ML050610410

Technical Specification Pages Re Dose Caclulation Methodology to Alternate Source Term

ML050610410
Person / Time
Site:	Ginna
Issue date:	02/25/2005
From:	NRC/NRR/DLPM
To:
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TAC MB9123
Download: ML050610410 (16)

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Definitions 1.1 CHANNEL OPERATIONAL TEST (COT)

CORE ALTERATIONS CORE OPERATING LIMITS REPORT (COLR)

DOSE EQUIVALENT 1-1 31 A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, Interlock, display, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

CORE ALTERATIONS shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

The COLR is the plant specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in Individual Specifications.

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132,1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in ICRP 30, Supplement to Part 1, pages 192-212, table entitled, uCommitted Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity."

E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies (in MeV) per disintegration for non-iodine isotopes, with half lives > 15 minutes, making up at least 95%o of the total non-iodine activity in the coolant.

I E-AVERAGE DISINTEGRATION ENERGY R.E. Ginna Nuclear Power Plant 1.1-2 Amendment 87

3.3 3.3.6 GREATS Actuation Instrumentation 3.3.6 INSTRUMENTATION Control Room Emergency Air Treatment System (GREATS) Actuation Instrumentation The CREATS actuation instrumentation for each Function in Table 3.3.6-1 shall be OPERABLE.

LCO 3.3.V I

APPLICABILITY:

According to Table 3.3.6-1.

I I

ACTIONS

-NOTE-Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION l_COMPLETION TIME A.

One or more Functions A.1 Place one GREATS train in 7 days with one channel or train emergency mode.

inoperable.

B.

One or more Functions B.1.1 Place one GREATS train in Immediately with two channels or two emergency mode.

trains inoperable.

AND B.1.2 Enter applicable Conditions Immediately and Required Actions for one GREATS train made inoperable by inoperable GREATS actuation Instrumentation.

B.2 Place both GREATS trains Immediately in emergency mode.

C.

Required Action and C.1 Be In MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND not met in MODE 1, 2,3, or4.

C.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> I

R.E. Ginna Nuclear Power Plant 3.3.6-1 Amendment 87

CREATS Actuation Instrumentation 3.3.6 CONDITION I

REQUIRED ACTION ICOMPLETION TIME I

D.

Required Action and associated Completion Time of Condition A or B not met during movement of irradiated fuel I

assemblies.

D.1 Suspend movement of irradiated fuel assemblies.

Immediately SURVEILLANCE REQUIREMENTS

-NOTE-Refer to Table 3.3.6-1 to determine which S Rs apply for each C REATS Actuation Function.

SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.6.2 Perform COT.

92 days SR 3.3.6.3

- NOTE -

Verification of setpoint is not required.

Perform TADOT.

24 months SR 3.3.6.4.

Perform CHANNEL CALIBRATION.

24 months SR 3.3.6.5 Perform ACTUATION LOGIC TEST.

24 months R.E. Ginna Nuclear Power Plant 3.3.6-2 Amendment 87

I I

I I

CREATS Actuation Instrumentation 3.3.6 Table 3.3.6-1 CREATS Actuation Instrumentation APPLICABLE LIMITING MODES OR SAFETY OTHER SPECIFIED REQUIRED SURVEILLANCE SYSTEM FUNCTION CONDITIONS CHANNELS REQUIREMENTS SETrINGSIa)

1.

Manual Initiation 1,2.3.4.

2 trains SR 3.3.6.3 NA (b)

2.

AutomaticActuation Logicand 1,2,3,4, 2 trains SR 3.3.6.5 NA Actuation Relays (b)

3.

Control Room Radiation Intake 1.2.3,4.

2 SR 3.3.6.1 s.57 mR/hr Monitors (b)

SR 3.3.6.2 SR 3.3.6.4

4.

Safety Injection Refer to LCO 3.3.2, 'ESFAS Instrumentation,' Function 1, for all Initiation functions and requirements.

(a)

A channel is OPERABLE when both of the following conditions are met:

1. The absolute difference between the as-found Trip Setpoint (TSP) and the previous as-left TSP Is within the COT Acceptance Criteria. The COT Acceptance Criteria is defined as:

las-found TSP - previous as-left TSPI S COT uncertainty The COT uncertainty shall not Include the calibration tolerance.

2. The as-left TSP is within the established calibration tolerance band about the nominal TSP. The nominal TSP Is the desired setting and shall not exceed the Limiting Safety System Setting (LSSS). The LSSS, COT uncertainty, and the established calibration tolerance band are defined in accordance with the Ginna instrument setpoint methodology. The channel is considered operable even If the as-left TSP is non-conservative with respect to the LSSS provided that the as-left TSP is within the established calibration tolerance band.

(b)

During movement of irradiated fuel assemblies R.E. Ginna Nuclear Power Plant 3.3.6-3 Amendment 87

RCS Specific Activity 3.4.16 3.4 3.4.16 REACTOR COOLANT SYSTEM (RCS)

RCS Specific Activity LCO 3.4.16 APPLICABILITY:

The specific activity of the reactor coolant shall be within limits.

MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) 2 500'F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DOSE EQUIVALENT 1-131 specific activity not CO-NOTE -

Ithn limit a

LCO 3.0.4 is not applicable.

within limit.

A.1 Verify DOSE EQUIVALENT Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1-131 c 60 pCi/gm.

AND A.2 Restore DOSE 7 days EQUIVALENT 1-131 to within limit.

B.

Required Action and B.1 Be in MODE 3 with Tavg 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> associated Completion

< 500'F.

Time of Condition A not met.

OR DOSE EQUIVALENT 1-131 specific activity > 60 PCi/gm.

C.

Gross specific activity not C.1 Be in MODE 3 with Tavg 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> within limit.

< 500 0F.

R.E. Ginna Nuclear Power Plant 3.4.16-1 Amendment 87

CS, CREC, and NaOH Systems 3.6 CS, CRFC, and NaOH Systems 3.6.6 CONTAINMENT SYSTEMS Containment Spray (CS), Containment Recirculation Fan Cooler (CRFC), and NaOH Systems 3.6.6 I

I LCO 3.6.6 Two CS trains, four CRFC units, and the NaOH system shall be OPERABLE.

-NOTE-In MODE 4, both CS pumps may be in pull-stop for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of interlock and valve testing of motor operated valves (MOVs) 857A, 857B, and 857C. Power may also be restored to MOVs 896A and 8958, and the valves placed in the closed position, for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the purpose of each test.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION l

REQUIRED ACTION l COMPLETION TIME A.

One CS train inoperable.

A.1 Restore CS train to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.

B.

NaOH system inoperable. B.1 Restore NaOH System to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.

C.

Required Action and C.A Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND not met.

C.2 Be in MODE 5.

84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> D.

One or two CRFC units D.1 Restore CRFC unit(s) to 7 days Inoperable.

OPERABLE status.

E.

Required Action and E.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition D not AND met.

E.2 Be In MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> R.E. Ginna Nuclear Power Plant 3.6.6-1 Amendment 87

CS, CRFC, and NaOH Systems 3.6.6 CONDITION I

REQUIRED ACTION l COMPLETION TIME F.

Two CS trains inoperable. F.1 Enter LCO 3.0.3.

Immediately OR Three or more CRFC units inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Perform SR 3.5.2.1 and SR 3.5.2.3 for valves 896A In accordance with and 896B.

applicable SRs.

SR 3.6.6.2 Verify each CS manual, power operated, and 31 days automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

SR 3.6.6.3 Verify each NaOH System manual, power operated, 31 days and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

SR 3.6.6.4 Operate each CRFC unit for > 15 minutes.

31 days SR 3.6.6.5 Verify cooling water flow through each CRFC unit.

31 days SR 3.6.6.6 Verify each CS pump's developed head at the flow In accordance with test point is greater than or equal to the required the Inservice developed head.

Testing Program SR 3.6.6.7 Verify NaOH System solution volume Is 2 3000 gal.

184 days SR 3.6.6.8 Verify NaOH System tank NaOH solution 184 days concentration is 2 30% and 5 35% by weight.

SR 3.6.6.9 Perform required CRFC unit testing in accordance In accordance with with the VFTP.

the VFTP SR 3.6.6.10 Verify each automatic CS valve in the flow path that is 24 months not locked, sealed, or otherwise secured in position actuates to the correct position on an actual or simulated actuation signal.

R.E. Ginna Nuclear Power Plant 3.6.6-2 Amendment 87

CS, CRFC, and NaOH Systems 3.6.6 I

SURVEILLANCE FREQUENCY SR 3.6.6.11 Verify each CS pump starts automatically on an actual 24 months or simulated actuation signal.

SR 3.6.6.12 Verify each CRFC unit starts automatically on an 24 months actual or simulated actuation signal.

SR 3.6.6.13 Verify each automatic NaOH System valve in the flow 24 months path that is not locked, sealed, or otherwise secured in position actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.6.14 Verify spray additive flow through each eductor path.

5 years SR 3.6.6.15 Verify each spray nozzle is unobstructed.

10 years I

I I

R.E. Ginna Nuclear Power Plant 3.6.6-3 Amendment 87

CREATS 3.7.9 3.7 3.7.9 PLANT SYSTEMS Control Room Emergency Air Treatment System (CREATS)

I LCO 3.7.9 APPLICABILITY:

I Two CREATS Trains shall be OPERABLE.

MODES 1, 2, 3, and 4, During movement of irradiated fuel assemblies.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One CREATS train A.1 Restore CREATS train to 7 days inoperable.

OPERABLE status.

B.

Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion rime of Condition A not AND met in MODE 1, 2, 3, or4.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C.

Required Action and C.1 Suspend movement of Immediately associated Completion irradiated fuel assemblies.

lime of Condition A not met during movement of irradiated fuel assemblies.

D.

Two CREATS trains D.1 Enter LCO 3.0.3.

Immediately inoperable in MODE 1, 2, 3, or4.

E.

Two CREATS trains E.1 Suspend movement of Immediately inoperable during Irradiated fuel assemblies.

movement of irradiated fuel assemblies.

R.E. Ginna Nuclear Power Plant 3.7.9-1 Amendment 87

CREATS 3.7.9 I

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Operate each CREATS filtration train 2 15 minutes.

31 days SR 3.7.9.2 Perform required CREATS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP).

VFTP SR 3.7.9.3 Verify each CR EATS train actuates on an actual or 24 months simulated actuation signal.

R.E. Ginna Nuclear Power Plant 3.7.9-2 Amendment 87

Programs and Manuals 5.5.

b.

SG tubes that have imperfections > 40% through wall, as indicated by eddy current, shall be repaired by plugging or sleeving.

c.

SG sleeves that have imperfections > 30% through wall, as indicated by eddy current, shall be repaired by plugging.

5.5.9 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. This program shall include:

a.

Identification of a sampling schedule for the critical variables and control points for these variables;

b.

Identification of the procedures used to measure the values of the critical variables;

c.

Identification of process sampling points;

d.

Procedures for the recording and management of data;

e.

Procedures defining corrective actions for all off control point chemistry conditions; and

f.

A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

5.5.10 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature filter ventilation systems and the Spent Fuel Pool (SFP) Charcoal Adsorber System. The test frequencies will be in accordance with Regulatory Guide 1.52, Revision 2, except that in lieu of 18 month test intervals, a 24 month interval will be implemented.

The test methods will be In accordance with Regulatory Guide 1.52, Revision 2, except as modified below.

a.

Containment Recirculation Fan Cooler System

1.

Demonstrate the pressure drop across the high efficiency particulate air (HEPA) filter bank is < 3 inches of water at a design flow rate (+/- 10%).

2.

Demonstrate that an In-place dioctylphthalate (DOP) test of the HEPA filter bank shows a penetration and system bypass

< 1.0%.

R.E. Ginna Nuclear Power Plant 5.5-5 Amendment 87

Programs and Manuals 5.5

b.

Control Room Emergency Air Treatment System (CREATS)

1.

Demonstrate the pressure drop across the combined HEPA filters, the prefilters, the charcoal adsorbers and the post-filters is < 11 inches of water at a design flow rate (+/- 10%).

2.

Demonstrate that an in-place DOP test of the HE PA filter bank shows a penetration and system bypass < 0.05%.

3.

Demonstrate that an in-place Freon test of the charcoal adsorber bank shows a penetration and system bypass

< 0.05%, when tested under ambient conditions.

4.

Demonstrate that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows a methyl iodide penetration of less than 1.5% when tested in accorda nce with ASTM D3803-1989 at a test temperature of 300C (860F), a relative humidity of 95%, and a face velocity of 61 ft/min.

c.

SFP Charcoal Adsorber System

1.

Demonstrate that the total airflow rate from the charcoal adsorbers shows at least 75% of that measured with a complete set of rew adsorbers.

2.

Demonstrate that an in-place Freon test of the charcoal adsorbers bank shows a penetration and system bypass

< 1.0%, when tested under ambient conditions.

3.

Demonstrate that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows a methyl Iodide penetration of less than 14.5% when tested in accord ance with ASTM D3803-1989 at a test temperature of 30QC (86 0F) and a relative humidity of 95%.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP frequencies.

5.5.11 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mIxtures contained in the waste gas decay tanks and the quantity of radioactivity contained in waste gas decay tanks. The gaseous radioactivity quantities shall be determined following the methodology in NU REG-0133.

The program shall include:

R.E. Ginna Nuclear Power Plant 5.5-6 Amendment 87

Programs and Manuals 5.5

a.

The limits for concentrations of hydrogen and oxygen in the waste gas decay tanks and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and

b.

A surveillance program to ensure that the quantity of radioactivity contained in each waste gas decay tank is less than the amount that would result in a whole body exposure of 2 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.12 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1.

an API gravity or an absolute specific gravity within limits,

2.

a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3.

a clear and bright appearance with proper color; and

b.

Within 31 days following addition of the new fuel to the storage tanks, verify that the properties of the new fuel oil, other than those addressed in a. above, are within limits for ASTM 2D fuel oil.

5.5.13 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a.

Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

R.E. Ginna Nuclear Power Plant 5.5-7 Amendment 87

Programs and Manuals 5.5

1.

A change in the TS incorporated in the license; or

2.

A change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d.

Proposed changes that meet the criteria of Specification 5.5.13.b.1 or Specification 5.5.13.b.2 shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71e.

5.5.14 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a.

Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

b.

Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c.

Provisions to ensure that an inoperable supported system's Completion lime Is not inappropriately extended as a result of multiple support system inoperabilities; and

d.

Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a.

A required system redundant to the supported system(s) is also inoperable; or

b.

A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or R.E. Ginna Nuclear Power Plant 5.5-8 Amendment 87

.3 Programs and Manuals 5.5

c.

A required system redundant to the inoperable supp5ort system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.15 Containment Leakaae Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 60 psig.

The maximum allowable primary containment leakage rate, La, at Pal shall be 0.2% of containment air weight per day.

Leakage Rate acceptance criteria are:

a.

Containment leakage rate acceptance criterion is

• 1.0 La. During the first plant startup following testing in accordance with this program, the leakage rate acceptance criteria are S 0.60 La for the Type B and Type C tests and

• 0.75 La for Type A tests;

b.

Air lock testing acceptance criteria are:

1.

For each air lock, overall leakage rate is

• 0.05 La when tested at 2 Pa, and

2.

For each door, leakage rate is s 0.01 La when tested at 2 Pa.

c.

Mini-purge valve acceptance criteria is s 0.05 La when tested at 2 Pa.

R.E. Ginna Nuclear Power Plant 5.5-9 Amendment 87

Programs and Manuals 5.5 The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

R.E. Ginna Nuclear Power Plant 5.5-1 0 Amendment 87