ML060180452: Difference between revisions

a. Reactor Vessel Water Level - Low Low Low, Level 1;

b. Drywell Pressure - High;

c. Reactor Steam Dome Pressure - Low (Injection Permissive and ECCS Initiation); and

d. Reactor Steam Dome Pressure - Low (Recirculation Discharge Valve Permissive).

a. Reactor Vessel Water Level - Low Low, Level 2; and

b. Drywell Pressure - High.

4. Automatic Depressurization System (ADS) Trip System A

a. Reactor Vessel Water Level - Low Low Low, Level 1;

b. Drywell Pressure - High; and

d. Reactor Vessel Water Level - Low, Level 3 (Confirmatory).

5. Automatic Depressurization System (ADS) Trip System B

a. Reactor Vessel Water Level - Low Low Low, Level 1;

b. Drywell Pressure - High; and

d. Reactor Vessel Water Level - Low, Level 3 (Confirmatory).

The footnote will be the same as above.

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C. In Units 1, 2, and 3 TS 3.3.5.2, Reactor Core Isolation Cooling (RCIC) System Instrumentation, Table 3.3.5.2-1, "Reactor Core Isolation Cooling System Instrumentation,"

a new footnote will be added to the following instrument function:

1. Reactor Vessel Water Level - Low Low, Level 2.

The footnote will be the same as above.

D. In Units 1, 2, and 3 TS 3.3.6.1, Primary Containment Isolation Instrumentation, Table 3.3.6.1-1, "Primary Containment Isolation Instrumentation," a new footnote will be added to the following Main Steam Line Isolation instrument function:

1.b Main Steam Line Pressure - Low.

The footnote will be the same as above.

A mark-up of the TS showing the proposed changes is provided in . A mark-up of the TS Bases showing the proposed changes is provided in Enclosure 3.

3.0 BACKGROUND

3.1 Reason for the Proposed Changes The underlying reason for the Units 1, 2, and 3 TS change is to resolve NRC concerns regarding TVA's setpoint methodology. NRC expressed concerns regarding TVA's setpoint methodology in Reference 1. NRC has stated that these concerns must be addressed as part of the review of several proposed TS amendments (References 1 and 2). Including this information in the TS will ensure control of critical instrument setpoints and compliance with 10 CFR 50.36.

Analytical Limits represent the values used in the safety analyses to demonstrate that automatic protective actions prevent the plant from exceeding a Safety Limit or 10 CFR 50.46 limit.

Typically, the Analytical Limits do not account for instrument characteristics such as drift, repeatability, accident induced error, etc. These phenomena are accounted for in the instrument setpoint calculations. Instrument setpoint and scaling calculations utilize the Analytical Limits to establish the nominal Trip Setpoint, Acceptable As Left (AAL) band, Acceptable As Found (AAF) band and the Allowable Value. The Allowable Value is the value in the TS. If the instrument actuates under normal plant conditions or during a surveillance test at or before the E1-4

Allowable Value, the setpoint calculation demonstrates that the instrument would actuate at or before the Analytical Limit under transient or accident conditions, and thus the Safety Limit or 10 CFR 50.46 limit would not be exceeded.

The AAL band accounts for uncertainties such as drift and repeatability so that an instrument would be expected to remain within the AAF band when tested at the end of the next surveillance interval. The AAF band represents the expected band of instrument performance. If an instrument is outside the AAF band, but conservative with respect to the Allowable Value, it would still perform its as designed function and thus could be considered operable, but would be evaluated to determine why its performance is outside the expected AAF band. The initial evaluation, which is performed prior to returning the instrument to operation, would be performed to show that the instrument is not degrading such that it might not function as designed during the next interval of operation.

Instruments which perform outside the Allowable Value during surveillance testing may not be able to perform its design function during a transient or an accident, and thus would be declared inoperable until actions are taken to ensure the channel will perform as designed.

Requiring the channel setpoint to be reset to a value that is within the acceptable as-left tolerance after any instrument calibration is necessary to ensure the channel is in conformance with the assumptions of the supporting instrument setpoint and scaling calculation.

3.2 Safety Limits and 10 CER 50.46 Requirements Safety Limits are defined in Section 2.1 of the TS as:

Reactor Core Safety Limits:

- With the reactor steam dome pressure less than 785 psig or core flow less than 10 percent rated core flow, thermal power shall be less than or equal to 25 percent rated thermal power (Safety Limit 2.1.1.1).

- With the reactor steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10 percent rated core flow, the minimum value of the critical power ratio (MCPR) shall be greater than or equal to (reload specific value) for two recirculation loop operation or greater than or equal to (reload specific value) for single loop operation (Safety Limit 2.1.1.2).

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- Reactor vessel water level shall be greater than the top of active irradiated fuel (Safety Limit 2.1.1.3).

Reactor Coolant System (RCS) Pressure Safety Limit:

- Reactor steam dome pressure shall be less than or equal to 1325 psig (Safety Limit 2.1.2).

With regards to the reactor core Safety Limits, operation with core thermal power below 25% of rated without thermal margin surveillance is conservatively acceptable for complying with the Safety Limits even for reactor operations at natural circulation.

Adequate fuel thermal margins are also maintained without further surveillance for the low power conditions that would be present if core natural circulation is below 10% of rated flow.

Critical power correlations are applicable for calculations at pressure greater than 785 psig and core flow greater than 10 percent rated core flow. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Although it is recognized that the onset of transition boiling would not result in damage to Boiling Water Reactor fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit.

The reactor vessel water level Safety Limit has been established at the top of the active fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

With regards to the RCS pressure Safety Limit, reactor steam dome pressure protects the reactor coolant system against overpressurization. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity.

10 CFR 50.46 provides the acceptance criteria for ECCS for light-water nuclear power reactors. 10 CFR 50.46(b)(1) requires the calculated maximum fuel element cladding temperature not exceed 22000F.

3.3 Protection Function Description Protection functions are designed to initiate appropriate responses from systems to ensure that Safety Limits are maintained in the event of a design basis accident or transient.

This is achieved by specifying Limiting Safety System Settings (LSSS), as well as Limiting Conditions for Operation (LCO) on other reactor system parameters, and equipment performance. TS El-6

are required by 10 CFR 50.36 to contain LSSS defined by the regulation as "...settings for automatic protective devices...so chosen that automatic protective action will correct the abnormal situation before a Safety Limit (SL) is exceeded." For BFN, the LSSS are defined in TS Bases 3.3.1.1 as the Allowable Values, which, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits, including Safety Limits during design basis accidents. The Analytical Limit is the limit of the process variable at which a safety action is initiated, as established by the safety analysis, to ensure that a Safety Limit is not exceeded. Any automatic protection action that occurs on or before reaching the Analytical Limit therefore ensures that the Safety Limit is not exceeded. However, in practice, the actual trip setpoints for automatic protective devices must be chosen to be more conservative than the Analytical Limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

The trip setpoint for a protective device is chosen to ensure automatic actuation prior to the process variable reaching the Analytical Limit and thus ensuring that the Safety Limit would not be exceeded. As such, the trip setpoint accounts for uncertainties in setting the device (e.g., calibration),

uncertainties in how the device might actually perform (e.g.,

repeatability), changes in the point of action of the device over time (e.g., drift during surveillance intervals), and any other factors which may influence its actual performance (e.g., harsh accident environments). In this manner, the trip setpoint ensures that Safety Limits are not exceeded.

At BFN, the nominal Trip Setpoints are specified by design documents, which require an evaluation under the provisions of 10 CFR 50.59 before they can be changed.

3.4 System Description The instrumentation affected by this TS change involves the following systems:

Reactor Protection System,

Emergency Core Cooling Systems,

Reactor Core Isolation Cooling, and

Primary Containment Isolation.

A brief description of each system and the affected instrumentation is provided below:

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Reactor Protection System The RPS initiates a reactor scram when one or more monitored parameters exceed their specified limits, to preserve the integrity of the fuel cladding and the RCS, and minimize the energy that must be absorbed following a loss of coolant accident. This can be accomplished either automatically or manually. The protection and monitoring functions of the RPS have been designed to ensure safe operation of the reactor. This is achieved by specifying LSSSs in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance. The LSSS are defined in this specification as the Allowable Values, which in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits, including Safety Limits during Design Basis Accidents and operational transients.

Additional information regarding the RPS is provided in UFSAR Section 7.2.

The drift susceptible RPS instrumentation affected by this TS change are:

Reactor Vessel Steam Dome Pressure - High, which maintains compliance with Safety Limit 2.1.1.2;

Reactor Vessel Water Level - Low, Level 3, which maintains compliance with Safety Limit 2.1.1.3 and the fuel peak cladding temperature acceptance criterion of 10 CFR 50.46; and

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low, which maintains compliance with Safety Limit 2.1.1.2.

Each is discussed below:

An increase in the Reactor Pressure Vessel (RPV) pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and thermal power to increase, which could challenge the integrity of the fuel cladding and the reactor coolant pressure boundary. The Reactor Vessel Steam Dome Pressure - High function initiates a scram for transients that result in a pressure increase, thereby counteracting the pressure increase by rapidly reducing core power.

Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel El-8

from fission. The Reactor Vessel Water Level - Low, Level 3 function is assumed in the accident analysis of the recirculation line break. The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Fast closure of the Turbine Control Valves (TCVs) results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves. The Turbine Control Valve Fast Closure, Trip Oil Pressure - Low function is the primary scram signal for the generator load rejection event. For this event, the reactor scram reduces the amount of energy required to be absorbed.

Emergency Core Cooling Systems As discussed above, the ECCS coupled with a reactor scram ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. As shown in Figure 1, the BFN ECCS consists of the following:

Core Spray;

Low Pressure Coolant Injection, which is an operating mode of Residual Heat Removal (RHR) (The other operating modes of RHR include shutdown cooling, containment spray and pool cooling, standby cooling, and supplemental fuel pool cooling.);

High Pressure Coolant Injection; and

Automatic Depressurization System.

These systems are designed to limit clad temperature over the complete spectrum of possible break sizes in the nuclear system process barrier, including the design basis break. The design basis break is defined as the complete and sudden circumferential rupture of the largest pipe connected to the reactor vessel (i.e., one of the recirculation loop pipes) with displacement of the ends so that blowdown occurs from both ends.

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FIGURE 1 LAYOUT OF THE UNIT I EMERGENCY CORE COOLING SYSTEM A

I ----- TO TORUS 10 LPCI LPCI INJECTION INJECTION VALVE VALVE 1-FCV-74-53 1-FCV-74-67 f I I

DISCHARCE VALVE I-FCV.68-79 1 FCV468-RECIRC RECIRC PUW 9 PUW A The Units 2 and 3 design is similar.

The ECCS affected by this TS change are described below.

1. The Core Spray System instruments, which maintain compliance with the fuel peak cladding temperature acceptance criterion of 10 CFR 50.46, are:

Reactor Vessel Water Level - Low Low Low, Level 1;

Drywell Pressure - High; and

Reactor Steam Dome Pressure - Low (Injection Permissive and ECCS Initiation).

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2. The LPCI instruments, which maintain compliance with the fuel peak cladding temperature acceptance criterion of 10 CFR 50.46, are:

Reactor Vessel Water Level - Low Low Low, Level 1;

Drywell Pressure - High;

Reactor Steam Dome Pressure - Low (Injection Permissive and ECCS Initiation); and

Reactor Steam Dome Pressure - Low (Recirculation Discharge Valve Permissive).

3. The following HPCI instruments:

Reactor Vessel Water Level - Low Low, Level 2, which maintains compliance with Safety Limit 2.1.1.3 and the fuel peak cladding temperature acceptance criterion of 10 CFR 50.46; and

Drywell Pressure - High, which maintains compliance with the fuel peak cladding temperature acceptance criterion of 10 CFR 50.46.

4. The ADS instruments, which maintain compliance with the fuel peak cladding temperature acceptance criterion of 10 CFR 50.46, are:

Reactor Vessel Water Level - Low Low Low, Level 1;

Drywell Pressure - High; and

Reactor Vessel Water Level - Low, Level 3 (Confirmatory).

Low RPV water level and high drywell pressure indicate that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. The low pressure ECCS are initiated to ensure that core spray and flooding functions are available to prevent or minimize fuel damage. Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS subsystems. This ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure.

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The Core Spray system consists of two independent loops. Each loop consists of two pumps, a spray sparger inside the core shroud and above the core, piping and valves to convey water from the pressure suppression pool to the sparger, and the associated controls and instrumentation. When the system is actuated, water is taken from the pressure suppression pool. Flow then passes through a normally open motor-operated valve in the suction line to each 50 percent capacity pump.

The LPCI is an operating mode of the RHR System, with two LPCI subsystems. During LPCI operation, the four RHR pumps take suction from the pressure suppression pool and discharge to the reactor vessel into the core region through both of the recirculation loops. Two pumps discharge to each recirculation loop. Once an initiation signal is received by the LPCI control circuitry, the signal is sealed in until manually reset.

The HPCI System consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger.

Suction piping for the system is provided from the Condensate Storage Tank (CST) and the suppression pool. The HPCI System may be initiated by either automatic or manual means.

In case the capability of the HPCI is not sufficient to maintain the reactor water level, the ADS functions to reduce the reactor pressure so that flow from the LPCI and the Core Spray System enters the reactor vessel in time to cool the core and limit fuel cladding temperature. The ADS uses six of the nuclear system main steam relief valves to relieve the high pressure steam to the pressure suppression pool.

Additional information regarding the ECCS is provided in UFSAR Chapter 6.

Reactor Core Isolation Cooling The RCIC System is designed to operate either automatically or manually following RPV isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of the RPV water level. Under these conditions, the HPCI and RCIC systems perform similar functions.

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The RCIC System consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger.

Suction piping is provided from the CST and the suppression pool.

Pump suction is normally aligned to the CST to minimize injection of suppression pool water into the RPV.

The RCIC instrumentation function affected by this TS change is the Reactor Vessel Water Level - Low Low, Level 2, which maintains compliance with Safety Limit 2.1.1.3. Additional information regarding the RCIC system is provided in UFSAR Section 4.7.

Primary Containment Isolation Low main steam line pressure with the reactor at power indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV cooling down more than 100'F/hr if the pressure loss is allowed to continue. The primary containment isolation system instrumentation affected by this TS change is the Main Steam Line Pressure - Low function, which is directly assumed in the analysis of the pressure regulator failure. For this event, the closure of the Main Steam Isolation Valves (MSIVs) ensures that the RPV temperature change limit (100 degrees F/hr) is not reached. In addition, this function supports actions to ensure that Safety Limit 2.1.1.1 is not exceeded. This function closes the MSIVs prior to pressure decreasing below 785 psig, which results in a scram.

Additional information regarding the Primary Containment Isolation system is provided in UFSAR Section 7.3.

4.0 TECHNICAL ANALYSIS

4.1 Setpoint Methodology As evidenced below, TVA's method for performing setpoint calculations and the implementing programmatic controls were previously explicitly reviewed and approved by NRC as part of the BFN docket.

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Prior to Unit 2 restart, TVA developed its current setpoint calculation methodology based on Method 3 of Instrument Society of America (ISA) S67.04.02. NRC (including NRR staff) performed an inspection (Reference 3) to assess the adequacy of the testing, calibration, maintenance, and configuration control of safety-related instrumentation. Section 5 of Inspection Report 89-06 states:

"The latest procedure used by the licensee for setpoint calculations is the Division of Nuclear Engineering (DNE),

Electrical Engineering Branch (EEB), instruction EEB-TI-28, Revision 1, dated October 24, 1988. ... Procedure EEB-TI-28 incorporates the guidance found in RG 1.105 and ISA Standard 67.04 and is acceptable for assuring that setpoints are established and held within specified limits for nuclear safety-related instruments used in nuclear power plants.

The guidance provided by this procedure was reflected in the setpoint calculations which were reviewed during this inspection and are identified in the scope paragraph. The methodology of determining instrument loop errors and using them in the accuracy calculation reviewed is acceptable."

In order to support the restart of BFN Unit 2, TVA submitted (Reference 4) a request to revise the TS low water level setpoint. On January 2, 1991, NRC-approved (Reference 5) the requested amendment, which stated:

"The amendment changes the Technical Specifications (TS) to incorporate a revised trip setpoint for the Level 1 low reactor pressure vessel (RPV) water level based on new calculation methodology."

In addition, the accompanying Safety Evaluation stated:

"TVA performed a Setpoint and Scaling Calculation to determine the accuracy of the instruments and loops. This accuracy was compared to the required accuracies to assure that there is sufficient margin between the setpoints and the operating limits, and the Safety Limits. The calculations reviewed by the staff at TVA's Rockville offices were as follows:

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Instrument No. Calculation No. Revision No.

2-LT-3-56A ED-Q2003-88122 3 2-LT-3-56B ED-Q2003-88123 3 2-LT-3-56C ED-Q2003-88124 3 2-LT-3-56D ED-Q2003-88125 3 2-LT-3-58A ED-Q2003-880126 4 2-LT-3-58B ED-Q2003-880127 4 2-LT-3-58C ED-Q2003-880128 4 2-LT-3-58D ED-Q2003-880129 4 The staff's review of the calculations verified that TVA addressed instrument and loop errors for normal operation and accident conditions. ... The methodology for determination of instrument setpoints used by TVA was in accordance with Regulatory Guide (RG) 1.105 that endorses Instrument Society of America (ISA) Standard ISA-S67.04 - 1982 "Setpoint for Nuclear Safety Related Instrumentation Used in Nuclear Power Plants". This standard provides guidance for ensuring that setpoints stay within TS limits. ...

The proposed changes to the LSSS (limiting safety system setting) and SL (Safety Limit) settings were deemed acceptable because they are based on a value derived by approved calculational means. This change ensures that trips occur within the analytical limit used to confirm the design bases of the plant."

This NRC approved setpoint methodology continues to be used and has formed the basis for subsequent NRC approval of TS changes.

For example, the NRC approved (Reference 6) a change in the reactor vessel water level Safety Limit and limiting safety system setting for BFN Units 1 and 3 by Amendments 222 and 196, respectively. The Safety Evaluation states:

"The methodology used by the licensee to determine the LSSS is in accordance with the Instrument Society of America Standard ISA-S67.04 - 1982 "Setpoints for Nuclear Safety Related Instrumentation Used in Nuclear Power Plants." This methodology is consistent with the guidance of Regulatory Guide 1.105. Therefore, the proposed LSSS is acceptable."

In addition, as part of NRC's review of the Units 1, 2, and 3 TS for a 24-month fuel cycle (Reference 7), NRC endorsed TVA's method of evaluation of as-found and as-left values in TVA's maintenance program and TVA's method of addressing failures through the corrective action program.

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4.2 Compliance with Current Regulations and Commitments As discussed above, NRC has previously concluded that the methodology used by TVA to determine the LSSS is in accordance with the Instrument Society of America Standard ISA-S67.04 - 1982 "Setpoints for Nuclear Safety Related Instrumentation Used in Nuclear Power Plants" and that this methodology is consistent with the guidance of Regulatory Guide 1.105.

TVA has reviewed its Licensing Basis and determined that no commitments are affected by this proposed change.

5.0 REGULATORY SAFETY ANALYSIS The Tennessee Valley Authority (TVA) is submitting an amendment request to licenses DPR-33, DPR-52, and DPR-68 for Units 1, 2, and 3, respectively. The scope of this proposed Technical Specification (TS) change includes those drift susceptible instruments, which are either necessary to ensure compliance with a Safety Limit or critical in ensuring the fuel peak cladding temperature acceptance criterion of 10 CFR 50.46 are met. The proposed change adds a footnote that specifies the actions to be taken for the applicable as-found instrument setpoints and references a discussion of the NRC-approved TVA setpoint methodology. The purpose of including this information in the TS is to control critical instrument setpoints and ensure compliance with 10 CFR 50.36.

5.1 No Significant Hazards Consideration TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment", as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No EI-16

Including references to TVA's methodology for determining, setting, and evaluating as-found instrument setpoints in the TS is an administrative change. There will be no change to the manner in which Safety Limits, Analytical Limits, or Allowable Values are determined. No changes are proposed in the manner in which the Reactor Protection System (RPS), Emergency Core Cooling System (ECCS),

Reactor Core Isolation Cooling (RCIC), or Primary Containment Isolation systems provide plant protection or which create new modes of plant operation.

The proposed request will not affect the probability of any event initiators. There will be no degradation in the performance of, or an increase in the number of challenges imposed on, safety-related equipment assumed to function during an accident situation. There will be no change to normal plant operating parameters or accident mitigation performance.

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No There are no hardware changes nor are there any changes in the method by which any plant system performs a safety function. This request does not affect the normal method of plant operation. The proposed amendment does not introduce new equipment, which could create a new or different kind of accident.

No new external threats, release pathways, or equipment failure modes are created. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this request. Therefore, the implementation of the proposed amendment will not create a possibility for an accident of a new or different type than those previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No EI-17

Including references to TVA's methodology for determining, setting, and evaluating as-found instrument setpoints in the TS is an administrative change. No changes are proposed in the manner in which the RPS, ECCS, RCIC, or Primary Containment Isolation systems satisfy the Updated Final Safety Analysis Report requirements for accident mitigation or unit safe shutdown. There will be no change to Safety Limits, Analytical Limits, Allowable Values, or post-Loss Of Coolant Accident peak clad temperatures. For these reasons, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, TVA concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria As discussed above, NRC has previously concluded that the methodology used by TVA to determine the LSSS is in accordance with the Instrument Society of America Standard ISA-S67.04 - 1982 "Setpoints for Nuclear Safety Related Instrumentation Used in Nuclear Power Plants" and that this methodology is consistent with the guidance of Regulatory Guide 1.105.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(9). Therefore, pursuant to 10 CFR 51.22(b), no El-18

environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. NRC letter to TVA, dated January 6, 2005, "Browns Ferry Units 1, 2, and 3 - Request for Information Regarding Status of Amendments Using Method 3 (TAC Nos. MC1330, MC1427, MC2305, MC3812, MC4070, MC4071, MC4072, MC4161, MC3743, and MC3744)."

2. NRC letter to TVA, dated April 19, 2005, "Browns Ferry Nuclear Plant, Unit 1 - Licensing Action Status and Interdependencies (TAC Nos. MC1330, MC1427, MC2305, MC3812, MC3813, MC3822, MC3960, MC4070, MC4161, MC4659, MC4797, MC5254, and MC5373)."

3. NRC letter to TVA, dated May 8, 1989, "Notice of Violation (NRC Inspection Report Nos. 50-259/89-06, 50-260/89-06 and 50-296/89-06) ."

4. TVA letter to NRC, dated August 6, 1990, "Browns Ferry Nuclear Plant (BFN) - Unit 2 - TVA BFN Technical Specification (TS) No. 291 - Revision to Level 1 Low Reactor Pressure Vessel (RPV) Water Level."

5. NRC letter to TVA, dated January 2, 1991, "Issuance of Amendment (TAC No. 77279) (TS 291)."

6. NRC letter to TVA, dated July 17, 1995, "Issuance of Technical Specification Amendments for the Browns Ferry Nuclear Plant Units 1, 2, and 3 (TAC NOS. M89248, M89249 and M89250) (TS 318) ."

7. NRC letter to TVA, dated November 30, 1998, "Issuance of Amendments - Browns Ferry Nuclear Plan Units 1, 2, and 3 (TAC Nos. MA2081, MA2082, and MA2083)."

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Enclosure 2 Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3 Technical Specifications (TS) Change TS-453 Instrument Setpoint Program Proposed Technical Specification Changes (mark-up)

The following footnote will be added where indicated on the TS pages:

During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable.

The nominal Trip Setpoint shall be specified on design output documentation. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band shall be specified in Chapter 7 of the Updated Final Safety Analysis Report.

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

2. Average Power Range Monitors (continued)

d. Downscale F SR 3.3.1.1.7 2 3% RTP SR 3.3.1.1.8 SR 3.3.1.1.14

e. Inop 1,2 G SR 3.3.1.1.7 NA SR 3.3.1.1.8 SR 3.3.1.1.14

3. Reactor Vessel earn Dome 1,2 G SR 3.3.1.1.1 s1055 psig SR 3.3.1.1.8 Pressure - Hig SR 3.3.1.1.10 I

4. Reactor VeKWater Level - 1,2 G SR 3.3.1.1.1 2 538 inches SR 3.3.1.1.8 above vessel Low, LevelId SR 3.3.1.1.13 zero I

5. Main Steam Isolation Valve - F SR 3.3.1.1.8

10% dosed Closure SR 3.3.1.1.13 SR 3.3.1.1.14

6. Drywell Pressure - High 1,2 G SR 3.3.1.1.8

2.5 psig SR 3.3.1.1.13 SR 3.3.1.1.14 1,2

7. Scram Discharge Volume Water Level - High

a. Resistance Temperature 1,2 G SR 3.3.1.1.8

50 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5 (a)

H SR 3.3.1.1.8

50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14

b. Float Switch 1,2 G SR 3.3.1.1.8

50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 5 (8) H SR 3.3.1.1.8

50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(d) INSERT FOOTNOTE I

B F N - NI T 3.3 7 Am nd me t No 312 BFN-UNIT 1 3.3-7 Amendment No. 234

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

8. Turbine Stop Valve - 230%RTP 4 E SR 3.3.1.1.8 s10% closed Closure SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.15

9. Turbine Control Valve Fast 2 30% RTP 2 E SR 3.3.1.1.8 2 550 psig Clg5c, Trip Oil Pressure - SR 3.3.1.1.13 LrLd) SR 3.3.1.1.14 I L SR 3.3.1.1.15

10. Reactor Mode Switch - 1,2 1 G SR 3.3.1.1.12 NA Shutdown Position SR 3.3.1.1.14 5 (a) 1 H SR 3.3.1.1.12 NA SR 3.3.1.1.14

11. Manual Scram 1,2 1 G SR 3.3.1.1.8 NA SR 3.3.1.1.14 5(a) I H SR 3.3.1.1.8 NA SR 3.3.1.1.14

12. RPSChannelTestSwitches 1,2 2 G SR 3.3.1.1.4 NA 5 (a) 2 H SR 3.3.1.1.4 NA (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(d) INSERT FOOTNOTE BFN-UNIT 1 3.3-8 Amendment No. 234

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION A.1

1. Core Spray System

a. Reactor Vessel Water 1,2,3 4 (b)

B SR 3.3.5.1.1 2!398 Inches Level -tow Low Low, 4 (a), 5 (a)

SR 3.3.5.1.2 above vessel Level 14e SR 3.3.5.1.5 zero I SR 3.3.5.1.6

b. DrynA Pressure 1,2,3 4 (b) B SR 3.3.5.1.2

2.5 psig Highe SR 3.3.5.1.5 I SR 3.3.5.1.6

c. Reactor Steam Dome 1,2,3 4 (b) C SR 3.3.5.1.2 2 435 psig Pressure - Low (injection 2 per trip SR 3.3.5.1.A and Permissiv' nd ECCS system SR 3.3.5.1.6 5 465 psig Initiation)ff I 4(a), 5(a) 4 B SR 3.3.5.1.2 > 435 psig 2 per trip SR 3.3.5.1.4 and system SR 3.3.5.1.6

465 psig

d. Core Spray Pump 1,2,3, 2 E SR 3.3.5.1.2 > 1647gpm Discharge Flow - Low 4(a) 5(a) 1 per SR 3.3.5.1.5 and (Bypass) subsystem

2910 gpm

e. Core Spray Pump Start -

Time Delay Relay Pumps AB,C,D (with 1,2,3, 4 C SR 3.3.5.1.5 t 6 seconds diesel power) 4(a), 5(a) I per pump SR 3.3.5.1.6 and

8 seconds Pump A (with normal 1,2,3, 1 C SR 3.3.5.1.5 2 0 seconds power) SR 3.3.5.1.6 and 4 (a), 5(a)

1 second Pump B (with normal 1,2,3, 1 C SR 3.3.5.1.5 26seconds power) 4(5), 5(a) SR 3.3.5.1.6 and

8 seconds Pump C (with normal 1,2,3, i C SR 3.3.5.1.5 212seconds power) 4(a), 5 (a) SR 3.3.5.1.6 and

16 seconds (continuedl (a) When associated subsystem(s) are required to be OPERABLE.

(b) Channels affect Common Accident Signal Logic. Refer to LCO 3.8.1, 'AC Sources - Operating."

(e) INSERT FOOTNOTE I

BFN-UNIT 1 3.342 Amendment No. 234

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION AI

1. Core Spray System (continued)

e. Core Spray Pump Start -

Time Delay Relay (continued)

Pump D (with normal power) 1,2,3, I C SR 3.3.5.1.5 218seconds 4(a), 5 (a) SR 3.3.5.1.6 and 5 24 seconds

a. Reactor Vessel Water Lel 1,2,3, 4 B SR 3.3.5.1.1 2 398 inches

- Low Low Low, Level i2) SR 3.3.5.1.2 above vessel I 4(a), 5 (a)

SR 3.3.5.1.5 zero SR 3.3.5.1.6

b. Drywell Pressure - HigG 1,2,3 4 B SR 3.3.5.1.2 *2.5psig SR 3.3.5.1.5 SR 3.3.5.1.6

c. Reactor Steam Dome 1,2,3 4 C SR 3.3.5.1.2 Ž 435 psig Pressure - Low (Injection SR 3.3.5.1.4 and Pemmissiynd ECCS SR 3.3.5.1.6

465 psig initiatiorUe 5(8) 4 B SR 3.3.5.1.2 2 435 psig 4 (a),

SR 3.3.5.1.4 and SR 3.3.5.1.6 5 465 psig

d. Reactor Steam Dome 1(C),2(C), 4 C SR 3.3.5.1.2 2 215 psig Pressure - Low SR 3.3.5.1.4 and (Recirculation DisApge 3 (c) SR 3.3.5.1.6

245 psig Valve Permissivef428P

e. Reactor Vessel Water Level 1,2.3 2 B SR 3.3.5.1.1 23125/16

- Level 0 1 per SR 3.3.5.1.2 inches above subsystem SR 3.3.5.1.5 vessel zero SR 3.3.5.1.6 (continued)

(a) When associated subsystem(s) are required to be OPERABLE.

(b) Deleted.

(c) Wth associated recirculation pump discharge valve open.

I (e) INSERT FOOTNOTE I I BFN-UNIT 1 3.3-43 Amendment No. 234

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION A.1

2. LPCI System (continued)

f. Low Pressure Coolant Injection Pump Start - Time Delay Relay Pump A,B,C.D (with diesel 1,2,3, 4 C SR 3.3.5.1.5 Ž0seconds power) 4 (a), 5 (a) SR 3.3.5.1.6 and

I second Pump A(with normal power) 1,2,3, 1 C SR 3.3.5.1.5 Ž0 seconds 4 (a), 5 (a) SR 3.3.5.1.6 and 5 1 second Pump 8 (with normal power) 1,2,3, 1 C SR 3.3.5.1.5 2 6 seconds 4 (a) 5 (a) SR 3.3.5.1.6 and

8 seconds Pump C (with normal power) 1,2,3, 1 C SR 3.3.5.1.5 2 12 seconds 4 (a), 5 (a) SR 3.3.5.1.6 and

16 seconds Pump D (with normal power) 1,2,3, 1 C SR 3.3.5.1.5 k18seconds 4 (a), 5 (a) SR 3.3.5.1.6 and 5 24 seconds

a. Reactor Vessel WakLevel 1, 4 B SR 3.3.5.1.1 Ž470inches

- Low Low, Level 2IV 2 (d), 3 (d) SR 3.3.5.1.2 above vessel SR 3.3.5.1.5 zero SR 3.3.5.1.6 (continued)

(a) When the associated subsystem(s) are required to be OPERABLE.

(d) Wth reactor steam dome pressure > 150 psig.

K(e) INSERT FOOTNOTE

-3 I I BFN-UNIT 1 3.3-44 Amendment No.-234 7 250 April 1, 2004

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 4 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION Al

3. HPCI System (continued)

b. Drywell Pressure - HighG 1, 4 B SR 3.3.5.1.2 *2.5 psig 2(d),3(d) SR 3.3.5.1.5 SR 3.3.5.1.6

c. Reactor Vessel Water Level 1, 2 C SR 3.3.5.1.1 s583inches

- High, Level 8 2 (d) 3 (d) SR 3.3.5.1.2 above vessel SR 3.3.5.1.5 zero SR 3.3.5.1.6

d. CondensateHeaderLevel - 1, 1 D SR 3.3.5.1.2 2 Elev. 551 Low 2 (d) 3 (d) SR 3.3.5.1.3 feet SR 3.3.5.1.6

e. Suppression Pool Water 1, 1 D SR 3.3.5.1.2 s7 Inches Level - High 2 (d) 3 (d) SR 3.3.5.1.3 above SR 3.3.5.1.6 instrument zero

f. High Pressure Coolant 1, 1 E SR 3.3.5.1.2 Ž671 gpm Injection Pump Discharge 2 (d) 3 (d) SR 3.3.5.1.5 Flow- Low (Bypass) SR 3.3.5.1.6

4. Automatic Depressurization System (ADS) Trip System A

a. ReactorVesselWater Level 1, 2 F SR 3.3.5.1.1 2398Inches

- Low Low Low, Level 11 2 (d) 3 (d) SR 3.3.5.1.2 above vessel SR 3.3.5.1.5 zero SR 3.3.5.1.6

b. DrywellPressure - Higto 1, 2 F SR 3.3.5.1.2 s2.5psig 2 (d) 3 (d) SR 3.3.5.1.5 SR 3.3.5.1.6

c. Automatic Depressurization 1, 1 G SR 3.3.5.1.5 S 115 System Initiation Timer 2 (d), 3 (d) SR 3.3.5.1.6 seconds (continued)

(d) With reactor steam dome pressure > 150 psig.

(e) INSERT FOOTNOTE I BFN-UNIT I 3.3-45 Amendment No. 234

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 5 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION A.1

4. ADS Trip System A (continued)

d. Reactor Vessel Water Level 1, 1 F SR 3.3.5.1.1 2:544 inches

- Low Level 3 SR 3.3.5.1.2 above vessel 2 (d), 3 (d)

(Confirmatory1 SR 3.3.5.1.5 zero I SR 3.3.5.1.6

e. Core Spray Pump Discharge 1, 4 G SR 3.3.5.1.2 Ž 175 psig Pressure - High SR 3.3.5.1.3 and 2 (d), 3 (d)

SR 3.3.5.1.6 195 psig

f. Low Pressure Coolant 1, 8 G SR 3.3.5.1.2 2 90 psig and Injection Pump Discharge 2 (d), 3 (d)

SR 3.3.5.1.3 S 110 psig Pressure - High SR 3.3.5.1.6

g. Automatic Depressurization 1, 2 G SR 3.3.5.1.5 *322 System High Drywell 2 (d), 3 (d)

SR 3.3.5.1.6 seconds Pressure Bypass Timer

5. ADS Trip System B

a. Reactor Vessel Water LeA 1, 2 F SR 3.3.5.1.1 2!398 nches

- Low Low Low, Level 1Ue 2 (d), 3 (d)

SR 3.3.5.1.2 above vessel I SR 3.3.5.1.5 zero SR 3.3.5.1.6

b. Drywell Pressure - HighE) 1, 2 F SR 3.3.5.1.2

2.5 psig 2 (d), 3 (d)

SR 3.3.5.1.5 SR 3.3.5.1.6

c. Automatic Depressurization 1, 1 G SR 3.3.5.1.5

115 System Initiation Timer SR 3.3.5.1.6 seconds 2 (d), 3 (d)

d. Reactor Vessel Water Level 1, I F SR 3.3.5.1.1 2 544 inches

- Low Level 3 SR 3.3.5.1.2 above vessel 2 (d), 3 (d)

(Confirmatory)G SR 3.3.5.1.5 zero I SR 3.3.5.1.6 (continued)

(d) With reactorsteam dome pressure> 150 psig.

I (e) INSERT FOOTNOTE BFN-UNIT 1 3.3-46 Amendment No. 234

RCIC System Instrumentation 3.3.5.2 Table 3.3.5.2-1 (page 1 of 1)

Reactor Core Isolation Cooling System Instrumentation CONDITIONS REQUIRED REFERENCED SURVEILLANCE ALLOWABLE FUNCTION CHANNELS PER FROM REQUIRED REQUIREMENTS VALUE FUNCTION ACTION A1

1. Reactor Vessel Wgr Level - 4 B SR 3.3.5.2.1 Ž470Inches Low Low, Level 2a SR 3.3.5.2.2 above vessel SR 3.3.5.2.3 zero SR 3.3.5.2.4

2. ReaclorVesselWaterLevel - 2 C SR 3.3.5.2.1 s 583inches High, Level8 SR 3.3.5.2.2 above vessel SR 3.3.5.2.3 zero SR 3.3.5.2.4 (a) INSERT FOOTNOTE BFN-UNIT 1 3.3-51 Amendment No. 234

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 1 of 3)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION C.1

1. Main Steam Une Isolation

a. Reactor Vessel Water 1,2,3 2 D SR 3.3.6.1.1 2 398 Inches Level - Low Low Low, SR 3.3.6.1.2 above vessel Level 1 SR 3.3.6.1.5 zero SR 3.3.6.1.6

b. Main Steam Unw, 2 E SR 3.3.6.1.2 2 825 psig SR 3.3.6.1.5 Pressure - Loai 1,, SR 3.3.6.1.6 I

c. Main Steam Une Flow - 1,2,3 2 per D SR 3.3.6.1.1 s 140% rated High MSL SR 3.3.6.1.2 steam low SR 3.3.6.1.5 SR 3.3.6.1.6 0

d. Main Steam Tunnel 1,2,3 8 D SR 3.3.6.1.2

200 F Temperature - High SR 3.3.6.1.5 SR 3.3.6.1.6

2. Primary Containment Isolation

a. Reactor Vessel Water 1,2,3 2 G SR 3.3.6.1.1 k 538 Inches Level - Low, Level 3 SR 3.3.6.1.2 above vessel SR 3.3.6.1.5 zero SR 3.3.6.1.6

b. Drywell Pressure - High 1,2,3 2 G SR 3.3.6.1.2

2.5 psig SR 3.3.6.1.5 SR 3.3.6.1.6

3. High Pressure Coolant Injection (HPCI) System Isolation

a. HPCI Steam Une Flow - 1,2,3 3 F SR 3.3.6.1.2

90 psi High SR 3.3.6.1.5 SR 3.3.6.1.6

b. HPCI Steam Supply Uine 1,2,3 3 F SR 3.3.6.1.2 2 100 psig Pressure - Low SR 3.3.6.1.5 SR 3.3.6.1.6

c. HPCI Turbine 1,2,3 3 F SR 3.3.6.1.2

20 psig Exhaust Diaphragm SR 3.3.6.1.5 Pressure - High SR 3.3.6.1.6 (continued)

(c) INSERT FOOTNOTE BFN-UNIT I 3.3-58 Amendment No. 234

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

2. AvragePowerRang

2. Average Power Range Monitors (continued)

d. Inop 1,2 3 (b) G SR 3.3.1.1.16 NA

e. 2-Out-Of-4 Voter 1,2 2 G SR 3.3.1.1.1 NA SR 3.3.1.1.14 SR 3.3.1.1.16

f. OPRM Upscale 1 3(b) I SR 3.3.1.1.1 NA SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.17

3. Reactor Vessel Sjram Dome 1.2 2 G SR 3.3.1.1.1 S 1090 psig SR 3.3.1.1.8 Pressure - Higt2dl SR 3.3.1.1.10 I SR 3.3.1.1.14

4. Reactor Vessl;LWater Level - 1,2 2 G SR 3.3.1.1.1 2!528 inches SR 3.3.1.1.8 above vessel Low, Level W SR 3.3.1.1.13 zero I

5. Main Steam Isolation Valve - 8 F SR 3.3.1.1.8 5 10% dosed Closure SR 3.3.1.1.13 SR 3.3.1.1.14

6. Drywell Pressure - High 11,2 2 G SR 3.3.1.1.8 S2.5 psig SR 3.3.1.1.13 SR 3.3.1.1.14

7. Scram Discharge Volume Water Level - High

a. Resistance Temperature 1,2 2 G SR 3.3.1.1.8 s 50 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5 (a) 2 H SR 3.3.1.1.8 s 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Each APRM channel provides Inputs to both trip systems.

(d) INSERT FOOTNOT E I

BFN-UNIT 2 3.3-8 Amendment No. 253, 254,28, 260 August 16, 1999

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

7. Scram Discharge Volume Water Level - High (continued)

b. FloatSwitch 1,2 2 G SR 3.3.1.1.8 s46gallons SR 3.3.1.1.13 SR 3.3.1.1.14 5 (a) 2 H SR 3.3.1.1.8 s46 gallons SR 3.3.1.1.13 SR 3.3.1.1.14

8. Turbine Stop Valve - Ž30% RTP 4 E SR 3.3.1.1.8 s10% dosed Closure SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.15

9. Turbine Control Valve Fast Ž30% RTP 2 E SR 3.3.1.1.8 2550 psig Clos e, Trip Oil Pressure - SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.15 I

10. ReactorModeSwitch - 1,2 1 G SR 3.3.1.1.12 NA Shutdown Position SR 3.3.1.1.14 5 (a) 1 H SR 3.3.1.1.12 NA SR 3.3.1.1.14

11. Manual Scram 1.2 1 G SR 3.3.1.1.8 NA SR 3.3.1.1.14 5 (a) 1 H SR 3.3.1.1.8 NA SR 3.3.1.1.14

12. RPS Channel TestSwitches 1,2 2 G SR 3.3.1.1.4 NA 5 (a) 2 H SR 3.3.1.1.4 NA 13.Deleted (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

l (d) INSERT FOOTNOTE BFN-UNIT 2 3.3-9 Amendment No. 2-8, 276 April 08, 2002

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION A.1

1. Core Spray System

a. Reactor Vessel Water 1,2,3, 4 (b) B SR 3.3.5.1.1 2 398 Inches Level ew Low Low, 4 (a), 5 (a) SR 3.3.5.1.2 above vessel Leel ) 'SR 3.3.5.1.5 zero I

LevelSR 3.3.5.1.6

b. DryvAlLPressure - 1,2,3 4 (b) B SR 3.3.5.1.2 S2.5 psig HigO SR 3.3.5.1.5 I 9 SR 3.3.5.1.6

c. Reactor Steam Dome 1,2,3 4 (b) C SR 3.3.5.1.2 2435 psig Pressure - Low (Injection 2 per trip SR 3.3.5.1.4 and Permissiv nd ECCS system SR 3.3.5.1.6 *465 psig I

4 (a), 5 (a) 4 B SR 3.3.5.1.2 Ž435psig 2pertrip SR 3.3.5.1.4 and system SR 3.3.5.1.6 s 465psig

d. Core Spray Pump 1,2,3, 2 E SR 3.3.5.1.2 21647 gpm Discharge Flow - Low 4(a), 5 (a) I per SR 3.3.5.1.5 and (Bypass) subsystem s 2910 gpm

e. Core Spray Pump Start -

Time Delay Relay Pumps A,BC,D (with 1,2,3, 4 C SR 3.3.5.1.5 26seconds diesel power) 4 (a)I 5 (a) per pump SR 3.3.5.1.6 and s 8 seconds Pump A (with normal 1,2,3, 1 C SR 3.3.5.1.5 20 seconds Per) 4 (a) 5 (a) SR 3.3.5.1.6 and S 1 second Pump B (with normal 1,2,3, 1 C SR 3.3.5.1.5 Ž6 seconds power) 4 (a) s(a) SR 3.3.5.1.6 and s 8 seconds Pump C (with normal 1,2,3, 1 C SR 3.3.5.1.5 2 12 seconds power) 4 (a) s(a) SR 3.3.5.1.6 and s 16 seconds (continued)

(a) When associated subsystem(s) are required to be OPERABLE.

(b) Channels affect Common Accident Signal Logic. Refer to LCO 3.8.1, 'AC Sources - Operating."

I (a) INSERT FOOTNOTE I I

BFN-UNIT 2 3.3-43 Amendment No. 253

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION Al

1. Core Spray System (continued)

e. Core Spray Pump Start -

Timne Delay Relay (continued)

Pump D (with normal power) 1,2,3, I C SR 3.3.5.1.5 Ž18seconds SR 3.3.5.1.6 and 4 (a), 5 (a)

24 seconds

a. Reactor Vessel Water Le I 1,2,3, 4 B SR 3.3.5.1.1 2 398 inches

- Low Low Low, Level Ie9 (a), 5 (a) SR 3.3.5.1.2 above vessel I 4

SR 3.3.5.1.5 zero SR 3.3.5.1.6

b. Drywell Pressure - Highl 1,2,3 4 B SR 3.3.5.1.2

2.5 psig SR 3.3.5.1.5 SR 3.3.5.1.6

c. Reactor Steam Dome 1,2,3 4 C SR 3.3.5.1.2 2 435 psig Pressure - Low (injection SR 3.3.5.1.4 and Permissh>nd ECCS SR 3.3.5.1.6

465 psig InitiationAV) I

.4 B SR 3.3.5.1.2 Ž 435 psig 4 (a), 5(a)

SR 3.3.5.1.4 and SR 3.3.5.1.6

465 psig

d. Reactor Steam Dome 1l(C),2(C),

4 C SR 3.3.5.1.2 2 215 psig Pressure - Low SR 3.3.5.1.4 and (Recirculation Disc)krge 3 (c) SR 3.3.5.1.6

245 psig Valve Permissive"" I

e. Reactor Vessel Water Level 1,2,3 2 B SR 3.3.5.1.1 Ž3125/16

- Level O 1 per SR 3.3.5.1.2 inches above subsystem SR 3.3.5.1.5 vessel zero SR 3.3.5.1.6 (continued)

(a) When associated subsystem(s) are required to be OPERABLE.

(b) Deleted.

(c) With associated recirculation pump discharge valve open.

L - I I (e) INSERT FOOTNOTE II BFN-UNIT 2 3.3-44 Amendment No. 253

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION A.1

2. LPCI System (continued)

f. Low Pressure Coolant Injection Pump Start - Time Delay Relay Pump A,B,C,D (with diesel 1,2.3, 4 C SR 3.3.5.1.5 Ž0 seconds power) 4 (a), 5 (a) SR 3.3.5.1.6 and s 1 second Pump A (with normal power) 1,2,3, 1 C SR 3.3.5.1.5 a0 seconds 4 (a) 5 (a) SR 3.3.5.1.6 and S 1 second Pump B (with normal power) 1,2,3. 1 C SR 3.3.5.1.5 26 seconds 4 (a), 5 (a) SR 3.3.5.1.6 and S 8 seconds Pump C (with normal power) 1,2,3, 1 C SR 3.3.5.1.5 a 12 seconds 4 (a), 5(a) SR 3.3.5.1.6 and s 16 seconds Pump D (with normal power) 1,2,3, 1 C SR 3.3.5.1.5 a 18 seconds 4 (a), 5 (a) SR 3.3.5.1.6 and s 24 seconds

a. ReactorVesselWaILevel 1, 4 B SR 3.3.5.1.1 Ž470inches

- Low Low, Level ;e 2 (d), 3 (d) SR 3.3.5.1.2 above vessel SR 3.3.5.1.5 zero SR 3.3.5.1.6 (continued)

(a) When the associated subsystem(s) are required to be OPERABLE.

(d) With reactor steam dome pressure > 150 psig.

I (e)INSERT FOOTNOTE

-1 I BFN-UNIT 2 3.3-45 Amendment No.-263, 289 April 1, 2004

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 4 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION A.1

3. HPCI System (continued)

b. Drywell Pressure - Higt0 1, B SR 3.3.5.1.2 S 2.5 psig SR 3.3.5.1.5 2 (d), 3 (d)

SR 3.3.5.1.6

c. Reactor Vessel Water Level 1, C SR 3.3.5.1.1 : 583 inches

- High, Level 8 2 (d), 3 (d)

SR 3.3.5.1.2 above vessel SR 3.3.5.1.5 zero SR 3.3.5.1.6

d. Condensate Header Level - 1, D SR 3.3.5.1.2 2 Elev. 551 Low SR 3.3.5.1.3 feet 2 (d), 3 (d)

SR 3.3.5.1.6

e. Suppression Pool Water 1, D SR 3.3.5.1.2 S 7 inches Level - High SR 3.3.5.1.3 above 2 (d), 3 (d)

SR 3.3.5.1.6 instrument zero

f. High Pressure Coolant 1, E SR 3.3.5.1.2 Ž 671 gpm Injection Pump Discharge (d), SR 3.3.5.1.5 Flow - Low (Bypass) 2 3 (d)

SR 3.3.5.1.6

4. Automatic Depressurization System (ADS) Trip System A

a. Reactor Vessel Water Level 1,. F SR 3.3.5.1.1 2 398 inches

- Low Low Low, Level 10 2 (d), 3 (d)

SR 3.3.5.1.2 above vessel I SR 3.3.5.1.5 zero SR 3.3.5.1.6

b. Drywell Pressure - HigtG 1, F SR 3.3.5.1.2

2.5 psig 2 (d), 3 (d)

SR 3.3.5.1.5 SR 3.3.5.1.6

c. Autormatic Depressurization 1, G SR 3.3.5.1.5 r 115 System Initiation Timer SR 3.3.5.1.6 seconds 2 (d)* 3 (d)

(continued)

(d) WIth reactor steam dome pressure > 150 psig.

F(e) INSERT FOOTNOTE I

BFN-UNIT 2 3.3-46 Amendment No. 253

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 5 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION A.1

4. ADS Trip System A (continued)

d. Reactor Vessel Water Level 1, 1 F SR 3.3.5.1.1 2 528 inches

- Low, Level 3 2(d), 3(d) SR 3.3.5.1.2 above vessel (Confirmatory)o SR 3.3.5.1.5 zero I SR 3.3.5.1.6

e. Core Spray Pump Discharge 1, 4 G SR 3.3.5.1.2 2 175 psig Pressure - High 2 (d), 3 (d)

SR 3.3.5.1.3 and SR 3.3.5.1.6 *195psig

f. Low Pressure Coolant 1, 8 G SR 3.3.5.1.2 2 90 psig and Injection Pump Discharge 2 (d), 3 (d)

SR 3.3.5.1.3 *110 psig Pressure - High SR 3.3.5.1.6

g. Automatic Depressurization 1, 2 G SR 3.3.5.1.5 *322 System High Drywell SR 3.3.5.1.6 seconds 2 (d), 3 (d)

Pressure Bypass wrier

5. ADS Trip System B

a. Reactor Vessel Water Leql 1, 2 F SR 3.3.5.1.1 2 398 inches

- Low Low Low, Level 1e (d), SR 3.3.5.1.2 above vessel I 2 3 (d)

SR 3.3.5.1.5 zero SR 3.3.5.1.6

b. Drywell Pressure - High 0 1, 2 F SR 3.3.5.1.2

2.5 psig 2 (d), 3 (d)

SR 3.3.5.1.5 SR 3.3.5.1.6

c. Automatic Depressurization 1, 1 G SR 3.3.5.1.5 *115 System Initiation Timer SR 3.3.5.1.6 seconds 2 (d), 3 (d)

d. Reactor Vessel Water Level 1. I F SR 3.3.5.1.1 2 528 inches

- Low. Level 3 (d), SR 3.3.5.1.2 above vessel 2 3 (d)

(Confirmatory)o SR 3.3.5.1.5 zero I SR 3.3.5.1.6 (continued)

(d) With reactor steam dome pressure > 150 psig.

I (e) INSERT FOOTNOTE BFN-UNIT 2 3.3-47 Amendment No. 23, 260 August 16, 1999

RCIC System Instrumentation 3.3.5.2 Table 3.3.5.2-1 (page 1 of 1)

Reactor Core Isolation Cooling System Instrumentation CONDITIONS REQUIRED REFERENCED SURVEILLANCE ALLOWABLE FUNCTION CHANNELS PER FROM REQUIRED REQUIREMENTS VALUE FUNCTION ACTION A.1

1. ReactorVesselW1 r Level - 4 B SR 3.3.5.2.1 470inches Low Low Level AV SR 3.3.5.2.2 above vessel I SR 3.3.5.2.3 zero SR 3.3.5.2.4

2. Reactor Vessel Water Level - 2 C SR 3.3.5.2.1

583 inches High, Level 8 SR 3.3.5.2.2 above vessel SR 3.3.5.2.3 zero SR 3.3.5.2.4 (a) INSERT FOOTNOTE l I

BFN-UNIT 2 3.3-52 Amendment No. 253

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 1 of 3)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION C.1

1. Main SteamL Une Isolation

a. Reactor Vessel Water 1,2,3 2 D SR 3.3.6.1.1 2398 inches Level - Low Low Low, SR 3.3.6.1.2 above vessel Level 1 SR 3.3.6.1.5 zero SR 3.3.6.1.6

b. Main Steam Une 1 2 E SR 3.3.6.1.2 > 825 psig SR 3.3.6.1.5 Pressure - Lobe SR 3.3.6.1.6 I

c. Main Steam Une Flow - 1,2,3 2 per D SR 3.3.6.1.1

140% rated High MSL SR 3.3.6.1.2 steam flow SR 3.3.6.1.5 SR 3.3.6.1.6

d. Main Steam Tunnel 1,2,3 8 D SR 3.3.6.1.2 s 2000 F Temperature - High SR 3.3.6.1.5 SR 3.3.6.1.6

2. Primary Containment Isolation

a. Reactor Vessel Water 1,2,3 2 G SR 3.3.6.1.1 2 528 inches Level - Low, Level 3 SR 3.3.6.1.2 above vessel SR 3.3.6.1.5 zero SR 3.3.6.1.6

b. Drywell Pressure - High 1,2,3 2 G SR 3.3.6.1.2

2.5 psig SR 3.3.6.1.5 SR 3.3.6.1.6

3. High Pressure Coolant Injection (HPCI) System Isolation

a. HPCI Steam Line Flow - 1.2,3 1 F SR 3.3.6.1.2

90 psi High SR 3.3.6.1.5 SR 3.3.6.1.6

b. HPCI Steam Supply Une 1,2,3 3 F SR 3.3.6.1.2 t 100 psig Pressure - Low SR 3.3.6.1.5 SR 3.3.6.1.6

c. HPCI Turbine 1,2,3 3 F SR 3.3.6.1.2

20 psig Exhaust Diaphragm SR 3.3.6.1.5 Pressure - High SR 3.3.6.1.6 (continued)

I (C) INSERT FOOTNOTE To I BFN-UNIT 2 3.3-59 Amendment No. 2;3, 260 August 16, 1999

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

2. Average Power Range Monitors (continued)

d. Inop 1,2 3 (b) G SR 3.3.1.1.16 NA

e. 2-Out-Of-4 Voter 1,2 2 G SR 3.3.1.1.1 NA SR 3.3.1.1.14 SR 3.3.1.1.16 f OPRM Upscale 3 (b) I SR 3.3.1.1.1 NA SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.17

3. Reactor Vessel ,tn Dome 1,2 2 G SR 3.3.1.1.1 s 1090 psig Pressure - Higd)T SR 3.3.1.1.8 SR 3.3.1.1.10 I SR 3.3.1.1.14

4. Reactor VesWLWater Level - 1,2 2 G SR 3.3.1.1.1 > 528 Inches SR 3.3.1.1.8 above vessel Low, Level 1)

SR 3.3.1.1.13 zero I

5. Main Steam Isolation Valve - 1 8 F SR 3.3.1.1.8 s 10% dosed Closure SR 3.3.1.1.13 SR 3.3.1.1.14

6. Drywell Pressure - High 1,2 2 G SR 3.3.1.1.8 s 2.5 psig SR 3.3.1.1.13 SR 3.3.1.1.14

7. Scram Discharge Volume Water Level - High

a. Resistance Temperature 1,2 2 G SR 3.3.1.1.8 s 50 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5 (a) 2 H SR 3.3.1.1.8 s 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (continued)

(a) WIth any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Each APRM channel provides inputs to both trip systems.

(d) INSERT FOOTNOTE BFN-UNIT 3 3.3-8 Amendment No. 221 242,213,214 219 September 27, 1999

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

7. Scram Discharge Volume Water Level - High

b. FloatSwitch 1,2 2 G SR 3.3.1.1.8 s46gallons SR 3.3.1.1.13 SR 3.3.1.1.14 5 (a) 2 H SR 3.3.1.1.8 s46gallons SR 3.3.1.1.13 SR 3.3.1.1.14

8. TurbineStop Valve - 30%RTP 4 E SR 3.3.1.1.8 s10% dosed Closure SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.15

9. TurbineControlValveFast 230%RTP 2 E SR 3.3.1.1.8 2550psig CIosyu Trip Oil Pressure - SR 3.3.1.1.13 LowQd) SR 3.3.1.1.14 I SR 3.3.1.1.15

10. Reactor Mode Switch - 1,2 1 G SR 3.3.1.1.12 NA Shutdown Position SR 3.3.1.1.14 5 (a) I H SR 3.3.1.1.12 NA SR 3.3.1.1.14

11. Manual Scram 1,2 1 G SR 3.3.1.1.8 NA SR 3.3.1.1.14 5 (a) 1 H SR 3.3.1.1.8 NA SR 3.3.1.1.14

12. RPS Channel Test Switches 12 2 G SR 3.3.1.1.4 NA 5 (a) 2 H SR 3.3.1.1.4 NA

13. Deleted (a) VMth any control rod withdrawn from a core cell containing one or more fuel assemblies.

1

,I(d) INSERT FOOTNOTE BFN-UNIT 3 3.3-9 Amendment No. 212, 243, 224, 235 April 08, 2002

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDmONS FUNCTION ACTION A.1

1. Core Spray System

a. Reactor Vessel Water Le=el 1,2,3, 4 (b) B SR 3.3.5.1.1 2 398 Inches

- Low Low Low, Level W 4 (a), 5 (a)

SR 3.3.5.1.2 above vessel I SR 3.3.5.1.5 zero SR 3.3.5.1.6

b. Drywell Pressure-High 1,2.3 4 (b) B SR 3.3.5.1.2

2.5 psg I SR 3.3.5.1.5 SR 3.3.5.1.6

c. Reactor Steam Dome 1,2,3 4 (b) C SR 3.3.5.1.2 2 435 psig and Pressure - Low (Injection 2 per trip SR 3.3.5.1.4

465 psig Perrmssi nd ECCS system SR 3.3.5.1.6 Initiationk) 4(a), 5(a) 4 B SR 3.3.5.1.2 Ž 435 psig and 2 per trip SR 3.3.5.1.4

465 psig system SR 3.3.5.1.6

d. Core Spray Pump Discharge 1,2,3, 2 E SR 3.3.5.1.2 Ž 1647 gpm Flow - Low (Bypass) 4 (a), 5 (a) 1 per SR 3.3.5.1.5 and subsystem

2910 gpm

e. Core Spray Pump Start -

lime Delay Relay Pumps A,BC,D (with diesel 1,2,3, 4 C SR 3.3.5.1.5 Ž 6 seconds power) 4 (a), 5 (a)

I per pump SR 3.3.5.1.6 and

8 seconds Pump A (with normal power) 1,2,3, 1 C SR 3.3.5.1.5 2 0 seconds a), 5 (a) SR 3.3.5.1.6 and 4

1 second Pump B (with normal power) 1,2,3, 1 C SR 3.3.5.1.5 Ž 6 seconds 4 (a), 5 (a)

SR 3.3.5.1.6 and

8 seconds Pump C (with normal power) 1,2,3, I C SR 3.3.5.1.5 2 12 seconds 4(8), 5(8) SR 3.3.5.1.6 and

16 seconds (continued)

(a) When associated subsystem(s) are required to be OPERABLE.

(b) Channels affect Common Accident Signal Logic. Refer to LCO 3.8.1, 'AC Sources - Operating."

I t INSERT FOOTNOTE IAl BFN-UNIT 3 3.3-43 Amendment No. 213 September 03,1998

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION A1

1. Core Spray System (continued)

e. Core Spray Pump Start -

Time Delay Relay (continued)

Pump D (with normal power) 1,2,3, I C SR 3.3.5.1.5 2 18 seconds 4 (a), 5 (a)

SR 3.3.5.1.6 and

24 seconds

a. Reactor Vessel WaterLe I 1,2,3, 4 B SR 3.3.5.1.1 2 398 inches

- Low Low Low, Level bJ 4 (a), 5 (a) SR 3.3.5.1.2 above vessel I SR 3.3.5.1.5 zero SR 3.3.5.1.6

b. Drywell Pressure - Hige 1,2,3 4 B SR 3.3.5.1.2

2.5 psig SR 3.3.5.1.5 SR 3.3.5.1.6

c. Reactor Steam Dome 1,2,3 4 C SR 3.3.5.1.2 2 435 psig and Pressure - Low (Injection SR 3.3.5.1.4

465 psig Permiss[vAand ECCS SR 3.3.5.1.6 InitiationJ 4 B SR 3.3.5.1.2 > 435 psig and 4 (a), 5 (a)

SR 3.3.5.1.4

465 psig SR 3.3.5.1.6

d. Reactor Steam Dome 1(c)2(C) 4 C SR 3.3.5.1.2 Ž215psig Pressure - Low SR 3.3.5.1.4 and *245 psig (Recirculation Disc)arge 3 (c)

SR 3.3.5.1.6 Valve Permissivek)J I

e. Reactor Vessel Water Level 1,2,3 2 B SR 3.3.5.1.1 k 312 5116

- Level 0 1 per SR 3.3.5.1.2 inches above subsystem SR 3.3.5.1.5 vessel zero SR 3.3.5.1.6 (continued)

(a) When associated subsystem(s) are required to be OPERABLE.

(b) Deleted.

(c) With associated recirculation pump discharge valve open.

I (f) INSERT FOOTNOTE I

BFN-UNIT 3 3.3-44 Amendment No. 213 September 03, 1998

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION A.1

2. LPCI System (continued)

f. Low Pressure Coolant Injection Pump Start -lime Delay Relay Pump AB,CD (with diesel 1,2,3, 8 (e) C SR 3.3.5.1.5 20seconds power) 4 (a) 5 (a) SR 3.3.5.1.6 and

1 second Pump A (with normal power) 1.2,3, 2 C SR 3.3.5.1.5 20 seconds 4 (a), (a) I per trip SR 3.3.5.1.6 and system

I second Pump B (with normal power) 1,2,3, 2 C SR 3.3.5.1.5 Ž6 seconds 4 (a), 5 (a) 1 per trip SR 3.3.5.1.6 and system s 8 seconds Pump C (with normal power) 1,2,3, 2 C SR 3.3.5.1.5 Ž 12 seconds 4 (a), 5 (a) 1 per trip SR 3.3.5.1.6 and system s 16 seconds Pump D (with normal power) 1,2,3, 2 C SR 3.3.5.1.5 Ž 18 seconds 4 (a) 5 (a) 1 pertrip SR 3.3.5.1.6 and system s 24 seconds