View source for BSEP 02-0093, Part 2 of 2 - Brunswick, Units 1 and 2 - Submittal of Technical Specification Bases Changes for Revisions 21 and 22 (Units 1 and 2) and Revision 23 (Unit 1 Only)

{{Adams
| number = ML021430051
| issue date = 05/08/2002
| title = Part 2 of 2 - Brunswick, Units 1 and 2 - Submittal of Technical Specification Bases Changes for Revisions 21 and 22 (Units 1 and 2) and Revision 23 (Unit 1 Only)
| author name = O'Neil E
| author affiliation = Carolina Power &amp; Light Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000324, 05000325
| license number = DPR-062, DPR-071, NPF-037, NPF-066
| contact person = 
| case reference number = BSEP 02-0093
| document type = Technical Specification, Bases Change
| page count = 117
}}

=Text=
{{#Wiki_filter:Secondary Containment Isolation Instrumentation B 3.3.6.2 B 3.3   INSTRUMENTATION B 3.3.6.2   Secondary Containment Isolation Instrumentation BASES BACKGROUND         The secondary containment isolation instrumentation automatically initiates closure of appropriate secondary containment isolation dampers (SCIDs) and starts the Standby Gas Treatment (SGT) System. The function of these systems, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Refs. 1, 2, and 3). Secondary containment isolation and establishment of vacuum with the SGT System ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be OPERABLE are maintained within applicable limits.
The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of secondary containment isolation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a secondary containment isolation signal to the isolation logic. Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logic are (1) reactor vessel water level, (2) drywell pressure, and (3) reactor building exhaust radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation.
The outputs of the channels associated with the Reactor Vessel-Water Level-Low Level 2 Function and the Drywell Pressure-High Function are arranged in one-out-of-two taken twice trip system logics. One trip system initiates isolation of one automatic secondary containment isolation damper in each penetration and starts both SGT subsystems while the other trip system initiates isolation of the other automatic secondary containment isolation damper in each penetration and starts both SGT subsystems. Each logic closes one of the two dampers in each penetration and starts both SGT subsystems, so that operation of either logic isolates the secondary containment and provides for the necessary filtration of fission products.
(continued)
Brunswick Unit I                     B 3.3-175                  Revision No. 21    I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES BACKGROUND        The outputs of the channels associated with the Reactor (continued)    Building Exhaust Radiation-High Function are arranged in two one-out-of-one trip system logics. Each trip system initiates isolation of one automatic secondary containment isolation damper in each penetration and starts both SGT subsystems while the other trip system initiates isolation of the other secondary containment isolation damper in each penetration and starts both SGT subsystems. Each logic closes one of the two dampers in each penetration and starts both SGT subsystems, so that operation of either logic isolates the secondary containment and provides for the necessary filtration of fission products.
APPLICABLE        The isolation signals generated by the secondary containment SAFETY ANALYSES,   isolation instrumentation are implicitly assumed in the LCO, and         safety analyses of References I and 3 to initiate APPLICABILITY     closure of dampers and start the SGT System to limit offsite doses.
Refer to LCO 3.6.4.2, "Secondary Containment Isolation Dampers (SCIDs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System," Applicable Safety Analyses Bases for more detail of the safety analyses.
The secondary containment isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).
Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.
The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have the required number of OPERABLE channels with their setpoints set within the specified Allowable Values, as shown in Table 3.3.6.2-1. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
Allowable Values are specified for each Function specified in the Table. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less (continued)
Brunswick Unit I                    B 3.3-176                  Revision No. 22   I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE        conservative than the trip setpoint, but within its SAFETY ANALYSES,  Allowable Value, is acceptable. A channel is inoperable if LCO, and          its actual trip setting is not within its required Allowable APPLICABILITY     Value.
(continued)
Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors.
The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors.
These calibration based errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components.
The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.
In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions when SCIDs and the SGT System are required.
The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.
: 1. Reactor Vessel Water Level-Low Level 2 Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened.      Should RPV water level decrease too far, fuel damage could result.
An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The Reactor Vessel (continued)
Brunswick Unit I                    B 3.3-177                  Revision No. 21    I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE        1. Reactor Vessel Water Level-Low Level 2     (continued)
SAFETY ANALYSES, LCO, and         Water Level--Low Level 2 Function is one of the Functions APPLICABILITY    assumed to be OPERABLE and capable of providing isolation and initiation signals. The isolation and initiation systems on Reactor Vessel Water Level-Low Level 2 support actions to ensure that any offsite releases are within the limits calculated in the safety analysis.
Reactor Vessel Water Level-Low Level 2 signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Level 2 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
The Reactor Vessel Water Level-Low Level 2 Allowable Value was chosen to be the same at the High Pressure Coolant Injection/Reactor Core Isolation Cooling (HPCI/RCIC) Reactor Vessel Water Level-Low Level 2 Allowable Value (LCO 3.3.5.1 and LCO 3.3.5.2), since this could indicate that the capability to cool the fuel is being threatened. The Allowable Value is referenced from reference level zero.
Reference level zero is 367 inches above the vessel zero point.
The Reactor Vessel Water Level-Low Level 2 Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, this Function is not required.
: 2. Drywell Pressure-High High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite (continued)
Brunswick Unit I                  B 3.3-178                  Revision No. 21     I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE        2. Drywell Pressure-High    (continued)
SAFETY ANALYSES, LCO, and         dose release. The isolation on high drywell pressure APPLICABILITY     supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis.
However, the Drywell Pressure-High Function associated with isolation is not assumed in any UFSAR accident or transient analyses. It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis.
High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation function.
The Allowable Value was chosen to be the same as the ECCS Drywell Pressure-High Function Allowable Value (LCO 3.3.5.1) since this is indicative of a loss of coolant accident (LOCA).
The Drywell Pressure-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.
: 3. Reactor Buildina Exhaust Radiation-Hiqh High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding.
The release may have originated from the primary containment due to a break in the RCPB or the refueling floor due to a fuel handling accident. When Reactor Building Exhaust Radiation-High is detected, secondary containment isolation and actuation of the SGT System are initiated to limit the release of fission products.
(contnued (continued)
Brunswick Unit I                   B 3.3-179                  Revision No. 22     I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE       3. Reactor Building Exhaust Radiation-High     (continued)
SAFETY ANALYSES, LCO, and         The Reactor Building Exhaust Radiation-High signals are APPLICABILITY    initiated from radiation detectors that are located in the ventilation exhaust ductwork plenum coming from the reactor building. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Two channels of Reactor Building Exhaust Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
The Allowable Values are chosen to promptly detect gross failure of the fuel cladding.
The Reactor Building Exhaust Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, the Function is not required. In addition, the Function is also required to be OPERABLE during OPDRVs and movement of recently irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation I
releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded. Due to radioactive decay, this function is only required   to isolate secondary containment during fuel handling  accidents  involving handling recently irradiated fuel (i.e.,   fuel that has occupied part of a critical  reactor  core within the previous 24 hours).
ACTIONS          A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels.
Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the (conti nued)
Brunswick Unit I                  B 3.3-180                   Revision No. 22     I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS          Condition. However, the Required Actions for inoperable (continued)       secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel.
A.I Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Function 2, and 24 hours for Functions other than Function 2, has been shown to be acceptable (Refs. 5 and 6) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases).
If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.I. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an isolation), Condition C must be entered and its Required Actions taken.
B.1 Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic isolation capability for the associated penetration flow path(s) or a complete loss of automatic initiation capability for the SGT System. A Function is considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from the given Function on a valid signal.      This ensures that one of the two SCIDs in the associated penetration flow path and both SGT subsystems can be initiated on an isolation signal from (continued)
Brunswick Unit I                    B 3.3-181                   Revision No. 22     I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS            B.1  (continued) the given Function. For the Functions with two one-out-of-two logic trip systems (Functions I and 2), this would require one trip system to have one channel OPERABLE or in trip. An inoperable channel need not be placed in the tripped condition where this would cause the trip Function to occur. In these cases, if the inoperable channel is not restored within the required Completion Time, Condition C shall be entered.
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
C.1.1, C.1.2, C.2.1,     and C.2.2 If any Required Action and associated Completion Time of Condition A or B are not met, the ability to isolate the secondary containment and start the SGT System cannot be ensured. Therefore, further actions must be performed to ensure the ability to maintain the secondary containment function. Isolating the associated dampers and starting the associated SGT subsystem (Required Actions C.1.1 and C.2.1) performs the intended function of the instrumentation and allows operation to continue.
Alternately, declaring the associated SCIDs or SGT subsystem(s) inoperable (Required Actions C.1.2 and C.2.2) is also acceptable since the Required Actions of the respective LCOs (LCO 3.6.4.2 and LCO 3.6.4.3) provide appropriate actions for the inoperable components.
One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components inoperable without unnecessarily challenging plant systems.
SURVEILLANCE     As noted at the beginning of the SRs, the SRs for each REQUIREMENTS     Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1.
(continued)
Brunswick Unit I                     B 3.3-182                  Revision No. 22     I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE     The Surveillances are modified by a Note to indicate that REQUIREMENTS     when a channel is placed in an inoperable status solely for (continued)     performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows:
(a) for up to 2 hours for Function 3 and (b) for up to 6 hours for Functions I and 2 provided the associated Function maintains isolation capability. Upon completion of the Surveillance, or expiration of the 2 hour allowance for Function 3 or the 6 hour allowance for Functions I and 2, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
This Note is based on the reliability analysis (Refs. 5 and 6) assumption of the average time required to perform channel surveillance. That analysis demonstrated the 2 hour testing allowance for Function 3 and the 6 hour testing allowance for Functions 1 and 2 do not significantly reduce the probability that the SCIDs will isolate the associated penetration flow paths and that the SGT System will initiate when necessary.
SR   3.3.6.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of (continued)
Brunswick Unit I                   B 3.3-183                  Revision No. 22     I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE      SR  3.3.6.2.1   (continued)
REQUIREMENTS channel status during normal operational use of the displays associated with channels required by the LCO.
SR  3.3.6.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Frequency of 92 days is based on the reliability analysis of References 5 and 6.
SR   3.3.6.2.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.2-1.      If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
The Frequency of 92 days is based on the reliability analysis of References 5 and 6.
SR   3.3.6.2.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
(continued)
Brunswick Unit I                   B 3.3-184                  Revision No. 22     I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE     SR  3.3.6.2.4    (continued)
REQUIREMENTS The Frequency is based on the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
SR  3.3.6.2.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed on SCIDs and the SGT System in LCO 3.6.4.2 and LCO 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function.
The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency.
REFERENCES       1. UFSAR, Section 15.6.4.
: 2. Not used.
: 3. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
: 4. 10 CFR 50.36(c)(2)(ii).
: 5. NEDC-31677P-A, Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation, July 1990.
: 6. NEDC-30851P-A Supplement 2, Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
Brunswick Unit I                    B 3.3-185                  Revision No. 22    I

CREV System Instrumentation B 3.3.7.1 B 3.3   INSTRUMENTATION B 3.3.7.1   Control Room Emergency Ventilation (CREV)   System Instrumentation BASES BACKGROUND         The CREV System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions. Two independent CREV subsystems are each capable of fulfilling the stated safety function. The instrumentation and controls for the CREV System automatically initiate action to pressurize the main control room (MCR) to minimize the consequences of radioactive material in the control room environment.
In the event of a Control Building Air Intake Radiation-High signal, the CREV System is automatically started in the radiation/smoke protection mode. Air is then recirculated through the charcoal filter, and sufficient outside air is drawn in through the normal intake to maintain the MCR slightly pressurized with respect to outside atmosphere.
The CREV System instrumentation has two trip systems, either of which can initiate the CREV System. Each trip system receives input from the two Control Building Air Intake Radiation-High Function channels. The Control Building Air Intake Radiation-High Function is arranged in a one-out-of-two logic for each trip system. The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a CREV System initiation signal to the initiation logic.
APPLICABLE         The ability of the CREV System to maintain the habitability SAFETY ANALYSES,   of the MCR is explicitly assumed for the design basis LCO, and           accident as discussed in the UFSAR safety analyses APPLICABILITY      (Refs. I and 2). CREV System operation ensures that the          I radiation exposure of control room personnel, through the duration of any one of the postulated accidents, does not exceed the limits set by GDC 19 of 10 CFR 50, Appendix A.
CREV System instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 3).
(contnued (continued)
Brunswick Unit I                    B 3.3-186                   Revision No. 22     I

CREV System Instrumentation B 3.3.7.1 BASES APPLICABLE         The OPERABILITY of the CREV System instrumentation is SAFETY ANALYSES,   dependent upon the OPERABILITY of the Control Building Air LCO, and           Intake Radiation-High instrumentation channel Function.
APPLICABILITY      The Function must have a required number of OPERABLE (continued)      channels, with their setpoints within the specified Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
Allowable Values are specified for Control Building Air Intake Radiation-High Function. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS.
Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value.      Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., control building air intake radiation), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors.
The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors.
These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.
The control building air intake radiation monitors measure radiation levels in the control building air intake plenum.
A high radiation level may pose a threat to MCR personnel; thus, automatically initiating the CREV System.
(continued)
Brunswick Unit I                    B 3.3-187                   Revision No. 21     1

CREV System Instrumentation B 3.3.7.1 BASES APPLICABLE        The Control Building Air Intake Radiation-High Function SAFETY ANALYSES,  consists of two independent monitors. Two channels per trip LCO, and          system of Control Building Air Intake Radiation-High are APPLICABILITY     available and are required to be OPERABLE to ensure that no (continued)     single instrument failure can preclude CREV System initiation. The Allowable Value was selected to ensure protection of the control room personnel.
The Control Building Air Intake Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 and during OPDRVs and movement or recently irradiated fuel assemblies in the secondary containment, to ensure that control room personnel are protected during a LOCA, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., OPDRVs), the probability of a LOCA, main steam line break accident, or control rod drop accident is low; thus, the Function is not required. Also due to radioactive decay, this Function is only required to initiate the CREV System during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours).
ACTIONS           A Note has been provided to modify the ACTIONS related to CREV System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable CREV System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable CREV System instrumentation channel.
A.1 Because of the redundancy of sensors available to provide initiation signals and the redundancy of the CREV System design, an allowable out of service time of 7 days is provided to permit restoration of any inoperable channel to (continued)
Brunswick Unit I                    B 3.3-188                   Revision No. 22     I

CREV System Instrumentation B 3.3.7.1 BASES ACTIONS          A.1    (continued)
OPERABLE status. This out of service time is only acceptable provided the Control Building Air Intake Radiation-High Function is still maintaining CREV System initiation capability (refer to Required Action B.1 Bases).
If the Function is not maintaining CREV System initiation capability, Condition B must be entered.
If the inoperable channel cannot be restored to OPERABLE status within the 7 day allowable out of service time, one CREV subsystem must be placed in the radiation/smoke protection mode of operation per Required Action A.1. The method used to place the CREV subsystem in operation must provide for automatically re-initiating the subsystem upon restoration of power following a loss of power to the CREV subsystem. Placing one CREV subsystem in the radiation/smoke protection mode of operation provides a suitable compensatory action to ensure that the automatic radiation protection function of the CREV System is not lost.
B.1 Required Action B.I is intended to ensure that appropriate action is taken if multiple, inoperable, untripped channels result in the Control Building Air Intake Radiation-High Function not maintaining CREV System initiation capability.
The Function is considered to be maintaining CREV System initiation capability when sufficient channels are OPERABLE or in trip such that one trip system will generate an initiation signal for one CREV subsystem from the Function on a valid signal. For the Control Building Air Intake Radiation-High Function, this would require one trip system to have one channel OPERABLE or in trip. With CREV System initiation capability not maintained, one CREV subsystem must be placed in the radiation/smoke protection mode of operation per Required Action B.1 to ensure that control room personnel will be protected in the event of a Design Basis Accident. The method used to place the CREV subsystem in operation must provide for automatically re-initiating the subsystem upon restoration of power following a loss of power to the CREV subsystem.
(continued)
Brunswick Unit I                    B 3.3-189                Revision No. 22    1

CREV System Instrumentation B 3.3.7.1 BASES ACTIONS           B.1   (continued)
The 1 hour Completion Time is intended to allow the operator time to place the CREV subsystem in operation. The I hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels, or for placing one CREV subsystem in operation.
SURVEILLANCE     The Surveillances are modified by a Note to indicate that REQUIREMENTS     when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains CREV System initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
This Note is based on the reliability analysis (Ref. 4)           I assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the CREV System will initiate when necessary.
SR   3.3.7.1.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
(continued)
Brunswick Unit I                     B 3.3-190                 Revision No. 22     I

CREV System Instrumentation B 3.3.7.1 BASES SURVEILLANCE     SR  3.3.7.1.1   (continued)
REQUIREMENTS The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with channels required by the LCO.
SR  3.3.7.1.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Frequency of 92 days is based on the reliability analyses of Reference 4.
SR   3.3.7.1.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
SR   3.3.7.1.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.7.3, "Control Room Emergency Ventilation (CREV)
System," overlaps this Surveillance to provide complete testing of the assumed safety function.
(continued)
Brunswick Unit I                   B 3.3-191                  Revision No. 22     I

CREV System Instrumentation B 3.3.7.1 BASES SURVIELLANCE     SR 3.3.7.1.4 (continued)
REQUIREMENTS While this surveillance can be performed with the reactor at power, operating experience has demonstrated these components will usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was found to be acceptable from a reliability standpoint.
REFERENCES       1. UFSAR, Section 15.6.4.5.5.
: 2. UFSAR, Section 15.7.1
: 3. 10 CFR 50.36(c)(2)(ii).
: 4. GENE-770-06-1-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, December 1992.
Brunswick Unit I                   B 3.3-192                  Revision No. 22   I

Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 B 3.3   INSTRUMENTATION B 3.3.7.2    Condenser Vacuum Pump Isolation Instrumentation BASES BACKGROUND          The condenser vacuum pump isolation instrumentation initiates a trip of the respective condenser vacuum pump and isolation of the common isolation valve following events in which main steam radiation monitors exceed a predetermined value. Tripping and isolating the condenser vacuum pumps limits control room doses in the event of a control rod drop accident (CRDA).
The condenser vacuum pump isolation instrumentation includes sensors, relays and switches that are-necessary to cause initiation of condenser vacuum pump isolation. The channels include electronic equipment that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an isolation signal to the condenser vacuum pump isolation logic.
The isolation logic consists of two independent trip systems, with two channels of the Main Steam Line Radiation-High Function in each trip system. Each trip system is a one-out-of-two logic for this Function. Thus, either channel of the Main Steam Line Radiation-High Function in each trip system are needed to trip a trip system. The outputs of the channels in a trip system are arranged in a logic so that both trip systems must trip to result in an isolation signal.
There are two condenser vacuum pumps and one isolation valve associated with this function.
APPLICABLE          The condenser vacuum pump isolation is assumed in the safety SAFETY ANALYSES     analysis for the CRDA. The condenser vacuum pump isolation instrumentation initiates an isolation of the condenser vacuum pumps to limit control room doses resulting from fuel cladding failure in a CRDA.
The condenser vacuum pump isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 1).
(continued)
Brunswick Unit I                     B 3.3-193                  Revision No. 21    I

Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES   (continued)
LCO                 The OPERABILITY of the condenser vacuum pump isolation instrumentation is dependent on the OPERABILITY of the individual Main Steam Line Radiation-High Function instrumentation channels, which must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 3.3.7.2.3. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Channel OPERABILITY also includes the condenser vacuum pump trip breakers and isolation valve.
Allowable Values are specified for the condenser vacuum pump isolation Function specified in the LCO. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the actual trip settings do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value.
Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (i.e., main steam line radiation), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors.
The Allowable Values are then determined, based on the trip setpoint values, by accounting for the calibration based errors. These calibration based errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerances of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.
(continued)
Brunswick Unit 1                     B 3.3-194                  Revision No. 21    I

Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES   (continued)
APPLICABILITY       The condenser vacuum pump isolation is required to be OPERABLE in MODES I and 2 when the condenser vacuum pump is in service to mitigate the consequences of a postulated CRDA. In this condition, fission products released during a CRDA could be discharged directly to the environment.
Therefore, the condenser vacuum pump isolation is necessary to assure conformance with the radiological evaluation of the CRDA. In MODE 3, 4 or 5 the consequences of a control rod drop are insignificant, and are not expected to result in any fuel damage or fission product releases. When the condenser vacuum pump is not in operation in MODE 1 or 2, fission product releases via this pathway would not occur.
ACTIONS             A Note has been provided to modify the ACTIONS related to condenser vacuum pump isolation instrumentation channels.
Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable condenser vacuum pump isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable condenser vacuum pump isolation instrumentation channel.
A.1 and A.2 With one or more channels inoperable, but with condenser vacuum pump isolation capability maintained (refer to Required Actions B.1, B.2, and B.3 Bases), the condenser vacuum pump isolation instrumentation is capable of performing the intended function. However, the reliability and redundancy of the condenser vacuum pump isolation instrumentation is reduced, such that a single failure in one of the remaining channels could result in the inability of the condenser vacuum pump isolation instrumentation to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to (continued)
Brunswick Unit 1                     B 3.3-195                  Revision No. 21    I

Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES ACTIONS          A.1 and A.2   (continued)
OPERABLE status. Because of the low probability of extensive number of inoperabilities affecting multiple channels, and the low probability of an event requiring the initiation of condenser vacuum pump isolation, 12 hours has been shown to be acceptable (Ref. 2) to permit restoration of any inoperable channel to OPERABLE status (Required Action A.1). Alternately, the inoperable channel, or associated trip system, may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable condenser vacuum pump trip breaker or isolation valve, since this may not adequately compensate for the inoperable condenser vacuum pump trip breaker or isolation valve (e.g., the trip breaker may be inoperable such that it will not trip). If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in loss of condenser vacuum), or if the inoperable channel is the result of an inoperable condenser vacuum pump trip breaker or isolation valve, Condition B must be entered and its Required Actions taken.
B.1, B.2, and B.3 With any Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE or other specified condition in which the LCO does not apply.
To achieve this status, the plant must be brought to at least MODE 3 with 12 hours (Required Action B.3).
Alternately, the associated condenser vacuum pumps may be removed from service since this performs the intended function of the instrumentation (Required Action B.1). An additional option is provided to isolate the main steam lines (Required Action B.2), which may allow operation to continue. Isolating the main steam lines effectively provides an equivalent level of protection by precluding fission product transport to the condenser. This isolation is accomplished by isolation of all main steam lines and main steam line drains which bypass the main steam isolation valves.
(continued)
Brunswick Unit 1                  B 3.3-196                   Revision No. 21    I

Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES ACTIONS          B.1, B.2, and B.3    (continued)
Condition B is also intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels result in the Function not maintaining condenser vacuum pump isolation capability. The Function is considered to be maintaining condenser vacuum pump isolation capability when sufficient channels are OPERABLE or in trip such that the condenser vacuum pump isolation instruments will generate a trip signal from a valid Main Steam Line Radiation-High signal, and the condenser vacuum pumps will trip. This requires one channel of the Function in each trip system to be OPERABLE or in trip, and the condenser vacuum pump trip breakers to be OPERABLE.
SURVEILLANCE     The Surveillances are modified by a Note to indicate that REQUIREMENTS     when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains condenser vacuum pump isolation trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 2) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the condenser vacuum pumps will isolate when necessary.
SR   3.3.7.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of (continued)
Brunswick Unit 1                   B 3.3-197                   Revision No. 21   I

Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES SURVEILLANCE     SR 3.3.7.2.1 (continued)
REQU IREMENTS    excessive instrument drift in one of the channels or something even   more serious. A CHANNEL  CHECK will   detect gross channel   failure; thus,  it is key  to verifying   the instrumentation   continues  to operate properly   between  each CHANNEL  CALIBRATION.
Agreement criteria are determined by the plant staff based on combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
The Frequency is based on the CHANNEL CHECK Frequency requirement of other instrumentation. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO.
SR   3.3.7.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Frequency of 92 days is based on the reliability analysis of Reference 2.
SR   3.3.7.2.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
The Frequency is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
(continued)
B 3.3-198                    Revision No. 21    I Brunswick Unit 1

Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES SURVEILLANCE     SR  3.3.7.2.3    (continued)
REQUIREMENTS For the purposes of this SR, background  is the dose level experienced at 100% RATED THERMAL POWER  with hydrogen water chemistry at the maximum injection rate. Under these conditions, an Allowable Value of
* 6 x  background will ensure that General Design Criterion 19  limits of 10 CFR 50, Appendix A, will not be exceeded in the   control room in the event of a CRDA.
SR 3.3.7.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers and actuation of the associated isolation valve are included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function. Therefore, if a breaker is incapable of operating or the isolation valve is incapable of actuating, the instrument channel would be inoperable.
The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
REFERENCES       1. 10 CFR 50.36(c)(2)(ii).
: 2. NEDC-30851P-A, Supplement 2, Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
Brunswick Unit I                   B 3.3-199                  Revision No. 21    I

LOP Instrumentation B 3.3.8.1 B 3.3   INSTRUMENTATION B 3.3.8.1    Loss of Power (LOP)   Instrumentation BASES BACKGROUND           Successful operation of the required safety functions of the Emergency Core Cooling Systems (ECCS) is dependent upon the availability of adequate power sources for energizing the various components such as pump motors, motor operated valves, and the associated control components. The LOP instrumentation monitors the 4.16 kV emergency buses.
Offsite power is the preferred source of power for the 4.16 kV emergency buses. If the monitors determine that insufficient power is available, the buses are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources.
Each 4.16 kV emergency bus has its own independent LOP instrumentation and associated trip logic. The voltage for each bus is monitored at two levels, which can be considered as two different undervoltage Functions: Loss of Voltage and 4.16 kV Emergency Bus Undervoltage Degraded Voltage.
Each Function causes various bus transfers and disconnects.
The Loss of Voltage Function is monitored by one inverse time delay undervoltage relay (27/59E) and the Degraded Voltage Function is monitored by three definite time undervoltage relays (27DVA, 27DVB, and 27DVC) for each emergency bus. The Loss of Voltage Function is a one-out-of-one logic configuration and the Degraded Voltage Function output is arranged as a two-out-of-three logic configuration. The channels include electronic equipment (e~g., internal relay contacts, coils, etc.) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a LOP trip signal to the trip logic.
APPLICABLE          The LOP instrumentation is required for Engineered Safety SAFETY ANALYSES,    Features to function in any accident with a loss of offsite LCO, and            power. The required channels of LOP instrumentation ensure APPLICABILITY       that the ECCS and other assumed systems powered from the DGs, provide plant protection in the event of any of the Reference I and 2 analyzed accidents in which a loss of offsite power is assumed. The initiation of the DGs on loss of offsite power, and subsequent initiation of the ECCS, ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
(continued)
Brunswick Unit I                      B 3.3-200                  Revision No. 21    I

LOP Instrumentation B 3.3.8.1 BASES APPLICABLE       Accident analyses credit the loading of the DG based on the SAFETY ANALYSES,  loss of offsite power during a loss of coolant accident.
LCO, and          The diesel starting and loading times have been included in APPLICABILITY     the delay time associated with each safety system component (continued)    requiring DG supplied power following a loss of offsite power.
The LOP instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 3).
The OPERABILITY of the LOP instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.8.1-1. Each Function must have a required number of OPERABLE channels per 4.16 kV emergency bus, with their setpoints within the specified Allowable Values. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
The Allowable Values are specified for each Function in the Table. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within the Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints.
are compared to the actual process parameter (e.g., degraded voltage), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g.,
trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for (continued)
Brunswick Unit 1                   B 3.3-201                  Revision No. 21    I

LOP Instrumentation B 3.3.8.1 BASES APPLICABLE       channels that must function in harsh environments as defined SAFETY ANALYSES,  by 10 CFR 50.49) are accounted for and appropriately applied LCO, and          for the instrumentation.
APPLICABILITY (continued)   The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.
: 1. 4.16 kV Emergency Bus Undervoltaqe (Loss of Voltage)
Loss of voltage on a 4.16 kV emergency bus indicates that offsite power may be completely lost to the respective emergency bus and is unable to supply sufficient power for proper operation of the applicable equipment. Therefore, the power supply to the bus is transferred from offsite power to DG power when the voltage on the bus drops below the Loss of Voltage Function Allowable Values (loss of voltage with a short time delay). This ensures that adequate power will be available to the required equipment.
The Bus Undervoltage Allowable Values are low enough to prevent inadvertent power supply transfer, but high enough to ensure that power is available to the required equipment.
The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to normal voltages, but short enough to ensure that power is available to the required equipment.
One channel of 4.16 kV Emergency Bus Undervoltage (Loss of Voltage) Function per associated emergency bus is only required to be OPERABLE when the associated DG is required to be OPERABLE to ensure that no single instrument failure can preclude the start of three of the four DGs. (One channel inputs to each of the four DGs.)   Refer to LCO 3.8.1, "AC Sources-Operating," and 3.8.2, "AC Sources-Shutdown," for Applicability Bases for the DGs.
: 2. 4.16 kV Emergency Bus Undervoltage (Degraded Voltage)
A reduced voltage condition on a 4.16 kV emergency bus indicates that, while offsite power may not be completely lost to the respective emergency bus, available power may be (continued)
Brunswick Unit 1                  B 3.3-202                  Revision No. 21    I

LOP Instrumentation B 3.3.8.1 BASES APPLICABLE        2. 4.16 kV Emergency Bus Undervoltaqe (Degraded Voltage)
SAFETY ANALYSES,  (continued)
LCO, and APPLICABILITY     insufficient for starting large ECCS motors without risking damage to the motors that could disable the ECCS function.
Therefore, the power supply to the bus is transferred from offsite power to onsite DG power when the voltage on the bus drops below the Degraded Voltage Function Allowable Values (degraded voltage with a time delay). This ensures that adequate power will be available to the required equipment.
The Bus Undervoltage Allowable Values are low enough to prevent inadvertent power supply transfer, but high enough to ensure that sufficient power is available to the required equipment. The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to normal voltages, but short enough to ensure that sufficient power is available to the required equipment.
Three channels of 4.16 kV Emergency Bus Undervoltage (Degraded Voltage) Function per associated bus are only required to be OPERABLE when the associated DG is required to be OPERABLE to ensure that no single instrument failure can preclude the DG function. (Three channels input to each of the four emergency buses and DGs.)    Refer to LCO 3.8.1 and LCO 3.8.2 for Applicability Bases for the DGs.
ACTIONS          A Note has been provided to modify the ACTIONS related to LOP instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable LOP instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable LOP instrumentation channel.
(continued)
Brunswick Unit 1                   B 3.3-203                  Revision No. 21    I

LOP Instrumentation B 3.3.8.1 BASES ACTIONS         A.1 (continued)
With one or more channels of a Function inoperable, the Function is not capable of performing the intended function.
Therefore, only I hour is allowed to restore the inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.I. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure (within the LOP instrumentation), and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the channel in trip would result in a DG initiation), Condition B must be entered and its Required Action taken.
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The I hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
B.1 If any Required Action and associated Completion Time are not met, the associated Function is not capable of performing the intended function. Therefore, the associated DG(s) is declared inoperable immediately. This requires entry into applicable Conditions and Required Actions of LCO 3.8.1 and LCO 3.8.2, which provide appropriate actions for the inoperable DG(s).
SURVEILLANCE     As noted at the beginning of the SRs, the SRs for each LOP REQUIREMENTS     instrumentation Function are located in the SRs column of Table 3.3.8.1-1.
The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to (continued)
Brunswick Unit I                  B 3.3-204                  Revision No. 21     I

LOP Instrumentation B 3.3.8.1 BASES SURVEILLANCE     2 hours provided: (a) for Function 1, the associated REQUIREMENTS     Function maintains initiation capability for three DGs; and (continued)    (b) for Function 2, the associate Function maintains DG initiation capability. For Function 1, the loss of function for one DG for this short period is appropriate since only three of four DGs are required to start within the required times and because there is no appreciable impact on risk.
Also, upon completion of the Surveillance, or expiration of the 2 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
SR   3.3.8.1.1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Frequency of 31 days is based on operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function is a rare event.
SR   3.3.8.1.2 and SR  3.3.8.1.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Frequencies of SR 3.3.8.1.2 and SR 3.3.8.1.3 are based upon the assumptions of 18 and 24 month calibration intervals, respectively, in the determination of the magnitude of equipment drift in the setpoint analyses.
(continued)
Brunswick Unit I                  B 3.3-205                  Revision No. 21     I

LOP Instrumentation B 3.3.8.1 BASES SURVEILLANCE     SR  3.3.8.1.4 REQUIREMENTS (continued)    The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed in LCO 3.8.1 and LCO 3.8.2 overlaps this Surveillance to provide complete testing of the assumed safety functions.
The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
Operating experience has demonstrated these components will pass the Surveillance when performed at the 24 month Frequency.
REFERENCES       1. UFSAR, Section 6.3.
: 2. UFSAR, Chapter 15.
: 3. 10 CFR 50.36(c)(2)(ii).
Brunswick Unit I                   B 3.3-206                  Revision No. 21   I

RPS Electric Power Monitoring B 3.3.8.2 B 3.3   INSTRUMENTATION B 3.3.8.2   Reactor Protection System (RPS)    Electric Power Monitoring BASES BACKGROUND          RPS Electric Power Monitoring System is provided to isolate the RPS bus from the motor generator (MG) set or an alternate power supply in the event of overvoltage, undervoltage, or underfrequency. This system protects the loads connected to the RPS bus against unacceptable voltage and frequency conditions (Ref. 1) and forms an important part of the primary success path of the essential safety circuits. Some of the essential equipment powered from the RPS buses includes the RPS logic and scram solenoids.
RPS electric power monitoring assembly will detect any abnormal high or low voltage or low frequency condition in the outputs of the two MG sets or the alternate power supply and will de-energize its respective RPS bus, thereby causing all safety functions normally powered by this bus to de-energize.
In the event of failure of an RPS Electric Power Monitoring System (e.g., both in series electric power monitoring assemblies), the RPS loads may experience significant effects from the unregulated power supply. Deviation from the nominal conditions can potentially cause damage to the scram solenoids and other Class IE devices.
In the event of a low voltage condition for an extended period of time, the scram solenoids can chatter and potentially lose their pneumatic control capability, resulting in a loss of primary scram action.
In the event of an overvoltage condition, the RPS logic relays and scram solenoids may experience a voltage higher than their design voltage. If the overvoltage condition persists for an extended time period, it may cause equipment degradation and the loss of plant safety function.
Two redundant Class IE circuit breakers are connected in series between each RPS bus and its MG set, and between each RPS bus and the alternate power supply. Each of these circuit breakers has an associated independent set of Class IE overvoltage, undervoltage, and underfrequency sensing logic. Together, a circuit breaker and its sensing logic constitute an electric power monitoring assembly. If (continued)
Brunswick Unit 1                     B 3.3-207                   Revision No. 21    I

RPS Electric Power Monitoring B 3.3.8.2 BASES BACKGROUND       the output of the MG set or the alternate power supply (continued)    exceeds predetermined limits of overvoltage, undervoltage, or underfrequency, a trip coil driven by this logic circuitry opens the circuit breaker, which removes the associated power supply from service.
APPLICABLE        The RPS electric power monitoring is necessary to meet the SAFETY ANALYSES   assumptions of the safety analyses by ensuring that the RPS equipment powered from the RPS buses can perform its intended function. RPS electric power monitoring provides protection to the RPS components, by acting to disconnect the RPS from the power supply under specified conditions that could damage the RPS equipment.
RPS electric power monitoring satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 2).
LCO              The OPERABILITY of each RPS electric power monitoring assembly is dependent on the OPERABILITY of the overvoltage, undervoltage, and underfrequency logic, as well as the OPERABILITY of the associated circuit breaker. Two electric power monitoring assemblies are required to be OPERABLE for each inservice power supply. This provides redundant protection against any abnormal voltage or frequency conditions to ensure that no single RPS electric power monitoring assembly failure can preclude the function of RPS components. Each of the inservice electric power monitoring assembly trip logic setpoints is required to be within the specified Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
Allowable Values are specified for each RPS electric power monitoring assembly trip logic (refer to SR 3.3.8.2.2 and SR 3.3.8.2.3). Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g.,
overvoltage), and when the measured output value of (continued)
Brunswick Unit I                  B 3.3-208                    Revision No. 21     I

RPS Electric Power Monitoring B 3.3.8.2 BASES LCO               the process parameter exceeds the setpoint, the associated (continued)    device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors.
The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors.
These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.
The Allowable Values for the instrument settings of the normal power supply (RPS MG set) electric power monitoring assembly are based on the RPS MG sets providing
* 57 Hz and 117 V +/- 10%. The Allowable Values for the instrument settings of the alternate power supply electric power monitoring assembly are based on the alternate power supply providing i 57 Hz and 120 V +/- 10%. The most limiting voltage requirement and associated line losses determine the settings of the electric power monitoring instrument channels. Thesettings are calculated based on the line resistance losses at the downstream locations of the solenoids and relays.
APPLICABILITY    The operation of the RPS electric power monitoring assemblies is essential to disconnect the RPS components from the MG set or alternate power supply during abnormal voltage or frequency conditions. Since the degradation of a nonclass IE source supplying power to the RPS bus can occur as a result of any random single failure, the OPERABILITY of the RPS electric power monitoring assemblies is required when the RPS components are required to be OPERABLE. This results in the RPS Electric Power Monitoring System OPERABILITY being required in MODES 1 and 2; and in MODES 3, 4, and 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies.
(continued)
Brunswick Unit 1                   B 3.3-209                  Revision No. 21    I

RPS Electric Power Monitoring B 3.3.8.2 BASES   (continued)
ACTIONS             A.1 If one RPS electric power monitoring assembly for an inservice power supply (MG set or alternate) is inoperable, or one RPS electric power monitoring assembly on each inservice power supply is inoperable, the OPERABLE assembly will still provide protection to the RPS components under degraded voltage or frequency conditions. However, the reliability and redundancy of the RPS Electric Power Monitoring System is reduced, and only a limited time (72 hours) is allowed to restore the inoperable assembly to OPERABLE status. If the inoperable assembly cannot be restored to OPERABLE status, the associated power supply(s) must be removed from service (Required Action A.1). This places the RPS bus in a safe condition. An alternate power supply with OPERABLE powering monitoring assemblies may then be used to power the RPS bus.
The 72 hour Completion Time takes into account the remaining OPERABLE electric power monitoring assembly and the low probability of an event requiring RPS electric power monitoring protection occurring during this period. It allows time for plant operations personnel to take corrective actions or to place the plant in the required condition in an orderly manner and without challenging plant systems.
Alternately, if it is not desired to remove the power supply from service (e.g., as in the case where removing the power supply(s) from service would result in a scram or isolation), Condition C or D, as applicable, must be entered and its Required Actions taken.
B.1 If both power monitoring assemblies for an inservice power supply (MG set or alternate) are inoperable or both power monitoring assemblies in each inservice power supply are inoperable, the system protective function is lost. In this condition, 1 hour is allowed to restore one assembly to OPERABLE status for each inservice power supply. If one inoperable assembly for each inservice power supply cannot be restored to OPERABLE status, the associated power (continued)
Brunswick Unit 1                      B 3.3-210                  Revision No. 21    I

RPS Electric Power Monitoring B 3.3.8.2 BASES ACTIONS            B.1  (continued) supply(s) must be removed from service within 1 hour (Required Action B.1). An alternate power supply with OPERABLE assemblies may then be used to power one RPS bus.
The I hour Completion Time is sufficient for the plant operations personnel to take corrective actions and is acceptable because it minimizes risk while allowing time for restoration or removal from service of the electric power monitoring assemblies.
Alternately, if it is not desired to remove the power supply(s) from service (e.g., as in the case where removing the power supply(s) from service would result in a scram or isolation), Condition C or D, as applicable, must be entered and its Required Actions taken.
C.1 and C.2' If any Required Action and associated Completion Time of Condition A or B are not met in MODE 1 or 2, a plant shutdown must be performed. This places the plant in a condition where minimal equipment, powered through the inoperable RPS electric power monitoring assembly(s), is required and ensures that the safety function of the RPS (e.g., scram of control rods) is not required. The plant shutdown is accomplished by placing the plant in MODE 3 within 12 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
D.1 If any Required Action and associated Completion Time of Condition A or B are not met in MODE 3, 4, or 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, the operator must immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Required Action D.1 results in the least reactive condition for the reactor core and ensures that the safety function of the RPS (e.g., scram of control rods) is not required.
(continued)
Brunswick Unit 1                    B 3.3-211                 Revision No. 21     I

RPS Electric Power Monitoring B 3.3.8.2 BASES  (continued)
SURVEILLANCE          SR   3.3.8.2.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed on each overvoltage, undervoltage, and underfrequency channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
As noted in the Surveillance, the CHANNEL FUNCTIONAL TEST is only required to be performed while the plant is in a condition in which the loss of the RPS bus will not jeopardize steady state power operation (the design of the system is such that the power source must be removed from service to conduct the Surveillance). The 24 hours is intended to indicate an outage of sufficient duration to allow for scheduling and proper performance of the Surveillance.
The 184 day Frequency and the Note in the Surveillance are based on guidance provided in Generic Letter 91-09 (Ref. 3).
SR   3.3.8.2.2 and SR   3.3.8.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
The Frequencies are based on the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
SR    3.3.8.2.4 Performance of a system functional test demonstrates that, with a required system actuation (simulated or actual) signal, the logic of the system will automatically trip open the associated power monitoring assembly. Only one signal (continued)
Brunswick Unit I                      B 3.3-212                    Revision No. 21     I

RPS Electric Power Monitoring B 3.3.8.2 BASES SURVEILLANCE     SR  3.3.8.2.4   (continued)
REQUIREMENTS per power monitoring assembly is required to be tested.
This Surveillance overlaps with the CHANNEL CALIBRATION to provide complete testing of the safety function. The system functional test of the Class IE circuit breakers is included as part of this test to provide complete testing of the safety function. If the breakers are incapable of operating, the associated electric power monitoring assembly would be inoperable.
The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency.
REFERENCES       1. UFSAR, Section 7.2.1.1.1.3.
: 2. 10 CFR 50.36(c)(2)(ii).
: 3. NRC Generic Letter 91-09, Modification of Surveillance Interval for the Electrical Protective Assemblies in Power Supplies for the Reactor Protection System.
Brunswick Unit I                  B 3.3-213                   Revision No. 21    I

Unit 1 Bases Book 2 Replacement Pages

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Brunswick Unit 1                     LOEP-1I                  Revision 22 1

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Brunswick Unit I                    LOEP-2                    Revision 22

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Brunswick Unit 1                    LOEP-3                   Revision 22 1

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Brunswick Unit I                    LOEP-4                   Revision 22 1

LIST OF EFFECTIVE PAGES - BASES (continued)
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Recirculation Loops Operating B 3.4.1 BASES APPLICABLE       A plant specific LOCA analysis has been performed assuming SAFETY ANALYSES  only one operating recirculation loop. This analysis has (continued)   demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, without the requirement to modify the APLHGR requirements (Ref. 3). However, the COLR may require APLHGR limits to restrict the peak clad temperature for a LOCA with a single recirculation loop operating below the corresponding temperature for both loops operating.
The transient analyses of Chapter 15 of the UFSAR have also been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed without the requirement to modify the MCPR requirements. During single recirculation loop operation, modification to the Reactor Protection.System (RPS) average power range monitor (APRM)
Simulated Thermal Power-High Allowable Value is required to account for the different relationships between recirculation drive flow and reactor core flow. The APRM channel subtracts the tAvalue from the measured recirculation drive flow and uses the adjusted recirculation drive flow value to determine the APRM Simulated Thermal Power-High Function trip setpoint.
Recirculation loops operating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).
LCO              Two recirculation loops are normally required to be in operation with their recirculation pump speeds matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM Simulated Thermal Power-High (contnued (continued)
Brunswick Unit 1                   B 3.4-3                     Revision No. 21 1

Recirculation Loops Operating B 3.4.1 BASES LCO               Allowable Value (LCO 3.3.1.1), as applicable, must be (continued)     applied to allow continued operation. The COLR defines adjustments or modifications required for the APLHGR and MCPR limits for the current operating cycle.
APPLICABILITY     In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.
In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.
ACTIONS          A.1 With the requirements of the LCO not met, the recirculation loops must be restored to operation with matched recirculation pump speeds within 6 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the difference in pump speeds of the two recirculation pumps is greater than the match criteria.
The loop with the lower recirculation pump speed must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the inoperable loop to operating status.
Alternatively, if the single loop requirements of the LCO are applied to operating limits and RPS setpoints, as applicable, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence.
The 6 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action (i.e., reset the applicable limits or setpoints for single recirculation loop operation), and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.
(continued)
Brunswick Unit I                    B 3.4-4                     Revision No. 21 I

Recirculation Loops Operating B 3.4.1 BASES SURVEILLANCE     SR   3.4.1.1   (continued)
REQUIREMENTS 20% match criterion in terms of recirculation pump speed conservatively equates to the 10% match criterion in terms of recirculation loop flow. The generator speed associated with the recirculation pump motor-generator set may be used to measure recirculation pump speed.
The match criteria are measured in terms of the percent difference between recirculation pump speeds. If the difference between the recirculation pump speeds exceeds the match criteria, the loop with the lower recirculation pump speed is considered not in operation. The SR is not required when both loops are not in operation since the match criteria are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The 24 hour Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal recirculation pump speeds in a timely manner.
REFERENCES       1. UFSAR, Section 5.4.1.3.
: 2. UFSAR, Chapter 15.
: 3. NEDC-31776P, Brunswick Steam Electric Plant Units 1       I and 2 Single Loop Operation, February 1990.
: 4. 10 CFR 50.36(c)(2)(ii).
Brunswick Unit I                     B 3.4-6                    Revision No. 21 1

Secondary Containment B 3.6.4.1 B 3.6   CONTAINMENT SYSTEMS B 3.6.4.1   Secondary Containment BASES BACKGROUND          The function of the secondary containment is to contain and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA). In conjunction with operation of the Standby Gas Treatment (SGT) System and closure of certain valves whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment.
The secondary containment is a structure that completely encloses the primary containment and those components that may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure. To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2, "Secondary Containment Isolation Dampers (SCIDs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System."
APPLICABLE         There are two principal accidents for which credit is taken SAFETY ANALYSES    for secondary containment OPERABILITY. These are a loss of coolant accident (LOCA) (Refs. I and 2) and a fuel handling accident involving handling recently irradiated fuel (i.e.,
fuel that has occupied part of a critical reactor core within the previous 24 hours) inside secondary containment.
The secondary containment performs no active function in I
response to each of these limiting events; however, its leak tightness is required to ensure that fission products entrapped within the secondary containment structure will be treated by the SGT System prior to discharge to the environment.
(continued)
Brunswick Unit I                      B 3.6-69                     Revision No. 22 I

Secondary Containment B 3.6.4.1 BASES APPLICABLE         Secondary containment satisfies Criterion 3 of SAFETY ANALYSES    10 CFR 50.36(c)(2)(ii) (Ref. 4).
(continued)
LCO               An OPERABLE secondary containment provides a control volume into which fission products that leak from primary containment, or are released from the reactor coolant pressure boundary components or irradiated fuel assemblies located in secondary containment, can be processed prior to release to the environment. For the secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained, at least one door in each access to the Reactor Building must be closed, and the sealing mechanism associated with each penetration (e.g., welds, bellows or O-rings) must be OPERABLE.
APPLICABILITY     In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.
In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay, secondary containment is only required to be OPERABLE during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours).
ACTIONS          A.1 If secondary containment is inoperable, it must be restored to OPERABLE status within 8 hours. The 8 hour Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary (continued)
Brunswick Unit I                     B 3.6-70                    Revision No. 22 1

Secondary Containment B 3.6.4.1 BASES ACTIONS           A.1   (continued) containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.
B.1 and B.2 If secondary containment cannot be restored to OPERABLE status within the required 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 12 hours and to MODE 4 within 36 hours. 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.
C.1 and C.2 Movement of recently irradiated fuel assemblies in the secondary containment and OPDRVs can be postulated to cause significant fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. Therefore, movement of recently irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperable. Suspension of this activity shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement (continued)
Brunswick Unit I                     B 3.6-71                   Revision No. 22 1

Secondary Containment B 3.6.4.1 BASES ACTIONS          C.1 and C.2   (continued)                                      I I
of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
SURVEILLANCE     SR   3.6.4.1.1 and SR    3.6.4.1.2 REQUIREMENTS Verifying that secondary containment equipment hatches and one secondary containment access door in each access opening are closed ensures that the infiltration of outside air of such magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containment will not occur.
In this application, the term "sealed" has no connotation of leak tightness. Maintaining secondary containment OPERABILITY requires verifying one door in each access opening is closed. The 24 month Frequency for these SRs has been shown to be adequate, based on operating experience, and is considered adequate in view of other indications of door and hatch status that are available to the operator.
SR 3.6.4.1.3 The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment.
To ensure that fission products are treated, SR 3.6.4.1.3 verifies that the SGT System will establish and maintain a negative pressure in the secondary containment. This is confirmed by demonstrating that one SGT subsystem can maintain &#x17d; 0.25 inches of vacuum water gauge for 1 hour at a flow rate - 3000 cfm. The 1 hour test period allows secondary containment to be in thermal equilibrium at steady state conditions. Therefore, this test is used to ensure secondary containment boundary integrity. Since this SR is a secondary containment test, it need not be performed with each SGT subsystem. The SGT subsystems are tested on a STAGGERED TEST BASIS, however, to ensure that in addition to the requirements of LCO 3.6.4.3, either SGT subsystem will perform this test. Operating experience has demonstrated these components will usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
(continued)
Brunswick Unit I                    B 3.6-72                   Revision No. 22 I

Secondary Containment B 3.6.4.1 BASES   (continued)
REFERENCES           I. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
: 2. UFSAR, Section 15.6.4.
: 3. Not used.                                                 I
: 4. 10 CFR 50.36(c)(2)(ii).
: 5. 10 CFR 50.36(c) (2) (ii).
: 6. Regulatory Guide 1.52, Revision 1.
z Brunswick Unit I                    B 3.6-73                    Revision No. 22 1

SCIDs B 3.6.4.2 B 3.6   CONTAINMENT SYSTEMS B 3.6.4.2   Secondary Containment Isolation Dampers (SCIDs)
BASES BACKGROUND          The function of the SCIDs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Refs. 1, 2, and 3). Secondary containment isolation within the time limits specified for those isolation dampers designed to close automatically ensures that fission products that leak from primary containment following a DBA, or that are released during certain operations when primary containment is not required to be OPERABLE or take place outside primary containment, are maintained within the secondary containment boundary.
The OPERABILITY requirements for SCIDs help ensure that an adequate secondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. These isolation devices consist of active (automatic) devices.
Automatic SCIDs close on a secondary containment isolation signal to establish a boundary for untreated radioactive material within secondary containment following a DBA or other accidents.
APPLICABLE          The SCIDs must be OPERABLE to ensure the secondary SAFETY ANALYSES     containment barrier to fission product releases is established. The principal accidents for which the secondary containment boundary is required are a loss of coolant accident (Refs. I and 2) and a fuel handling accident involving handling recently irradiated fuel (i.e.,
fuel that has occupied part of a critical reactor core within the previous 24 hours) inside secondary containment.
The secondary containment performs no active function in response to either of these limiting events, but the boundary established by SCIDs is required to ensure that leakage from the primary containment is processed by the Standby Gas Treatment (SGT) System before being released to the environment.
(continued)
Brunswick Unit I                     B 3.6-74                     Revision No. 22 1

SCIDs B 3.6.4.2 BASES APPLICABLE         Maintaining SCIDs OPERABLE with isolation times within SAFETY ANALYSES   limits ensures that fission products will remain trapped (continued)     inside secondary containment so that they can be treated by the SGT System prior to discharge to the environment.
SCIDs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii)
(Ref. 4).
LCO                SCIDs form a part of the secondary containment boundary.
The SCID safety function is related to control of offsite radiation releases resulting from DBAs.
The isolation dampers are considered OPERABLE when their associated accumulators are pressurized, their isolation times are within limits, and the dampers are capable of actuating on an automatic isolation signal. The dampers covered by this LCO, along with their associated stroke times, are listed in Reference 5.
APPLICABILITY     In MODES 1, 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIDs is required.
In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIDs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. Moving recently irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3. Due to radioactive decay, SCIDs are only required to be OPERABLE during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours).
ACTIONS           The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator, who is (continued)
Brunswick Unit I                    B 3.6-75                    Revision No. 22 1

SCIDs B 3.6.4.2 BASES ACTIONS           in continuous communication with the control room, at the (continued)     controls of the isolation device. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated.
The second Note provides clarification that for the purpose of this LCO separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable SCID. Complying with the Required Actions may allow for continued operation, and subsequent inoperable SCIDs are governed by subsequent Condition entry and application of associated Required Actions.
The third Note ensures appropriate remedial actions are taken, if necessary, if the affected system(s) are rendered inoperable by an inoperable SCID.
A.1 and A.2 In the event that there are one or more penetration flow paths with one SCID inoperable, the affected penetration flow path(s) must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.
Isolation barriers that meet this criterion are a closed and de-activated automatic SCID, a closed manual damper, and a blind flange. For penetrations isolated in accordance with Required Action A.1, the device used to isolate the penetration should be the closest available device to secondary containment. The Required Action must be completed within the 8 hour Completion Time. The specified time period is reasonable considering the time required to isolate the penetration, and the probability of a DBA, which requires the SCIDs to close, occurring during this short time is very low.
For affected penetrations that have been isolated in accordance with Required Action A.1, the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that secondary containment penetrations required to be isolated following an accident, but no longer capable of being automatically isolated, will be in the isolation position should an event occur. The Completion Time of once per 92 days is (continued)
Brunswick Unit I                    .B 3.6-76                    Revision No. 22 1

SCIDs B 3.6.4.2 BASES ACTIONS           A.1 and A.2   (continued) appropriate because the devices are operated under administrative controls and the probability of their misalignment is low. This Required Action does not require any testing or device manipulation. Rather, it involves verification that the affected penetration remains isolated.
Required Action A.2 is modified by a Note that applies to devices located in high radiation areas and allows them to be verified closed by use of administrative controls.
Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment, once they have been verified to be in the proper position, is low.
B.1 With two SCIDs in one or more penetration flow paths inoperable, the affected penetration flow path must be isolated within 4 hours. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.
Isolation barriers that meet this criterion are a closed and de-activated automatic damper, a closed manual damper, and a blind flange. The 4 hour Completion Time is reasonable considering the time required to isolate the penetration and the probability of a DBA, which requires the SCIDs to close, occurring during this short time, is very low.
The Condition has been modified by a Note stating that Condition B is only applicable to penetration flow paths with two isolation dampers. This clarifies that only Condition A is entered if one SCID is inoperable in each of two penetrations.
C.1 and C.2 If any Required Action and associated Completion Time cannot be met in MODE 1, 2, or 3, 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 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating (continued)
Brunswick Unit I                    B 3.6-77                    Revision No. 22 1

SCIDs B 3.6.4.2 BASES ACTIONS           C.1 and C.2   (continued) experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
D.1 and D.2 If any Required Action and associated Completion Time are not met, the plant must be placed in a condition in which the LCO does not apply. If applicable, the movement of recently irradiated fuel assemblies in the secondary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release.
Actions must continue until OPDRVs are suspended.
LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action D.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
SURVEILLANCE     SR   3.6.4.2.1 REQUIREMENTS Verifying that the isolation time of each automatic SCID is within limits, by cycling each SCID through one complete cycle of full travel and measuring the isolation time, is required to demonstrate OPERABILITY. The isolation time test ensures that the SCID will isolate in the required time period. The Frequency of this SR is once per 24 months.
Operating experience has demonstrated these components will usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
(continued)
Brunswick Unit I                    B 3.6-78                   Revision No. 22 I

SCIDs B 3.6.4.2 BASES SURVEILLANCE     SR  3.6.4.2.2 REQUIREMENTS (continued)   Verifying that each automatic SCID closes on a secondary containment isolation signal is required to minimize leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SCID will actuate to the isolation position on a secondary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation," overlaps this SR to provide complete testing of the safety function. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated these components will usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES      1. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
: 2. UFSAR, Section 15.6.4.
: 3. Not used.                                                  I
: 4. 10 CFR 50.36(c)(2)(ii).
: 5. Technical Requirements Manual.
Brunswick Unit I                   B 3.6-79                     Revision No. 22 1

SGT System B 3.6.4.3 BASES BACKGROUND        filter removes fine particulate matter and protects the (continued)    charcoal from fouling. The charcoal adsorber beds remove gaseous elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhausted from the charcoal adsorber.
The SGT System automatically starts and operates in response to actuation signals indicative of conditions or an accident that could require operation of the system. Following an initiation signal, both SGT charcoal filter train fans start.
APPLICABLE        The design basis for the SGT System is to mitigate the SAFETY ANALYSES   consequences of a loss of coolant accident and fuel handling accidents involving handling recently irradiated fuel (i.e.,
fuel that has occupied part of a critical reactor core within the previous 24 hours) (Refs. 2 and 3). For all events analyzed, the SGT System is shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive material released to the environment.
The SGT System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 5).
LCO              Following a DBA, a minimum of one SGT subsystem is required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LCO requirements for two OPERABLE subsystems ensures operation of at least one SGT subsystem in the event of a single active failure.
APPLICABILITY     In MODES 1, 2, and 3, a DBA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, SGT System OPERABILITY is required during these MODES.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT System in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (continued)
Brunswick Unit I                     B 3.6-81                    Revision No. 22 1

SGT System B 3.6.4.3 BASES APPLICABILITY      (OPDRVs) or during movement of recently irradiated fuel (continued)      assemblies in the secondary containment. Due to radioactive decay, the SGT System is only required to be OPERABLE during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours).
ACTIONS            A.1 With one SGT subsystem inoperable in MODE 1, 2, or 3, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT subsystem and the low probability of a DBA occurring during this period.
B.1 and B.2 In MODE 1, 2, or 3, if one SGT subsystem cannot be restored to OPERABLE status within the required Completion Time or both SGT subsystems are inoperable, 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 12 hours and to MODE 4 within 36 hours. 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.
C.1 With one SGT subsystem inoperable during movement of recently irradiated fuel assemblies in secondary containment or during OPDRVs, the inoperable subsystem must be restored to OPERABLE status in 31 days. In this condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is (continued)
Brunswick Unit I                     B 3.6-82                    Revision No. 22 I

SGT System B 3.6.4.3 BASES ACTIONS           C.1   (continued) reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 31 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT subsystem and the probability and consequences of an event requiring the radioactivity release control function during this period.
D.1,  D.2.1,  and D.2.2 During movement of recently irradiated fuel assemblies, in the secondary containment or during OPDRVs, when Required Action C.1 cannot be completed within the required Completion Time, the OPERABLE SGT subsystem should immediately be placed in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected.
An alternative to Required Action D.1 is to immediately suspend activities that represent a potential for releasing a significant amount of radioactive material to the secondary containment, thus placing the plant in a condition that minimizes risk. If applicable, movement of recently irradiated fuel assemblies must immediately be suspended.
Suspension of these activities must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
LCO 3.0.3 is not applicable in MODE 4 or 5.       However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the Required Actions of Condition D have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while I in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.      Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies (continued)
Brunswick Unit I                     B 3.6-83                     Revision No. 22 1

SGT System B 3.6.4.3 BASES ACTIONS           D.1,   D.2.1,  and D.2.2   (continued) would not be a sufficient reason to require a reactor shutdown.
E.1 and E.2 When two SGT subsystems are inoperable, if applicable, movement of recently irradiated fuel assemblies in secondary containment must immediately be suspended.       Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
LCO 3.0.3 is not applicable while in MODE 4 or 5.        However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action E.1 has been modified by a Note stating that LCO 3.0.3 is not applicable.       If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action.      If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
SURVEILLANCE     SR     3.6.4.3.1 REQUIREMENTS Operating each SGT subsystem, by initiating (from the control room) flow through the HEPA filters and charcoal adsorbers, for > 10 continuous hours ensures that both subsystems are OPERABLE and that all associated controls are functioning properly.       It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action.      Operation with the heaters on automatic control for &#x17d; 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters.       The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system.
(continued)
Brunswick Unit 1                      B 3.6-84                     Revision No. 22 1

SGT System B 3.6.4.3 BASES SURVEILLANCE     SR   3.6.4.3.2 REQUIREMENTS (continued)    This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The SGT System filter tests are in accordance with Regulatory Guide 1.52 (Ref. 6), except as specified in Specification 5.5.7, "Ventilation Filter Testing Program (VFTP)". The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). It is noted that, per the basis provided by ESR 99-00055 (Ref. 7), system flow rate is determined using installed calibrated flow orifice plates. Specific test frequencies and additional information are discussed in detail in the VFTP.
SR   3.6.4.3.3 This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal. While this Surveillance can be performed with the reactor at power, operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation," overlaps this SR-to provide complete testing of the safety function.      Therefore, the Frequency was found to be acceptable from a reliability standpoint.
REFERENCES        1. UFSAR, Section 6.5.1.
: 2. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units I and 2, September 1995.
: 3. UFSAR Section 15.6.4.
: 4. Not used.                                                   I
: 5. 10 CFR 50.36(c)(2)(ii).
: 6. Regulatory Guide 1.52, Revision 1.
: 7. ESR 99-00055, SBGT and CBEAF Technical Specification Surveillance Flow Measurement.
Brunswick Unit I                     B 3.6-85                     Revision No. 22 1

Spent Fuel Storage Pool Water Level B 3.7.7 B 3.7   PLANT SYSTEMS B 3.7.7   Spent Fuel Storage Pool Water Level BASES BACKGROUND          The minimum water level in the spent fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.
A general description of the spent fuel storage pool design is found in the UFSAR, Section 9.1.2 (Ref. 1). The assumptions of the fuel handling accident are found in the UFSAR, Section 15.7.1 (Ref. 2).
APPLICABLE          The water level above the irradiated fuel assemblies is an SAFETY ANALYSES     explicit assumption of the fuel handling accident (Ref. 2).
A fuel handling accident is evaluated to ensure that the radiological consequences (calculated control room operator dose and doses at the exclusion area and low population zone boundaries) are well below the 10 CFR 50.67 (Ref. 3) exposure guidelines. A fuel handling accident could release a fraction of the fission product inventory by breaching the fuel rod cladding as discussed in the Regulatory Guide 1.183 I (Ref. 4).
The fuel handling accident is evaluated for the dropping of an irradiated fuel assembly onto the reactor core. The consequences of a fuel handling accident over the spent fuel storage pool are no more severe than those of the fuel handling accident over the reactor core. The water level in the spent fuel storage pool provides for absorption of water soluble fission product gases and transport delays of soluble and insoluble gases that must pass through the water before being released to the secondary containment atmosphere. This absorption and transport delay reduces the potential radioactivity of the release during a fuel handling accident.
The spent fuel storage pool water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 5).
(continued)
Brunswick Unit I                      B 3.7-40                    Revision No. 22 I

Spent Fuel Storage Pool Water Level B 3.7.7 BASES   (continued)
REFERENCES          1. UFSAR, Section 9.1.2.
: 2. UFSAR, Section 15.7.1.
: 3. 10 CFR 50.67.                                            l I
: 4. Regulatory Guide 1.183, July 2000.
: 5. 10 CFR 50.36(c)(2)(ii).
Brunswick Unit I                   B 3.7-42                    Revision No. 22 1

RPV Water Level B 3.9.6 B 3.9   REFUELING OPERATIONS B 3.9.6   Reactor Pressure Vessel (RPV)   Water Level BASES BACKGROUND          The movement of fuel assemblies or handling of control rods within the RPV requires a minimum water level of 23 ft above the top of irradiated fuel assemblies seated within the RPV.
During refueling, this maintains a sufficient water level in the reactor vessel. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 1 and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to well below the 10 CFR 50.67 exposure guidelines I (Ref. 3).
APPLICABLE          During movement of fuel assemblies or handling of control SAFETY ANALYSES     rods, the water level in the RPV is an initial condition design parameter in the analysis of a fuel handling accident in containment postulated by Regulatory Guide 1.183 (Ref. 1). A minimum water level of 23 ft allows a decontamination factor of 200 to be used in the accident analysis for iodine (Ref. 1). This relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the damaged fuel assembly rods is retained by the water.                                         I Analysis of the fuel handling accident inside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 24 hours prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and that offsite doses are maintained well below the allowable limits of Reference 3.
RPV water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).
(continued)
Brunswick Unit I                      B 3.9-19                     Revision No. 22 1

RPV Water Level B 3.9.6 BASES   (continued)
REFERENCES          1. Regulatory Guide 1.183, July 2000.
: 2. UFSAR, Section 15.7.1.
: 3. 10 CFR 50.67.
: 4. 10 CFR 50.36(c)(2)(ii).
Brunswick Unit I                   B 3.9-21                Revision No. 22 1

SDM Test-Refueling B 3.10.8 BASES APPLICABLE       CRDA analyses assume that the reactor operator follows SAFETY ANALYSES  prescribed withdrawal sequences. For SDM tests performed (continued)    within these defined sequences, the analyses of Reference 1 are applicable. However, for some sequences developed for the SDM testing, the control rod patterns assumed in the safety analysis of Reference 1 may not be met. Therefore, special CRDA analyses, performed in accordance with an NRC approved methodology, are required to demonstrate the SDM test sequence will not result in unacceptable consequences should a CRDA occur during the testing. For the purpose of this test, the protection provided by the normally required MODE 5 applicable LCOs, in addition to the requirements of this LCO, will maintain normal test operations as well as postulated accidents within the bounds of the appropriate safety analyses (Ref. 1). In addition to the added requirements for the RWM, APRM, and control rod coupling, the notch out mode is specified for out of sequence withdrawals. Requiring the notch out mode limits withdrawal steps to a single notch, which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may occur during the test.
As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of 10 CFR 50.36(c)(2)(ii) (Ref. 2) apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.
LCO              As described in LCO3.0.7, compliance with this Special Operations:LCO is optional. SDM tests may be performed while in MODE 2, in accordance with Table 1.1-1, without meeting this Special Operations LCO or its ACTIONS. For SDM tests performed while in MODE 5, additional requirements must be met to ensure that adequate protection against potential reactivity excursions is available. To provide additional scram protection, beyond the normally required IRMs, the APRMs are also required to be OPERABLE (LCO 3.3.1.1, Functions 2.a, 2.d, and 2.e) as though the         I reactor were in MODE 2. Because multiple control rods will be withdrawn and the reactor will potentially become critical, the approved control rod withdrawal sequence must be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2), or must be verified by a second licensed operator or other (continued)
Brunswick Unit I                  B 3.10-34                    Revision No. 21    I

SDM Test-Refueling B 3.10.8 BASES LCO              qualified member of the technical staff. To provide (continued)    additional protection against an inadvertent criticality, control rod withdrawals that do not conform to the banked position withdrawal sequence specified in LCO 3.1.6, "Rod Pattern Control," (i.e., out of sequence control rod withdrawals) must be made in the individual notched withdrawal mode to minimize the potential reactivity insertion associated with each movement. Coupling integrity of withdrawn control rods is required to minimize the probability of a CRDA and ensure proper functioning of the withdrawn control rodsi if they are required to scram.
Because the reactor vessel head may be removed during these tests, no other CORE ALTERATIONS may be in progress.
Furthermore, since the control rod scram function with the RCS at atmospheric pressure relies solely on the CRD accumulator, it is essential that the CRD charging water header remain pressurized. This Special Operations LCO then allows changing the Table 1.1-1 reactor mode switch position requirements to include the startup/hot standby position, such that the SDM tests may be performed while in MODE 5.
APPLICABILITY    These SDM test Special Operations requirements are only applicable if the SDM tests are to be performed while in MODE 5 with the reactor vessel head removed or the head bolts not fully tensioned. Additional requirements during these tests to enforce control rod withdrawal sequences and restrict other CORE ALTERATIONS provide protection against potential reactivity excursions. Operations in all other MODES are unaffected by this LCO.
ACTIONS          A.I With one or more control rods discovered uncoupled during this Special Operation, a controlled insertion of each uncoupled control rod is required; either to attempt recoupling, or to preclude a control rod drop. This controlled insertion is preferred since, if the control rod fails to follow the drive as it is withdrawn (i.e., is "stuck" in an inserted position), placing the reactor mode switch in the shutdown position per Required Action B.1 could cause substantial secondary damage. If recoupling is not accomplished, operation may continue, provided the control rods are fully inserted within 3 hours and disarmed (electrically or hydraulically) within 4 hours. Inserting a control rod ensures the shutdown and scram capabilities are (continued)
Brunswick Unit I                  B 3.10-35                    Revision No. 21    I

SDM Test-Refueling B 3.10.8 BASES ACTIONS           A.1   (continued) not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations.
The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves.
Electrically the control rods can be disarmed by disconnecting power from all four directional control valve solenoids. Required Action A.1 is modified by a Note that allows the inoperable control rods to be bypassed or the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation.
LCO 3.3.2.1, "Control Rod Block Instrumentation," Actions provide additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.
The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems.
Condition A is modified by a Note allowing separate Condition entry for each uncoupled control rod. This is acceptable since the Required Actions for this Condition provide appropriate compensatory actions for each uncoupled control rod. Complying with the Required Actions may allow for continued operation. Subsequent uncoupled control rods are governed by subsequent entry into the Condition and application of the Required Actions.
B.1 With one or more of the requirements of this LCO not met for reasons other than an uncoupled control rod, the testing should be immediately stopped by placing the reactor mode switch in the shutdown or refuel position. This results in a condition that is consistent with the requirements for MODE 5 where the provisions of this Special Operations LCO are no longer required.
SURVEILLANCE     SR    3.10.8.1, SR    3.10.8.2. and SR 3.10.8.3 REQUIREMENTS LCO 3.3.1.1, Functions 2.a, 2.d, and 2.e, made applicable in I this Special Operations LCO, are required to have applicable Surveillances met to establish that this Special Operations (continued)
Brunswick Unit I                     B 3.10-36                   Revision No. 21    I

Unit 2 Bases Book 1 Replacement Pages

BASES TO THE FACILITY OPERATING LICENSE DPR-62 TECHNICAL SPECIFICATIONS FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 CAROLINA POWER & LIGHT COMPANY REVISION 22

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Brunswick Unit 2                  LOEP- I                    Revision 22 1

LIST OF EFFECTIVE PAGES - BASES (continued)
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Control Rod Scram Times B 3.1.4 BASES   (continued)
SURVEILLANCE         The four SRs of this LCO are modified by a Note stating that REQUIREMENTS         during a single control rod scram time surveillance, the CRD pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated, (i.e., charging valve closed) the influence of the CRD pump head does not affect the single control rod scram times. During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times.
SR  3.1.4.1 The scram reactivity used in DBA and transient analyses is based on an assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure L 800 psig demonstrates acceptable scram times for the transients analyzed in Reference 4.
Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure > 800 psig ensures that the measured scram times will be within the specified limits at higher pressures. This test is performed for each control rod from its fully withdrawn position. Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed.
To ensure that scram time testing is performed within a reasonable time following a shutdown &#x17d; 120 days, all control rods are required to be tested before exceeding 40% RTP following the shutdown. The specified Frequencies are acceptable considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by fuel movement within the associated core cell and by work on control rods or the CRD System.
(continued)
Brunswick Unit 2                        B 3.1-24                    Revision No. 22    1

Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE     SR 3.1.4.3 (continued)
REQU IREMENTS    scram time sequence is verified. The limits for reactor pressures < 800 psig are established based on a high probability of meeting the acceptance criteria at reactor pressures &#xfd; 800 psig. Limits for &#x17d; 800 psig are found in Table 3.1.4-1 and do not apply for testing performed at
                 < 800 psig. If testing demonstrates the affected control rod does not meet these limits, but is within the 7-second limit of Note 2 to Table 3.1.4-1, the control rod can be considered OPERABLE and "slow."
Specific examples of work that could affect the scram times are (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram.
The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.
SR   3.1.4.4 When work that could affect the scram insertion time is performed on a control rod or CRD System, or when fuel movement within the reactor pressure vessel occurs, testing must be performed to demonstrate each affected control rod is still within the scram time limits of Table 3.1.4-1 with the reactor steam dome pressure &#x17d; 800 psig. Where work has been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 can be satisfied with one test.
For a control rod affected by work performed while shut down, however, a zero pressure and high pressure test may be required. This testing ensures that, prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions. This test is performed for each affected control rod from its fully withdrawn position.
Alternatively, a control rod scram test during hydrostatic pressure testing could also satisfy both criteria. When fuel movement within the reactor pressure vessel occurs, (continued)
Brunswick Unit 2                  B 3.1-26                     Revision No. 22 I

Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE     SR 3.1.4.4    (continued)
REQUIREMENTS only those control rods associated with the control cells affected by the fuel movement are required to be scram time tested. During a routine refueling outage, it is expected that all control rods will be affected. The Frequency of prior to exceeding 40% RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.
REFERENCES      1. USFAR, Section 3.1.2.2.1.
: 2. UFSAR, Section 4.2.1.1.8.
: 3. UFSAR, Section 4.3.2.
: 4. UFSAR, Chapter 15.
: 5. Letter from R.F. Janecek (BWROG) to R.W. Starostecki (NRC), BWR Owners Group Revised Reactivity Control System Technical Specifications, BWROG-8754, September 17, 1987.
: 6. 10 CFR 50.36(c)(2)(ii).
Brunswick Unit 2                   B 3.1-27                    Revision No. 22 I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES BACKGROUND         The outputs of the channels associated with the Reactor (continued)      Building Exhaust Radiation-High Function are arranged in two one-out-of-one trip system logics. Each trip system initiates isolation of one automatic secondary containment isolation damper in each penetration and starts both SGT subsystems while the other trip system initiates isolation of the other secondary containment isolation damper in each penetration and starts both SGT subsystems. Each logic closes one of the two dampers in each penetration and starts both SGT subsystems, so that operation of either logic isolates the secondary containment and provides for the necessary filtration of fission products.
APPLICABLE         The isolation signals generated by the secondary containment SAFETY ANALYSES,   isolation instrumentation are implicitly assumed in the LCO, and           safety analyses of References I and 3 to initiate                I APPLICABILITY      closure of dampers and start the SGT System to limit offsite doses.
Refer to LCO 3.6.4.2, "Secondary Containment Isolation Dampers (SCIDs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System," Applicable Safety Analyses Bases for more detail of the safety analyses.
The secondary containment isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).
Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.
The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have the required number of OPERABLE channels with their setpoints set within the specified Allowable Values, as shown in Table 3.3.6.2-1.      The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
Allowable Values are specified for each Function specified in the Table. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less (continued)
Brunswick Unit 2                     B 3.3-176                     Revision No. 21 1

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE          2. Drywell Pressure-High   (continued)
SAFETY ANALYSES, LCO, and           dose release. The isolation on high drywell pressure APPLICABILITY      supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis.
However, the Drywell Pressure-High Function associated with isolation is not assumed in any UFSAR accident or transient analyses. It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis.
High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation function.
The Allowable Value was chosen to be the same as the ECCS Drywell Pressure-High Function Allowable Value (LCO 3.3.5.1) since this is indicative of a loss of coolant accident (LOCA).
The Drywell Pressure-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.
: 3. Reactor Building Exhaust Radiation-High High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding.
The release may have originated from the primary containment due to a break in the RCPB or the refueling floor due to a fuel handling accident. When Reactor Building Exhaust Radiation-High is detected, secondary containment isolation and actuation of the SGT System are initiated to limit the release of fission products.                                     I (continued)
Brunswick Unit 2                    B 3.3-179                    Revision No. 21 1

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE         3. Reactor Building Exhaust Radiation-High    (continued)
SAFETY ANALYSES, LCO, and           The Reactor Building Exhaust Radiation-High signals are APPLICABILITY      initiated from radiation detectors that are located in the ventilation exhaust ductwork plenum coming from the reactor building. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Two channels of Reactor Building Exhaust Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
The Allowable Values are chosen to promptly detect gross failure of the fuel cladding.
The Reactor Building Exhaust Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, the Function is not required. In addition, the Function is also required to be OPERABLE during OPDRVs and movement of recently irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation I
releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded. Due to radioactive decay, this function is only required to isolate secondary containment during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours).
ACTIONS           A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels.
Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the (continued)
Brunswick Unit 2                    B 3.3-180                    Revision No. 21 1

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS            Condition. However, the Required Actions for inoperable (continued)     secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel.
A.1 Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Function 2, and 24 hours for Functions other than Function 2, has been shown to be acceptable (Refs. 5 and 6) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases).
If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an isolation), Condition C must be entered and its Required Actions taken.
B.1 Required Action B.I is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic isolation capability for the associated penetration flow path(s) or a complete loss of automatic initiation capability for the SGT System. A Function is considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from the given Function on a valid signal. This ensures that one of the two SCIDs in the associated penetration flow path and both SGT subsystems can be initiated on an isolation signal from (continued)
Brunswick Unit 2                    B 3.3-181                    Revision No. 21 1

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS            B.1   (continued) the given Function. For the Functions with two one-out-of-two logic trip systems (Functions I and 2), this would require one trip system to have one channel OPERABLE or in trip. An inoperable channel need not be placed in the tripped condition where this would cause the trip Function to occur. In these cases, if the inoperable channel is not restored within the required Completion Time, Condition C shall be entered.
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The I hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
C.1.1, C.1.2, C.2.1, and C.2.2 If any Required Action and associated Completion Time of Condition A or B are not met, the ability to isolate the secondary containment and start the SGT System cannot be ensured. Therefore, further actions must be performed to ensure the ability to maintain the secondary containment function. Isolating the associated dampers and starting the associated SGT subsystem (Required Actions C.1.1 and C.2.1) performs the intended function of the instrumentation and allows operation to continue.
Alternately, declaring the associated SCIDs or SGT subsystem(s) inoperable (Required Actions C.1.2 and C.2.2) is also acceptable since the Required Actions of the respective LCOs (LCO 3.6.4.2 and LCO 3.6.4.3) provide appropriate actions for the inoperable components.
One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components inoperable without unnecessarily challenging plant systems.
SURVEILLANCE     As noted at the beginning of the SRs, the SRs for each REQUIREMENTS     Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1.
(continued)
Brunswick Unit 2                     B 3.3-182                    Revision No. 21 I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE       The Surveillances are modified by a Note to indicate that REQUIREMENTS       when a channel is placed in an inoperable status solely for (continued)     performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows:
(a) for up to 2 hours for Function 3 and (b) for up to 6 hours for Functions I and 2 provided the associated Function maintains isolation capability. Upon completion of the Surveillance, or expiration of the 2 hour allowance for Function 3 or the 6 hour allowance for Functions I and 2, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
This Note is based on the reliability analysis (Refs. 5 and 6) assumption of the average time required to perform channel surveillance. That analysis demonstrated the 2 hour testing allowance for Function 3 and the 6 hour testing allowance for Functions I and 2 do not significantly reduce the probability that the SCIDs will isolate the associated penetration flow paths and that the SGT System will initiate when necessary.
SR   3.3.6.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of (continued)
Brunswick Unit 2                    B 3.3-183                    Revision No. 21 1

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE      SR   3.3.6.2.1   (continued)
REQUIREMENTS channel status during normal operational use of the displays associated with channels required by the LCO.
SR   3.3.6.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Frequency of 92 days is based on the reliability analysis of References 5 and 6.
SR 3.3.6.2.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.2-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
The Frequency of 92 days is based on the reliability analysis of References 5 and 6.
SR    3.3.6.2.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
(continued)
Brunswick Unit 2                    B 3.3-184                    Revision No. 21 I

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE      SR  3.3.6.2.4    (continued)
REQUIREMENTS The Frequency is based on the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
SR  3.3.6.2.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed on SCIDs and the SGT System in LCO 3.6.4.2 and LCO 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function.
The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency.
REFERENCES       1. UFSAR, Section 15.6.4.
: 2. Not used.
: 3. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units I and 2, September 1995.
: 4. 10 CFR 50.36(c)(2)(ii).
: 5. NEDC-31677P-A, Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation, July 1990.
: 6. NEDC-30851P-A Supplement 2, Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
Brunswick Unit 2                     B 3.3-185                    Revision No. 21 I

CREV System Instrumentation B 3.3.7.1 B 3.3   INSTRUMENTATION B 3.3.7.1   Control Room Emergency Ventilation (CREV)   System Instrumentation BASES BACKGROUND           The CREV System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions. Two independent CREV subsystems are each capable of fulfilling the stated safety function. The instrumentation and controls for the CREV System automatically initiate action to pressurize the main control room (MCR) to minimize the consequences of radioactive material in the control room environment.
In the event of a Control Building Air Intake Radiation-High signal, the CREV System is automatically started in the radiation/smoke protection mode. Air is then recirculated through the charcoal filter, and sufficient outside air is drawn in through the normal intake to maintain the MCR slightly pressurized with respect to outside atmosphere.
The CREV System instrumentation has two trip systems, either of which can initiate the CREV System. Each trip system receives input from the two Control Building Air Intake Radiation-High Function channels. The Control Building Air Intake Radiation-High Function is arranged in a one-out-of-two logic for each trip system. The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a CREV System initiation signal to the initiation logic.
APPLICABLE          The ability of the CREV System to maintain the habitability SAFETY ANALYSES,    of the MCR is explicitly assumed for the design basis LCO, and            accident as discussed in the UFSAR safety analyses (Refs. 1 APPLICABILITY       and 2). CREV System operation ensures that the radiation exposure of control room personnel, through the duration of I
any one of the postulated accidents, does not exceed the limits set by GDC 19 of 10 CFR 50, Appendix A.
CREV System instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 3).                                 I (continued)
Brunswick Unit 2                      B 3.3-186                    Revision No. 21 1

CREV System Instrumentation B 3.3.7.1 BASES APPLICABLE         The Control Building Air Intake Radiation-High Function SAFETY ANALYSES,   consists of two independent monitors. Two channels per trip LCO, and           system of Control Building Air Intake Radiation-High are APPLICABILITY      available and are required to be OPERABLE to ensure that no (continued)      single instrument failure can preclude CREV System initiation. The Allowable Value was selected to ensure protection of the control room personnel.
The Control Building Air Intake Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 and during OPDRVs and movement of recently irradiated fuel assemblies in the secondary containment, to ensure that control room personnel are protected during a LOCA, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., OPDRVs), the probability of a LOCA, main steam line break accident, or control rod drop accident is low; thus, the Function is not required. Also due to radioactive decay, this Function is only required to initiate the CREV System during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours).
ACTIONS           A Note has been provided to modify the ACTIONS related to CREV System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable CREV System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable CREV System instrumentation channel.
A.1 Because of the redundancy of sensors available to provide initiation signals and the redundancy of the CREV System design, an allowable out of service time of 7 days is provided to permit restoration of any inoperable channel to (continued)
Brunswick Unit 2                    B 3.3-188                    Revision No. 21 1

CREV System Instrumentation B 3.3.7.1 BASES ACTIONS            A.1   (continued)
OPERABLE status. This out of service time is only acceptable provided the Control Building Air Intake Radiation-High Function is still maintaining CREV System initiation capability (refer to Required Action B.1 Bases).
If the Function is not maintaining CREV System initiation capability, Condition B must be entered.
If the inoperable channel cannot be restored to OPERABLE status within the 7 day allowable out of service time, one CREV subsystem must be placed in the radiation/smoke protection mode of operation per Required Action A.I. The method used to place the CREV subsystem in operation must provide for automatically re-initiating the subsystem upon restoration of power following a loss of power to the CREV subsystem. Placing one CREV subsystem in the radiation/smoke protection mode of operation provides a suitable compensatory action to ensure that the automatic radiation protection function of the CREV System is not lost.
B.1 Required Action B.1 is intended to ensure that appropriate action is taken if multiple, inoperable, untripped channels result in the Control Building Air Intake Radiation-High Function not maintaining CREV System initiation capability.
The Function is considered to be maintaining CREV System initiation capability when sufficient channels are OPERABLE or in trip such that one trip system will generate an initiation signal for one CREV subsystem from the Function on a valid signal. For the Control Building Air Intake Radiation-High Function, this would require one trip system to have one channel OPERABLE or in trip. With CREV System initiation capability not maintained, one CREV subsystem must be placed in the radiation/smoke protection mode of operation per Required Action B.1 to ensure that control room personnel will be protected in the event of a Design Basis Accident.      The method used to place the CREV subsystem in operation must provide for automatically re-initiating the subsystem upon restoration of power following a loss of power to the CREV subsystem.
(continued)
Brunswick Unit 2                     B 3.3-189                    Revision No. 21 1

CREV System Instrumentation B 3.3.7.1 BASES ACTIONS           B.1   (continued)
The 1 hour Completion Time is intended to allow the operator time to place the CREV subsystem in operation. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels, or for placing one CREV subsystem in operation.
SURVEILLANCE      The Surveillances are modified by a Note to indicate that REQUIREMENTS      when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains CREV System initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
This Note is based on the reliability analysis (Ref. 4)          I assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the CREV System will initiate when necessary.
SR   3.3.7.1.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
(continued)
Brunswick Unit 2                     B 3.3-190                   Revision No. 21 I

CREV System Instrumentation B 3.3.7.1 BASES SURVEILLANCE      SR  3.3.7.1.1   (continued)
REQUIREMENTS The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with channels required by the LCO.
SR  3.3.7.1.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Frequency of 92 days is based on the reliability analyses of Reference 4.
SR   3.3.7.1.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
SR   3.3.7.1.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.7.3, "Control Room Emergency Ventilation (CREV)
System," overlaps this Surveillance to provide complete testing of the assumed safety function.
(continued)
Brunswick Unit 2                   B 3.3-191                    Revision No. 21 1

CREV System Instrumentation B 3.3.7.1 BASES SURVEILLANCE    SR  3.3.7.1.4   (continued)
REQUIREMENTS While this surveillance can be performed with the reactor at power, operating experience has demonstrated these components will usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was found to be acceptable from a reliability standpoint.
REFERENCES      1. UFSAR, Section 15.6.4.5.5.
: 2. UFSAR, Section 15.7.1
: 3. 10 CFR 50.36(c)(2)(ii).
: 4. GENE-770-06-1-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, December 1992.
Brunswick Unit 2                 B 3.3-192                    Revision No. 21   I

Unit 2 Bases Book 2 Replacement Pages

LIST OF EFFECTIVE PAGES - BASES Page No.            Revision No.         Page No. Revision No.
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Brunswick Unit 2                   LOEP-1                   Revision 21 1

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Brunswick Unit 2                    LOEP-2                     Revision 21 1

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Brunswick Unit 2                     LOEP-3                     Revision 21 1

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Brunswick Unit 2                     LOEP-4                   Revision 21 1

Secondary Containment B 3.6.4.1 B 3.6  CONTAINMENT SYSTEMS B 3.6.4.1   Secondary Containment BASES BACKGROUND          The function of the secondary containment is to contain and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA). In conjunction with operation of the Standby Gas Treatment (SGT) System and closure of certain valves whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment.
The secondary containment is a structure that completely encloses the primary containment and those components that may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure. To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2, "Secondary Containment Isolation Dampers (SCIDs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System."
APPLICABLE         There are two principal accidents for which credit is taken SAFETY ANALYSES    for secondary containment OPERABILITY. These are a loss of coolant accident (LOCA) (Refs. I and 2) and a fuel handling accident involving handling recently irradiated fuel (i.e.,
fuel that has occupied part of a critical reactor core within the previous 24 hours) inside secondary containment.
The secondary containment performs no active function in response to each of these limiting events; however, its leak tightness is required to ensure that fission products entrapped within the secondary containment structure will be treated by the SGT System prior to discharge to the environment.
(continued)
Brunswick Unit 2                     B 3.6-69                     Revision No. 21 1

Secondary Containment B 3.6.4.1 BASES APPLICABLE          Secondary containment satisfies Criterion 3 of SAFETY ANALYSES     10 CFR 50.36(c)(2)(ii) (Ref. 4).
(continued)
LCO                An OPERABLE secondary containment provides a control volume into which fission products that leak from primary containment, or are released from the reactor coolant pressure boundary components or irradiated fuel assemblies located in secondary containment, can be processed prior to release to the environment. For the secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained, at least one door in each access to the Reactor Building must be closed, and the sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) must be OPERABLE.
APPLICABILITY      In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.
In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay, secondary containment is only required to be OPERABLE during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours).
ACTIONS           A.1 If secondary containment is inoperable, it must be restored to OPERABLE status within 8 hours. The 8 hour Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary (continued)
Brunswick Unit 2                      B 3.6-70                    Revision No. 21 I

Secondary Containment B 3.6.4.1 BASES ACTIONS           A.1   (continued) containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.
B.1 and B.2 If secondary containment cannot be restored to OPERABLE status within the required 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 12 hours and to MODE 4 within 36 hours. 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.
C.1 and C.2 Movement of irradiated fuel assemblies in the secondary containment and OPDRVs can be postulated to cause significant fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. Therefore, movement of recently irradiated fuel assemblies must be immediately suspended if the secondary containment is operable. Suspension of this activity shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement (continued)
Brunswick Unit 2                     B 3.6-71                    Revision No. 21 I

Secondary Containment B 3.6.4.1 BASES ACTIONS           C.1 and C.2   (continued)                                      I of recently irradiated fuel assemblies would not be a I
sufficient reason to require a reactor shutdown.
SURVEILLANCE       SR  3.6.4.1.1 and SR   3.6.4.1.2 REQUIREMENTS Verifying that secondary containment equipment hatches and one secondary containment access door in each access opening are closed ensures that the infiltration of outside air of such magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containment will not occur.
In this application, the term "sealed" has no connotation of leak tightness. Maintaining secondary containment OPERABILITY requires verifying one door in each access opening is closed. The 24 month Frequency for these SRs has been shown to be adequate, based on operating experience, and is considered adequate in view of other indications of door and hatch status that are available to the operator.
SR   3.6.4.1.3 The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment.
To ensure that fission products are treated, SR 3.6.4.1.3 verifies that the SGT System will establish and maintain a negative pressure in the secondary containment. This is confirmed by demonstrating that one SGT subsystem can maintain i 0.25 inches of vacuum water gauge for 1 hour at a flow rate - 3000 cfm. The I hour test period allows secondary containment to be in thermal equilibrium at steady state conditions. Therefore, this test is used to ensure secondary containment boundary integrity. Since this SR is a secondary containment test, it need not be performed with each SGT subsystem. The SGT subsystems are tested on a STAGGERED TEST BASIS, however, to ensure that in addition to the requirements of LCO 3.6.4.3, either SGT subsystem will perform this test. Operating experience has demonstrated these components will usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
(continued)
Brunswick Unit 2                    B 3.6-72                    Revision No. 21 1

Secondary Containment B 3.6.4.1 BASES   (continued)
REFERENCES          1. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
: 2. UFSAR, Section 15.6.4.
: 3. Not used.                                                I
: 4. 10 CFR 50.36(c)(2)(ii).
Brunswick Unit 2                    B 3.6-73                   Revision No. 21 I

SCIDs B 3.6.4.2 B 3.6   CONTAINMENT SYSTEMS B 3.6.4.2   Secondary Containment Isolation Dampers (SCIDs)
BASES BACKGROUND          The function of the SCIDs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Refs. 1, 2, and 3). Secondary containment isolation within the time limits specified for those isolation dampers designed to close automatically ensures that fission products that leak from primary containment following a DBA, or that are released during certain operations when primary containment is not required to be OPERABLE or take place outside primary containment, are maintained within the secondary containment boundary.
The OPERABILITY requirements for SCIDs help ensure that an adequate secondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. These isolation devices consist of active (automatic) devices.
Automatic SCIDs close on a secondary containment isolation signal to establish a boundary for untreated radioactive material within secondary containment following a DBA or other accidents.
APPLICABLE          The SCIDs must be OPERABLE to ensure the secondary SAFETY ANALYSES     containment barrier to fission product releases is established. The principal accidents for which the secondary containment boundary is required are a loss of coolant accident (Refs. 1 and 2) and a fuel handling accident involving handling recently irradiated fuel (i.e.,
fuel that has occupied part of a critical reactor core within the previous 24 hours) inside secondary containment.
The secondary containment performs no active function in response to either of these limiting events, but the boundary established by SCIDs is required to ensure that leakage from the primary containment is processed by the Standby Gas Treatment (SGT) System before being released to the environment.
(continued)
Brunswick Unit 2                      B 3.6-74                     Revision No. 21 I

SCIDs B 3.6.4.2 BASES APPLICABLE         Maintaining SCIDs OPERABLE with isolation times within SAFETY ANALYSES    limits ensures that fission products will remain trapped (continued)      inside secondary containment so that they can be treated by the SGT System prior to discharge to the environment.
SCIDs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii)
(Ref. 4).
LCO                SCIDs form a part of the secondary containment boundary.
The SCID safety function is related to control of offsite radiation releases resulting from DBAs.
The isolation dampers are considered OPERABLE when their associated accumulators are pressurized, their isolation times are within limits, and the dampers are capable of actuating on an automatic isolation signal. The dampers covered by this LCO, along with their associated stroke times, are listed in Reference 5.
APPLICABILITY      In MODES 1, 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIDs is required.
In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIDs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as duringoperationswith a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. Moving recently irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3. Due to radioactive decay, SCIDs are only required to be OPERABLE during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours).
ACTIONS           The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator, who is (continued)
Brunswick Unit 2                     B 3.6-75                    Revision No. 21 1

SCIDs B 3.6.4.2 BASES ACTIONS           in continuous communication with the control room, at the (continued)     controls of the isolation device. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated. The second Note provides clarification that for the purpose of this LCO separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable SCID. Complying with the Required Actions may allow for continued operation, and subsequent inoperable SCIDs are governed by subsequent Condition entry and application of associated Required Actions.
The third Note ensures appropriate remedial actions are taken, if necessary, if the affected system(s) are rendered inoperable by an inoperable SCID.
A.1 and A.2 In the event that there are one or more penetration flow paths with one SCID inoperable, the affected penetration flow path(s) must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.
Isolation barriers that meet this criterion are a closed and de-activated automatic SCID, a closed manual damper, and a blind flange. For penetrations isolated in accordance with Required Action A.1, the device used to isolate the penetration should be the closest available device to secondary containment. The Required Action must be completed within the 8 hour Completion Time. The specified time period is reasonable considering the time required to isolate the penetration, and the probability of a DBA, which requires the SCIDs to close, occurring during this short time is very low.
For affected penetrations that have been isolated in accordance with Required Action A.1, the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that secondary containment penetrations required to be isolated following an accident, but no longer capable of being automatically isolated, will be in the isolation position should an event occur. The Completion Time of once per 92 days is (continued)
Brunswick Unit 2                    B 3.6-76                    Revision No. 21 I

SCIDs B 3.6.4.2 BASES ACTIONS           A.1 and A.2   (continued) appropriate because the devices are operated under administrative controls and the probability of their misalignment is low. This Required Action does not require any testing or device manipulation. Rather, it involves verification that the affected penetration remains isolated.
Required Action A.2 is modified by a Note that applies to devices located in high radiation areas and allows them to be verified closed by use of administrative controls.
Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment, once they have been verified to be in the proper position, is low.
B.1 With two SCIDs in one or more penetration flow paths inoperable, the affected penetration flow path must be isolated within 4 hours. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.
Isolation barriers that meet this criterion are a closed and de-activated automatic damper, a closed manual damper, and a blind flange. The 4 hour Completion Time is reasonable considering the time required to isolate the penetration and the probability of a DBA, which requires the SCIDs to close, occurring during this short time, is very low.
The Condition has been modified by a Note stating that Condition B is only applicable to penetration flow paths with two isolation dampers. This clarifies that only Condition A is entered if one SCID is inoperable in each of two penetrations.
C.1 and C.2 If any Required Action and associated Completion Time cannot be met in MODE 1, 2, or 3, 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 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating (continued)
Brunswick Unit 2                    B 3.6-77                     Revision No. 21 1

SCIDs B 3.6.4.2 BASES ACTIONS           C.1 and C.2   (continued) experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
D.1 and D.2 If any Required Action and associated Completion Time are not met, the plant must be placed in a condition in which the LCO does not apply. If applicable, the movement of recently irradiated fuel assemblies in the secondary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release.
Actions must continue until OPDRVs are suspended.
LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action D.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
SURVEILLANCE     SR    3.6.4.2.1 REQUIREMENTS Verifying that the isolation time of each automatic SCID is within limits, by cycling each SCID through one complete cycle of full travel and measuring the isolation time, is required to demonstrate OPERABILITY. The isolation time test ensures that the SCID will isolate in the required time period. The Frequency of this SR is once per 24 months.
Operating experience has demonstrated these components will usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
(continued)
Brunswick Unit 2                    B 3.6-78                    Revision No. 21 I

SCIDs B 3.6.4.2 BASES SURVEILLANCE     SR  3.6.4.2.2 REQUIREMENTS (continued)    Verifying that each automatic SCID closes on a secondary containment isolation signal is required to minimize leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SCID will actuate to the isolation position on a secondary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation," overlaps this SR to provide complete testing of the safety function. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated these components will usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES       1. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units I and 2, September 1995.
: 2. UFSAR, Section 15.6.4.
: 3. Not used.                                                  I
: 4. 10 CFR 50.36(c)(2)(ii).
: 5. Technical Requirements Manual.
Brunswick Unit 2                   B 3.6-79                    Revision No. 21 I

SGT System B 3.6.4.3 BASES BACKGROUND        filter removes fine particulate matter and protects the (continued)   charcoal from fouling. The charcoal adsorber beds remove gaseous elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhausted from the charcoal adsorber.
The SGT System automatically starts and operates in response to actuation signals indicative of conditions or an accident that could require operation of the system. Following an initiation signal, both SGT charcoal filter train fans start.
APPLICABLE        The design basis for the SGT System is to mitigate the SAFETY ANALYSES   consequences of a loss of coolant accident and fuel handling accidents involving handling recently irradiated fuel (i.e.,
fuel that has occupied part of a critical reactor core within the previous 24 hours) (Refs. 2 and 3). For all       I events analyzed, the SGT System is shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive material released to the environment.
The SGT System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 5).
LCO              Following a DBA, a minimum of one SGT subsystem is required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LCO requirements for two OPERABLE subsystems ensures operation of at least one SGT subsystem in the event of a single active failure.
APPLICABILITY     In MODES 1, 2, and 3, a DBA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, SGT System OPERABILITY is required during these MODES.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT System in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (continued)
Brunswick Unit 2                    B 3.6-81                   Revision No. 21 I

SGT System B 3.6.4.3 BASES APPLICABILITY      (OPDRVs) or during movement of recently irradiated fuel (continued)     assemblies in the secondary containment. Due to radioactive decay, the SGT System is only required to be OPERABLE during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours).
ACTIONS           A.1 With one SGT subsystem inoperable in MODE 1, 2, or 3, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT subsystem and the low probability of a DBA occurring during this period.
B.1 and B.2 In MODE 1, 2, or 3, if one SGT subsystem cannot be restored to OPERABLE status within the required Completion Time or both SGT subsystems are inoperable, 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 12 hours and to MODE 4 within 36 hours. 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.
C.1 With one SGT subsystem inoperable during movement of recently irradiated fuel assemblies in secondary containment, or during OPDRVs, the inoperable subsystem must be restored to OPERABLE status in 31 days. In this condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is (conti nued)
Brunswick Unit 2                    B 3.6-82                    Revision No. 21 1

SGT System B 3.6.4.3 BASES ACTIONS           C.1   (continued) reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 31 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT subsystem and the probability and consequences of an event requiring the radioactivity release control function during this period.
D.1, D.2.1 and D.2.2 During movement of recently irradiated fuel assemblies, in the secondary containment or during OPDRVs, when Required Action C.1 cannot be completed within the required Completion Time, the OPERABLE SGT subsystem should immediately be placed in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected.
An alternative to Required Action D.1 is to immediately suspend activities that represent a potential for releasing a significant amount of radioactive material to the secondary containment, thus placing the plant in a condition that minimizes risk. If applicable, movement of recently irradiated fuel assemblies must immediately be suspended.
Suspension of these activities must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
LCO 3.0.3 is not applicable in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the Required Actions of Condition D have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while I in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies (continued)
Brunswick Unit 2                    B 3.6-83                     Revision No. 21 I

SGT System B 3.6.4.3 BASES ACTIONS           D.1, D.2.1 and D.2.2   (continued) would not be a sufficient reason to require a reactor shutdown.
E.1 and E.2 When two SGT subsystems are inoperable, if applicable, movement of recently irradiated fuel assemblies in secondary containment must immediately be suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action E.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
SURVEILLANCE     SR    3.6.4.3.1 REQUIREMENTS Operating each SGT subsystem, by initiating (from the control room) flow through the HEPA filters and charcoal adsorbers, for -10 continuous hours ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on automatic control for
* 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system.
(continued)
Brunswick Unit 2                    B 3.6-84                    Revision No. 21 I

SGT System B 3.6.4.3 BASES SURVEILLANCE      SR   3.6.4.3.2 REQUIREMENTS (continued)     This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The SGT System filter tests are in accordance with Regulatory Guide 1.52 (Ref. 6), except as specified in Specification 5.5.7, "Ventilation Filter Testing Program (VFTP)".      The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). It is noted that, per the basis provided by ESR 99-00055 (Ref. 7), system flow rate is determined using installed calibrated flow orifice plates. Specific test frequencies and additional information are discussed in detail in the VFTP.
SR    3.6.4.3.3 This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal. While this Surveillance can be performed with the reactor at power, operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation," overlaps this SR to provide complete testing of the safety function. Therefore, the Frequency was found to be acceptable from a reliability standpoint.
REFERENCES       I. UFSAR, Section 6.5.1.
: 2. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units I and 2, September 1995.
: 3.      UFSAR Section 15.6.4.
: 4.      Not used.                                                    I
: 5.      10 CFR 50.36(c)(2)(ii).
: 6. Regulatory Guide 1.52, Revision 1.
: 7. ESR 99-00055, SBGT and CBEAF Technical Specification Surveillance Flow Measurement.
Brunswick Unit 2                       B 3.6-85                    Revision No. 21 I

Spent Fuel Storage Pool Water Level B 3.7.7 B 3.7   PLANT SYSTEMS B 3.7.7   Spent Fuel Storage Pool Water Level BASES BACKGROUND           The minimum water level in the spent fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.
A general description of the spent fuel storage pool design is found in the UFSAR, Section 9.1.2 (Ref. 1). The assumptions of the fuel handling accident are found in the UFSAR, Section 15.7.1 (Ref. 2).
APPLICABLE          The water level above the irradiated fuel assemblies is an SAFETY ANALYSES     explicit assumption of the fuel handling accident (Ref. 2).
A fuel handling accident is evaluated to ensure that the radiological consequences (calculated control room operator dose and doses at the exclusion area and low population zone J boundaries) are well below the 10 CFR 50.67 (Ref. 3)
I exposure guidelines. A fuel handling accident could release a fraction of the fission product inventory by breaching the I fuel rod cladding as discussed in the Regulatory Guide 1.183 (Ref. 4).
The fuel handling accident is evaluated for the dropping of an irradiated fuel assembly onto the reactor core. The consequences of a fuel handling accident over the spent fuel storage pool are no more severe than those of the fuel handling accident over the reactor core. The water level in the spent fuel storage pool provides for absorption of water soluble fission product gases and transport delays of soluble and insoluble gases that must pass through the water before being released to the secondary containment atmosphere. This absorption and transport delay reduces the potential radioactivity of the release during a fuel handling accident.
The spent fuel storage pool water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 5).
(continued)
Brunswick Unit 2                      B 3.7-40                     Revision No. 21 I

Spent Fuel Storage Pool Water Level B 3.7.7 BASES   (continued)
REFERENCES          1. UFSAR, Section 9.1.2.
: 2. UFSAR, Section 15.7.1.
: 3. 10 CFR 50.67.                                             I I
: 4. Regulatory Guide 1.183, July 2000.
: 5. 10 CFR 50.36(c)(2)(ii).
Brunswick Unit 2                   .B 3.7-42                    Revision No. 21 1

RPV Water Level B 3.9.6 B 3.9   REFUELING OPERATIONS B 3.9.6   Reactor Pressure Vessel (RPV)    Water Level BASES BACKGROUND           The movement of fuel assemblies or handling of control rods within the RPV requires a minimum water level of 23 ft above the top of irradiated fuel assemblies seated within the RPV.
During refueling, this maintains a sufficient water level in the reactor vessel. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. I and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to well below the 10 CFR 50.67 exposure guidelines I (Ref. 3).
APPLICABLE          During movement of fuel assemblies or handling of control SAFETY ANALYSES     rods, the water level in the RPV is an initial condition design parameter in the analysis of a fuel handling accident in containment postulated by Regulatory Guide 1.183 (Ref. 1). A minimum water level of 23 ft allows a decontamination factor of 200 to be used in the accident analysis for iodine (Ref. 1). This relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the damaged fuel assembly rods is retained by the water.                                         I Analysis of the fuel handling accident inside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 24 hours prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and that offsite doses are maintained well below the allowable limits of Reference 3.
RPV water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).
(continued)
Brunswick Unit 2                       B 3.9-19                     Revision No. 21 1

RPV Water Level B 3.9.6 BASES  (continued)
REFERENCES         1. Regulatory Guide 1.183, July 2000
: 2. UFSAR, Section 15.7.1.
: 3. 10 CFR 50.67.
: 4. 10 CFR 50.36(c)(2)(ii).
Brunswick Unit 2                   B 3.9-21               Revision No. 21 1}}