ML082490835
| ML082490835 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 09/05/2008 |
| From: | Plant Licensing Branch 1 |
| To: | |
| Shared Package | |
| ML08232057 | List: |
| References | |
| TAC MD6807 | |
| Download: ML082490835 (16) | |
Text
Definitions 1.1 1.1 Definitions (continued)
CHANNEL FUNCTIONAL TEST CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)
DOSE EQUIVALENT 1-131 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarm, interlock, display, and trip functions, and channel failure trips.
The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested.
CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel wi th the vessel head removed and fuel in the vessel.
The following exceptions are not considered to be CORE ALTERATIONS:
- a.
Movement of wide range neutron monitors, local power range monitors, traversing incore probes, or speci a1 movabl e detectors (including undervessel replacement); and
- b.
Control rod movement, provided there are no fuel assemblies in the associated core cell.
Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle.
These cycle specific 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 dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present.
The dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites,"
or Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.
(continued)
PBAPS UNIT 3 1.1-2 Amendment No. 273
Definitions 1.1 1.1 Definitions PHYSICS TESTS (continued)
RATED THERMAL POWER (RTP)
REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME RECENTLY IRRADIATED FUEL SHUTDOWN MARGIN (SDM)
STAGGERED TEST BASIS THERMAL POWER
- b.
Authorized under 10 CFR 50.59; or the provisions of
- c.
Otherwise approve Commission.
d by the Nuclear Regulatory RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3514 MWt.
The RPS RESPONSE TIME shall be that time interval from the opening of the sensor contact up to and including the opening of the trip actuator contacts.
RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
When using this definition to suspend the Applicability of LCOs, secondary containment ground-level hatches H20, H21, H22, H23, H24, and H34 shall be closed during the movement of any irradiated fuel in Secondary Containment.
SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:
- a.
The reactor is xenon free;
- b. The moderator temperature is 68°F; and
- c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.
A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
(continued)
PBAPS UNIT 3 1.1-5 Amendment No. 273
SLC System 3.1. 7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQU I RED ACTI ON COMPLETION TIME A.
Concentration of boron in solution> 9.82%
weight.
A.l AND A.2 Verify the concentration and temperature of boron in solution and pump suction piping temperature are within the limits of Figure 3.1.7-1.
Restore concentration of boron in solution to
~ 9.82% weight.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND 10 days from discovery of failure to meet the LCO B.
One SLC subsystem inoperable for reasons other than Condition A.
B.l Restore SLC sUbsystem to OPERABLE status.
7 days AND 10 days from discovery of failure to meet the LCO (continued)
PBAPS UN IT 3 3.1-20 Amendment No. 273
SLC System 3.1. 7 ACTIONS (continued)
CONDITION REQU I RED ACTI ON COMPLETION TIME C.
Two SLC subsystems inoperable for reasons other than Condition A.
C.1 Restore one SLC subsystem to OPERABLE status.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D.
Required Action and associated Completion Time not met.
D.1 AND D.2 Be in MODE 3.
Be in MODE 4.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify level of sodium penta borate solution in the SLC tank is ~ 46%.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.1.7.2 Verify temperature of sodium penta borate solution is ~ 53°F.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.1.7.3 Verify temperature of pump suction piping i s ~ 53" F.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.1.7.4 Verify continuity of explosive charge.
31 days (continued)
PBAPS UN IT 3 3.1-21 Amendment No.273
Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 3 of 3l Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIREO REFERENCED OTHER CHANNELS FROM SPEC I FlED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION COND IT IONS SYSTEM ACTION C.l REQUIREMENTS VALUE
- 5.
Reactor Water (RWCU) System Cleanup Isolation
- a.
RWCU Flow-High 1,2,3 SR SR SR 3.3.6.1.1 3.3.6.1.3 3.3.6.1.7
$ 125% rated flow (23.0 in-we)
- b.
SLC System I niti at ion 1,2,3 H
SR 3.3.6.1.7 NA
- c.
Reactor Vessel Water Level - Low (Level 3) 1,2,3 SR SR SR SR 3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.7
~ 1.0 inches
- 6.
RHR Shutdown Isolation Cooling System
- a.
Reactor Pressure Hi gh 1, 2,3 SR SR 3.3.6.1.3 3.3.6.1.7
$ 70.0 psig
- b.
Reactor Vessel Water Level - Low (Level 3) 3,4,5 2(a 1 SR SR SR SR 3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.7
~ 1.0 inches
- 7.
Feedwater Isolation Recirculation
- a.
Reactor Pressure Hi gh 1,2,3 SR SR SR SR 3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.7
$ 600 psig
- 8.
Traversing Isolation Incore Probe
- a.
Reactor Vessel Water Level-Low (Level 3) 1,2,3 J
SR SR SR SR 3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.7
~ 1.0 inches
- b.
Drywell Pressure-High 1,2,3 J
SR SR SR SR 3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.7
$ 2.0 psig (a ) In MODES 4 and 5, provided RHR Shutdown Cooling System integrity is maintained. only one channel per trip system with an isolation signal available to one shutdown cooling pump suction isolation valve is required.
PBAPS UN IT 3 3.3-54 Amendment NO.2 73
Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)
Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEI LLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE
- 1.
Reactor Vessel Water 1.2,3.
- 2 1.0 inches Level--Low (Level 3)
(a)
SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.5
- 2. Drywell Pressure--High 1.2,3 SR 3.3.6.2.1 s 2.0 psig SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.5
- 3. Reactor Building 1.2,3.
2 SR 3.3.6.2.1 s 16.0 mR/hr Ventilation Exhaust (a) *(b)
SR 3.3.6.2.3 Radiation--High SR 3.3.6.2.5
- 4. Refueling Floor 1, 2,3.
SR 3.3.6.2.1 s 16.0 mR/hr Ventilation Exhaust (a) *(b)
SR 3.3.6.2.3 Radiation--High SR 3.3.6.2.5 (a)
During operations with a potential for draining the reactor vessel.
(b)
During movement of RECENTLY 1RRADIATED FUEL assemblies in secondary containment.
PBAPS UNIT 3 3.3-58 Amendment No. 273
PCIVs 3.6.1.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME E.
Purge/Vent flowpath open for an accumulated time greater than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> for the calendar year while in MODE 1 or 2 with Reactor Pressure greater than 100 psig.
E.1 OR E.2.1 E.2.2 Isolate the penetration.
Be in MODE 3.
Be in Mode 4.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 12 hours 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> F.
Required Action and associated Completion Time of Condition A, B, C, or D not met in MODE 1, 2, or 3.
F.1 AND F.2 Be Be in in MODE MODE
- 3.
- 4.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours G.
Required Action and associated Completion Time of Condition A, B, C, or D not met for PCIV(s) required to be OPERABLE during MODE 4 or 5.
G.1 OR G.2 Initiate action to suspend operations with a potential for draining the reactor vessel.
Initiate action to restore valve(s) to OPERABLE status.
Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.6.1.3.1 Verify Containment Atmospheric Dilution (CAD) System liquid nitrogen storage tank level is ~ 16 inches water column.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.6.1.3.2 Verify Safety Grade Instrument Gas (SGIG)
System header pressure is ~ 80 psig.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (continued)
PBAPS UNIT 3 3.6-12 Amendment No.2 73
PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.1.3.14 SURVEILLANCE Verify combined MSIV leakage rate for all four main steam lines is ~ 204 scfh, and
~ 116 scfh for anyone steam line, when tested at ~ 25 psig.
FREQUENCY In accordance with the Primary Containment Leakage Rate Testing Program SR 3.6.1.3.15 Verify each 6 inch and 18 inch primary containment purge valve and each 18 inch primary containment exhaust valve is blocked to restrict opening greater than the required maximum opening angle.
24 months SR 3.6.1.3.16 Replace the inflatable seal of each 6 inch and 18 inch primary containment purge valve and each 18 inch primary containment exhaust valve.
96 months PBAPS UN IT 3 3.6-16 Amendment No. 273
Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.
APPLICABI LITY:
MODES 1,2, and 3, During movement of RECENTLY IRRADIATED FUEL assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Secondary containment inoperable in MODE 1, 2, or 3.
A.1 Restore secondary containment to OPERABLE status.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B.
Required Action and associated Completion Time of Condition A not met.
B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C.
Secondary containment inoperable during movement of RECENTLY IRRADIATED FUEL assemblies in the secondary containment or during OPDRVs.
C.1
- - - - - -- -NOT E- - -- - - -
LCO 3.0.3 is not applicable.
Suspend movement of RECENTLY IRRADIATED FUEL assemblies in the secondary containment.
Immediately (continued)
PBAPS UNIT 3 3.6-34 Amendment No.2 73
Secondary Containment 3.6.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.
(continued)
C.2 Initiate action suspend OPDRVs.
to Immediately SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.6.4.1.1 Verify all secondary containment equipment hatches are closed and sealed.
31 days SR 3.6.4.1.2 Verify one secondary containment access door in each access opening is closed.
31 days SR 3.6.4.1.3 Verify secondary containment can be drawn down to
~ 0.25 inch of vacuum water gauge in ~ 180 seconds using one standby gas treatment (SGT) subsystem.
24 months on a STAGGERED TEST BASIS for each subsystem SR 3.6.4.1.4 Verify the secondary containment can be maintained ~ 0.25 inch of vacuum water gauge for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using one SGT subsystem at a flow rate ~ 10,500 cfm.
24 months on a STAGGERED TEST BASIS for each subsystem PBAPS UNIT 3 3.6-35 Amendment No. 273
SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)
LCO 3.6.4.2 Each SCIV shall be OPERABLE.
APPLICABI LITY:
MODES 1, 2, and 3, During movement of RECENTLY IRRADIATED FUEL assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS
--- - - - - - - - - - -- - - - -- - - - - -- -- - -- - - - - - --NOTES -- - - - -- - - - - - - - - -- -- - - - -- - - - - - -- - - - -
- 1. Penetration flow paths may be unisolated intermittently under administrative controls.
- 2. Separate Condition entry is allowed for each penetration flow path.
- 3. Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more penetration flow paths with one SCIV inoperable.
A.l Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (continued)
PBAPS UNIT 3 3.6-36 Amendment No.273
SCIVs 3.6.4.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D.
Required Action and associated Completion Time of Condition A or B not met during movement of RECENTLY IRRADIATED FUEL assemblies in the secondary containment or during OPDRVs.
0.1 0.2
-- - - - -- -NOTE - - -- - - -
LCO 3.0.3 is not applicable.
Suspend movement of RECENTLY IRRADIATED FUEL assemblies in the secondary containment.
Initiate action to suspend OPDRVs.
Immediately Immediately PBAPS UN IT 3 3.6-38 Amendment No. 273
SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.
APPLICABI LITY:
MODES 1, 2, and 3, During movement of RECENTLY IRRADIATED FUEL assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One SGT subsystem inoperable.
A.1 Restore SGT subsystem to OPERABLE status.
7 days B.
Required Action and associated Completion Time of Condition A not met in MODE 1, 2, or 3.
B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C.
Required Action and associated Completion Time of Condition A not met during movement of RECENTLY IRRADIATED FUEL assemblies in the secondary containment or during OPDRVs.
- - - - - - - -- - - -NOT E- -- - - - - - - - - -
LCO 3.0.3 is not applicable.
C.1 Place OPERABLE SGT subsystem in operation.
Immediately (continued)
PBAPS L1NIT 3 3.6-40 Amendment No.273
SGT System 3.6.4.3 ACT IONS CONDITION REQU IRED ACTI ON COMPLETION TIME C.
(continued)
C. 2.1 C.2.2 Suspend movement of RECENTLY IRRADIATED FUEL assemblies in secondary containment.
Initiate action to suspend OPDRVs.
Immediately Immediately D.
Two SGT subsystems inoperable in MODE 1, 2, or 3.
0.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> E.
Two SGT subsystems inoperable during movement of RECENTLY I RRADIATED FUEL assemblies in the secondary containment or during OPDRVs.
E.1
NOTE--------
LCO 3.0.3 is not applicable.
Suspend movement of RECENTLY IRRADIATED FUEL assemblies in secondary containment.
Immediately E.2 Initiate action to suspend OPDRVs.
Immediately PBAPS UN IT 3 3.6-41 Amendment No. 273
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)
- 1.
A required system redundant to system(s) supported by the inoperable support system is also inoperable; or
- 2.
A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or
- 3.
A required system redundant to support system(s) for the supported systems (b.1) and (b.2) above is also inoperable.
- c.
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.
5.5.12 Primary Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) 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, as modified by the following exceptions to NEI 94-01, Rev. 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J":
- a.
Section 10.2:
MSIV leakage is excluded from the combined tot a1 of 0.6 La for the Type Band C tests.
- b.
Section 9.2.3:
The first Type A test performed after the December, 1991 Type A test sha 11 be performed no 1ater than December, 2006.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 49.1 psig.
The maximum allowable primary containment leakage rate, La' at POl shall be 0.7% of primary containment air weight per day.
Leakage Rate acceptance criteria are:
- a.
Primary Containment leakage rate acceptance criterion is ~
1.0 La.
During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are ~ 0.60 La for the Type B and Type C tests and ~
0.75 La for Type A tests; (continued)
PBAPS UNIT 3 5.0-17 Amendment No. 273
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 5.5.13 Primary Containment Leakage Rate Testing Program (continued)
- b.
Air lock testing acceptance criteria are:
- 1)
Overall air lock leakage rate is s 9000 scc/min when tested at ~ Pa *
- c.
MSIV leakage acceptance criteria are as specified in SR 3.6.1.3.14.
The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.
The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.
Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Main Control Room Emergency Ventilation (MCREV) System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release as applicable, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditio~s without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident.
The program shall include the following elements:
- a.
The definition of the CRE and the CRE boundary.
- b.
Requirements for maintaining the CRE boundary in its design condition including configuration control and preventative maintenance.
- c.
Requirements of (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Section C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Section C.1 and C.2 of Regulatory Guide 1.197, Revision O.
(continued)