant for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to and during the test. (In Mode 3 or 4 not required unti112 hours of steady state operation.)

C. ~

3.4 The pressurizer level and pressure control systems are in automatic or are in manual control and are stable.

C. ~

3.5 The level of the VCT is in the normal operating band high enough to prevent the occurrence of an Auto Makeup during the test. Version 42.0

08/08/08 09:39:04 1

I 1

FNP-I-STP-9.0 L~3.6 L~3.7 t{,k 3.8 The CVCS system is aligned per FNP-I-SOP-2.1A, CHEMICAL & VOLUME CONTROL SYSTEM.

Notify the Shift Chemist and Shift Radiochemist of the performance of the test to ensure that no sampling of the RCS or CVCS will be done during this test.

IF required for step 5.2, THEN ensure the following instrument is in calibration.

Calibrated Digital Voltmeter FNP J.D. #

Cal Due Date Version 42.0

08/08/08 09:39:04 1 I FNP-1-STP-9.0 4.0 Precautions And Limitations 4.1 No sampling of the RCS or CVCS shall be done during this test.

4.2 Any of the following will render this test void:

4.2.1 Emergency boration 4.2.2 Diversion ofletdown to the recycle holdup tanks.

4.2.3 Make up from any source which does not go through the boric acid blender.

4.2.4 Boration ofless than 10 gpm, due to Batch Integrator counter inaccuracies.

4.3 To minimize the inaccuracy introduced into the calculation by RCS temperature changes, RCS temperature should be maintained as follows:

4.3.1 IF RCS temp is < 545 of, THEN the RCS temperature should not change by more than 0.1 OF during the test.

4.3.2 IF RCS temp is:::: 545 OF, THEN the RCS temperature should not change by more than 0.3 OF during the test.

4.3.3 IF required to maintain RCS temperature, THEN control rods, turbine load or boron concentration should be adjusted as necessary.

4.4 The calculation assumes that changes in RCS volume due to PZR temperature 1 pressure fluctuations are negligible. Pressurizer parameters should be maintained stable to minimize inaccuracy.

4.5 The following guidelines should be followed to maximize precision:

IF available, THEN computer points should be used for obtaining data.

Otherwise, the available indications are to be read as accurately as possible.

For RCS Tavg, the computer point data should be entered to include three decimal places(i.e., 572.204 OF).

For other computer points and RCDT level, the data should be entered to include at least one decimal place (i.e., 50.1 %).

Identified and unidentified leakage rates are to be reported in two decimal places (e.g., 0.07 gpm).

IF possible, THEN normal makeup to the VeT should be avoided.

4.6 IF the RCDT or PRT level indication is invalid, THEN use 0 gpm for RCDT or PRT portion of identified leakage unless leakage into the RCDT or PRT is to be determined using another approved method.

4.7 To ensure that the STP-9.0 Computer Program remains current, the Engineering Support Group should be notified of any revision or TCN to the Data Sheet 1. Version 42.0

08/08/08 09:39:04 FNP-I-STP-9.0 5.0 Instructions 5.1 NOTE:

I&C N"f\\ 5.2 CV c...~ 5.3 NOTE:

5.4 NOTE:

5.5 5.6 The RCDT system is aligned as follows:

5.1.1 RCDT level is in the normal operating band.

5.1.2 Close RCDT PUMPS DISCH LINE ISO Q1G21HV7136 The following step is only required if increased accuracy is necessary for determination of leak rate into PRT or the MCB PRT level indicator has a problem.

IF required, THEN have I&C connect a calibrated digital voltmeter across the output ofLQY-470, location C5-231.

Place VCT HI LVL DIVERT VLV, QIE2ILCVI15A, in the VCT position.

Batch Integrator readings will be taken prior to and at the conclusion of each make up evolution.

Read and record initial readings on data sheet I.

A time span of at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months should be used during normal steady state plant operations, however if plant conditions dictate, a shorter time span may be used. (30 minutes minimum).

After the desired time span (normally 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) record final values on data sheet 1.

Record the following readings (recorded for trending purposes):

Rad Monitor R-I1 CPM Rad Monitor R-12 CPM Ctmt. Sump lvl. Q1G21LI3282A Inches Ctmt. Sump lvl. QIG21LI3282B Inches Version 42.0

08/08/08 09:39:04 1

FNP-1-STP-9.0 NA. 5.7 IF the RCS leakrate program is to be used, THEN verify that the program is revision 3.

NOTE:

If the RCS leakrate program is used, then the remainder of data sheet 1 may be left blank.

5.8 Calculate identified and unidentified leakages using the RCS leakrate program or formulas on data sheet 1.

ACCEPTANCE CRITERIA:

Identified Leakage ~ 10 gpm Unidentified Leakage ~ 1 gpm 5.9 IF unidentified leakage is more negative than -0.2, THEN re-perform leak rate measurement.

_1_ 5.10 Open RCDT PUMPS DISCH LINE ISO QlG2lHV7136.

IV

_I _ 5.11 Place VCT HI LVL DIVERT VLV, QlE2lLCVl15A in the AUTO position.

IV 5.12 IF computer point LC0500 is available, THEN review the RCS leakrate trend (last 30 days if possible) on IPC to determine if any abnormal trends exist.

5.13 IF unidentified leakage is >0.15 gpm, THEN re-perform leak rate measurement to confirm the results.

5.14 IF unidentified leakage is confirmed to be >0.15 gpm, THEN perform the following: (steps may be performed in any order) 5.14.1 Perform inspection to identify the leakage path(s) (AOP-l.O, attachments 2 through 5).

5.14.2 Perform evaluation including any recent maintenance, plant evolutions or filter alignments to locate source of leakage, determine corrective actions and the effects of the leakage.

5.14.3 IF leakage is NOT known to be outside CTMT, THEN request chemistry sample CTMT via R-67 for iron analysis.

5.14.4 Submit CR to document the leakage and actions taken. Version 42.0

08/08/08 09:39:04 1

FNP-I-STP-9.0 I&e

_1_ 5.15 IF applicable, THEN have I&e remove the calibrated digital voltmeter IV installed in step 5.2.

5.16 Update OPS home page (ULR Data spreadsheet) with unidentified leakage rate.

5.17 IF used for Res leakrate calculation, THEN attach the computer generated Data Sheet 1 to this procedure.

6.0 References 6.1 P&ID D-175037 - ReS, sheet 2 6.2 P&ID D-175039 - eves, sheet 2 6.3 P&ID D-175042 - Waste Processing System, sheet 1 Version 42.0

08/08/08 09:39:04 INSTRUMENT NAME Computer TIME (MCB)

TE0453 LIQ PRZR TEMP (TI0453)

PC0482, PT0455, PT0456 or PT0457 PRZRPRESS (PI 455, PI 456 (Note 1) or PI 457)

TC0484 (preferred),

OR TY0412K, OR TY0422K,OR RCSTAVG TY0432K (Note 7)

(Note 1)

(Average ofTI 412D, 422D &

432D)

RCS Temperature CF (Note 5)

Correction Factor LC 1600 OR Average ofLT0459, PRZRLVL L T0460 & LT0461 (LI-459, 460, 461)

LT0115 VCTLVL (LI 115)

LI 1003 Waste Pnl or RCDTLVL BOP LS261 Pos 6 LT0470 PRTLVL (LI 470)

(Note 2)

FIS 168 TOTAL FLOW BATCH INTEG

From Tank Curve Book DATA SHEET 1 RCS Leakage INITIAL of psig of

Gal

Gal Gal.

Gal.

Page 1 of3 1 I FINAL of psig of N/A

Gal

Gal Gal.

Gal.

A=

FNP-I-STP-9.0 DATA SHEET 1 FINAL - INITIAL Minutes No significant change t:: 1°F)

No significant change t::5 psig) dT =

of Maximum change of 0.3°F allowed ifTAVG is 545°F or greater, 0.1 OF if TAVG is less than 545°F.

B=dT x CF =

Gal.

C = 56.3 x

%=

Gal.

D=14.18x

%=

Gal.

E=

Gal.

(Enter 0 if negative)

F=

Gal.

{Enter 0 if n~ative)

G=

Gal. Dilution and Blended Makeup Version 42.0

08/08/08 09:39:04 Total Leakage 1

FNP-I-STP-9.0 DATA SHEET 1

= B - C - D + G = -'-.( _--,--) ----'.(_---::--')_--.::...( --;:-------,--) +_(""----'-) _ ---GPM A

()

(Note 6)

Identified Leakage

=_E +_F =....:...( _--,--:c)_+--O.(-..,...-_..:..-.) +

= ---GPM A

(

)

Other leakage (Note 6)

Other Leakage:

Source Rate CGPM)

Total Other Unidentified Leakage =

GPM Total Leakage Identified Leakage (Notes 3, 4, & 6)

ACCEPTANCE CRITERIA:

Identified Leakage ~ 10 gpm Unidentified Leakage ~ 1 gpm Page 2 of3 Version 42.0

08/08/08 09:39:04 NOTES:

1 FNP-I-STP-9.0 DATA SHEET 1 1 IF TAVG < 530°F, THEN use: PI-402A (PT0402) and PI-403A (PT0403), lC and lA Loop RCS WR PRESS (Avg. of Readings)

AND TR-4l0 (TE04l0) and TR-413 (TE04l3), RCS COLD AND HOT LEG TEMP (Avg. of Readings) 2 Calibrated fluke may be used for PRT level determination if deemed necessary.

3 For reporting purposes values between -0.2 and 0 gpm shall be reported as 0 gpm. Values more negative than -0.2 gpm indicate a potential problem and therefore shall be reported as is.

4 If unidentified leakage> 0.9 but < 1 gpm, test should be reperformed with ZAS secured. At maximum injection rate, ZAS can introduce -6.03 gpm error into calculation.

5 Obtain CF from Table 1 using the nearest value of RCS temperature. N/A ifRCS Leakrate program is used.

6 Leakage calculations are to be reported in two decimal places (e.g.,

0.07 gpm).

7 TC0484 is preferred for RCS Tavg, but an individual loop temperature may be used if desired due to instability in the average reading.

TABLE 1 RCSTemp p

Corr-=....... llnn Page 3 of3 Version 42.0

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VOLUME n CURVE 27A PRESSURIZER RELIEF TANK CAPACITY CAPACITY (GAL) VS LEVEL (FEE'!')

REV. 1 March 8,1978 GAF APPROVED:

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VOLUME II CURVE 27C PRESSURIZER RELIEF TANK CAPACITY TABLE NIB32TOOl CAPACITY (GAL) VS % LEVEL REV. 0 May 7, 1980 GAF

(

APPROVED:

fC8-I-V{)i.'J.." (.J.V'J.1C

~!M~ k)."CUIM~~

r;lt~o TECHNICAL SUPERINTENDENT DATE

% LEVEL GALLONS

~.~ LEVEL GALLOl'l::;

... LEVEL GALLON~3

>, LEVEL GALU)t,~S 0.0 222.12 26.0 2272.2'3 52.0 5086.:38 78.13 7842.44

1. ~3 272.10 27.13 2372.49 53.0 51'38.21 7':~. 13 7936.83 2.0 325. :3'~

28.0 2473.73 54.13 5:309. ',2

30.0 8029.90 3.0 381. 81 29.0 2575.91 55.0 5421. 45 81.0 8121.58 4.13 441. 16 30.0 2678.98 56.0 5532.7'3 82.0 8211. 80 5.0 503.28 31.0 2782.92

57. ~J 564:3.89 83.0 8300.53 6.0 568.e5

32.0 2887.70 58.13 5754.6:3 84.0 8387.71 7.0 635.34

3:3.13 2993.24 59.0 5865.14 85.0 847:3.2:3

8. (1 705.02 34.0

3099.49 60.0 5975.22 86.13 8557.136 9.13 776. '3',

35.0 3206.4:3 61.0 61384. 9~)

87.0 8639.11

,.' ~

113.0 851,,16 36.0 3314.00 62.0 6194.11 88.0 871".2'3 11.0 927.4:3 37.0 3422.18 63.0 6:302.82 89.0

3797.57 12.0 1005.71

38.0 35:30.89 64.0 641 L 0f1 90.0 8873. :~4 13.0 1085.89 39.0 3640.10 65.0 6518.57 91.0 894::3.01 14.0 1167.94 40.0

3749.78 66.10 6625.51 92.0 9101'3.98 15.0 1251. 77 41.0 3859.86 67.0 6731.76 9:3.0 908'3.66 16.0 1337.29 42.0 3970.32 68.0 68:37.30 94.0

',156. '35 17.0 1424.47 43.0 4081. 11 69.0 6',42.08 95.121 9221.72 18.10 151:3.2121 44.0 4192.21 70.121 7046.82 96.0 9283.84 19.0 1603.42 45.121 4303.55 71.8 714'31.09 97.10 934:3.19 20.0 1695.10 46.121 4415.08 72.0 7251. 27 98.0 9399.6i 21.121 1788.17 47.10 4526.79 7:3.0 7:352.51

'39.0 9452.. 9121 22.0 1882.56 48.0 4638.62 74.0 7452.71 te0.a 13502. ~38 23.0 1*~78. 22 49.0 4750.54 75.0 7551. 86 24.8 212175.11 5121.0 4862.50 76.0 764',. :3 f;t 25.0 2173.14 51.0 4,,74.46 77.10 7746.78

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i, VOI..UB II cmlVI 28A UACTOll COOLANT DRAIN TANK CAPACITY N1G21TOOl

, '" CAPACITY (GAL) VS LEVEL (FEET)

"REV 2 June ll, 1986 JMR

"" APPROVED:

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,~

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.. ~..

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GALLONS BIGHT 0.00 0.25 0.50 0.75 1.00 1.25 1.47 1.72 1.97 2.23 2.47 2.72 VOLUME 0.0 1.8 4.8 8.6 12.9 17.4 21.4 26.0 30.4 34.6 38.3 41.3 1"".~""'

GALLONS 0.0 13.7 38.2 69.0 103.9 141.5 175.1 213.9 251.2 285.7 315.8 339.0 c,.,...... +.....,......,'+-.....,..j".........,.. TO;,.;,.T:,:,:AL,'" VOLllME IN GALLONS 349. 9 I t

Unit 1 Volume II Curve 28B Reactor Coolant Drain Tank Capacity* PlB-I-lIot;;,,-Cl<'V;A813 NIG21TOOl Capacity (Gallons) vs % Level Rev. 2, December 14, 1981, C.A.P.

Approved:

) -;).i{ - a.i=::

Technical Superintendent Date

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3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAG E RCS Operational LEAKAGE 3.4.13 LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and

d.

150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION A.

RCS operational A.1 Reduce LEAKAGE to LEAKAGE not within limits within limits.

for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B.

Required Action and B.1 Be in MODE 3.

associated Completion Time of Condition A not AND met.

B.2 Be in MODES.

OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

Farley Units 1 and 2 3.4.13-1 COMPLETION TIME 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 6 hours 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Amendment No. 163 (Unit 1)

Amendment No. 156 (Unit 2)

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SR 3.4.13.1 SR 3.4.13.2 SURVEILLANCE

NOTES--------------------------------

1. Not required to be performed in MODE 3 or 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance.

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

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is S 150 gallons per day through anyone SG.

FREQUENCY

NOTE--------

Only required to be performed during steady state operation 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 72 hours Farley Units 1 and 2 3.4.13-2 Amendment No. 163 (Unit 1)

Amendment No. 156 (Unit 2)

RCS Operational LEAKAGE B 3.4.13 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.13 RCS Operational LEAKAGE BASES BACKGROUND Farley Units 1 and 2 Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary.

Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

B 3.4.13-1 Revision 0

BASES RCS Operational LEAKAGE B 3.4.13 APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do not SAFETY ANALYSES address operational LEAKAGE. However, other operational LEAKAGE is typically seen as a precursor to a LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) is 1 gpm as a result of accident induced conditions. The LCO requirement to limit primary to secondary LEAKAGE through anyone SG to less than or equal to 150 gpd (Le. total leakage less than or equal to 450 gpd) is significantly less than the conditions assumed in the safety analysis (with leakage assumed to occur at room temperature in both cases).

Farley Units 1 and 2 Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident.

To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is released via the main steam safety valves. The majority of the activity released to the atmosphere results from the tube rupture. Therefore, the 1 gpm primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential.

The SLB is more limiting for primary to secondary LEAKAGE. The safety analysis for the SLB assumes 500 gpd and 470 gpd primary to secondary LEAKAGE in the faulted and intact steam generators respectively as an initial condition. The dose consequences resulting from the SLB accident are bounded by a small fraction (Le., 10%) of the limits defined in 10 CFR 100. The RCS specific activity assumed was 0.5 JlCi/gm DOSE EQUIVALENT 1-131 at a conservatively high letdown flow of 145 gpm, with either a pre-existing or an accident initiated iodine spike. These values bound the Technical Specifications values.

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

B 3.4.13-2 Revision 24

BASES Leo Farley Units 1 and 2 Res Operational LEAKAG E B 3.4.13 Res operational LEAKAGE shall be limited to:

a.

Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LeO could result in continued degradation of the RePB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

b.

Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LeO could result in continued degradation of the RePB, if the LEAKAGE is from the pressure boundary.

c.

Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the ReS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (Rep) sealleakoff (a normal function not considered LEAKAGE). Violation of this LeO could result in continued degradation of a component or system.

d.

Primary to Secondary LEAKAGE Through Any One SG The limit of 150 gpd per each SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 4). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, "The ReS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

B 3.4.13-3 Revision 24

BASES APPLICABILITY ACTIONS Farley Units 1 and 2 RCS Operational LEAKAGE B 3.4.13 In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

LCO 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight.

If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

Unidentified LEAKAGE or identified LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

B.1 and B.2 If any pressure boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within limit, or if unidentified or identified LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

B 3.4.13-4 Revision 24

BASES SURVEILLANCE REQUIREMENTS Farley Units 1 and 2 SR 3.4.13.1 RCS Operational LEAKAGE B 3.4.13 Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

The RCS water inventory balance must be met with the reactor at steady state operating conditions and near operating pressure. The I Surveillance is modified by two Notes. Note 1 states that this SR is not required to be performed in MODES 3 and 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation near operating pressure have been established.

Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful and a Note requires the Surveillance to be met when steady state is established.

For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment air cooler condensate flow rate. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation."

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE. This is because LEAKAGE of 150 gpd cannot be measured accurately by an RCS water inventory balance.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. A Note under the Frequency column states that this SR is required to be performed during steady state operation.

( continued)

B 3.4.13-5 Revision 24

BASES SURVEILLANCE REQUIREMENTS REFERENCES Farley Units 1 and 2 SR 3.4.13.2 RCS Operational LEAKAGE B 3.4.13 This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gpd through anyone SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, "Steam Generator Tube Integrity," should be evaluated. The 150 gpd limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through anyone SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from oneSG.

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

The Surveillance Frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. During normal operation the primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with EPRI guidelines.

1.

10 CFR 50, Appendix A, GDC 30.

2.

Regulatory Guide 1.45, May 1973.

3.

FSAR, Section 3.1.2.6, 5.2.7,10.4,11.0, 12.0 and 15.0.

4.

NEI97-06, "Steam Generator Program Guidelines."

5.

EPRI TR-104788, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

B 3.4.13-6 Revision 24

(

HLT-32 ADMIN exam A.3 SRO&RO Page 1 of 6 A.3 SRO&RO Radiation Control ADMIN G2.3.4 - SRO and RO TITLE: Determine If Any Radiation Dose Limits Will Be Exceeded TASK STANDARD: Calculate Dose expected for two workers, and determine that the job cannot be performed by one worker due to FNP Admin dose limit, and the job cannot be performed by the other worker due to the RWP Digital Alarming Dosimeter (DAD) Alarm limit.

PROGRAM APPLICABLE: SOT SOCT OLT~LOCT __

ACCEPTABLE EVALUATION METHOD: ~

PERFORM SIMULATE DISCUSS EVALUATION LOCATION:

SIMULATOR CONTROL ROOM ~

CLASSROOM PROJECTED TIME:

30 MIN SIMULATOR IC NUMBER: ---:..;N~A~ __ _

ALTERNATE PATH TIME CRITICAL PRA __

Examinee:

Overall JPM Performance:

Satisfactory

[]

Unsatisfactory []

Evaluator Comments (attach additional sheets if necessary)

EXAN.IT.NER: ______________ __

HLT-32 ADMIN exam A.3 SRO&RO Page 2 of 6 CONDITIONS When I tell you to begin, you are to DETERMINE IF ANY RADIATION DOSE LIMITS WILL BE EXCEEDED during a containment entry to inspect and take pictures at the 1A RCP seal area. The conditions under which this task is to be performed are:

a. A power reduction to 12% has been performed on Unit 1.

b. The ED and the lIP Supervisor on-call have approved personnel entry inside the 105' Missile Barrier.

c. The transit route is <2 mrlhr except as noted on the provided survey maps.

d. lIP has determined the lowest dose route is down the containment stairwell to the 105' level, outside the bio shield, go inside the bio shield at the south east entrance, proceed past the C loop to the A loop area, and ascend the ladder to the A RCP LOWER platform to access the 1A RCP Seal area (See survey maps).

e. lIP estimates that it will take 3 minutes to travel inside the bio shield and ascend the ladder to the 1A RCP.

f.

The workers estimate that it will take 36 minutes at the RCP seal area for the inspection.

g. NO contact with any RCP Cubical surfaces will be needed for the inspection.

h. Worker A year to date accumulated dose is 1700 mr.

i.

Worker B year to date accumulated dose is 1650 mr.

j. The containment survey maps have been marked during the pre-job brief with the expected transit route in red.

k. A radiological pre-job brief has been performed.

1.

Your task is to prepare for the seal inspection on the 2A RCP. You are to determine:

1.) what predicted dose both workers will receive, and 2.) if workers A and B will be able to perform the task without exceeding any dose limits.

EVALUATION CHECKLIST ELEMENTS:

STANDARDS:

START TIME Calculates dose that 25mr will be received, RESULTS:

(CIRCLE)

1.

Calculates dose that will be received from entry into the bio shield to the RCP.

inside the bio shield traveling to the RCP for a S / U total of 50mr for the entry and return trips.

(500mrlhr)(3minltrip)(2trips)(1hr/60mins)

=50mr

2.

Calculates dose that will be received near the RCP.

(450 mrlhr)(36min) (1hr/60mins)=270mr Calculates dose that 270mr will be received by each worker near the RCP.

S / U

HLT-32 ADMIN exam A.3 SRO&RO EVALUATION CHECKLIST ELEMENTS:

3.

Calculates total dose that will be received by each worker.

270nrr+50rrrr=320nrr

4.

Calculates yearly dose which would be accumulated if the job was performed and dose accumulated as estimated.

320+ 1700=2020nrr>ADMIN limit 320+ 1650=1970nrr< ADMIN limit, but>

DAD limit STANDARDS:

Calculates dose that will be received by each worker during the entire entry for a total of 320nrr.

Calculates the estimated yearly dose after job would be 2020nrr for worker A and 1,970nrr for worker B.

Page 3 of 6 RESULTS:

(CIRCLE)

S / U S / U NOTE TO EV ALUATOR:

Examinee may indicate that permission from HP is required for both workers to raise their DAD dose alarm setpoints to greater than 320mr AND/OR an admin dose extension above 1220 mr is required for worker A to perform the job.

5.

Determines that the DAD Alarm limit is exceeded by BOTH workers, and the annual admin dose limit would be exceeded by worker A ONLY.

_STOP TIME Determines that the DAD limit (above 300nrr for the job) would be exceeded by BOTH workers, and the ADMIN limit would be exceeded by worker A only (above the limit of 2,000 nrr).

S / U Terminate when both worker doses have been determined and evaluation of limits is complete.

CRITICAL ELEMENTS: Critical Elements are denoted with an asterisk (*) preceding the element number.

HL T-32 ADMIN exam A.3 SRO&RO Page 4 of 6 GENERAL

REFERENCES:

1. FNP-0-M-001, Version 18.0

2. KA: G2.3.4 RO-3.2 SRO-3.7 GENERAL TOOLS AND EQUIPMENT Provide:

1. FNP-O-M-OOl, HP Manual, Version 18.0

2. Containment Survey Maps

3. RWP

4. Calculator (or the Applicant may supply a calculator)

COMMENTS:

HLT-32 ADMIN exam A.3 SRO&RO Page 5 of 6 KEY:

Worker A WorkerB Initial Dose 1700mr 1650mr Trip to pump 3minx500mrlhr = 25mr 3minx500mrlhr = 25mr At pump 36minx450mrlhr = 270mr 36minx450mrlhr = 270mr Trip from pump 3minx500mrlhr = 25mr 3minx500mrlhr = 25mr Total dose for job 320mr 320mr Margin left to 2000mr/ qtr admin 300mr 350mr limit Margin left to 300mr DAD limit 300mr 300mr Total IF Job was performed 1700+320=2020mr 1650+320=1970mr Limiting dose which could be 300mr due to the ADMIN AND 300mr due to the DAD limit received the DAD limits ONLY Worker A cannot perform the task due to exceeding the FNP ADMIN 2000mr/qtr limit, AND the DAD alarm dose limit of 300mr.

Worker B cannot perform the task due to exceeding the DAD dose limit of 300mr, but does NOT exceed the FNP ADMIN dose limit of 2000mr/qtr.

ONLY with an increased DAD dose limit approval by Health Physics (HP) AND an FNP admin dose limit extension could Worker A perform the job.

(must take into account the 25mr on the return trip or it will appear that Worker A can perform the task without exceeding his ADMIN Yearly margin or DAD limit)

ONLY with an increased DAD dose limit approval by Health Physics (HP) could Worker B perform the job.

(must take into account the 25mr on the return trip or it will appear that Worker B can perform the task without exceeding his DAD dose margin)

Limiting dose which could be 300mr due to the ADMIN or 300mr due to the DAD limit received DAD limit

(

A.3SRO & RO (1 page)

HANDOUT CONDITIONS When I tell you to begin, you are to DETERMINE IF ANY RADIATION DOSE LIMITS WILL BE EXCEEDED during a containment entry to inspect and take pictures at the lA RCP seal area. The conditions under which this task is to be performed are:

a. A power reduction to 12% has been performed on Unit 1.

b. The ED and the HP Supervisor on-call have approved personnel entry inside the 105' Missile Barrier.

c. The transit route is <2 rnrlhr except as noted on the provided survey maps.

d. HP has determined the lowest dose route is down the containment stairwell to the 105' level, outside the bio shield, go inside the bio shield at the south east entrance, proceed past the C loop to the A loop area, and ascend the ladder to the A RCP LOWER platform to access the IA RCP Seal area (See survey maps).

e. HP estimates that it will take 3 minutes to travel inside the bio shield and ascend the ladder to the IA RCP.

f.

The workers estimate that it will take 36 minutes at the RCP seal area for the inspection.

g. NO contact with any RCP Cubical surfaces will be needed for the inspection.

h. Worker A year to date accumulated dose is 1700 rnr.

i. Worker B year to date accumulated dose is 1650 rnr.

j. The containment survey maps have been marked during the pre-job brief with the expected transit route in red.

k. A radiological pre-job brief has been performed.

1.

Your task is to prepare for the seal inspection on the 2A RCP. You are to determine:

1.) what predicted dose both workers will receive, and 2.) if workers A and B will be able to perform the task without exceeding any dose limits.

Plant Farley U-1155 ft. Containment (1CB155)

Survey #43180 - 0712312008 16:48 -

Rx Power Level: 0%

1 Caution-Contaminated Area Danger - Locked High Radiation Area Notify HP prior to entry 2 Caution-Contaminated Area Danger - Locked High Radiation Area Notify HP prior to entry

(

Plant Farley U-1 105ft Containment (1 CB1 05)

Survey #43166 - 07/231200810:00 -

Rx Power Level: 0%

RWP required for entry HP Escort Required OR Alarming Dosimeter Req'd Danger - High Radiation Area 2 RWP required for entry HP Escort Required OR Alarming Dosimeter Req'd Danger - High Radiation Area 3 RWP required for entry HP Escort Required OR Alarming Dosimeter Req'd Danger - High Radiation Area 4 RWP required for entry Caution -

Contaminated Area Grave Danger - Very High Radiation Area Dosimetry Required for Entry Notify HP prior to entry 5 Grave Danger - Very High Radiation Area 6 RWP required for entry HP Escort Required OR Alarming Dosimeter Req'd Danger - High Radiation Area

Entrance posted to CTMT -

LO\\¥erPlatform Entrance posted to CTMT -

Lower Platform Plant Farley U1 A RCP Cay (1 CB129)

Survey #43141 - 07/221200817:39 -

Rx Power Level: 0%

ASIG

~I Dose rates at the Regen to access this platform were 25 - 100 mRern!hr.

Upper Platform Dose rates at the Regen to access this platform were 25 - 100 mRem/hr.

Upper Platform Grave Danger - Very High Radiation Area Caution -

Contaminated Area

HP Information Radiation Work Permit Plant Farley 08-3493 AC.IIY.E D

J~bt' All work associated with lA RCP Seal inspection in Unit One CTMT.

escnp Ion Location lUI CONTAINMENT Page 1 of2 Unit Rev 2

1 HP Coverage

,AuJb.m:.izatic;m lJrwfigg INDIVIDUAL Start Date 7122/2008 End Date 1213112008 CONTINUOUS INDIVIDUAL Job Supv. CHRIS THORNELL Ext. 4768

~

_________ Ra m_*o_lO~g~ic_a_I_C_o_n_d_iti_*o_n_s __________ ~11

~ _________________ T_a_s_ks ________________ ~

AIRBORNE LEVELS: <.3 DAC Part,Iodine & <2 DAC Noble Gas or Tritium Description CONTAMINATION LEVELS:<200 MRAD/lOOCM2 BETA-GAMMA &<20 DPMllooCM2 ALPHA CTMT INSPECTIONS, W ALKDOWNS, MAINTENANCE < THAN 15% POWER RADIATION LEVELS: <10000 mRemIhr@ 30 CM I

Dosimetry

'TLDANDDAD PAM OR PEA REQUIRED I

Protective Clothing Requirements

'THE FOLLOWING DRESSOUT REQUIREMENTS ARE ALLOWED DEPENDING ON CONDITIONS PARTIAL ENTRY, STANDARD LABCOAT, STANDARD COVERALL FULL FACE HOOD Respirators Usage is Conditional per HP Instructions This RWP involves work identified as Radiologically Risk Significant activities.

No entries allowed inside the 105' Missle Barrier >15% Power; DAD Alarms Dose (mr) Rate (mrlh) 300 600 Entry inside the 105' Missile Barrier following Rx Trip and/or at Power levels of 15% or less requires the prior approval of both the ED and the Health Physics Supervisor On Call, Prior to entry ensure each worker has an adequate exposure margin. Account for any accumulated Neutron Exposure.

A radiological pre..job briefmg shall be conducted by a HP Foreman Qualified Individual and the Work Supervisor with personnel involved in the job.

Worker(s) shall periodically check their DAD.

Prior to exceeding the Accumulated Dose Alarm, worker(s) shall exit the RCA. Upon receipt of the Dose Rate Alarm, back out of the area until the alarm clears & contact Health Physics.

The Health Physics Technician shall ensure that Locked High Rad Area controls are implemented for entry into any posted Locked High Rad Area.

file:IIE:\\NRC Exam development\\HLT-32 exam 5-6-2008\\4 HLT-32 ADMIN\\3 admin radiation control\\...

8111/2008

HP Information Page 2 of2 The HP Technician providing job coverage shall ensure Neutron calculation worksheets (DOS form 933) are completed for entries made in a Neutron Radiation area.

Wearing personal items such as chains, rings, etc. should be minimized when entering Neutron Radiation fields due to possible Neutron activation of the objects.

Specific Designated Controls are incorporated into this RWP and may only be deviated from with the HP Manager or his designee approval.

The following special instructions may be deviated from with the Health Physics Supervisor's permission.

No work is allowed on this RWP in areas where radiological conditions are greater than those listed in the expected "Radiological Conditions".

The Normal Dose and Dose Rate Alarm values may be adjusted by Health Physics based on expected conditions.

Prepared 7/2312008 06:47 by BLACK Approved 7123/2008 07:00 by PETERS Suspended Terminated file:IIE:\\NRC Exam development\\HLT-32 exam 5-6-2008\\4 HLT-32 ADMIN\\3 admin radiation control\\,..

8/1112008

02124/04 14: 13 :02 FNP-O-M-OOI December 12,2002 Version 18.0 SOUTHERN NUCLEAR COMPANY JOSEPH M. FARLEY NUCLEAR PLANT HEALTH PHYSICS MANUAL PROCEDURE USAGE REQUIREMENTS PER FNP-0-AP-6 SECTIONS Continuous Use Reference Use Information Use ALL D. E. GRISSETTE Nuclear Plant General Manager S

A F

E T

Y R

E L

A T

E D

Date Issued --=.2-"-1=2-'-0'-'4 ______ _

HLT-32 ADMIN exam AA SRO Page 1 of 6 A.4 SRO Emergency Plan ADMIN G2.4.44-SRO TITLE: Evaluate plant conditions during a site area emergency to determine if a follow-up message or an upgrade notification is warranted, and complete all required forms.

TASK STANDARD: Classify an emergency event and determine an upgrade from SAE to GE is required, fill out all forms for emergency notification, and initiate correct Protective Action Recommendations (PARS) within the time allowed.

PROGRAM APPLICABLE: SOT SOCT OLT~LOCT __

ACCEPTABLE EVALUATION METHOD: ~

PERFORM ~

SIMULATE DISCUSS EVALUATION LOCATION: ~

SIMULATOR ~

CONTROL ROOM..x.. CLASSROOM PROJECTED TIME:

30 MIN SIMULATOR IC NUMBER:

N/A ALTERNATE PATH TIMECRITICAL~ PRA

TIDS JPM IS TIME CRITICAL*

Examinee:

Overall JPM Performance:

Satisfactory 0

Unsatisfactory 0 Evaluator Comments (attach additional sheets if necessary)

EXANUNER: ________________ _

HLT-32 ADMIN exam A.4 SRO Page 2 of 6 CONDITIONS When I tell you to begin, you are to EVALUATE PLANT CONDITIONS DURING A SITE AREA EMERGENCY TO DETERMINE IF A FOLLOWUP MESSAGE OR AN UPGRADE NOTIFICATION IS WARRANTED, AND COMPLETE ALL REQUIRED FORMS. The conditions under which this task is to be performed are:

a. Unit 1 was at 100% power when a Large Break LOCA occurred.

b. A Site Area Emergency (SAE) has been declared for Unit 1, FSI - Loss or Potential Loss of ANY Two Barriers due to:

Potential loss of the Fuel Clad Barrier (4. RVLS Plenum LEVEL less than 0%).

Loss of the RCS Barrier (2. RCS subcooling less than 16°F {less than 45° F Adverse}).

c. Unit 2 is unaffected and has remained at 100%.

d. EIP-9.0, Guideline 2, Site Area Emergency, has been performed up to step D.3.

e. The Control Room reports the following current conditions on Unit 1:

RVLS lights are all red RCS subcooling is 12°F RE-27A and RE-27B, Containment Radiation Monitors, are both 100 RIhr and rising.

RE-14, Plant Vent, RE-21, Vent Stack Particulate, and RE-22, Vent Stack Gas, are in alarm.

The Shift Radio Chemist has projected dose to be 1.3 REM TEDE at the site boundary.

Wind Direction

= from 95°.

Wind Speed

= -1:L mph.

ilT

= +1.5°F.

f. Another SRO is standing by to make any requested announcements, callouts, or notifications.

g. This JPM contains time critical elements.

h. A pre-job brief is not required.

i.

You are the ED and are required to evaluate plant conditions and determine which is warranted:

an upgrade in classification per EIP-9.0, Step 4.0, OR a follow-up message per EIP-9.0, Step 6.0, AND then fill out all applicable forms and paperwork.

EXAMINER NOTES: DO NOT START THE TIME UNTIL THE APPLICANT UNDERST ANDS THE TASK.

An EXAMINER'S KEY is available for all forms On Guideline 1, boxes are circled instead of checked on the KEY to ensure the content of the boxes are legible. Circling the boxes, checking inside the boxes, or x'ing inside the boxes are all acceptable for the applicant.

EVALUATION CHECKLIST ELEMENTS:

STANDARDS:

RESULTS:

(CIRCLE)

HL T-32 ADMIN exam A.4 SRO EVALUATION CHECKLIST ELEMENTS:

__ TIME CRITICAL START TIME STANDARDS:

Page 3 of 6 RESULTS:

(CIRCLE)

NOTE:* THE TIME IT TAKES TO CLASSIFY THE EVENT IS TIME CRITICAL AND MUST BE COMPLETED IN 15 MINUTES.

1.

Classify the event using Data Sheet 1 from EIP-9.2.

Data Sheet 1 completed through step 9, including signature, date and time. Event classified as a GENERAL EMERGENCY - RGI and/or FGl:

RGI #2. >1000mrTEDE FGl:

o Fuel clad barrier loss - #5, Containment Radiation Monitoring, oRCS barrier loss - #2, RCS Leak Rate, and o

CTMT barrier loss - 7, Other indications.

__ TIME CRITICAL STOP / START TIME SIU NOTE:

THE TIME IT TAKES TO COMPLETE AND APPROVE THE DECLARATION FORM PER THE FOLLOWING ELEMENTS IS TIME CRITICAL AND MUST BE COMPLETED IN 15 MINUTES.

2.

Directs notification of personnel on Directs notification of personnel on site.

S I U site.

NOTE:

ELEMENT 3 WAS ACCOMPLISHED WHEN THE SITE AREA EMERGENCY WAS DECLARED. EVEN THOUGH THE PROCEDURE STATES TO PERFORM IT, THE APPLICANT MAY REALIZE IT DOESN'T NEED TO BE PERFORMED AGAIN. THERE ARE NO ADVERSE CONSEQUENCES IF PERFORMED AGAIN OR NOT. TIDS IS NOT A CRITICAL STEP.

ELEMENT 3 MAY BE DELEGATED TO THE EXTRA SRO.

3.

Directs callout the ERO staff.

Individual requested to activate the ERO callout system per FNP-O-EIP-8.3, Table 2.

S I U (Cue: The request to initiate ERO callout is acknowledged. )

HL T -32 ADMIN exam A.4 SRO EVALUATION CHECKLIST Page 4 of 6 RESULTS:

ELEMENTS:

STANDARDS:

(CIRCLE)

NOTE: ACCURATE COMPLETION OF CERTAIN STEPS OF EIP-9.0, GENERAL EMERGENCY NOTIFICATION FORM, IS ESSENTIAL TO ENSURE ADEQUATE NOTIFICATION OF STATE AND LOCAL AGENCIES. THESE STEPS ARE SHOWN AS THE STANDARDS FOR ELEMENT NUMBER 4 (critical tasks are based on shaded portions of Guideline 1 form which annotate the items which affect EP Performance Indicators (PIs).

4.

Complete EIP-9.0, Guideline 1 General Emergency Red Verbal Notification Form.

Correct form selected S I U LINE 1 - Indicates Drill OR Actual Event S I U LINE 4 - Indicates General and identifies RGI and/or FGl as criteria for EAL#

S I U EXAMINER'S CUE IF ASKED: EP WEATHER IS NOT AVAILABLE. MET TOWER DATA IS AS STATED IN THE INITIAL CONDITIONS.

LINE 5 - Evaluates PARs and determines that PAR 3 is appropriate, and based on 95° wind direction, the following zones should be evacuated: A, B5, C5, D5, E5, F5, IS, J5, KS, CIO, DIO and EIO.

S I U NO zones are sheltered LINE 6 - Evaluates emergency release and marks - Is Occuring LINE 9 - Accurately completes met tower wind direction and wind speed data.

LINE 10 - Completes declaration time/date

- Matches Step 8 on EIP-9.2 Data Sheet 1 S I U S/U S I U S I U NOTE:

IF EXAMINEE ASKS FOR TIME OF SHUTDOWN, SUBTRACT 25 MINUTES FROM THE START OF TIDS TASK AND PROVIDE THAT TIME.

LINE 11 - Indicates Unit 1 S/U

HLT-32 ADMIN exam A.4 SRO EVALUATION CHECKLIST ELEMENTS:

STANDARDS:

NOTE:

STOP TIME IS AFTER LINE 17 IS COMPLETE:

"APPROVED BY" SIGNATURE, TIME AND DATE IS FILLED IN.

TIME CRITICAL STOP TIME Terminate JPM when initial notification form is completed Page 5 of 6 RESULTS:

(CIRCLE)

CRITICAL ELEMENTS: Critical Elements are denoted with an Asterisk (*) before the element number.

GENERAL

REFERENCES:

1. FNP-0-EIP-9.0 Version 59

2. FNP-0-EIP-9.2 Version 7

3. NMP-EP-109 Version 2.0

4. KA: G2,4,44 RO-2,4 SRO-4,4 GENERAL TOOLS AND EQUIPMENT:

Provide:

1. FNP-0-EIP-9.0, Guideline 1 and Figure 6

2. FNP-0-EIP-9.2, Data Sheet 1

3. NMP-EP-109 Version 2.0

4. Handouts of SAE paperwork already filled out.

COMMENTS:

ML091480143

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