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{{#Wiki_filter:TENNESSEE 1\14        VALLEY AUTHORITY Post Office Box 2000, Decatur, Alabama 35609-2000 November 26, 2021 10 CFR 50.4 10 CFR 50.59(d)(2) 10 CFR 50.71(e)
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ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296
 
==Subject:==
Summary Report for 10 CFR 50.59 Evaluations, Technical Specifications Bases Changes, Technical Requirements Manual Changes, and NRC Commitment Revisions The purpose of this letter is to provide the NRC with the following required periodic submittals:  contains the summary report of the changes, tests, and experiments at Browns Ferry Nuclear Plant (BFN) for the period of July 1, 2019, through July 1, 2021, in accordance with 10 CFR 50.59(d)(2). contains copies of changes to the BFN Technical Specifications (TS) Bases in accordance with BFN TS Section 5.5.10, Technical Specifications (TS) Bases Control Program. contains copies of changes to the BFN Technical requirements Manual (TRM) in accordance with TRM Section TR 5.1.4, Technical Requirements Control Program. contains the revised NRC commitment change that was found to meet the criteria for inclusion in the Commitment Change Summary Report. BFN revised these commitments in accordance with the Nuclear Energy Institute's (NEI) 99-04, Revision 0, "Guidelines for Managing NRC Commitment Changes," as endorsed in the NRC Regulatory Issue Summary 2000-17.
There are no new regulatory commitments contained in this letter. If you have any questions regarding this report, please contact C. L. Vaughn, Nuclear Site Licensing Manager, at (256) 729-2636.
 
U.S. Nuclear Regulatory Commission Page 2 November 26, 2021 I declare under penalty of perjury that the foregoing is true and correct. Executed on this 26th day of November 2021.
Respec ctf tful u ly, Respectfully, Maatthew Rasmussen Matthew    Rasm  mus usse sen n
Sit  Vi President Site Vice  P id t
 
==Enclosures:==
 
Enclosure 1 - 10 CFR 50.59 Summary Report Enclosure 2 - Technical Specifications (TS) Bases Changes and Additions Enclosure 3 - Technical Requirements Manual (TRM) Changes and Additions Enclosure 4 - Summary of Revised Commitments cc (w/Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

: 1. DCN 69532, REPLACE UNIT 1/2 AND U3 Emergency Diesel Generator Governors, Revision 4

: 2. DCN 70515, REPLACE TRAVELING WATER SCREENS FOR UNIT 1, 2, and 3, Revision 5
: 3. BFN-19-598, R0 & R1, Add a Second PCIV and Test Connections to the Unit 1, 2, and 3 TIP System Purge Lines, BFN-19-703, BFN-19-740, Revision 1

: 4. BFN-19-660, Replace Unit 1/2 Electronic Expansion Valves and Associated Control Module 1U3 on Control Bay Chillers, Revision 0

: 5. BFN-19-695 R1 EHC Software Modifications, Revision 1

: 6. BFN-20-1124, Replacement of Unit 0 Control Bay Chiller A Control Module and Associated Devices, Revision 0

: 7. Compensatory Measures for FE 1342280, Revision 1

: 8. EOIPM 0-TOC R118, EOI Program Manual, Revision 0

: 9. EOIPM 0-TOC R127, EOI Program Manual, Revision 0

: 10. 1-EOI Appendix 7L, 2-EOI Appendix 7L, 3-EOI Appendix 7L, Revision 0
: 11. ECP 71506, Upgrade Existing U2 Foxboro I/A System to Address Obsolescence, Revision 1
: 12. ECP BFN-19-918, U1, U2, AND U3 HPCI Gland Seal Check Valve Addition, Revision 0

: 13. BFN-20-1462, Intermediate Range Monitor (IRM) HI HI Trip Time Delay Relay Installation, Revision 0

: 14. ECP BFN-20-1747, CCW Cross-Tie Abandonment for Units 1, 2, AND 3, Revision 0
: 15. DC 19-382, ALT Boundary Valve Addition, Revision 0

: 16. DCP BFN 19-816, Intermediate Range Monitor (IRM) Mean Square Analog Module Signal Filter Circuit Change, Revision 0

: 17. TM BFN-2-2021-068-001, R1-U2 RPV Access Hole Cover Repair, Revisions 0, 1, and 2 E1 of E31
                                                                    
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

: 18. BFN-1-2020-068-003, Temporary Replacement of Recirculation Loop 1B Flow Measurement (T-Mod), Revision 0

: 19. TVA-COLR-BF1C14, Core Operating Limits Report, Revision 0

: 20. TVA-COLR-BF2C22, Core Operating Limits Report, Revision 0 E2 of E31
                                                              
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

DCN 69532 -- REPLACE UNIT 1/2 AND U3 EMERGENCY DIESEL GENERATOR GOVERNORS, Revision 4 Activity
 
== Description:==
 
DCN 69532 replaces the obsolete governor control equipment for all 8 Emergency Diesel Generators (EDGs). The existing EGA governor, Motor Operated Potentiometer (MOP), and EGB-10 governor actuators are replaced with the new 2301A Electronic Load Sharing and Speed Control (ELSSC), Digital Reference Unit (DRU), and EGB-13 governor actuator. DCN 69532 also installs a new magnetic speed pickup at the engine flywheel as the speed sensing input to the 2301A ELSSC. This DCN removes the existing resistor box which is no longer required for the new 2301A ELSSC. This DCN rewires the existing raise/lower governor control switch to the DRU to retain its local functionality from the engine control cabinet for rated speed operation. This DCN installs new idle start and idle release push buttons on the diesel engine control cabinet. In addition, this DCN installs a new voltage relay in the diesel electrical control cabinet to prevent the early diesel output breaker closure prior to the diesel achieving minimum required bus voltage.
As discussed in Question 1 of the Screening Review, the two aspects of this change which are considered adverse are the impact of the new 2301A ELSSC with respect to EDG breaker closing at the required voltage and the impact of the DCN with respect to affected failure modes.
The new 2301A ELSSC installed in this DCN, will ramp up slower than the existing EGA governor.
The existing interlock scheme could result in the breaker closing before voltage is above 3940 V as required by BFN Technical Specifications. Voltage relays are incorporated into the design to preclude this occurrence. Regarding affected failure modes and malfunctions with respect to the governor, no new failure modes or malfunctions with a different result are introduced.
In addition to above, the two aspects of this change which are potentially adverse are as below:
: 1. The diodes added by DCN 69532 fails due to Short in Start Circuit 1 or Start Circuit 2 and impacts diesel paralleling mode of operation.
: 2. The failure due to open circuit of filtering circuit added by DCN 69532 could result in noise issue at the EDG governor where filtering circuit is open circuited due to no filtering of noise.
Summary of Evaluation:
DCN 69532 does not present an unreviewed safety question as the replacement of the obsolete governor components are essentially a parts replacement that will enhance diesel generator reliability by replacing obsolete parts with the OEM's recommended replacements. The EDGs are not an initiator of any accident. The bounding accident the EDGs would help mitigate is the Events Resulting in a Reactor Vessel Coolant Inventory Decrease initiated by a Loss of Auxiliary Power. Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

DCN 70515 -- REPLACE TRAVELING WATER SCREENS FOR UNIT 1, 2, AND 3, Revision 5 Activity
 
== Description:==
 
The proposed activity also involves the removal of a motor-operated valve (3-FCV-027-0100), the installation and use of a manual-operated valve and renaming the valve to 3-SHV-027-0100. 3-FCV-027-0100 aligns the spare screen wash pump discharge to Unit 3 traveling screens. 3-FCV-027-0100 and the spare screen wash pump exist to allow maintenance to be performed on the Unit 3 screen wash pump. This change is needed because the operator is obsolete, and no spare parts exist to repair it.
The spare screen wash pump discharge isolation valve (3-FCV-027-0100) is designed to be manually or automatically opened. Substituting a manual-operated valve, in lieu of the motor-operated valve, renders the valve incapable of automatic or remote operation. This is considered an adverse effect on a UFSAR design function. Therefore, a 50.59 evaluation is required.
1-AOl-47-3, "Loss of Condenser Vacuum", does not discuss the screen wash system as means of mitigation of a loss of condenser vacuum transient. The only mention of the CCW system is to confirm CCW pumps are running and to start additional CCW pumps if any are not running.
Therefore, the screen wash system is not credited for a loss of condenser vacuum transient.
3-Ol-27A, "Screen Wash System" states that the Travelling Water Screens will be manned when the screen wash system is operating. Therefore if 3-SHV-27-100 needed to be opened, the time delay due to conversion to a manual valve is minimal.
The new valve (3-SHV-027-0100) is not designed to function during or after any design basis accidents. The valve is part of the traveling screen system. Water for the Residual Heat Removal Service Water (RHRSW) System, a safety-related system, passes through the Unit 3 traveling screens. However, the Unit 3 traveling screen system is not designed to function during loss of power conditions. Thus, no design basis accidents are involved with or affected by this activity.
The credible failure modes associated with this activity are the failure of the valve to close, failure of the valve to open, or failure of the valve in a mid-position. With the automatic operators removed, a failure of one of these valves to close would not automatically prohibit alignment of the spare screen wash pump to any Unit, if the spare was needed to support one of those units.
However, discharge flow and pressure from the spare screen wash pump to the other Units would be reduced, which could result in decreased effectiveness of the screen wash system on the affected Unit. Failure of one of these valves to open could possibly prohibit alignment of the spare screen wash pump to the associated Unit, if the spare was needed to support the Unit.
This could result in decreased effectiveness of the screen wash system on the associated Unit.
Failure of the valve in a mid position could result in decreased effectiveness of the screen wash system on the associated Unit, or decreased effectiveness of the screen wash system on the other Units, if the spare screen wash pump was needed to support one of those Units. No new operational failure modes are created by use of a manual valve which would threaten the E4 of E31
                                                                            
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

proper function of the CCW or RHRSW systems. A failure of the valve would lead to a degradation of flow to the screen wash system, not to a failure of the screen wash system.
As the existing valve (3-FCV-027-0100) is used in manual mode most of the time, the use of a manual-operated valve is considered a similar change without the automatic function. The lack of the automatic function on the manual-operated valve will remove the possibility of a failure and unwanted actuations. Therefore, the new manual-operated valve is no less dependable than the existing motor-operated valve.
The proposed activity also involves the replacement of the existing Y type screen wash strainers with Hellan strainers (3-STN-027-0905, -0905 & -0906). This change is needed because the existing Y type strainers have a history of clogging due to debris. The proposed Hellan strainers have baskets with 1/16-inch perforations vs. 1/4-inch perforations in the current baskets, this is potentially an adverse effect on a UFSAR design function. Therefore, a 50.59 evaluation is required.
The new Hellan strainers are designed to allow debris flushing while the screen wash system is in operation. As they are manual strainers this would involve opening a drain valve (3-SHV-588, -591, -593) and turning the strainer basket handles for 30 seconds each, then closing the drain valve. As the TWS system is manned during screen wash operations, there would be minimal delay in performing this operation.
The new strainers are not designed to function during or after any design basis accidents.
The strainer is part of the traveling screen system. Water for the Residual Heat Removal Service Water (RHRSW) System, a safety-related system, passes, trough the Unit 3 traveling screens. However, the Unit 3 traveling screen system is not designed to function during loss of power conditions, nor is the screen wash system. Thus, no design basis accidents are involved with or affected by this activity.
The proposed modification affects the CCW system by replacing the Unit 1, 2, and 3 screen wash system strainers and associated piping. This replacement will not adversely impact the screen wash systems' performance of its design function as shown in DCN 70515, nor will it impact the RHRSW or CCW systems.
The replacement strainers will improve the Unit 1, 2, and 3 screen wash system function because the pressure drop across the strainer is less than the Y strainers being replaced. The replacement strainers have baskets with 1/16 inch perforations vs. 1/4 inch perforations in the current baskets, however, this change is not adverse because the smaller perforation will better protect downstream equipment. Increased cleaning requirements based on the smaller basket perforations are possible, but not expected based on operating experience, and strainer cleaning frequency is not a UFSAR described design function. A strainer failure would lead to screen wash degradation, not screen wash system failure.

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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

Summary of Evaluation:
The proposed modifications do not increase the frequency or likelihood of accidents or malfunctions or create a new type of accident. A hardware-related (credible failure mode) was used from the technical support work in addition to previous evaluations which were performed via Units 1 & 2 valves replacements (1, 2-FCV-027-0100) which are similar. This DCN does not install digital components, therefore NEI 01-01 is not applicable. The new screen wash strainers are as or more reliable than the existing strainers and create no new failure modes.
As a result of this evaluation. it is concluded that this activity does not meet any of the criteria of 10 CFR 50.59(c)(2), and therefore obtaining prior NRC approval is not required to implement this activity.
BFN-19-598, R0 & R1 -- ADD A SECOND PCIV AND TEST CONNECTIONS TO THE UNIT 1, 2,
& 3 TIP SYSTEM PURGE LINES, BFN-19-703, BFN-19-740, Revision 1 Activity
 
== Description:==
 
This design change will improve the method for Appendix J testing by adding test connections and valves inside the TIP Room. This will eliminate the need for drywell entry and the need to break apart threaded fittings while performing Appendix J tests on the subject TIP purge line.
Compression fittings are used for all tubing connection which will eliminate the need for thread sealant. Furthermore, a second PCIV (check valve) Is added to the TIP purge line. This will eliminate the single point vulnerability where a failure of a single TIP Purge check valve could result in containment leakage exceeding the Technical Specification limits.
Summary of Evaluation:
The following evaluation was warranted based on Screening Question 1 being answered YES.
The proposed TIP purge line Primary Containment Isolation Valve (PCIV) configuration does not specifically meet the FSAR PCIV criteria as it applies to 10 CFR 50 Appendix A, General Design Criterion 56. This proposed PCIV configuration (two check valves in series) would need to be documented as an exception in the FSAR Section 5.2 and the new PCIV would need to be added to FSAR Table 5.2-2. There are no changes required to the Technical Specification of the plant.
In addition, the reliability to provide the purge function in support of the Neutron Monitoring System has been reduced by the addition of the second check valve, however, this addition will not increase more than minimally the probability of a malfunction of equipment important to safety.
The proposed activity was evaluated against credible malfunctions and UFSAR described accidents that are relevant to the TIP purge line. Malfunctions in the subject purge line can result in either loss of nitrogen flow path or a loss of primary containment boundary. Applicable accidents and transients were determined to include Control Rod Drop Accident, Pipe Break Inside Containment, and a Design Basis Seismic Event (Earthquake).
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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

There are no new accidents, malfunction types, or consequences created, and the likelihood of such malfunctions or accidents is not increased more than a minimum amount. Additionally, the consequences of existing malfunction types or accidents are determined to not increase. UFSAR described methods of evaluation for design basis and safety analysis of the plant remain unchanged.
Therefore, the proposed activity shall be implemented per plant procedures without obtaining a License Amendment.
BFN-19-660 -- REPLACE UNIT 1/2 ELECTRONIC EXPANSION VALVES AND ASSOCIATED CONTROL MODULE 1U3 ON CONTROL BAY CHILLERS Revision 0 Activity
 
== Description:==
 
The Common Cause failure vulnerability of digital systems due to software errors could be considered as a special cause of single failure vulnerability. Since the same software resides in the control module 1U3 that is being installed on both the A and the B chillers, a single undetected design error in the software could lead to a common cause failure of redundant circuits. This element of the activity will be discussed further in evaluation (Section 2h & 6b).
Summary of Evaluation:
This evaluation has determined that the systems and equipment affected by the proposed modification system will continue to meet their design and licensing bases requirements following the implementation of the proposed modification. Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
BFN-19-695 -- EHC SOFTWARE MODIFICATIONS, Revision 1 Activity
 
== Description:==
 
ECP BFN-19-695 makes no hardware changes. The Turbine Supervisory Instrumentation (TSI) system for BFN Unit 2 is being replaced with an updated system from GE/Bently Nevada by ECP BFN-18-285. This replacement necessitates changes to the Electro-Hydraulic Controls (EHC) system inputs, outputs and computations. The EHC software must be modified to accommodate the hardware changes made by the TSI Upgrade. ECP BFN-18-285 makes all associated hardware changes. ECP BFN-19-695 establishes, validates and tests the EHC software changes necessary to implement the hardware changes made by ECP BFN-18-285. This document addresses only the 10 CFR 50.59 implications of the software changes themselves.

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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

CURRENT DESIGN The TSI signals below are currently brought into the EHC system. Of these signals, all except the thrust bearing wear contacts are provided to control room recorders on panel 2-9-7. All TSI signals are made available to the Integrated Computer System.
Vibration - velocity probes on bearings 1-12 Thrust Bearing Wear - pressure switch and limit switch contacts Eccentricity - proximity probe Shell Expansion - Linear Variable Differential Transformer (LVDT)
Rotor Expansion - LVDT Differential Expansion - LVDT The existing EHC system provides interface and controls to perform Thrust Bearing Wear Detector (TBWD) testing. The EHC system initiates turbine trips based on the thrust bearing wear pressure switch inputs or on a confirmed high vibration signal. Vibration confirmation logic is built into the EHC software.
PROPOSED DESIGN The TSI signals in the current design are removed from the EHC system to be monitored and processed by the upgraded TSI system. EHC will no longer process inputs to provide a turbine trip on high vibration. Dry contact trip signals will be provided from the TSI for thrust bearing wear.
Existing pressure switches will provide input of low bearing oil pressure. These triplicated inputs are combined in a 2-out-of-3 logic scheme to initiate a turbine trip. The high vibration trip is eliminated by the TSI upgrade. The EHC software for Unit 2 is modified to address the changes in inputs, outputs, and signal processing.
Summary or Evaluation:
This evaluation has determined that the EHC system will continue to meet its design and licensing bases requirements following the implementation of the purposed software modifications.
Triplicated TBWD trip inputs and signal selection logic results in a software and hardware control system that is as reliable as the existing components. Existing detection of low bearing oil pressure is unchanged. No new system level failure mode effects are introduced. The proposed modification does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.
The revised software will not initiate any new system malfunctions. The software modifications will not adversely impact any of the systems that have a dynamic interface with the EHC System.
Therefore, the modification does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
Transients impacted by the EHC system do not result in fuel cladding failure and result in no E8 of E31
                                                                          
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

radiological consequences. The EHC systems ability to mitigate any postulated events will not be decreased. The new software will not initiate any new events. The modification will not impair or prevent ECCS from mitigating the consequences of any design basis events. Therefore, this activity does not result in more than a minimal increase the consequence of an accident previously evaluated in the UFSAR.
Failure or malfunction of the revised software will not prevent or affect the ability of safety related systems or systems important to safety to respond to the accidents describe in the UFSAR.
Therefore, implementation of the proposed modification does not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR. The potential malfunctions of the modified software are bounded at a system level in the UFSAR. Therefore, the possibility for an unanalyzed malfunction of an SSC important to safety or an accident of a different type than any previously evaluated in the UFSAR are not created.
As described in the UFSAR accident analysis, no malfunction of the EHC system can cause a transient sufficient to damage the fuel barrier or exceed the nuclear limits as required by the safety design basis. No new failure modes are created by the software modifications. The proposed modification does not adversely impact the technical attributes supporting this conclusion.
Therefore, the possibility for an unanalyzed malfunction of an SSC important to safety or an accident of a different type than any previously evaluated in the UFSAR is not created.
The software modification does not necessitate a revision or replacement of any currently used evaluation methodology for the TDAFW System. The modification does not result in a departure from the method of evaluation described in the UFSAR in establishing the design bases or in the safety analyses.
Guidance for evaluation of digital upgrades is contained in NEI 01-01, Guideline on Licensing Digital Upgrades, March 2002. NRC Information Notice (IN) 2010-10 discusses a digital modification to the LaSalle nonsafety-related control rod drive system. The NRC addressed LaSalle's application of 10 CFR 50.59 not answering all the questions in Appendix A of NEI 01-01 in the associated 50.59 Evaluation. In the Criteria Evaluation below, the questions in Appendix A of NEI 01-01 are provided in italics and answered for the proposed digital upgrade. Following each the NEI 01-01 question the criterion question is answered for the analyzed activity.
This evaluation concludes that implementation of the modification does not require a Technical Specification change, does not require a License Amendment, and therefore may proceed without NRC approval.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

BFN-20-1124 -- REPLACEMENT OF UNIT 0 CONTROL BAY CHILLER A CONTROL MODULE AND ASSOCIATED DEVICES, Revision 0 Activity
 
== Description:==
 
The Common Cause failure vulnerability of digital systems due to software errors could be considered as a special cause of single failure vulnerability. Since the same software resides in the control module 1U1 that is being installed on both the A and the B chillers, a single undetected design error in the software could lead to a common cause failure of redundant circuits. The same scenario exists for the new Clear Language Display module being installed on both the A train and B train control bay chiller as well. This element of the activity will be discussed further in evaluation (Section 2h & 6b).
The failure of new Clear Language Display module 1U6 and new control power transformer (CPT) will result in loss of display of parameters human interface at local panel. The failure of Clear Language Display module 1U6 and new CPT will result in indirect failure of control module 1U1.
This activity is bounded and not creating unanalyzed failure. However, this creates a new mechanism of failure of control module 1U1.
This element of the activity will be discussed further in evaluation (Section 6h).
Summary of Evaluation:
This evaluation has determined that the systems and equipment affected by the proposed modification system will continue to meet their design and licensing bases requirements following the implementation of the proposed modification.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
Compensatory Measures for FE 1342280, Revision 1 Activity
 
== Description:==
 
This review evaluates the following compensatory measures discussed in the FE for CR 1342280 that were determined to be adverse and require further evaluation:
Preventative Compensatory Measures:
: 1. Once per hour verify no water leakage from RSW piping in the A 4KV Electric Board Room (EBR 1A) or B 4W Electric Board Room (EBR 1B).
: 2. Once per hour verify no water leakage from RBCCW piping in the A 4KV Electric Board Room (EBR 1A) or C 4W Electric Board Room (EBR 2A).
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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

These preventative measures ensure timely responses to postulated breaks within these rooms to ensure the safe shutdown of the affected Unit 1 or 2.
Mitigating Compensatory Measures (taken within one hour upon the discovery of a RSW leak):
: 1. Place RSW pumps in Manual
: 2. Isolate RSW Tanks and header by the following
: a. Close RSW Tank valves 0-25-728 and 0-25-729
: b. Close FP header valves 0-26-817 and 0-26-6149
: c. Establish appropriate FPIP for isolation of Hose Stations and Sprinkler Systems These mitigating measures will limit the accumulation of water in EBR 1A or 1B following a detected line break, to permit safe shutdown of Unit 1.
All of these compensatory measures will remain in place until installation of Emergency Drain Lines from these rooms by DCN BFN-18-043 stages 1, 2, and 3.
The screening review has concluded that these compensatory actions can be viewed as an adverse change to how FSAR described functions are performed or controlled as discussed in Question 2. Therefore, a 10 CFR 50.59 Evaluation is performed. Actions involving RBCCW leak mitigation strategies screened out.
Summary of Evaluation:
This evaluation concludes that this change does not have more than a minimal adverse impact on the probability or consequences of an accident or malfunction of equipment important to safety. In addition, this activity does not create a new accident or new malfunction of equipment important to safety. Finally, this activity does not result in a design basis limit for a fission product barrier being exceeded or altered and therefore may be performed without prior NRC approval. Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
EOIPM 0-TOC R118, EOI PROGRAM MANUAL, Revision 0 Activity
 
== Description:==
 
EOIPM 0-TOC R118 implements generic BWROG EPG/SAG Revision 4 into the BFN specific EOIPM sections for PSTG and SATG. Four of the changes from EPG/SAG Revision 3 to EPG/SAG Revision 4 required an Evaluation to determine if prior NRC approval is required for implementation. Those changes are listed following:
Change 3: The Change 3 Tables of Systems include operating details permitting bypass of interlocks. These changes must be looked at in the scenarios they might be used.
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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

Change 5: Changes to Reactor Control and to Contingency 1.
The concerns with this change are (1) does the change in RC/L affect the SBLOCA, and (2) does the permission in RC/P and in C1/P to exceed the TS cooldown limit, if necessary, impact Tech Spec.
Change 6: Changes to Address ATWS.
The ATWS Stability SER was received by the BWROG on June 6, 1996. This SER addressed changes in operator instruction to mitigate the consequences of power instabilities during anticipated transients without scram (ATWS). NRC focused its review on two additional aspects of the EPGs: (1) the optimal water level strategy and (2) the effectiveness of boron remixing. The staff concluded that rapid boron remixing will likely occur with flow rates above 15 percent (not 5-10 percent as previously assumed) of rated flow. The staff also concluded that licensees should consider strengthening the procedure guidance by specific instructions to ensure early boron injection after confirmation of ATWS and initiation of actions to lower water level, and by a high level control strategy (targeting TAF + 5 feet (1.52 m), but not to exceed 2 feet (0.61 m) below the feedwater sparger) for plants that injection boron through a standpipe below the core. The BWROG guidance is acceptable, but deviations from the guidance that are consistent with the staffs recommendations are encouraged. ... Control at any level between the minimum steam cooling water level and 2 feet below the feedwater sparger was found to be acceptable.
Change 9: Changes to ATWS when RPV water level cannot be determined.
Contingency 4 is changed due to issue 1610. It addresses conditions when RPV water level cannot be determined, both shutdown and ATWs. The major changes added in C4 due to issue 1610 are in C4-1 which transfer control to RC-1 if level can be determined. The actions C4-1 calls for are similar to the actions in C5/L and C5/P in that ADS is prevented, and attempts are made to maintain the main condenser as a heat sink. Injection is terminated if power is above 20 percent (the Large Oscillation Power Threshold) until power is below the threshold or injection is below the Minimum Core Steam Flow (MCSF). This addresses the possibility of large irregular power oscillations when RPV water level cannot be determined. After that injection is controlled to maintain the MCSF or above and RPV pressure below the scram set point. By controlling RPV pressure and RPV injection assurance of adequate core cooling can be maintained while minimizing the impact on the containment.
Summary of Evaluation:
Changes 3, 5, 6, and 9 are shown to not require prior NRC approval before implementation.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
E12 of E31
                                                                          
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

EOIPM 0-TOC R127, EOI PROGRAM MANUAL, Revision 0 Activity
 
== Description:==
 
EOIPM 0-TOC R127 implements generic BWROG EPG/SAG Revision 4 into the BFN specific EOI and SAMG flowcharts. Four of the changes from EPG/SAG Revision 3 to EPG/SAG Revision 4 required an Evaluation to determine if prior NRC approval is required for implementation. Those changes are listed following:
Change 3: The change 3 Tables of Systems include operating details permitting bypass of interlocks. These changes must be looked at in the scenarios they might be used.
Change 5: Changes to Reactor Control and to Contingency 1.
The concerns with this change are (1) does the change in RC/L affect the SBLOCA, and (2) does the permission in RC/P and in C1/P to exceed the TS cooldown limit, if necessary, impact Tech Specs.
Change 6: Changes to Address ATWS The ATWS Stability SER was received by the BWROG on June 6, 1996. This SER addressed changes in operator instruction "to mitigate the consequences of power instabilities during anticipated transients without scram (ATWS). NRC focused its review on two additional aspects of the EPGs: (1) the optimal water level strategy and (2) the effectiveness of boron remixing. The staff concluded that rapid boron remixing will likely occur with flow rates above 15 percent (not 5-10 percent as previously assumed) of rated flow. The staff also concluded that licensees should consider strengthening the procedure guidance by specific instructions to ensure early boron injection after confirmation of ATWS and initiation of actions to lower water level, and by a high level control strategy (targeting TAF + 5 feet (1.52 m), but not to exceed 2 feet (0.61 m) below the feedwater sparger) for plants that inject boron through a standpipe below the core. The BWROG guidance is acceptable, but deviations from the guidance that are consistent with the staff's recommendations are encouraged. Control at any level between the minimum steam cooling water level and 2 feet below the feedwater sparger was found to be acceptable."
Change 9: Changes to ATWS when RPV water level cannot be determined.
Contingency 4 is changed to address conditions when RPV water level cannot be determined, both shutdown and ATWS. The major changes added in C-4A due are in C4-1 which transfer control to EOl-1A if level can be determined. The actions C-4A calls for are similar to the actions in EOl-1A in that ADS is prevented, and attempts are made to maintain the main condenser as a heat sink. Injection is terminated if power is above 20 percent (the Large Oscillation Power Threshold) until power is below the threshold or injection is below the Minimum Core Steam Flow (MCSF). This addresses the possibility of large irregular power oscillations when RPV water level cannot be determined. After that injection is controlled to maintain the MCSF or above and RPV pressure below the scram set point. By controlling RPV pressure and RPV injection assurance of adequate core cooling can be maintained while minimizing the impact on the containment.
E13 of E31
                                                                      
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

Change 17: Changes to EOI Appendix SC and allows the bypass of the high RPV water level trip for RCIC.
This would be used when RPV water level needs to be maintained high as in Contingency 1 or in EOl-1 when a higher level is needed to promote natural circulation or when raw water injection is used. These conditions would require level above +51" the high RPV water level trip for RCIC. The conditions which require this are outside both the design basis conditions (that high a water level is not required) and outside the licensing basis SBO (power is restored at 4 hours does not require a high water level necessarily). The ELAP where it is needed in C-1 is outside the both the design basis and licensing basis. Therefore, prior NRC approval is not needed.
Summary of Evaluation:
Changes 3, 5, 6, 9 and 17 are shown to not require prior NRC approval before implementation.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
1-EOI Appendix 7L, 2-EOI Appendix 7L, 3-EOI Appendix 7L, Revision 0 Activity
 
== Description:==
 
This change will add operating instructions to 1, 2, 3-EOI Appendix-7L (Alternate RPV Injection System Lineup EHPM System) that will give operators the ability to inject water into the reactor using a different units EHPM pump. This feature will only be utilized when other normal and emergency reactor vessel inventory control sources for the receiving unit are non-functional or ineffective. When crosstie operation is required, a manual crosstie valve is opened aligning the discharge of the pump to the RPV injection valves (FCV-007-0008) on each unit. Flow from the pump is directed to either unit by throttling CLOSED the injection valve to one unit and throttling OPEN the injection valve for the other unit. These instructions include cross- tie of the Unit 1, 2, and 3 CSTs at the tank bottom connections such that the donor unit CST is not depleted.
Summary of Evaluation:
The screening review for this change determined that a 50.59 evaluation was required due to the impact on the donor unit CST inventory and impacts on operator actions for the donor unit. This evaluation concludes that these impacts are either unchanged or are already recognized in the FSAR. Impacts on operator actions were found to be either outside design basis or a minimal impact to design basis events. Impacts on CST inventory are addressed by instructions for cross-tie of the CSTs when EHPM systems are cross-tied.
Impact on CST inventory is found to be minimal. Impacts that are specific to fire events were found to be outside the scope of 50.59.
E14 of E31
                                                                          
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
ECP 71506 -- UPGRADE EXISTING U2 FOXBORO I/A SYSTEM TO ADDRESS OBSOLECENCE, Revision 1 Activity
 
== Description:==
 
==Background:==
 
This change impacts the BFN Unit 2 System 046 Feedwater Level Control System (FWLCS),
System 006 Reactor Feedwater Heater Control System (RFWHCS), System 068/096 Reactor Recirculation Control System (RRCS), and System 035 Generator Temperature Monitoring System (GTMS).
This modification will remove and replace the existing Foxboro I/A system on Unit 2 with current version Foxboro I/A 9.4 hardware and software. This upgrade will impact the existing Foxboro I/A version 4.3 Feedwater Level Control System (FWLCS), Reactor Feedwater Heater Control System (RFWHCS), Reactor Recirculation Control System (RRCS), and Generator Temperature Monitoring System (GTMS). The existing pneumatic Moisture Separator Control System (MSCS) will be replaced with current version Foxboro I/A 9.4 hardware and software, similar to BFN, Units 1 and 3.
This modification has previously been implemented on BFN, Unit 3 by ECP 71507. See Screening Review Section I of this 10 CFR 50.59 document for a detailed description of this modification.
Combining Multiple Functions into a Single Digital Device:
The proposed modification combines multiple functions into a single digital device in two instances.
The first is the two existing FWLCS Control Processor pairs are combined into a single Control Processor pair. The second instance of combining multiple functions into a single digital device is where the existing discrete pneumatic control loops for the Moisture Separator control system are combined into three Control Processor pairs.
Summary of Evaluation:
This evaluation has determined that the systems and equipment affected by the proposed modification system will continue to meet their design and licensing bases requirements following the implementation of the proposed modification.
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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

Guidance for evaluation of digital upgrades is contained in NEI 01-01, Guideline on Licensing Digital Upgrades, March 2002. In the Criteria Evaluation below, the questions in Appendix A of NEI 01-01 are provided in italics and answered for the proposed digital upgrade.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
ECP BFN-19-918 -- U1, U2, AND U3 HPCI GLAND SEAL CHECK VALVE ADDITION, Revision 0 Activity
 
== Description:==
 
This review evaluates the installation of 2-inch piston type check valves (one each for Units 1, 2, and 3) in the cooling water return lines which route effluent from the HPCI Gland Seal Condenser and HPCI Lube Oil Cooler to the suction side of the HPCI Booster Pump. These changes are made to mitigate the effects of pressure transients due to interactions with the Feedwater System by preventing the propagated pressure wave from reaching components which have lower design pressures.
The screening review has concluded that these modifications can be viewed as having an adverse effect on UFSAR described functions as discussed in Question 1 of the Screening Review.
Therefore, a 10 CFR 50.59 Evaluation is performed.
Summary of Evaluation:
This evaluation concludes that this change does not have more than a minimal adverse impact on the probability or consequences of an accident or malfunction of equipment important to safety. In addition, this activity does not create a new accident or new malfunction of equipment important to safety. Finally, this activity does not result in a design basis limit for a fission product barrier being exceeded or altered and therefore may be performed without prior NRC approval.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
BFN-20-1462 -- INTERMEDIATE RANGE MONITOR (IRM) HI HI TRIP TIME DELAY RELAY INSTALLATION, Revision 0 Activity
 
== Description:==
 
The Intermediate Range Monitor (IRM) monitors neutron flux from the upper portion of the Source Range Monitor range to the lower portion of the power range monitoring subsystems of the E16 of E31
                                                                              
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

Neutron Monitoring System. The IRM signal represents neutron flux level and is used to generate trip and rod block signals.
The proposed modification increases the total time from an IRM signal HI-HI detection to an RPS trip signal to nominally less than one second to filter out noise spikes. These spikes have resulted in IRM signal spikes that have produced IRM trips at BFN and throughout the BWR fleet. This response time is modified by a combination of signal filtering and a time delay relay for each IRM.
Rod blocks are generated from the IRM downscale, inoperable or HI trip circuit. The time delay relay is placed in the HI-HI trip circuit, having no impact on the rod block function.
This change will return the filter time constant for the IRM signals to the original equipment values of 110 ms (ranges 1-6) and 10 ms (ranges 7-10) from 560 ms (installed by ECP BFN-19-816) for all ranges. This filter is on the input of the MEAN SQUARE Analog module within each IRM chassis (Reference 4). The time constant change is being accomplished by adding the C1 capacitor back to the filter circuit replacing the C2 capacitor in the filter circuit with a smaller capacitance value. A time delay relay is added to the HI-HI trip circuitry to generate the remainder of the signal delay.
The introduction of this interposing relay represents an additional circuit failure mechanism. The failure of the relay could result in the failure of the trip signal actuation. This change applies to all 8 IRMs on each of Units 1, 2 and 3.
This change increases the response time of the IRM SCRAM function which is credited in the Rod Withdrawal Error described in the UFSAR. The additional failure mechanism and increased neutron peak flux are considered adverse and screen in for further evaluation under 10 CFR 50.59.
This 50.59 Evaluation will address these aspects. Other aspects of the change are screened out.
Summary of Evaluation:
This evaluation concludes that peak fuel enthalpy in a Rod Withdrawal Error (RWE) event during startup operation will increase as a result of increasing the IRM trip response time. Analysis performed by Framatome (reference 5) shows that peak fuel enthalpy remains well below the UFSAR limit of 170 Cal/gm.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
ECP BFN-20-1747 -- CCW CROSS-TIE ABANDONMENT FOR UNITS 1, 2, AND 3, Revision 0 Activity
 
== Description:==
 
ECP BFN-20-1747 and child ECPs BFN-20-1747-01, -02 and -03 provide for the effective removal of the CCW Cross-tie design feature for Units 1, 2 and 3, respectively. These changes include the blanking off of the piping at selected locations, the removal of power to the MOVs and the removal of hand switches and indication from control room panels 1/2/3-PNLA-009-0020.
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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

The screening review has concluded that these modifications can be viewed as having an adverse effect on UFSAR described functions and Procedures as discussed in Questions 1 and 2 of the Screening Review. Therefore a 10 CFR 50.59 Evaluation is performed.
Summary of Evaluation:
This evaluation concludes that this change does not have more than a minimal adverse impact on the probability or consequences of an accident or malfunction of equipment important to safety. In addition, this activity does not create a new accident or new malfunction of equipment important to safety. Finally, this activity does not result in a design basis limit for a fission product barrier being exceeded or altered and therefore may be performed without prior NRC approval.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
DC 19-382 -- ALT BOUNDARY VALVE ADDITION, Revision 0 Activity
 
== Description:==
 
Design Change (DC) BFN-19-382 adds the following check valves to the Main Steam/Main Steam Drain piping in the Moisture Separator room of each unit:
x CKV-006-0874: 2" check valve for the main steam stop valve above the seat drain line header.
x CKV-001-0875: 4" check valve for high pressure auxiliary steam turbine seal steam.
x CKV-001-0876: 6" check valve for High Pressure Auxiliary Steam Supply to Reactor Feed Pumps (RFP), Steam Jet Air Ejectors (SJAE) & Off Gas.
These valves will function as the new Alternate Leakage Treatment (ALT) pathway boundary, and therefore the ISI/IST boundary. The three new boundary check valves are described below:
CKV-006-0874: Due to the piping layout and space constraints, the new boundary valve will be added downstream of control valves FCV-6-100, FCV-6-101, FCV-6-102, and FCV-6-103. These control valves are currently identified as having a manual action to be closed or confirm closed for establishing the ALT boundary following MSIV closure. By the new MS ALT path boundary being relocated to valve CKV-006-0874, the manual action requirement for the control valves to close or be verified closed following MSIV closure will no longer be required. This part of the MS ALT path boundary will be automatically established by the closure of the spring-loaded check valve.
CKV-001-0875: Four valves will be removed from the boundary due to the new 4" check valve (FCV-1-146, FCV-1-145, FCV-1-154, & PCV-1-147). Valves FCV-1-145, FCV-1-154
        & PCV-1-147 are currently identified as having a manual action to be closed or confirm closed for establishing the ALT boundary following MSIV closure. By the new MS ALT path boundary being relocated to valve CKV-006-0875, the manual action requirement for the E18 of E31
                                                                                
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

above notes valves to close or be verified closed following MSIV closure will no longer be required. This part of the MS ALT path boundary will be automatically established by the closure of the spring-loaded check valve.
CKV-001-0876: Seventeen valves will be removed from the boundary due to the new 6" check valve (FCV-1-127, FCV-6-153, FCV-6-122, FCV-6-155, FCV-1-135, FCV-6-127, FCV-6-157, FCV-6-132, FCV-1-143, PCV-1-151, PCV-1-166, PCV-1-167, PCV-1-153, CKV-1-742, CKV-1-742, CKV-6-822, CKV-6-826). Valves FCV-6-153, FCV-6-122, FCV-6-155, FCV-6-127, FCV-6-157 & FCV-6-132 are currently identified as having a manual action to be closed or confirm closed for establishing the ALT boundary following MSIV closure. By the new MS ALT path boundary being relocated to valve CKV-006-0875, the manual action requirement for the above notes valves to close or be verified closed following MSIV closure will no longer be required. This part of the MS ALT path boundary will be automatically established by the closure of the spring-loaded check valve.
The main functions of these main steam valves (System 001) and drain line valves (System 006) and the associated piping are:
x Provide a boundary for the MS ALT pathway to ensure MSIV leakage following a DBA from the main steam lines is directed through the MS drain lines (primary pathway) to the condenser.
x Provide a continuous drain path from above the seat main steam stop valve leakage to the main condenser (CKV-006-0874).
x Provide high pressure seal steam for the main turbines and the Reactor Feed Pump Turbines (RFPT) (CKV-001-0875).
x Provide high pressure steam to the RFPT, Steam Jet Air Ejectors (SJAE), and the Off-gas system (CKV-001-0876).
The scope of this 50.59 evaluation is limited to the screening in of the change for establishing the main steam alternate leak path boundary from manual actions associated with several of the current boundary valves to an automatic action for the new check valves. All other portions of this change screen out and are covered by the 50.59 screening for DC BFN-19-382.
Summary of Evaluation:
The 50.59 evaluation for DC BFN-19-382, Revision 0 addressed the change in establishing portions of the MS Alternate Leak Path boundary from motor and hydraulically operated valves that that require manual action to close or confirm closed to check valves that will automatically close following closure of the MSIVs.
The change in automatic action from manual action required to establish the MS Alternate Leakage path requires a full 50.59 evaluation. The remaining activities associated with this change screened out and were addressed by the 50.59 Screening Review.
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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

The results of the 50.59 evaluation have shown that the MS Alternate Leak Path valve changes do not increase the frequency of occurrence of an accident previously evaluated and do not result in more than a minimal increase in the likelihood of a malfunction previously evaluated. The changes also do not result in more than a minimal increase in the consequences of a previously evaluated accident or a previously evaluated malfunction. The changes also do not introduce a new accident or event that has not been previously evaluated or introduce a different result due to a malfunction previously evaluated. The changes also do not impact a fission product barrier. The changes also do not change any method of evaluation. Therefore, the changes per DC BFN-19-382 do not require NRC approval via a License Amendment prior to implementation.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
DCP BFN 19-816 -- INTERMEDIATE RANGE MONITOR (IRM) MEAN SQUARE ANALOLOG MODULE SIGNAL FILTERCIRCUIT CHANGE, Revision 0 Activity
 
== Description:==
 
This change will increase the filter time constant for the IRM signals from 110 ms in the low ranges (1-6) and 10 ms in the high ranges (7-10) to up to 1000 ms for all ranges. This change is intended to filter out noise spikes that have resulted in IRM trips at BFN and throughout the BWR fleet. The IRM signal represents increases in neutron flux and is used to generate trip and Rod Block signals.
The increased time constant is being accomplished by abandoning the filter circuit having the longer time constant (110 ms) and replacing the capacitor in the remaining filter circuit with one of a larger capacitance value. This will create a single filter for all IRM ranges that has a larger time constant and thus is more effective at filtering out noise. This change applies to all 8 IRMs on Units 1,2 and 3.
This change increases the response time of the IRM SCRAM function which is credited in the Rod Withdrawal Error described in the FSAR. This 50.59 Evaluation will address this aspect. Other aspects of the change are screened out.
Summary of Evaluation:
This evaluation concludes that peak fuel enthalpy in a Rod Withdrawal Error (RWE) event during startup operation will increase as a result of increasing the IRM filter time constant. Analysis performed by Framatome (Reference 5) shows that peak fuel enthalpy remains well below the FSAR limit of 170 Cal/gm.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
E20 of E31
                                                                            
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

TM BFN-2-2021-068-001 -- R1-U2 RPV ACCESS HOLE COVER REPAIR, Revision 0, 1, and 2 Activity
 
== Description:==
 
TMOD BFN-2-2021-068-001 improves the reliability of the joint integrity of the affected Unit 2 Shroud Access Hole Covers to the shroud support plate. This will be accomplished by:
: 1. Electro Discharge Machining (EDM) a 4 by 5 box in the middle of each shroud access hole cover (AHC).
: 2. Insert the lower support (beam & pin) through the hole and swing up to lock under each AHC below the support plate.
: 3. Install bushing and clamping plate above access hole cover and tighten and torque the nut to clamp each AHC to the underside of the support plate and crimp the nut collar.
This temporary modification is intended to be installed for one operating cycle, at which time a permanent design can be implemented which includes removal of the flawed weld for each access cover and the installation of new, multi-bolted designs for each location.
The screening review has concluded that this temporary modification can be viewed as having an adverse effect on UFSAR described functions as discussed in Question 1 of the Screening Review. Therefore, a 10 CFR 50.59 Evaluation is performed.
Summary of Evaluation:
This evaluation concludes that this change does not have more than a minimal adverse impact on the probability or consequences of an accident or malfunction of equipment important to safety. In addition, this activity does not create a new accident or new malfunction of equipment important to safety. Finally, this activity does not result in a design basis limit for a fission product barrier being exceeded or altered and therefore may be performed without prior NRC approval.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
BFN-1-2020-068-003, Temporary Replacement of Recirculation Loop 1B Flow Measurement (T-Mod), Revision 0 Activity
 
== Description:==
 
Upon restart of Unit 1 from the U1R13 outage, flow indication for B train reactor recirculation (RR) pump discharge flow indicated zero flow after start of the 1B RR pump. Troubleshooting resulted in a determination that an internal mechanical failure was present in the venturi used to measure RR pump discharge flow. This resulted in zero differential pressure (dP) measured by the four venturi dP transmitters. The transmitters provide a dP signal proportional to core flow to the four channels of Average Power Range Monitors (APRMs) for flow bias scram trip and rod block logic, and for Oscillation Power Range Monitor (OPRM) functions.
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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

To compensate for the degraded condition, Compensatory Actions are required as described in Condition Report (CR) 1649288 Prompt Determination of Operability (PDO). The Compensatory Actions include the issuance of a temporary modification (TMOD) to replace and repurpose the dP transmitters to measure system dP as a means of system flow measurement, as well as manual actions to aid in instances where the system dP measurement would not perform as well as the dP measurement using the venturi.
CR 1649288 PDO Compensatory Measures that did not screen in for further evaluation:
: 1. Implement TMOD (BFN-1-2020-068-003) to substitute the 1B Recirculation Pump differential pressure signal for the failed B Recirculation Loop Flow Element signal:
: a. Lowering the APRM Simulated Thermal Power (STP) Scram and Rod Block Setpoints (as implemented by TMOD BFN-1-2020-068-003)
: b. Adjusting OPRM Trip Enabled Setpoints (as implemented by TMOD BFN-1-2020-068-003)
: 2. Automated Backup Stability Protection (ABSP) Prohibited from Use.
CR 1649288 PDO Compensatory Measures that screened in for further evaluation:
: 3. 1 B Reactor Recirculation Pump Trip, Pump Start, and Discharge Valve Closure Events
: a. 1B Reactor Recirculation Pump Trip
: i. Procedures are revised to create actions to valve out 1-FT-68-81A, -81B, -81C,
                -81D and open associated equalizing valves after pump trip.
: b. 1 B Reactor Recirculation Pump Start
: i. Procedures are revised to create actions to valve out 1-FT-68-81A, -81 B, -81 C,
                -81 D and open associated equalizing valves prior to pump start.
: c. 1 B Reactor Recirculation Pump Discharge Valve Closure Events
: i. Procedures are revised to create actions to valve out 1-FT-68-81A, -81B, -81C,
                -81D and open associated equalizing valves after valve closure events Compensatory Actions 3a, 3b, and 3c result in the creation of the manual actions for the APRM function. Compensatory Action 3b results in the creation of a manual action for the OPRM function. Changes that fundamentally alter the existing means of controlling design functions are adverse; therefore, these Compensatory Actions screen in for further evaluation.
Summary of Evaluation:
The eight questions evaluation resulted in all "No" answers. The Design Functions of the APRMs and OPRMs are not accident initiators, and these compensatory measures will maintain the function of the APRMs and OPRMs; thus, the compensatory measures are not accident initiators.
All UFSAR described dose analyses have been reviewed and have been found bounding.
Malfunction of these SSCs are not considered in the UFSAR. No new design interactions have been created. The Compensatory Measures are reflected in plant procedures, were evaluated to either be not time critical or can be accomplished in the time required, have been evaluated to be capable of being recoverable from credible errors, and do not have an effect on plant systems.
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ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

Impacts were evaluated and found to be acceptable to implement the activity per plant procedure without obtaining a License Amendment.
TVA-COLR-BF-BF1C14, Core Operating Limits Report, Revision 0 Activity
 
== Description:==
 
A core operating limits report is generated each cycle as a result of updated safety analyses.
Cycle specific reload licensing safety analyses results are summarized in Reference 15; the report is added to FSAR, Appendix N. Cycle specific operating limits for the core design are identified in the new COLR, per Reference 1.
The new core is similar to the previous cycle utilizing the approved MELLLA+ operating domain.
The core is licensed for operation over the MELLLA+ flow range of 85 percent to 105 percent (ICF) at rated power conditions. Lower flows, at reduced powers, are available along the MELLLA rod line. Core flow above 105 percent rated flow is analytically allowed by following the constant pump speed line during coastdown starting from the old 105% OLTP 100 percent P / 105 percent F state point. In other words, flow above 105 percent rated is not allowed for absolute power levels greater than 3458 MWt, (87.5 percent Rated Power on the 120 percent OLTP basis)
The Unit 1 Cycle 14 will consist of ATRIUM-10XM. The ATRIUM-10XM was generically evaluated per Reference per Reference 6. Previously exposed fuel utilizes some BLEU. BLEU material was evaluated per References 3, 5, & 6.
The core is licensed for RPTOOS, TBVOOS, FHOOS, PLUOOS, and Dual or Single Recirculation Loop operation, with appropriate thermal limits provided in Reference 1. Out-Of-Service options may be combined, as long as appropriate limit sets are implemented. FFTR and ICF may be used in any order for cycle extension. ICF and FHOOS are available at any point in the cycle.
CPR limits are provided for NEOC, EOC, and FFTR/Coastdown exposure windows. NEOC limits may be used until a core average exposure of 30,758.8 MWd/MTU is reached. EOC limits may be used at any point in the cycle up to a maximum core average exposure of 34,078.5 MWd/MTU.
FFTR/Coastdown limits may be used at any point in the cycle up to a maximum core average exposure of 35,767.8 MWd/MTU.
Pressurization transients use a minimum power for operability for the Power Load Unbalance (PLU) feature of the Digital EHC system. The PLU will generate a TCV fast closure for a load rejection at or above 40 percent generator power. Reload analyses assume PLU operation above 50 percent core thermal power to conservatively account for efficiency loss when operating at low power conditions with FHOOS.
The hydraulic stability is now performed using the DSS-CD system based on the MELLLA+ NRC approval.
E23 of E31
                                                                          
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

The GE ULTRA HD and ULTRA MD blade designs will be used for any replacements; CR82M-1 is also acceptable. The blade types are currently in use at all three BFN units. The NA-300 LPRM design will be used for any planned replacements. The LPRM type is currently in use at all three BFN units.
Summary of Evaluation:
The core design uses ATRIUM-10XM. This is the third reload of ATRIUM-10XM for Unit 1.
Previously exposed ATRIUM-10XM fuel utilizes some BLEU material. Fresh ATRIUM-10XM only uses CGU material. Control Blade replacements use the previously approved GE ULTRA HD/MD or Westinghouse CR82M-1 designs. LRPM replacements will be made using the previously approved NA-300 design. All potential changes to the core and core components used in this reload have been evaluated and approved for use.
Cycle specific reload evaluations have been performed as required to establish the cycle specific core operating limits supplied in the COLR. Operation of the core within required limits will ensure the fuel performs without fuel failures expected for normal operation and releases within 10 CFR 20 guidelines for AOTs, and within 10 CFR 50.67 guidelines for DBAs. All reload analyses were determined to have been performed in accordance with approved methodologies.
Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
TVA-COLR-BF2C22, Core Operating Limits Report, Revision 0 Activity
 
== Description:==
 
A core operating limits report is generated each cycle as a result of updated safety analyses. Cycle specific reload licensing safety analyses results are summarized in Reference 15; the report is added to FSAR, Appendix N. Cycle specific operating limits for the core design are identified in the new COLR, per Reference 1.
The new core is similar to the previous cycle with the exception of utilizing the approved MELLLA+
operating domain. The core is licensed for operation over the MELLLA+ flow range of 85 percent to 105 percent (ICF) at rated power conditions. Lower flows, at reduced powers, are available along the MELLLA rod line. Core flow above 105% rated flow is analytically allowed by following the constant pump speed line during coastdown starting from the old 105 percent OLTP, 100 percent P / 105 percent F state point. In other words, flow above 105 percent rated is not allowed for absolute power levels greater than 3458 MWt, (87.5 percent Rated Power on the 120 percent OLTP basis)
E24 of E31
                                                                          
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

The Unit 2 Cycle 22 will consist of ATRIUM-10XM; ATRIUM-11 LTAs. The ATRIUM-10XM was generically evaluated per Reference 6; ATRIUM-11 LTAs per Reference 7. Some previously exposed ATRIUM-10XM fuel utilizes some BLEU. BLEU material was evaluated per References 3, 5, & 6.
The core is licensed for RPTOOS, TBVOOS, FHOOS, PLUOOS, and Dual or Single Recirculation Loop operation, with appropriate thermal limits provided in Reference 1. Out-Of-Service options may be combined, as long as appropriate limit sets are implemented. FFTR and ICF may be used in any order for cycle extension. ICF and FHOOS are available at any point in the cycle.
CPR limits are provided for NEOC, EOC, and FFTR/Coastdown exposure windows. NEOC limits may be used until a core average exposure of 31,055.6 MWd/MTU is reached. EOC limits may be used at any point in the cycle up to a maximum core average exposure of 34,233.1 MWd/MTU.
FFTR/Coastdown limits may be used at any point in the cycle up to a maximum core average exposure of 35,926.7 MWd/MTU.
Pressurization transients use a minimum power for operability for the Power Load Unbalance (PLU) feature of the Digital EHC system. The PLU will generate a TCV fast closure for a load rejection at or above 40 percent generator power. Reload analyses assume PLU operation above 50 percent core thermal power to conservatively account for efficiency loss when operating at low power conditions with FHOOS.
The hydraulic stability is now performed using the DSS-CD system based on the MELLLA+ NRC approval.
The GE ULTRA HD and ULTRA MD blade designs will be used for any replacements; CR82M-1 is also acceptable. The blade types are currently in use at all three BFN units. The NA-300 LPRM design will be used for any planned replacements. The LPRM type is currently in use at all three BFN units.
Summary of Evaluation:
The core design uses ATRIUM-10XM and ATRIUM-11 LTAs. Previously exposed ATRIUM-10XM fuel utilizes some BLEU material. Fresh ATRIUM-10XM only uses CGU material. Control Blade replacements use the previously approved GE ULTRA HD/MD or Westinghouse CR82M-1 designs. LRPM replacements will be made using the previously approved NA-300 design. All potential changes to the core and core components used in this reload have been evaluated and approved for use.
Cycle specific reload evaluations have been performed as required to establish the cycle specific core operating limits supplied in the COLR. Operation of the core within required limits will ensure the fuel performs without fuel failures expected for normal operation and releases within 10 CFR 20 guidelines for AOTs, and within 10 CFR 50.67 guidelines for DBAs. All reload analyses were determined to have been performed in accordance with approved methodologies.
E25 of E31
                                                                        
 
ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 10 CFR 50.59(d)(2) Summary Report

Based upon the results of this evaluation, implement the activity per plant procedures without obtaining a License Amendment.
E26 of E31
                                                                        
 
Acronym                                Description AC        Alternating Current ADSP      Advanced Design Steam Path AOTs      Abnormal Operational Transients APRM      Average Power Range Monitor ARTS      APRM and RBM Technical Specification ASME      American Society of Mechanical Engineers ATWS-RPT  Anticipated Transient Without a Scram Reactor Protection Trip B&PV      ASME B&PV Code BFN      Browns Ferry Nuclear BFNP      Browns Ferry Nuclear Plant BLEU      Blended Low Enriched Uranium BWROG    Bowling Water Reactor Owners Group BWRs      Bowling Water Reactors CARV      Cross-Around Relief Valve CAVEX    Core Average Exposure CCW      Closed Cooling Water CGU      Commercial Grade Uranium CKV      Check Valve COLR      Core Operating Limits Report C1/P      Contingency 1 Pressure Step CPR      Critical Power Ratio CPU      Central Processing Unit CRs      Condition Reports CSTs      Condensate Storage Tanks DBA      Design Basis Accidents DBLFPB    Design Basis Limit for a Fission Product Barrier DCN      Design Change Notice DC        Design Change DG        Diesel Generator DIVOM    Delta CPR over Initial CPR vs Oscillation Magnitude DRV      Drain Valve DTPG      Design Test Plan Group DSS-CD    Detect and Suppress Solution - Confirmation Density ECCS      Emergency Core Cooling System ECN      Engineering Change Notice ECP      Engineering Change Package EG-M      Electronic Governor-Modulator EHC      Electro-Hydraulic Controls ELAP      Extended loss of AC Power EOC      End of Cycle EOC-RPT  End of Cycle Recirculation Pump Trip EOIs      Emergency Operating Instruction EMI      Electro Magnetic Interference EPG      Emergency Procedure Guideline EPG/ASG  Emergency Procedures Guidelines/Severe Accident Guidelines EPROM    Erasable Programmable Read-Only Memory
 
Acronym                                Description EPU      Extended Power Up rate FCV      Flow Control Valve FE        Functional Evaluation FDST      Floor Drain Sample Tank FFTR      Final Feedwater Temperature Reduction FHOOS    Feedwater Heaters Out Of Service FME      Failure Modes Evaluation FPIP      Fire Protection Impairment Permit FSAR      Final Safety Analysis Report FTR      Final Task Report FW        Feedwater FWCF      Feedwater Controller Failure FWCS      Feedwater Control System GAC      Granular Activated Carbon GE        General Electric gpm      Gallons per Minute GSF      Generic Shape Function HCTL      Heat Capacity Temperature Limit HCVS      Hardened Containment Venting System HFE      Human Factors Evaluation HICs      High Integrity Containers HMI      Human Machine Interface HP        High Pressure HPCI      High Pressure Coolant Injection System HPT      High Pressure Turbine HWWV      Hardened Wetwell Vent I&C      Instrumentation and Controls ICA      Interim Corrective Action ICF      Increased Core Flow INOP      Inoperable ISTA      lnservice Testing Program Subsection - General Requirements ISTD      lnservice Testing Subsection - Snubbers ISV      Isolation Valve kV        1000 Volts LAR      License Amendment Request LCO      TS Limiting Condition Of Operation LHGR      Linear Heat Generation Ratio LLSLC    Low Leakage Scatter Loaded Core LOCA      Loss Of Coolant Accident LPCI      Low-Pressure Coolant Injection LPRM      Local Power Range Monitor Generator Load Reject (TCV Fast Closure) with Turbine Bypass Valve LRNBP Failure LRW      Liquid Radwaste System LT        Level Transmitter LUAs      Lead Use Assemblies MAPLHGR  Maximum Average Planar Linear Heat Generation Ratio
 
Acronym                                Description MCPR      Minimum Critical Power Ratio Safety Limit MELLL    Maximum Extended Load Line Limit MELLLA    Maximum Extended Load Line Limit Analysis MCR      Main Control Room MG        Motor Generator MSIV      Main Steam Isolation Valve MOV      Motor Operated Vavle MS        Main Steam MSRV      Main Steam Relief Valve MWd/MTU  Megawatt Day Per Metric Ton Uranium MWt      Megawatt (Thermal)
NEOC      Near End Of Cycle NRC      Nuclear Regulatory Commission NUMAC    Nuclear Measurement, Analysis, and Control OEM      Original Equipment Manufacturer OIP      Overall Integrated Plan OLMCPRs  MCPR operating limits OLTP      Original Licensed Thermal Power OM        Operation and Maintenance OPRM      Oscillation Power Range Monitor OPRMS    Oscillation Power Range Monitor System PCIS      Primary Containment Isolation System PCIVs    Primary Containment Isolation Valve PCT      Peak Cladding Temperature PER      Problem Evaluation Report PLC      Programmable Logic Controller PLU      Power Load Unbalance PLUOOS    Power Load Unbalance Out Of Service PRNM      Power Range Neutron Monitoring PS        Pressure Switch PSP      Pressure Suppression Pressure PSTG      Plant Specific Technical Guideline RBCCW    Reactor Building Closed Cooling Water RBM      Rod Block Monitor RCIC      Reactor Core Isolation Cooling RC/L      RPV Control/Level RC/P      RPV Control/Pressure RCPOOS    Recirculation Pump Out Of Service RFI      Radio-Frequency Interference RFPT      Reactor Feedpump Turbine RFWCS    Reactor Feedwater Control System RHR      Residual Heat Removal RMOV      Reactor Motor-Operated Valve RPV      Reactor Pressure Vessel RO        Reverse Osmosis RPT      Recirculation Pump Trip RPTOOS    Recirculation Pump Trip Out Of Service
 
Acronym                              Description RSW      Raw Service Water RTO      Return to Operations RW        Radwaste RWM      Rod Worth Minimizer SAG      Severe Accident Guidelines SATG      Severe Accident Technical Guideline SBO      Station Black Out SCRAM    Safety Control Rod Axe Man SER      Safety Evaluation Report SHV      Sutoff Valve SLMCPR    Safety Limit and the Minimum Critical Power Ratio Safety Limit SLO      Single Loop Operation SQAP      Software Quality Assurance Plan SR        Surveillance Procedure SRLR      Supplemental Reload Licensing Report SSC      Structure, System or Component TAF      Top of Active Fuel TBVOOS    Turbine Bypass Valves Out Of Service TCV      Turbine Control Valve TDA      Particulate Detection Apparatus TIP      Traversing Incore Probe TRM      Technical Requirements Manual TS        Temperature Switch TTNBP    Turbine Bypass Valve Failure Following Turbine Trip, High Power TV        Test Valve TVA      Tennessee Valley Authority UFSAR    Updated Final Safety Analysis Report V        Volt VA        Volt Ampere VFDs      Variable Frequency Drives XFA      Transformer ZS        Position Switch
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions BFN Units 1, 2, and 3 - TS Bases Revision 116 Unit 1 TS Bases, Revision 116 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.1-53                          Revision 116            07/09/2019 B 3.1-54                          Revision 116            07/09/2019 B 3.1-54a                        Revision 116            07/09/2019 B 3.1-55                          Revision 116            07/09/2019 Unit 2 TS Bases, Revision 116 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.1-53                          Revision 116            07/09/2019 B 3.1-54                          Revision 116            07/09/2019 B 3.1-54a                        Revision 116            07/09/2019 B 3.1-55                          Revision 116            07/09/2019 Unit 3 TS Bases, Revision 116 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.1-53                          Revision 116            07/09/2019 B 3.1-54                          Revision 116            07/09/2019 B 3.1-54a                        Revision 116            07/09/2019 B 3.1-55                          Revision 116            07/09/2019 BFN Units 1, 2, and 3 - TS Bases Revision 117 Unit 1 TS Bases, Revision 117 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.6-90                          Revision 117            08/22/2019 Unit 2 TS Bases, Revision 117 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.6-90                          Revision 117            08/22/2019 Unit 3 TS Bases, Revision 117 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.6-90                          Revision 117            08/22/2019
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions BFN Units 1, 2, and 3 - TS Bases Revision 118 Unit 1 TS Bases, Revision 118 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-253                        Revision 118            12/23/2019 B 3.3-254                        Revision 118            12/23/2019 B 3.3-257                        Revision 118            12/23/2019 B 3.3-257a                        Revision 118            12/23/2019 B 3.3-259                        Revision 118            12/23/2019 B 3.3-260                        Revision 118            12/23/2019 B 3.3-261                        Revision 118            12/23/2019 B 3.3-262                        Revision 118            12/23/2019 B 3.3-263a                        Revision 118            12/23/2019 B 3.3-264                        Revision 118            12/23/2019 B 3.3-265                        Revision 118            12/23/2019 B 3.8-3a                          Revision 118            12/23/2019 B 3.8-4                          Revision 118            12/23/2019 Unit 2 TS Bases, Revision 118 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-256                        Revision 118            12/23/2019 B 3.3-257                        Revision 118            12/23/2019 B 3.3-260a                        Revision 118            12/23/2019 B 3.3-262                        Revision 118            12/23/2019 B 3.3-263                        Revision 118            12/23/2019 B 3.3-264                        Revision 118            12/23/2019 B 3.3-265                        Revision 118            12/23/2019 B 3.3-265a                        Revision 118            12/23/2019 B 3.3-266                        Revision 118            12/23/2019 B 3.3-267                        Revision 118            12/23/2019 B 3.3-268                        Revision 118            12/23/2019 B 3.8-3a                          Revision 118            12/23/2019 B 3.8-4                          Revision 118            12/23/2019
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 3 TS Bases, Revision 118 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-256                        Revision 118            12/23/2019 B 3.3-257                        Revision 118            12/23/2019 B 3.3-260                        Revision 118            12/23/2019 B 3.3-260a                        Revision 118            12/23/2019 B 3.3-262                        Revision 118            12/23/2019 B 3.3-263                        Revision 118            12/23/2019 B 3.3-264                        Revision 118            12/23/2019 B 3.3-265                        Revision 118            12/23/2019 B 3.3-266                        Revision 118            12/23/2019 B 3.3-267                        Revision 118            12/23/2019 B 3.3-268                        Revision 118            12/23/2019 B 3.8-3                          Revision 118            12/23/2019 BFN Units 1, 2, and 3 - TS Bases Revision 119 Unit 1 TS Bases, Revision 119 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-16a                        Revision 119            01/03/2020 B 3.3-16b                        Revision 119            01/03/2020 B 3.3-16c                        Revision 119            01/03/2020 B 3.3-33                          Revision 119            01/03/2020 B 3.3-34                          Revision 119            01/03/2020 B 3.3-34a                        Revision 119            01/03/2020 B 3.3-34b                        Revision 119            01/03/2020 B 3.3-34c                        Revision 119            01/03/2020 B 3.3-34d                        Revision 119            01/03/2020 B 3.3-43a                        Revision 119            01/03/2020 B 3.3-44                          Revision 119            01/03/2020 B 3.4-4                          Revision 119            01/03/2020 B 3.4-5                          Revision 119            01/03/2020 B 3.4-6                          Revision 119            01/03/2020 B 3.4-5                          Revision 119            01/03/2020 B 3.4-6                          Revision 119            01/03/2020 B 3.4-7                          Revision 119            01/03/2020 B 3.4-8                          Revision 119            01/03/2020 B 3.4-10                          Revision 119            01/03/2020
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 2 TS Bases, Revision 119 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-15a                        Revision 119            04/27/2020 B 3.3-15b                        Revision 119            04/27/2020 B 3.3-15c                        Revision 119            04/27/2020 B 3.3-34                          Revision 119            04/27/2020 B 3.3-35                          Revision 119            04/27/2020 B 3.3-35a                        Revision 119            04/27/2020 B 3.3-35b                        Revision 119            04/27/2020 B 3.3-35c                        Revision 119            04/27/2020 B 3.3-45a                        Revision 119            04/27/2020 B 3.3-46                          Revision 119            04/27/2020 B 3.4-4                          Revision 119            04/27/2020 B 3.4-5                          Revision 119            04/27/2020 B 3.4-6                          Revision 119            04/27/2020 B 3.4-7                          Revision 119            04/27/2020 B 3.4-8                          Revision 119            04/27/2020 B 3.4-10                          Revision 119            04/27/2020 Unit 3 TS Bases, Revision 119 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-15a                        Revision 119            03/10/2020 B 3.3-15b                        Revision 119            03/10/2020 B 3.3-15c                        Revision 119            03/10/2020 B 3.3-34                          Revision 119            03/10/2020 B 3.3-35                          Revision 119            03/10/2020 B 3.3-35a                        Revision 119            03/10/2020 B 3.3-35b                        Revision 119            03/10/2020 B 3.3-35c                        Revision 119            03/10/2020 B 3.3-45a                        Revision 119            03/10/2020 B 3.3-46                          Revision 119            03/10/2020 B 3.4-4                          Revision 119            03/10/2020 B 3.4-5                          Revision 119            03/10/2020 B 3.4-6                          Revision 119            03/10/2020 B 3.4-7                          Revision 119            03/10/2020 B 3.4-8                          Revision 119            03/10/2020 B 3.4-10                          Revision 119            03/10/2020
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions BFN Units 1, and 3 - TS Bases Revision 120 Unit 1 TS Bases, Revision 120 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.0-2                          Revision 120            03/13/2020 B 3.0-3                          Revision 120            03/13/2020 B 3.0-4                          Revision 120            03/13/2020 Unit 3 TS Bases, Revision 120 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.0-2                          Revision 120            03/13/2020 B 3.0-3                          Revision 120            03/13/2020 B 3.0-4                          Revision 120            03/13/2020 BFN Units 1, 2, and 3 - TS Bases Revision 121 Unit 1 TS Bases, Revision 121 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-103                        Revision 121            09/02/2020 Unit 2 TS Bases, Revision 121 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-106                        Revision 121            09/02/2020 Unit 3 TS Bases, Revision 121 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-106                        Revision 121            09/02/2020
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions BFN Units 1, 2, and 3 - TS Bases Revision 122 Unit 1 TS Bases, Revision 122 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-229                        Revision 122            01/20/2021 B 3.3-242                        Revision 122            01/20/2021 Unit 2 TS Bases, Revision 121 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-232                        Revision 122            01/20/2021 B 3.3-245                        Revision 122            01/20/2021 Unit 3 TS Bases, Revision 121 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-232                        Revision 122            01/20/2021 B 3.3-245                        Revision 122            01/20/2021 BFN Units 1, 2, and 3 - TS Bases Revision 123 Unit 1 TS Bases, Revision 123 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.1-23                          Revision 123            04/12/2021 B 3.1-31                          Revision 123            04/12/2021 B 3.1-40                          Revision 123            04/12/2021 B 3.1-45                          Revision 123            04/12/2021 B 3.1-51                          Revision 123            04/12/2021 B 3.1-52                          Revision 123            04/12/2021 B 3.1-53                          Revision 123            04/12/2021 B 3.1-54                          Revision 123            04/12/2021 B 3.1-54a                        Revision 123            04/12/2021 B 3.1-55                          Revision 123            04/12/2021 B 3.1-56                          Revision 123            04/12/2021 B 3.1-57                          Revision 123            04/12/2021 B 3.1-62                          Revision 123            04/12/2021 B 3.1-63                          Revision 123            04/12/2021 B 3.2-4                          Revision 123            04/12/2021 B 3.2-10                          Revision 123            04/12/2021 B 3.2-15                          Revision 123            04/12/2021 B 3.3-28                          Revision 123            04/12/2021 B 3.3-35                          Revision 123            04/12/2021
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 1 TS Bases, Revision 123 Changes (continued):
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-36                          Revision 123            04/12/2021 B 3.3-37                          Revision 123            04/12/2021 B 3.3-39                          Revision 123            04/12/2021 B 3.3-41                          Revision 123            04/12/2021 B 3.3-42                          Revision 123            04/12/2021 B 3.3-43                          Revision 123            04/12/2021 B 3.3-51                          Revision 123            04/12/2021 B 3.3-52                          Revision 123            04/12/2021 B 3.3-28                          Revision 123            04/12/2021 B 3.3-65                          Revision 123            04/12/2021 B 3.3-66                          Revision 123            04/12/2021 B 3.3-71                          Revision 123            04/12/2021 B 3.3-71a                        Revision 123            04/12/2021 B 3.3-85                          Revision 123            04/12/2021 B 3.3-86                          Revision 123            04/12/2021 B 3.3-87                          Revision 123            04/12/2021 B 3.3-96                          Revision 123            04/12/2021 B 3.3-97                          Revision 123            04/12/2021 B 3.3-147                        Revision 123            04/12/2021 B 3.3-148                        Revision 123            04/12/2021 B 3.3-190                        Revision 123            04/12/2021 B 3.3-191                        Revision 123            04/12/2021 B 3.3-192                        Revision 123            04/12/2021 B 3.3-204                        Revision 123            04/12/2021 B 3.3-205                        Revision 123            04/12/2021 B 3.3-206                        Revision 123            04/12/2021 B 3.3-234                        Revision 123            04/12/2021 B 3.3-245                        Revision 123            04/12/2021 B 3.3-246                        Revision 123            04/12/2021 B 3.4-9                          Revision 123            04/12/2021 B 3.4-15                          Revision 123            04/12/2021 B 3.4-22                          Revision 123            04/12/2021 B 3.4-28                          Revision 123            04/12/2021 B 3.4-29                          Revision 123            04/12/2021 B 3.4-35                          Revision 123            04/12/2021 B 3.4-36                          Revision 123            04/12/2021 B 3.4-41                          Revision 123            04/12/2021 B 3.4-48                          Revision 123            04/12/2021 B 3.4-54                          Revision 123            04/12/2021
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 1 TS Bases, Revision 123 Changes (continued):
Bases Page No.        Amendment / Revision No.      Effective Date B 3.4-62                          Revision 123            04/12/2021 B 3.4-65                          Revision 123            04/12/2021 B 3.4-69                          Revision 123            04/12/2021 B 3.5-12                          Revision 123            04/12/2021 B 3.5-13                          Revision 123            04/12/2021 B 3.5-14                          Revision 123            04/12/2021 B 3.5-17                          Revision 123            04/12/2021 B 3.5-18                          Revision 123            04/12/2021 B 3.5-19                          Revision 123            04/12/2021 B 3.5-21                          Revision 123            04/12/2021 B 3.5-34                          Revision 123            04/12/2021 B 3.5-35                          Revision 123            04/12/2021 B 3.5-36                          Revision 123            04/12/2021 B 3.5-37                          Revision 123            04/12/2021 B 3.6-6                          Revision 123            04/12/2021 B 3.6-16                          Revision 123            04/12/2021 B 3.6-17                          Revision 123            04/12/2021 B 3.6-30                          Revision 123            04/12/2021 B 3.6-33                          Revision 123            04/12/2021 B 3.6-34                          Revision 123            04/12/2021 B 3.6-35                          Revision 123            04/12/2021 B 3.6-36                          Revision 123            04/12/2021 B 3.6-39                          Revision 123            04/12/2021 B 3.6-47                          Revision 123            04/12/2021 B 3.6-48                          Revision 123            04/12/2021 B 3.6-55                          Revision 123            04/12/2021 B 3.6-56                          Revision 123            04/12/2021 B 3.6-63                          Revision 123            04/12/2021 B 3.6-67                          Revision 123            04/12/2021 B 3.6-72                          Revision 123            04/12/2021 B 3.6-79                          Revision 123            04/12/2021 B 3.6-84                          Revision 123            04/12/2021 B 3.6-85                          Revision 123            04/12/2021 B 3.6-89                          Revision 123            04/12/2021 B 3.6-94                          Revision 123            04/12/2021 B 3.6-95                          Revision 123            04/12/2021 B 3.6-100                        Revision 123            04/12/2021 B 3.6-105                        Revision 123            04/12/2021 B 3.6-106                        Revision 123            04/12/2021
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 1 TS Bases, Revision 123 Changes (continued):
Bases Page No.        Amendment / Revision No.      Effective Date B 3.6-113                        Revision 123            04/12/2021 B 3.6-120                        Revision 123            04/12/2021 B 3.6-121                        Revision 123            04/12/2021 B 3.7-9                          Revision 123            04/12/2021 B 3.7-14                          Revision 123            04/12/2021 B 3.7-15                          Revision 123            04/12/2021 B 3.7-21a                        Revision 123            04/12/2021 B 3.7-22                          Revision 123            04/12/2021 B 3.7-28                          Revision 123            04/12/2021 B 3.7-32                          Revision 123            04/12/2021 B 3.7-33                          Revision 123            04/12/2021 B 3.7-36                          Revision 123            04/12/2021 B 3.8-29                          Revision 123            04/12/2021 B 3.8-30                          Revision 123            04/12/2021 B 3.8-31                          Revision 123            04/12/2021 B 3.8-33                          Revision 123            04/12/2021 B 3.8-34                          Revision 123            04/12/2021 B 3.8-35                          Revision 123            04/12/2021 B 3.8-37                          Revision 123            04/12/2021 B 3.8-38                          Revision 123            04/12/2021 B 3.8-55                          Revision 123            04/12/2021 B 3.8-56b                        Revision 123            04/12/2021 B 3.8-64                          Revision 123            04/12/2021 B 3.8-65                          Revision 123            04/12/2021 B 3.8-66                          Revision 123            04/12/2021 B 3.8-67                          Revision 123            04/12/2021 B 3.8-78                          Revision 123            04/12/2021 B 3.8-100                        Revision 123            04/12/2021 B 3.8-108                        Revision 123            04/12/2021 B 3.9-5                          Revision 123            04/12/2021 B 3.9-9                          Revision 123            04/12/2021 B 3.9-10                          Revision 123            04/12/2021 B 3.9-13                          Revision 123            04/12/2021 B 3.9-21                          Revision 123            04/12/2021 B 3.9-13                          Revision 123            04/12/2021 B 3.9-25                          Revision 123            04/12/2021 B 3.10-12                        Revision 123            04/12/2021 B 3.10-18                        Revision 123            04/12/2021 B 3.10-25                        Revision 123            04/12/2021
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 1 TS Bases, Revision 123 Changes (continued):
Bases Page No.        Amendment / Revision No.      Effective Date B 3.10-31                        Revision 123            04/12/2021 B 3.10-35                        Revision 123            04/12/2021 B 3.10-47                        Revision 123            04/12/2021 B 3.10-48                        Revision 123            04/12/2021 Unit 2 TS Bases, Revision 123 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.1-23                          Revision 123            04/12/2021 B 3.1-31                          Revision 123            04/12/2021 B 3.1-40                          Revision 123            04/12/2021 B 3.1-44                          Revision 123            04/12/2021 B 3.1-50                          Revision 123            04/12/2021 B 3.1-51                          Revision 123            04/12/2021 B 3.1-52                          Revision 123            04/12/2021 B 3.1-55                          Revision 123            04/12/2021 B 3.2-4                          Revision 123            04/12/2021 B 3.2-10                          Revision 123            04/12/2021 B 3.2-15                          Revision 123            04/12/2021 B 3.3-27                          Revision 123            04/12/2021 B 3.3-36                          Revision 123            04/12/2021 B 3.3-37                          Revision 123            04/12/2021 B 3.3-38                          Revision 123            04/12/2021 B 3.3-40                          Revision 123            04/12/2021 B 3.3-47                          Revision 123            04/12/2021 B 3.3-48                          Revision 123            04/12/2021 B 3.3-49                          Revision 123            04/12/2021 B 3.3-50                          Revision 123            04/12/2021 B 3.3-51                          Revision 123            04/12/2021 B 3.3-52                          Revision 123            04/12/2021 B 3.3-63                          Revision 123            04/12/2021 B 3.3-64                          Revision 123            04/12/2021 B 3.3-65                          Revision 123            04/12/2021 B 3.3-66                          Revision 123            04/12/2021 B 3.3-67                          Revision 123            04/12/2021 B 3.3-69                          Revision 123            04/12/2021 B 3.3-75                          Revision 123            04/12/2021 B 3.3-76                          Revision 123            04/12/2021 B 3.3-77                          Revision 123            04/12/2021 B 3.3-90                          Revision 123            04/12/2021
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 2 TS Bases, Revision 123 Changes (continued):
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-91                          Revision 123            04/12/2021 B 3.3-97                          Revision 123            04/12/2021 B 3.3-98                          Revision 123            04/12/2021 B 3.3-112                        Revision 123            04/12/2021 B 3.3-113                        Revision 123            04/12/2021 B 3.3-114                        Revision 123            04/12/2021 B 3.3-118                        Revision 123            04/12/2021 B 3.3-123                        Revision 123            04/12/2021 B 3.3-124                        Revision 123            04/12/2021 B 3.3-125                        Revision 123            04/12/2021 B 3.3-170                        Revision 123            04/12/2021 B 3.3-171                        Revision 123            04/12/2021 B 3.3-172                        Revision 123            04/12/2021 B 3.3-217                        Revision 123            04/12/2021 B 3.3-218                        Revision 123            04/12/2021 B 3.3-232                        Revision 123            04/12/2021 B 3.3-233                        Revision 123            04/12/2021 B 3.3-234                        Revision 123            04/12/2021 B 3.3-261                        Revision 123            04/12/2021 B 3.3-262                        Revision 123            04/12/2021 B 3.3-271                        Revision 123            04/12/2021 B 3.3-272                        Revision 123            04/12/2021 B 3.4-9                          Revision 123            04/12/2021 B 3.4-15                          Revision 123            04/12/2021 B 3.4-22                          Revision 123            04/12/2021 B 3.4-28                          Revision 123            04/12/2021 B 3.4-29                          Revision 123            04/12/2021 B 3.4-30                          Revision 123            04/12/2021 B 3.4-31                          Revision 123            04/12/2021 B 3.4-36                          Revision 123            04/12/2021 B 3.4-43                          Revision 123            04/12/2021 B 3.4-49                          Revision 123            04/12/2021 B 3.4-57                          Revision 123            04/12/2021 B 3.4-60                          Revision 123            04/12/2021 B 3.4-65                          Revision 123            04/12/2021 B 3.5-12                          Revision 123            04/12/2021 B 3.5-13                          Revision 123            04/12/2021 B 3.5-14                          Revision 123            04/12/2021 B 3.5-17                          Revision 123            04/12/2021 B 3.5-18                          Revision 123            04/12/2021
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 2 TS Bases, Revision 123 Changes (continued):
Bases Page No.        Amendment / Revision No.      Effective Date B 3.5-19                          Revision 123            04/12/2021 B 3.5-21                          Revision 123            04/12/2021 B 3.5-34                          Revision 123            04/12/2021 B 3.5-35                          Revision 123            04/12/2021 B 3.5-36                          Revision 123            04/12/2021 B 3.5-37                          Revision 123            04/12/2021 B 3.6-6                          Revision 123            04/12/2021 B 3.6-16                          Revision 123            04/12/2021 B 3.6-30                          Revision 123            04/12/2021 B 3.6-33                          Revision 123            04/12/2021 B 3.6-34                          Revision 123            04/12/2021 B 3.6-35                          Revision 123            04/12/2021 B 3.6-36                          Revision 123            04/12/2021 B 3.6-39                          Revision 123            04/12/2021 B 3.6-47                          Revision 123            04/12/2021 B 3.6-48                          Revision 123            04/12/2021 B 3.6-55                          Revision 123            04/12/2021 B 3.6-56                          Revision 123            04/12/2021 B 3.6-63                          Revision 123            04/12/2021 B 3.6-67                          Revision 123            04/12/2021 B 3.6-72                          Revision 123            04/12/2021 B 3.6-79                          Revision 123            04/12/2021 B 3.6-84                          Revision 123            04/12/2021 B 3.6-85                          Revision 123            04/12/2021 B 3.6-89                          Revision 123            04/12/2021 B 3.6-95                          Revision 123            04/12/2021 B 3.6-96                          Revision 123            04/12/2021 B 3.6-100                        Revision 123            04/12/2021 B 3.6-105                        Revision 123            04/12/2021 B 3.6-106                        Revision 123            04/12/2021 B 3.6-113                        Revision 123            04/12/2021 B 3.6-120                        Revision 123            04/12/2021 B 3.6-121                        Revision 123            04/12/2021 B 3.7-9                          Revision 123            04/12/2021 B 3.7-14                          Revision 123            04/12/2021 B 3.7-15                          Revision 123            04/12/2021 B 3.7-16                          Revision 123            04/12/2021 B 3.7-24                          Revision 123            04/12/2021 B 3.7-25                          Revision 123            04/12/2021 B 3.7-31                          Revision 123            04/12/2021
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 2 TS Bases, Revision 123 Changes (continued):
Bases Page No.        Amendment / Revision No.      Effective Date B 3.7-35                          Revision 123            04/12/2021 B 3.7-36                          Revision 123            04/12/2021 B 3.7-39                          Revision 123            04/12/2021 B 3.8-29                          Revision 123            04/12/2021 B 3.8-30                          Revision 123            04/12/2021 B 3.8-31                          Revision 123            04/12/2021 B 3.8-33                          Revision 123            04/12/2021 B 3.8-34                          Revision 123            04/12/2021 B 3.8-35                          Revision 123            04/12/2021 B 3.8-37                          Revision 123            04/12/2021 B 3.8-38                          Revision 123            04/12/2021 B 3.8-55                          Revision 123            04/12/2021 B 3.8-56b                        Revision 123            04/12/2021 B 3.8-64                          Revision 123            04/12/2021 B 3.8-65                          Revision 123            04/12/2021 B 3.8-66                          Revision 123            04/12/2021 B 3.8-67                          Revision 123            04/12/2021 B 3.8-78                          Revision 123            04/12/2021 B 3.8-100                        Revision 123            04/12/2021 B 3.8-108                        Revision 123            04/12/2021 B 3.9-5                          Revision 123            04/12/2021 B 3.9-9                          Revision 123            04/12/2021 B 3.9-10                          Revision 123            04/12/2021 B 3.9-13                          Revision 123            04/12/2021 B 3.9-21                          Revision 123            04/12/2021 B 3.9-25                          Revision 123            04/12/2021 B 3.9-30                          Revision 123            04/12/2021 B 3.9-35                          Revision 123            04/12/2021 B 3.10-12                        Revision 123            04/12/2021 B 3.10-18                        Revision 123            04/12/2021 B 3.10-25                        Revision 123            04/12/2021 B 3.10-31                        Revision 123            04/12/2021 B 3.10-35                        Revision 123            04/12/2021 B 3.10-47                        Revision 123            04/12/2021 B 3.10-48                        Revision 123            04/12/2021
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 3 TS Bases, Revision 123 Changes:
Bases Page No.        Amendment / Revision No.      Effective Date B 3.1-23                          Revision 123            04/12/2021 B 3.1-31                          Revision 123            04/12/2021 B 3.1-40                          Revision 123            04/12/2021 B 3.1-45                          Revision 123            04/12/2021 B 3.1-51                          Revision 123            04/12/2021 B 3.1-52                          Revision 123            04/12/2021 B 3.1-53                          Revision 123            04/12/2021 B 3.1-54                          Revision 123            04/12/2021 B 3.1-54a                        Revision 123            04/12/2021 B 3.1-55                          Revision 123            04/12/2021 B 3.1-56                          Revision 123            04/12/2021 B 3.1-57                          Revision 123            04/12/2021 B 3.1-62                          Revision 123            04/12/2021 B 3.1-63                          Revision 123            04/12/2021 B 3.2-4                          Revision 123            04/12/2021 B 3.2-10                          Revision 123            04/12/2021 B 3.2-13a                        Revision 123            04/12/2021 B 3.3-27                          Revision 123            04/12/2021 B 3.3-37                          Revision 123            04/12/2021 B 3.3-38                          Revision 123            04/12/2021 B 3.3-39                          Revision 123            04/12/2021 B 3.3-41                          Revision 123            04/12/2021 B 3.3-42                          Revision 123            04/12/2021 B 3.3-43                          Revision 123            04/12/2021 B 3.3-44                          Revision 123            04/12/2021 B 3.3-45                          Revision 123            04/12/2021 B 3.3-54                          Revision 123            04/12/2021 B 3.3-55                          Revision 123            04/12/2021 B 3.3-56                          Revision 123            04/12/2021 B 3.3-57                          Revision 123            04/12/2021 B 3.3-58                          Revision 123            04/12/2021 B 3.3-69                          Revision 123            04/12/2021 B 3.3-82                          Revision 123            04/12/2021 B 3.3-88                          Revision 123            04/12/2021 B 3.3-89                          Revision 123            04/12/2021 B 3.3103                          Revision 123            04/12/2021
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 3 TS Bases, Revision 123 Changes (continued):
Bases Page No.        Amendment / Revision No.      Effective Date B 3.3-104                        Revision 123            04/12/2021 B 3.3-105                        Revision 123            04/12/2021 B 3.3-113                        Revision 123            04/12/2021 B 3.3-114                        Revision 123            04/12/2021 B 3.3-115                        Revision 123            04/12/2021 B 3.3-160                        Revision 123            04/12/2021 B 3.3-161                        Revision 123            04/12/2021 B 3.3-162                        Revision 123            04/12/2021 B 3.3-207                        Revision 123            04/12/2021 B 3.3-208                        Revision 123            04/12/2021 B 3.3-209                        Revision 123            04/12/2021 B 3.3-222                        Revision 123            04/12/2021 B 3.3-223                        Revision 123            04/12/2021 B 3.3-224                        Revision 123            04/12/2021 B 3.3-252                        Revision 123            04/12/2021 B 3.3-262                        Revision 123            04/12/2021 B 3.3-263                        Revision 123            04/12/2021 B 3.4-9                          Revision 123            04/12/2021 B 3.4-15                          Revision 123            04/12/2021 B 3.4-22                          Revision 123            04/12/2021 B 3.4-28                          Revision 123            04/12/2021 B 3.4-29                          Revision 123            04/12/2021 B 3.4-35                          Revision 123            04/12/2021 B 3.4-36                          Revision 123            04/12/2021 B 3.4-41                          Revision 123            04/12/2021 B 3.4-55a                        Revision 123            04/12/2021 B 3.4-56b                        Revision 123            04/12/2021 B 3.5-12                          Revision 123            04/12/2021 B 3.5-13                          Revision 123            04/12/2021 B 3.5-14                          Revision 123            04/12/2021 B 3.5-17                          Revision 123            04/12/2021 B 3.5-18                          Revision 123            04/12/2021 B 3.5-19                          Revision 123            04/12/2021 B 3.5-21                          Revision 123            04/12/2021 B 3.5-34                          Revision 123            04/12/2021 B 3.5-35                          Revision 123            04/12/2021 B 3.5-36                          Revision 123            04/12/2021 B 3.5-37                          Revision 123            04/12/2021
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 3 TS Bases, Revision 123 Changes (continued):
Bases Page No.        Amendment / Revision No.      Effective Date B 3.6-6                          Revision 123            04/12/2021 B 3.6-16                          Revision 123            04/12/2021 B 3.6-30                          Revision 123            04/12/2021 B 3.6-33                          Revision 123            04/12/2021 B 3.6-34                          Revision 123            04/12/2021 B 3.6-35                          Revision 123            04/12/2021 B 3.6-36                          Revision 123            04/12/2021 B 3.6-39                          Revision 123            04/12/2021 B 3.6-47                          Revision 123            04/12/2021 B 3.6-48                          Revision 123            04/12/2021 B 3.6-55                          Revision 123            04/12/2021 B 3.6-56                          Revision 123            04/12/2021 B 3.6-63                          Revision 123            04/12/2021 B 3.6-67                          Revision 123            04/12/2021 B 3.6-72                          Revision 123            04/12/2021 B 3.6-79                          Revision 123            04/12/2021 B 3.6-84                          Revision 123            04/12/2021 B 3.6-85                          Revision 123            04/12/2021 B 3.6-89                          Revision 123            04/12/2021 B 3.6-95                          Revision 123            04/12/2021 B 3.6-96                          Revision 123            04/12/2021 B 3.6-100                        Revision 123            04/12/2021 B 3.6-105                        Revision 123            04/12/2021 B 3.6-106                        Revision 123            04/12/2021 B 3.6-113                        Revision 123            04/12/2021 B 3.6-120                        Revision 123            04/12/2021 B 3.6-121                        Revision 123            04/12/2021 B 3.7-9                          Revision 123            04/12/2021 B 3.7-14                          Revision 123            04/12/2021 B 3.7-15                          Revision 123            04/12/2021 B 3.7-16                          Revision 123            04/12/2021 B 3.7-24                          Revision 123            04/12/2021 B 3.7-25                          Revision 123            04/12/2021 B 3.7-31                          Revision 123            04/12/2021 B 3.7-35                          Revision 123            04/12/2021 B 3.7-36                          Revision 123            04/12/2021 B 3.7-39                          Revision 123            04/12/2021
 
ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Specifications (TS) Bases Changes and Additions Unit 3 TS Bases, Revision 123 Changes (continued):
Bases Page No.        Amendment / Revision No.      Effective Date B 3.8-29                          Revision 123            04/12/2021 B 3.8-30                          Revision 123            04/12/2021 B 3.8-31                          Revision 123            04/12/2021 B 3.8-33                          Revision 123            04/12/2021 B 3.8-34                          Revision 123            04/12/2021 B 3.8-35                          Revision 123            04/12/2021 B 3.8-37                          Revision 123            04/12/2021 B 3.8-38                          Revision 123            04/12/2021 B 3.8-55                          Revision 123            04/12/2021 B 3.8-56b                        Revision 123            04/12/2021 B 3.8-64                          Revision 123            04/12/2021 B 3.8-65                          Revision 123            04/12/2021 B 3.8-66                          Revision 123            04/12/2021 B 3.8-67                          Revision 123            04/12/2021 B 3.8-78                          Revision 123            04/12/2021 B 3.8-101                        Revision 123            04/12/2021 B 3.8-109                        Revision 123            04/12/2021 B 3.9-5                          Revision 123            04/12/2021 B 3.9-9                          Revision 123            04/12/2021 B 3.9-10                          Revision 123            04/12/2021 B 3.9-13                          Revision 123            04/12/2021 B 3.9-21                          Revision 123            04/12/2021 B 3.9-25                          Revision 123            04/12/2021 B 3.9-30                          Revision 123            04/12/2021 B 3.10-12                        Revision 123            04/12/2021 B 3.10-18                        Revision 123            04/12/2021 B 3.10-25                        Revision 123            04/12/2021 B 3.10-31                        Revision 123            04/12/2021 B 3.10-35                        Revision 123            04/12/2021 B 3.10-47                        Revision 123            04/12/2021 B 3.10-48                        Revision 123            04/12/2021
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.4 REQUIREMENTS (continued) SR 3.1.7.4 requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank. The concentration is dependent upon the volume of water and quantity of boron in the storage tank.
The sodium pentaborate solution (SPB) concentration is allowed to be > 9.2 weight percent provided the concentration and temperature of the sodium pentaborate solution are verified to be within the limits of Figure 3.1.7-1 . This ensures that unwanted precipitation of the sodium pentaborate does not occur.
SR 3.1.7.4 must be performed every 31 days or within 24 hours of when boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.
SR 3.1.7.4 must be performed within 8 hours of discovery that the concentration is> 9.2 weight percent and every 12 hours thereafter until the concentration is verified to be~ 9.2 weight percent. This Frequency is appropriate under these conditions taking into consideration the SLC System design capability still exists for vessel injection under these conditions and the low probability of the temperature and concentration limits of Figure 3.1.7-1 not being met.
(continued)
BFN-UNIT 1                    83.1-53                  Revision G, 2-9, ~ . 116 June 23, 2019
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1 .7.5 REQUIREMENTS (continued) This Surveillance requires the amount of Boron- 10 in the SLC solution tank to be determined every 31 days. The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water. Since the chemicals used have known Boron-10 quantities, the Boron-10 quantity in the sodium pentaborate solution formed can be calculated. This parameter is used as input to determine the volume requirements for reactivity control encompassed by SR 3.1.7.1. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances .
SR 3.1.7 .6 SR 3.1 .7 .6 requires verification thatthe SLC system conditions satisfy the following equation:
( C )( Q )(          E    )      > = 1.0 (8.7 WT%)( 50 GPM )(94 ATOM%)
C = sodium pentaborate solution weight percent concentration Q = SLC system pump flow rate in gpm E = Boron-10 atom percent enrichment in the sodium pentaborate solution To meet 10 CFR 50 .62, the SLC System must have a minimum flow capacity and boron content equivalent in control capacity to 86 gpm of 13 weight percent natural sodium pentaborate solution . The purpose of this injection rate is to ensure that during an ATWS condition with the MSIVs closed, sufficient B-10 is injected into the RPV to bring the reactor subcritical (Hot Shutdown) prior to suppression pool temperature exceeding it's heat capacity temperature limit. The atom percentage of natural B-10 is 19.8%. This equivalency requirement is met when the equation given above is satisfied.
(continued)
BFN-UNIT 1                    B 3.1-54                        Revision G, ~. 116 June 23, 2019
 
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SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7 .7 REQUIREMENTS (continued) Demonstrating that each SLC System pump develops a flow rate~ 39 gpm at a discharge pressure::?: 1325 psig ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration and enrichment requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance. The 24 month Frequency is acceptable since inservice testing of the pumps, performed every 92 days, will detect any adverse trends in pump performance.
The pump flow rate of 39 gpm is based on the original licensing bases for SLC for an alternate reactivity insertion system.
SR 3.1 .7.8 and SR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired . Additionally, replacement charges shall be selected such that the age of charge in service shall not exceed five years from the manufacturer's assembly date. The pump and explosive valve tested should be alternated such that both complete flow paths are tested every 48 months at alternating 24 month intervals.
(continued)
BFN-UNIT 1                    B 3.1-55              Revision G, 29, 4-J, W, 116 June 23, 2019
 
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SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.5 REQUIREMENTS (continued) This Surveillance requires the amount of Boron-10 in the SLC solution tank to be determined every 31 days. The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water. Since the chemicals used have known Boron-10 quantities, the Boron-10 quantity in the sodium pentaborate solution formed can be calculated. This parameter is used as input to determine the volume requirements for reactivity control encompassed by SR 3.1.7.1. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.
SR 3.1.7.6 SR 3.1.7.6 requires verification that the SLC system conditions satisfy the following equation:
_ _(_C_)(_Q~)_(_ _        E_....;)_ _ > = 1. 0 (8 .7WT%)(50GPM )(94ATOM%)
C = sodium pentaborate solution weight percent concentration Q = SLC system pump flow rate in gpm E = 8oron-10 atom percent enrichment in the sodium pentaborate solution To meet 10 CFR 50.62, the SLC System must have a minimum flow capacity and boron content equivalent in control capacity to 86 gpm of 13 weight percent natural sodium pentaborate solution . The purpose of this injection rate is to ensure that during an ATWS condition with the MSIVs closed, sufficient 8-10 is injected into the RPV to bring the reactor subcritical (Hot Shutdown) prior to suppression pool temperature exceeding it's heat capacity temperature limit. The atom percentage of natural 8-10 is 19.8%. This equivalency requirement is met when the equation given above is satisfied.
(continued)
BFN-UNIT 2                    B 3.1-54                        Revision Q, 29, 116 June 23, 2019
 
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SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1 .7.7 REQUIREMENTS (continued) Demonstrating that each SLC System pump develops a flow rate :2: 39 gpm at a discharge pressure :2: 1325 psig ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration and enrichment requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance. The 24 month Frequency is acceptable since inservice testing of the pumps, performed every 92 days, will detect any adverse trends in pump performance .
The pump flow rate of 39 gpm is based on the original licensing basis for SLC for an alternate reactivity insertion system.
SR 3.1.7.8andSR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. Additionally, replacement charges shall be selected such that the age of charge in service shall not exceed five years from the manufacturer's assembly date. The pump and explosive valve tested should be alternated such that both complete flow paths are tested every 48 months at alternating 24 month intervals.
(continued)
BFN-UNIT 2                    B 3.1-55                        Revision ~. 116 Amendment No. 255 June 23, 2019
 
SLC System 8 3.1 .7 BASES SURVEILLANCE SR 3.1.7.4 REQUIREMENTS (continued) SR 3.1.7.4 requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank. The concentration is dependent upon the volume of water and quantity of boron in the storage tank.
The sodium pentaborate solution (SPB) concentration is allowed to be > 9.2 weight percent provided the concentration and temperature of the sodium pentaborate solution are verified to be within the limits of Figure 3.1.7-1. This ensures that unwanted precipitation of the sodium pentaborate does not occur.
SR 3.1.7.4 must be performed every 31 days or within 24 hours of when boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.
SR 3.1.7.4 must be performed within 8 hours of discovery that the concentration is> 9.2 weight percent and every 12 hours thereafter until the concentration is verified to be ~ 9.2 weight percent. This Frequency is appropriate under these conditions taking into consideration the SLC System design capability still exists for vessel injection under these conditions and the low probability of the temperature and concentration limits of Figure 3.1.7-1 not being met.
(continued)
BFN-UNIT 3                    B 3.1-53                  Revision G, 2-9, 44-Q, 116 June 23, 2019
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.5 REQUIREMENTS This Surveillance requires the amount of Boron-10 in the SLC solution tank to be determined every 31 days. The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water. Since the chemicals used have known Boron-10 quantities, the Boron-10 quantity in the sodium pentaborate solution formed can be calculated. This parameter is used as input to determine the volume requirements for reactivity control encompassed by SR 3.1.7.1. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.
SR 3.1.7.6 SR 3.1 .7.6 requires verification that the SLC system conditions satisfy the following equation:
_ _(_C_)_(____;;_Q---'-)(_ _E_....;)_ _ > = 1. 0 (8.7 WT%)( SOGPM )(94ATOM % )
C = sodium pentaborate solution weight percent concentration Q = SLC system pump flow rate in gpm E = Boron-10 atom percent enrichment in the sodium pentaborate solution To meet 10 CFR 50.62, the SLC System must have a minimum flow capacity and boron content equivalent in control capacity to 86 gpm of 13 weight percent natural sodium pentaborate solution. The purpose of this injection rate is to ensure that during an ATWS condition with the MSIVs closed, sufficient 8-10 is injected into the RPV to bring the reactor subcritical (Hot Shutdown) prior to suppression pool temperature exceeding it's heat capacity temperature limit. The atom percentage of natural 8-10 is 19.8%. This equivalency requirement is met when the equation given above is satisfied.
(continued)
BFN-UNIT 3                    B 3.1-54                      Revision Q, 28, 116 June 23, 2019
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.6 (continued)
REQUIREMENTS (continued) The equation can be satisfied by adjusting the solution concentration, pump flow rate or Boron-10 enrichment. If the results of the equation are < 1, the SLC System is no longer capable of shutting down the reactor with the margin described in Reference 2. As described in Reference 2, the BFN analysis assumes a flow capacity and boron content equivalent to 50 gpm of 8.7 weight percent and 94 atom percent B-10 enriched sodium pentaborate solution. This exceeds the requirement of 10 CFR 50 .62, and the equation is adjusted to reflect the BFN requirements. The quantity of stored boron includes an additional margin (25%) beyond the amount needed to shut down the reactor to allow for possible imperfect mixing of the chemical solution in the reactor water, leakage, and the volume in other piping connected to the reactor system.
SR 3.1 .7.6 must be performed every 31 days or within 24 hours of when boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.
(continued)
BFN-UNIT 3                    B 3.1-54a                    Revision Q, ~. 116 June 23, 2019
 
SLC System B 3.1 .7 BASES SURVEILLANCE SR 3.1.7.7 REQUIREMENTS (continued) Demonstrating that each SLC System pump develops a flow rate:?. 39 gpm at a discharge pressure~ 1325 psig ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration and enrichment requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance. The 24 month Frequency is acceptable since inservice testing of the pumps, performed every 92 days, will detect any adverse trends in pump performance.
The pump flow rate of 39 gpm is based on the original licensing basis for SLC for an alternate reactivity insertion system.
SR 3.1.7.8 and SR 3.1.7 .9 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. Additionally, replacement charges shall be selected such that the age of charge in service shall not exceed five years from the manufacturer's assembly date. The pump and explosive valve tested should be alternated such that both complete flow paths are tested every 48 months at alternating 24 month intervals.
(continued)
BFN-UNIT 3                    83.1-55                          Revision ~. 116 Amendment No. 215 June 23, 2019
 
CAD System B 3.6.3.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.3.1 Containment Atmosphere Dilution (CAD) System BASES BACKGROUND          The CAD System functions to maintain combustible gas concentrations within the primary containment at or below the flammability limits following a postulated loss of coolant accident (LOCA) by diluting hydrogen and oxygen with nitrogen. To ensure that a combustible gas mixture does not occur, oxygen concentration is kept< 5.0 volume percent (v/o), or hydrogen concentration is kept< 4.0 v/o.
The CAD System is manually initiated and consists of two independent, 100% capacity subsystems, each of which is capable of supplying nitrogen through separate piping systems to the drywell and suppression chamber of each unit. Each subsystem includes a liquid nitrogen supply tank, ambient vaporizer, electric heater (unqualified), and a manifold with branches to each primary containment (for Units 1, 2, and 3).
The nitrogen storage tanks each contain 2 2615 gal, which is adequate for 7 days of CAD subsystem operation (Ref. 4).
The CAD System operates in conjunction with emergency operating procedures that are used to reduce primary containment pressure periodically during CAD System operation. This combination results in a feed and bleed approach to maintaining hydrogen and oxygen concentrations below combustible levels.
(continued)
BFN-UNIT 1                            B 3.6-90                    Revision Q, ..:t-+2, 117 August16,2019
 
CAD System B 3.6.3.1 8 3.6 CONTAINMENT SYSTEMS B 3.6.3.1 Containment Atmosphere Dilution (CAD) System BASES BACKGROUND          The CAD System functions to maintain combustible gas concentrations within the primary containment at or below the flammability limits following a postulated loss of coolant accident (LOCA) by diluting hydrogen and oxygen with nitrogen. To ensure that a combustible gas mixture does not occur, oxygen concentration is kept< 5.0 volume percent (v/o), or hydrogen concentration is kept< 4 .0 v/o.
The CAD System is manually initiated and consists of two independent, 100% capacity subsystems, each of which is capable of supplying nitrogen through separate piping systems to the drywell and suppression chamber of each unit. Each subsystem includes a liquid nitrogen supply tank, ambient vaporizer, electric heater (unqualified), and a manifold with branches to each primary conta inment (for Units 1, 2, and 3).
The nitrogen storage tanks each contain~ 2615 gal, which is adequate for 7 days of CAD subsystem operation (Ref. 4).
The CAD System operates in conjunction with emergency operating procedures that are used to reduce primary containment pressure periodically during CAD System operation . This combination results in a feed and bleed approach to maintaining hydrogen and oxygen concentrations below combustible levels.
(continued)
BFN-UNIT 2                            B 3.6-90                    Revision Q, 44-J, 117 August 16, 2019
 
CAD System B 3.6.3.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.3.1 Containment Atmosphere Dilution (CAD) System BASES BACKGROUND          The CAD System functions to maintain combustible gas concentrations within the primary containment at or below the flammability limits following a postulated loss of coolant accident (LOCA) by diluting hydrogen and oxygen with nitrogen. To ensure that a combustible gas mixture does not occur, oxygen concentration is kept< 5.0 volume percent (v/o), or hydrogen concentration is kept < 4.0 v/o.
The CAD System is manually initiated and consists of two independent, 100% capacity subsystems, each of which is capable of supplying nitrogen through separate piping systems to the drywell and suppression chamber of each unit. Each subsystem includes a liquid nitrogen supply tank, ambient vaporizer, electric heater (unqualified), and a manifold with branches to each primary conta inment (for Units 1, 2, and 3).
The nitrogen storage tanks each contain ;;:: 2615 gal, which is adequate for 7 days of CAD subsystem operation (Ref. 4) .
The CAD System operates in conjunction with emergency operating procedures that are used to reduce primary containment pressure periodically during CAD System operation. This combination results in a feed and bleed approach to maintaining hydrogen and oxygen concentrations below combustible levels.
(continued)
BFN-UNIT 3                            B 3.6-90                    Revision G, 4-+G, 117 August 16, 2019
 
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 shutdown boards. Offsite power is the preferred source of power for the 4.16 kV shutdown boards. If the monitors determine that insufficient power is available, the boards are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources.
Each 4.16 kV shutdown board has its own independent LOP instrumentation and associated trip logic. The voltage for each board is monitored at three levels, which can be considered as three different undervoltage Functions: Loss of Voltage, Unbalanced Voltage, and 4.16 kV Shutdown Board Undervoltage Degraded Voltage. Each Function causes various board transfers and disconnects.
The Degraded Voltage Function is monitored by three undervoltage relay channels for each shutdown board, whose outputs are arranged in a two-out-of-three logic configuration (Ref. 1). The channels compare measured input signals with pre-established setpoints. When the setpoint is exceeded for two-of-three degraded voltage channels, the logic energizes timers which provides a LOP trip signal to the shutdown board logic.
The unbalanced voltage function is monitored by three unbalanced voltage relays (UVRs) for each shutdown board, whose outputs are arranged in a permissive one-out-of-two logic configuration. The UVRs operate on an unbalanced voltage detection signal dependent on the length of time the signal (continued)
BFN-UNIT 1                            B 3.3-253                      Revision Q., 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES BACKGROUND      is detected. If the permissive one-out-of-two logic is met, the (continued)    relays energize auxiliary relays to provide the trip signal to the shutdown board logic. A permissive one-out-of-two trip logic is defined as a trip of the "Alarm" relay and either the "High" or "Low" relay.
The Loss of Voltage Function is monitored by two undervoltage relay pairs for each shutdown board, where outputs are arranged in a two-out-of-two logic configuration (Ref. .1). The channels include four electro-mechanical relays, two of which must deenergize to start the associated diesel generator and another two which must deenergize to initiate load shed of the associated 4.16 kV shutdown board.
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 2, 3, and 4 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.
Accident analyses credit the loading of the DG based on the loss of offsite power concurrent with a loss of coolant accident.
The diesel starting and loading times have been included in the delay time associated with each safety system component requiring DG supplied power following a loss of offsite power.
The LOP instrumentation satisfies Criterion 3 of the NRG Policy Statement (Ref. 5).
(continued)
BFN-UNIT 1                        B 3.3-254                          Revision 0, 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES APPLICABLE      2. 4.16 kV Shutdown Board Undervoltage (Degraded Voltage)
SAFETY ANALYSES, LCO, and        A reduced voltage condition on a 4.16 kV shutdown board APPLICABILITY    indicates that, while offsite power may not be completely lost (continued)    to the respective shutdown board, available power maybe insufficient for starting large ECCS motors without risking damage to the motors that could disable the ECCS function.
Therefore, power supply to the board is transferred from offsite power to onsite DG power when the voltage on the board 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 Board 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 Shutdown Board Undervoltage (Degraded Voltage) Function per associated board are 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. Refer to LCO 3.8.1 and LCO 3.8.2 for Applicability Bases for the DGs.
: 3. 4.16 kV Shutdown Board Voltage Unbalanced (Unbalanced Voltage Relay)
An unbalanced voltage condition on a 4.16kV shutdown board indicates that, while offsite power may not be completely degraded to the board undervoltage level, available power may be insufficient for starting and running ECCS motors without risking damage to the motors that could disable the ECCS function. Therefore, power supply to the board is transferred from offsite power to onsite DG power when the unbalanced voltage level increases above the Unbalanced Voltage Function (continued)
BFN-UNIT 1                        B 3.3-257                        Revision 0, 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 APPLICABLE      3. 4.16 kV Shutdown BoardVoltage Unbalanced (Unbalanced SAFETY ANALYSES, Voltage Relay) (continued)
LCO, and APPLICABILITY    Allowable Values (unbalanced voltage level with an associated (continued)    time delay). This ensures adequate power will be available to the required equipment. The Board Unbalanced Voltage Allowable Values are high enough to prevent inadvertent power supply transfer, but low 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 voltage balance, but short enough to ensure power is available to the required equipment.
Three UVRs are provided on each 4.16 kV Shutdown Board for detecting an unbalanced voltage condition. The relays are combined in a permissive one-out-of-two logic configuration to generate a supply breaker trip. Three UVRs are required to be OPERABLE when the associated DG is required to be OPERABLE to ensure that no signal instrument failure can preclude a DG function. Refer to LCO 3.8.1 and LCO 3.8.2 for Applicability Bases for the DGs.
(continued)
BFN-UNIT 1                        B 3.3-257a                          Revision 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS    A.1 and A.2 (continued)
Condition C or D, as applicable, must be entered immediately.
The 15 day allowable out of service time is justified based on the two-out-of-three permissive logic scheme provided for these relays. If the inoperable relay channel cannot be restored to OPERABLE status within the allowable out of service time, the degraded voltage relay channel must be placed in the tripped condition per Required Action A.2. 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 F must be entered and its Required Action taken.
B.1 With two or more degraded voltage relay channels or one or more associated timers inoperable on one or more shutdown boards, the Function is not capable of performing the intended function. Required Action B.1 provides a 10 day allowable out of service time provided the loss of voltage relay channels on the affected shutdown board(s) are OPERABLE.
The 10 day allowable out of service time is justified since the loss of voltage relay channels on the same shutdown board are independent of the degraded voltage relay channel(s) and will continue to function and start the diesel generators on a complete loss of voltage. If the inoperable channel(s) cannot (continued)
BFN-UNIT 1                    B 3.3-259                          Revision Q, 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS    B.1 (continued) be restored to OPERABLE status within the allowable out of service time, the channel(s) must be placed in the tripped condition per Required Action B.1. Placing the inoperable channel(s) 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(s) in trip (e.g., as in the case where placing the channel(s) in trip would result in a DG initiation), Condition F must be entered and its Required Action taken.
C.1 With one or more loss of voltage relay channels inoperable on one or more shutdown boards, the Function is not capable of performing the intended function. Required Action C.1 provides a 10 day allowable out of service time provided two or more degraded voltage relay channels and associated timers on the affected shutdown board(s) are OPERABLE. The 10 day allowable out of service time is justified since the degraded voltage relay channels on the same shutdown board are independent of the loss of voltage relay channels and will continue to function and start the diesel generators on a complete loss of voltage. If the inoperable channels cannot be restored to OPERABLE status within the allowable out of service time, the channel(s) must be placed in the tripped condition per Required Action C.1. Placing the inoperable channel(s) 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(s) in trip (e.g., as in the case where placing the channel(s) in trip would result in a DG initiation), Condition F must be entered and its Required Action taken.
(continued)
BFN-UNIT 1                    B 3.3-260                          Revision 0, 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS      D.1 and D.2 (continued)
With two or more degraded voltage relay channels or one or more associated timers and the loss of voltage relay channel(s) inoperable on the same shutdown board, the associated diesel generator will not automatically start upon degraded voltage or complete loss of voltage on that shutdown board. In this situation, Required Action D.2 provides a 5 day allowable out of service time provided the other shutdown boards and undervoltage relay channels are OPERABLE. Immediate verification of the OPERABILITY of the other shutdown boards and undervoltage relay channels is therefore required (Required Action D.1 ). This may be performed as an administrative check by examining logs or other information to determine if this equipment is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate OPERABILITY of this equipment. If the OPERABILITY of this equipment cannot be verified, however, Condition F must be entered immediately. The 5 day allowable out of service time is justified based on the remaining redundancy of the 4.16 kV Shutdown Boards. The 4.16 kV Shutdown Boards have a similar allowable out of service time.
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 D.2.
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 F must be entered and its Required Action taken.
(continued)
BFN-UNIT 1                    B 3.3-261                          Revision 0, 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS      E.1 and E.2 (continued)
The Unbalanced Voltage function generates an LOP signal if the permissive alarm relay and either the Low or High relay actuates to the predetermined unbalanced voltage setting. With one or more UVRs inoperable, the associated diesel generator will not automatically start upon an Unbalanced Voltage signal.
In this situation, Required Action E.2 provides a 5 day allowable out of service time provided the other shutdown boards and Unbalanced Voltage relays are OPERABLE. Immediate verification of the OPERABILITY of the other shutdown boards and UVRs is required (Required Action E.1 ). This action may be performed as an administrative check by examining logs or other information to determine if this equipment is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate OPERABILITY of this equipment. If the OPERABILITY of this equipment cannot be verified, however, Condition F must be entered immediately.
The 5 day allowable out of service time is justified based on the remaining redundancy of the 4.16 kV shutdown boards. The 4.16 kV shutdown boards have a similar allowable out of service time. If the inoperable relay cannot be restored to OPERABLE status within the allowable out of service time, the relay must be placed in the tripped condition. Placing the inoperable relay 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 relay in trip (e.g., as in the case where placing the relay in trip would result in a DG initiation), Condition F must be entered and its Required Action taken.
(continued)
BFN-UNIT 1                      B 3.3-262                          Revision Q., 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES REFERENCES 1. FSAR, Figure 8.4-4.
: 2. FSAR, Section 6.5.
: 3. FSAR, Section 8.5.4.
: 4. FSAR, Chapter 14.
: 5. NRG No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 6. EDQ0000002016000556, "Determination of Unbalance Voltage Relay Analytical Limits."
: 7. EDQ0009992016000564, "Evaluation of 60Q Voltage Unbalanced Relays for Class 1E 4kV Shutdown Boards A, B, C, D, 3EA, 3EB, 3EC, and 3ED."
BFN-UNIT 1                B 3.3-263a                        Revision 118 Amendment No. 235 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 Table B 3.3.8.1-1 (Page 1 of 2)
Loss of Power Instrumentation Channel Device Identification BOARD AND FUNCTIONS                                              CHANNEL DEVICES (UNIDs) 4.16 kV Shutdown Board A
{Loss of Voltage) 1.a    Board Undervoltage - Board Load Shedding                            27SA 0A and 27SA 0C (27-211-000A/12E & /12F) 1.b    Board Undervoltage - Diesel Start Time Delay                        27DA 0A and 27DA 0C (27-211-000A/12A & /12B)
(Degraded Voltage) 2.a    Board Undervoltage                                                  27-211-1A, 27-211-1B, and 27-211-1C (27-211-000A/23A, /23B, & /23C) 2.b.1  Initial Diesel Start and Load Shedding Time Delay                  2-211-1A (02-211-0001A) 2.b.2  Diesel Start Time Delay                                            2-211-2A (02-211-0002A) 2.b.3  Board Load Shedding Time Delay                                      2-211-3A (02-211-0003A) 2.b.4  Diesel Generator Breaker Closure Time Delay                        2-211-4A (02-211-0004A)
(Unbalanced Voltage) 3.a    Board Voltage Unbalanced - Board Load Shedding                      RLY-211-60A1 3.b    Board Voltage Unbalanced - Diesel Start Time Delay                  RLY-211-60A2 RLY-211-60A3 RLY-211-A60A1 RLY-211-A60A2 4.16 kV Shutdown Board B (Loss of Voltage) 1.a    Board Undervoltage - Board Load Shedding                            27SB 0A and 27SB 0C (27-211-000B/12E & /12F) 1.b    Board Undervoltage - Diesel Start Time Delay                        27DB 0A and 27DB 0C (27-211-000B/12A & /12B)
(Degraded Voltage) 2.a    Board Undervoltage                                                27-211-2A, 27-211-2B, and 27-211-2C (27-211-000B/21A, /21 B, /21C) 2.b.1  Initial Diesel Start and Load Shedding Time Delay                  2-211-1 B (02-211-0001 B) 2.b.2  Diesel Start Time Delay                                            2-211-2B (02-211-0002B) 2.b.3  Board Load Shedding Time Delay                                    2-211-3B (02-211-0003B) 2.b.4  Diesel Generator Breaker Closure Time Delay                        2-211-4B (02-211-0004B)
(Unbalanced Voltage) 3.a    Board Voltage Unbalanced - Board Load Shedding                    RLY-211-60B1 3.b    Board Voltage Unbalanced - Diesel Start Time Delay                RLY-211-60B2 RLY-211-60B3 RLY-211-A60B1 RLY-211-A60B2 BFN-UNIT 1                                              B 3.3-264                                    Revision 0, 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 Table B 3.3.8.1-1 (Page 2 of 2)
Loss of Power Instrumentation Channel Device Identification BOARD AND FUNCTIONS                                              CHANNEL DEVICES (UNIDs) 4.16 kV Shutdown Board C (Loss of Voltage)                                                          27SC 0A and 27SC 0C 1.a    Board Undervoltage - Board Load Shedding                            (27-211-000C/11 E & /11 F) 27DC 0A and 27DC 0C 1.b    Board Undervoltage - Diesel Start Time Delay                        (27-211-000C/11A & /11 B)
(Degraded Voltage)                                                        27-211-3A, 27-211-3B, and 27-211-3C 2.a    Board Undervoltage                                                  (27-211-000C/25A, /25B, /25C) 2-211-1C (02-211-0001C) 2.b.1  Initial Diesel Start and Load Shedding Time Delay                  2-211-2C (02-211-0002C) 2.b.2  Diesel Start Time Delay                                            2-211-3C (02-211-0003C) 2.b.3  Board Load Shedding Time Delay                                      2-211-4C (02-211-0004C) 2.b.4  Diesel Generator Breaker Closure Time Delay (Unbalanced Voltage) 3.a    Board Voltage Unbalanced - Board Load Shedding                      RLY-211-60C1 3.b    Board Voltage Unbalanced - Diesel Start Time Delay                  RLY-211-60C2 RLY-211-60C3 RLY-211-A60C1 4.16 kV Shutdown Board D                                                        RLY-211-A60C2 (Loss of Voltage) 1.a    Board Undervoltage - Board Load Shedding                            27SD 0A and 27SD 0C (27-211-000D/11 E & /11 F) 1.b    Board Undervoltage - Diesel Start Time Delay                        27DD 0A and 27DD 0C (27-211-000D/11A & /11 B)
(Degraded Voltage) 2.a    Board Undervoltage                                                27-211-4A, 27-211-4B, and 27-211-4C (27-211-000D/21A, /21 B, /21 C) 2.b.1  Initial Diesel Start and Load Shedding Time Delay                  2-211-10 (02-211-00010) 2.b.2  Diesel Start Time Delay                                            2-211-2D (02-211-0002D) 2.b.3  Board Load Shedding Time Delay                                    2-211-3D (02-211-0003D) 2.b.4  Diesel Generator Breaker Closure Time Delay                        2-211-4D (02-211-0004D)
(Unbalanced Voltage) 3.a    Board Voltage Unbalanced - Board Load Shedding                    RLY-211-60D1 3.b    Board Voltage Unbalanced - Diesel Start Time Delay                RLY-211-60D2 RLY-211-60D3 RLY-211-A60D1 RLY-211-A60D2 BFN-UNIT 1                                              B 3.3-265                                    Revision Q., 118 December 12, 2019
 
AC Sources - Operating B 3.8.1 BASES BACKGROUND  sufficient capacity to support the automatic transfer of all Unit 1 (continued) non-safety related loads when there are existing loads aligned to the CSSTs from Units 2 or 3.
This is addressed by manually disabling the automatic transfer of selected 4.16 kV Unit Boards and/or4.16 kV Common Boards. With the most restrictive manual actions in place, upon a loss of the normal 500 kV offsite circuit coincident with a LOCA, the diesel generators would supply the associated safety-related ESF loads in both divisions needed to mitigate the immediate consequences of a LOCA.
The 161 kV supplied CSSTs can still be credited as part of a qualified alternate offsite circuit for Unit 1. However, access to the 161 kV circuit will require a delayed manual transfer when operators can manually control the loads on the 4.16 kV Start Buses to support long term post accident recovery and shutdown. Operators can restore the de-energized 4.16 kV Unit Boards by manually transferring them to the CSST supplied 4.16 kV Start Buses as desired. The 4.16 kV Shutdown Boards could then be manually transferred from the diesel generators to the CSST supplied 4.16 kV Unit Boards as desired.
The onsite standby power source for 4.16 kV shutdown boards A, B, C, and D consists of four Unit 1 and 2 DGs, each dedicated to a shutdown board. Each DG starts automatically on a LOCA signal (i.e., low reactor water level signal or high drywell pressure signal), or on its respective 4.16 kV shutdown board degraded voltage, unbalanced voltage, or undervoltage signal. In addition to starting all diesel generators, the CAS logic trips the alternate feeder breakers to 4.16 kV Shutdown Boards A, B, C, D. After the DG has started , it automatically ties to its respective bus after offsite power is tripped as a consequence of 4.16 kV shutdown board undervoltage, unbalanced voltage, or degraded voltage, independent of or coincident with a LOCA signal. The DGs also start and operate in the standby mode without tying to the 4.16 kV shutdown board on a LOCA signal alone. Following (continued)
BFN-UNIT 1                      B 3.8-3a                          Revision ~. 118 December 12, 2019
 
AC Sources - Operating B 3.8.1 BASES BACKGROUND  the trip of offsite power, an undervoltage, unbalanced voltage, (continued) or degraded voltage activated load shed logic strips all loads from the 4.16 kV Shutdown Board. Feeder breakers to transformers supplying auxiliary power system distribution boards are not load shed on undervoltage. When the DG is tied to the 4.16 kV shutdown board, large loads are then sequentially connected to its respective 4.16 kV shutdown board by individual pump timers. The individual pump timers control the permissive and starting signals to motor breakers to prevent overloading the DG.
In the event of a loss of offsite power, the ESF electrical loads are automatically connected to the DGs in sufficient time to provide for safe reactor shutdown and to mitigate the consequences of a Design Basis Accident (OBA) such as a LOCA.
Certain required plant loads are returned to service in a predetermined sequence in order to prevent overloading of the DGs in the process. Within 40 seconds after the initiating signal (DG breaker closure with accident signal) is received, all automatic and permanently connected loads needed to recover the unit or maintain it in a safe condition are returned to service.
In the event that the DGs were already running and loaded on the receipt of a spurious or real common accident signal (GAS A/GAS B) from Unit 3, any diesel generator output breakers which are closed are signaled to open to load shed the running loads off of the DG. After the DG breaker closing springs recharge, the DG breakers will reclose and tie the DG to the 4.16 kV shutdown board. Loads are then sequentially connected to its respective 4.16 kV shutdown board by individual pump timers as described above. Any subsequent common accident signal DG breaker trip signals are blocked.
Should a second RHR initiation signal be received (i.e., from a spurious or real accident signal from Unit 1), the Unit 1/2 diesel generator output breakers will be reopened on a unit priority (continued)
BFN-UNIT 1                      B 3.8-4          Revision 0, 40, JO, ~. 47, 118 December 12, 2019
 
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 shutdown boards. Offsite power is the preferred source of power for the 4.16 kV shutdown boards. If the monitors determine that insufficient power is available, the boards are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources.
Each 4.16 kV shutdown board has its own independent LOP instrumentation and associated trip logic. The voltage for each board is monitored at three levels, which can be considered as three different undervoltage Functions: Loss of Voltage, Unbalanced Voltage, and 4.16 kV Shutdown Board Undervoltage Degraded Voltage. Each Function causes various board transfers and disconnects.
The Degraded Voltage Function is monitored by three undervoltage relay channels for each shutdown board, whose outputs are arranged in a two-out-of-three logic configuration (Ref. 1). The channels compare measured input signals with pre-established setpoints. When the setpoint is exceeded for two-of-three degraded voltage channels, the logic energizes timers which provides a LOP trip signal to the shutdown board logic.
(continued)
BFN-UNIT 2                            B 3.3-256                      Revision Q., 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES BACKGROUND      The unbalanced voltage function is monitored by three (continued)    unbalanced voltage relays (UVRs) for each shutdown board, whose outputs are arranged in a permissive one-out-of-two logic configuration. The UVRs operate on an unbalanced voltage detection signal dependent on the length of time the signal is detected. If the permissive one-out-of-two logic is met, the relays energize aux relays to provide the trip signal to the shutdown board logic. A permissive one-out-of-two trip logic is defined as a trip of the "Alarm" relay and either the "High" or "Low" relay.
The Loss of Voltage Function is monitored by two undervoltage relay pairs for each shutdown board, where outputs are arranged in a two-out-of-two logic configuration (Ref. 1). The channels include four electro-mechanical relays, two of which must deenergize to start the associated diesel generator and another two which must deenergize to initiate load shed of the associated 4.16 kV shutdown board.
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 2, 3, and 4 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.
Accident analyses credit the loading of the DG based on the loss of offsite power concurrent with a loss of coolant accident.
The diesel starting and loading times have been included in the delay time associated with each safety system component requiring DG supplied power following a loss of offsite power.
The LOP instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref. 5).
(continued)
BFN-UNIT 2                        B 3.3-257                          Revision 0, 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES APPLICABLE      3. 4.16 kV Shutdown Board Voltage Unbalanced (Unbalanced SAFETY ANALYSES, Voltage Relay)
LCO, and APPLICABILITY    An unbalanced voltage condition on a 4.16kV shutdown board (continued)    indicates that, while offsite power may not be completely degraded to the board undervoltage level, available power may be insufficient for starting and running ECCS motors without risking damage to the motors that could disable the ECCS function. Therefore, power supply to the board is transferred from offsite power to onsite DG power when the unbalanced voltage level increases above the Unbalanced Voltage Function Allowable Values (unbalanced voltage level with an associated time delay). This ensures adequate power will be available to the required equipment. The Board Unbalanced Voltage Allowable Values are high enough to prevent inadvertent power supply transfer, but low 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 voltage balance, but short enough to ensure power is available to the required equipment.
Three UVRs are provided on each 4.16 kV Shutdown Board for detecting an unbalanced voltage condition. The relays are combined in a permissive one-out-of-two logic configuration to generate a supply breaker trip. Three UVRs are required to be OPERABLE when the associated DG is required to be OPERABLE to ensure that no single instrument failure can preclude a DG function. Refer to LCO 3.8.1 and LCO 3.8.2 for Applicability Bases for the DGs.
(continued)
BFN-UNIT 2                        B 3.3-260a                          Revision 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS    A.1 and A.2 (continued)
Condition C or D, as applicable, must be entered immediately.
The 15 day allowable out of service time is justified based on the two-out-of-three permissive logic scheme provided for these relays. If the inoperable relay channel cannot be restored to OPERABLE status within the allowable out of service time, the degraded voltage relay channel must be placed in the tripped condition per Required Action A.2. 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 F must be entered and its Required Action taken.
B.1 With two or more degraded voltage relay channels or one or more associated timers inoperable on one or more shutdown boards, the Function is not capable of performing the intended function. Required Action B.1 provides a 10 day allowable out of service time provided the loss of voltage relay channels on the affected shutdown board(s) are OPERABLE.
The 10 day allowable out of service time is justified since the loss of voltage relay channels on the same shutdown board are independent of the degraded voltage relay channel(s) and will continue to function and start the diesel generators on a complete loss of voltage. If the inoperable channel(s) cannot (continued)
BFN-UNIT 2                    B 3.3-262                          Revision Q., 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS    B.1 (continued) be restored to OPERABLE status within the allowable out of service time, the channel(s) must be placed in the tripped condition per Required Action B.1. Placing the inoperable channel(s) 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(s) in trip (e.g., as in the case where placing the channel(s) in trip would result in a DG initiation), Condition F must be entered and its Required Action taken.
C.1 With one or more loss of voltage relay channels inoperable on one or more shutdown boards, the Function is not capable of performing the intended function. Required Action C.1 provides a 10 day allowable out of service time provided two or more degraded voltage relay channels and associated timers on the affected shutdown board(s) are OPERABLE. The 10 day allowable out of service time is justified since the degraded voltage relay channels on the same shutdown board are independent of the loss of voltage relay channels and will continue to function and start the diesel generators on a complete loss of voltage. If the inoperable channels cannot be restored to OPERABLE status within the allowable out of service time, the channel(s) must be placed in the tripped condition per Required Action C.1 . Placing the inoperable channel(s) 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(s) in trip (e.g., as in the case where placing the channel(s) in trip would result in a DG initiation), Condition F must be entered and its Required Action taken.
(continued)
BFN-UNIT 2                    B 3.3-263                          Revision 0, 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS      D.1 and D.2 (continued)
With two or more degraded voltage relay channels or one or more associated timers and the loss of voltage relay channel(s) inoperable on the same shutdown board, the associated diesel generator will not automatically start upon degraded voltage or complete loss of voltage on that shutdown board. In this situation, Required Action D.2 provides a 5 day allowable out of service time provided the other shutdown boards and undervoltage relay channels are OPERABLE. Immediate verification of the OPERABILITY of the other shutdown boards and undervoltage relay channels is therefore required (Required Action D.1 ). This may be performed as an administrative check by examining logs or other information to determine if this equipment is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate OPERABILITY of this equipment. If the OPERABILITY of this equipment cannot be verified, however, Condition F must be entered immediately. The 5 day allowable out of service time is justified based on the remaining redundancy of the 4.16 kV Shutdown Boards. The 4.16 kV Shutdown Boards have a similar allowable out of service time.
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 D.2.
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 F must be entered and its Required Action taken.
(continued)
BFN-UNIT 2                    B 3.3-264                          Revision 0, 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS      E.1 and E.2 (continued)
The Unbalanced Voltage function generates an LOP signal if the permissive alarm relay and either the Low or High relay actuates to the predetermined unbalanced voltage setting. With one or more UVRs inoperable, the associated diesel generator will not automatically start upon an Unbalanced Voltage signal.
In this situation, Required Action E.2 provides a 5 day allowable out of service time provided the other shutdown boards and Unbalanced Voltage relays are OPERABLE. Immediate verification of the OPERABILITY of the other shutdown boards and UVRs is required (Required Action E.1 ). This action may be performed as an administrative check by examining logs or other information to determine if this equipment is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate OPERABILITY of this equipment. If the OPERABILITY of this equipment cannot be verified, however, Condition F must be entered immediately.
The 5 day allowable out of service time is justified based on the remaining redundancy of the 4.16 kV shutdown boards. The 4.16kV shutdown boards have a similar allowable out of service time. If the inoperable relay cannot be restored to OPERABLE status within the allowable out of service time, the relay must be placed in the tripped condition. Placing the inoperable relay 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 relay in trip (e.g., as in the case where placing the relay in trip would result in a DG initiation), Condition F must be entered and its Required Action taken.
(continued)
BFN-UNIT 2                      B 3.3-265                          Revision 0, 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 SURVEILLANCE As noted (Note 1) at the beginning of the SRs, the SRs for REQUIREMENTS each LOP instrumentation Function are located in the SRs column of Table 3.3.8.1-1.
SR 3.3.8.1.1 and SR 3.3.8.1.2 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 Frequency is based upon the calibration interval assumed in the determination of the magnitude of equipment drift in the setpoint analysis.
{continued)
BFN-UNIT 2                  B 3.3-265a                          Revision Q., 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES SURVEILLANCE SR 3.3.8.1.3 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specific 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 with these components supports performance of the Surveillance at the 24 month Frequency.
REFERENCES  1. FSAR, Figure 8.4-4.
: 2. FSAR, Section 6.5.
: 3. FSAR, Section 8.5.4.
: 4. FSAR, Chapter 14.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 6. EDQ0000002016000556, "Determination of Unbalance Voltage Relay Analytical Limits."
: 7. EDQ0009992016000564, "Evaluation of 60Q Voltage Unbalance Relays for Class 1E 4kV Shutdown Boards A, B, C, D, 3EA, 3EB, 3EC, and 3ED."
BFN-UNIT 2                    B 3.3-266                          Revision 118 Amendment No. 255 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 Table B 3.3.8.1-1 (Page 1 of 2)
Loss of Power Instrumentation Channel Device Identification BOARD AND FUNCTIONS                                              CHANNEL DEVICES (UNIDs) 4.16 kV Shutdown Board A (Loss of Voltage) 1.a    Board Undervoltage - Board Load Shedding                            27SA 0A and 27SA 0C (27-211-000A/12E & /12F) 1.b    Board Undervoltage - Diesel Start Time Delay                        27DA 0A and 27DA 0C (27-211-000A/12A & /12B)
(Degraded Voltage) 2.a    Board Undervoltage                                                  27-211-1A, 27-211-1B, and 27-211-1C (27-211-000A/23A, /23B, & /23C) 2-211-1A (02-211-0001A) 2.b.1  Initial Diesel Start and Load Shedding Time Delay 2-211-2A (02-211-0002A) 2.b.2  Diesel Start Time Delay                                            2-211-3A (02-211-0003A) 2.b.3  Board Load Shedding Time Delay                                      2-211-4A (02-211-0004A) 2.b.4  Diesel Generator Breaker Closure Time Delay (Unbalanced Voltage)                                                      RLY-211-60A1 3.a    Board Voltage Unbalance - Board Load Shedding                      RLY-211-60A2 RLY-211-60A3 3.b    Board Voltage Unbalance - Diesel Start Time Delay                  RLY-211-A60A1 RLT-211-A60A2 4.16 kV Shutdown Board B (Loss of Voltage) 1.a    Board Undervoltage - Board Load Shedding                            27SB 0A and 27SB 0C (27-211-000B/12E & /12F) 1.b    Board Undervoltage - Diesel Start Time Delay                        27DB 0A and 27DB 0C (27-211-000B/12A & /12B)
(Degraded Voltage) 27-211-2A, 27-211-2B, and 27-211-2C 2.a    Board Undervoltage (27-211-000B/21A, /21B, /21C) 2-211-1 B (02-211-0001 B) 2.b.1  Initial Diesel Start and Load Shedding Time Delay 2-211-2B (02-211-0002B) 2.b.2  Diesel Start Time Delay                                            2-211-3B (02-211-0003B) 2.b.3  Board Load Shedding Time Delay                                    2-211-4B (02-211-0004B) 2.b.4  Diesel Generator Breaker Closure Time Delay (Unbalanced Voltage) 3.a    Board Voltage Unbalance - Board Load Shedding                      RLY-211-60B1 RLY-211-60B2 3.b    Board Voltage Unbalance - Diesel Start Time Delay                  RLY-211-60B3 RLY-211-A60B1 RLT-211-A60B2 BFN-UNIT 2                                              B 3.3-267                                      Revision 0, 118 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 Table B 3.3.8.1-1 (Page 2 of 2)
Loss of Power Instrumentation Channel Device Identification BOARD AND FUNCTIONS                                              CHANNEL DEVICES (UNIDs) 4.16 kV Shutdown Board C (Loss of Voltage) 1.a    Board Undervoltage - Board Load Shedding                          27SC 0A and 27SC 0C (27-211-000C/11 E & /11 F) 1.b    Board Undervoltage - Diesel Start Time Delay                      27DC 0A and 27DC 0C (27-211-000C/11A & /11 B)
(Degraded Voltage) 2.a    Board Undervoltage                                                27-211-3A, 27-211-38, and 27-211-3C (27-211-000C/25A, /258, /25C) 2-211-1C (02-211-0001C) 2.b.1  Initial Diesel Start and Load Shedding Time Delay 2-211-2C (02-211-0002C) 2.b.2  Diesel Start Time Delay                                            2-211-3C (02-211-0003C) 2.b.3  Board Load Shedding Time Delay                                    2-211-4C (02-211-0004C) 2.b.4  Diesel Generator Breaker Closure Time Delay (Unbalanced Voltage) 3.a    Board Voltage Unbalance - Board Load Shedding                      RLY-211-60C1 RLY-211-60C2 3.b    Board Voltage Unbalance - Diesel Start Time Delay                  RLY-211-60C3 RLY-211-A60C1 RLT-211-A60C2 4.16 kV Shutdown Board D (Loss of Voltage) 1.a    Board Undervoltage - Board Load Shedding                          27SD 0A and 27SD 0C (27-211-000D/11E &/11F) 1.b    Board Undervoltage - Diesel Start Time Delay                      27DD 0A and 27DD 0C (27-211-000D/11A & /118)
(Degraded Voltage) 27-211-4A, 27-211-48, and 27-211-4C 2.a    Board Undervoltage (27-211-000D/21A, /21 B, /21 C) 2-211-10 (02-211-00010) 2.b.1  Initial Diesel Start and Load Shedding Time Delay 2-211-2D (02-211-00020) 2.b.2  Diesel Start Time Delay                                          2-211-3D (02-211-0003D) 2.b.3  Board Load Shedding Time Delay                                    2-211-4D (02-211-0004D) 2.b.4  Diesel Generator Breaker Closure Time Delay (Unbalanced Voltage) 3.a    Board Voltage Unbalance - Board Load Shedding                      RLY-211-60D1 RLY-211-60D2 3.b    Board Voltage Unbalance - Diesel Start Time Delay                  RLY-211-60D3 RLY-211-A60D1 RLT-211-A60D2 BFN-UNIT 2                                              B 3.3-268                                      Revision 0, 118 December 12, 2019
 
AC Sources - Operating B 3.8.1 BASES BACKGROUND  sufficient capacity to support the automatic transfer of all Unit 2 (continued) non-safety related loads when there are existing loads aligned to the CSSTs from Units 1 or 3.
This is addressed by manually disabling the automatic transfer of selected 4.16 kV Unit Boards and/or 4.16 kV Common Boards. With the most restrictive manual actions in place, upon a loss of the normal 500 kV offsite circuit coincident with a LOCA, the diesel generators would supply the associated safety-related ESF loads in both divisions needed to mitigate the immediate consequences of a LOCA.
The 161 kV supplied CSSTs can still be credited as part of a qualified alternate offsite circuit for Unit 2. However, access to the 161 kV circuit will require a delayed manual transfer when operators can manually control the loads on the 4.16 kV Start Buses to support long term post accident recovery and shutdown. Operators can restore the de-energized 4.16 kV Unit Boards by manually transferring them to the CSST supplied 4.16 kV Start Buses as desired. The 4.16 kV Shutdown Boards could then be manually transferred from the diesel generators to the CSST supplied 4.16 kV Unit Boards as desired.
The onsite standby power source for 4.16 kV shutdown boards A, B, C, and D consists of four Unit 1 and 2 DGs, each dedicated to a shutdown board. Each DG starts automatically on a LOCA signal (i.e., low reactor water level signal or high drywell pressure signal), or on its respective 4.16 kV shutdown board degraded voltage, unbalanced voltage, or undervoltage signal. In addition to starting all diesel generators, the GAS logic trips the alternate feeder breakers to 4.16 kV Shutdown Boards A, B, C, D. After the DG has started, it automatically ties to its respective bus after offsite power is tripped as a consequence of 4.16 kV shutdown board undervoltage, unbalanced voltage, or degraded voltage, independent of or coincident with a LOCA signal. The DGs also start and operate in the standby mode without tying to the 4.16 kV shutdown board on a LOCA signal alone. Following (continued)
BFN-UNIT 2                      B 3.8-3a                          Revision ~. 118 December 12, 2019
 
AC Sources - Operating B 3.8.1 BASES BACKGROUND  the trip of offsite power, an undervoltage, unbalanced voltage, (continued) or degraded voltage activated load shed logic strips all loads from the 4.16 kV Shutdown Board. Feeder breakers to transformers supplying auxiliary power system distribution boards are not load shed on undervoltage. When the DG is tied to the 4.16 kV shutdown board, large loads are then sequentially connected to its respective 4.16 kV shutdown board by individual pump timers. The individual pump timers control the permissive and starting signals to motor breakers to prevent overloading the DG In the event of a loss of offsite power, the ESF electrical loads are automatically connected to the DGs in sufficient time to provide for safe reactor shutdown and to mitigate the consequences of a Design Basis Accident (OBA) such as a LOCA.
Certain required plant loads are returned to service in a predetermined sequence in order to prevent overloading of the DGs in the process. Within 40 seconds _after the initiating signal (DG breaker closure with accident signal) is received, all automatic and permanently connected loads needed to recover the unit or maintain it in a safe condition are returned to service.
In the event that the DGs were already running and loaded on the receipt of a spurious or real common accident signal (CAS A/CAS B) from Unit 3, any diesel generator output breakers which are closed are signaled to open to load shed the running loads off of the DG. After the DG breaker closing springs recharge, the DG breakers will reclose and tie the DG to the 4.16 kV shutdown board. Loads are then sequentially connected to its respective 4.16 kV shutdown board by individual pump timers as described above. Any subsequent common accident signal DG breaker trip signals are blocked.
Should a second RHR initiation signal be received (i.e., from a spurious or real accident signal from Unit 2), the Unit 1/2 diesel generator output breakers will be reopened on a unit priority (continued)
BFN-UNIT 2                      B 3.8-4          Revision 0, 4-0, JG, 42c, 4-7, 118 December 12, 2019
 
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 shutdown boards. Offsite power is the preferred source of power for the 4.16 kV shutdown boards. If the monitors determine that insufficient power is available, the boards are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources.
Each 4.16 kV shutdown board has its own independent LOP instrumentation and associated trip logic. The voltage for each board is monitored at three levels, which can be considered as three different undervoltage Functions: Loss of Voltage, Unbalanced Voltage, and 4.16 kV Shutdown Board Undervoltage Degraded Voltage. Each Function causes various board transfers and disconnects.
The Degraded Voltage Function is monitored by three undervoltage relay channels for each shutdown board, whose outputs are arranged in a two-out-of-three logic configuration (Ref. 1}. The channels compare measured input signals with pre-established setpoints. When the setpoint is exceeded for two-of-three degraded voltage channels, the logic energizes timers which provides a LOP trip signal to the shutdown board logic.
The unbalanced voltage function is monitored by three unbalanced voltage relays (UVRs) for each shutdown board, whose outputs are arranged in a permissive one-out-of-two logic configuration. The UVRs operate on an unbalanced voltage (continued)
BFN-UNIT 3                              B 3.3-256                            Revision 118 Amendment No. 213 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES BACKGROUND      detection signal dependent on the length of time the signal is (continued)    detected. If the permissive one-out-of-two logic is met, the relays energize auxiliary relays to provide the trip signal to the shutdown board logic. A permissive one-out-of-two trip logic is defined as a trip of the "Alarm" relay and either the "High" or "Low" relay.
The Loss of Voltage Function is monitored by two undervoltage relay pairs for each shutdown board, where outputs are arranged in a two-out-of-two logic configuration (Ref. 1). The channels include four electro-mechanical relays, two of which must deenergize to start the associated diesel generator and another two which must deenergize to initiate load shed of the associated 4.16 kV shutdown board.
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 2, 3, and 4 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.
Accident analyses credit the loading of the DG based on the loss of offsite power concurrent with a loss of coolant accident.
The diesel starting and loading times have been included in the delay time associated with each safety system component requiring DG supplied power following a loss of offsite power.
The LOP instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref. 5).
(continued)
BFN-UNIT 3                        B 3.3-257                            Revision 118 Amendment No. 213 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES APPLICABLE      2. 4.16 kV Shutdown Board Undervoltage (Degraded Voltage)
SAFETY ANALYSES, LCO, and        A reduced voltage condition on a 4.16 kV shutdown board APPLICABILITY    indicates that, while offsite power may not be completely lost (continued)    to the respective shutdown board, available power maybe insufficient for starting large ECCS motors without risking damage to the motors that could disable the ECCS function. Therefore, power supply to the board is transferred from offsite power to onsite DG power when the voltage on the board 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 Board 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 Shutdown Board Undervoltage (Degraded Voltage) Function per associated board are 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. Refer to LCO 3.8.1 and LCO 3.8.2 for Applicability Bases for the DGs.
: 3. 4.16 kV Shutdown Board Voltage Unbalance (Unbalanced Voltage Relay)
An unbalanced voltage condition on a 4.16kV shutdown board indicates that, while offsite power may not be completely degraded to the board undervoltage level, available power may be insufficient for starting and running ECCS motors without risking damage to the motors that could disable the ECCS function.
Therefore, power supply to the board is transferred from offsite power to onsite DG power when the unbalanced voltage level increases above the Unbalanced Voltage Function Allowable Values (unbalanced voltage level with an associated time delay).
(continued)
BFN-UNIT 3                          B 3.3-260                            Revision 118 Amendment No. 213 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES APPLICABLE      3. 4.16 kV Shutdown Board Voltage Unbalance (Unbalanced SAFETY ANALYSES, Voltage Relay)
LCO, and APPLICABILITY    This ensures adequate power will be available to the required equipment. The Board Unbalanced Voltage Allowable Values are high enough to prevent inadvertent power supply transfer, but low 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 voltage balance, but short enough to ensure power is available to the required equipment.
Three UVRs are provided on each 4.16 kV Shutdown Board for detecting an unbalanced voltage condition. The relays are combined in a permissive one-out-of-two logic configuration to generate a supply breaker trip. Three UVRs are required to be OPERABLE when the associated DG is required to be OPERABLE to ensure that no single instrument failure can preclude a DG function. Refer to LCO 3.8.1 and LCO 3.8.2 for Applicability Bases for the DGs.
(continued)
BFN-UNIT 3                      B 3.3-260a                              Revision 118 Amendment No. 213 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS    A.1 and A.2 (continued)
Condition C or D, as applicable, must be entered immediately.
The 15 day allowable out of service time is justified based on the two-out-of-three permissive logic scheme provided for these relays. If the inoperable relay channel cannot be restored to OPERABLE status within the allowable out of service time, the degraded voltage relay channel must be placed in the tripped condition per Required Action A.2. 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 F must be entered and its Required Action taken.
With two or more degraded voltage relay channels or one or more associated timers inoperable on one or more shutdown boards, the Function is not capable of performing the intended function. Required Action B.1 provides a 10 day allowable out of service time provided the loss of voltage relay channels on the affected shutdown board(s) are OPERABLE.
The 10 day allowable out of service time is justified since the loss of voltage relay channels on the same shutdown board are independent of the degraded voltage relay channel(s) and will continue to function and start the diesel generators on a complete loss of voltage. If the inoperable channel(s) cannot (continued)
BFN-UNIT 3                    B 3.3-262                            Revision 118 Amendment No. 213 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS    B.1 (continued) be restored to OPERABLE status within the allowable out of service time, the channel(s) must be placed in the tripped condition per Required Action B.1. Placing the inoperable channel(s) 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(s) in trip (e.g., as in the case where placing the channel(s) in trip would result in a DG initiation), Condition F must be entered and its Required Action taken.
With one or more loss of voltage relay channels inoperable on one or more shutdown boards, the Function is not capable of performing the intended function. Required Action C.1 provides a 10 day allowable out of service time provided two or more degraded voltage relay channels and associated timers on the affected shutdown board(s) are OPERABLE. The 10 day allowable out of service time is justified since the degraded voltage relay channels on the same shutdown board are independent of the loss of voltage relay channels and will continue to function and start the diesel generators on a complete loss of voltage. If the inoperable channels cannot be restored to OPERABLE status within the allowable out of service time, the channel(s) must be placed in the tripped condition per Required Action C.1. Placing the inoperable channel(s) 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(s) in trip (e.g., as in the case where placing the channel(s) in trip would result in a DG initiation), Condition F must be entered and its Required Action taken.
(continued)
BFN-UNIT 3                    B 3.3-263                            Revision 118 Amendment No. 213 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS      D.1 and D.2 (continued)
With two or more degraded voltage relay channels or one or more associated timers and the loss of voltage relay channel(s) inoperable on the same shutdown board, the associated diesel generator will not automatically start upon degraded voltage or complete loss of voltage on that shutdown board. In this situation, Required Action D.2 provides a 5 day allowable out of service time provided the other shutdown boards and undervoltage relay channels are OPERABLE. Immediate verification of the OPERABILITY of the other shutdown boards and undervoltage relay channels is therefore required (Required Action D.1 ). This may be performed as an administrative check by examining logs or other information to determine if this equipment is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate OPERABILITY of this equipment. If the OPERABILITY of this equipment cannot be verified, however, Condition F must be entered immediately. The 5 day allowable out of service time is justified based on the remaining redundancy of the 4.16 kV Shutdown Boards. The 4.16 kV Shutdown Boards have a similar allowable out of service time.
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 D.2.
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 F must be entered and its Required Action taken.
(continued)
BFN-UNIT 3                    B 3.3-264                            Revision 118 Amendment No. 213 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS      E.1 and E.2 (continued)
The Unbalanced Voltage function generates an LOP signal if the permissive alarm relay and either the Low or High relay actuates to the predetermined unbalanced voltage setting. With one or more UVRs inoperable, the associated diesel generator will not automatically start upon an Unbalanced Voltage signal.
In this situation, Required Action E.2 provides a 5 day allowable out of service time provided the other shutdown boards and Unbalanced Voltage relays are OPERABLE. Immediate verification of the OPERABILITY of the other shutdown boards and UVRs is required (Required Action E.1 ). This action may be performed as an administrative check by examining logs or other information to determine if this equipment is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate OPERABILITY of this equipment. If the OPERABILITY of this equipment cannot be verified, however, Condition F must be entered immediately.
The 5 day allowable out of service time is justified based on the remaining redundancy of the 4.16 kV shutdown boards. The 4.16kV shutdown boards have a similar allowable out of service time. If the inoperable relay cannot be restored to OPERABLE status within the allowable out of service time, the relay must be placed in the tripped condition. Placing the inoperable relay 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 relay in trip (e.g., as in the case where placing the relay in trip would result in a DG initiation), Condition F must be entered and its Required Action taken.
(continued)
BFN-UNIT 3                      B 3.3-265                              Revision 118 Amendment No. 213 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES REFERENCES 1. FSAR, Figure 8.4-4.
: 2. FSAR, Section 6.5.
: 3. FSAR, Section 8.5.4.
: 4. FSAR, Chapter 14.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 6. EDQ0000002016000556, "Determination of Unbalance Voltage Relay Analytical Limits."
: 7. EDQ0009992016000564, "Evaluation of 6Q Voltage Unbalance Relays for Class 1E 4kV Shutdown Boards A, B, C, D, 3EA, 3EB, 3EC, and 3ED.
BFN-UNIT 3                B 3.3-266                        Revision 118 Amendment No. 215 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 Table B 3.3.8.1-1 (Page 1 of 2)
Loss of Power Instrumentation Channel Device Identification BOARD AND FUNCTIONS                                              CHANNEL DEVICES (UNIDs) 4.16 kV Shutdown Board 3EA (Loss of Voltage) 1.a    Board Undervoltage - Board Load Shedding                          27S3A A0 and 27S3A C0 (27-211-O3EA/O8E & /OBF) 1.b    Board Undervoltage - Diesel Start Time Delay                      27D3A A0 and 27D3A C0 (27-211-O3EA/O8A & OBB)
(Degraded Voltage) 2.a    Board Undervoltage                                                27-211-1A3, 27-211-1 B3, and 27-211-1C3 (27-211-O3EA/O3A, /O3B, & /O3C) 2.b.1  Initial Diesel Start and Load Shedding Time Delay                2-211-1A3 (O2-211-O3EA/O3A) 2.b.2  Diesel Start Time Delay                                          2-211-2A3 (O2-211-O3EA/O3B) 2.b.3  Board Load Shedding Time Delay                                    2-211-3A3 (O2-211-O3EA/O3C) 2.b.4  Diesel Generator Breaker Closure Time Delay                      2-211-4A3 (O2-211-O3EA/O3D)
(Unbalance Voltage) 3.a    Board Voltage Unbalance - Board Load Shedding                    3-RLY-211-6O3EA 1 3-RLY-211-6O3EA2 3.b    Board Voltage Unbalance - Diesel Start Time Delay                3-RLY-211-6O3EA3 3-RLY-211-A6O3EA 1 3-RLY-211-A6O3EA2 4.16 kV Shutdown Board 3EB (Loss of Voltage) 1.a    Board Undervoltage - Board Load Shedding                          27S3B A0 and 27S3B C0 (27-211-O3EB/O7E & /O7F) 1.b    Board Undervoltage - Diesel Start Time Delay                      27D3B A0 and 27D3B C0 (27-211-O3EB/O7A & /O?B)
(Degraded Voltage) 2.a    Board Undervoltage                                                27-211-2A3, 27-211-2B3, and 27-211-2C3 (27-211-O3EB/12A, /12B, & /12C) 2.b.1  Initial Diesel Start and Load Shedding Time Delay                2-211-1 B3 (O2-211-O3EB/12A) 2.b.2  Diesel Start Time Delay                                          2-211-2B3 (O2-211-O3EB/12B) 2.b.3  Board Load Shedding Time Delay                                    2-211-3B3 (O2-211-O3EB/12C) 2.b.4  Diesel Generator Breaker Closure Time Delay                      2-211-4B3 (O2-211-O3EB/12D)
(Unbalanced Voltage) 1.a    Board Voltage Unbalance - Board Load Shedding                    3-RLY-211-6O3EB1 3-RLY-211-6O3EB2 1.b    Board Voltage Unbalance - Diesel Start Time Delay                3-RLY-211-6O3EB3 3-RLY-211-A6O3EB1 3-RLY-211-A6O3EB2 BFN-UNIT 3                                              B 3.3-267                                      Revision 118 Amendment No. 213 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 Table B 3.3.8.1-1 (Page 2 of 2)
Loss of Power Instrumentation Channel Device Identification BOARD AND FUNCTIONS                                              CHANNEL DEVICES (UNIDs) 4.16 kV Shutdown Board 3EC (Loss of Voltage) 1.a    Board Undervoltage - Board Load Shedding                          27S3C A0 and 27S3C C0 (27-211-03EC/11 E & /11 F) 1.b    Board Undervoltage - Diesel Start Time Delay                      2703C A0 and 27D3C C0 (27-211-03EC/11A & /118)
(Degraded Voltage) 2.a    Board Undervoltage                                                27-211-3A3, 27-211-3B3, and 27-211-3C3 (27-211-03EC/05A, /05B, & /05C) 2.b.1  Initial Diesel Start and Load Shedding Time Delay                  2-211-1C3 (02-211-03EC/05A) 2.b.2  Diesel Start Time Delay                                            2-211-2C3 (02-211-03EC/05B) 2.b.3  Board Load Shedding Time Delay                                    2-211-3C3 (02-211-03EC/05C) 2.b.4  Diesel Generator Breaker Closure Time Delay                        2-211-4C3 (02-211-03EC/050)
(Unbalanced Voltage) 1.a    Board Voltage Unbalance - Board Load Shedding                      3-RLY-211-603EC1 3-RLY-211-603EC2 1.b    Board Voltage Unbalance - Diesel Start Time Delay                  3-RLY-211-603EC3 3-RLY-211-A603EC1 3-RLY-211-A603EC2 4.16 kV Shutdown Board 3ED (Loss of Voltage) 1.a    Board Undervoltage - Board Load Shedding                          27S3D A0 and 27S30 C0 (27-211-03ED/09E & /09F) 1.b    Board Undervoltage - Diesel Start Time Delay                      27030 A0 and 27030 C0 (27-211-03ED/09A & /09B)
(Degraded Voltage) 2.a    Board Undervoltage                                                27-211-4A3, 27-211-4B3, and 27-211-4C3 (27-211-03ED/03A, /03B, & /03C) 2.b.1  Initial Diesel Start and Load Shedding Time Delay                2-211-103 (02-211-03ED/03A) 2.b.2  Diesel Start Time Delay                                            2-211-203 (02-211-03ED/03B) 2.b.3  Board Load Shedding Time Delay                                    2-211-303 (02-211-03ED/03C) 2.b.4  Diesel Generator Breaker Closure Time Delay                        2-211-403 (02-211-03ED/03D)
(Unbalanced Voltage)                                                      3-RLY-211-603ED1 1.a    Board Voltage Unbalance - Board Load Shedding                    3-RLY-211-603ED2 3-RLY-211-603ED3 1.b    Board Voltage Unbalance - Diesel Start Time Delay                  3-RLY-211-A603ED1 3-RLY-211-A603ED2 BFN-UNIT 3                                              B 3.3-268                                      Revision 118 Amendment No. 213 December 12, 2019
 
AC Sources - Operating B 3.8.1 BASES BACKGROUND    The onsite standby power source for 4.16 kV shutdown boards (continued)  3EA, 3EB, 3EC, and 3ED consists of four Unit 3 DGs, each dedicated to a shutdown board. Each DG starts automatically
            .on a loss of coolant accident (LOCA) signal (i.e., low reactor water level signal or high drywell pressure signal), or on its respective 4.16 kV shutdown board degraded voltage, unbalance voltage, or undervoltage signal. Common Accident Signal Logic (CAS A/CAS B) actuates on high drywell pressure with low reactor pressure, or low water level.
After the DG has started, it automatically ties to its respective bus after offsite power is tripped as a consequence of 4.16 kV shutdown board undervoltage, unbalanced voltage, or degraded voltage, independent of or coincident with a LOCA signal. The DGs also start and operate in the standby mode without tying to the 4.16 kV shutdown board on a LOCA signal alone. Following the trip of offsite power, an undervoltage, unbalanced voltage, or degraded voltage activated load shed logic strips all loads from the 4.16 kV Shutdown Board except transformer feeds.
When the DG is tied to the 4.16 kV shutdown board, large loads are then sequentially connected to its respective 4.16 kV shutdown board by individual pump timers. The individual pump timers control the permissive and starting signals to motor breakers to prevent overloading the DG.
In the event of a loss of offsite power, the ESF electrical loads are automatically connected to the DGs in sufficient time to provide for safe reactor shutdown and to mitigate the consequences of a Design Basis Accident (OBA) such as a LOCA.
(continued)
BFN-UNIT 3                        B 3.8-3                          Revision Q., 118 December 12, 2019
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, LCO, and        The OPRM Upscale Function provides compliance with APPLICABILITY    10 CFR 50, Appendix A, General Design Criteria (GDC) 10 and (continued)    12, thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations.
Reference 13 describes the Detect and Suppress - Confirmation Density (DSS-CD) long-term stability solution and the licensing basis Confirmation Density Algorithm (CDA). Reference 13 also describes the DSS-CD Armed Region and the three additional algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm (PBDA), the amplitude based algorithm (ABA), and the growth rate algorithm (GRA). All four algorithms are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the CDA. The remaining three algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY is based only on the CDA.
The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into cells for evaluation by the OPRM algorithms.
DSS-CD OPERABILITY requires at least 8 responsive OPRM cells per channel. The DSS-CD software includes a self-check for the responsive OPRM cells; therefore, no SR is necessary.
The OPRM Upscale Function is required to be OPERABLE above a power level set at 5% of rated power below the lower boundary of the Armed Region, which is ~18% RTP. This requirement is designed to encompass the region of power-flow operation where anticipated events could lead to thermal-hydraulic instability and related neutron flux oscillations. The OPRM Upscale Function is automatically trip-enabled when (continued)
BFN-UNIT 1                        B 3.3-16a                  Revision 4-a, ~ . 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY    THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 2:23% RTP corresponding to the MCPR monitoring threshold and reactor recirculation drive flow, is less than 75% of rated flow. This region is the OPRM Armed Region.
Note (f) allows for entry into the DSS-CD Armed Region without automatic arming of DSS-CD prior to completely passing through the DSS-CD Armed Region during the first startup and the first shutdown following DSS-CD implementation.
If any OPRM auto-enable setpoint is in a non-conservative condition, i.e., the OPRM Upscale is not auto-enabled with RTP 2: 23% and reactor recirculation drive flows 75% of rated, the associated channel is considered inoperable for the OPRM Upscale function. Alternatively, the auto-enable setpoint may be adjusted to place the channel in a conservative condition (armed). If placed in the armed condition, the channel is considered OPERABLE.
Note (f) reflects the need for plant data collection in order to test the DSS-CD equipment. Testing the DSS-CD equipment ensures its proper operation and prevents spurious reactor trips. Entry into the DSS-CD Armed Region without automatic arming of DSS-CD during this initial testing phase also allows for changes in plant operations to address maintenance or other operational needs. However, during this initial testing period, the OPRM upscale function is OPERABLE and DSS-CD operability and capability to automatically arm shall be maintained at recirculation drive flow rates above the DSS-CD Armed Region flow boundary.
An OPRM Upscale trip is issued from an OPRM channel when the confirmation density algorithm in that channel detects oscillatory changes in the neutron flux, indicated by periodic confirmations and amplitude exceeding specified setpoints for a specified number of OPRM cells in the channel. An OPRM (continued)
BFN-UNIT 1                        B 3.3-16b                        Revision 4&sect;, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY    Upscale trip is also issued from the channel if any of the defense-in-depth algorithms (PBDA, ABA, GRA) exceed their trip condition for one or more cells in that channel.
Three of the four channels are required to be OPERABLE. Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the SLMCPR is exceeded. There is no Allowable Value for this function.
The OPRM Upscale Function is not LSSS SL-related (Reference 13) and Reference 14 confirms that the OPRM Upscale Function settings based on DSS-CD also do not have traditional instrumentation setpoints determined under an instrument setpoint methodology.
(continued)
BFN-UNIT 1                        B 3.3-16c                        Revision 4-a, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS    C.1 (continued)
For Function 8 (Turbine Stop Valve - Closure), this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated trip system in trip).
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.
D.1 Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition.
E.1, F.1, and G.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply.
The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Action E.1 is consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)."
(continued)
BFN-UNIT 1                  B 3.3-33                      Revision 0, 4&sect;, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS      H.1 (continued)
If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply.
This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
u If OPRM Upscale trip capability is not maintained, Condition I exists and Backup Stability Protection (BSP) is required. The Manual BSP Regions are described in Reference 13. The Manual BSP Regions are procedurally established consistent with the guidelines identified in Reference 13 and require specified manual operator actions if certain predefined operational conditions occur. The Completion Time of immediate is based on the importance of limiting the period of time during which no automatic or alternate detect and suppress trip capability is in place.
1.2 and 1.3 Required Actions 1.2 and 1.3 are both required to be taken in conjunction with Required Action 1.1 if OPRM Upscale trip capability is not maintained. As described in Section 7.4 of Reference 13, the Automated BSP Scram Region is designed to avoid reactor instability by automatically preventing entry into the region of the power and flow operating map that is susceptible to reactor instability. The reactor trip would be
{continued)
BFN-UNIT 1                      B 3.3-34                      Revision 0, 4-a, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS    1.2 and 1.3 (continued) initiated by the modified APRM Simulated Thermal Power-High scram setpoints for flow reduction events that would have terminated in the Manual BSP Region I. The Automated BSP Scram Region ensures an early scram and SLMCPR protection.
The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, and based on the importance of restoring an automatic reactor trip for thermal-hydraulic instability events. Backup Stability Protection is intended as a temporary means to protect against thermal-hydraulic instability events. The action should be initiated immediately to document the situation and prepare the report. The reporting requirements of Specification 5.6.7 document the corrective actions and schedule to restore the required channels to an OPERABLE status. The Completion Time of 90 days shown in Specification 5.6. 7 is adequate to allow time to evaluate the cause of the inoperability and to determine the appropriate corrective actions and schedule to restore the required channels to OPERABLE status.
4:.1 If the Required Action I is not completed within the associated Completion Time, then Action J is required. The Bases for the Manual BSP Regions and associated Completion Time is addressed in the Bases for 1.1. The Manual BSP Regions are required in conjunction with the BSP Boundary.
The BSP Boundary, as described in Section 7.3 of Reference 13, defines an operating domain where potential instability events can be effectively addressed by the specified BSP manual operator actions. The BSP Boundary is constructed such that a flow reduction event initiated from this boundary (continued)
BFN-UNIT 1                  B 3.3-34a                    Revision 0, 4-a, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS    J.2 (continued) and terminated at the core natural circulation line (NCL) would not exceed the Manual BSP Region I stability criterion. Potential instabilities would develop slowly as a result of the feedwater temperature transient (Reference 13).
The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, to reach the specific condition from full power conditions in an orderly manner and without challenging plant systems.
J.3 Backup Stability Protection (BSP) is a temporary means for protection against thermal-hydraulic instability events. An extended period of inoperability without automatic trip capability is not justified. Consequently, the required channels are required to be restored to OPERABLE status within 120 days.
Based on engineering judgment, the likelihood of an instability event that could not be adequately handled by the use of the BSP Regions (See Required Action J.1) and the BSP Boundary (See Required Action J.2) during a 120-day period is negligibly small. The 120-day period is intended to allow for the case where limited design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment. This action is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to normally be accomplished within the completion times allowed for Actions for Conditions A and B.
A Note is provided to indicate that LCO 3.0.4 is not applicable.
The intent of the note is to allow plant startup while operating within the 120 day Completion Time for Required Action J.3.
The primary purpose of this exclusion is to allow an orderly (continued)
BFN-UNIT 1                  B 3.3-34b                      Revision 0, 4a, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS      J.3 (continued)
(continued) completion of design and verification activities, in the event of a required design change, without undue impact on plant operation.
K.1 If the required channels are not restored to OPERABLE status and the Required Actions of Condition J are not met within the associated Completion Times, then the plant must be placed in an operating condition in which the LCO does not apply. To achieve this status, the plant must be brought to less than 18%
RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the specified operating power level from full power conditions in an orderly manner and without challenging plant systems.
(continued)
BFN-UNIT 1                    B 3.3-34c                      Revision 0, 4-a, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES (continued)
SURVEILLANCE      As noted at the beginning of the SRs, the SRs for each RPS REQUIREMENTS      instrumentation Function are located in the SRs column of Table 3.3.1.1-1.
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 6 hours, provided the associated Function maintains RPS 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. 3) 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 RPS will trip when necessary.
(continued)
BFN-UNIT 1                        B 3.3-34d                      Revision 4-a, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.17 REQUIREMENTS (continued) (Deleted)
(continued)
BFN-UNIT 1                B 3.3-43a Revision 4-a, ~ . 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES (continued)
REFERENCES        1. FSAR, Section 7.2.
: 2. FSAR, Chapter 14.
: 3. NEDO-23842, "Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
: 4. FSAR, Appendix N.
: 5. FSAR, Section 14.6.2.
: 6. FSAR, Section 6.5.
: 7. FSAR, Section 14.5.
: 8. P. Check (NRG) letter to G. Lainas (NRG), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.
: 9. NEDC-30851-P-A , "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.
: 10. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 11. MED-32-0286, "Technical Specification Improvement Analysis for Browns Ferry Nuclear Plant, Unit 2," October 1995.
: 12. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option Ill Stability Trip Function," October 1995.
: 13. GE Hitachi Nuclear Energy, "GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density," NEDC-33075P-A, Revision 8, November 2013.
: 14. GEH letter to NRC, "NEDC-33075P-A, Detect and Suppress Solution - Confirmation Density (DCC-CD), Analytical Limit (TAC No. MD0277)," October 29, 2008, (ADAMS Accession No. ML083040052).
BFN-UNIT 1                          B 3.3-44                    Revision G, 4-0, 4&, 119 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES APPLICABLE      Plant specific LOCA analyses have been performed assuming SAFETY ANALYSES only one operating recirculation loop. These analyses have (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, provided the APLHGR requirements are modified accordingly (Refs. 7 and 8). The Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain has not been analyzed for single recirculation loop operation. Therefore, single loop operation is prohibited in the MELLLA+ operating domain (Ref. 10).
The transient analyses of Chapter 14 of the FSAR have also been performed for single recirculation loop operation (Ref. 7) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoint is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Flow Biased Simulated Thermal Power-High setpoint is in LCO 3.3.1.1, "Reactor Protection System (RPS)
Instrumentation."
Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 6).
(continued)
BFN-UNIT 1                        B 3.4-4                      Revision 48, aO, 119 Amendment No. 236 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
LCO              Two recirculation loops are required to be in operation with their flows 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. With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation. 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 Flow Biased Simulated Thermal Power-High Setpoint (LCO 3.3.1.1) may be applied to allow continued operation consistent with the assumptions of References 7 and 8. Single loop operation is prohibited in the MELLLA+ operating domain. This prohibition is consistent with Reference 10.
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.
(continued)
BFN-UNIT 1                          B 3.4-5                          Revision ~. 119 Amendment No. 236 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
ACTIONS          A.1 With the requirements of the LCO not met, the recirculation loops must be restored to operation with matched flows within 24 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow 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 the operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence. Note that single loop operation is prohibited in the MELLLA+
operating domain per Reference 10.
(continued)
BFN-UNIT 1                            B 3.4-6                      Revision 0, 4-e, 119 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES ACTIONS    A.1 (continued)
The 24 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, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.
This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump.
B.1 The Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain is not analyzed for single recirculation loop operation. Therefore, single loop operation is prohibited in the MELLLA+ operating domain (Ref. 10). Action shall be taken to immediately exit the MELLLA+ operating domain in order to return to operation at an analyzed condition.
(continued)
BFN-UNIT 1                  B 3.4-7                        Revision 4&, 119 Amendment No. 236 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES ACTIONS      C.1 (continued)
With no recirculation loops in operation while in MODES 1 or 2 or the Required Action and associated Completion Time of Condition A or B not met, 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 MODE 3 within 12 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics.
The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
(continued)
BFN-UNIT 1                    B 3.4-8                        Revision 4&sect;, 119 Amendment No. 236 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
REFERENCES        1. FSAR, Section 14.6.3.
: 2. FSAR, Section 4.3.5.
: 3. Deleted.
: 4. Deleted.
: 5. Deleted.
: 6. NRG No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 7. NEDO-24236, "Browns Ferry Nuclear Plant Units 1, 2, and 3, Single-Loop Operation," May 1981.
: 8. NEDC-32484P, "Browns Ferry Nuclear Plant Units 1, 2, and 3, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," Revision 6, February 2005.
: 9. ANP-3015(P), "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis," Revision 0, September 2011.
: 10. NEDC-33006P-A, "General Electrical Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report," Revision 3, June 2009.
BFN-UNIT 1                        B 3.4-10                    Revision 4-a, 8-7, 119 Amendment No. 236 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, LCO, and        The OPRM Upscale Function provides compliance with APPLICABILITY    10 CFR 50, Appendix A, General Design Criteria (GDC) 10 and (continued)    12, thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal hydraulic power oscillations.
Reference 13 describes the Detect and Suppress - Confirmation Density (DSS-CD) long-term stability solution and the licensing basis Confirmation Density Algorithm (CDA). Reference 13 also describes the DSS-CD Armed Region and the three additional algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm (PBDA), the amplitude based algorithm (ABA), and the growth rate algorithm (GRA). All four algorithms are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the CDA. The remaining three algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY is based only on the CDA.
The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation of the OPRM algorithms.
DSS-CD OPERABILITY requires at least 8 responsive OPRM cells per channel. The DSS-CD software includes a self-check for the responsive OPRM cells; therefore, no SR is necessary.
The OPRM Upscale Function is required to be OPERABLE above a power level set at 5% of rated power below the lower boundary of the Armed Region, which is 18% RTP. This requirement is designed to encompass the region of power-flow operation where anticipated events could lead to thermal-hydraulic instability and related neutron flux oscillations. The (continued)
BFN-UNIT 2                        B 3.3-15a                      Revision 113, 119 Amendment No. 258 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY    OPRM Upscale Function is automatically trip-enabled when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 23% RTP corresponding to the MCPR monitoring threshold and reactor recirculation drive flow, is less than 75% of rated flow. This region is the OPRM Armed Region. Note (f) allows for entry into the DSS-CD Armed Region without automatic arming of DSS-CD prior to completely passing through the DSS-CD Armed Region during the first startup and the first shutdown following DSS-CD implementation.
If any OPRM auto-enable setpoint is in a non-conservative condition, i.e., the OPRM Upscale is not auto-enabled with RTP 23% and reactor recirculation drive flow  75% of rated, the associated channel is considered inoperable for the OPRM Upscale function. Alternatively, the auto-enable setpoint may be adjusted to place the channel in a conservative condition (armed). If placed in the armed condition, the channel is considered OPERABLE.
Note (f) reflects the need for plant data collection in order to test the DSS-CD equipment. Testing the DSS-CD equipment ensures its proper operation and prevents spurious reactor trips.
Entry into the DSS-CD Armed Region without automatic arming of DSS-CD during this initial testing phase also allows for changes in plant operations to address maintenance or other operational needs. However, during this initial testing period, the OPRM Upscale Function is OPERABLE and DSS-CD OPERABILITY and capability to automatically arm shall be maintained at recirculation drive flow rates above the DSS-CD Armed Region flow boundary.
(continued)
BFN-UNIT 2                        B 3.3-15b                            Revision 119 Amendment No. 258 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY    An OPRM Upscale trip is issued from an OPRM channel when the CDA in that channel detects oscillatory changes in the neutron flux, indicated by periodic confirmations and amplitude exceeding specified setpoints for a specified number of OPRM cells in the channel. An OPRM Upscale trip is also issued from the channel if any of the defense-in-depth algorithms (PBDA, ABA, GRA) exceed their trip condition for one or more cells in that channel.
Three of the four channels are required to be OPERABLE.
Each channel is capable of detecting thermal hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the SLMCPR is exceeded.
There is no Allowable Value for this function.
The OPRM Upscale Function is not LSSS SL-related (Reference 13) and Reference 14 confirms that the OPRM Upscale Function settings based on DSS-CD also do not have traditional instrumentation setpoints determined under an instrument setpoint methodology.
(continued)
BFN-UNIT 2                        B 3.3-15c                            Revision 119 Amendment No. 258 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS      D.1 (continued)
Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition.
E.1, F.1, and G.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply.
The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Action E.1 is consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)."
(continued)
BFN-UNIT 2                    B 3.3-34                              Revision 119 Amendment No. 258 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS      H.1 (continued)
If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply.
This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
I.1 If OPRM Upscale trip capability is not maintained, Condition I exists and Backup Stability Protection (BSP) is required. The Manual BSP Regions are described in Reference 13. The Manual BSP Regions are procedurally established consistent with the guidelines identified in Reference 13 and require specified manual operator actions if certain predefined operational conditions occur. The Completion Time of immediate is based on the importance of limiting the period of time during which no automatic or alternate detect and suppress trip capability is in place.
I.2 and I.3 Required Actions I.2 and I.3 are both required to be taken in conjunction with Required Action I.1 if OPRM Upscale trip capability is not maintained. As described in Section 7.4 of Reference 13, the Automated BSP Scram Region is designed to avoid reactor instability by automatically preventing entry into the region of the power and flow operating map that is susceptible to reactor instability. The reactor trip would be initiated by the modified APRM Simulated Thermal Power-High scram setpoints for flow reduction events that would have (continued)
BFN-UNIT 2                      B 3.3-35                        Revision 14, 119 Amendment No. 258 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS      terminated in the Manual BSP Region I. The Automated BSP (continued) Scram Region ensures an early scram and SLMCPR protection.
The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, and based on the importance of restoring an automatic reactor trip for thermal hydraulic instability events. Backup Stability Protection is intended as a temporary means to protect against thermal-hydraulic instability events. The action should be initiated immediately to document the situation and prepare the report. The reporting requirements of Specification 5.6.7 document the corrective actions and schedule to restore the required channels to an OPERABLE status. The Completion Time of 90 days shown in Specification 5.6.7 is adequate to allow time to evaluate the cause of the inoperability and to determine the appropriate corrective actions and schedule to restore the required channels to OPERABLE status.
J.1 If the Required Action I is not completed within the associated Completion Time, then Action J is required. The Bases for the Manual BSP Regions and associated Completion Time is addressed in the Bases for I.1. The Manual BSP Regions are required in conjunction with the BSP Boundary.
J.2 The BSP Boundary, as described in Section 7.3 of Reference 13, defines an operating domain where potential instability events can be effectively addressed by the specified BSP manual operator actions. The BSP Boundary is constructed such that a flow reduction event initiated from this boundary and terminated at the core natural circulation line (NCL) would not exceed the Manual BSP Region I stability criterion. Potential instabilities would develop slowly as a result of the feedwater temperature transient (Reference 13).
(continued)
BFN-UNIT 2                    B 3.3-35a                        Revision 14, 119 Amendment No. 258 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS      The Completion Time of 12 hours to complete the specified (continued) actions is reasonable, based on operational experience, to reach the specific condition from full power conditions in an orderly manner and without challenging plant systems.
J.3 Backup Stability Protection (BSP) is a temporary means for protection against thermal-hydraulic instability events. An extended period of inoperability without automatic trip capability is not justified. Consequently, the required channels are required to be restored to OPERABLE status within 120 days.
Based on engineering judgment, the likelihood of an instability event that could not be adequately handled by the use of the BSP Regions (See Required Action J.1) and the BSP Boundary (See Required Action J.2) during a 120-day period is negligibly small. The 120-day period is intended to allow for the case where limited design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment. This action is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to normally be accomplished within the Completion Times allowed for Actions for Conditions A and B.
A Note is provided to indicate that LCO 3.0.4 is not applicable.
The intent of the note is to allow plant startup while operating within the 120 day Completion Time for Required Action J.3.
The primary purpose of this exclusion is to allow an orderly completion of design and verification activities, in the event of a required design change, without undue impact on plant operation.
(continued)
BFN-UNIT 2                    B 3.3-35b                        Revision 14, 119 Amendment No. 258 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS      K.1 (continued)
If the required channels are not restored to OPERABLE status and the Required Actions of Condition J are not met within the associated Completion Times, then the plant must be placed in an operating condition in which the LCO does not apply. To achieve this status, the plant must be brought to less than 18%
RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the specified operating power level from full power conditions in an orderly manner and without challenging plant systems.
(continued)
BFN-UNIT 2                    B 3.3-35c                      Revision 14, 119 Amendment No. 258 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.17 REQUIREMENTS (continued) (Deleted)
(continued)
BFN-UNIT 2                B 3.3-45a  Revision 113, 119 Amendment No. 258 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES (continued)
REFERENCES        1. FSAR, Section 7.2.
: 2. FSAR, Chapter 14.
: 3. NEDO-23842, "Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
: 4. FSAR, Appendix N.
: 5. FSAR, Section 14.6.2.
: 6. FSAR, Section 6.5.
: 7. FSAR, Section 14.5.
: 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.
: 9. NEDC-30851-P-A , "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.
: 10. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 11. MED-32-0286, "Technical Specification Improvement Analysis for Browns Ferry Nuclear Plant, Unit 2," October 1995.
: 12. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Trip Function," October 1995.
: 13. GE Hitachi Nuclear Energy, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, NEDC-33075P-A, Revision 8, November 2013.
: 14. GEH letter to NRC, "NEDC-33075P-A, Detect and Suppress Solution - Confirmation Density (DSS-CD) Analytical Limit (TAC No. MD0277)," October 29, 2008. (ADAMS Accession No. ML083040052).
BFN-UNIT 2                          B 3.3-46                              Revision 119 Amendment No. 258 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES APPLICABLE      Plant specific LOCA analyses have been performed assuming SAFETY ANALYSES only one operating recirculation loop. These analyses have (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, provided the APLHGR requirements are modified accordingly (Refs. 7 and 8). The Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain has not been analyzed for single recirculation loop operation. Therefore, single loop operation is prohibited in the MELLLA+ operating domain (Ref. 10).
The transient analyses of Chapter 14 of the FSAR have also been performed for single recirculation loop operation (Ref. 7) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoint is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Flow Biased Simulated Thermal Power-High setpoint is in LCO 3.3.1.1, Reactor Protection System (RPS)
Instrumentation.
Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 6).
(continued)
BFN-UNIT 2                        B 3.4-4                          Revision 3, 119 Amendment No. 258 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
LCO              Two recirculation loops are required to be in operation with their flows 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. With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation. 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 Flow Biased Simulated Thermal Power-High Setpoint (LCO 3.3.1.1) may be applied to allow continued operation consistent with the assumptions of References 7 and 8. Single loop operation is prohibited in the MELLLA+ operating domain. This prohibition is consistent with Reference 10.
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.
(continued)
BFN-UNIT 2                          B 3.4-5                            Revision 119 Amendment No. 258 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
ACTIONS          A.1 With the requirements of the LCO not met, the recirculation loops must be restored to operation with matched flows within 24 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow 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 the operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence. Note that single loop operation is prohibited in the MELLLA+
operating domain per Reference 10.
(continued)
BFN-UNIT 2                            B 3.4-6                            Revision 119 Amendment No. 258 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES ACTIONS    A.1 (continued)
The 24 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, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.
This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump.
B.1 The Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain is not analyzed for single recirculation loop operation. Therefore, single loop operation is prohibited in the MELLLA+ operating domain (Ref. 10). Action shall be taken to immediately exit the MELLLA+ operating domain in order to return to operation at an analyzed condition.
(continued)
BFN-UNIT 2                  B 3.4-7                            Revision 119 Amendment No. 258 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES ACTIONS      C.1 (continued)
With no recirculation loops in operation while in MODES 1 or 2 or the Required Action and associated Completion Time of Condition A or B not met, 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 MODE 3 within 12 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics.
The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
(continued)
BFN-UNIT 2                    B 3.4-8                            Revision 119 Amendment No. 258 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
REFERENCES        1. FSAR, Section 14.6.3.
: 2. FSAR, Section 4.3.5.
: 3. Deleted.
: 4. Deleted.
: 5. Deleted.
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 7. NEDO-24236, Browns Ferry Nuclear Plant Units 1, 2, and 3, Single-Loop Operation, May 1981.
: 8. NEDC-32484P, Browns Ferry Nuclear Plant Units 1, 2, and 3, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis, Revision 2, December 1997.
: 9. ANP-3015(P), "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis," Revision 0, September 2011.
: 10. NEDC-33006P-A, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report, Revision 3, June 2009.
BFN-UNIT 2                        B 3.4-10                  Amendment No. 258 Revision 87, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, LCO, and        The OPRM Upscale Function provides compliance with APPLICABILITY    10 CFR 50, Appendix A, General Design Criteria (GDC) 10 and (continued)    12, thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations.
Reference 13 describes the Detect and Suppress -
Confirmation Density (DSS-CD) long-term stability solution and the licensing basis Confirmation Density Algorithm (CDA).
Reference 13 also describes the DSS-CD Armed Region and the three additional algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm (PBDA), the amplitude based algorithm (ABA), and the growth rate algorithm (GRA). All four algorithms are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the CDA. The remaining three algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY is based only on the CDA.
The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into cells for evaluation by the OPRM algorithms.
DSS-CD operability requires at least 8 responsive OPRM cells per channel. The DSS-CD software eludes a self-check for the responsive OPRM cells; therefore, no SR is necessary.
The OPRM Upscale Function is required to be OPERABLE above a power level set at 5% of rated power below the lower boundary of the Armed Region, which is ~18% RTP. This requirement is designed to encompass the region of power-flow operation where anticipated events could lead to thermal-hydraulic instability and related neutron flux oscillations. The OPRM Upscale Function is automatically trip-enabled (continued)
BFN-UNIT 3                        B 3.3-15a                              Revision 119 Amendment No. 221 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY    when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is ~23% RTP corresponding to the MCPR monitoring threshold and reactor recirculation drive flow, is less than 75% of rated flow. This region is the OPRM Armed Region. Note (f) allows for entry into the DSS-CD Armed Region without automatic arming of DSS-CD prior to completely passing through the DSS-CD Armed Region during the first startup and the first shutdown following DSS-CD implementation.
If any OPRM auto-enabled setpoint is in a non-conservative condition, i.e., the OPRM Upscale is not auto-enabled with RTP
                ~ 23% and reactor recirculation drive flow :s 75% of rated, the associated channel is considered inoperable for the OPRM Upscale function. Alternatively, the auto-enabled setpoint may be adjusted to place the channel in a conservative condition (armed). If placed in the armed condition, the channel is considered OPERABLE.
Note (f) reflects the need for plant data collection in order to test the DSS-CD equipment. Testing the DSS-CD equipment ensures its proper operation and prevents spurious reactor trips. Entry into the DSS-CD Armed Region without automatic arming of DSS-CD during this initial testing phase also allows for changes in plant operations to address maintenance or other operational needs. However, during this initial testing period, the OPRM Upscale function is OPERABLE and DSS-CD operability and capability to automatically arm shall be maintained at recirculation drive flow rates above the DSS-CD Armed Region flow boundary.
An OPRM Upscale trip is issued from an OPRM channel when the CDA in that channel detects oscillatory changes in the neutron flux, indicated by periodic confirmations and amplitude (continued)
BFN-UNIT 3                        B 3.3-15b                            Revision 119 Amendment No. 221 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY    exceeding specified setpoints for a specified number of OPRM cells in the channel. An OPRM Upscale trip is also issued from the channel if any of the defense-in-depth algorithms (PBDA, ABA, GRA) exceed their trip condition for one or more cells in that channel.
Three of the four channels are required to be OPERABLE.
Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the SLMCPR is exceeded.
There is no Allowable Value for this function.
The OPRM Upscale Function is not LSSS SL-related (Reference 13) and Reference 14 confirms that the OPRM Upscale Function settings based on DSS-CD also do not have traditional instrumentation setpoints determined under an instrument setpoint methodology.
(continued)
BFN-UNIT 3                        B3.3-15c                            Revision 119 Amendment No. 221 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS      D.1 (continued)
Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition.
E.1, F.1, and G.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply.
The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Action E.1 is consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)."
(continued)
BFN-UNIT 3                    B 3.3-34                Amendment No. ~ 224 Revision 0, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS      H.1 (continued)
If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply.
This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
11 If OPRM Upscale trip capability is not maintained, Condition I exists and Backup Stability Protection (BSP) is required. The Manual BSP Regions are described in Reference 13. The Manual BSP Regions are procedurally established consistent with the guidelines identified in Reference 13 and require specified manual operator actions if certain predefined operational conditions occur. The Completion Time of immediate is based on the importance of limiting the period of time during which no automatic or alternate detect and suppress trip capability is in place.
1.2 and 1.3 Required Actions. 1.2 and 1.3 are both required to be taken in conjunction with Required Action 1.1 if OPRM Upscale trip capability is not maintained. As described in Section 7.4 of Reference 13, the Automated BSP Scram Region is designed to avoid reactor instability by automatically preventing entry into the region of the power and flow operating map that is susceptible to reactor instability. The reactor trip would be (continued)
BFN-UNIT 3                      B 3.3-35              Amendment No. ~ . ~
Revision 0, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1
: BASES, ACTIONS      1.2 and 1.3 (continued) initiated by the modified APRM Simulated Thermal Power-High scram setpoints for flow reduction events that would have terminated in the Manual BSP Region I. The Automated BSP Scram Region ensures an early scram and SLMCPR protection.
The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, and based on the importance of restoring an automatic reactor trip for thermal hydraulic instability events. Backup Stability Protection is intended as a temporary means to protect against thermal-hydraulic instability events. The action should be initiated immediately to document the situation and prepare the report. The reporting requirements of Specification 5.6.7 document the corrective actions and schedule to restore the required channels to an OPERABLE status. The Completion Time of 90 days shown in Specification 5.6. 7 is adequate to allow time to evaluate the cause of the inoperability and to determine the appropriate corrective actions and schedule to restore the required channels to OPERABLE status.
If the Required Action I is not completed within the associated Completion Time, then Action J is required. The Bases for the Manual BSP Regions and associated Completion Time is addressed in the Bases for 1.1. The Manual BSP Regions are required in conjunction with the BSP Boundary.
(continued)
BFN-UNIT 3                    B 3.3-35a                Amendment No. ~ . ~
Revision 0, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS      J.2 (continued)
The BSP Boundary, as described in Section 7 .3 of Reference 13, defines an operating domain where potential instability events can be effectively addressed by the specified BSP manual operator actions. The BSP Boundary is constructed such that a flow reduction event initiated from this boundary and terminated at the core natural circulation line (NCL) would not exceed the Manual BSP Region I stability criterion.
Potential instabilities would develop slowly as a result of the feedwater temperature transient (Reference 13).
The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, to reach the specific condition from full power conditions in an orderly manner and without challenging plant systems.
J.3 Backup Stability Protection (BSP) is a temporary means for protection against thermal-hydraulic instability events. An extended period of inoperability without automatic trip capability is not justified. Consequently, the required channels are required to be restored to OPERABLE status within 120 days.
Based on engineering judgment, the likelihood of an instability event that could not be adequately handled by the use of the BSP Regions (See Required Action J.1) and the BSP Boundary (See Required Action J.2) during a 120-day period is negligibly small. The 120-day period is intended to allow for the case where limited design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment. This action is not intended and was not evaluated as a routine alternative to returning failed or inoperable (continued)
BFN-UNIT 3                    B 3.3-35b                Amendment No. ~ . 234 Revision 0, 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS      J.3 (continued)
(continued) equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to normally be accomplished within the completion times allowed for Actions for Conditions A and B.
A Note is provided to indicate that LCO 3.0.4 is not applicable.
The intent of the note is to allow plant startup while operating within the 120 day Completion Time for Required Action J.3.
The primary purpose of this exclusion is to allow an orderly completion of design and verification activities, in the event of a required design change, without undue impact on plant operation.
K.1 If the required channels are not restored to OPERABLE status and the Required Actions of Condition J are not met within the associated Completion Times, then the plant must be placed in an operating condition in which the LCO does not apply. To achieve this status, the plant must be brought to less than 18%
RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the specified operating power level from full power conditions in an orderly manner and without challenging plant systems.
(continued)
BFN-UNIT 3                    B 3.3-35c                Amendment No. ~ . ~
Revision Q., 119 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.17 REQUIREMENTS (continued) (Deleted)
(continued)
BFN-UNIT 3                B 3.3-45a  Revision 4-1-Q, 119 Amendment No. 221 January 8, 2020
 
RPS Instrumentation B 3.3.1.1 BASES (continued)
REFERENCES        1. FSAR, Section 7.2.
: 2. FSAR, Chapter 14.
: 3. NED0-23842, "Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
: 4. FSAR, Appendix N.
: 5. FSAR, Section 14.6.2.
: 6. FSAR, Section 6.5.
: 7. FSAR, Section 14.5.
: 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.
: 9. NEDC-30851-P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.
: 10. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 11. MED-32-0286, "Technical Specification Improvement Analysis for Browns Ferry Nuclear Plant, Unit 2," October 1995.
: 12. NEDC-3241 OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function," October 1995.
: 13. GE Hitachi Nuclear Energy, "GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density," NEDC-33075P-A, Revision 8, November 2013.
: 14. GEH letter to NRC, "NEDC-33075P-A, Detect and Suppress Solution - Confirmation Density (DSS-CD) Analytical Limit (TAC No. MD0277," October 29, 2008. (ADAMS Access No.
ML083040052).
BFN-UNIT 3                          B 3.3-46            Amendment No. 213, 215, 221 Revision 0, 119 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES APPLICABLE      Plant specific LOCA analyses have been performed assuming SAFETY ANALYSES only one operating recirculation loop. These analyses have (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, provided the APLHGR requirements are modified accordingly (Refs. 7 and 8). The Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain has not been analyzed for single recirculation loop operation. Therefore, single loop operation is prohibited in the MELLLA+ operating domain (Ref. 10).
The transient analyses of Chapter 14 of the FSAR have also been performed for single recirculation loop operation (Ref. 7) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoint is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Flow Biased Simulated Thermal Power-High setpoint is in LCO 3.3.1.1, "Reactor Protection System (RPS)
Instrumentation."
Recirculation loops operating satisfies Criterion 2 of the NRG Policy Statement (Ref. 6).
(continued)
BFN-UNIT 3                        B 3.4-4              Amendment No. 216, 221 Revision 0, ~. 119 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
LCO              Two recirculation loops are required to be in operation with their flows 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. With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation. 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 Flow Biased Simulated Thermal Power-High Setpoint (LCO 3.3.1.1) may be applied to allow continued operation consistent with the assumptions of References 7 and 8. Single loop operation is prohibited in the MELLLA+ operating domain. This prohibition is consistent with Reference 10.
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.
(continued)
BFN-UNIT 3                          B 3.4-5    Amendment No. 213,214,216,221 Revision 0, 119 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
ACTIONS          A.1 With the requirements of the LCO not met, the recirculation loops must be restored to operation with matched flows within 24 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow 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 the operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence. Note that single loop operation is prohibited in the MELLLA+
operating domain per Reference 10.
(continued) .
BFN-UNIT 3                            B 3.4-6              Amendment No. 216, 221 Revision 0, 119 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES ACTIONS    A.1 (continued)
The 24 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, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.
This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump.
B.1 The Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain is not analyzed for single recirculation loop operation. Therefore, single loop operation is prohibited in the MELLLA+ operating domain (Ref. 10). Action shall be taken to immediately exit the MELLLA+ operating domain in order to return to operation at an analyzed condition.
(continued)
BFN-UNIT 3                  B 3.4-7              Amendment No. 216, 221 Revision 0, 119 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES ACTIONS      C.1 (continued)
With no recirculation loops in operation while in MODES 1 or 2 or the Required Action and associated Completion Time of Condition A or B not met, 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 MODE 3 within 12 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics.
The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
(continued)
BFN-UNIT 3                    B 3.4-8              Amendment No. 216, 221 Revision Q, 119 January 8, 2020
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
REFERENCES        1. FSAR, Section 14.6.3.
: 2. FSAR, Section 4.3.5.
: 3. Deleted.
: 4. Deleted.
: 5. Deleted.
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 7. NEDO-24236, "Browns Ferry Nuclear Plant Units 1, 2, and 3, Single-Loop Operation," May 1981.
: 8. NEDC-32484P, "Browns Ferry Nuclear Plant Units 1, 2, and 3, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," Revision 2, December 1997.
: 9. ANP-3015(P), "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis," Revision 0, September 2011.
: 10. NEDC-33006P-A, "General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report," Revision 3, June 2009.
BFN-UNIT 3                        B 3.4-10        Amendment No. 213, 216, 221 Revision Q, &7-, 119 January 8, 2020
 
LCO Applicability B 3.0 BASES LCO 3.0.2    There are two basic types of Required Actions. The first type of (continued) Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.) The second type of Required Action specifies the remedial measures that permit continued operation of the unit that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.
Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.
The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the case. An example of this is in LCO 3.4.9, "RCS Pressure and Temperature Limits."
The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally.
The ACTIONS for not meeting a single LCO adequately manage any increase in the plant risk, provided any unusual external conditions (e.g., severe weather, offsite power instability) are considered. In addition, the increased risk associated with simultaneous removal of multiple structures, systems, trains, or components from service is assessed and managed in accordance with 10 CFR 50.65(a)(4).
(continued)
BFN-UNIT 1                        B 3.0-2                          Revision 0, 120 February 3, 2020
 
LCO Applicability B 3.0 BASES LCO 3.0.2    Individual Specifications may specify a time limit for performing an (continued) SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.
When a change in MODE or other specified condition is required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition(s) are entered.
LCO 3.0.3    LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and:
: a. An associated Required Action and Completion Time is not met and no other Condition applies; or
: b. The condition of the unit is not specifically addressed by the associated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit. Sometimes, possible combinations of Conditions are such that entering LCO 3.0.3 is warranted; in such cases, the ACTIONS specifically state a Condition corresponding to such combinations and also that LCO 3.0.3 be entered immediately.
(continued)
BFN-UNIT 1                        B 3.0-3                          Revision 0, 120 February 3, 2020
 
LCO Applicability B 3.0 BASES LCO 3.0.3    This Specification delineates the time limits for placing the unit in a (continued) safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS. Planned entry into LCO 3.0.3 should be avoided. If it is not practical to avoid planned entry into LCO 3.0.3, plant risk should be assessed and managed in accordance with 10 CFR 50.65(a)(4), and the planned entry into LCO 3.0.3 should have less effect on plant safety than other practicable alternatives.
Upon entering LCO 3.0.3, 1 hour is allowed to prepare for an orderly shutdown before initiating a change in unit operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The LCO phrase, Action shall be initiated within 1 hour  does not mean that a change in load must be commenced by the end of the 1 hour period. The action initiated at the end of the 1 hour period may be administrative in nature, such as preparing shutdown procedures. If corrective measures which would allow exiting LCO 3.0.3 are not complete at the end of 1 hour, but there is reasonable assurance that they will be completed with enough time remaining to allow for an orderly unit shutdown, if required, commencing a load decrease may be delayed until that time. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the unit, assuming that only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the Reactor Coolant System and the potential for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 are consistent with the discussion of Section 1.3, Completion Times.
(continued)
BFN-UNIT 1                        B 3.0-4                      Revision 0, 93, 120 February 3, 2020
 
LCO Applicability B 3.0 BASES LCO 3.0.2    There are two basic types of Required Actions. The first type of (continued) Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.) The second type of Required Action specifies the remedial measures that permit continued operation of the unit that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.
Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.
The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the case. An example of this is in LCO 3.4.9, "RCS Pressure and Temperature Limits."
The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally.
The ACTIONS for not meeting a single LCO adequately manage any increase in plant risk, provided any unusual external conditions (e.g., severe weather, offsite power instability) are considered. In addition, the increased risk associated with simultaneous removal of multiple structures, systems, trains, or components from service is assessed and managed in accordance with 10 CFR 50.65(a)(4).
(continued)
BFN-UNIT 3                        B 3.0-2                          Revision 0, 120 February 3, 2020
 
LCO Applicability B 3.0 BASES LCO 3.0.2    Individual Specifications may specify a time limit for performing an (continued) SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.
When a change in MODE or other specified condition is required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition(s) are entered.
LCO 3.0.3    LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and:
: a. An associated Required Action and Completion Time is not met and no other Condition applies; or
: b. The condition of the unit is not specifically addressed by the associated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit. Sometimes, possible combinations of Conditions are such that entering LCO 3.0.3 is warranted; in such cases, the ACTIONS specifically state a Condition corresponding to such combinations and also that LCO 3.0.3 be entered immediately.
(continued)
BFN-UNIT 3                        B 3.0-3                          Revision 0, 120 February 3, 2020
 
LCO Applicability B 3.0 BASES LCO 3.0.3    This Specification delineates the time limits for placing the unit in a (continued) safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS. Planned entry into LCO 3.0.3 should be avoided. If it is not practical to avoid planned entry into LCO 3.0.3, plant risk should be assessed and managed in accordance with 10 CFR 50.65(a)(4), and the planned entry into LCO 3.0.3 should have less effect on plant safety than other practicable alternatives.
Upon entering LCO 3.0.3, 1 hour is allowed to prepare for an orderly shutdown before initiating a change in unit operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The LCO phrase, Action shall be initiated within 1 hour  does not mean that a change in load must be commenced by the end of the 1 hour period. The action initiated at the end of the 1 hour period may be administrative in nature, such as preparing shutdown procedures. If corrective measures which would allow exiting LCO 3.0.3 are not complete at the end of 1 hour, but there is reasonable assurance that they will be completed with enough time remaining to allow for an orderly unit shutdown, if required, commencing a load decrease may be delayed until that time. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the unit, assuming that only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the Reactor Coolant System and the potential for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 are consistent with the discussion of Section 1.3, Completion Times.
(continued)
BFN-UNIT 3                        B 3.0-4                      Revision 0, 93, 120 February 3, 2020
 
Backup Control System B 3.3.3.2 Table B 3.3.3.2-1 (Page 2 of 4)
Backup Control System Instrumentation and Controls NUMBER FUNCTION                                      REQUIRED Transfer/Control Parameter (continued)
: 14. RHRSW Pumps                                                            note e (0-43-23-15, -19, -23, -27) (0-HS-23-15C, -19C, -23C, -27C);
EECW Pumps (north header)
(0-43-23-1, -85, -8, -91) (0-HS-23-1C, -85C, -8C, -91C);
EECW Pumps (south header)
(0-43-23-15, -88, -23, -94) (0-HS-23-15C, -88C, -23C, -94C)
: 15. RHRSW Discharge Valves for RHR Loop II Heat Exchangers                2, note f (1-XS-23-46, -52) (1-HS-23-46C, -52C)
: 16. RCW Pumps 1D and 3D (Trip Function Only)                              2, note g (1-XS-24-16, 3-XS-24-16) (1-HS-24-16C, 3-HS-24-16C)
: 17. Recirculation System Sample Line Isolation Valves                      1, note i (1-XS-43-13, -14) (1-HS-43-13C, -14C)
: 18. (Deleted)
: 19. (Deleted)
: 20. Recirculation Pump 1A Discharge Valve                                      1 (1-XS-68-3) (1-HS-68-3C)
: 21. RWCU Drain to Main Condenser Hotwell Isolation Valve                      1 (1-XS-69-16) (1-HS-69-16C) note e:    There are 12 RHRSW pumps. All are equipped with emergency transfer switches. Backup Control must be available for 2 OPERABLE pumps aligned for EECW service (supports all units).
Backup control for an additional 2 OPERABLE pumps must be available for RHRSW for RHR Loop II.
note f:    1 Discharge Valve per RHR Loop II Heat Exchanger for a total of 2.
note g:    1 per pump. Trip function necessary to prevent spurious start overloading 4-kV Buses/Diesel Generators.
note h:    (Deleted) note i:    1 Recirculation System Sample Line Isolation Valve required, may be either inboard valve or outboard valve.
note j:    (Deleted)
BFN-UNIT 1                                      B 3.3-103      Revision 4, 53, 78, 82, 88, 96, 121 August 12, 2020
 
Backup Control System B 3.3.3.2 Table B 3.3.3.2-1 (Page 2 of 4)
Backup Control System Instrumentation and Controls NUMBER FUNCTION                                      REQUIRED Transfer/Control Parameter (continued)
: 14. RHRSW Pumps                                                          note e (0-43-23-1, -5, -8, -12) (0-HS-23-1C, -5C, -8C, -12C);
EECW Pumps (north header)
(0-43-23-1, -85, -8, -91) (0-HS-23-1C, -85C, -8C, -91C);
EECW Pumps (south header)
(0-43-23-15, -88, -23, -94) (0-HS-23-15C, -88C, -23C, -94C)
: 15. RHRSW Discharge Valves for RHR Loop I Heat Exchangers                2, note f (2-XS-23-34, -40) (2-HS-23-34C, -40C)
: 16. RCW Pumps 1D and 3D (Trip Function Only)                            2, note g (1-XS-24-16, 3-XS-24-16) (1-HS-24-16C, 3-HS-24-16C)
: 17. Recirculation System Sample Line Isolation Valves                    1, note i (2-XS-43-13, -14) (2-HS-43-13C, -14C)
: 18. (Deleted)
: 19. (Deleted)
: 20. Recirculation Pump 2B Discharge Valve                                    1 (2-XS-68-79) (2-HS-68-79C)
: 21. RWCU Drain to Main Condenser Hotwell Isolation Valve                    1 (2-XS-69-16) (2-HS-69-16C) note e:    There are 12 RHRSW pumps. All are equipped with emergency transfer switches. Backup Control must be available for 2 OPERABLE pumps aligned for EECW service (supports all units). Backup control for an additional 2 OPERABLE pumps must be available for RHRSW for RHR Loop I.
note f:    1 Discharge Valve per RHR Loop I Heat Exchanger for a total of 2.
note g:    1 per pump. Trip function necessary to prevent spurious start overloading 4-kV Buses/Diesel Generators.
note h:    (Deleted) note i:    1 Recirculation System Sample Line Isolation Valve required, may be either inboard valve or outboard valve.
Note j:    (Deleted)
BFN-UNIT 2                                    B 3.3-106                    Revision 4, 88, 96, 121 August 12, 2020
 
Backup Control System B 3.3.3.2 Table B 3.3.3.2-1 (Page 2 of 4)
Backup Control System Instrumentation and Controls NUMBER FUNCTION                                      REQUIRED Transfer/Control Parameter (continued)
: 14. RHRSW Pumps                                                            note e (0-43-23-1, -5, -8, -12) (0-HS-23-1C, -5C, -8C, -12C)
EECW Pumps (north header)
(0-43-23-1, -85, -8, -91) (0-HS-23-1C, -85C, -8C, -91C)
EECW Pumps (south header)
(0-43-23-15, -88, -23, -94) (0-HS-23-15C, -88C, -23C, -94C)
: 15. RHRSW Discharge Valves for RHR Loop I Heat Exchangers                2, note f (3-XS-23-34, -40) (3-HS-23-34C, -40C)
: 16. RCW Pumps 1D and 3D (Trip Function Only)                            2, note g (1-XS-24-16, 3-XS-24-16) (1-HS-24-16C, 3-HS-24-16C)
: 17. Recirculation System Sample Line Isolation Valves                    1, note i (3-XS-43-13, -14) (3-HS-43-13C, 14C)
: 18. (Deleted)
: 19. (Deleted)
: 20. Recirculation Pump 3B Discharge Valve                                    1 (3-XS-68-79) (3-HS-68-79C)
: 21. RWCU Drain to Main Condenser Hotwell Isolation Valve                      1 (3-XS-69-16) (3-HS-69-16C) note e:    There are 12 RHRSW pumps. All are equipped with emergency transfer switches. Backup Control must be available for 2 OPERABLE pumps aligned for EECW service (supports all units). Backup control for an additional 2 OPERABLE pumps must be available for RHRSW for RHR Loop I.
note f:    1 Discharge Valve per RHR Loop I Heat Exchanger for a total of 2.
note g:    1 per pump. Trip function necessary to prevent spurious start overloading 4-kV Buses/Diesel Generators.
note h:    (Deleted) note i:    1 Recirculation System Sample Line Isolation Valve required, may be either inboard valve or outboard valve.
note j:    (Deleted)
BFN-UNIT 3                                    B 3.3-106                  Revision 4, 88, 96, 121 August 12, 2020
 
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE      3, 4. Reactor Zone Exhaust and Refueling Floor Radiation -
SAFETY ANALYSES, High (RM-90-140, 141, 142, 143) (continued)
LCO, and APPLICABILITY    ventilation exhaust both of which must be OPERABLE or tripped for the channel to be OPERABLE. Both radiation elements must provide a High signal to trip the associated channel (two-out-of-two). However, the output relays from the divisional trip systems are arranged in logic systems such that if either channel for a zone trips, a secondary containment isolation signal is initiated (one-out-of-two). Two channels of Reactor Zone Exhaust Radiation - High Function and two channels of Refueling Floor 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 provide timely detection of nuclear system process barrier leaks inside containment but are far enough above background levels to avoid spurious isolation.
The Reactor Zone Exhaust and Refueling Floor Radiation -
High Functions are 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, these Functions are not required. In addition, the Functions are also required to be OPERABLE during OPDRVs because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery) must be provided to ensure that offsite dose limits are not exceeded.
(continued)
BFN-UNIT 1                        B 3.3-229            Revision 0, 21, 29, 35, 122 December 12, 2020
 
CREV System Instrumentation B 3.3.7.1 BASES APPLICABLE      3., 4. Reactor Zone Exhaust and Refueling Floor SAFETY ANALYSES, Radiation - High (RM-90-140, 141, 142, 143)
LCO, and APPLICABILITY    High secondary containment exhaust radiation is an indication (continued)    of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the RCPB. A reactor zone or refueling floor exhaust high radiation signal will automatically initiate the CREV System, since this radiation release could result in radiation exposure to control room personnel.
The reactor zone and refueling floor exhaust radiation monitors provide two independent channels for each ventilation exhaust path coming from the reactor zones and the refueling zone.
There are two radiation monitors (each monitor provides one channel of each Function) and two divisional trip systems for each unit (Units 1, 2, and 3). Two channels of each function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude CREV System initiation.
The Allowable Value was selected to ensure that the Function will promptly detect high activity that could threaten exposure to control room personnel.
The Reactor Zone Exhaust and Refueling Floor Radiation -
High Functions are required to be OPERABLE in MODES 1, 2, and 3 and during operations with a potential for draining the reactor vessel (OPDRVs), to ensure that control room personnel are protected during a LOCA 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 or fuel damage is low; thus, the Function is not required.
(continued)
BFN-UNIT 1                        B 3.3-242                    Revision 0, 29, 35, 122 December 21, 2020
 
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE      3, 4. Reactor Zone Exhaust and Refueling Floor Radiation -
SAFETY ANALYSES, High (RM-90-140, 141, 142, 143) (continued)
LCO, and APPLICABILITY    ventilation exhaust both of which must be OPERABLE or tripped for the channel to be OPERABLE. Both radiation elements must provide a High signal to trip the associated channel (two-out-of-two). However, the output relays from the divisional trip systems are arranged in logic systems such that if either channel for a zone trips, a secondary containment isolation signal is initiated (one-out-of-two). Two channels of Reactor Zone Exhaust Radiation - High Function and two channels of Refueling Floor 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 provide timely detection of nuclear system process barrier leaks inside containment but are far enough above background levels to avoid spurious isolation.
The Reactor Zone Exhaust and Refueling Floor Radiation -
High Functions are 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, these Functions are not required. In addition, the Functions are also required to be OPERABLE during OPDRVs because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery) must be provided to ensure that offsite dose limits are not exceeded.
(continued)
BFN-UNIT 2                        B 3.3-232            Revision 0, 21, 29, 35, 122 December 21, 2020
 
CREV System Instrumentation B 3.3.7.1 BASES APPLICABLE      3., 4. Reactor Zone Exhaust and Refueling Floor SAFETY ANALYSES, Radiation - High (RM-90-140, 141, 142, 143)
LCO, and APPLICABILITY    High secondary containment exhaust radiation is an indication (continued)    of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the RCPB. A reactor zone or refueling floor exhaust high radiation signal will automatically initiate the CREV System, since this radiation release could result in radiation exposure to control room personnel.
The reactor zone and refueling floor exhaust radiation monitors provide two independent channels for each ventilation exhaust path coming from the reactor zones and the refueling zone.
There are two radiation monitors (each monitor provides one channel of each Function) and two divisional trip systems for each unit (Units 1, 2, and 3). Two channels of each function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude CREV System initiation.
The Allowable Value was selected to ensure that the Function will promptly detect high activity that could threaten exposure to control room personnel.
The Reactor Zone Exhaust and Refueling Floor Radiation -
High Functions are required to be OPERABLE in MODES 1, 2, and 3 and during operations with a potential for draining the reactor vessel (OPDRVs), to ensure that control room personnel are protected during a LOCA 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 or fuel damage is low; thus, the Function is not required.
(continued)
BFN-UNIT 2                        B 3.3-245                    Revision 0, 29, 35, 122 December 21, 2020
 
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE      3, 4. Reactor Zone Exhaust and Refueling Floor Radiation -
SAFETY ANALYSES, High (RM-90-140, 141, 142, 143) (continued)
LCO, and APPLICABILITY    ventilation exhaust both of which must be OPERABLE or tripped for the channel to be OPERABLE. Both radiation elements must provide a High signal to trip the associated channel (two-out-of-two). However, the output relays from the divisional trip systems are arranged in logic systems such that if either channel for a zone trips, a secondary containment isolation signal is initiated (one-out-of-two). Two channels of Reactor Zone Exhaust Radiation - High Function and two channels of Refueling Floor 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 provide timely detection of nuclear system process barrier leaks inside containment but are far enough above background levels to avoid spurious isolation.
The Reactor Zone Exhaust and Refueling Floor Radiation -
High Functions are 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, these Functions are not required. In addition, the Functions are also required to be OPERABLE during OPDRVs because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery) must be provided to ensure that offsite dose limits are not exceeded.
(continued)
BFN-UNIT 3                        B 3.3-232                Revision 21, 29, 35, 122 Amendment No. 213 December 21, 2020
 
CREV System Instrumentation B 3.3.7.1 BASES APPLICABLE      3., 4. Reactor Zone Exhaust and Refueling Floor SAFETY ANALYSES, Radiation - High (RM-90-140, 141, 142, 143)
LCO, and APPLICABILITY    High secondary containment exhaust radiation is an indication (continued)    of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the RCPB. A reactor zone or refueling floor exhaust high radiation signal will automatically initiate the CREV System, since this radiation release could result in radiation exposure to control room personnel.
The reactor zone and refueling floor exhaust radiation monitors provide two independent channels for each ventilation exhaust path coming from the reactor zones and the refueling zone.
There are two radiation monitors (each monitor provides one channel of each Function) and two divisional trip systems for each unit (Units 1, 2, and 3). Two channels of each function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude CREV System initiation.
The Allowable Value was selected to ensure that the Function will promptly detect high activity that could threaten exposure to control room personnel.
The Reactor Zone Exhaust and Refueling Floor Radiation -
High Functions are required to be OPERABLE in MODES 1, 2, and 3 and during operations with a potential for draining the reactor vessel (OPDRVs), to ensure that control room personnel are protected during a LOCA 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 or fuel damage is low; thus, the Function is not required.
(continued)
BFN-UNIT 3                        B 3.3-245                      Revision 29, 35, 122 Amendment No. 213 December 21, 2020
 
Control Rod OPERABILITY B 3.1.3 BASES (continued)
SURVEILLANCE      SR 3.1.3.1 REQUIREMENTS The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.3.2 (Deleted).
SR 3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. This surveillance is not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of banked position withdrawal sequence (BPWS) (LCO 3.1.6) and the RWM (LCO 3.3.2.1). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                        B 3.1-23                          Revision 0, 123 Amendment 274 March 22, 2020
 
Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE SR 3.1.4.2 REQUIREMENTS (continued) Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods. This sample remains representative if no more than 7.5% of the control rods in the sample tested are determined to be "slow." With more than 7.5% of the sample declared to be "slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 7.5% criterion (i.e., 7.5% of the entire sample) is satisfied, or until the total number of "slow" control rods (throughout the core from all Surveillances) exceeds the LCO limit. For planned testing, the control rods selected for the sample should be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.1-31                      Revision 0, 9, 94, 123 Amendment No. 239 March 22, 2020
 
Control Rod Scram Accumulators B 3.1.5 BASES (continued)
SURVEILLANCE      SR 3.1.5.1 REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be checked periodically to ensure adequate accumulator pressure exists to provide sufficient scram force. An automatic accumulator monitor may be used to continuously satisfy this requirement.
The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940 psig is well below the expected pressure of 1100 psig (Ref. 1). Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant degradation in scram times does not occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 3.4.6.
: 2. FSAR, Section 14.5.
: 3. FSAR, Section 14.6.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                        B 3.1-40                        Revision 0, 123 March 22, 2020
 
Rod Pattern Control B 3.1.6 BASES ACTIONS      B.1 and B.2 (continued)
LCO 3.3.2.1 requires verification of control rod movement by a second licensed operator or a qualified member of the technical staff.
When nine or more OPERABLE control rods are not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within 1 hour. With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.
SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is periodically verified to be in compliance with the BPWS to ensure the assumptions of the CRDA analyses are met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The RWM provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at d 10% RTP.
(continued)
BFN-UNIT 1                    B 3.1-45                            Revision 0, 123 March 22, 2020
 
SLC System B 3.1.7 BASES ACTIONS      B.1 (continued)
If both SLC subsystems are inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours. The allowed Completion Time of 8 hours is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor.
C.1 and C.2 If any Required Action and associated Completion Time is not met, 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.
SURVEILLANCE SR 3.1.7.1 REQUIREMENTS SR 3.1.7.1 verifies the volume of the borated solution in the storage tank, thereby ensuring SLC System OPERABILITY without disturbing normal plant operation. This Surveillance ensures that the proper borated solution volume is maintained for reactivity control and post-LOCA suppression pool pH control. The tank volume requirement of 4000 gallons is established by the amount of boron at 8.0% by weight concentration required for the radiological dose analysis for post-LOCA suppression pool pH control. The tank volume requirement for reactivity control is encompassed by the requirement for post LOCA pH control. For reactivity control, the sodium (continued)
BFN-UNIT 1                    B 3.1-51                Revision 0, 29, 112, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.1 (continued)
REQUIREMENTS pentaborate solution concentration requirements (d 9.2% by weight) and the required quantity of Boron-10 (t 203 lbs) establish the tank volume requirement. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.7.2 SR 3.1.7.2 verifies the continuity of the explosive charges in the injection valves to ensure that proper operation will occur if required. An automatic continuity monitor may be used to continuously satisfy this requirement. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.7.3 SR 3.1.7.3 requires an examination of sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank for post-LOCA suppression pool pH control. This parameter is used as input to determine the volume requirements for SR 3.1.7.1. The concentration is dependent upon the volume of water and quantity of boron in the storage tank.
SR 3.1.7.3 must be performed according to the Surveillance Frequency Control Program or within 24 hours of when boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits.
(continued)
BFN-UNIT 1                    B 3.1-52                  Revision 0, 29, 50, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.4 REQUIREMENTS (continued) SR 3.1.7.4 requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank. The concentration is dependent upon the volume of water and quantity of boron in the storage tank.
The sodium pentaborate solution (SPB) concentration is allowed to be > 9.2 weight percent provided the concentration and temperature of the sodium pentaborate solution are verified to be within the limits of Figure 3.1.7-1. This ensures that unwanted precipitation of the sodium pentaborate does not occur.
SR 3.1.7.4 must be performed according to the Surveillance Frequency Control Program or within 24 hours of when boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits.
SR 3.1.7.4 must be performed within 8 hours of discovery that the concentration is > 9.2 weight percent and every 12 hours thereafter until the concentration is verified to be d 9.2 weight percent. This Frequency is appropriate under these conditions taking into consideration the SLC System design capability still exists for vessel injection under these conditions and the low probability of the temperature and concentration limits of Figure 3.1.7-1 not being met.
(continued)
BFN-UNIT 1                    B 3.1-53          Revision 0, 29, 112, 116, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.5 REQUIREMENTS (continued) This Surveillance requires the amount of Boron-10 in the SLC solution tank to be determined periodically. The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water. Since the chemicals used have known Boron-10 quantities, the Boron-10 quantity in the sodium pentaborate solution formed can be calculated. This parameter is used as input to determine the volume requirements for reactivity control encompassed by SR 3.1.7.1. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.7.6 SR 3.1.7.6 requires verification that the SLC system conditions satisfy the following equation:
( C )( Q )(        E    )
                                                        > = 1.0
( 8.7 WT % )( 50 GPM )( 94 ATOM % )
C = sodium pentaborate solution weight percent concentration Q = SLC system pump flow rate in gpm E = Boron-10 atom percent enrichment in the sodium pentaborate solution To meet 10 CFR 50.62, the SLC System must have a minimum flow capacity and boron content equivalent in control capacity to 86 gpm of 13 weight percent natural sodium pentaborate solution. The purpose of this injection rate is to ensure that during an ATWS condition with the MSIVs closed, sufficient B-10 is injected into the RPV to bring the reactor subcritical (Hot Shutdown) prior to suppression pool temperature exceeding its heat capacity temperature limit. The atom percentage of natural B-10 is 19.8%. This equivalency requirement is met when the equation given above is satisfied.
(continued)
BFN-UNIT 1                    B 3.1-54                Revision 0, 29, 116, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.6 REQUIREMENTS (continued) The equation can be satisfied by adjusting the solution concentration, pump flow rate or Boron-10 enrichment. If the results of the equation are < 1, the SLC System is no longer capable of shutting down the reactor with the margin described in Reference 2. As described in Reference 2, the BFN analysis assumes a flow capacity and boron content equivalent to 50 gpm of 8.7 weight percent and 94 atom percent B-10 enriched sodium pentaborate solution. This exceeds the requirement of 10 CFR 50.62, and the equation is adjusted to reflect the BFN requirements. The quantity of stored boron includes an additional margin (25%) beyond the amount needed to shut down the reactor to allow for possible imperfect mixing of the chemical solution in the reactor water, leakage, and the volume in other piping connected to the reactor system.
SR 3.1.7.6 must be performed according to the Surveillance Frequency Control Program or within 24 hours of when boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specific limits.
(continued)
BFN-UNIT 1                  B 3.1-54a                Revision 0, 29, 116, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.7 REQUIREMENTS (continued) Demonstrating that each SLC System pump develops a flow rate t 39 gpm at a discharge pressure t 1325 psig ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration and enrichment requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The pump flow rate of 39 gpm is based on the original licensing bases for SLC for an alternate reactivity insertion system.
SR 3.1.7.8 and SR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. Additionally, replacement charges shall be selected such that the age of charge in service shall not exceed five years from the manufacturer's assembly date.
(continued)
BFN-UNIT 1                    B 3.1-55        Revision 0, 29, 43, 50, 116, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.8 and SR 3.1.7.9 (continued)
REQUIREMENTS The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Demonstrating that all piping between the boron solution storage tank and the suction inlet to the injection pumps is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution. An acceptable method for verifying that the suction piping is unblocked is to pump from the storage tank to the storage tank. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.7.10 The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water. Isotopic tests on these chemicals to verify the actual B-10 enrichment must be performed according to the Surveillance Frequency Control Program and after addition of boron to the SLC tank in order to ensure that the proper B-10 atom percentage is being used and SR 3.1.7.6 will be met. The sodium pentaborate enrichment must be calculated within 24 hours and verified by analysis within 30 days.
(continued)
BFN-UNIT 1                    B 3.1-56                  Revision 0, 29, 43, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.11 REQUIREMENTS (continued) SR 3.1.7.11 verifies that each valve in the system is in its correct position, but does not apply to the squib (i.e., explosive) valves. Verifying the correct alignment for manual, power operated, and automatic valves in the SLC System Flowpath provides assurance that the proper flow paths will exist for system operation. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position from the control room, or locally by a dedicated operator at the valve control. This is acceptable since the SLC System is a manually initiated system. This surveillance also does not apply to valves that are locked, sealed, or otherwise secured in position since they are verified to be in the correct position prior to locking, sealing or securing. This verification of valve alignment does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. 10 CFR 50.62.
: 2. NEDC-33860P, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate, Section 2.8.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 4. FSAR, Section 14.6.
BFN-UNIT 1                    B 3.1-57                  Revision 0, 29, 112, 123 March 22, 2020
 
SDV Vent and Drain Valves B 3.1.8 BASES (continued)
SURVEILLANCE      SR 3.1.8.1 REQUIREMENTS During normal operation, the SDV vent and drain valves should be in the open position (except when performing SR 3.1.8.2) to allow for drainage of the SDV piping. Verifying that each valve is in the open position ensures that the SDV vent and drain valves will perform their intended functions during normal operation. This SR does not require any testing or valve manipulation; rather, it involves verification that the valves are in the correct position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.8.2 During a scram, the SDV vent and drain valves should close to contain the reactor water discharged to the SDV piping. Cycling each valve through its complete range of motion (closed and open) ensures that the valve will function properly during a scram. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.8.3 SR 3.1.8.3 is an integrated test of the SDV vent and drain valves to verify total system performance. After receipt of a simulated or actual scram signal, the closure of the SDV vent and drain valves is verified. The closure time of 60 seconds after receipt of a scram signal is acceptable based on the (continued)
BFN-UNIT 1                          B 3.1-62                      Revision 0, 29, 123 March 22, 2020
 
SDV Vent and Drain Valves B 3.1.8 BASES SURVEILLANCE SR 3.1.8.3 (continued)
REQUIREMENTS bounding analysis for release of reactor coolant outside containment (Ref. 2). Similarly, after receipt of a simulated or actual scram reset signal, the opening of the SDV vent and drain valves is verified. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1 and the scram time testing of control rods in LCO 3.1.3 overlap this Surveillance to provide complete testing of the assumed safety function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 3.4.5.3.1.
: 2. FSAR, Section 14.6.5.
: 3. 10 CFR 50.67.
: 4. FSAR, Section 6.5.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                    B 3.1-63                  Revision 0, 29, 43, 123 March 22, 2020
 
APLHGR B 3.2.1 BASES (continued)
ACTIONS          A.1 If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA and transient analyses may not be met. Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 2 hour Completion Time is sufficient to restore the APLHGR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification.
B.1 If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 23% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 23% RTP in an orderly manner and without challenging plant systems.
SURVEILLANCE      SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 12 hours after THERMAL POWER is t 23% RTP and periodically thereafter. They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER t 23% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                          B 3.2-4                    Revision 0, 113, 123 March 22, 2020
 
MCPR B 3.2.2 BASES (continued)
SURVEILLANCE      SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours after THERMAL POWER is t 23% RTP and then periodically thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER t 23% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.2.2.2 Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis. SR 3.2.2.2 determines the actual scram speed distribution and compares it with the assumed distribution. The MCPR operating limit is determined based either on the applicable limit associated with scram times of LCO 3.1.4, Control Rod Scram Times, or the nominal scram times. The scram speed-dependent MCPR limits are contained in the COLR. This determination must be performed within 72 hours after each set of control rod scram time tests required by SR 3.1.4.1 and SR 3.1.4.2 because the effective scram speed distribution may change during the cycle. The 72-hour Completion Time is acceptable due to the relatively minor changes in the actual control rod scram speed distribution expected during the fuel cycle.
(continued)
BFN-UNIT 2                          B 3.2-10                Revision 0, 31, 113, 123 March 22, 2020
 
LHGR B 3.2.3 BASES (continued)
SURVEILLANCE      SR 3.2.3.1 REQUIREMENTS The LHGR is required to be initially calculated within 12 hours after THERMAL POWER is t 23% RTP and then periodically thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER t 23% RTP is achieved is acceptable given the large inherent margin to operating limits at lower power levels. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Chapter 14.
: 2. FSAR, Chapter 3.
: 3. NUREG-0800, Standard Review Plan 4.2, Section II.A.2(g),
Revision 2, July 1981.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                          B 3.2-15                    Revision 0, 113, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      11. Manual Scram (continued)
SAFETY ANALYSES, LCO, and        Two channels of Manual Scram with one channel in each APPLICABILITY    manual scram trip system are available and required to be OPERABLE in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
: 12. RPS Channel Test Switches There are four RPS Channel Test Switches, one associated with each of the four automatic scram logic channels (A1, A2, B1, and B2). These keylock switches allow the operator to test the OPERABILITY of each individual logic channel without the necessity of using a scram function trip. When the RPS Channel Test Switch is placed in test, the associated scram logic channel is deenergized and OPERABILITY of the channel's scram contactors can be confirmed. The RPS Channel Test Switches are not specifically credited in the accident analysis. However, because the Manual Scram Function at Browns Ferry Nuclear Plant is not configured the same as the generic model in Reference 9, the RPS Channel Test Switches are included in the analysis in Reference 11.
(continued)
BFN-UNIT 1                        B 3.3-28                        Revision 0, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.1 REQUIREMENTS (continued) Performance of the CHANNEL CHECK 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 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
(continued)
BFN-UNIT 1                    B 3.3-35                            Revision 0, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.2 REQUIREMENTS (continued) To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
A restriction to satisfying this SR when < 23% RTP is provided that requires the SR to be met only at t 23% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when
            < 23% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At t 23% RTP, the Surveillance is required to have been satisfactorily performed, in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 23% if the Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 23% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
(continued)
BFN-UNIT 1                    B 3.3-36                Revision 0, 40, 112, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.3 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.
Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
As noted, SR 3.3.1.1.3 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.3-37                      Revision 0, 40, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.5 and SR 3.3.1.1.6 (continued)
REQUIREMENTS If overlap for a group of channels is not demonstrated (e.g.,
IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel(s) declared inoperable. Only those appropriate channels that are required in the current MODE or condition should be declared inoperable.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.1.7 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.1.8, SR 3.3.1.1.12 (Deleted)
(continued)
BFN-UNIT 1                    B 3.3-39                    Revision 0, 40, 43, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.9, SR 3.3.1.1.10 and SR 3.3.1.1.13 (continued)
REQUIREMENTS The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.
SR 3.3.1.1.11 (Deleted).
SR 3.3.1.1.14 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel.
The functional testing of control rods (LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.
The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM trip and OPRM conditions at the 2-out-of-4 voter channel inputs to check all combinations of two tripped inputs to the 2-out-of-4 logic in the voter channels and APRM related redundant RPS relays.
(continued)
BFN-UNIT 1                    B 3.3-41                Revision 0, 40, 43, 45, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 REQUIREMENTS (continued) This SR ensures that scrams initiated from the Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is t 26% RTP. This involves calibration of the bypass channels (PIS-1-81A, PIS-1-81B, PIS-1-91A, and PIS-1-91B). Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.
If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at t 26% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are enabled), this SR is met and the channel is considered OPERABLE.
The Frequency of 24 months 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 3.3.1.1.16 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
(continued)
BFN-UNIT 1                    B 3.3-42                    Revision 0, 40, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.16 (continued)
REQUIREMENTS The SR 3.3.1.1.16 Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing - applicable to Function 2.b, only), the 2-out-of-4 voter channels, and the interface connections into the RPS trip systems from the voter channels.
Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
(NOTE: The actual voting logic of the 2-out-of-4 Voter Function is tested as part of SR 3.3.1.1.14.) A Note for SR 3.3.1.1.16 is provided that requires the APRM Function 2.a SR to be performed within 12 hours of entering MODE 2 from MODE 1.
Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. This Note allows entry into MODE 2 from MODE 1 if the associated frequency is not met per SR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
(continued)
BFN-UNIT 1                    B 3.3-43                Revision 0, 43, 112, 123 March 22, 2020
 
SRM Instrumentation B 3.3.1.2 BASES ACTIONS      E.1 and E.2 (continued)
With one or more required SRM inoperable in MODE 5, the ability to detect local reactivity changes in the core during refueling is degraded. CORE ALTERATIONS must be immediately suspended and action must be immediately initiated to insert all insertable control rods in core cells containing one or more fuel assemblies. Suspending CORE ALTERATIONS prevents the two most probable causes of reactivity changes, fuel loading and control rod withdrawal, from occurring. Inserting all insertable control rods ensures that the reactor will be at its minimum reactivity given that fuel is present in the core. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative position.
Action (once required to be initiated) to insert control rods must continue until all insertable rods in core cells containing one or more fuel assemblies are inserted.
SURVEILLANCE As noted at the beginning of the SRs, the SRs for each SRM REQUIREMENTS Applicable MODE or other specified conditions are found in the SRs column of Table 3.3.1.2-1.
SR 3.3.1.2.1 Performance of the CHANNEL CHECK 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 another channel. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
(continued)
BFN-UNIT 1                      B 3.3-51                          Revision 0, 123 March 22, 2020
 
SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.1 (continued)
REQUIREMENTS 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
SR 3.3.1.2.2 To provide adequate coverage of potential reactivity changes in the core when the fueled region encompasses more than one SRM, one SRM is required to be OPERABLE in the quadrant where CORE ALTERATIONS are being performed, and the other OPERABLE SRM must be in an adjacent quadrant (continued)
BFN-UNIT 1                    B 3.3-52                            Revision 0, 123 March 22, 2020
 
Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS    C.1, C.2.1.1, C.2.1.2, and C.2.2 (continued) accordance with the restrictions imposed by Required Action C.2.2. Required Action C.2.2 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff (e.g.,
a qualified shift technical advisor or reactor engineer).
The RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LCO 3.1.3 and LCO 3.1.6 may require bypassing the RWM, during which time the RWM must be considered inoperable with Condition C entered and its Required Actions taken.
D.1 With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence. Required Action D.1 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff. The RWM may be bypassed under these conditions to allow the reactor shutdown to continue.
(continued)
BFN-UNIT 1                    B 3.3-65                              Revision 0
 
Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS      E.1 and E.2 (continued)
With one Reactor Mode Switch - Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function. However, since the Required Actions are consistent with the normal action of an OPERABLE Reactor Mode Switch - Shutdown Position Function (i.e., maintaining all control rods inserted), there is no distinction between having one or two channels inoperable.
In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies will ensure that the core is subcritical with adequate SDM ensured by LCO 3.1.1. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
SURVEILLANCE As noted at the beginning of the SRs, the SRs for each Control REQUIREMENTS Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1.
The Surveillances are modified by a second Note (Note 2) to indicate that when an RBM 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 control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable (continued)
BFN-UNIT 1                    B 3.3-66                                Revision 0
 
Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)
REFERENCES        1. FSAR, Section 7.5.8.2.3.
: 2. FSAR, Section 7.16.5.3.1.k.
: 3. NEDC-32433P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Unit 1, 2 and 3," April 1995.
: 4. NEDE-24011-P-A-US, "General Electrical Standard Application for Reload Fuel," Supplement for United States, (revision specified in the COLR).
: 5. "Modifications to the Requirements for Control Rod Drop Accident Mitigating Systems," BWR Owners' Group, July 1986.
: 6. NEDO-21231, "Banked Position Withdrawal Sequence,"
January 1977.
: 7. NRC SER, "Acceptance of Referencing of Licensing Topical Report NEDE-24011-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17,"
December 27, 1987.
: 8. NEDC-30851-P-A, Supplement 1, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," October 1988.
: 9. GENE-770-06-1, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.
: 10. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 11. Deleted.
(continued)
BFN-UNIT 1                        B 3.3-71                  Revision 0, 40, 68, 123 Amendment 276 March 22, 2020
 
Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES  12. NEDO 33091-A, Revision 2, Improved BPWS Control Rod (continued)    Insertion Process, July 2004.
: 13. XN-NF-80-19(P)(A), Volume 1, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, (as identified in the COLR).
: 14. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, (as identified in the COLR).
(continued)
BFN-UNIT 1                    B 3.3-71a                      Revision 0, 40, 68 Amendment 276 October 18, 2012
 
PAM Instrumentation B 3.3.3.1 BASES LCO          3. Suppression Pool Water Level (continued) (LI-64-159A and XR-64-159)
Suppression pool water level is a Category 1 variable provided to detect a breach in the reactor coolant pressure boundary (RCPB). This variable is also used to verify and provide long term surveillance of ECCS function. The wide range suppression pool water level measurement provides the operator with sufficient information to assess the status of both the RCPB and the water supply to the ECCS. The wide range water level indicators monitor the suppression pool water level from two feet from the bottom of the pool to five feet above normal water level. Two wide range suppression pool water level signals are transmitted from separate differential pressure transmitters and are continuously recorded and displayed on one recorder and one indicator in the control room. The recorder and indicator are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
: 4. Drywell Pressure (PI-64-67B, XR-64-50, PI-64-160A, and XR-64-159)
Drywell pressure is a Category 1 variable provided to detect breach of the RCPB and to verify ECCS functions that operate to maintain RCS integrity. Two different ranges of drywell pressure channels (normal and wide range) receive signals that are transmitted from separate pressure transmitters and are continuously recorded and displayed on two control room recorders and two control room indicators. These recorders and indicators are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
(continued)
BFN-UNIT 1                    B 3.3-85                              Revision 0
 
PAM Instrumentation B 3.3.3.1 BASES LCO          5. Primary Containment Area Radiation (High Range)
(continued) (RR-90-272 and RR-90-273)
Primary containment area radiation (high range) is provided to monitor the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans. Two high range primary containment area radiation signals (RM-90-272A and RM-90-273A) are transmitted from separate radiation detectors and are continuously recorded and displayed on two control room recorders. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
: 6. Primary Containment Isolation Valve (PCIV) Position PCIV position is provided for verification of containment integrity. In the case of PCIV position, the important information is the isolation status of the containment penetration. The LCO requires one channel of valve position indication in the control room to be OPERABLE for each active PCIV in a containment penetration flow path, i.e., two total channels of PCIV position indication for a penetration flow path with two active valves. For containment penetrations with only one active PCIV having control room indication, Note (b) requires a single channel of valve position indication to be OPERABLE. This is sufficient to redundantly verify the isolation status of each isolable penetration via indicated status of the active valve, as applicable, and prior knowledge of passive valve or system boundary status. If a penetration flow path is isolated, position indication for the PCIV(s) in the associated penetration flow path is not needed to determine status. Therefore, the position indication for valves in an isolated penetration flow path is not required to be OPERABLE.
(continued)
BFN-UNIT 1                    B 3.3-86                            Revision 0, 53 May 18, 2007
 
PAM Instrumentation B 3.3.3.1 BASES LCO        6. Primary Containment Isolation Valve (PCIV) Position (continued)
The PCIV position PAM indication instrumentation consists of the category 1 PCIV position indications identified in Reference
: 4. The indication for each PCIV consists of green and red indicator lights that illuminate to indicate whether the PCIV is fully open, fully closed, or in a mid-position. Therefore, the PAM specification deals specifically with this portion of the instrument channel.
: 7. (Deleted)
(continued)
BFN-UNIT 1                  B 3.3-87                            Revision 0, 32 April 6, 2005
 
Backup Control System B 3.3.3.2 B 3.3 INSTRUMENTATION B 3.3.3.2 Backup Control System BASES BACKGROUND          The Backup Control System provides the control room operator with sufficient instrumentation and controls to place and maintain the plant in a safe shutdown condition from a location other than the control room. This capability is necessary to protect against the possibility of the control room becoming inaccessible. A safe shutdown condition is defined as MODE 3.
With the plant in MODE 3, the Reactor Core Isolation Cooling (RCIC) System, the safety/relief valves, and the Residual Heat Removal System can be used to remove core decay heat and meet all safety requirements. The long term supply of water for the RCIC and the ability to operate the RHR System for decay heat removal from outside the control room allow extended operation in MODE 3.
In the event that the control room becomes inaccessible, the operators can establish control at the backup control panel and place and maintain the plant in MODE 3. Not all controls and necessary transfer switches are located at the backup control panel. Some controls and transfer switches will have to be operated locally at the switchgear, motor control panels, or other local stations. The plant automatically reaches MODE 3 following a plant shutdown and can be maintained safely in MODE 3 for an extended period of time.
The OPERABILITY of the Backup Control System control and instrumentation Functions ensures that there is sufficient information available on selected plant parameters to place and maintain the plant in MODE 3 should the control room become inaccessible.
(continued)
BFN-UNIT 1                            B 3.3-96                              Revision 0
 
Backup Control System B 3.3.3.2 BASES (continued)
APPLICABLE        The Backup Control System is required to provide equipment SAFETY ANALYSES  at appropriate locations outside the control room with a design capability to promptly shut down the reactor to MODE 3, including the necessary instrumentation and controls, to maintain the plant in a safe condition in MODE 3.
The criteria governing the design and the specific system requirements of the Backup Control System are located in 10 CFR 50, Appendix A, GDC 19 (Ref. 1) and Reference 2.
The Backup Control System is considered an important contributor to reducing the risk of accidents; as such, it meets Criterion 4 of the NRC Policy Statement (Ref. 3).
LCO              The Backup Control System LCO provides the requirements for the OPERABILITY of the instrumentation and controls necessary to place and maintain the plant in MODE 3 from a location other than the control room. The instrumentation and controls typically required are listed in Table B 3.3.3.2-1.
The controls, instrumentation, and transfer switches are those required for:
x    Reactor pressure vessel (RPV) pressure control; x    Decay heat removal; x    RPV inventory control; and x    Safety support systems for the above functions, including Residual Heat Removal (RHR) Service Water, Emergency Equipment Cooling Water, and onsite power, including the diesel generators.
(continued)
BFN-UNIT 1                          B 3.3-97                                Revision 0
 
ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE      2.e. Reactor Vessel Water Level - Level 0 SAFETY ANALYSES, (LIS-3-52 and 62A) (continued)
LCO, and APPLICABILITY    reactor water level is below Level 0. Reactor Vessel Water Level - Level 0 signals are initiated from two 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. The Reactor Vessel Water Level - Level 0 Allowable Value is chosen to allow the low pressure core flooding systems to activate and provide adequate cooling before allowing a manual transfer.
Two channels of the Reactor Vessel Water Level - Level 0 Function are only required to be OPERABLE in MODES 1, 2, and 3. In MODES 4 and 5, the specified initiation time of the LPCI subsystems is not assumed, and other administrative controls are adequate to control the valves that this Function isolates (since the systems that the valves are opened for are not required to be OPERABLE in MODES 4 and 5 and are normally not used).
HPCI System 3.a. Reactor Vessel Water Level - Low Low, Level 2 (LIS-3-58A-D)
Low 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. Therefore, the HPCI System is initiated at Level 2 to maintain level above the top of the active fuel. The Reactor Vessel Water Level - Low Low, Level 2 is (continued)
BFN-UNIT 1                        B 3.3-147                                Revision 0
 
ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE      3.a. Reactor Vessel Water Level - Low Low, Level 2 SAFETY ANALYSES, (LIS-3-58A-D) (continued)
LCO, and APPLICABILITY    one of the Functions assumed to be OPERABLE and capable of initiating HPCI during the transients analyzed in References 1, 2, and 3. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
Reactor Vessel Water Level - Low Low, Level 2 signals are initiated from four 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.
The Reactor Vessel Water Level - Low Low, Level 2 Allowable Value is high enough such that for complete loss of feedwater flow, the Reactor Core Isolation Cooling (RCIC) System flow with HPCI assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Reactor Vessel Water Level - Low Low Low, Level 1.
Four channels of Reactor Vessel Water Level - Low Low, Level 2 Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI initiation. Refer to LCO 3.5.1 for HPCI Applicability Bases.
For this instrument function, the nominal trip setpoint including the as-left tolerances is defined as the LSSS. The acceptable as-found band is based on a statistical combination of possible measurable uncertainties (i.e., setting tolerance, drift, temperature effects, and measurement and test equipment).
During instrument calibrations, if the as-found setpoint is found (continued)
BFN-UNIT 1                        B 3.3-148                          Revision 0, 41 November 09, 2006
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND 3, 4. High Pressure Coolant Injection System Isolation and Reactor Core Isolation Cooling System Isolation (continued)
The Steam Supply Line Pressure - Low and Turbine Exhaust Diaphragm Pressure - High Functions for HPCI and RCIC each contain four channels in a single trip system. The Steam Supply Line Pressure - Low channels are arranged in a series of logic parallel pairs to form one-out-of-two taken twice logic. Each HPCI isolation valve receives a single isolation signal from this logic. Each RCIC isolation valve receives an isolation signal from this logic through redundant logic systems. The trip system for the Turbine Exhaust Diaphragm Pressure - High Function contains two trip channels. Each trip channel contains two instrument channels (logic parallel pair). The output relays for the trip channels are arranged in logic systems (redundant logic systems for most isolation valves) such that both trip channels must trip (effectively one-out-of-two taken twice logic for the instrument channels) to cause an isolation.
The HPCI and RCIC Area Temperature - High Functions each contain sixteen channels, four Pump Room Area and twelve Torus Area channels (four channels for each area monitored).
Each trip system contains two trip channels; Logic A trip channel 1 (trip channel output relay 23A-K34 for HPCI and 13A-K10 for RCIC) and trip channel 2 (trip channel output relay 23A-K35 for HPCI and 13A-K11 for RCIC) and Logic B trip channel 1 (trip channel output relay 23A-K6 for HPCI and 13A-K30 for RCIC) and trip channel 2 (trip channel output relay 23A-K8 for HPCI and 13A-K31 for RCIC). Each trip channel receives one input from each of the four areas monitored. Any (continued)
BFN-UNIT 1                  B 3.3-190                              Revision 0
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND 3, 4. High Pressure Coolant Injection System Isolation and Reactor Core Isolation Cooling System Isolation (continued) one of these inputs will trip the associated trip channel. The trip channel output relays are arranged in logic systems (redundant logic systems for most isolation valves) such that trip channel 1 of either the A or B Logic and trip channel 2 of either the A or B Logic must trip (one-out-of-two taken twice logic) to cause an isolation.
HPCI and RCIC Functions isolate the Group 4 and 5 valves.
: 5. Reactor Water Cleanup System Isolation The RWCU Isolation Reactor Vessel Water Level - Low, Level 3 Function contains four channels. Each of the six Area Temperature - High Functions contain four channels which monitor the area associated with the Function. One channel for each of these RWCU Isolation Functions are provided in each of the four PCIS trip channels (trip channels A1 and A2 for PCIS trip system A and trip channels B1 and B2 for PCIS trip system B). Any one of these inputs will trip the associated PCIS trip channel. The PCIS trip channel output relays are arranged in logic systems (one logic system for the inboard valve and one logic system for the outboard valve) such that PCIS trip channels A1 or A2 and B1 or B2 must trip (one-out-of-two taken twice logic) to cause an isolation. The SLC System Initiation Function provides an isolation signal to close both RWCU isolation valves.
RWCU Isolation Functions are required for isolation of the Group 3 valves.
(continued)
BFN-UNIT 1                  B 3.3-191                              Revision 0
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND  6. Shutdown Cooling System Isolation (continued)
The Shutdown Cooling System Isolation Reactor Vessel Water Level - Low, Level 3 and Drywell Pressure - High Functions each contain four channels. One channel for each Function is provided in each of the four PCIS trip channels (trip channels A1 and A2 for PCIS trip system A and trip channels B1 and B2 for PCIS trip system B). Any one of these inputs will trip the associated PCIS trip channel. The PCIS trip channel output relays are arranged in logic systems (each division of logic provides a signal for one RHR LPCI to Reactor isolation valve and one RHR SDC Supply isolation valve) such that PCIS trip channels A1 or A2 and B1 or B2 must trip (one-out-of-two taken twice logic) to cause an isolation. Isolation of the RHR LPCI to Reactor isolation valves from these functions are enabled only when both RHR SDC Supply isolation valves are open.
The Reactor Steam Dome Pressure - High Function consists of two channels, one per trip system. The output relays from these channels are arranged in logic systems to provide one-out-of-two isolation logic to each RHR SDC isolation valve.
The Shutdown Cooling System Isolation Reactor Vessel Water Level - Low, Level 3 and Drywell Pressure - High Functions are required for isolation of the Group 2 RHR LPCI to Reactor and RHR SDC Supply isolation valves. The Reactor Steam Dome Pressure - High Function also isolates the Group 2 RHR SDC Supply isolation valves.
(continued)
BFN-UNIT 1                    B 3.3-192                              Revision 0
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE      3.d., 3.e., 3.f., 3.g., 4.d., 4.e., 4.f., 4.g. Area Temperature - High SAFETY ANALYSES, (TS-71-2A-H, J-N, P, R, S and TS-73-2A-H, J-N, P, R, S)
LCO, and APPLICABILITY    Area Temperature Functions are provided to detect a leak from (continued)    the associated system steam piping. The isolation occurs when a very small leak has occurred and is diverse to the high flow instrumentation. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. These Functions are not assumed in any FSAR transient or accident analysis, since bounding analyses are performed for large breaks such as recirculation or MSL breaks.
Area Temperature - High signals are initiated from bimetallic temperature switches that are appropriately located to protect the system that is being monitored. Four instruments monitor each area. HPCI and RCIC each have sixteen total channels of Area Temperature - High Function available. All of which are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
The Allowable Values are set low enough to detect a leak equivalent to 25 gpm.
These Functions isolate the Group 4 and 5 valves, as appropriate.
(continued)
BFN-UNIT 1                          B 3.3-204                                Revision 0
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE      Reactor Water Cleanup System Isolation SAFETY ANALYSES, LCO, and        5.a., 5.b., 5.c., 5.d., 5.e., 5.f. Area Temperature - High APPLICABILITY    (TIS-69-834A-D, 835A-D, 836A-D, 837A-D, 838A-D, 839A-D)
(continued)
RWCU Area Temperature Functions are provided to detect a leak from the RWCU System. The isolation occurs even when very small leaks have occurred. If the small leak continues without isolation, offsite dose limits may be reached. Credit for these instruments is not taken in any transient or accident analysis in the FSAR, since bounding analyses are performed for large breaks such as recirculation or MSL breaks.
Area temperature signals are initiated from temperature elements that are located in the areas monitored. Four sensors in each of the six monitored areas are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
The Area Temperature - High Allowable Values are set based on the maximum abnormal operating temperature for each area.
These Functions isolate the Group 3 valves.
(continued)
BFN-UNIT 1                        B 3.3-205                              Revision 0
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE      5.g. SLC System Initiation SAFETY ANALYSES, LCO, and        The isolation of the RWCU System is required when the SLC APPLICABILITY    System has been initiated to prevent dilution and removal of the (continued)    boron solution by the RWCU System (Ref. 4). An isolation signal for both RWCU isolation valves is initiated when the SLC pump start handswitch is not in the stop position.
There is no Allowable Value associated with this Function since the channels are mechanically actuated based solely on the position of the SLC System initiation switch.
The SLC System Initiation Function is required to be OPERABLE in MODES 1 and 2, because these are the only MODES where the reactor can be critical, and in MODE 3 because this MODE uses the SLC System sodium pentaborate as a buffering solution to maintain the pH level at or above 7 in the suppression pool in the event of a LOCA. These MODES are consistent with the Applicability for the SLC System (LCO 3.1.7).
As noted (footnote (a) to Table 3.3.6.1-1), the SLC initiation signal provides input to the isolation logic for both RWCU isolation valves.
5.h. Reactor Vessel Water Level - Low, Level 3 (LIS-3-203A-D)
Low 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. Therefore, isolation of some interfaces with the reactor vessel occurs to isolate the potential sources of a break. The isolation of the RWCU System on Level 3 supports actions to ensure that the fuel peak cladding (continued)
BFN-UNIT 1                        B 3.3-206                            Revision 0, 98 February 5, 2016
 
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.1 REQUIREMENTS (continued) Performance of the CHANNEL CHECK 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.3-234                          Revision 0, 123 March 22, 2020
 
CREV System Instrumentation B 3.3.7.1 BASES ACTIONS      B.1 and B.2 (continued)
Because of the diversity of sensors available to provide initiation signals and the redundancy of the CREV System design, an allowable out of service time of 12 hours has been shown to be acceptable (Refs. 3 and 4) to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the associated Function is still maintaining CREV System initiation capability.
A Function is considered to be maintaining CREV System initiation capability when sufficient channels are OPERABLE or in trip such that an initiation signal from the given Function will be generated on a valid signal. For Functions 1 and 2, this would require both PCIS trip systems to have at least one channel of the Function OPERABLE or in trip. In this situation (loss of CREV System initiation capability), the 12 hour allowance of Required Action B.2 is not appropriate. If the Function is not maintaining CREV System initiation capability, the CREV System must be declared inoperable within 1 hour of discovery of the loss of CREV System initiation capability.
The 1 hour Completion Time (B.1) is acceptable because it minimizes risk while allowing time for restoring or tripping of channels.
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 B.2.
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 initiation), Condition E must be entered and its Required Action taken.
(continued)
BFN-UNIT 1                      B 3.3-245                                Revision 0
 
CREV System Instrumentation B 3.3.7.1 BASES ACTIONS      C.1 and C.2 (continued)
Because of the diversity of sensors available to provide initiation signals and the redundancy of the CREV System design, an allowable out of service time of 24 hours is provided to permit restoration of any inoperable channel to OPERABLE status.
However, this out of service time is only acceptable provided the associated Function is still maintaining CREV System initiation capability. A Function is considered to be maintaining CREV System initiation capability when sufficient channels are OPERABLE or in trip such that an initiation signal from the given Function will be generated on a valid signal. For Functions 3 and 4, this would require each unit to have at least one channel of the Function OPERABLE or in trip. In this situation (loss of CREV System initiation capability), the 24 hour allowance of Required Action C.2 is not appropriate. If the Function is not maintaining CREV System initiation capability, the CREV System must be declared inoperable within 1 hour of discovery of the loss of CREV System initiation capability.
The 1 hour Completion Time (C.1) is acceptable because it minimizes risk while allowing time for restoring or tripping of channels.
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 C.2.
Placing the inoperable channel in trip performs the intended function of the channel (starts the selected CREV subsystem in the pressurization mode). Alternately, if it is not desired to place the channel in trip (e.g., as in the case where it is not desired to start the subsystem), Condition E must be entered and its Required Action taken.
(continued)
BFN-UNIT 1                    B 3.3-246                                Revision 0
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
SURVEILLANCE      SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 70% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is
                  < 70% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.
The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered inoperable. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                          B 3.4-9                      Revision 0, 45, 123 March 22, 2020
 
Jet Pumps B 3.4.2 BASES SURVEILLANCE SR 3.4.2.1 (continued)
REQUIREMENTS Individual jet pumps in a recirculation loop normally do not have the same flow. The unequal flow is due to the drive flow manifold, which does not distribute flow equally to all risers.
The flow (or jet pump diffuser to lower plenum differential pressure) pattern or relationship of one jet pump to the loop average is repeatable. An appreciable change in this relationship is an indication that increased (or reduced) resistance has occurred in one of the jet pumps. This may be indicated by an increase in the relative flow for a jet pump that has experienced beam cracks.
The deviations from normal are considered indicative of a potential problem in the recirculation drive flow or jet pump system (Ref. 2). Normal flow ranges and established jet pump flow and differential pressure patterns are established by plotting historical data as discussed in Reference 2.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by two Notes. Note 1 allows this Surveillance not to be performed until 4 hours after the associated recirculation loop is in operation, since these checks can only be performed during jet pump operation. The 4 hours is an acceptable time to establish conditions appropriate for data collection and evaluation.
(continued)
BFN-UNIT 1                    B 3.4-15                          Revision 0, 123 March 22, 2020
 
S/RVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 4.4.6.
: 2. FSAR, Section 14.5.1.
: 3. ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code).
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                  B 3.4-22                Revision 0, 43, 81, 123 March 22, 2020
 
RCS Operational LEAKAGE B 3.4.4 BASES ACTIONS      C.1 and C.2 (continued)
If any Required Action and associated Completion Time of Condition A or B is not met or if pressure boundary LEAKAGE exists, 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 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 safety systems.
SURVEILLANCE SR 3.4.4.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE. Leakage detection instrumentation is discussed in more detail in the Bases for LCO 3.4.5, "RCS Leakage Detection Instrumentation." Sump level and flow rate are typically monitored to determine actual LEAKAGE rates; however, other methods may be used to quantify LEAKAGE. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.4-28                        Revision 0, 123 March 22, 2020
 
RCS Operational LEAKAGE B 3.4.4 BASES (continued)
REFERENCES        1. 10 CFR 50.2.
: 2. 10 CFR 50.55a(c).
: 3. 10 CFR 50, Appendix A, GDC 55.
: 4. GEAP-5620, "Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws," April 1968.
: 5. NUREG-75/067, "Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping in Boiling Water Reactors,"
October 1975.
: 6. FSAR, Section 4.10.3.2.
: 7. Deleted.
: 8. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                        B 3.4-29                        Revision 0, 123 March 22, 2020
 
RCS Leakage Detection Instrumentation B 3.4.5 BASES (continued)
SURVEILLANCE      SR 3.4.5.1 REQUIREMENTS This SR is for the performance of a CHANNEL CHECK of the required primary containment atmospheric monitoring system instrumentation. The check gives reasonable confidence that the channel is operating properly. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.5.2 This SR is for the performance of a CHANNEL FUNCTIONAL TEST of the required primary containment atmospheric monitoring system instrumentation. The test ensures that the monitors can perform their function in the desired manner. The test also verifies the alarm setpoint and relative accuracy of the instrument string. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.5.3 This SR is for the performance of a CHANNEL CALIBRATION of required drywell floor drain sump flow integrator instrumentation channels. The calibration verifies the accuracy of the instrument string. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                          B 3.4-35                          Revision 0, 123 March 22, 2020
 
RCS Leakage Detection Instrumentation B 3.4.5 BASES SURVEILLANCE SR 3.4.5.4 REQUIREMENTS (continued) This SR is for the performance of a CHANNEL CALIBRATION of required leakage detection system instrumentation channels.
The calibration verifies the accuracy of the instrument string.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 30.
: 2. FSAR, Section 4.10.3.
: 3. GEAP-5620, "Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws," April 1968.
: 4. NUREG-75/067, "Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping in Boiling Water Reactors,"
October 1975.
: 5. FSAR, Section 4.10.3.2.
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                    B 3.4-36                      Revision 0, 43, 123 March 22, 2020
 
RCS Specific Activity B 3.4.6 BASES (continued)
SURVEILLANCE      SR 3.4.6.1 REQUIREMENTS This Surveillance is performed to ensure iodine remains within limit during normal operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that requires this Surveillance to be performed only in MODE 1 because the level of fission products generated in other MODES is much less.
REFERENCES        1. 10 CFR 50.67.
: 2. FSAR, Section 14.6.5.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                        B 3.4-41                    Revision 0, 29, 123 March 22, 2020
 
RHR Shutdown Cooling System - Hot Shutdown B 3.4.7 BASES (continued)
SURVEILLANCE      SR 3.4.7.1 REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This Surveillance is modified by a Note allowing sufficient time to align the RHR System for shutdown cooling operation after clearing the pressure interlock that isolates the system, or for placing a recirculation pump in operation. The Note takes exception to the requirements of the Surveillance being met (i.e., forced coolant circulation is not required for this initial 2 hour period), which also allows entry into the Applicability of this Specification in accordance with SR 3.0.4 since the Surveillance will not be "not met" at the time of entry into the Applicability.
REFERENCES        1. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                          B 3.4-48                            Revision 0, 123 March 22, 2020
 
RHR Shutdown Cooling System - Cold Shutdown B 3.4.8 BASES ACTIONS      B.1 and B.2 (continued)
During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem or recirculation pump), the reactor coolant temperature and pressure must be periodically monitored to ensure proper function of the alternate method.
The once per hour Completion Time is deemed appropriate.
SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This Surveillance verifies that one required RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                    B 3.4-54                        Revision 0, 123 March 22, 2020
 
RCS P/T Limits B 3.4.9 BASES ACTIONS      C.1 and C.2 (continued) analyses, or inspection of the components. ASME Code, Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.
Condition C is modified by a Note requiring Required Action C.2 be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
SURVEILLANCE SR 3.4.9.1 REQUIREMENTS Verification that operation is within limits is required when RCS pressure and temperature conditions are undergoing planned changes. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.
(continued)
BFN-UNIT 1                    B 3.4-62                            Revision 0, 123 March 22, 2020
 
RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7 REQUIREMENTS (continued) Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.
The flange temperatures must be verified to be above the limits before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied.
When in MODE 4 with RCS temperature d 85&deg;F, checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature d 100&deg;F, monitoring of the flange temperature is required to ensure the temperature is > 83qF.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.9.5 is modified by two Notes. Note 1 requires the Surveillance to be performed only when tensioning the reactor vessel head bolting studs. Note 2 allows the reactor vessel head bolts to be partially tensioned (four sequences of the seating pass) provided the studs and flange materials are
            > 70qF. SR 3.4.9.6 is modified by a Note that requires the (continued)
BFN-UNIT 1                    B 3.4-65                    Revision 0, 38, 123 March 22, 2020
 
Reactor Steam Dome Pressure B 3.4.10 BASES (continued)
ACTIONS          A.1 With the reactor steam dome pressure greater than the limit, prompt action should be taken to reduce pressure to below the limit and return the reactor to operation within the bounds of the analyses. The 15 minute Completion Time is reasonable considering the importance of maintaining the pressure within limits. This Completion Time also ensures that the probability of an accident occurring while pressure is greater than the limit is minimized. If the operator is unable to restore the reactor steam dome pressure to below the limit, then the reactor should be placed in MODE 3 to be operating within the assumptions of the transient analyses.
B.1 If the reactor steam dome pressure cannot be restored to within the limit within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE      SR 3.4.10.1 REQUIREMENTS Verification that reactor steam dome pressure is d 1050 psig ensures that the initial conditions of the design basis accidents and transients are met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                          B 3.4-69                      Revision 0, 50, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES ACTIONS      G.1 and G.2 (continued)
If any Required Action and associated Completion Time of Condition C, D, E, or F is not met, or if two or more ADS valves are inoperable, the plant must be brought to a condition 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 reactor steam dome pressure reduced to d 150 psig 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.
H.1 When multiple ECCS subsystems are inoperable, as stated in Condition H, the plant is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCI System, CS System, and LPCI subsystems full of water ensures that the ECCS will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.5-12                          Revision 0, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.2 REQUIREMENTS (continued) Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the HPCI System, this SR also includes the steam flow path for the turbine and the flow controller position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the RHR low pressure permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. This allows operation in the RHR shutdown cooling mode during MODE 3, if necessary.
(continued)
BFN-UNIT 1                    B 3.5-13                      Revision 0, 109, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.3 REQUIREMENTS (continued) Verification that ADS air supply header pressure is t 81 psig ensures adequate air pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least two valve actuations can occur with the drywell at 62.5% of design pressure plus three additional actuations at 0 psig drywell pressure (Ref. 10). The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of t 81 psig is provided by the Drywell Control Air System. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.1.4 Deleted.
(continued)
BFN-UNIT 1                    B 3.5-14                    Revision 0, 46, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.6, SR 3.5.1.7, and SR 3.5.1.8 (continued)
REQUIREMENTS pressure and flow are achieved to perform these tests. Reactor startup is allowed prior to performing the low pressure Surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance test is short. Alternately, the low pressure Surveillance test may be performed prior to startup using an auxiliary steam supply. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure test has been satisfactorily completed and there is no indication or reason to believe that HPCI is inoperable.
Therefore, SR 3.5.1.7 and SR 3.5.1.8 are modified by Notes that state the Surveillances are not required to be performed until 12 hours after the reactor steam pressure and flow are adequate to perform the test.
The Frequency for SR 3.5.1.6 is in accordance with the INSERVICE TESTING PROGRAM requirements. The Frequencies for SR 3.5.1.7 and SR 3.5.1.8 are controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.5-17                Revision 0, 43, 109, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.9 REQUIREMENTS (continued) The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This SR also ensures that the HPCI System will automatically restart on an RPV low-low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool.
The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
(continued)
BFN-UNIT 1                    B 3.5-18                      Revision 0, 43, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.10 REQUIREMENTS (continued) The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e., solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that excludes valve actuation.
This prevents an RPV pressure blowdown.
(continued)
BFN-UNIT 1                    B 3.5-19                      Revision 0, 43, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.11 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                  B 3.5-21                  Revision 0, 33, 43, 123 March 22, 2020
 
RCIC System B 3.5.3 BASES (continued)
SURVEILLANCE      SR 3.5.3.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge line of the RCIC System full of water ensures that the system will perform properly, injecting its full capacity into the Reactor Coolant System upon demand. This will also prevent a water hammer following an initiation signal. One acceptable method of ensuring the line is full is to vent at the high points. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.3.2 Verifying the correct alignment for manual, power operated, and automatic valves in the RCIC flow path provides assurance that the proper flow path will exist for RCIC operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are (continued)
BFN-UNIT 1                          B 3.5-34                            Revision 0, 123 March 22, 2020
 
RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.2 (continued)
REQUIREMENTS in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the RCIC System, this SR also includes the steam flow path for the turbine and the flow controller position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.3.3 and SR 3.5.3.4 The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated. The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Reactor steam pressure must be t 950 psig to perform SR 3.5.3.3 and t 150 psig to perform SR 3.5.3.4. Adequate steam flow is represented by at least one turbine bypass valve full open for SR 3.5.3.3 and at least one turbine bypass valve > 50% open for SR 3.5.3.4. Therefore, sufficient time is allowed (continued)
BFN-UNIT 1                    B 3.5-35            Revision 0, 50, 53, 109, 123 March 22, 2020
 
RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.3 and SR 3.5.3.4 (continued)
REQUIREMENTS after adequate pressure and flow are achieved to perform these SRs. Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance is short. Alternately, the low pressure Surveillance test may be performed prior to startup using an auxiliary steam supply. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure Surveillance has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable. Therefore, these SRs are modified by Notes that state the Surveillances are not required to be performed until 12 hours after the reactor steam pressure and flow are adequate to perform the test.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                  B 3.5-36                  Revision 0, 43, 109, 123 March 22, 2020
 
RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.5 REQUIREMENTS (continued) The RCIC System is required to actuate automatically in order to perform its design function satisfactorily. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of the RCIC System will cause the system to operate as designed, including actuation of the system throughout its emergency operating sequence; that is, automatic pump startup and actuation of all automatic valves to their required positions. This test also ensures the RCIC System will automatically restart on an RPV low-low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.2 overlaps this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that excludes vessel injection during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
(continued)
BFN-UNIT 1                    B 3.5-37                      Revision 0, 43, 123 March 22, 2020
 
Primary Containment B 3.6.1.1 BASES SURVEILLANCE SR 3.6.1.1.2 (continued)
REQUIREMENTS Satisfactory performance of this SR can be achieved by establishing a known differential pressure between the drywell and the suppression chamber and verifying that the pressure in either the suppression chamber or the drywell does not change by more than 0.25 inch of water per minute over a 10 minute period. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. NEDC-33960P, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate, Section 2.6.
: 2. FSAR, Section 14.6.
: 3. 10 CFR 50, Appendix J, Option B.
: 4. NEI 94-01, Revision 3A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J."
: 5. ANSI/ANS-56.8-2002, "American National Standard for Containment System Leakage Testing Requirements."
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                    B 3.6-6          Revision 0, 43, 112, 114, 123 March 22, 2020
 
Primary Containment Air Lock B 3.6.1.2 BASES SURVEILLANCE SR 3.6.1.2.1 (continued)
REQUIREMENTS The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 requires the results of airlock leakage tests be evaluated against the acceptance criteria of the Primary Containment Leakage Rate Testing Program, 5.5.12. This ensures that the airlock leakage is properly accounted for in determining the combined Type B and C primary containment leakage.
SR 3.6.1.2.2 The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment pressure, closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment.
Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur.
(continued)
BFN-UNIT 1                    B 3.6-16                          Revision 0, 123 March 22, 2020
 
Primary Containment Air Lock B 3.6.1.2 BASES SURVEILLANCE SR 3.6.1.2.2 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 5.2.3.4.5.
: 2. 10 CFR 50, Appendix J, Option B.
: 3. FSAR, Section 5.2.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                  B 3.6-17                        Revision 0, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.1 (continued)
REQUIREMENTS following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.2 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured, and is required to be closed during accident conditions is closed.
The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.
This SR does not require any testing or valve manipulation.
Rather, it involves verification that those PCIVs outside primary containment, and capable of being mispositioned, are in the correct position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.6-30                          Revision 0, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.4 REQUIREMENTS (continued) The traversing incore probe (TIP) shear isolation valves are actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that TIP valves will actuate when required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.5 Verifying the isolation time of each power operated, automatic PCIV is within limits is required to demonstrate OPERABILITY.
MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.6. The isolation time test ensures that the valve will isolate in a time period less than or equal to that assumed in the safety analyses. The isolation time and Frequency of this SR are in accordance with the requirements of the INSERVICE TESTING PROGRAM.
SR 3.6.1.3.6 Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. The isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the DBA analyses. This ensures that the calculated radiological consequences of these events remain within 10 CFR 50.67 limits. The Frequency of this SR is in accordance with the requirements of the INSERVICE TESTING PROGRAM.
(continued)
BFN-UNIT 1                    B 3.6-33                  Revision 0, 29, 109, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.7 REQUIREMENTS (continued) Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.1 overlaps this SR to provide complete testing of the safety function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.8 This SR requires a demonstration that a representative sample of reactor instrumentation line excess flow check valves (EFCV) are OPERABLE by verifying that the valves actuate to the isolation position on an actual or simulated instrument line break signal. This SR provides assurance that the instrumentation line EFCVs will perform so that the radiological consequences will not exceed the predicted radiological consequences during events evaluated in Reference 5. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.6-34                  Revision 0, 40, 43, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.9 REQUIREMENTS (continued) The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.10 The analyses in References 1 and 5 are based on leakage that is less than the specified leakage rate. Leakage through each MSIV must be d 100 scfh when tested at t Pt (25 psig). The combined leakage rate for all four main steam lines must be d 150 scfh when tested at t 25 psig in accordance with the Primary Containment Leakage Rate Testing Program. If the leakage rate through an individual MSIV exceeds 100 scfh, the leakage rate shall be restored below the alarm limit value as specified in the Containment Leakage Rate Testing Program referenced in TS 5.5.12. This ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate. The Frequency is specified in the Primary Containment Leakage Rate Testing Program.
(continued)
BFN-UNIT 1                    B 3.6-35                Revision 0, 43, 62, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES REFERENCES 1. FSAR, Section 14.6.
: 2. BFN Technical Instruction (TI), 0-TI-360.
: 3. 10 CFR 50, Appendix J, Option B.
: 4. FSAR, Section 5.2.
: 5. FSAR, Section 14.6.5.
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 7. FSAR Table 5.2-2.
: 8. Deleted.
: 9. MDQ0000012016000566, Revision 0, Main Steam Isolation Valve (MSIV) Loss of Coolant Accident (LOCA) Closure Analysis, dated September 2016.
BFN-UNIT 1                B 3.6-36        Revision 0, 40, 43, 62, 104, 123 March 22, 2020
 
Drywell Air Temperature B 3.6.1.4 BASES ACTIONS      B.1 and B.2 (continued)
If the drywell average air temperature cannot be restored to within limit 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.
SURVEILLANCE SR 3.6.1.4.1 REQUIREMENTS Verifying that the drywell average air temperature is within the LCO limit ensures that operation remains within the limits assumed for the primary containment analyses. Drywell air temperature is monitored in various quadrants and at various elevations (referenced to mean sea level). Due to the shape of the drywell, a volumetric average is used to determine an accurate representation of the actual average temperature.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.6-39                        Revision 0, 123 March 22, 2020
 
Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.5 BASES (continued)
SURVEILLANCE      SR 3.6.1.5.1 REQUIREMENTS Each vacuum breaker is verified to be closed to ensure that a potential breach in the primary containment boundary is not present. This Surveillance is performed by observing local or control room indications of vacuum breaker position or by verifying a differential pressure of 0.5 psid is maintained between the reactor building and suppression chamber. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Two Notes are added to this SR. The first Note allows reactor building-to-suppression chamber vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable breakers. A second Note is included to clarify that vacuum breakers open due to an actual differential pressure, are not considered as failing this SR.
SR 3.6.1.5.2 Each vacuum breaker must be cycled to ensure that it opens properly to perform its design function and returns to its fully closed position. This ensures that the safety analysis assumptions are valid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                          B 3.6-47                      Revision 0, 109, 123 March 22, 2020
 
Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.5 BASES SURVEILLANCE SR 3.6.1.5.3 REQUIREMENTS (continued) Demonstration of vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of d 0.5 psid is valid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. TVA Calculation ND-Q0064-900040.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                    B 3.6-48                    Revision 0, 43, 123 March 22, 2020
 
Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.6 BASES (continued)
SURVEILLANCE      SR 3.6.1.6.1 REQUIREMENTS Each vacuum breaker is verified closed to ensure that this potential large bypass leakage path is not present. This Surveillance is performed by observing the vacuum breaker position indication or by verifying that the rate of increase in suppression chamber pressure is less than 0.25 inches of water per minute over a ten minute period at a differential pressure of at least 1.0 psi. Note 2 specifies that vacuum breaker may be nonfully closed provided it is not more than 3&deg; open as indicated by position indication lights. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Note 1 has been added to this SR which allows suppression chamber-to-drywell vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers.
SR 3.6.1.6.2 Each required (i.e., required to be OPERABLE for opening) vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position. This ensures that the safety analysis assumptions are valid. The INSERVICE TESTING PROGRAM Frequency is based on operating experience that has demonstrated that the Frequency is adequate to assure OPERABILITY.
(continued)
BFN-UNIT 1                          B 3.6-55                      Revision 0, 109, 123 March 22, 2020
 
Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.6 BASES SURVEILLANCE SR 3.6.1.6.3 REQUIREMENTS (continued) Verification of the differential pressure required to open the vacuum breaker is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of 0.5 psid is valid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 5.2.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 3. Technical Requirements Manual.
BFN-UNIT 1                    B 3.6-56                      Revision 0, 43, 123 March 22, 2020
 
Suppression Pool Average Temperature B 3.6.2.1 BASES (continued)
SURVEILLANCE      SR 3.6.2.1.1 REQUIREMENTS The suppression pool average temperature is regularly monitored to ensure that the required limits are satisfied. The average temperature is determined by taking an arithmetic average of OPERABLE suppression pool water temperature channels. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The 5 minute Frequency during testing is justified by the rates at which tests will heat up the suppression pool, has been shown to be acceptable based on operating experience, and provides assurance that allowable pool temperatures are not exceeded.
The Frequency is further justified in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature condition.
REFERENCES        1. FSAR, Section 5.2.
: 2. FSAR, Section 14.6.
: 3. NUREG-0783, Suppression Pool Temperature Limits for BWR Containments, November 1981.
: 4. NUREG-0661, "Safety Evaluation Report Mark I Containment Long Term Program - Resolution of Generic Technical Activity A-7," July 1980.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 6. NEDC-22004-P, Browns Ferry Nuclear Plant Units 1, 2, and 3 Suppression Pool Temperature Response, October 1981.
BFN-UNIT 1                        B 3.6-63                  Revision 0, 66, 85, 123 March 22, 2020
 
Suppression Pool Water Level B 3.6.2.2 BASES ACTIONS      B.1 and B.2 (continued)
If suppression pool water level cannot be restored to within limits 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.
SURVEILLANCE SR 3.6.2.2.1 REQUIREMENTS Verification of the suppression pool water level is to ensure that the required limits are satisfied. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Sections 5.2 and 14.6.3.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                    B 3.6-67                        Revision 0, 123 March 22, 2020
 
RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS      D.1 and D.2 (continued)
If any Required Action and associated Completion Time cannot be met, 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.
SURVEILLANCE SR 3.6.2.3.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.6-72                      Amendment No. 241 Revision 0, 123 March 22, 2020
 
RHR Suppression Pool Spray B 3.6.2.4 BASES SURVEILLANCE SR 3.6.2.4.1 (continued)
REQUIREMENTS sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool spray mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.2.4.2 This Surveillance is performed using air or water to verify that the spray nozzles are not obstructed and that flow will be provided when required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Sections 5.2 and 14.6.3.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                    B 3.6-79                          Revision 0, 123 March 22, 2020
 
RHR Drywell Spray B 3.6.2.5 BASES ACTIONS      D.1 and D.2 (continued)
If any Required Action and the associated Completion Time cannot be met, 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 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.
SURVEILLANCE SR 3.6.2.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR drywell spray mode flow path provides assurance that the proper flow paths will exist for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR drywell cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.6-84                            Revision 0, 123 March 22, 2020
 
RHR Drywell Spray B 3.6.2.5 BASES SURVEILLANCE SR 3.6.2.5.2 REQUIREMENTS (continued) This Surveillance is performed using air to verify that the spray nozzles are not obstructed and that flow will be provided when required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Sections 5.2 and 14.6.3.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                  B 3.6-85                          Revision 3, 123 March 22, 2020
 
Drywell-to-Suppression Chamber Differential Pressure B 3.6.2.6 BASES (continued)
SURVEILLANCE      SR 3.6.2.6.1 REQUIREMENTS The drywell-to-suppression chamber differential pressure is regularly monitored to ensure that the required limits are satisfied. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 5.2.3.9.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                        B 3.6-89                        Revision 0, 123 March 22, 2020
 
CAD System B 3.6.3.1 BASES ACTIONS      B.1 and B.2 (continued)
The Completion Time of 7 days is a reasonable time to allow continued reactor operation with two CAD subsystems inoperable because the hydrogen control function is maintained (via the Primary Containment Inerting System) and because of the low probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flammability limit.
C.1 If any Required Action cannot be met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.3.1.1 REQUIREMENTS Verifying that there is t 2615 gal of liquid nitrogen supply in each nitrogen storage tank will ensure at least 7 days of post-LOCA CAD operation. This minimum volume of liquid nitrogen represents the analytical limit assumed in the analysis of the primary containment atmosphere following a postulated LOCA and does not include allowance for potential nitrogen boiloff and tank level instrumentation inaccuracies. This minimum volume of liquid nitrogen allows sufficient time after an accident to replenish the nitrogen supply for long term inerting. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.6-94                Revision 0, 34, 112, 123 October 24, 2018
 
CAD System B 3.6.3.1 BASES SURVEILLANCE SR 3.6.3.1.2 REQUIREMENTS (continued) Verifying the correct alignment for manual, power operated, and automatic valves in each of the CAD subsystem flow paths provides assurance that the proper flow paths exist for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing.
A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable because the CAD System is manually initiated. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.6-95                            Revision 0, 123 March 22, 2020
 
Primary Containment Oxygen Concentration B 3.6.3.2 BASES (continued)
SURVEILLANCE      SR 3.6.3.2.1 REQUIREMENTS The primary containment (drywell and suppression chamber) must be determined to be inert by verifying that oxygen concentration is < 4.0 v/o. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 5.2.6.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                        B 3.6-100                        Revision 0, 123 March 22, 2020
 
Secondary Containment B 3.6.4.1 BASES (continued)
SURVEILLANCE      SR 3.6.4.1.1 and SR 3.6.4.1.2 REQUIREMENTS Verifying that secondary containment equipment hatches and one access door in each access opening are closed ensures that the infiltration of outside air of such a 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 the access opening is closed. An access opening contains one inner and one outer door. In some cases, secondary containment access openings are shared such that a secondary containment barrier may have multiple inner doors. The main Equipment Access Lock (EAL) has a smaller sub-door on each of the large inner and outer main EAL doors. For the EAL, maintaining secondary containment OPERABILITY requires verifying that a large door and its integral sub-door are both closed. The intent is to not breach the secondary containment at any time when secondary containment is required. This is achieved by maintaining the inner or outer portion of the barrier closed at all times. However, all secondary containment access doors are normally kept closed, except when the access opening is being used for entry and exit or when maintenance is being performed on an access opening. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                          B 3.6-105                  Revision 0, 23, 29, 123 Amendment No. 238 March 22, 2020
 
Secondary Containment B 3.6.4.1 BASES SURVEILLANCE SR 3.6.4.1.3 and SR 3.6.4.1.4 REQUIREMENTS (continued) The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment. To ensure that all fission products are treated, SR 3.6.4.1.3 verifies that the SGT System will rapidly establish and maintain a pressure in the secondary containment that is less than the lowest postulated pressure external to the secondary containment boundary. This is confirmed by demonstrating that two SGT subsystems will draw down the secondary containment to t 0.25 inches of vacuum water gauge in d 120 seconds. This cannot be accomplished if the secondary containment boundary is not intact. SR 3.6.4.1.4 demonstrates that two SGT subsystems can maintain t 0.25 inches of vacuum water gauge at a stable flow rate d 12,000 cfm. Both of these SRs are performed under neutral
( 5 mph) wind conditions. Therefore, these two tests are used to ensure secondary containment boundary integrity. Since these SRs are secondary containment tests, they need not be performed with each combination of SGT subsystems. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 5.3.
: 2. FSAR, Section 14.6.3.
: 3. Deleted.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                  B 3.6-106                      Revision 29, 123 Amendment No. 235 March 22, 2020
 
SCIVs B 3.6.4.2 BASES (continued)
SURVEILLANCE      SR 3.6.4.2.1 REQUIREMENTS Verifying that the isolation time of each power operated, automatic SCIV is within limits is required to demonstrate OPERABILITY. The isolation time test ensures that the SCIV will isolate in a time period less than or equal to that assumed in the safety analyses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.4.2.2 Verifying that each automatic SCIV closes on a secondary containment isolation signal is required to prevent leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SCIV 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 14.6.3.
: 2. Deleted.
: 3. Technical Requirements Manual.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                          B 3.6-113                        Revision 29, 123 Amendment No. 235 March 22, 2020
 
SGT System B 3.6.4.3 BASES (continued)
SURVEILLANCE      SR 3.6.4.3.1 REQUIREMENTS Operating each SGT subsystem for t 15 continuous minutes with heaters on ensures that the 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. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.4.3.2 This SR verifies that the required SGT filter testing is performed in accordance with the 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). This SR will also include a chemical smoke test to check the sealing of gaskets for filter housing doors.
Specific test frequencies and additional information are discussed in detail in the VFTP.
(continued)
BFN-UNIT 1                        B 3.6-120                    Revision 0, 108, 123 March 22, 2020
 
SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.3 REQUIREMENTS (continued) This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.4.3.4 This SR verifies that the SGT decay heat discharge dampers are in the correct position. This ensures that the decay heat removal mode of SGT System operation is available. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 41.
: 2. FSAR, Section 5.3.3.7.
: 3. FSAR, Section 14.6.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                  B 3.6-121                            Revision 123 Amendment No. 235 March 22, 2020
 
RHRSW System B 3.7.1 BASES ACTIONS      G.1 and G.2 (continued)
If the RHRSW subsystem(s) or the RHRSW pump(s) cannot be restored to OPERABLE status within the associated Completion Times, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.1.1 REQUIREMENTS Verifying the correct alignment for each manual and power operated valve in each RHRSW subsystem flow path provides assurance that the proper flow paths will exist for RHRSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be realigned to its accident position. This is acceptable because the RHRSW System is a manually initiated system.
This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                      B 3.7-9                        Revision 0, 73, 123 March 22, 2020
 
EECW System and UHS B 3.7.2 BASES ACTIONS      A.1 (continued)
The 7 day Completion Time is based on the redundant EECW System capabilities afforded by the remaining OPERABLE pumps, the low probability of an accident occurring during this time period and is consistent with the allowed Completion Time for restoring an inoperable DG.
B.1 and B.2 If the required EECW pump cannot be restored to OPERABLE status within the associated Completion Time, or two or more EECW pumps are inoperable or the UHS is determined inoperable, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.2.1 REQUIREMENTS Verification of the UHS temperature ensures that the heat removal capability of the EECW System is within the assumptions of the DBA analysis (Ref. 5) and is sufficient for removal of heat from supported equipment to maintain OPERABILITY of that equipment. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.7-14                    Revision 0, 69, 123 March 22, 2020
 
EECW System and UHS B 3.7.2 BASES SURVEILLANCE SR 3.7.2.2 REQUIREMENTS (continued) Verifying the correct alignment for each manual and power operated valve in the EECW System flow paths provide assurance that the proper flow paths will exist for EECW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be automatically realigned to its accident position within the required time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
This SR is modified by a Note indicating that isolation of the EECW System to components or systems may render those components or systems inoperable, but does not affect the OPERABILITY of the EECW System. As such, when required EECW pumps, valves, and piping are OPERABLE, but a branch connection off the main header is isolated, the EECW System is still OPERABLE.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.7-15                            Revision 0, 123 March 22, 2020
 
CREV System B 3.7.3 BASES ACTIONS    B.1, B.2 and B.3 (continued)
Completion Time is reasonable based on low probability of a DBA occurring during this time period and the use of mitigating actions. The 90 day Completion Time is reasonable based on the determination that the mitigating actions will ensure protection of CRE occupants within analyzed limits while limiting the probability that CRE occupants will have to implement protective measures that may adversely affect their ability to control the reactor and maintain it in a safe shutdown condition in the event of a DBA. In addition, the 90 day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most problems within the CRE boundary.
C.1 With two CREV subsystems inoperable due to a HEPA filter inoperable or both CREV subsystems charcoal adsorbers inoperable, which do not impact the ability of the CREV subsystems to meet flowrate requirements specified in the VFTP, the HEPA filter and one of the charcoal adsorbers must be restored within 7 days. The 7 day Completion Time is based on the analysis for radiological dose to CRE occupants (Ref.
10), which has determined that the CRE 30 day dose after a DBA does not exceed 5 rem (TEDE) without credit for either the HEPA filter or the charcoal adsorbers.
D.1 With one CREV subsystem inoperable due to one inoperable charcoal adsorber, which does not impact the ability of the associated CREV subsystem to meet flowrate requirements specified in the VFTP, the charcoal adsorber must be restored within 14 days. The 14 day Completion Time is based on the analysis for radiological dose to CRE occupants (Ref. 10),
(continued)
BFN-UNIT 1                  B 3.7-21a                      Revision 0, 29, 67 Amendment No. 246, 275 August 10, 2012
 
CREV System B 3.7.3 BASES ACTIONS    D.1 (continued) which has determined that the CRE 30 day dose after the DBA does not exceed 5 rem (TEDE) without credit for either the HEPA filter or the charcoal adsorbers, and the capability of the remaining OPERABLE CREV subsystem.
E.1 and E.2 In MODE 1, 2, or 3, if the inoperable CREV subsystem or the CRE boundary cannot be restored to OPERABLE status within the required Completion Time, the unit must be placed in a MODE that minimizes accident risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
F.1 and F.2 During OPDRVs, if the inoperable CREV subsystem cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE CREV subsystem may be placed in the pressurization mode. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active failure will be readily detected.
An alternative to Required Action F.1 is to immediately initiate actions to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.
(continued)
BFN-UNIT 1                  B 3.7-22                      Revision 0, 29, 67 Amendment No. 246, 275 August 10, 2012
 
Control Room AC System B 3.7.4 BASES (continued)
APPLICABILITY    In MODE 1, 2, or 3, the Control Room AC System must be OPERABLE to ensure that the control room temperature will not exceed equipment OPERABILITY limits following control room isolation.
In MODES 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Control Room AC System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:
: a. During operations with a potential for draining the reactor vessel (OPDRVs);
: b. During CORE ALTERATIONS; and
: c. During movement of irradiated fuel assemblies in the secondary containment.
ACTIONS          A.1 With one Unit 1 and 2 control room AC subsystem inoperable, the inoperable Unit 1 and 2 control room AC subsystem must be restored to OPERABLE status within 30 days. With the unit in this condition, the remaining OPERABLE Unit 1 and 2 control room AC subsystem is adequate to perform the Unit 1 and 2 control room air conditioning function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in loss of the Unit 1 and 2 control room air conditioning function. The 30 day Completion Time is based on the low probability of an event occurring requiring control room isolation, the consideration that the remaining subsystem can provide the required protection, and the availability of alternate safety and nonsafety cooling methods.
(continued)
BFN-UNIT 1                          B 3.7-28                              Revision 0
 
Main Turbine Bypass System B 3.7.5 B 3.7 PLANT SYSTEMS B 3.7.5 Main Turbine Bypass System BASES BACKGROUND            The Main Turbine Bypass System is designed to control steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and cooldown. It allows excess steam flow from the reactor to the condenser without going through the turbine. The bypass capacity of the system is 21.3% of the Nuclear Steam Supply System rated steam flow. Sudden load reductions within the capacity of the steam bypass can be accommodated without reactor scram. The Main Turbine Bypass System consists of nine valves connected to the main steam lines between the main steam isolation valves and the turbine stop valve bypass valve chest. Each of these nine valves is operated by hydraulic cylinders. The bypass valves are controlled by the pressure regulation function of the Pressure Regulator and Turbine Generator Control System, as discussed in the FSAR, Section 7.11 (Ref. 1). The bypass valves are normally closed, and the pressure regulator controls the turbine control valves that direct all steam flow to the turbine. If the speed governor or the load limiter restricts steam flow to the turbine, the pressure regulator controls the system pressure by opening the bypass valves. When the bypass valves open, the steam flows from the bypass chest, through connecting piping, to the pressure breakdown assemblies, where a series of orifices are used to further reduce the steam pressure before the steam enters the condenser.
(continued)
BFN-UNIT 1                              B 3.7-32                    Revision 0, 103, 112 October 24, 2018
 
Main Turbine Bypass System B 3.7.5 BASES (continued)
APPLICABLE        The Main Turbine Bypass System is assumed to function during SAFETY ANALYSES  abnormal operational transients (e.g., the feedwater controller failure-maximum demand event), as discussed in the FSAR, Section 14.5.1.1 (Ref. 2). Opening the bypass valves during the event mitigates the increase in reactor vessel pressure, which affects the MCPR during the event. An inoperable Main Turbine Bypass System may result in APLHGR, MCPR, and LHGR penalties.
The Main Turbine Bypass System satisfies Criterion 3 of the NRC Policy Statement (Ref. 3).
LCO              The Main Turbine Bypass System is required to be OPERABLE to limit peak pressure in the main steam lines and maintain reactor pressure within acceptable limits during events that cause rapid pressurization, so that the APLHGR limits, MCPR Safety Limit, and LHGR limits are not exceeded. With the Main Turbine Bypass System inoperable, modifications to the APLHGR limits (LCO 3.2.1, AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)), the MCPR limits (LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR)),
and LHGR limits (LCO 3.2.3, LINEAR HEAT GENERATION RATE (LHGR)), may be applied to allow this LCO to be met.
The APLHGR, MCPR, and LHGR limits for the inoperable Main Turbine Bypass System are specified in the COLR. An OPERABLE Main Turbine Bypass System requires the bypass valves to open in response to increasing main steam line pressure. This response is within the assumptions of the applicable analysis (Ref. 2).
(continued)
BFN-UNIT 1                        B 3.7-33                          Revision 0, 68 October 18, 2012
 
Main Turbine Bypass System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.3 REQUIREMENTS (continued) This SR ensures that the TURBINE BYPASS SYSTEM RESPONSE TIME is in compliance with the assumptions of the appropriate safety analysis. The response time limits are specified in the cycle specific transient analyses performed to support the preparation of FSAR, Appendix N, Supplemental Reload Licensing Report (Ref. 4). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 7.11.
: 2. FSAR, Section 14.5.1.1.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 4. FSAR, Appendix N.
BFN-UNIT 1                    B 3.7-36                  Revision 0, 43, 103, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.1 and SR 3.8.1.4 (continued)
REQUIREMENTS SR 3.8.1.4 requires the DG starts from standby conditions and achieves required voltage and frequency within 10 seconds.
The 10 second start requirement supports the assumptions in the design basis LOCA analysis of FSAR, Section 14.6.3 (Ref. 10). The 10 second start requirement is not applicable to SR 3.8.1.1 (see the Note for SR 3.8.1.1), when a modified start procedure as described above is used. If a modified start is not used, the 10 second start requirement of SR 3.8.1.4 applies.
Since SR 3.8.1.4 does require a 10 second start, it is more restrictive than SR 3.8.1.1, and it may be performed in lieu of SR 3.8.1.1. This procedure is the intent of Note 1 of SR 3.8.1.1.
In addition to the SR requirements, the time for the DG to reach steady state operation, unless the modified DG start method is employed, is periodically monitored and the trend evaluated to identify degradation of governor performance.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.8-29                          Revision 0, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 REQUIREMENTS (continued) This Surveillance demonstrates that the DGs are capable of synchronizing and accepting greater than 90 percent of the continuous rating. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source.
Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Note 1 modifies this Surveillance to indicate that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized.
Note 2 modifies this Surveillance by stating that momentary transients because of changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test.
Note 3 indicates that this Surveillance should be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations.
Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.
(continued)
BFN-UNIT 1                    B 3.8-30                            Revision 0, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.3 REQUIREMENTS (continued) This Surveillance demonstrates that each required fuel oil transfer pump operates and transfers fuel oil from its associated 7-day storage tank to its associated engine fuel oil tank. It is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE.
The design of fuel transfer systems is such that pumps that transfer the fuel oil operate automatically in order to maintain an adequate volume of fuel oil in the engine tank during or following DG operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.1.4 See SR 3.8.1.1.
(continued)
BFN-UNIT 1                    B 3.8-31                          Revision 0, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.5 (continued)
REQUIREMENTS The voltage tolerances specified in this SR are based on the degraded voltage and overvoltage relay settings. The frequency tolerances specified in this SR are derived from Safety Guide 9 (Ref. 3) recommendations for response during load sequence intervals. The voltage and frequency specified are consistent with the design range of the equipment powered by the DG. SR 3.8.1.5.a corresponds to the maximum frequency excursion, while SR 3.8.1.5.b and 3.8.1.5.c are steady state voltage and frequency values to which the system must recover following load rejection. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, the Note requires that, if synchronized to offsite power, testing must be performed using a power factor d 0.9. This power factor is chosen to be representative of the actual design basis inductive loading that the DG would experience.
SR 3.8.1.6 This Surveillance demonstrates that the DG automatically starts from the design basis actuation signal (LOCA signal). This test will also verify the start of the Unit 3 DGs aligned to the SGT and CREV Systems on an accident signal from Unit 1. In order to minimize the number of DGs involved in testing, demonstration of automatic starts of the Unit 3 DGs on an accident signal from Unit 1 may be performed in conjunction (continued)
BFN-UNIT 1                      B 3.8-33                            Revision 123 Amendment No. 235 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.6 (continued)
REQUIREMENTS with testing to demonstrate automatic starts of the Unit 3 DGs on an accident signal from Unit 3. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
To minimize wear and tear on the DGs, this SR has been modified by a Note which permits DG starts to be preceded by an engine prelube period followed by a warmup period.
SR 3.8.1.7 Demonstration periodically that the DGs can start and run continuously at full load capability for an interval of not less than 24 hours - 22 hours of which is at a load equivalent to the continuous rating of the DG, and 2 hours of which is at a load equivalent to 105 percent to 110 percent of the continuous duty rating of the DG. The DG starts for this Surveillance can be performed either from standby or hot conditions. The provisions for prelube and warmup, discussed in SR 3.8.1.1, and for gradual loading, discussed in SR 3.8.1.2, are applicable to this SR.
In order to ensure that the DG is tested under load conditions that are as close to design conditions as possible, testing must be performed using a power factor d 0.9. This power factor is chosen to be representative of the actual design basis inductive loading that the DG could experience. A load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.
(continued)
BFN-UNIT 1                    B 3.8-34                              Revision 123 Amendment No. 235 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.7 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This Surveillance has been modified by a Note that states that momentary transients due to changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test.
SR 3.8.1.8 Under accident conditions (and loss of offsite power) loads are sequentially connected to the shutdown boards by automatic individual pump timers. The individual pump timers control the permissive and starting signals to motor breakers to prevent overloading of the DGs due to high motor starting currents. This SR is demonstrated by performance of SR 3.3.5.1.5 for the Core Spray and LPCI pump timers, SR 3.7.2.3 for the EECW pump timers, and SR 3.8.1.9.b for the 480 V load shed logic timers. The allowable values for these timers ensure that sufficient time exists for the DG to restore frequency and voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not violated. Reference 2 provides a summary of the automatic loading of ESF shutdown boards.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.8-35                            Revision 123 Amendment No. 235 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.9 (continued)
REQUIREMENTS mode of operation. In lieu of actual demonstration of the connection and loading of these loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note. The reason for the Note is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil being continuously circulated and temperature maintained consistent with manufacturer recommendations.
SR 3.8.1.10 This Surveillance is provided to direct that the appropriate Surveillances for the required Unit 3 DGs are governed by the Unit 3 Technical Specifications. Performance of the applicable Unit 3 Surveillances will satisfy any Unit 3 requirements, as well as this Unit 1 and 2 Surveillance requirement. The Frequency required by the applicable Unit 3 SR also governs performance of that SR for both Units.
(continued)
BFN-UNIT 1                    B 3.8-37                        Revision 42, 123 Amendment No. 235 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES (continued)
REFERENCES        1. 10 CFR 50, Appendix A, GDC 17.
: 2. FSAR, Chapter 8.
: 3. Safety Guide 9.
: 4. FSAR, Chapter 6.
: 5. FSAR, Chapter 14.
: 6. Regulatory Guide 1.93.
: 7. Generic Letter 84-15.
: 8. Deleted.
: 9. ANSI C84.1, 1982.
: 10. FSAR, Section 14.6.3.
: 11. IEEE Standard 308.
: 12. FSAR, Section 8.5, Table 8.5-6.
: 13. FSAR, Section 8.5.2.
: 14. TVA Design Criteria BFN-50-7082.
: 15. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                        B 3.8-38                      Revision 0, 123 March 22, 2020
 
Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES (continued)
SURVEILLANCE      SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 7 days at full load. The fuel oil level equivalent to a 7-day supply is 35,280 gallons when calculated in accordance with References 2 and 6. The required fuel storage volume is determined using the most limiting energy content of the stored fuel. Using the known correlation of diesel fuel oil absolute specific gravity or API gravity to energy content, the required diesel generator output, and the corresponding fuel consumption rate, the on site fuel storage volume required for 7 days of operation can be determined. SR 3.8.3.3 requires new fuel to be tested to verify that the absolute specific gravity or API gravity is within the range assumed in the diesel fuel oil consumption calculations. The 7-day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.3.2 This Surveillance ensures that sufficient lubricating oil inventory is available to support at least 7 days of full load operation for each DG. The lube oil inventory equivalent to a 7-day supply is 175 gallons and is based on the DG manufacturer's consumption values for the run time of the DG.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                          B 3.8-55                  Revision 0, 63, 95, 123 March 22, 2020
 
Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.4 REQUIREMENTS (continued) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.3.5 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Periodic removal of water from the fuel storage tanks eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling.
In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 8.5.3.4.
: 2. Regulatory Guide 1.137, Revision 1, October 1979.
: 3. FSAR, Chapter 6.
: 4. FSAR, Chapter 14.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                    B 3.8-56b                      Revision 0, 95, 123 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES (continued)
SURVEILLANCE      SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or a battery cell) in a fully charged state, while supplying adequate power to the connected DC loads. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.4.2 and SR 3.8.4.5 Battery charger capability requirements are based on the design capacity of the chargers (Ref. 4). According to Regulatory Guide 1.32 (Ref. 8), the battery charger supply is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and verification of the charger's ability to recharge the battery ensures that these requirements can be satisfied.
(continued)
BFN-UNIT 1                          B 3.8-64                            Revision 0, 123 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.2 and SR 3.8.4.5 (continued)
REQUIREMENTS SR 3.8.4.2 verifies that the chargers are capable of charging the batteries after their designed duty cycle testing and ensures that the chargers will perform their design function. This SR is modified by a Note that allows the performance of SR 3.8.4.5 in lieu of this Surveillance requirement. SR 3.8.4.5 verifies that the chargers are capable of charging the batteries after each discharge test and ensures that the chargers are capable of performing at maximum output. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.4.5 is modified by a Note. The Note is added to this SR to acknowledge that credit may be taken for unplanned events that satisfy the Surveillance.
SR 3.8.4.3 A battery service test is a special test of the battery's capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length corresponds to the design duty cycle requirements as specified in Reference 4.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.8-65                              Revision 123 Amendment No. 235 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.3 (continued)
REQUIREMENTS This SR is modified by a Note that allows the performance of a modified performance discharge test in lieu of a service test.
The modified performance discharge test is a simulated duty cycle consisting of just two periods (the one minute rate, followed by the test rate employed for the performance test) or three periods (the one minute rate, followed by the second minute rate followed by the test rate employed for the performance test) both of which envelope the duty cycle of the service test. Since the ampere-hours removed by the rated one or two minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.
A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity.
Initial conditions for the modified performance discharge test should be identical to those specified for a service test.
(continued)
BFN-UNIT 1                      B 3.8-66                      Revision 0, 58, 123 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.4 REQUIREMENTS (continued) A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.
A battery modified performance discharge test is described in the Bases for SR 3.8.4.3. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.4; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.4 while satisfying the requirements of SR 3.8.4.3 at the same time.
The acceptance criteria for this Surveillance is consistent with IEEE-450 (Ref. 7) and IEEE-485 (Ref. 10). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. If the battery shows degradation, or if the battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity t 100% of the manufacturer's rating. Degradation is indicated, according to IEEE-450 (Ref. 7), when the battery capacity drops by more than 10%
relative to its capacity on the previous performance test or when it is 10% below the manufacturer's rating. All these Frequencies are consistent with the recommendations in IEEE-450 (Ref. 7).
(continued)
BFN-UNIT 1                    B 3.8-67                          Revision 0, 123 March 22, 2020
 
Battery Cell Parameters B 3.8.6 BASES (continued)
SURVEILLANCE      SR 3.8.6.1 REQUIREMENTS This SR verifies that Category A battery cell parameters are consistent with IEEE-450 (Ref. 3), including voltage, specific gravity, and electrolyte temperature of pilot cells. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.6.2 The inspection of specific gravity and voltage is consistent with IEEE-450 (Ref. 3). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.6.3 This Surveillance verification that the average temperature of representative cells is within limits is consistent with a recommendation of IEEE-450 (Ref. 3) that states that the temperature of electrolytes in representative (10 percent of) cells should be determined. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Lower than normal temperatures act to inhibit or reduce battery capacity. This SR ensures that the operating temperatures remain within an acceptable operating range. This limit is based on manufacturer's recommendations.
(continued)
BFN-UNIT 1                        B 3.8-78                            Revision 0, 123 March 22, 2020
 
Distribution Systems - Operating B 3.8.7 BASES ACTIONS      G.1 and G.2 (continued)
If the inoperable distribution subsystem cannot be restored to OPERABLE status within the associated Completion Time, the unit 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.
H.1 Condition H corresponds to a level of degradation in the electrical distribution system that causes a required safety function to be lost. When more than one AC or DC electrical power distribution subsystem is lost, and this results in the loss of a required function, the plant is in a condition outside the accident analysis. Therefore, no additional time is justified for continued operation. LCO 3.0.3 must be entered immediately to commence a controlled shutdown.
SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the AC and DC electrical power distribution subsystem is functioning properly, with the buses energized. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                    B 3.8-100                      Revision 0, 33, 123 March 22, 2020
 
Distribution Systems - Shutdown B 3.8.8 BASES (continued)
SURVEILLANCE      SR 3.8.8.1 REQUIREMENTS This Surveillance verifies that the AC and DC electrical power distribution subsystem is functioning properly, with the buses energized. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Chapter 6.
: 2. FSAR, Chapter 14.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                        B 3.8-108                          Revision 0, 123 March 22, 2020
 
Refueling Equipment Interlocks B 3.9.1 BASES ACTIONS      A.1, A.2.1, and A.2.2 (continued) remain inserted). Required Action A.2.2 is normally performed after placing the rod withdrawal block in effect, and provides a verification that all control rods are fully inserted. This verification that all control rods are fully inserted is in addition to the periodic verifications required by SR 3.9.3.1. Like Required Action A.1, Required Actions A.2.1 and A.2.2 ensure unacceptable operations are blocked (e.g., loading fuel into a cell with the control rod withdrawn). It is not the intent of Actions A.2 to eliminate the first performance of SR 3.9.1.1 prior to in-vessel fuel movement. It is expected that the refueling interlocks would be operable except for equipment failure or expiration of the required surveillance interval, and Actions A.2 would not be entered as a convenience for avoiding the first performance of SR 3.9.1.1.
SURVEILLANCE SR 3.9.1.1 REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST demonstrates each required refueling equipment interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested. This SR is only required for refueling equipment in use.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                      B 3.9-5                      Amendment No. 242 Revision 0, 16, 123 March 22, 2020
 
Refuel Position One-Rod-Out Interlock B 3.9.2 BASES (continued)
SURVEILLANCE      SR 3.9.2.1 REQUIREMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel. During control rod withdrawal in MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock will be OPERABLE when required. By "locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel),
an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested.
(continued)
BFN-UNIT 1                          B 3.9-9                          Revision 0, 123 March 22, 2020
 
Refuel Position One-Rod-Out Interlock B 3.9.2 BASES SURVEILLANCE SR 3.9.2.2 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. To perform the required testing, the applicable condition must be entered (i.e., a control rod must be withdrawn from its full-in position). Therefore, SR 3.9.2.2 has been modified by a Note that states the CHANNEL FUNCTIONAL TEST is not required to be performed until 1 hour after any control rod is withdrawn.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 26.
: 2. FSAR, Section 7.6.3.
: 3. FSAR, Section 14.5.4.3.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                    B 3.9-10                      Revision 0, 60, 123 March 22, 2020
 
Control Rod Position B 3.9.3 BASES (continued)
ACTIONS          A.1 With all control rods not fully inserted during the applicable conditions, an inadvertent criticality could occur that is not analyzed in the FSAR. All fuel loading operations must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE      SR 3.9.3.1 REQUIREMENTS During refueling, to ensure that the reactor remains subcritical, all control rods must be fully inserted prior to and during fuel loading. Periodic checks of the control rod position ensure this condition is maintained.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. 10 CFR 50, Appendix A, GDC 26.
: 2. FSAR, Section 14.5.4.3.
: 3. FSAR, Section 14.5.4.4.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                          B 3.9-13                      Revision 0, 60, 123 March 22, 2020
 
Control Rod OPERABILITY - Refueling B 3.9.5 BASES (continued)
ACTIONS          A.1 With one or more withdrawn control rods inoperable, action must be immediately initiated to fully insert the inoperable control rod(s). Inserting the control rod(s) ensures the shutdown and scram capabilities are not adversely affected. Actions must continue until the inoperable control rod(s) is fully inserted.
SURVEILLANCE      SR 3.9.5.1 and SR 3.9.5.2 REQUIREMENTS During MODE 5, the OPERABILITY of control rods is primarily required to ensure a withdrawn control rod will automatically insert if a signal requiring a reactor shutdown occurs. Because no explicit analysis exists for automatic shutdown during refueling, the shutdown function is satisfied if the withdrawn control rod is capable of automatic insertion and the associated CRD scram accumulator pressure is t 940 psig.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. An automatic accumulator monitor may be used to continuously satisfy SR 3.9.5.2.
SR 3.9.5.1 is modified by a Note that allows 7 days after withdrawal of the control rod to perform the Surveillance. This acknowledges that the control rod must first be withdrawn before performance of the Surveillance, and therefore avoids potential conflicts with SR 3.0.3 and SR 3.0.4.
(continued)
BFN-UNIT 1                          B 3.9-21                          Revision 0, 123 March 22, 2020
 
Control Rod Position B 3.9.3 BASES (continued)
ACTIONS          A.1 With all control rods not fully inserted during the applicable conditions, an inadvertent criticality could occur that is not analyzed in the FSAR. All fuel loading operations must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE      SR 3.9.3.1 REQUIREMENTS During refueling, to ensure that the reactor remains subcritical, all control rods must be fully inserted prior to and during fuel loading. Periodic checks of the control rod position ensure this condition is maintained.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. 10 CFR 50, Appendix A, GDC 26.
: 2. FSAR, Section 14.5.4.3.
: 3. FSAR, Section 14.5.4.4.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                          B 3.9-13                      Revision 0, 60, 123 March 22, 2020
 
RPV Water Level B 3.9.6 BASES (continued)
ACTIONS          A.1 If the water level is < 22 ft above the top of the RPV flange, all operations involving movement of fuel assemblies and handling of control rods within the RPV shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement and control rod handling shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE      SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 22 ft above the top of the RPV flange ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. 2).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. Regulatory Guide 1.183.
: 2. FSAR, Section 14.6.4.
: 3. NUREG-0800, Section 15.0.1.
: 4. 10 CFR 50.67.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1                          B 3.9-25                      Revision 0, 29, 123 March 22, 2020
 
Reactor Mode Switch Interlock Testing B 3.10.2 BASES (continued)
SURVEILLANCE      SR 3.10.2.1 and SR 3.10.2.2 REQUIREMENTS Meeting the requirements of this Special Operations LCO maintains operation consistent with or conservative to operating with the reactor mode switch in the shutdown position (or the refuel position for MODE 5). The functions of the reactor mode switch interlocks that are not in effect, due to the testing in progress, are adequately compensated for by the Special Operations LCO requirements. The administrative controls are to be periodically verified to ensure that the operational requirements continue to be met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 7.2.3.7.
: 2. FSAR, Section 14.5.3.3.
: 3. FSAR, Section 14.5.3.4.
BFN-UNIT 1                        B 3.10-12                          Revision 0, 123 March 22, 2020
 
Single Control Rod Withdrawal - Hot Shutdown B 3.10.3 BASES (continued)
SURVEILLANCE      SR 3.10.3.1, SR 3.10.3.2, and SR 3.10.3.3 REQUIREMENTS The other LCOs made applicable in this Special Operations LCO are required to have their Surveillances met to establish that this Special Operations LCO is being met. If the local array of control rods is inserted and disarmed (electrically or hydraulically) while the scram function for the withdrawn rod is not available, periodic verification in accordance with SR 3.10.3.2 is required to preclude the possibility of criticality.
SR 3.10.3.2 has been modified by a Note, which clarifies that this SR is not required to be met if SR 3.10.3.1 is satisfied for LCO 3.10.3.d.1 requirements, since SR 3.10.3.2 demonstrates that the alternative LCO 3.10.3.d.2 requirements are satisfied.
Also, SR 3.10.3.3 verifies that all control rods other than the control rod being withdrawn are fully inserted. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 14.5.3.3.
BFN-UNIT 1                        B 3.10-18                          Revision 0, 123 March 22, 2020
 
Single Control Rod Withdrawal - Cold Shutdown B 3.10.4 BASES (continued)
SURVEILLANCE      SR 3.10.4.1, SR 3.10.4.2, SR 3.10.4.3, and SR 3.10.4.4 REQUIREMENTS The other LCOs made applicable by this Special Operations LCO are required to have their associated Surveillances met to establish that this Special Operations LCO is being met. If the local array of control rods is inserted and disarmed (electrically or hydraulically) while the scram function for the withdrawn rod is not available, periodic verification is required to ensure that the possibility of criticality remains precluded. Verification that all the other control rods are fully inserted is required to meet the SDM requirements. Verification that a control rod withdrawal block has been inserted ensures that no other control rods can be inadvertently withdrawn under conditions when position indication instrumentation is inoperable for the affected control rod. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.10.4.2 and SR 3.10.4.4 have been modified by Notes, which clarify that these SRs are not required to be met if the alternative requirements demonstrated by SR 3.10.4.1 are satisfied.
REFERENCES        1. FSAR, Section 14.5.3.3.
BFN-UNIT 1                        B 3.10-25                            Revision 0, 123 March 22, 2020
 
Single CRD Removal - Refueling B 3.10.5 BASES SURVEILLANCE SR 3.10.5.1, SR 3.10.5.2, SR 3.10.5.3, SR 3.10.5.4, REQUIREMENTS and SR 3.10.5.5 (continued) control rod. The Surveillance for LCO 3.1.1, which is made applicable by this Special Operations LCO, is required in order to establish that this Special Operations LCO is being met.
Verification that no other CORE ALTERATIONS are being made is required to ensure the assumptions of the safety analysis are satisfied.
Periodic verification of the administrative controls established by this Special Operations LCO is prudent to preclude the possibility of an inadvertent criticality. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 14.5.3.3.
BFN-UNIT 1                    B 3.10-31                          Revision 0, 123 March 22, 2020
 
Multiple Control Rod Withdrawal - Refueling B 3.10.6 BASES (continued)
ACTIONS          A.1, A.2, A.3.1, and A.3.2 If one or more of the requirements of this Special Operations LCO are not met, the immediate implementation of these Required Actions restores operation consistent with the normal requirements for refueling (i.e., all control rods inserted in core cells containing one or more fuel assemblies) or with the exceptions granted by this Special Operations LCO. The Completion Times for Required Action A.1, Required Action A.2, Required Action A.3.1, and Required Action A.3.2 are intended to require that these Required Actions be implemented in a very short time and carried through in an expeditious manner to either initiate action to restore the affected CRDs and insert their control rods, or initiate action to restore compliance with this Special Operations LCO.
SURVEILLANCE      SR 3.10.6.1, SR 3.10.6.2, and SR 3.10.6.3 REQUIREMENTS Periodic verification of the administrative controls established by this Special Operations LCO is prudent to preclude the possibility of an inadvertent criticality. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. SR 3.10.6.3 is modified by a Note stating that the SR is only required to be met during refueling.
REFERENCES        1. FSAR, Section 14.5.3.3.
BFN-UNIT 1                          B 3.10-35                          Revision 0, 123 March 22, 2020
 
SDM Test -Refueling B 3.10.8 BASES (continued)
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 this Special Operations LCO, are required to have applicable Surveillances met to establish that this Special Operations LCO is being met. However, the control rod withdrawal sequences during the SDM tests may be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2 requirements) or by a second licensed operator or other qualified member of the technical staff (i.e., personnel trained in accordance with an approved training program for this test). As noted, either the applicable SRs for the RWM (LCO 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.8.2), or the proper movement of control rods must be verified (SR 3.10.8.3).
This latter verification (i.e., SR 3.10.8.3) must be performed during control rod movement to prevent deviations from the specified sequence. These Surveillances provide adequate assurance that the specified test sequence is being followed.
SR 3.10.8.4 Periodic verification of the administrative controls established by this LCO will ensure that the reactor is operated within the bounds of the safety analysis. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 1                          B 3.10-47                      Revision 0, 40, 123 March 22, 2020
 
SDM Test -Refueling B 3.10.8 BASES SURVEILLANCE      SR 3.10.8.5 REQUIREMENTS (continued)      Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the "full out" notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events.
SR 3.10.8.6 CRD charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram signal. Since the reactor is depressurized in MODE 5, there is insufficient reactor pressure to scram the control rods. Verification of charging water pressure ensures that if a scram is required, capability for rapid control rod insertion would exist. The minimum pressure of 940 psig, which is well below the expected pressure of approximately 1100 psig, ensures sufficient pressure for rapid control rod insertion. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. NEDE-24011-P-A, Rev. 16, "General Electric Standard Application for Reactor Fuel," October 2007.
: 2. Letter from T. Pickens (BWROG) to G. C. Lainas, NRC, "Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," August 15, 1986.
______________________________________                                      (continued)
BFN-UNIT 1                          B 3.10-48                      Revision 0, 68, 123 March 22, 2020
 
Control Rod OPERABILITY B 3.1.3 BASES (continued)
SURVEILLANCE      SR 3.1.3.1 REQUIREMENTS The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.3.2 (Deleted.)
SR 3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal.
This surveillance is not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of banked position withdrawal sequence (BPWS) (LCO 3.1.6) and the RWM (LCO 3.3.2.1). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                        B 3.1-23                            Revision 0, 123 Amendment 301 March 22, 2020
 
Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE SR 3.1.4.2 REQUIREMENTS (continued) Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods. This sample remains representative if no more than 7.5% of the control rods in the sample tested are determined to be "slow." With more than 7.5% of the sample declared to be "slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 7.5% criterion (i.e., 7.5% of the entire sample) is satisfied, or until the total number of "slow" control rods (throughout the core from all Surveillances) exceeds the LCO limit. For planned testing, the control rods selected for the sample should be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.1-31                  Revision 0, 9, 35, 55, 123 Amendment No. 266 March 22, 2020
 
Control Rod Scram Accumulators B 3.1.5 BASES (continued)
SURVEILLANCE      SR 3.1.5.1 REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be checked periodically to ensure adequate accumulator pressure exists to provide sufficient scram force. An automatic accumulator monitor may be used to continuously satisfy this requirement.
The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940 psig is well below the expected pressure of 1100 psig (Ref. 1). Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant degradation in scram times does not occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 3.4.6.
: 2. FSAR, Section 14.5.
: 3. FSAR, Section 14.6.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                        B 3.1-40                        Revision 0, 123 March 22, 2020
 
Rod Pattern Control B 3.1.6 BASES ACTIONS    A.1 and A.2 (continued) operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence. When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement must be stopped except for moves needed to correct the rod pattern, or scram if warranted.
Required Action A.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a second licensed operator or a qualified member of the technical staff. This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2. The allowed Completion Time of 8 hours is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a CRDA occurring during the time the control rods are out of sequence.
B.1 and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence (Ref. 8). Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod worth than withdrawals have. Required Action B.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position.
(continued)
BFN-UNIT 2                  B 3.1-44                              Revision 0
 
SLC System B 3.1.7 BASES (continued)
APPLICABILITY    In MODES 1 and 2, shutdown capability is required. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. In MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate SDM (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") ensures that the reactor will not become critical. Therefore, the SLC System shutdown capability is not required to be OPERABLE when only a single control rod can be withdrawn.
In MODES 1, 2, and 3, the SLC System must be OPERABLE to ensure offsite doses remain within 10 CFR 50.67, Accident Source Term, limits following a LOCA involving significant fission product releases. The SLC System is used to maintain suppression pool pH above 7 following a LOCA to ensure iodine is retained in the suppression pool water.
ACTIONS          A.1 If one SLC subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE subsystem is adequate to perform the shutdown function. However, the overall reliability is reduced because a single failure in the remaining OPERABLE subsystem could result in reduced SLC System shutdown capability. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of performing the intended SLC System function and the low probability of a Design Basis Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive (CRD)
System to shut down the plant.
(continued)
BFN-UNIT 2                          B 3.1-50                        Revision 0, 29 January 25, 2005
 
SLC System B 3.1.7 BASES ACTIONS      B.1 (continued)
If both SLC subsystems are inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours. The allowed Completion Time of 8 hours is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor.
C.1 and C.2 If any Required Action and associated Completion Time is not met, 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.
SURVEILLANCE SR 3.1.7.1 REQUIREMENTS SR 3.1.7.1 verifies the volume of the borated solution in the storage tank, thereby ensuring SLC System OPERABILITY without disturbing normal plant operation. This Surveillance ensures that the proper borated solution volume is maintained for reactivity control and post-LOCA suppression pool pH control. The tank volume requirement of 4000 gallons is established by the amount of boron at 8.0% by weight concentration required for the radiological dose analysis for post-LOCA suppression pool pH control. The tank volume requirement for reactivity control is encompassed by the requirement for post LOCA pH control. For reactivity control, the sodium (continued)
BFN-UNIT 2                    B 3.1-51                Revision 0, 29, 113, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.1 (continued)
REQUIREMENTS pentaborate solution concentration requirements (d 9.2% by weight) and the required quantity of Boron-10 (t 203 lbs) establish the tank volume requirement. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.7.2 SR 3.1.7.2 verifies the continuity of the explosive charges in the injection valves to ensure that proper operation will occur if required. An automatic continuity monitor may be used to continuously satisfy this requirement. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.7.3 SR 3.1.7.3 requires an examination of sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank for post-LOCA suppression pool pH control. This parameter is used as input to determine the volume requirements for SR 3.1.7.1. The concentration is dependent upon the volume of water and quantity of boron in the storage tank.
SR 3.1.7.3 must be performed according to the Surveillance Frequency Control Program or within 24 hours of when boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits.
(continued)
BFN-UNIT 2                    B 3.1-52                  Revision 0, 29, 113, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.7 REQUIREMENTS (continued) Demonstrating that each SLC System pump develops a flow rate t 39 gpm at a discharge pressure t 1325 psig ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration and enrichment requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The pump flow rate of 39 gpm is based on the original licensing basis for SLC for an alternate reactivity insertion system.
SR 3.1.7.8 and SR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. Additionally, replacement charges shall be selected such that the age of charge in service shall not exceed five years from the manufacturer's assembly date.
(continued)
BFN-UNIT 2                    B 3.1-55                  Revision 29, 116, 123 Amendment No. 255 March 22, 2020
 
APLHGR B 3.2.1 BASES (continued)
ACTIONS          A.1 If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA and transient analyses may not be met. Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 2 hour Completion Time is sufficient to restore the APLHGR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification.
B.1 If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 23% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 23% RTP in an orderly manner and without challenging plant systems.
SURVEILLANCE      SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 12 hours after THERMAL POWER is t 23% RTP and periodically thereafter. They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER t 23% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                          B 3.2-4                    Revision 0, 113, 123 March 22, 2020
 
MCPR B 3.2.2 BASES (continued)
SURVEILLANCE      SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours after THERMAL POWER is t 23% RTP and then periodically thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER t 23% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.2.2.2 Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis. SR 3.2.2.2 determines the actual scram speed distribution and compares it with the assumed distribution. The MCPR operating limit is determined based either on the applicable limit associated with scram times of LCO 3.1.4, Control Rod Scram Times, or the nominal scram times. The scram speed-dependent MCPR limits are contained in the COLR. This determination must be performed within 72 hours after each set of control rod scram time tests required by SR 3.1.4.1 and SR 3.1.4.2 because the effective scram speed distribution may change during the cycle. The 72-hour Completion Time is acceptable due to the relatively minor changes in the actual control rod scram speed distribution expected during the fuel cycle.
(continued)
BFN-UNIT 2                          B 3.2-10                Revision 0, 31, 113, 123 March 22, 2020
 
LHGR B 3.2.3 BASES (continued)
SURVEILLANCE      SR 3.2.3.1 REQUIREMENTS The LHGR is required to be initially calculated within 12 hours after THERMAL POWER is t 23% RTP and then periodically thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER t 23% RTP is achieved is acceptable given the large inherent margin to operating limits at lower power levels. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Chapter 14.
: 2. FSAR, Chapter 3.
: 3. NUREG-0800, Standard Review Plan 4.2, Section II.A.2(g),
Revision 2, July 1981.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                          B 3.2-15                    Revision 0, 113, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      11. Manual Scram (continued)
SAFETY ANALYSES, LCO, and        Two channels of Manual Scram with one channel in each APPLICABILITY    manual scram trip system are available and required to be OPERABLE in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
: 12. RPS Channel Test Switches There are four RPS Channel Test Switches, one associated with each of the four automatic scram logic channels (A1, A2, B1, and B2). These keylock switches allow the operator to test the OPERABILITY of each individual logic channel without the necessity of using a scram function trip. When the RPS Channel Test Switch is placed in test, the associated scram logic channel is deenergized and OPERABILITY of the channel's scram contactors can be confirmed. The RPS Channel Test Switches are not specifically credited in the accident analysis. However, because the Manual Scram Function at Browns Ferry Nuclear Plant is not configured the same as the generic model in Reference 9, the RPS Channel Test Switches are included in the analysis in Reference 11.
(continued)
BFN-UNIT 2                        B 3.3-27                    Revision 0, 41, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.1 REQUIREMENTS (continued) Performance of the CHANNEL CHECK 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 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
(continued)
BFN-UNIT 2                    B 3.3-36                            Revision 0, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.2 REQUIREMENTS (continued) To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
A restriction to satisfying this SR when < 23% RTP is provided that requires the SR to be met only at t 23% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when
            < 23% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At t 23% RTP, the Surveillance is required to have been satisfactorily performed in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 23% if the Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 23% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
(continued)
BFN-UNIT 2                    B 3.3-37                  Revision 0, 113, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.3 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.
Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
As noted, SR 3.3.1.1.3 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.3-38                          Revision 0, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.5 and SR 3.3.1.1.6 (continued)
REQUIREMENTS If overlap for a group of channels is not demonstrated (e.g.,
IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel(s) declared inoperable. Only those appropriate channels that are required in the current MODE or condition should be declared inoperable.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.1.7 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.1.8 and SR 3.3.1.1.12 Deleted.
(continued)
BFN-UNIT 2                    B 3.3-40                          Revision 0, 123 March 22, 2020
 
SRM Instrumentation B 3.3.1.2 BASES BACKGROUND      During refueling, shutdown, and low power operations, the (continued)    primary indication of neutron flux levels is provided by the SRMs or special movable detectors connected to the normal SRM circuits. The SRMs provide monitoring of reactivity changes during fuel or control rod movement and give the control room operator early indication of subcritical multiplication that could be indicative of an approach to criticality.
APPLICABLE      Prevention and mitigation of prompt reactivity excursions during SAFETY ANALYSES refueling and low power operation is provided by LCO 3.9.1, "Refueling Equipment Interlocks;" LCO 3.1.1, "SHUTDOWN MARGIN (SDM)"; LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"; IRM Neutron Flux - High and Average Power Range Monitor (APRM) Neutron Flux - High, (Setdown)
Functions; and LCO 3.3.2.1, "Control Rod Block Instrumentation."
The SRMs have no safety function and are not assumed to function during any FSAR design basis accident or transient analysis. However, the SRMs provide the only on scale monitoring of neutron flux levels during startup and refueling.
Therefore, they are being retained in Technical Specifications.
LCO            During startup in MODE 2, three of the four SRM channels are required to be OPERABLE to monitor the reactor flux level prior to and during control rod withdrawal, subcritical multiplication and reactor criticality, and neutron flux level and reactor period until the flux level is sufficient to maintain the IRMs on Range 3 or above. All but one of the channels are required in order to provide a representation of the overall core response during those periods when reactivity changes are occurring throughout the core.
(continued)
BFN-UNIT 2                        B 3.3-47                                Revision 0
 
SRM Instrumentation B 3.3.1.2 BASES LCO          In MODES 3 and 4, with the reactor shut down, two SRM (continued) channels provide redundant monitoring of flux levels in the core.
In MODE 5, during a spiral offload or reload, an SRM outside the fueled region will no longer be required to be OPERABLE, since it is not capable of monitoring neutron flux in the fueled region of the core. Thus, CORE ALTERATIONS are allowed in a quadrant with no OPERABLE SRM in an adjacent quadrant provided the Table 3.3.1.2-1, footnote (b), requirement that the bundles being spiral reloaded or spiral offloaded are all in a single fueled region containing at least one OPERABLE SRM is met. Spiral reloading and offloading encompass reloading or offloading a cell on the edge of a continuous fueled region (the cell can be reloaded or offloaded in any sequence).
In nonspiral routine operations, two SRMs are required to be OPERABLE to provide redundant monitoring of reactivity changes occurring in the reactor core. Because of the local nature of reactivity changes during refueling, adequate coverage is provided by requiring one SRM to be OPERABLE in the quadrant of the reactor core where CORE ALTERATIONS are being performed, and the other SRM to be OPERABLE in an adjacent quadrant containing fuel. These requirements ensure that the reactivity of the core will be continuously monitored during CORE ALTERATIONS.
(continued)
BFN-UNIT 2                    B 3.3-48                              Revision 0
 
SRM Instrumentation B 3.3.1.2 BASES LCO          Special movable detectors, according to footnote (c) of (continued)  Table 3.3.1.2-1, may be used in place of the normal SRM nuclear detectors. These special detectors must be connected to the normal SRM circuits in the NMS, such that the applicable neutron flux indication can be generated. These special detectors provide more flexibility in monitoring reactivity changes during fuel loading, since they can be positioned anywhere within the core during refueling. They must still meet the location requirements of SR 3.3.1.2.2 and all other required SRs for SRMs.
For an SRM channel to be considered OPERABLE, it must be providing neutron flux monitoring indication.
APPLICABILITY The SRMs are required to be OPERABLE in MODES 2, 3, 4, and 5 prior to the IRMs being on scale on Range 3 to provide for neutron monitoring. In MODE 1, the APRMs provide adequate monitoring of reactivity changes in the core; therefore, the SRMs are not required. In MODE 2, with IRMs on Range 3 or above, the IRMs provide adequate monitoring and the SRMs are not required.
ACTIONS      A.1 and B.1 In MODE 2, with the IRMs on Range 2 or below, SRMs provide the means of monitoring core reactivity and criticality. With any number of the required SRMs inoperable, the ability to monitor neutron flux is degraded. Therefore, a limited time is allowed to restore the inoperable channels to OPERABLE status.
(continued)
BFN-UNIT 2                    B 3.3-49                                Revision 0
 
SRM Instrumentation B 3.3.1.2 BASES ACTIONS    A.1 and B.1 (continued)
Provided at least one SRM remains OPERABLE, Required Action A.1 allows 4 hours to restore the required SRMs to OPERABLE status. This time is reasonable because there is adequate capability remaining to monitor the core, there is limited risk of an event during this time, and there is sufficient time to take corrective actions to restore the required SRMs to OPERABLE status or to establish alternate IRM monitoring capability. During this time, control rod withdrawal and power increase is not precluded by this Required Action. Having the ability to monitor the core with at least one SRM, proceeding to IRM Range 3 or greater (with overlap required by SR 3.3.1.1.5),
and thereby exiting the Applicability of this LCO, is acceptable for ensuring adequate core monitoring and allowing continued operation.
With three required SRMs inoperable, Required Action B.1 allows no positive changes in reactivity (control rod withdrawal must be immediately suspended) due to inability to monitor the changes. Required Action A.1 still applies and allows 4 hours to restore monitoring capability prior to requiring control rod insertion. This allowance is based on the limited risk of an event during this time, provided that no control rod withdrawals are allowed, and the desire to concentrate efforts on repair, rather than to immediately shut down, with no SRMs OPERABLE.
(continued)
BFN-UNIT 2                  B 3.3-50                                Revision 0
 
SRM Instrumentation B 3.3.1.2 BASES ACTIONS      C.1 (continued)
In MODE 2, if the required number of SRMs is not restored to OPERABLE status within the allowed Completion Time, the reactor shall be placed in MODE 3. With all control rods fully inserted, the core is in its least reactive state with the most margin to criticality. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 in an orderly manner and without challenging plant systems.
D.1 and D.2 With one or more required SRMs inoperable in MODE 3 or 4, the neutron flux monitoring capability is degraded or nonexistent. The requirement to fully insert all insertable control rods ensures that the reactor will be at its minimum reactivity level while no neutron monitoring capability is available. Placing the reactor mode switch in the shutdown position prevents subsequent control rod withdrawal by maintaining a control rod block. The allowed Completion Time of 1 hour is sufficient to accomplish the Required Action, and takes into account the low probability of an event requiring the SRM occurring during this interval.
(continued)
BFN-UNIT 2                    B 3.3-51                                  Revision 0
 
SRM Instrumentation B 3.3.1.2 BASES ACTIONS      E.1 and E.2 (continued)
With one or more required SRM inoperable in MODE 5, the ability to detect local reactivity changes in the core during refueling is degraded. CORE ALTERATIONS must be immediately suspended and action must be immediately initiated to insert all insertable control rods in core cells containing one or more fuel assemblies. Suspending CORE ALTERATIONS prevents the two most probable causes of reactivity changes, fuel loading and control rod withdrawal, from occurring. Inserting all insertable control rods ensures that the reactor will be at its minimum reactivity given that fuel is present in the core. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative position.
Action (once required to be initiated) to insert control rods must continue until all insertable rods in core cells containing one or more fuel assemblies are inserted.
SURVEILLANCE As noted at the beginning of the SRs, the SRs for each SRM REQUIREMENTS Applicable MODE or other specified conditions are found in the SRs column of Table 3.3.1.2-1.
SR 3.3.1.2.1 Performance of the CHANNEL CHECK 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 another channel. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
(continued)
BFN-UNIT 2                      B 3.3-52                          Revision 0, 123 March 22, 2020
 
Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE      2. Rod Worth Minimizer (continued)
SAFETY ANALYSES, LCO, and        d 10% RTP. When THERMAL POWER is > 10% RTP, there is APPLICABILITY    no possible control rod configuration that results in a control rod (continued)    worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Refs. 5 and 7). In MODES 3 and 4, all control rods are required to be inserted into the core; therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.
: 3. Reactor Mode Switch - Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch -
Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.
The Reactor Mode Switch - Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).
Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required.
There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position.
(continued)
BFN-UNIT 2                        B 3.3-63        Revision 0, Amendment 303 November 19, 2009
 
Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE      3. Reactor Mode Switch - Shutdown Position (continued)
SAFETY ANALYSES, LCO, and        During shutdown conditions (MODE 3, 4, or 5), no positive APPLICABILITY    reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod withdrawal block is required to be OPERABLE. During MODE 5 with the reactor mode switch in the refueling position, the refuel position one-rod-out interlock (LCO 3.9.2, "Refuel Position One-Rod-Out Interlock")
provides the required control rod withdrawal blocks.
ACTIONS          A.1 With one RBM channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod block function; however, overall reliability is reduced because a single failure in the remaining OPERABLE channel can result in no control rod block capability for the RBM. For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status. The Completion Time of 24 hours is based on the low probability of an event occurring coincident with a failure in the remaining OPERABLE channel.
(continued)
BFN-UNIT 2                        B 3.3-64                                Revision 0
 
Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS      B.1 (continued)
If Required Action A.1 is not met and the associated Completion Time has expired, the inoperable channel must be placed in trip within 1 hour. If both RBM channels are inoperable, the RBM is not capable of performing its intended function; thus, one channel must also be placed in trip. This initiates a control rod withdrawal block, thereby ensuring that the RBM function is met.
The 1 hour Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or tripping of inoperable channels.
C.1, C.2.1.1, C.2.1.2, and C.2.2 With the RWM inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence.
However, the overall reliability is reduced because a single operator error can result in violating the control rod sequence.
Therefore, control rod movement must be immediately suspended except by scram. Alternatively, startup may continue if at least 12 control rods have already been withdrawn, or a reactor startup with an inoperable RWM during withdrawal of one or more of the first 12 rods was not performed in the last 12 months. These requirements minimize the number of reactor startups initiated with the RWM inoperable.
Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant logs and control room indications. Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily completed, control rod withdrawal may proceed in (continued)
BFN-UNIT 2                    B 3.3-65                                Revision 0
 
Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS    C.1, C.2.1.1, C.2.1.2, and C.2.2 (continued) accordance with the restrictions imposed by Required Action C.2.2. Required Action C.2.2 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff (e.g.,
a qualified shift technical advisor or reactor engineer).
The RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LCO 3.1.3 and LCO 3.1.6 may require bypassing the RWM, during which time the RWM must be considered inoperable with Condition C entered and its Required Actions taken.
D.1 With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence. Required Action D.1 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff. The RWM may be bypassed under these conditions to allow the reactor shutdown to continue.
(continued)
BFN-UNIT 2                    B 3.3-66                              Revision 0
 
Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS      E.1 and E.2 (continued)
With one Reactor Mode Switch - Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function. However, since the Required Actions are consistent with the normal action of an OPERABLE Reactor Mode Switch - Shutdown Position Function (i.e., maintaining all control rods inserted), there is no distinction between having one or two channels inoperable.
In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies will ensure that the core is subcritical with adequate SDM ensured by LCO 3.1.1. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
SURVEILLANCE As noted at the beginning of the SRs, the SRs for each Control REQUIREMENTS Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1.
The Surveillances are modified by a second Note (Note 2) to indicate that when an RBM 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 control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable (continued)
BFN-UNIT 2                    B 3.3-67                                Revision 0
 
Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.2 and SR 3.3.2.1.3 (continued)
REQUIREMENTS any control rod is withdrawn at d 10% RTP in MODE 2. As noted, SR 3.3.2.1.3 is not required to be performed until 1 hour after THERMAL POWER is reduced to d 10% RTP in MODE 1.
This allows entry into MODE 2 for SR 3.3.2.1.2, and THERMAL POWER reduction to d 10% RTP for SR 3.3.2.1.3, to perform the required Surveillance if the Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.2.1.4 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.
As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.7.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.3-69                              Revision 123 Amendment No. 255 March 22, 2020
 
Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES (continued)
APPLICABLE          The feedwater and main turbine high water level trip SAFETY ANALYSES    instrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref. 1). The reactor vessel high water level trip indirectly initiates a reactor scram from the main turbine trip (above 26% RTP) and trips the feedwater pumps, thereby terminating the event. The reactor scram mitigates the reduction in MCPR.
Feedwater and main turbine high water level trip instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref. 3).
LCO                The LCO requires two channels of the Reactor Vessel Water Level - High instrumentation per trip system to be OPERABLE to ensure that no single instrument failure will prevent the feedwater pump turbines and main turbine trip on a valid Reactor Vessel Water Level - High signal. Both channels in either trip system are needed to provide trip signals in order for the feedwater and main turbine trips to occur. Each channel must have its setpoint set within the specified Allowable Value of SR 3.3.2.2.3. The Allowable Value is set to ensure that the thermal limits are not exceeded during the event. The actual setpoint is calibrated to be consistent with the applicable setpoint methodology assumptions. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.
(continued)
BFN-UNIT 2                            B 3.3-75                            Revision 0, 113 October 24, 2018
 
Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES LCO            Trip setpoints are those predetermined values of output at (continued)    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 Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.
APPLICABILITY  The feedwater and main turbine high water level trip instrumentation is required to be OPERABLE at t 23% RTP to ensure that the fuel cladding integrity Safety Limit and the cladding 1% plastic strain limit are not violated during the feedwater controller failure, maximum demand event. As discussed in the Bases for LCO 3.2.1, "Average Planar Linear Heat Generation Rate (APLHGR)," LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," and LCO 3.2.3, LINEAR HEAT GENERATION RATE (LHGR), sufficient margin to these limits exists below 23% RTP; therefore, these requirements are only necessary when operating at or above this power level.
(continued)
BFN-UNIT 2                        B 3.3-76                      Revision 0, 31, 113 October 24, 2018
 
Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES (continued)
ACTIONS            A Note has been provided to modify the ACTIONS related to feedwater and main turbine high water level trip 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 feedwater and main turbine high water level trip 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 feedwater and main turbine high water level trip instrumentation channel.
A.1 With one or more channels inoperable in one trip system, the remaining two OPERABLE or in trip channels in the other trip system can provide the required trip signal. However, overall instrumentation reliability is reduced because a single failure in one of the two channels in the OPERABLE trip system concurrent with feedwater controller failure, maximum demand event, may result in the instrumentation not being able to perform its intended function. Therefore, continued operation is only allowed for a limited time with channels inoperable. If the inoperable channel(s) cannot be restored to OPERABLE status (continued)
BFN-UNIT 2                          B 3.3-77                                Revision 0
 
PAM Instrumentation B 3.3.3.1 BASES LCO          8. Suppression Pool Water Temperature (continued) (TI-64-161, TR-64-161, TI-64-162, and TR-64-162)
Suppression pool water temperature is a Category 1 variable provided to detect a condition that could potentially lead to containment breach and to verify the effectiveness of ECCS actions taken to prevent containment breach. The suppression pool water temperature instrumentation allows operators to detect trends in suppression pool water temperature in sufficient time to take action to prevent steam quenching vibrations in the suppression pool. Sixteen temperature sensors are arranged in two groups of two independent and redundant channels, located such that they are sufficient to provide a reasonable measure of bulk pool temperature. For a channel to be OPERABLE, at least 7 of its 8 sensors must be OPERABLE. The outputs for the sensors are recorded on two independent recorders in the control room. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channels.
: 9. Drywell Atmosphere Temperature (TI-64-52AB and XR-64-50)
Drywell atmosphere temperature is a Category 1 variable provided to detect a condition that could potentially lead to containment breach and to verify the effectiveness of ECCS actions taken to prevent containment breach. Two wide range drywell atmosphere temperature signals are transmitted from separate temperature sensors and are continuously recorded and displayed on one control room recorder and one control room indicator. The recorder and indicator are the primary indications used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
(continued)
BFN-UNIT 2                    B 3.3-90                                Revision 0
 
PAM Instrumentation B 3.3.3.1 BASES (continued)
APPLICABILITY    The PAM instrumentation LCO is applicable in MODES 1 and 2.
These variables are related to the diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1 and 2. In MODES 3, 4, and 5, plant conditions are such that the likelihood of an event that would require PAM instrumentation is extremely low; therefore, PAM instrumentation is not required to be OPERABLE in these MODES.
ACTIONS          Notes 1 and 2 have been provided to modify the ACTIONS related to PAM 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 PAM instrumentation channels provide appropriate compensatory measures for separate Functions. As such, Note 1 has been provided to allow separate Condition entry for each inoperable PAM Function. Note 2 has been provided for Function 6 to allow separate Condition entry for each penetration flow path.
(continued)
BFN-UNIT 2                        B 3.3-91                      Amendment No. 286 Revision 0 December 1, 2003
 
PAM Instrumentation B 3.3.3.1 BASES (continued)
REFERENCES        1. Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 3, May 1983.
: 2. TVA Letter from L. M. Mills to H. R. Denton (NRC) dated April 30, 1984.
: 3. NRC Letter from S. C. Black to S. A. White (TVA), NRC Regulatory Guide 1.97 SER letter, dated June 23, 1988.
: 4. TVA General Design Criteria No. BFN-50-7307, Revision 4, "Post-Accident Monitoring," dated June 22, 1993.
: 5. NRC Letter from Joseph F. Williams to Oliver D. Kingsley, Jr., "Regulatory Guide 1.97 - Boiling Water Reactor Neutron Flux Monitoring For the Browns Ferry Nuclear Plant, Units 1, 2, and 3," dated May 3, 1994.
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                        B 3.3-97                            Revision 0
 
Backup Control System B 3.3.3.2 B 3.3 INSTRUMENTATION B 3.3.3.2 Backup Control System BASES BACKGROUND          The Backup Control System provides the control room operator with sufficient instrumentation and controls to place and maintain the plant in a safe shutdown condition from a location other than the control room. This capability is necessary to protect against the possibility of the control room becoming inaccessible. A safe shutdown condition is defined as MODE 3.
With the plant in MODE 3, the Reactor Core Isolation Cooling (RCIC) System, the safety/relief valves, and the Residual Heat Removal System can be used to remove core decay heat and meet all safety requirements. The long term supply of water for the RCIC and the ability to operate the RHR System for decay heat removal from outside the control room allow extended operation in MODE 3.
In the event that the control room becomes inaccessible, the operators can establish control at the backup control panel and place and maintain the plant in MODE 3. Not all controls and necessary transfer switches are located at the backup control panel. Some controls and transfer switches will have to be operated locally at the switchgear, motor control panels, or other local stations. The plant automatically reaches MODE 3 following a plant shutdown and can be maintained safely in MODE 3 for an extended period of time.
The OPERABILITY of the Backup Control System control and instrumentation Functions ensures that there is sufficient information available on selected plant parameters to place and maintain the plant in MODE 3 should the control room become inaccessible.
(continued)
BFN-UNIT 2                            B 3.3-98                              Revision 0
 
EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE      Turbine Stop Valve - Closure (continued)
SAFETY ANALYSES, LCO, and        Closure of the TSVs is determined by measuring the position of APPLICABILITY    each valve. There are two separate position signals associated with each stop valve, the signal from each switch being assigned to a separate trip channel. The logic for the TSV -
Closure Function is such that two or more TSVs must be closed to produce an EOC-RPT. This Function must be enabled at THERMAL POWER t 26% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this function. To consider this function OPERABLE, bypass of the function must not occur when bypass valves are open. Four channels of TSV - Closure, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TSV -
Closure Allowable Value is selected to detect imminent TSV closure.
This protection is required, consistent with the safety analysis assumptions, whenever THERMAL POWER is t 26% RTP.
Below 26% RTP, the Reactor Vessel Steam Dome Pressure -
High and the Average Power Range Monitor (APRM) Fixed Neutron Flux - High Functions of the Reactor Protection System (RPS) are adequate to maintain the necessary margin to the MCPR SL and LHGR limits.
(continued)
BFN-UNIT 2                      B 3.3-112                  Revision 0, 31, 61, 113 October 24, 2018
 
EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE      Turbine Control Valve Fast Closure, Trip Oil Pressure - Low SAFETY ANALYSES, (PS-47-142, PS-47-144, PS-47-146, and PS-47-148)
LCO, and APPLICABILITY    Fast closure of the TCVs during a generator load rejection (continued)    results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.
Therefore, an RPT is initiated on TCV Fast Closure, Trip Oil Pressure - Low in anticipation of the transients that would result from the closure of these valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL and LHGR limits are not exceeded during the worst case transient.
Fast closure of the TCVs is determined by measuring the electrohydraulic control fluid pressure at each control valve.
There is one pressure switch associated with each control valve, and the signal from each switch is assigned to a separate trip channel. The logic for the TCV Fast Closure, Trip Oil Pressure - Low Function is such that two or more TCVs must be closed (pressure switch trips) to produce an EOC-RPT. This Function must be enabled at THERMAL POWER t 26% RTP.
This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this function. To consider this function OPERABLE, bypass of the function must not occur when bypass valves are open. Four channels of TCV Fast Closure, Trip Oil Pressure - Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TCV Fast Closure, Trip Oil Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure.
(continued)
BFN-UNIT 2                        B 3.3-113                      Revision 0, 31, 113 October 24, 2018
 
EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE      Turbine Control Valve Fast Closure, Trip Oil Pressure - Low SAFETY ANALYSES, (PS-47-142, PS-47-144, PS-47-146, and PS-47-148)
LCO, and        (continued)
APPLICABILITY This protection is required consistent with the safety analysis whenever THERMAL POWER is t 26% RTP. Below 26% RTP, the Reactor Vessel Steam Dome Pressure - High and the APRM Fixed Neutron Flux - High Functions of the RPS are adequate to maintain the necessary safety margins.
ACTIONS          A Note has been provided to modify the ACTIONS related to EOC-RPT 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 EOC-RPT 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 EOC-RPT instrumentation channel.
(continued)
BFN-UNIT 2                        B 3.3-114                            Revision 0, 113 October 24, 2018
 
EOC-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.2 REQUIREMENTS (continued) This SR ensures that an EOC-RPT initiated from the TSV -
Closure and TCV Fast Closure, Trip Oil Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is t 26% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at t 26% RTP, either due to open main turbine bypass valves or other reasons), the affected TSV - Closure and TCV Fast Closure, Trip Oil Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met with the channel considered OPERABLE.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.4.1.3 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                  B 3.3-118                      Revision 113, 123 Amendment No. 255 March 22, 2020
 
ATWS-RPT Instrumentation B 3.3.4.2 BASES APPLICABLE      The specific Applicable Safety Analyses and LCO discussions SAFETY ANALYSES, are listed below on a Function by Function basis.
LCO, and APPLICABILITY    a. Reactor Vessel Water Level - Low Low, Level 2 (continued)        (LS-3-58A1, LS-3-58B1, LS-3-58C1, and LS-3-58D1)
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, the ATWS-RPT System is initiated at Level 2 to aid in maintaining level above the top of the active fuel. The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.
Reactor vessel water level signals are initiated from four 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 Low, Level 2 with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Vessel Water Level - Low Low, Level 2 Allowable Value is chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation.
(continued)
BFN-UNIT 2                      B 3.3-123                                Revision 0
 
ATWS-RPT Instrumentation B 3.3.4.2 BASES APPLICABLE      b. Reactor Steam Dome Pressure - High SAFETY ANALYSES,    (PIS-3-204A, PIS-3-204B, PIS-3-204C, and PIS-3-204D)
LCO, and APPLICABILITY      Excessively high RPV pressure may rupture the RCPB. An (continued)        increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Steam Dome Pressure - High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety/relief valves, limits the peak RPV pressure to less than the ASME Section III Code limits.
The Reactor Steam Dome Pressure - High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure. Four channels of Reactor Steam Dome Pressure - High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Steam Dome Pressure - High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code limits.
(continued)
BFN-UNIT 2                      B 3.3-124                              Revision 0
 
ATWS-RPT Instrumentation B 3.3.4.2 BASES (continued)
ACTIONS          A Note has been provided to modify the ACTIONS related to ATWS-RPT 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 ATWS-RPT 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 ATWS-RPT instrumentation channel.
A.1 and A.2 With one or more channels inoperable, but with ATWS-RPT capability for each Function maintained (refer to Required Actions B.1 and C.1 Bases), the ATWS-RPT System is capable of performing the intended function. However, the reliability and redundancy of the ATWS-RPT instrumentation is reduced, such that a single failure in the remaining trip system could result in the inability of the ATWS-RPT System to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to OPERABLE status. Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of ATWS-RPT, 14 days is provided to restore the inoperable channel (Required Action A.1). Alternately, the inoperable channel may be placed in trip (Required Action A.2),
since this would conservatively (continued)
BFN-UNIT 2                          B 3.3-125                                Revision 0
 
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS    E.1 and E.2 (continued)
For Required Action E.1, the Completion Time only begins upon discovery that a redundant feature in the same system (e.g.,
both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.
If the instrumentation that controls the CS pump minimum flow valve is inoperable, such that the valve will not automatically open, extended CS pump operation with no injection path available could lead to pump overheating and failure. If there were a failure of the instrumentation, such that the valve would not automatically close, a portion of the pump flow could be diverted from the reactor vessel injection path, causing insufficient core cooling. These consequences can be averted by the operator's manual control of the valve, which would be adequate to maintain ECCS pump protection and required flow.
Furthermore, other ECCS pumps would be sufficient to complete the assumed safety function if no additional single failure were to occur. The 7 day Completion Time of Required Action E.2 to restore the inoperable channel to OPERABLE status is reasonable based on the remaining capability of the associated ECCS subsystems, the redundancy available in the ECCS design, and the low probability of a DBA occurring during the allowed out of service time. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel in all events.
(continued)
BFN-UNIT 2                  B 3.3-170                                Revision 0
 
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS      F.1 and F.2 (continued)
Required Action F.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within similar ADS trip system A and B Functions result in redundant automatic initiation capability being lost for the ADS.
Redundant automatic initiation capability is lost if either (a) one or more Function 4.a channels and one or more Function 5.a channels are inoperable and untripped, (b) one or more Function 4.b channels and one or more Function 5.b channels are inoperable and untripped, or (c) one or more Function 4.d channels and one or more Function 5.d channels are inoperable and untripped.
In this situation (loss of automatic initiation capability), the 96 hour or 8 day allowance, as applicable, of Required Action F.2 is not appropriate and all ADS valves must be declared inoperable within 1 hour after discovery of loss of ADS initiation capability.
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Required Action F.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable, untripped channels within similar ADS trip system Functions as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
(continued)
BFN-UNIT 2                    B 3.3-171                                Revision 0
 
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS    F.1 and F.2 (continued)
Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable channel to OPERABLE status if both HPCI and RCIC are OPERABLE. If either HPCI or RCIC is inoperable, the time is shortened to 96 hours. If the status of HPCI or RCIC changes such that the Completion Time changes from 8 days to 96 hours, the 96 hours begins upon discovery of HPCI or RCIC inoperability.
However, the total time for an inoperable, untripped channel cannot exceed 8 days. If the status of HPCI or RCIC changes such that the Completion Time changes from 96 hours to 8 days, the "time zero" for beginning the 8 day "clock" begins upon discovery of the inoperable, untripped channel. 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 F.2. 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 initiation), Condition H must be entered and its Required Action taken.
(continued)
BFN-UNIT 2                  B 3.3-172                                Revision 0
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTION    B.1 (continued) channels OPERABLE or in trip (combined total channels of at least eight for HPCI and eight for RCIC). The required channels would be one channel inputting to trip channel 1 of either the A or B Logic and one channel inputting to trip channel 2 of either the A or B Logic. For Function 5.g, this would require the SLC System initiation switch to be capable of generating an isolation signal to at least one of the RWCU isolation valves.
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.
The second Completion Time for Function 1.d when normal ventilation is not available is provided to allow the plant to avoid an MSL isolation transient when recovering from a temporary loss of ventilation in the MSL tunnel area (e.g., during performance of the secondary containment leak rate tests). As allowed by LCO 3.0.2 (and discussed in the Bases for LCO 3.0.2), the plant may intentionally enter this condition to avoid an MSL isolation transient and bypass the high temperature channels during restoration of ventilation flow.
However, during the period that multiple Main Steam Tunnel Temperature - High Function channels are inoperable due to this intentional action, an additional compensatory measure is deemed necessary and shall be taken: an operator shall observe control room indications of the affected space temperatures for indications of small steam leaks. In the event of rapid increases in temperature (indicative of a steam line break), the operator shall promptly close the MSIVs. The 4 hour Completion Time is acceptable because along with the compensatory measures described above it minimizes risk while allowing time for restoration or tripping of channels.
(continued)
BFN-UNIT 2                  B 3.3-217                                Revision 0
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS      C.1 (continued)
Required Action C.1 directs entry into the appropriate Condition referenced in Table 3.3.6.1-1. The applicable Condition specified in Table 3.3.6.1-1 is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A or B and the associated Completion Time has expired, Condition C will be entered for that channel and provides for transfer to the appropriate subsequent Condition.
D.1, D.2.1, and D.2.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours (Required Actions D.2.1 and D.2.2). Alternately, the associated MSLs may be isolated (Required Action D.1), and, if allowed (i.e., plant safety analysis allows operation with an MSL isolated), operation with that MSL isolated may continue.
Isolating the affected MSL accomplishes the safety function of the inoperable channel. The 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.
(continued)
BFN-UNIT 2                    B 3.3-218                              Revision 0
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS      E.1 (continued)
If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant in at least MODE 2 within 6 hours.
The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems.
F.1 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, plant operations may continue if the affected penetration flow path(s) is isolated.
Isolating the affected penetration flow path(s) accomplishes the safety function of the inoperable channels.
For the RWCU Area Temperature - High Functions, the affected penetration flow path(s) may be considered isolated by isolating only that portion of the system in the associated room monitored by the inoperable channel. That is, if the RWCU pump room A area channel is inoperable, the pump room A area can be isolated while allowing continued RWCU operation utilizing the B RWCU pump.
Alternately, if it is not desired to isolate the affected penetration flow path(s) (e.g., as in the case where isolating the penetration flow path(s) could result in a reactor scram), Condition G must be entered and its Required Actions taken.
(continued)
BFN-UNIT 2                      B 3.3-219                                Revision 0
 
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE      3, 4. Reactor Zone Exhaust and Refueling Floor Radiation -
SAFETY ANALYSES, High (RM-90-140, 141, 142, 143) (continued)
LCO, and APPLICABILITY    ventilation exhaust both of which must be OPERABLE or tripped for the channel to be OPERABLE. Both radiation elements must provide a High signal to trip the associated channel (two-out-of-two). However, the output relays from the divisional trip systems are arranged in logic systems such that if either channel for a zone trips, a secondary containment isolation signal is initiated (one-out-of-two). Two channels of Reactor Zone Exhaust Radiation - High Function and two channels of Refueling Floor 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 provide timely detection of nuclear system process barrier leaks inside containment but are far enough above background levels to avoid spurious isolation.
The Reactor Zone Exhaust and Refueling Floor Radiation -
High Functions are 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, these Functions are not required. In addition, the Functions are also required to be OPERABLE during OPDRVs because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery) must be provided to ensure that offsite dose limits are not exceeded.
(continued)
BFN-UNIT 2                        B 3.3-232            Revision 0, 21, 29, 35, 122 December 21, 2020
 
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES (continued)
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 Condition. However, the Required Actions for inoperable 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 Functions 1 and 2, and 24 hours for Functions other than Functions 1 and 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 (continued)
BFN-UNIT 2                          B 3.3-233                                Revision 0
 
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS    A.1 (continued) 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 secondary containment penetration flow path(s) or a complete loss of automatic initiation capability for the SGT System. A Function is considered to be maintaining secondary containment isolation capability when sufficient channels are OPERABLE or in trip, such that at least one of the two SCIVs in the associated penetration flow path(s) and two SGT subsystems can be initiated on an isolation signal from the given Function. For Functions 1 and 2, this would require both PCIS trip systems to have at least one channel of the Function OPERABLE or in trip.
For Functions 3 and 4, this would require each unit to have at least one channel of the Function OPERABLE or in trip.
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.
(continued)
BFN-UNIT 2                  B 3.3-234                                Revision 0
 
LOP Instrumentation B 3.3.8.1 BASES (continued)
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.
A.1 and A.2 With one of the degraded voltage relay channels inoperable on one or more shutdown boards and with the loss of voltage relay channels on the affected shutdown board(s) OPERABLE, Required Action A.2 provides a 15 day allowable out of service time to restore the relay channel to OPERABLE status provided the other two degraded voltage relay channels and associated timers are OPERABLE. Immediate verification of the OPERABILITY of the other degraded voltage relay channels and associated timers is therefore required (Required Action A.1). This may be performed as an administrative check by examining logs or other information to determine if this equipment is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate OPERABILITY of this equipment. If the OPERABILITY of this equipment cannot be verified, however, (continued)
BFN-UNIT 2                          B 3.3-261                              Revision 0
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS    A.1 and A.2 (continued)
Condition C or D, as applicable, must be entered immediately.
The 15 day allowable out of service time is justified based on the two-out-of-three permissive logic scheme provided for these relays. If the inoperable relay channel cannot be restored to OPERABLE status within the allowable out of service time, the degraded voltage relay channel must be placed in the tripped condition per Required Action A.2. 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 F must be entered and its Required Action taken.
B.1 With two or more degraded voltage relay channels or one or more associated timers inoperable on one or more shutdown boards, the Function is not capable of performing the intended function. Required Action B.1 provides a 10 day allowable out of service time provided the loss of voltage relay channels on the affected shutdown board(s) are OPERABLE.
The 10 day allowable out of service time is justified since the loss of voltage relay channels on the same shutdown board are independent of the degraded voltage relay channel(s) and will continue to function and start the diesel generators on a complete loss of voltage. If the inoperable channel(s) cannot (continued)
BFN-UNIT 2                    B 3.3-262                          Revision 0, 118 December 12, 2019
 
RPS Electric Power Monitoring B 3.3.8.2 BASES (continued)
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 contactor. 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 bus powered components. Each inservice electric power monitoring assembly's trip logic setpoints are required to be within the specified Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint procedures (nominal trip setpoint).
Allowable Values are specified for each RPS electric power monitoring assembly trip logic (refer to SR 3.3.8.2.2). Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected based on engineering judgment and operational experience to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint 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 the process parameter exceeds the setpoint, the associated device (e.g., trip relay) changes state.
(continued)
BFN-UNIT 2                        B 3.3-271                                Revision 0
 
RPS Electric Power Monitoring B 3.3.8.2 BASES LCO          The Allowable Values for the instrument settings are based on (continued)  the RPS continuously providing t 56 Hz, 120 V +/- 10% (to all equipment), and 115 V +/- 10 V (to scram and MSIV solenoids).
The most limiting voltage requirement and associated line losses determine the settings of the electric power monitoring instrument channels. The settings are calculated based on the loads on the buses and RPS MG set or alternate power supply being 120 VAC and 60 Hz.
APPLICABILITY The operation of the RPS electric power monitoring assemblies is essential to disconnect the RPS bus powered components from the MG set or alternate power supply during abnormal voltage or frequency conditions. Since the degradation of a nonclass 1E 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 bus powered components are required to be OPERABLE. This results in the RPS Electric Power Monitoring System OPERABILITY being required in MODES 1, 2, and 3; and in MODES 4 and 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies (a control rod withdrawn in MODE 4 is only allowed by Special Operations LCO 3.10.4, "Single Control Rod Withdrawal - Cold Shutdown").
(continued)
BFN-UNIT 2                    B 3.3-272                            Revision 0
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
SURVEILLANCE      SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 70% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is
                  < 70% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.
The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered inoperable. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                          B 3.4-9                            Revision 123 Amendment No. 258 March 22, 2020
 
Jet Pumps B 3.4.2 BASES SURVEILLANCE SR 3.4.2.1 (continued)
REQUIREMENTS Individual jet pumps in a recirculation loop normally do not have the same flow. The unequal flow is due to the drive flow manifold, which does not distribute flow equally to all risers.
The flow (or jet pump diffuser to lower plenum differential pressure) pattern or relationship of one jet pump to the loop average is repeatable. An appreciable change in this relationship is an indication that increased (or reduced) resistance has occurred in one of the jet pumps. This may be indicated by an increase in the relative flow for a jet pump that has experienced beam cracks.
The deviations from normal are considered indicative of a potential problem in the recirculation drive flow or jet pump system (Ref. 2). Normal flow ranges and established jet pump flow and differential pressure patterns are established by plotting historical data as discussed in Reference 2.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by two Notes. Note 1 allows this Surveillance not to be performed until 4 hours after the associated recirculation loop is in operation, since these checks can only be performed during jet pump operation. The 4 hours is an acceptable time to establish conditions appropriate for data collection and evaluation.
(continued)
BFN-UNIT 2                    B 3.4-15                          Revision 0, 123 March 22, 2020
 
S/RVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 4.4.6.
: 2. FSAR, Section 14.5.1.
: 3. ASME Code for Operation and Maintenance for Nuclear Power Plants (ASME OM Code).
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                  B 3.4-22                    Revision 0, 81, 123 Amendment No. 255 March 22, 2020
 
RCS Operational LEAKAGE B 3.4.4 BASES ACTIONS      C.1 and C.2 (continued)
If any Required Action and associated Completion Time of Condition A or B is not met or if pressure boundary LEAKAGE exists, 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 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 safety systems.
SURVEILLANCE SR 3.4.4.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE. Leakage detection instrumentation is discussed in more detail in the Bases for LCO 3.4.5, "RCS Leakage Detection Instrumentation." Sump level and flow rate are typically monitored to determine actual LEAKAGE rates; however, other methods may be used to quantify LEAKAGE. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.4-28                        Revision 0, 123 March 22, 2020
 
RCS Operational LEAKAGE B 3.4.4 BASES (continued)
REFERENCES        1. 10 CFR 50.2.
: 2. 10 CFR 50.55a(c).
: 3. 10 CFR 50, Appendix A, GDC 55.
: 4. GEAP-5620, "Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws," April 1968.
: 5. NUREG-75/067, "Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping in Boiling Water Reactors,"
October 1975.
: 6. FSAR, Section 4.10.3.2.
: 7. Deleted.
: 8. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                        B 3.4-29                        Revision 0, 123 March 22, 2020
 
RCS Leakage Detection Instrumentation B 3.4.5 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.5 RCS Leakage Detection Instrumentation BASES BACKGROUND          GDC 30 of 10 CFR 50, Appendix A (Ref. 1), requires means for detecting and, to the extent practical, identifying the location of the source of RCS LEAKAGE.
Limits on LEAKAGE from the reactor coolant pressure boundary (RCPB) are required so that appropriate action can be taken before the integrity of the RCPB is impaired. Leakage detection systems for the RCS are provided to alert the operators when leakage rates above normal background levels are detected and also to supply quantitative measurement of leakage rates. The Bases for LCO 3.4.4, "RCS Operational LEAKAGE," discuss the limits on RCS LEAKAGE rates.
Systems for separating the LEAKAGE of an identified source from an unidentified source are necessary to provide prompt and quantitative information to the operators to permit them to take immediate corrective action.
LEAKAGE from the RCPB inside the drywell is detected by at least one of two or three independently monitored variables, such as sump level changes and drywell gaseous and particulate radioactivity levels. The primary means of quantifying LEAKAGE in the drywell is the drywell floor drain sump monitoring system.
(continued)
BFN-UNIT 2                          B 3.4-30                                Revision 0
 
RCS Leakage Detection Instrumentation B 3.4.5 BASES BACKGROUND  The drywell floor drain sump monitoring system monitors the (continued) LEAKAGE collected in the floor drain sump. This unidentified LEAKAGE consists of LEAKAGE from control rod drives, valve flanges or packings, floor drains, the Reactor Building Closed Cooling Water System, and drywell air cooling unit condensate drains, and any LEAKAGE not collected in the drywell equipment drain sump. The drywell floor drain sump has transmitters that supply level indications locally.
The floor drain sump level indicators have switches that start and stop the sump pumps when required. A timer starts each time the sump is pumped down to the low level setpoint. If the sump fills to the high level setpoint before the timer ends, an alarm sounds in the control room, indicating a LEAKAGE rate into the sump in excess of a preset limit.
A flow transmitter in the discharge line of the drywell floor drain sump pumps provides flow indication in the control room. The pumps can also be started from the control room.
The primary containment air monitoring systems continuously monitor the primary containment atmosphere for airborne particulate and gaseous radioactivity. A sudden increase of radioactivity, which may be attributed to RCPB steam or reactor water LEAKAGE, is annunciated in the control room. The primary containment atmosphere particulate and gaseous radioactivity monitoring systems are not capable of quantifying LEAKAGE rates, but are sensitive enough to indicate increased LEAKAGE rates. This system is capable of detecting radiation levels in containment atmosphere of three times background (Ref. 2).
(continued)
BFN-UNIT 2                    B 3.4-31                                Revision 0
 
RCS Leakage Detection Instrumentation B 3.4.5 BASES SURVEILLANCE SR 3.4.5.4 REQUIREMENTS (continued) This SR is for the performance of a CHANNEL CALIBRATION of required leakage detection system instrumentation channels.
The calibration verifies the accuracy of the instrument string.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 30.
: 2. FSAR, Section 4.10.3.
: 3. GEAP-5620, "Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws," April 1968.
: 4. NUREG-75/067, "Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping in Boiling Water Reactors,"
October 1975.
: 5. FSAR, Section 4.10.3.2.
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                    B 3.4-36                            Revision 123 Amendment No. 255 March 22, 2020
 
RHR Shutdown Cooling System - Hot Shutdown B 3.4.7 BASES (continued)
APPLICABLE        Decay heat removal by operation of the RHR System in the SAFETY ANALYSES  shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result. The RHR Shutdown Cooling System meets Criterion 4 of the NRC Policy Statement (Ref. 1).
LCO              Two RHR shutdown cooling subsystems are required to be OPERABLE, and when no recirculation pump is in operation, one RHR shutdown cooling subsystem must be in operation.
An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, one RHRSW pump capable of providing cooling to the heat exchanger, and the associated piping and valves. The subsystems have a common suction source and are allowed to have common discharge piping. Since the piping is a passive component that is assumed not to fail, it is allowed to be common to the subsystems. Each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. In MODE 3, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. Operation of one subsystem can maintain or reduce the reactor coolant temperature as required.
However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required.
(continued)
BFN-UNIT 2                        B 3.4-43                              Revision 0
 
RHR Shutdown Cooling System - Cold Shutdown B 3.4.8 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown BASES BACKGROUND            Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant. This decay heat must be removed to maintain the temperature of the reactor coolant d 212&deg;F. This decay heat removal is in preparation for performing refueling or maintenance operations, or for keeping the reactor in the Cold Shutdown condition.
The RHR System has two loops with each loop consisting of two motor driven pumps, two heat exchangers, and associated piping and valves. There are two shutdown cooling subsystems per RHR System loop. Both loops have a common suction from the same recirculation loop. The four redundant, manually controlled shutdown cooling subsystems of the RHR System provide decay heat removal. Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via the associated recirculation loop. The RHR heat exchangers transfer heat to the RHR Service Water System. Any one of the four RHR shutdown cooling subsystems can provide the required decay heat removal function.
(continued)
BFN-UNIT 2                              B 3.4-49                              Revision 0
 
RCS P/T Limits B 3.4.9 BASES BACKGROUND      The criticality limits include the Reference 1 requirement that (continued)    they be at least 40&deg;F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.
The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. ASME Code, Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
APPLICABLE      The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed. Reference 7 establishes the methodology for determining the P/T limits. Since the P/T limits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 9).
(continued)
BFN-UNIT 2                        B 3.4-57                      Revision 0, 15, 26 March 17, 2004
 
RCS P/T Limits B 3.4.9 BASES (continued)
ACTIONS          A.1 and A.2 Operation outside the P/T limits while in MODE 1, 2, or 3 must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.
The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity remains acceptable and must be completed if continued operation is desired.
Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
ASME Code, Section XI, Appendix E (Ref. 6), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.
The 72 hour Completion Time is reasonable to accomplish the evaluation of a mild violation. More severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed if continued operation is desired.
Condition A is modified by a Note requiring Required Action A.2 be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
(continued)
BFN-UNIT 2                          B 3.4-60                              Revision 0
 
RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7 REQUIREMENTS (continued) Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.
The flange temperatures must be verified to be above the limits before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied.
When in MODE 4 with RCS temperature d 85&deg;F, checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature d 100&deg;F, monitoring of the flange temperature is required to ensure the temperature is > 83qF.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.9.5 is modified by two Notes. Note 1 requires the Surveillance to be performed only when tensioning the reactor vessel head bolting studs. Note 2 allows the reactor vessel head bolts to be partially tensioned (four sequences of the seating pass) provided the studs and flange materials are
            > 70qF. SR 3.4.9.6 is modified by a Note that requires the (continued)
BFN-UNIT 2                    B 3.4-65                    Revision 0, 26, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES ACTIONS      G.1 and G.2 (continued)
If any Required Action and associated Completion Time of Condition C, D, E, or F is not met, or if two or more ADS valves are inoperable, the plant must be brought to a condition 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 reactor steam dome pressure reduced to d 150 psig 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.
H.1 When multiple ECCS subsystems are inoperable, as stated in Condition H, the plant is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCI System, CS System, and LPCI subsystems full of water ensures that the ECCS will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.5-12                          Revision 0, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.2 REQUIREMENTS (continued) Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the HPCI System, this SR also includes the steam flow path for the turbine and the flow controller position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the RHR low pressure permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. This allows operation in the RHR shutdown cooling mode during MODE 3, if necessary.
(continued)
BFN-UNIT 2                    B 3.5-13                      Revision 0, 109, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.3 REQUIREMENTS (continued) Verification that ADS air supply header pressure is t 81 psig ensures adequate air pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least two valve actuations can occur with the drywell at 62.5% of design pressure plus three additional actuations at 0 psig drywell pressure (Ref. 10). The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of t 81 psig is provided by the Drywell Control Air System. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.1.4 Verification that the LPCI cross tie valve is closed and power to its operator is disconnected ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI subsystem. Acceptable methods of removing power to the operator include de-energizing breaker control power, racking out or removing the breaker, or disconnecting the motor leads.
If the LPCI cross tie valve is open or power has not been removed from the valve operator, both LPCI subsystems must be considered inoperable. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.5-14                          Revision 0, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.6, SR 3.5.1.7, and SR 3.5.1.8 (continued)
REQUIREMENTS pressure and flow are achieved to perform these tests. Reactor startup is allowed prior to performing the low pressure Surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance test is short. Alternately, the low pressure Surveillance test may be performed prior to startup using an auxiliary steam supply. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure test has been satisfactorily completed and there is no indication or reason to believe that HPCI is inoperable.
Therefore, SR 3.5.1.7 and SR 3.5.1.8 are modified by Notes that state the Surveillances are not required to be performed until 12 hours after the reactor steam pressure and flow are adequate to perform the test.
The Frequency for SR 3.5.1.6 is in accordance with the INSERVICE TESTING PROGRAM requirements. The Frequencies for SR 3.5.1.7 and SR 3.5.1.8 are controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.5-17                      Revision 109, 123 Amendment No. 255 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.9 REQUIREMENTS (continued) The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This SR also ensures that the HPCI System will automatically restart on an RPV low-low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool.
The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
(continued)
BFN-UNIT 2                    B 3.5-18                              Revision 123 Amendment No. 255 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.10 REQUIREMENTS (continued) The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e., solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that excludes valve actuation.
This prevents an RPV pressure blowdown.
(continued)
BFN-UNIT 2                    B 3.5-19                            Revision 123 Amendment No. 255 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.11 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.1.12 (deleted)
(continued)
BFN-UNIT 2                  B 3.5-21                        Revision 76, 123 Amendment No. 255 March 22, 2020
 
RCIC System B 3.5.3 BASES (continued)
SURVEILLANCE      SR 3.5.3.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge line of the RCIC System full of water ensures that the system will perform properly, injecting its full capacity into the Reactor Coolant System upon demand. This will also prevent a water hammer following an initiation signal. One acceptable method of ensuring the line is full is to vent at the high points. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.3.2 Verifying the correct alignment for manual, power operated, and automatic valves in the RCIC flow path provides assurance that the proper flow path will exist for RCIC operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are (continued)
BFN-UNIT 2                          B 3.5-34                            Revision 0, 123 March 22, 2020
 
RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.2 (continued)
REQUIREMENTS in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the RCIC System, this SR also includes the steam flow path for the turbine and the flow controller position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.3.3 and SR 3.5.3.4 The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated. The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Reactor steam pressure must be t 950 psig to perform SR 3.5.3.3 and t 150 psig to perform SR 3.5.3.4. Adequate steam flow is represented by at least one turbine bypass valve full open for SR 3.5.3.3 and at least one turbine bypass valve > 50% open for SR 3.5.3.4. Therefore, sufficient time is allowed (continued)
BFN-UNIT 2                    B 3.5-35                    Revision 53, 109, 123 Amendment No. 254 March 22, 2020
 
RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.3 and SR 3.5.3.4 (continued)
REQUIREMENTS after adequate pressure and flow are achieved to perform these SRs. Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance is short. Alternately, the low pressure Surveillance test may be performed prior to startup using an auxiliary steam supply. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure Surveillance has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable. Therefore, these SRs are modified by Notes that state the Surveillances are not required to be performed until 12 hours after the reactor steam pressure and flow are adequate to perform the test.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                  B 3.5-36                        Revision 109, 123 Amendment No. 255 March 22, 2020
 
RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.5 REQUIREMENTS (continued) The RCIC System is required to actuate automatically in order to perform its design function satisfactorily. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of the RCIC System will cause the system to operate as designed, including actuation of the system throughout its emergency operating sequence; that is, automatic pump startup and actuation of all automatic valves to their required positions. This test also ensures the RCIC System will automatically restart on an RPV low-low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.2 overlaps this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that excludes vessel injection during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
(continued)
BFN-UNIT 2                    B 3.5-37                              Revision 123 Amendment No. 255 March 22, 2020
 
Primary Containment B 3.6.1.1 BASES SURVEILLANCE SR 3.6.1.1.2 (continued)
REQUIREMENTS Satisfactory performance of this SR can be achieved by establishing a known differential pressure between the drywell and the suppression chamber and verifying that the pressure in either the suppression chamber or the drywell does not change by more than 0.25 inch of water per minute over a 10 minute period. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. NEDC-33860P, Safety Analysis Report for Browns Ferry Nuclear Plant nits 1, 2, and 3 Extended Power Uprate, Section 2.6.
: 2. FSAR, Section 14.6.
: 3. 10 CFR 50, Appendix J, Option B.
: 4. NEI 94-01, Revision 3A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J."
: 5. ANSI/ANS-56.8-2002, "American National Standard for Containment System Leakage Testing Requirements."
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                    B 3.6-6                Revision 113, 114, 123 Amendment No. 255 March 22, 2020
 
Primary Containment Air Lock B 3.6.1.2 BASES SURVEILLANCE SR 3.6.1.2.1 (continued)
REQUIREMENTS The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 requires the results of airlock leakage tests be evaluated against the acceptance criteria of the Primary Containment Leakage Rate Testing Program, 5.5.12. This ensures that the airlock leakage is properly accounted for in determining the combined Type B and C primary containment leakage.
SR 3.6.1.2.2 The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment pressure, closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment.
Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.6-16                          Revision 0, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.1 (continued)
REQUIREMENTS following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.2 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured, and is required to be closed during accident conditions is closed.
The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.
This SR does not require any testing or valve manipulation.
Rather, it involves verification that those PCIVs outside primary containment, and capable of being mispositioned, are in the correct position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.6-30                          Revision 0, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.4 REQUIREMENTS (continued) The traversing incore probe (TIP) shear isolation valves are actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that TIP valves will actuate when required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.5 Verifying the isolation time of each power operated, automatic PCIV is within limits is required to demonstrate OPERABILITY.
MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.6. The isolation time test ensures that the valve will isolate in a time period less than or equal to that assumed in the safety analyses. The isolation time and Frequency of this SR are in accordance with the requirements of the INSERVICE TESTING PROGRAM.
SR 3.6.1.3.6 Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. The isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the DBA analyses. This ensures that the calculated radiological consequences of these events remain within 10 CFR 50.67 limits. The Frequency of this SR is in accordance with the requirements of the INSERVICE TESTING PROGRAM.
(continued)
BFN-UNIT 2                    B 3.6-33                  Revision 0, 29, 109, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.7 REQUIREMENTS (continued) Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.1 overlaps this SR to provide complete testing of the safety function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.8 This SR requires a demonstration that a representative sample of reactor instrumentation line excess flow check valves (EFCVs) are OPERABLE by verifying that the valves actuate to the isolation position on an actual or simulated instrument line break signal. This SR provides assurance that the instrumentation line EFCVs will perform so that the radiological consequences will not exceed the predicted radiological consequences during events evaluated in Reference 5.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.6-34                          Revision 0, 123 Amendment No. 255, 268 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.9 REQUIREMENTS (continued) The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.10 The analyses in References 1 and 5 are based on leakage that is less than the specified leakage rate. Leakage through each MSIV must be d 100 scfh when tested at t Pt (25 psig). The combined leakage rate for all four main steam lines must be d 150 scfh when tested at t 25 psig in accordance with the Primary Containment Leakage Rate Testing Program. If the leakage rate through an individual MSIV exceeds 100 scfh, the leakage rate shall be restored below the alarm limit value as specified in the Containment Leakage Rate Testing Program referenced in TS 5.5.12. This ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate. The Frequency is specified in the Primary Containment Leakage Rate Testing Program.
(continued)
BFN-UNIT 2                    B 3.6-35                        Revision 62, 123 Amendment No. 255, 263, 267 March 22, 2020
 
PCIVs B 3.6.1.3 BASES REFERENCES 1. FSAR, Section 14.6.
: 2. BFN Technical Instruction (TI), 0-TI-360.
: 3. 10 CFR 50, Appendix J, Option B.
: 4. FSAR, Section 5.2.
: 5. FSAR, Section 14.6.5.
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 7. FSAR Table 5.2-2.
: 8. Deleted.
: 9. MDQ0000012016000566, Revision 0, Main Steam Isolation Valve (MSIV) Loss of Coolant Accident (LOCA) Closure Analysis, dated September 2016.
BFN-UNIT 2                B 3.6-36              Amendment No. 263, 268 Revision 0, 62, 105, 123 March 22, 2020
 
Drywell Air Temperature B 3.6.1.4 BASES ACTIONS      B.1 and B.2 (continued)
If the drywell average air temperature cannot be restored to within limit 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.
SURVEILLANCE SR 3.6.1.4.1 REQUIREMENTS Verifying that the drywell average air temperature is within the LCO limit ensures that operation remains within the limits assumed for the primary containment analyses. Drywell air temperature is monitored in various quadrants and at various elevations (referenced to mean sea level). Due to the shape of the drywell, a volumetric average is used to determine an accurate representation of the actual average temperature.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.6-39                        Revision 0, 123 March 22, 2020
 
Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.5 BASES (continued)
SURVEILLANCE      SR 3.6.1.5.1 REQUIREMENTS Each vacuum breaker is verified to be closed to ensure that a potential breach in the primary containment boundary is not present. This Surveillance is performed by observing local or control room indications of vacuum breaker position or by verifying a differential pressure of 0.5 psid is maintained between the reactor building and suppression chamber. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Two Notes are added to this SR. The first Note allows reactor building-to-suppression chamber vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable breakers. A second Note is included to clarify that vacuum breakers open due to an actual differential pressure, are not considered as failing this SR.
SR 3.6.1.5.2 Each vacuum breaker must be cycled to ensure that it opens properly to perform its design function and returns to its fully closed position. This ensures that the safety analysis assumptions are valid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                          B 3.6-47                      Revision 0, 109, 123 March 22, 2020
 
Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.5 BASES SURVEILLANCE SR 3.6.1.5.3 REQUIREMENTS (continued) Demonstration of vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of d 0.5 psid is valid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. TVA Calculation ND-Q0064-900040.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                    B 3.6-48                          Revision 123 Amendment No. 255 March 22, 2020
 
Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.6 BASES (continued)
SURVEILLANCE      SR 3.6.1.6.1 REQUIREMENTS Each vacuum breaker is verified closed to ensure that this potential large bypass leakage path is not present. This Surveillance is performed by observing the vacuum breaker position indication or by verifying that the rate of increase in suppression chamber pressure is less than 0.25 inches of water per minute over a ten minute period at a differential pressure of at least 1.0 psi. Note 2 specifies that vacuum breaker may be nonfully closed provided it is not more than 3&deg; open as indicated by position indication lights. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Note 1 has been added to this SR which allows suppression chamber-to-drywell vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers.
SR 3.6.1.6.2 Each required (i.e., required to be OPERABLE for opening) vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position. This ensures that the safety analysis assumptions are valid. The INSERVICE TESTING PROGRAM Frequency is based on operating experience that has demonstrated that the Frequency is adequate to assure OPERABILITY.
(continued)
BFN-UNIT 2                          B 3.6-55                      Revision 0, 109, 123 March 22, 2020
 
Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.6 BASES SURVEILLANCE SR 3.6.1.6.3 REQUIREMENTS (continued) Verification of the differential pressure required to open the vacuum breaker is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of 0.5 psid is valid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 5.2.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 3. Technical Requirements Manual.
BFN-UNIT 2                    B 3.6-56                            Revision 123 Amendment No. 255 March 22, 2020
 
Suppression Pool Average Temperature B 3.6.2.1 BASES (continued)
SURVEILLANCE      SR 3.6.2.1.1 REQUIREMENTS The suppression pool average temperature is regularly monitored to ensure that the required limits are satisfied. The average temperature is determined by taking an arithmetic average of OPERABLE suppression pool water temperature channels. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The 5 minute Frequency during testing is justified by the rates at which tests will heat up the suppression pool, has been shown to be acceptable based on operating experience, and provides assurance that allowable pool temperatures are not exceeded.
The Frequency is further justified in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature condition.
REFERENCES        1. FSAR, Section 5.2.
: 2. FSAR, Section 14.6.
: 3. NUREG-0783, Suppression Pool Temperature Limits for BWR Containments, November 1981.
: 4. NUREG-0661, "Safety Evaluation Report Mark I Containment Long Term Program - Resolution of Generic Technical Activity A-7," July 1980.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 6. NEDC-22004-P, Browns Ferry Nuclear Plant Units 1, 2, and 3 Suppression Pool Temperature Response, October 1981.
BFN-UNIT 2                        B 3.6-63                  Revision 0, 66, 85, 123 March 22, 2020
 
Suppression Pool Water Level B 3.6.2.2 BASES ACTIONS      B.1 and B.2 (continued)
If suppression pool water level cannot be restored to within limits 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.
SURVEILLANCE SR 3.6.2.2.1 REQUIREMENTS Verification of the suppression pool water level is to ensure that the required limits are satisfied. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Sections 5.2 and 14.6.3.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                    B 3.6-67                        Revision 0, 123 March 22, 2020
 
RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS      D.1 and D.2 (continued)
If any Required Action and associated Completion Time cannot be met, 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.
SURVEILLANCE SR 3.6.2.3.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.6-72                            Revision 0, 123 Amendment No. 272 March 22, 2020
 
RHR Suppression Pool Spray B 3.6.2.4 BASES SURVEILLANCE SR 3.6.2.4.1 (continued)
REQUIREMENTS sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool spray mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.2.4.2 This Surveillance is performed using air or water to verify that the spray nozzles are not obstructed and that flow will be provided when required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Sections 5.2 and 14.6.3.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                    B 3.6-79                            Revision 0, 123 March 22, 2020
 
RHR Drywell Spray B 3.6.2.5 BASES ACTIONS      D.1 and D.2 (continued)
If any Required Action and the associated Completion Time cannot be met, 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 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.
SURVEILLANCE SR 3.6.2.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR drywell spray mode flow path provides assurance that the proper flow paths will exist for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR drywell cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.6-84                            Revision 0, 123 March 22, 2020
 
RHR Drywell Spray B 3.6.2.5 BASES SURVEILLANCE SR 3.6.2.5.2 REQUIREMENTS (continued) This Surveillance is performed every 5 years using air to verify that the spray nozzles are not obstructed and that flow will be provided when required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Sections 5.2 and 14.6.3.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
U BFN-UNIT 2                  B 3.6-85                        Revision 3, 123 March 22, 2020
 
Drywell-to-Suppression Chamber Differential Pressure B 3.6.2.6 BASES (continued)
SURVEILLANCE      SR 3.6.2.6.1 REQUIREMENTS The drywell-to-suppression chamber differential pressure is regularly monitored to ensure that the required limits are satisfied. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 5.2.3.9.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                        B 3.6-89                        Revision 0, 123 March 22, 2020
 
CAD System B 3.6.3.1 BASES (continued)
SURVEILLANCE      SR 3.6.3.1.1 REQUIREMENTS Verifying that there is t 2615 gal of liquid nitrogen supply in each nitrogen storage tank will ensure at least 7 days of post-LOCA CAD operation. This minimum volume of liquid nitrogen represents the analytical limit assumed in the analysis of the primary containment atmosphere following a postulated LOCA and does not include allowance for potential nitrogen boiloff and tank level instrumentation inaccuracies. This minimum volume of liquid nitrogen allows sufficient time after an accident to replenish the nitrogen supply for long term inerting.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.3.1.2 Verifying the correct alignment for manual, power operated, and automatic valves in each of the CAD subsystem flow paths provides assurance that the proper flow paths exist for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing.
A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable because the CAD System is manually initiated. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.
(continued)
BFN-UNIT 2                        B 3.6-95                      Revision 0, 113, 123 Amendment No. 265 March 22, 2020
 
CAD System B 3.6.3.1 BASES SURVEILLANCE SR 3.6.3.1.2 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. AEC Safety Guide 7, Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident, March 10, 1971.
: 2. FSAR, Section 5.2.6.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 4. ANP-3403P, Fuel Uprate Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3, Section 2.6.4.
BFN-UNIT 2                  B 3.6-96                    Revision 0, 113, 123 Amendment No. 265 March 22, 2020
 
Primary Containment Oxygen Concentration B 3.6.3.2 BASES (continued)
SURVEILLANCE      SR 3.6.3.2.1 REQUIREMENTS The primary containment (drywell and suppression chamber) must be determined to be inert by verifying that oxygen concentration is < 4.0 v/o. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 5.2.6.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                        B 3.6-100                        Revision 0, 123 March 22, 2020
 
Secondary Containment B 3.6.4.1 BASES (continued)
SURVEILLANCE      SR 3.6.4.1.1 and SR 3.6.4.1.2 REQUIREMENTS Verifying that secondary containment equipment hatches and one access door in each access opening are closed ensures that the infiltration of outside air of such a 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 the access opening is closed. An access opening contains one inner and one outer door. In some cases, secondary containment access openings are shared such that a secondary containment barrier may have multiple inner doors. The main Equipment Access Lock (EAL) has a smaller sub-door on each of the large inner and outer main EAL doors. For the EAL, maintaining secondary containment OPERABILITY requires verifying that a large door and its integral sub-door are both closed. The intent is to not breach the secondary containment at any time when secondary containment is required. This is achieved by maintaining the inner or outer portion of the barrier closed at all times. However, all secondary containment access doors are normally kept closed, except when the access opening is being used for entry and exit or when maintenance is being performed on an access opening. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                          B 3.6-105                  Revision 0, 23, 29, 123 Amendment No. 264 March 22, 2020
 
Secondary Containment B 3.6.4.1 BASES SURVEILLANCE SR 3.6.4.1.3 and SR 3.6.4.1.4 REQUIREMENTS (continued) The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment. To ensure that all fission products are treated, SR 3.6.4.1.3 verifies that the SGT System will rapidly establish and maintain a pressure in the secondary containment that is less than the lowest postulated pressure external to the secondary containment boundary. This is confirmed by demonstrating that two SGT subsystems will draw down the secondary containment to t 0.25 inches of vacuum water gauge in d 120 seconds. This cannot be accomplished if the secondary containment boundary is not intact. SR 3.6.4.1.4 demonstrates that two SGT subsystems can maintain t 0.25 inches of vacuum water gauge at a stable flow rate d 12,000 cfm. Both of these SRs are performed under neutral
( 5 mph) wind conditions. Therefore, these two tests are used to ensure secondary containment boundary integrity. Since these SRs are secondary containment tests, they need not be performed with each combination of SGT subsystems. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 5.3.
: 2. FSAR, Section 14.6.3.
: 3. Deleted.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                  B 3.6-106                      Revision 29, 123 Amendment No. 255 March 22, 2020
 
SCIVs B 3.6.4.2 BASES (continued)
SURVEILLANCE      SR 3.6.4.2.1 REQUIREMENTS Verifying that the isolation time of each power operated, automatic SCIV is within limits is required to demonstrate OPERABILITY. The isolation time test ensures that the SCIV will isolate in a time period less than or equal to that assumed in the safety analyses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.4.2.2 Verifying that each automatic SCIV closes on a secondary containment isolation signal is required to prevent leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SCIV 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 14.6.3.
: 2. Deleted.
: 3. Technical Requirements Manual.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                          B 3.6-113                        Revision 29, 123 Amendment No. 255 March 22, 2020
 
SGT System B 3.6.4.3 BASES (continued)
SURVEILLANCE      SR 3.6.4.3.1 REQUIREMENTS Operating each SGT subsystem for t 15 continuous minutes with heaters on ensures that the 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. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.4.3.2 This SR verifies that the required SGT filter testing is performed in accordance with the 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). This SR will also include a chemical smoke test to check the sealing of gaskets for filter housing doors.
Specific test frequencies and additional information are discussed in detail in the VFTP.
(continued)
BFN-UNIT 2                        B 3.6-120                    Revision 0, 108, 123 March 22, 2020
 
SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.3 REQUIREMENTS (continued) This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.4.3.4 This SR verifies that the SGT decay heat discharge dampers are in the correct position. This ensures that the decay heat removal mode of SGT System operation is available. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 41.
: 2. FSAR, Section 5.3.3.7.
: 3. FSAR, Section 14.6.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                  B 3.6-121                            Revision 123 Amendment No. 255 March 22, 2020
 
RHRSW System B 3.7.1 BASES ACTIONS      G.1 and G.2 (continued)
If the RHRSW subsystem(s) or the RHRSW pump(s) cannot be restored to OPERABLE status within the associated Completion Times, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.1.1 REQUIREMENTS Verifying the correct alignment for each manual and power operated valve in each RHRSW subsystem flow path provides assurance that the proper flow paths will exist for RHRSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be realigned to its accident position. This is acceptable because the RHRSW System is a manually initiated system.
This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                      B 3.7-9                    Revision 73, 113, 123 Amendment No. 254 March 22, 2020
 
EECW System and UHS B 3.7.2 BASES ACTIONS      A.1 (continued)
The 7 day Completion Time is based on the redundant EECW System capabilities afforded by the remaining OPERABLE pumps, the low probability of an accident occurring during this time period and is consistent with the allowed Completion Time for restoring an inoperable DG.
B.1 and B.2 If the required EECW pump cannot be restored to OPERABLE status within the associated Completion Time, or two or more EECW pumps are inoperable or the UHS is determined inoperable, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.2.1 REQUIREMENTS Verification of the UHS temperature ensures that the heat removal capability of the EECW System is within the assumptions of the DBA analysis. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.7-14                Revision 0, 69, 113, 123 March 22, 2020
 
EECW System and UHS B 3.7.2 BASES SURVEILLANCE SR 3.7.2.2 REQUIREMENTS (continued) Verifying the correct alignment for each manual and power operated valve in the EECW System flow paths provide assurance that the proper flow paths will exist for EECW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be automatically realigned to its accident position within the required time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
This SR is modified by a Note indicating that isolation of the EECW System to components or systems may render those components or systems inoperable, but does not affect the OPERABILITY of the EECW System. As such, when required EECW pumps, valves, and piping are OPERABLE, but a branch connection off the main header is isolated, the EECW System is still OPERABLE.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.7-15                            Revision 0, 123 March 22, 2020
 
EECW System and UHS B 3.7.2 BASES SURVEILLANCE SR 3.7.2.3 REQUIREMENTS (continued) This SR verifies that the EECW System pumps will automatically start to provide cooling water to the required safety related equipment during an accident event. This is demonstrated by the use of an actual or simulated initiation signal. This SR includes a functional test of the initiation logic and a functional test and calibration of the EECW pump timers (both normal power and diesel power).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Chapter 5.
: 2. FSAR, Chapter 14.
: 3. FSAR, Section 10.10.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                  B 3.7-16                              Revision 123 Amendment No. 255 March 22, 2020
 
CREV System B 3.7.3 BASES (continued)
SURVEILLANCE      SR 3.7.3.1 REQUIREMENTS This SR verifies that a subsystem in a standby mode starts on demand and continues to operate. Standby systems should be checked periodically to ensure that they start and function properly. As the environmental and normal operating conditions of this system are not severe, testing each subsystem once every month provides an adequate check on this system.
Operation with the heaters on for  15 continuous minutes demonstrates OPERABILITY of the system. Periodic operation ensures that heater failure, blockage, fan or motor failure, or excessive vibration can be detected for corrective action. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.3.2 This SR verifies that the required CREV testing is performed in accordance with the 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).
Specific test frequencies and additional information are discussed in detail in the VFTP.
(continued)
BFN-UNIT 2                        B 3.7-24                Revision 0, 67, 108, 123 March 22, 2020
 
CREV System B 3.7.3 BASES SURVEILLANCE SR 3.7.3.3 REQUIREMENTS (continued) This SR verifies that on an actual or simulated initiation signal, each CREV subsystem starts and operates. This SR includes verification that dampers necessary for proper CREV operation function as required. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.7.1.4 and SR 3.3.7.1.6 overlaps this SR to provide complete testing of the safety function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.3.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.
The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more that 5 REM TEDE and the CRE occupants are protected from hazardous chemicals and smoke.
There is no automatic CREV actuation for hazardous chemical releases or smoke and there are no Surveillance Requirements to verify the OPERABILITY in cases of hazardous chemicals or smoke. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analysis of DBA consequences. When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered. Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for occupants following an accident.
Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 6) which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 7).
These compensatory measures may also be used as mitigating actions as required by Required Action B.2. Temporary analytical methods may also be used as compensatory BFN-UNIT 2                      B 3.7-25                          Revision 123 Amendment No. 255, 302 March 22, 2020
 
Control Room AC System B 3.7.4 BASES (continued)
SURVEILLANCE      SR 3.7.4.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the control room heat load assumed in the safety analyses. The SR consists of a combination of testing and calculation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 10.12.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                        B 3.7-31                            Revision 123 Amendment No. 255 March 22, 2020
 
Main Turbine Bypass System B 3.7.5 BASES ACTIONS      B.1 (continued)
Turbine Bypass System is not required to protect fuel integrity during abnormal operational transients. The 4 hour Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Cycling each main turbine bypass valve through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and will function when required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.5.2 The Main Turbine Bypass System is required to actuate automatically to perform its design function. This SR demonstrates that, with the required system initiation signals, the valves will actuate to their required position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.7-35                            Revision 123 Amendment No. 255 March 22, 2020
 
Main Turbine Bypass System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.3 REQUIREMENTS (continued) This SR ensures that the TURBINE BYPASS SYSTEM RESPONSE TIME is in compliance with the assumptions of the appropriate safety analysis. The response time limits are specified in the cycle specific transient analyses performed to support the preparation of FSAR, Appendix N, Supplemental Reload Licensing Report (Ref. 4). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 7.11.
: 2. FSAR, Section 14.5.1.1 and 14.5.1.2.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 4. FSAR, Appendix N.
BFN-UNIT 2                    B 3.7-36                    Revision 31, 103, 123 Amendment No. 255 March 22, 2020
 
Spent Fuel Storage Pool Water Level B 3.7.6 BASES (continued)
SURVEILLANCE      SR 3.7.6.1 REQUIREMENTS This SR verifies that sufficient water is available in the event of a fuel handling accident. The water level in the spent fuel storage pool must be checked periodically. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 10.3.
: 2. FSAR, Section 14.6.4.
: 3. NUREG-0800, Section 15.0.1.
: 4. 10 CFR 50.67.
: 5. Regulatory Guide 1.183.
: 6. FSAR, Section 14.6.4.5.
: 7. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                        B 3.7-39                      Revision 0, 29, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.1 and SR 3.8.1.4 (continued)
REQUIREMENTS SR 3.8.1.4 requires that the DG starts from standby conditions and achieves required voltage and frequency within 10 seconds.
The 10 second start requirement supports the assumptions in the design basis LOCA analysis of FSAR, Section 14.6.3 (Ref. 10). The 10 second start requirement is not applicable to SR 3.8.1.1 (see the Note for SR 3.8.1.1), when a modified start procedure as described above is used. If a modified start is not used, the 10 second start requirement of SR 3.8.1.4 applies.
Since SR 3.8.1.4 does require a 10 second start, it is more restrictive than SR 3.8.1.1, and it may be performed in lieu of SR 3.8.1.1. This procedure is the intent of Note 1 of SR 3.8.1.1.
In addition to the SR requirements, the time for the DG to reach steady state operation, unless the modified DG start method is employed, is periodically monitored and the trend evaluated to identify degradation of governor performance.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.8-29                          Revision 0, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 REQUIREMENTS (continued) This Surveillance demonstrates that the DGs are capable of synchronizing and accepting greater than 90 percent of the continuous rating. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source.
Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Note 1 modifies this Surveillance to indicate that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized.
Note 2 modifies this Surveillance by stating that momentary transients because of changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test.
Note 3 indicates that this Surveillance should be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations.
Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.
(continued)
BFN-UNIT 2                    B 3.8-30                            Revision 0, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.3 REQUIREMENTS (continued) This Surveillance demonstrates that each required fuel oil transfer pump operates and transfers fuel oil from its associated 7-day storage tank to its associated engine fuel oil tank. It is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE.
The design of fuel transfer systems is such that pumps that transfer the fuel oil operate automatically in order to maintain an adequate volume of fuel oil in the engine tank during or following DG operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.1.4 See SR 3.8.1.1.
(continued)
BFN-UNIT 2                    B 3.8-31                          Revision 0, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.5 (continued)
REQUIREMENTS The voltage tolerances specified in this SR are based on the degraded voltage and overvoltage relay settings. The frequency tolerances specified in this SR are derived from Safety Guide 9 (Ref. 3) recommendations for response during load sequence intervals. The voltage and frequency specified are consistent with the design range of the equipment powered by the DG. SR 3.8.1.5.a corresponds to the maximum frequency excursion, while SR 3.8.1.5.b and 3.8.1.5.c are steady state voltage and frequency values to which the system must recover following load rejection. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, the Note requires that, if synchronized to offsite power, testing must be performed using a power factor d 0.9. This power factor is chosen to be representative of the actual design basis inductive loading that the DG would experience.
SR 3.8.1.6 This Surveillance demonstrates that the DG automatically starts from the design basis actuation signal (LOCA signal). This test will also verify the start of the Unit 3 DGs aligned to the SGT and CREV Systems on an accident signal from Unit 2. In order to minimize the number of DGs involved in testing, demonstration of automatic starts of the Unit 3 DGs on an accident signal from Unit 2 may be performed in conjunction (continued)
BFN-UNIT 2                      B 3.8-33                            Revision 123 Amendment No. 255 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.6 (continued)
REQUIREMENTS with testing to demonstrate automatic starts of the Unit 3 DGs on an accident signal from Unit 3. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
To minimize wear and tear on the DGs, this SR has been modified by a Note which permits DG starts to be preceded by an engine prelube period followed by a warmup period.
SR 3.8.1.7 Demonstration periodically that the DGs can start and run continuously at full load capability for an interval of not less than 24 hours - 22 hours of which is at a load equivalent to the continuous rating of the DG, and 2 hours of which is at a load equivalent to 105 percent to 110 percent of the continuous duty rating of the DG. The DG starts for this Surveillance can be performed either from standby or hot conditions. The provisions for prelube and warmup, discussed in SR 3.8.1.1, and for gradual loading, discussed in SR 3.8.1.2, are applicable to this SR.
In order to ensure that the DG is tested under load conditions that are as close to design conditions as possible, testing must be performed using a power factor d 0.9. This power factor is chosen to be representative of the actual design basis inductive loading that the DG could experience. A load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.
(continued)
BFN-UNIT 2                    B 3.8-34                              Revision 123 Amendment No. 255 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.7 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This Surveillance has been modified by a Note that states that momentary transients due to changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test.
SR 3.8.1.8 Under accident conditions (and loss of offsite power) loads are sequentially connected to the shutdown boards by automatic individual pump timers. The individual pump timers control the permissive and starting signals to motor breakers to prevent overloading of the DGs due to high motor starting currents. This SR is demonstrated by performance of SR 3.3.5.1.5 for the Core Spray and LPCI pump timers, SR 3.7.2.3 for the EECW pump timers, and SR 3.8.1.9.b for the 480 V load shed logic timers. The allowable values for these timers ensure that sufficient time exists for the DG to restore frequency and voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not violated. Reference 2 provides a summary of the automatic loading of ESF shutdown boards.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.8-35                            Revision 123 Amendment No. 255 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.9 (continued)
REQUIREMENTS mode of operation. In lieu of actual demonstration of the connection and loading of these loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note. The reason for the Note is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil being continuously circulated and temperature maintained consistent with manufacturer recommendations.
SR 3.8.1.10 This Surveillance is provided to direct that the appropriate Surveillances for the required Unit 3 DGs are governed by the Unit 3 Technical Specifications. Performance of the applicable Unit 3 Surveillances will satisfy any Unit 3 requirements, as well as this Unit 1 and 2 Surveillance requirement. The Frequency required by the applicable Unit 3 SR also governs performance of that SR for both Units.
(continued)
BFN-UNIT 2                    B 3.8-37                        Revision 42, 123 Amendment No. 255 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES (continued)
REFERENCES        1. 10 CFR 50, Appendix A, GDC 17.
: 2. FSAR, Chapter 8.
: 3. Safety Guide 9.
: 4. FSAR, Chapter 6.
: 5. FSAR, Chapter 14.
: 6. Regulatory Guide 1.93.
: 7. Generic Letter 84-15.
: 8. Deleted.
: 9. ANSI C84.1, 1982.
: 10. FSAR, Section 14.6.3.
: 11. IEEE Standard 308.
: 12. FSAR, Section 8.5, Table 8.5-6.
: 13. FSAR, Section 8.5.2.
: 14. TVA Design Criteria BFN-50-7082.
: 15. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                        B 3.8-38                      Revision 0, 123 March 22, 2020
 
Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES (continued)
SURVEILLANCE      SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 7 days at full load. The fuel oil level equivalent to a 7-day supply is 35,280 gallons when calculated in accordance with References 2 and 6. The required fuel storage volume is determined using the most limiting energy content of the stored fuel. Using the known correlation of diesel fuel oil absolute specific gravity or API gravity to energy content, the required diesel generator output, and the corresponding fuel consumption rate, the on site fuel storage volume required for 7 days of operation can be determined. SR 3.8.3.3 requires new fuel to be tested to verify that the absolute specific gravity or API gravity is within the range assumed in the diesel fuel oil consumption calculations. The 7-day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.3.2 This Surveillance ensures that sufficient lubricating oil inventory is available to support at least 7 days of full load operation for each DG. The lube oil inventory equivalent to a 7-day supply is 175 gallons and is based on the DG manufacturer's consumption values for the run time of the DG.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                          B 3.8-55                  Revision 0, 63, 95, 123 March 22, 2020
 
Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.4 REQUIREMENTS (continued) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.3.5 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Periodic removal of water from the fuel storage tanks eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling.
In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 8.5.3.4.
: 2. Regulatory Guide 1.137, Revision 1, October 1979.
: 3. FSAR, Chapter 6.
: 4. FSAR, Chapter 14.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                    B 3.8-56b                      Revision 0, 95, 123 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES (continued)
SURVEILLANCE      SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or a battery cell) in a fully charged state, while supplying adequate power to the connected DC loads. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.4.2 and SR 3.8.4.5 Battery charger capability requirements are based on the design capacity of the chargers (Ref. 4). According to Regulatory Guide 1.32 (Ref. 8), the battery charger supply is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and verification of the charger's ability to recharge the battery ensures that these requirements can be satisfied.
(continued)
BFN-UNIT 2                          B 3.8-64                            Revision 0, 123 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.2 and SR 3.8.4.5 (continued)
REQUIREMENTS SR 3.8.4.2 verifies that the chargers are capable of charging the batteries after their designed duty cycle testing and ensures that the chargers will perform their design function. This SR is modified by a Note that allows the performance of SR 3.8.4.5 in lieu of this Surveillance requirement. SR 3.8.4.5 verifies that the chargers are capable of charging the batteries after each discharge test and ensures that the chargers are capable of performing at maximum output. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.4.5 is modified by a Note. The Note is added to this SR to acknowledge that credit may be taken for unplanned events that satisfy the Surveillance.
SR 3.8.4.3 A battery service test is a special test of the battery's capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length corresponds to the design duty cycle requirements as specified in Reference 4.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.8-65                              Revision 123 Amendment No. 255 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.3 (continued)
REQUIREMENTS This SR is modified by a Note that allows the performance of a modified performance discharge test in lieu of a service test.
The modified performance discharge test is a simulated duty cycle consisting of just two periods (the one minute rate, followed by the test rate employed for the performance test) or three periods (the one minute rate, followed by the second minute rate followed by the test rate employed for the performance test) both of which envelope the duty cycle of the service test. Since the ampere-hours removed by the rated one or two minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.
A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity.
Initial conditions for the modified performance discharge test should be identical to those specified for a service test.
(continued)
BFN-UNIT 2                      B 3.8-66                      Revision 0, 58, 123 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.4 REQUIREMENTS (continued) A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.
A battery modified performance discharge test is described in the Bases for SR 3.8.4.3. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.4; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.4 while satisfying the requirements of SR 3.8.4.3 at the same time.
The acceptance criteria for this Surveillance is consistent with IEEE-450 (Ref. 7) and IEEE-485 (Ref. 10). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. If the battery shows degradation, or if the battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity t 100% of the manufacturer's rating. Degradation is indicated, according to IEEE-450 (Ref. 7), when the battery capacity drops by more than 10%
relative to its capacity on the previous performance test or when it is 10% below the manufacturer's rating. All these Frequencies are consistent with the recommendations in IEEE-450 (Ref. 7).
(continued)
BFN-UNIT 2                    B 3.8-67                          Revision 0, 123 March 22, 2020
 
Battery Cell Parameters B 3.8.6 BASES (continued)
SURVEILLANCE      SR 3.8.6.1 REQUIREMENTS This SR verifies that Category A battery cell parameters are consistent with IEEE-450 (Ref. 3), including voltage, specific gravity, and electrolyte temperature of pilot cells. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.6.2 The inspection of specific gravity and voltage is consistent with IEEE-450 (Ref. 3). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.6.3 This Surveillance verification that the average temperature of representative cells is within limits is consistent with a recommendation of IEEE-450 (Ref. 3) that states that the temperature of electrolytes in representative (10 percent of) cells should be determined. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Lower than normal temperatures act to inhibit or reduce battery capacity. This SR ensures that the operating temperatures remain within an acceptable operating range. This limit is based on manufacturer's recommendations.
(continued)
BFN-UNIT 2                        B 3.8-78                            Revision 0, 123 March 22, 2020
 
Distribution Systems - Operating B 3.8.7 BASES ACTIONS      H.1 and H.2 (continued)
If the inoperable distribution subsystem cannot be restored to OPERABLE status within the associated Completion Time, the unit 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.
I.1 Condition I corresponds to a level of degradation in the electrical distribution system that causes a required safety function to be lost. When more than one AC or DC electrical power distribution subsystem is lost, and this results in the loss of a required function, the plant is in a condition outside the accident analysis. Therefore, no additional time is justified for continued operation. LCO 3.0.3 must be entered immediately to commence a controlled shutdown.
SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the AC and DC electrical power distribution subsystem is functioning properly, with the buses energized. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                    B 3.8-100                          Revision 0, 123 March 22, 2020
 
Distribution Systems - Shutdown B 3.8.8 BASES (continued)
SURVEILLANCE      SR 3.8.8.1 REQUIREMENTS This Surveillance verifies that the AC and DC electrical power distribution subsystem is functioning properly, with the buses energized. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Chapter 6.
: 2. FSAR, Chapter 14.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                        B 3.8-108                          Revision 0, 123 March 22, 2020
 
Refueling Equipment Interlocks B 3.9.1 BASES ACTIONS      A.1, A.2.1, and A.2.2 (continued) remain inserted). Required Action A.2.2 is normally performed after placing the rod withdrawal block in effect, and provides a verification that all control rods are fully inserted. This verification that all control rods are fully inserted is in addition to the periodic verifications required by SR 3.9.3.1. Like Required Action A.1, Required Actions A.2.1 and A.2.2 ensure unacceptable operations are blocked (e.g., loading fuel into a cell with the control rod withdrawn). It is not the intent of Actions A.2 to eliminate the first performance of SR 3.9.1.1 prior to in-vessel fuel movement. It is expected that the refueling interlocks would be operable except for equipment failure or expiration of the required surveillance interval, and Actions A.2 would not be entered as a convenience for avoiding the first performance of SR 3.9.1.1.
SURVEILLANCE SR 3.9.1.1 REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST demonstrates each required refueling equipment interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested. This SR is only required for refueling equipment in use.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                      B 3.9-5                        Revision 0, 16, 123 Amendment No. 274 March 22, 2020
 
Refuel Position One-Rod-Out Interlock B 3.9.2 BASES (continued)
SURVEILLANCE      SR 3.9.2.1 REQUIREMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel. During control rod withdrawal in MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock will be OPERABLE when required. By "locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel),
an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested.
(continued)
BFN-UNIT 2                          B 3.9-9                          Revision 0, 123 March 22, 2020
 
Refuel Position One-Rod-Out Interlock B 3.9.2 BASES SURVEILLANCE SR 3.9.2.2 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. To perform the required testing, the applicable condition must be entered (i.e., a control rod must be withdrawn from its full-in position). Therefore, SR 3.9.2.2 has been modified by a Note that states the CHANNEL FUNCTIONAL TEST is not required to be performed until 1 hour after any control rod is withdrawn.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 26.
: 2. FSAR, Section 7.6.3.
: 3. FSAR, Section 14.5.4.3.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                    B 3.9-10                      Revision 0, 60, 123 March 22, 2020
 
Control Rod Position B 3.9.3 BASES (continued)
ACTIONS          A.1 With all control rods not fully inserted during the applicable conditions, an inadvertent criticality could occur that is not analyzed in the FSAR. All fuel loading operations must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE      SR 3.9.3.1 REQUIREMENTS During refueling, to ensure that the reactor remains subcritical, all control rods must be fully inserted prior to and during fuel loading. Periodic checks of the control rod position ensure this condition is maintained.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. 10 CFR 50, Appendix A, GDC 26.
: 2. FSAR, Section 14.5.4.3.
: 3. FSAR, Section 14.5.4.4.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                          B 3.9-13                      Revision 0, 60, 123 March 22, 2020
 
Control Rod OPERABILITY - Refueling B 3.9.5 BASES (continued)
ACTIONS          A.1 With one or more withdrawn control rods inoperable, action must be immediately initiated to fully insert the inoperable control rod(s). Inserting the control rod(s) ensures the shutdown and scram capabilities are not adversely affected. Actions must continue until the inoperable control rod(s) is fully inserted.
SURVEILLANCE      SR 3.9.5.1 and SR 3.9.5.2 REQUIREMENTS During MODE 5, the OPERABILITY of control rods is primarily required to ensure a withdrawn control rod will automatically insert if a signal requiring a reactor shutdown occurs. Because no explicit analysis exists for automatic shutdown during refueling, the shutdown function is satisfied if the withdrawn control rod is capable of automatic insertion and the associated CRD scram accumulator pressure is t 940 psig.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. An automatic accumulator monitor may be used to continuously satisfy SR 3.9.5.2.
SR 3.9.5.1 is modified by a Note that allows 7 days after withdrawal of the control rod to perform the Surveillance. This acknowledges that the control rod must first be withdrawn before performance of the Surveillance, and therefore avoids potential conflicts with SR 3.0.3 and SR 3.0.4.
(continued)
BFN-UNIT 2                          B 3.9-21                          Revision 0, 123 March 22, 2020
 
RPV Water Level B 3.9.6 BASES (continued)
ACTIONS          A.1 If the water level is < 22 ft above the top of the RPV flange, all operations involving movement of fuel assemblies and handling of control rods within the RPV shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement and control rod handling shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE      SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 22 ft above the top of the RPV flange ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. 2).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. Regulatory Guide 1.183.
: 2. FSAR, Section 14.6.4.
: 3. NUREG-0800, Section 15.0.1.
: 4. 10 CFR 50.67.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                          B 3.9-25                      Revision 0, 29, 123 March 22, 2020
 
RHR-High Water Level B 3.9.7 BASES ACTIONS      C.1 and C.2 (continued)
If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour. This alternative method may utilize forced or natural circulation. The Completion Time is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation.
During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR Shutdown Cooling System), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.
SURVEILLANCE SR 3.9.7.1 REQUIREMENTS This Surveillance demonstrates that the RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 34.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                    B 3.9-30                          Revision 0, 123 March 22, 2020
 
RHR-Low Water Level B 3.9.8 BASES ACTIONS      C.1 and C.2 (continued)
If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour. This alternative method may utilize forced or natural circulation. The Completion Time is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation.
During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.
SURVEILLANCE SR 3.9.8.1 REQUIREMENTS This Surveillance demonstrates that one RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 34.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 2                    B 3.9-35                          Revision 0, 123 March 22, 2020
 
Reactor Mode Switch Interlock Testing B 3.10.2 BASES (continued)
SURVEILLANCE      SR 3.10.2.1 and SR 3.10.2.2 REQUIREMENTS Meeting the requirements of this Special Operations LCO maintains operation consistent with or conservative to operating with the reactor mode switch in the shutdown position (or the refuel position for MODE 5). The functions of the reactor mode switch interlocks that are not in effect, due to the testing in progress, are adequately compensated for by the Special Operations LCO requirements. The administrative controls are to be periodically verified to ensure that the operational requirements continue to be met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 7.2.3.7.
: 2. FSAR, Section 14.5.3.3.
: 3. FSAR, Section 14.5.3.4.
BFN-UNIT 2                        B 3.10-12                          Revision 0, 123 March 22, 2020
 
Single Control Rod Withdrawal - Hot Shutdown B 3.10.3 BASES (continued)
SURVEILLANCE      SR 3.10.3.1, SR 3.10.3.2, and SR 3.10.3.3 REQUIREMENTS The other LCOs made applicable in this Special Operations LCO are required to have their Surveillances met to establish that this Special Operations LCO is being met. If the local array of control rods is inserted and disarmed (electrically or hydraulically) while the scram function for the withdrawn rod is not available, periodic verification in accordance with SR 3.10.3.2 is required to preclude the possibility of criticality.
SR 3.10.3.2 has been modified by a Note, which clarifies that this SR is not required to be met if SR 3.10.3.1 is satisfied for LCO 3.10.3.d.1 requirements, since SR 3.10.3.2 demonstrates that the alternative LCO 3.10.3.d.2 requirements are satisfied.
Also, SR 3.10.3.3 verifies that all control rods other than the control rod being withdrawn are fully inserted. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 14.5.3.3.
BFN-UNIT 2                        B 3.10-18                          Revision 0, 123 March 22, 2020
 
Single Control Rod Withdrawal - Cold Shutdown B 3.10.4 BASES (continued)
SURVEILLANCE      SR 3.10.4.1, SR 3.10.4.2, SR 3.10.4.3, and SR 3.10.4.4 REQUIREMENTS The other LCOs made applicable by this Special Operations LCO are required to have their associated Surveillances met to establish that this Special Operations LCO is being met. If the local array of control rods is inserted and disarmed (electrically or hydraulically) while the scram function for the withdrawn rod is not available, periodic verification is required to ensure that the possibility of criticality remains precluded. Verification that all the other control rods are fully inserted is required to meet the SDM requirements. Verification that a control rod withdrawal block has been inserted ensures that no other control rods can be inadvertently withdrawn under conditions when position indication instrumentation is inoperable for the affected control rod. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.10.4.2 and SR 3.10.4.4 have been modified by Notes, which clarify that these SRs are not required to be met if the alternative requirements demonstrated by SR 3.10.4.1 are satisfied.
REFERENCES        1. FSAR, Section 14.5.3.3.
BFN-UNIT 2                        B 3.10-25                            Revision 0, 123 March 22, 2020
 
Single CRD Removal - Refueling B 3.10.5 BASES SURVEILLANCE SR 3.10.5.1, SR 3.10.5.2, SR 3.10.5.3, SR 3.10.5.4, REQUIREMENTS and SR 3.10.5.5 (continued) control rod. The Surveillance for LCO 3.1.1, which is made applicable by this Special Operations LCO, is required in order to establish that this Special Operations LCO is being met.
Verification that no other CORE ALTERATIONS are being made is required to ensure the assumptions of the safety analysis are satisfied.
Periodic verification of the administrative controls established by this Special Operations LCO is prudent to preclude the possibility of an inadvertent criticality. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 14.5.3.3.
BFN-UNIT 2                    B 3.10-31                          Revision 0, 123 March 22, 2020
 
Multiple Control Rod Withdrawal - Refueling B 3.10.6 BASES (continued)
ACTIONS          A.1, A.2, A.3.1, and A.3.2 If one or more of the requirements of this Special Operations LCO are not met, the immediate implementation of these Required Actions restores operation consistent with the normal requirements for refueling (i.e., all control rods inserted in core cells containing one or more fuel assemblies) or with the exceptions granted by this Special Operations LCO. The Completion Times for Required Action A.1, Required Action A.2, Required Action A.3.1, and Required Action A.3.2 are intended to require that these Required Actions be implemented in a very short time and carried through in an expeditious manner to either initiate action to restore the affected CRDs and insert their control rods, or initiate action to restore compliance with this Special Operations LCO.
SURVEILLANCE      SR 3.10.6.1, SR 3.10.6.2, and SR 3.10.6.3 REQUIREMENTS Periodic verification of the administrative controls established by this Special Operations LCO is prudent to preclude the possibility of an inadvertent criticality. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. SR 3.10.6.3 is modified by a Note stating that the SR is only required to be met during refueling.
REFERENCES        1. FSAR, Section 14.5.3.3.
BFN-UNIT 2                          B 3.10-35                          Revision 0, 123 March 22, 2020
 
SDM Test -Refueling B 3.10.8 BASES (continued)
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 this Special Operations LCO, are required to have applicable Surveillances met to establish that this Special Operations LCO is being met. However, the control rod withdrawal sequences during the SDM tests may be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2 requirements) or by a second licensed operator or other qualified member of the technical staff (i.e., personnel trained in accordance with an approved training program for this test). As noted, either the applicable SRs for the RWM (LCO 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.8.2), or the proper movement of control rods must be verified (SR 3.10.8.3).
This latter verification (i.e., SR 3.10.8.3) must be performed during control rod movement to prevent deviations from the specified sequence. These Surveillances provide adequate assurance that the specified test sequence is being followed.
SR 3.10.8.4 Periodic verification of the administrative controls established by this LCO will ensure that the reactor is operated within the bounds of the safety analysis. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 2                          B 3.10-47                          Revision 0, 123 March 22, 2020
 
SDM Test -Refueling B 3.10.8 BASES SURVEILLANCE SR 3.10.8.5 REQUIREMENTS (continued) Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the "full out" notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events.
SR 3.10.8.6 CRD charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram signal. Since the reactor is depressurized in MODE 5, there is insufficient reactor pressure to scram the control rods. Verification of charging water pressure ensures that if a scram is required, capability for rapid control rod insertion would exist. The minimum pressure of 940 psig, which is well below the expected pressure of approximately 1100 psig, ensures sufficient pressure for rapid control rod insertion. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. NEDE-24011-P-A-13, "General Electric Standard Application for Reactor Fuel," August 1996.
: 2. Letter from T. Pickens (BWROG) to G. C. Lainas, NRC, "Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," August 15, 1986.
(continued)
BFN-UNIT 2                    B 3.10-48                        Revision 0, 31, 123 March 22, 2020
 
Control Rod OPERABILITY B 3.1.3 BASES (continued)
SURVEILLANCE      SR 3.1.3.1 REQUIREMENTS The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.3.2 and SR 3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. These Surveillances are not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of banked position withdrawal sequence (BPWS) (LCO 3.1.6) and the RWM (LCO 3.3.2.1). The 7 day Frequency of SR 3.1.3.2 is based on operating experience related to the changes in CRD performance and the ease of performing notch testing for fully withdrawn control rods. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                        B 3.1-23                          Revision 0, 123 March 22, 2020
 
Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE SR 3.1.4.2 REQUIREMENTS (continued) Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods. This sample remains representative if no more than 7.5% of the control rods in the sample tested are determined to be "slow." With more than 7.5% of the sample declared to be "slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 7.5% criterion (i.e., 7.5% of the entire sample) is satisfied, or until the total number of "slow" control rods (throughout the core from all Surveillances) exceeds the LCO limit. For planned testing, the control rods selected for the sample should be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.1-31                  Revision 0, 9, 35, 55, 123 Amendment No. 226 March 22, 2020
 
Control Rod Scram Accumulators B 3.1.5 BASES (continued)
SURVEILLANCE      SR 3.1.5.1 REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be checked periodically to ensure adequate accumulator pressure exists to provide sufficient scram force. An automatic accumulator monitor may be used to continuously satisfy this requirement.
The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940 psig is well below the expected pressure of 1100 psig (Ref. 1). Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant degradation in scram times does not occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 3.4.6.
: 2. FSAR, Section 14.5.
: 3. FSAR, Section 14.6.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                        B 3.1-40                        Revision 0, 123 March 22, 2020
 
Rod Pattern Control B 3.1.6 BASES ACTIONS      B.1 and B.2 (continued)
LCO 3.3.2.1 requires verification of control rod movement by a second licensed operator or a qualified member of the technical staff.
When nine or more OPERABLE control rods are not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within 1 hour. With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.
SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is periodically verified to be in compliance with the BPWS to ensure the assumptions of the CRDA analyses are met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The RMW provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at d 10% RTP.
(continued)
BFN-UNIT 3                    B 3.1-45                            Revision 0, 123 March 22, 2020
 
SLC System B 3.1.7 BASES ACTIONS      B.1 (continued)
If both SLC subsystems are inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours. The allowed Completion Time of 8 hours is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor.
C.1 and C.2 If any Required Action and associated Completion Time is not met, 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.
SURVEILLANCE SR 3.1.7.1 REQUIREMENTS SR 3.1.7.1 verifies the volume of the borated solution in the storage tank, thereby ensuring SLC System OPERABILITY without disturbing normal plant operation. This Surveillance ensures that the proper borated solution volume is maintained for reactivity control and post-LOCA suppression pool pH control. The tank volume requirement of 4000 gallons is established by the amount of boron at 8.0% by weight concentration required for the radiological dose analysis for post-LOCA suppression pool pH control. The tank volume requirement for reactivity control is encompassed by the requirement for post LOCA pH control. For reactivity control, the sodium (continued)
BFN-UNIT 3                    B 3.1-51                Revision 0, 29, 110, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.1 (continued)
REQUIREMENTS pentaborate solution concentration requirements (d 9.2% by weight) and the required quantity of Boron-10 (t 203 lbs) establish the tank volume requirement. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.7.2 SR 3.1.7.2 verifies the continuity of the explosive charges in the injection valves to ensure that proper operation will occur if required. An automatic continuity monitor may be used to continuously satisfy this requirement. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.7.3 SR 3.1.7.3 requires an examination of sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank for post-LOCA suppression pool pH control. This parameter is used as input to determine the volume requirements for SR 3.1.7.1. The concentration is dependent upon the volume of water and quantity of boron in the storage tank.
SR 3.1.7.3 must be performed according to the Surveillance Frequency Control Program or within 24 hours of when boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits.
(continued)
BFN-UNIT 3                    B 3.1-52                  Revision 0, 29, 110, 123 March 1, 2018
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.4 REQUIREMENTS (continued) SR 3.1.7.4 requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank. The concentration is dependent upon the volume of water and quantity of boron in the storage tank.
The sodium pentaborate solution (SPB) concentration is allowed to be > 9.2 weight percent provided the concentration and temperature of the sodium pentaborate solution are verified to be within the limits of Figure 3.1.7-1. This ensures that unwanted precipitation of the sodium pentaborate does not occur.
SR 3.1.7.4 must be performed according to the Surveillance Frequency Control Program or within 24 hours of when boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits.
SR 3.1.7.4 must be performed within 8 hours of discovery that the concentration is > 9.2 weight percent and every 12 hours thereafter until the concentration is verified to be d 9.2 weight percent. This Frequency is appropriate under these conditions taking into consideration the SLC System design capability still exists for vessel injection under these conditions and the low probability of the temperature and concentration limits of Figure 3.1.7-1 not being met.
(continued)
BFN-UNIT 3                    B 3.1-53            Revision 0, 29, 110, 116, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.5 REQUIREMENTS This Surveillance requires the amount of Boron-10 in the SLC solution tank to be determined periodically. The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water. Since the chemicals used have known Boron-10 quantities, the Boron-10 quantity in the sodium pentaborate solution formed can be calculated. This parameter is used as input to determine the volume requirements for reactivity control encompassed by SR 3.1.7.1. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.7.6 SR 3.1.7.6 requires verification that the SLC system conditions satisfy the following equation:
( C )( Q )(        E    )
                                                        > = 1.0
( 8.7 WT % )( 50 GPM )( 94 ATOM % )
C = sodium pentaborate solution weight percent concentration Q = SLC system pump flow rate in gpm E = Boron-10 atom percent enrichment in the sodium pentaborate solution To meet 10 CFR 50.62, the SLC System must have a minimum flow capacity and boron content equivalent in control capacity to 86 gpm of 13 weight percent natural sodium pentaborate solution. The purpose of this injection rate is to ensure that during an ATWS condition with the MSIVs closed, sufficient B-10 is injected into the RPV to bring the reactor subcritical (Hot Shutdown) prior to suppression pool temperature exceeding its heat capacity temperature limit. The atom percentage of natural B-10 is 19.8%. This equivalency requirement is met when the equation given above is satisfied.
(continued)
BFN-UNIT 3                    B 3.1-54                Revision 0, 29, 116, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.6 (continued)
REQUIREMENTS (continued) The equation can be satisfied by adjusting the solution concentration, pump flow rate or Boron-10 enrichment. If the results of the equation are < 1, the SLC System is no longer capable of shutting down the reactor with the margin described in Reference 2. As described in Reference 2, the BFN analysis assumes a flow capacity and boron content equivalent to 50 gpm of 8.7 weight percent and 94 atom percent B-10 enriched sodium pentaborate solution. This exceeds the requirement of 10 CFR 50.62, and the equation is adjusted to reflect the BFN requirements. The quantity of stored boron includes an additional margin (25%) beyond the amount needed to shut down the reactor to allow for possible imperfect mixing of the chemical solution in the reactor water, leakage, and the volume in other piping connected to the reactor system.
SR 3.1.7.6 must be performed according to the Surveillance Frequency Control Program or within 24 hours of when boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits.
(continued)
BFN-UNIT 3                  B 3.1-54a                Revision 0, 29, 116, 123 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.7 REQUIREMENTS (continued) Demonstrating that each SLC System pump develops a flow rate t 39 gpm at a discharge pressure t 1325 psig ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration and enrichment requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The pump flow rate of 39 gpm is based on the original licensing basis for SLC for an alternate reactivity insertion system.
SR 3.1.7.8 and SR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. Additionally, replacement charges shall be selected such that the age of charge in service shall not exceed five years from the manufacturer's assembly date.
(continued)
BFN-UNIT 3                    B 3.1-55                  Revision 29, 116, 123 Amendment No. 215 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.8 and SR 3.1.7.9 (continued)
REQUIREMENTS The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Demonstrating that all piping between the boron solution storage tank and the suction inlet to the injection pumps is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution. An acceptable method for verifying that the suction piping is unblocked is to pump from the storage tank to the storage tank. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.7.10 The enriched sodium pentaborate solution is made by combining stoichiometric quantities of borax and boric acid in demineralized water. Isotopic tests on these chemicals to verify the actual B-10 enrichment must be performed according to the Surveillance Frequency Control Program and after addition of boron to the SLC tank in order to ensure that the proper B-10 atom percentage is being used and SR 3.1.7.6 will be met. The sodium pentaborate enrichment must be calculated within 24 hours and verified by analysis within 30 days.
(continued)
BFN-UNIT 3                    B 3.1-56                        Revision 29, 123 Amendment No. 215 March 22, 2020
 
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.11 REQUIREMENTS (continued) SR 3.1.7.11 verifies that each valve in the system is in its correct position, but does not apply to the squib (i.e., explosive) valves. Verifying the correct alignment for manual, power operated, and automatic valves in the SLC System Flowpath provides assurance that the proper flow paths will exist for system operation. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position from the control room, or locally by a dedicated operator at the valve control. This is acceptable since the SLC System is a manually initiated system. This surveillance also does not apply to valves that are locked, sealed, or otherwise secured in position since they are verified to be in the correct position prior to locking, sealing or securing. This verification of valve alignment does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. 10 CFR 50.62.
: 2. NEDC-33860P, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate, Section 2.8.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 4. FSAR, Section 14.6.
BFN-UNIT 3                    B 3.1-57                  Revision 0, 29, 110, 123 March 22, 2020
 
SDV Vent and Drain Valves B 3.1.8 BASES (continued)
SURVEILLANCE      SR 3.1.8.1 REQUIREMENTS During normal operation, the SDV vent and drain valves should be in the open position (except when performing SR 3.1.8.2) to allow for drainage of the SDV piping. Verifying that each valve is in the open position ensures that the SDV vent and drain valves will perform their intended functions during normal operation. This SR does not require any testing or valve manipulation; rather, it involves verification that the valves are in the correct position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.8.2 During a scram, the SDV vent and drain valves should close to contain the reactor water discharged to the SDV piping. Cycling each valve through its complete range of motion (closed and open) ensures that the valve will function properly during a scram. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.8.3 SR 3.1.8.3 is an integrated test of the SDV vent and drain valves to verify total system performance. After receipt of a simulated or actual scram signal, the closure of the SDV vent and drain valves is verified. The closure time of 60 seconds after receipt of a scram signal is acceptable based on the (continued)
BFN-UNIT 3                          B 3.1-62                      Revision 0, 29, 123 March 22, 2020
 
SDV Vent and Drain Valves B 3.1.8 BASES SURVEILLANCE SR 3.1.8.3 (continued)
REQUIREMENTS bounding analysis for release of reactor coolant outside containment (Ref. 2). Similarly, after receipt of a simulated or actual scram reset signal, the opening of the SDV vent and drain valves is verified. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1 and the scram time testing of control rods in LCO 3.1.3 overlap this Surveillance to provide complete testing of the assumed safety function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 3.4.5.3.1.
: 2. FSAR, Section 14.6.5.
: 3. 10 CFR 50.67.
: 4. FSAR, Section 6.5.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                    B 3.1-63                        Revision 29, 123 Amendment No. 215 March 22, 2020
 
APLHGR B 3.2.1 BASES (continued)
ACTIONS          A.1 If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA and transient analyses may not be met. Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 2 hour Completion Time is sufficient to restore the APLHGR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification.
B.1 If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 23% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 23% RTP in an orderly manner and without challenging plant systems.
SURVEILLANCE      SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 12 hours after THERMAL POWER is t 23% RTP and periodically thereafter. They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER t 23% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                          B 3.2-4                      Revision 110, 123 Amendment No. 213 March 22, 2020
 
MCPR B 3.2.2 BASES (continued)
SURVEILLANCE      SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours after THERMAL POWER is t 23% RTP and then periodically thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER t 23% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.2.2.2 Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis. SR 3.2.2.2 determines the actual scram speed distribution and compares it with the assumed distribution. The MCPR operating limit is determined based either on the applicable limit associated with scram times of LCO 3.1.4, Control Rod Scram Times, or the nominal scram times. The scram speed-dependent MCPR limits are contained in the COLR. This determination must be performed within 72 hours after each set of control rod scram time tests required by SR 3.1.4.1 and SR 3.1.4.2 because the effective scram speed distribution may change during the cycle. The 72-hour Completion Time is acceptable due to the relatively minor changes in the actual control rod scram speed distribution expected during the fuel cycle.
(continued)
BFN-UNIT 3                          B 3.2-10                  Revision 25, 110, 123 Amendment No. 213 March 22, 2020
 
LHGR B 3.2.3 BASES (continued)
LCO              The LHGR is a basic assumption in the fuel design analysis.
The fuel has been designed to operate at rated core power with sufficient design margin to the LHGR calculated to cause a 1%
fuel cladding plastic strain. The operating limit to accomplish this objective is specified in the COLR.
Additional LHGR operating limits adjustments may be provided in the COLR to support analyzed equipment out-of-service operation.
APPLICABILITY    The LHGR limits are derived from fuel design analysis that is limiting at high power level conditions. At core thermal power levels < 23% RTP, the reactor is operating with a substantial margin to the LHGR limits and, therefore, the Specification is only required when the reactor is operating at t 23% RTP.
(continued)
BFN-UNIT 3                        B 3.2-13a              Revision 25, 61, 106, 110 March 1, 2018
 
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE      11. Manual Scram (continued)
SAFETY ANALYSES, LCO, and        Two channels of Manual Scram with one channel in each APPLICABILITY    manual scram trip system are available and required to be OPERABLE in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
: 12. RPS Channel Test Switches There are four RPS Channel Test Switches, one associated with each of the four automatic scram logic channels (A1, A2, B1, and B2). These keylock switches allow the operator to test the OPERABILITY of each individual logic channel without the necessity of using a scram function trip. When the RPS Channel Test Switch is placed in test, the associated scram logic channel is deenergized and OPERABILITY of the channel's scram contactors can be confirmed. The RPS Channel Test Switches are not specifically credited in the accident analysis. However, because the Manual Scram Function at Browns Ferry Nuclear Plant is not configured the same as the generic model in Reference 9, the RPS Channel Test Switches are included in the analysis in Reference 11.
(continued)
BFN-UNIT 3                        B 3.3-27                          Revision 123 Amendment No. 213 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.1 REQUIREMENTS (continued) Performance of the CHANNEL CHECK 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 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
(continued)
BFN-UNIT 3                    B 3.3-37                            Revision 0, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.2 REQUIREMENTS (continued) To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
A restriction to satisfying this SR when < 23% RTP is provided that requires the SR to be met only at t 23% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when
            < 23% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At t 23% RTP, the Surveillance is required to have been satisfactorily performed in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 23% if the Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 23% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
(continued)
BFN-UNIT 3                    B 3.3-38                    Revision 110, 123 Amendment No. 213 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.3 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.
Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
As noted, SR 3.3.1.1.3 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.3-39                            Revision 123 Amendment No. 213 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.5 and SR 3.3.1.1.6 (continued)
REQUIREMENTS If overlap for a group of channels is not demonstrated (e.g.,
IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel(s) declared inoperable. Only those appropriate channels that are required in the current MODE or condition should be declared inoperable.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.1.7 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.1.8 and SR 3.3.1.1.12 Deleted.
SR 3.3.1.1.9, SR 3.3.1.1.10 and SR 3.3.1.1.13 A 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. For the APRM Simulated Thermal Power - High Function, SR 3.3.1.1.13 also includes calibrating the associated recirculation loop flow channel. For MSIV - Closure, SDV Water Level - High (Float Switch), and TSV - Closure Functions, SR 3.3.1.1.13 includes physical inspection and actuation of the switches.
(continued)
BFN-UNIT 3                    B 3.3-41                          Revision 0, 123 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.9, SR 3.3.1.1.10 and SR 3.3.1.1.13 (continued)
REQUIREMENTS A Note to SR 3.3.1.1.9 and SR 3.3.1.1.13 states that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) and the 1000 MWD/T LPRM calibration against the TIPs (SR 3.3.1.1.7). A second Note for SR 3.3.1.1.9 is provided that requires the IRM SRs to be performed within 12 hours of entering MODE 2 from MODE 1. Testing of the MODE 2 IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2.
Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.3-42                          Revision 123 Amendment No. 215 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.11 REQUIREMENTS (continued) (Deleted)
SR 3.3.1.1.14 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel.
The functional testing of control rods (LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM and OPRM trip conditions at the 2-out-of-4 voter channel inputs to check all combinations of two tripped inputs to the 2-out-of-4 logic in the voter channels and APRM related redundant RPS relays.
(continued)
BFN-UNIT 3                    B 3.3-43                          Revision 0, 123 Amendment No. 213, 215, 221 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 REQUIREMENTS (continued) This SR ensures that scrams initiated from the Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is t 26% RTP. This involves calibration of the bypass channels (PIS-1-81A, PIS-1-81B, PIS-1-91A, and PIS-1-91B). Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.
If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at t 26% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are enabled), this SR is met and the channel is considered OPERABLE.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.1.16 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
(continued)
BFN-UNIT 3                    B 3.3-44                    Revision 110, 123 Amendment No. 215 March 22, 2020
 
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.16 (continued)
REQUIREMENTS (continued) The SR 3.3.1.1.16 Surveillance Frequency is controlled under the Surveillance Frequency Control Panel. The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing - applicable to Function 2.b only), the 2-out-of-4 voter channels, and the interface connections into the RPS trip systems from the voter channels.
Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
(NOTE: The actual voting logic of the 2-out-of-4 Voter Function is tested as part of SR 3.3.1.1.14.) A Note for SR 3.3.1.1.16 is provided that requires the APRM Function 2.a SR to be performed within 12 hours of entering MODE 2 from MODE 1.
Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. This Note allows entry into MODE 2 from MODE 1 if the associated frequency is not met per SR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
SR 3.3.1.1.17 (Deleted)
(continued)
BFN-UNIT 3                    B 3.3-45                        Revision 110, 123 Amendment No. 215 March 22, 2020
 
SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.1 (continued)
REQUIREMENTS 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
SR 3.3.1.2.2 To provide adequate coverage of potential reactivity changes in the core when the fueled region encompasses more than one SRM, one SRM is required to be OPERABLE in the quadrant where CORE ALTERATIONS are being performed, and the other OPERABLE SRM must be in an adjacent quadrant (continued)
BFN-UNIT 3                    B 3.3-54                              Revision 123 Amendment No. 213 March 22, 2020
 
SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.2 (continued)
REQUIREMENTS containing fuel. Note 1 states that the SR is required to be met only during CORE ALTERATIONS. It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a review of plant logs to ensure that SRMs required to be OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE. In the event that only one SRM is required to be OPERABLE (when the fueled region encompasses only one SRM), per Table 3.3.1.2-1, footnote (b),
only the a. portion of this SR is required. Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE SRM. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.2.3 Deleted.
SR 3.3.1.2.4 This Surveillance consists of a verification of the SRM instrument readout to ensure that the SRM reading is greater than a specified minimum count rate, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR.
Therefore, allowances are made for loading sufficient "source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate.
(continued)
BFN-UNIT 3                    B 3.3-55                            Revision 123 Amendment No. 213 March 22, 2020
 
SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.4 (continued)
REQUIREMENTS To accomplish this, the SR is modified by a Note that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated core quadrant.
With four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be critical.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.2.5 and SR 3.3.1.2.6 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly.
SR 3.3.1.2.5 is required in MODE 5, and ensures that the channels are OPERABLE while core reactivity changes could be in progress. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.3-56                            Revision 123 Amendment No. 213 March 22, 2020
 
SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.5 and SR 3.3.1.2.6 (continued)
REQUIREMENTS SR 3.3.1.2.6 is required in MODE 2 with IRMs on Range 2 or below, and in MODES 3 and 4. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Verification of the signal to noise ratio also ensures that the detectors are inserted to an acceptable operating level. In a fully withdrawn condition, the detectors are sufficiently removed from the fueled region of the core to essentially eliminate neutrons from reaching the detector. Any count rate obtained while the detectors are fully withdrawn is assumed to be "noise" only.
The Note to SR 3.3.1.2.6 allows the Surveillance to be delayed until entry into the specified condition of the Applicability (THERMAL POWER decreased to IRM Range 2 or below). The SR must be performed within 12 hours after IRMs are on Range 2 or below. The allowance to enter the Applicability with the Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels.
Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour allowance is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.
(continued)
BFN-UNIT 3                      B 3.3-57                            Revision 123 Amendment No. 213 March 22, 2020
 
SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.7 REQUIREMENTS (continued) Performance of a CHANNEL CALIBRATION verifies the performance of the SRM detectors and associated circuitry.
The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the system or component status. The neutron detectors are excluded from the CHANNEL CALIBRATION (Note 1) because they cannot readily be adjusted. The detectors are fission chambers that are designed to have a relatively constant sensitivity over the range and with an accuracy specified for a fixed useful life. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Note 2 to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the Applicability. The SR must be performed in MODE 2 within 12 hours of entering MODE 2 with IRMs on Range 2 or below. The allowance to enter the Applicability with the Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level.
In this event, the 12 hour allowance is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e.,
satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.
REFERENCES  1. FSAR, Section 7.5.4.
BFN-UNIT 3                      B 3.3-58                            Revision 123 Amendment No. 213 March 22, 2020
 
Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE Condition entered and Required Actions taken. This Note is REQUIREMENTS based on the reliability analysis (Ref. 9) assumption of the (continued) average time required to perform a channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary.
SR 3.3.2.1.1 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control System input.
Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function.
The CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs. This test is performed as soon as possible after the applicable conditions are entered. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until 1 hour after (continued)
BFN-UNIT 3                    B 3.3-69                            Revision 123 Amendment No. 213 March 22, 2020
 
Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES SURVEILLANCE  SR 3.3.2.2.2 REQUIREMENTS (continued)  A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.2.2.3 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                      B 3.3-82                            Revision 123 Amendment No. 215 March 22, 2020
 
PAM Instrumentation B 3.3.3.1 BASES LCO          3. Suppression Pool Water Level (continued) (LI-64-159A and XR-64-159)
Suppression pool water level is a Category 1 variable provided to detect a breach in the reactor coolant pressure boundary (RCPB). This variable is also used to verify and provide long term surveillance of ECCS function. The wide range suppression pool water level measurement provides the operator with sufficient information to assess the status of both the RCPB and the water supply to the ECCS. The wide range water level indicators monitor the suppression pool water level from two feet from the bottom of the pool to five feet above normal water level. Two wide range suppression pool water level signals are transmitted from separate differential pressure transmitters and are continuously recorded and displayed on one recorder and one indicator in the control room. The recorder and indicator are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
: 4. Drywell Pressure (PI-64-67B, XR-64-50, PI-64-160A, and XR-64-159)
Drywell pressure is a Category 1 variable provided to detect breach of the RCPB and to verify ECCS functions that operate to maintain RCS integrity. Two different ranges of drywell pressure channels (normal and wide range) receive signals that are transmitted from separate pressure transmitters and are continuously recorded and displayed on two control room recorders and two control room indicators. These recorders and indicators are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
(continued)
BFN-UNIT 3                    B 3.3-88                      Amendment No. 213 September 03, 1998
 
PAM Instrumentation B 3.3.3.1 BASES LCO          5. Primary Containment Area Radiation (High Range)
(continued) (RR-90-272 and RR-90-273)
Primary containment area radiation (high range) is provided to monitor the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans. Two high range primary containment area radiation signals (RM-90-272A and RM-90-273A) are transmitted from separate radiation detectors and are continuously recorded and displayed on two control room recorders. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
: 6. Primary Containment Isolation Valve (PCIV) Position PCIV position is provided for verification of containment integrity. In the case of PCIV position, the important information is the isolation status of the containment penetration. The LCO requires one channel of valve position indication in the control room to be OPERABLE for each active PCIV in a containment penetration flow path, i.e., two total channels of PCIV position indication for a penetration flow path with two active valves. For containment penetrations with only one active PCIV having control room indication, Note (b) requires a single channel of valve position indication to be OPERABLE. This is sufficient to redundantly verify the isolation status of each isolable penetration via indicated status of the active valve, as applicable, and prior knowledge of passive valve or system boundary status. If a penetration flow path is isolated, position indication for the PCIV(s) in the associated penetration flow path is not needed to determine status. Therefore, the position indication for valves in an isolated penetration flow path is not required to be OPERABLE.
(continued)
BFN-UNIT 3                    B 3.3-89                                Revision 2 March 02, 1999
 
Backup Control System B 3.3.3.2 BASES ACTIONS      B.1 (continued)
If the Required Action and associated Completion Time of Condition A are not met, 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. The allowed Completion Time is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.3.3.2.1 REQUIREMENTS SR 3.3.3.2.1 verifies each required Backup Control System transfer switch and control circuit performs the intended function. This verification is performed from the backup control panel and locally, as appropriate. Operation of the equipment from the backup control panel is not necessary. The Surveillance can be satisfied by performance of a continuity check. This will ensure that if the control room becomes inaccessible, the plant can be placed and maintained in MODE 3 from the backup control panel and the local control stations. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                  B 3.3-103                            Revision 123 Amendment No. 215 March 22, 2020
 
Backup Control System B 3.3.3.2 BASES SURVEILLANCE SR 3.3.3.2.2 REQUIREMENTS (continued) CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies the channel responds to measured parameter values with the necessary range and accuracy.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.3.2.3 Deleted REFERENCES  1. 10 CFR 50, Appendix A, GDC 19.
: 2. FSAR Section 7.18.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                  B 3.3-104                          Revision 123 Amendment No. 215 March 22, 2020
 
Backup Control System B 3.3.3.2 Table B 3.3.3.2-1 (Page 1 of 4)
Backup Control System Instrumentation and Controls NUMBER FUNCTION                                      REQUIRED Instrument Parameter
: 1. Reactor Water Level Indication (3-LI-3-46A, -46B)                        1
: 2. Reactor Pressure Indication (3-PI-3-79)                                  1
: 3. Suppression Pool Temperature Indication (3-TI-64-55B)                    1
: 4. Suppression Pool Level Indication (3-LI-64-54B)                          1
: 5. Drywell Pressure Indication (3-PI-64-50)                                1
: 6. RHR Flow Indication (3-FI-74-79)                                        1
: 7. RCIC Flow Indication (3-FIC-71-36B)                                  1, note a
: 8. RCIC Turbine Speed Indication (3-SI-71-42B)                              1
: 9. Drywell Temperature Indication (3-TI-64-52AA)                            1
: 10. RHRSW Header Pressure (0-PI-23-58/3, -59/3)                              2 Transfer/Control Parameter
: 11. Main Steam Relief Valve (MSRV) Transfer & Control                    3, note b (3-XS-1-22, -5, -41, -34) (3-HS-1-22C, -5C, -41C, -34C)
: 12. Main Steam Isolation Valve (MSIV) Transfer & Control (Closure)        4, note c inboard: (3-XS-1-14, -26, -37, -51)
(3-HS-1-14C, -26C, -37C, -51C);
outboard: (3-XS-1-15, -27, -38, -52)
(3-HS-1-15C, -27C, -38C, -52C)
: 13. Main Steam Drain Line Isolation Valves                                1, note d (3-XS-1-55, -56) (3-HS-1-55C, -56C) note a:    RCIC flow indication may be obtained from the Flow Indicating Controller.
note b:    1 required for each of 3 MSRVs.
note c:    1 MSIV required per penetration, may be either inboard valve or outboard valve.
note d:    1 Main Steam Drain Line isolation valve required, may be either inboard valve or outboard valve.
note o:    Note not used.
note p:    Note not used.
BFN-UNIT 3                                    B 3.3-105                                Revision 4 April 09, 1999
 
EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE      Turbine Control Valve Fast Closure, Trip Oil Pressure - Low SAFETY ANALYSES, (PS-47-142, PS-47-144, PS-47-146, and PS-47-148)
LCO, and APPLICABILITY    Fast closure of the TCVs during a generator load rejection (continued)    results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.
Therefore, an RPT is initiated on TCV Fast Closure, Trip Oil Pressure - Low in anticipation of the transients that would result from the closure of these valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL and LHGR limits are not exceeded during the worst case transient.
Fast closure of the TCVs is determined by measuring the electrohydraulic control fluid pressure at each control valve.
There is one pressure switch associated with each control valve, and the signal from each switch is assigned to a separate trip channel. The logic for the TCV Fast Closure, Trip Oil Pressure - Low Function is such that two or more TCVs must be closed (pressure switch trips) to produce an EOC-RPT. This Function must be enabled at THERMAL POWER t 26% RTP.
This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this function. To consider this function OPERABLE, bypass of the function must not occur when bypass valves are open. Four channels of TCV Fast Closure, Trip Oil Pressure - Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TCV Fast Closure, Trip Oil Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure.
(continued)
BFN-UNIT 3                        B 3.3-113                        Revision 25, 110 Amendment No. 213 March 1, 2018
 
EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE      Turbine Control Valve Fast Closure, Trip Oil Pressure - Low SAFETY ANALYSES, (PS-47-142, PS-47-144, PS-47-146, and PS-47-148)
LCO, and        (continued)
APPLICABILITY This protection is required consistent with the safety analysis whenever THERMAL POWER is t 26% RTP. Below 26% RTP, the Reactor Vessel Steam Dome Pressure - High and the APRM Fixed Neutron Flux - High Functions of the RPS are adequate to maintain the necessary safety margins.
ACTIONS          A Note has been provided to modify the ACTIONS related to EOC-RPT 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 EOC-RPT 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 EOC-RPT instrumentation channel.
(continued)
BFN-UNIT 3                        B 3.3-114                              Revision 110 Amendment No. 213 March 1, 2018
 
EOC-RPT Instrumentation B 3.3.4.1 BASES ACTIONS      A.1 (continued)
With one or more channels inoperable, but with EOC-RPT trip capability maintained (refer to Required Actions B.1 and B.2 Bases), the EOC-RPT System is capable of performing the intended function. However, the reliability and redundancy of the EOC-RPT instrumentation is reduced such that a single failure in the remaining trip system could result in the inability of the EOC-RPT System to perform the intended function.
Therefore, only a limited time is allowed to restore compliance with the LCO. Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of an EOC-RPT, 72 hours is provided to restore the inoperable channels (Required Action A.1) or apply the EOC-RPT inoperable MCPR and LHGR limits. Alternately, the inoperable channels 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 breaker, since this may not adequately compensate for the inoperable breaker (e.g., the breaker may be inoperable such that it will not open). 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 RPT, or if the inoperable channel is the result of an inoperable breaker),
Condition C must be entered and its Required Actions taken.
(continued)
BFN-UNIT 3                    B 3.3-115                          Revision 25, 61 Amendment No. 213 December 7, 2010
 
ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE      4.e, 4.f, 5.e, 5.f. Core Spray and Low Pressure Coolant SAFETY ANALYSES, Injection Pump Discharge Pressure - High LCO, and        (PS-75-7, 16, 35, 44 and PS-74-8A and B, -19A and B, -31A APPLICABILITY    and B, -42A and B)
(continued)
The Pump Discharge Pressure - High signals from the CS and LPCI pumps are used as permissives for ADS initiation, indicating that there is a source of low pressure cooling water available once the ADS has depressurized the vessel. Pump Discharge Pressure - High is one of the Functions assumed to be OPERABLE and capable of permitting ADS initiation during the events analyzed in Reference 2 with an assumed HPCI failure. For these events the ADS depressurizes the reactor vessel so that the low pressure ECCS can perform the core cooling functions. This core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
Pump discharge pressure signals are initiated from twelve pressure switches, two on the discharge side of each RHR (LPCI) pump and one on the discharge side of each CS pump.
There are two ADS low pressure ECCS pump permissives in each trip system. Each of these permissives receives inputs from all four RHR (LPCI) pumps (different signals for each permissive) and two CS pumps, one from each subsystem (different pumps for each permissive). In order to generate an ADS permissive in one trip system, it is necessary that only one LPCI pump or two CS pumps (CS pumps A or B and either C or D) indicate the high discharge pressure condition. The Pump Discharge Pressure - High Allowable Value is less than the pump discharge pressure when the pump is operating in a full flow mode and high enough to avoid any condition that results in a discharge pressure permissive when the CS and LPCI pumps are aligned for injection and the pumps are not running.
(continued)
BFN-UNIT 3                          B 3.3-160                  Amendment No. 213 September 03, 1998
 
ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE      4.e, 4.f, 5.e, 5.f. Core Spray and Low Pressure Coolant SAFETY ANALYSES, Injection Pump Discharge Pressure - High LCO, and        (PS-75-7, 16, 35, 44 and PS-74-8A and B, -19A and B, -31A APPLICABILITY    and B, -42A and B) (continued)
The actual operating point of this function is not assumed in any transient or accident analysis. However, this function is indirectly assumed to operate (in Reference 2) to provide the ADS permissive to depressurize the RCS to allow the ECCS low pressure systems to operate.
Twelve channels of Core Spray and Low Pressure Coolant Injection Pump Discharge Pressure - High Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Four CS channels associated with CS pumps A through D and eight LPCI channels associated with LPCI pumps A through D are required for both trip systems.
Refer to LCO 3.5.1 for ADS Applicability Bases.
4.g, 5.g. Automatic Depressurization System High Drywell Pressure Bypass Timer One of the signals required for ADS initiation is Drywell Pressure - High. However, if the event requiring ADS initiation occurs outside the drywell (e.g., main steam line break outside containment), a high drywell pressure signal may never be present. Therefore, the Automatic Depressurization System High Drywell Pressure Bypass Timer is used to bypass the Drywell Pressure - High Function after a certain time period has elapsed. Operation of the Automatic Depressurization System High Drywell Pressure Bypass Timer Function is not assumed in any accident analysis. The instrumentation was installed to meet requirements of NUREG-0737, Item II.K.3.18 (Ref. 6) and is retained in the TS because ADS is part of the primary success path for mitigation of a DBA.
(continued)
BFN-UNIT 3                          B 3.3-161                  Amendment No. 213 September 03, 1998
 
ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE      4.g, 5.g. Automatic Depressurization System High Drywell SAFETY ANALYSES, Pressure Bypass Timer (continued)
LCO, and APPLICABILITY    There are four Automatic Depressurization System High Drywell Pressure Bypass Timer relays, two in each of the two ADS trip systems. The Allowable Value for the Automatic Depressurization System High Drywell Pressure Bypass Timer is chosen to ensure that there is still time after depressurization for the low pressure ECCS subsystems to provide adequate core cooling.
Two channels in each trip system of the Automatic Depressurization System High Drywell Pressure Bypass Timer Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Refer to LCO 3.5.1 for ADS Applicability Bases.
ACTIONS          A Note has been provided to modify the ACTIONS related to ECCS 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 ECCS instrumentation channels provide appropriate compensatory measures for separate inoperable Condition entry for each inoperable ECCS instrumentation channel.
(continued)
BFN-UNIT 3                        B 3.3-162                      Amendment No. 213 September 03, 1998
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE      3.d., 3.e., 3.f., 3.g., 4.d., 4.e., 4.f., 4.g. Area Temperature - High SAFETY ANALYSES, (TS-71-2A-H, J-N, P, R, S and TS-73-2A-H, J-N, P, R, S)
LCO, and APPLICABILITY    Area Temperature Functions are provided to detect a leak from (continued)    the associated system steam piping. The isolation occurs when a very small leak has occurred and is diverse to the high flow instrumentation. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. These Functions are not assumed in any FSAR transient or accident analysis, since bounding analyses are performed for large breaks such as recirculation or MSL breaks.
Area Temperature - High signals are initiated from bimetallic temperature switches that are appropriately located to protect the system that is being monitored. Four instruments monitor each area. HPCI and RCIC each have sixteen total channels of Area Temperature - High Function available. All of which are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
The Allowable Values are set low enough to detect a leak equivalent to 25 gpm.
These Functions isolate the Group 4 and 5 valves, as appropriate.
(continued)
BFN-UNIT 3                          B 3.3-207                      Amendment No. 213 September 03, 1998
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE      Reactor Water Cleanup System Isolation SAFETY ANALYSES, LCO, and        5.a., 5.b., 5.c., 5.d., 5.e., 5.f. Area Temperature - High APPLICABILITY    (TIS-69-834A-D, 835A-D, 836A-D, 837A-D, 838A-D, 839A-D)
(continued)
RWCU Area Temperature Functions are provided to detect a leak from the RWCU System. The isolation occurs even when very small leaks have occurred. If the small leak continues without isolation, offsite dose limits may be reached. Credit for these instruments is not taken in any transient or accident analysis in the FSAR, since bounding analyses are performed for large breaks such as recirculation or MSL breaks.
Area temperature signals are initiated from temperature elements that are located in the areas monitored. Four sensors in each of the six monitored areas are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
The Area Temperature - High Allowable Values are set based on the maximum abnormal operating temperature for each area.
These Functions isolate the Group 3 valves.
(continued)
BFN-UNIT 3                        B 3.3-208                    Amendment No. 213 September 03, 1998
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE      5.g. SLC System Initiation SAFETY ANALYSES, LCO, and        The isolation of the RWCU System is required when the SLC APPLICABILITY    System has been initiated to prevent dilution and removal of the (continued)    boron solution by the RWCU System (Ref. 4). An isolation signal for both RWCU isolation valves is initiated when the SLC pump start handswitch is not in the stop position.
There is no Allowable Value associated with this Function since the channels are mechanically actuated based solely on the position of the SLC System initiation switch.
The SLC System Initiation Function is required to be OPERABLE in MODES 1 and 2, because these are the only MODES where the reactor can be critical, and in MODE 3 because this MODE uses the SLC System sodium pentaborate as a buffering solution to maintain the pH level at or above 7 in the suppression pool in the event of a LOCA. These MODES are consistent with the Applicability for the SLC System (LCO 3.1.7).
As noted (footnote (a) to Table 3.3.6.1-1), the SLC initiation signal provides input to the isolation logic for both RWCU isolation valves.
5.h. Reactor Vessel Water Level - Low, Level 3 (LIS-3-203A-D)
Low 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. Therefore, isolation of some interfaces with the reactor vessel occurs to isolate the potential sources of a break. The isolation of the RWCU System on Level 3 supports actions to ensure that the fuel peak cladding (continued)
BFN-UNIT 3                        B 3.3-209                              Revision 98 Amendment 213 February 5, 2016
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.1 REQUIREMENTS (continued) Performance of the CHANNEL CHECK 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 of 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.3-222                              Revision 123 Amendment No. 213 March 22, 2020
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.2 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.
Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.6.1.3 and SR 3.3.6.1.4 Deleted SR 3.3.6.1.5 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                  B 3.3-223                            Revision 123 Amendment No. 213 March 22, 2020
 
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.6 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel. The system functional testing performed on PCIVs in LCO 3.6.1.3 overlaps this Surveillance to provide complete testing of the assumed safety function. The LOGIC SYSTEM FUNCTIONAL TEST shall include a calibration of time delay relays and timers necessary for proper functioning of the logic.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 6.5.
: 2. FSAR, Chapter 14.
: 3. NEDO-31466, "Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
: 4. FSAR, Section 4.9.3.
: 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.
(continued)
BFN-UNIT 3                    B 3.3-224                          Revision 123 Amendment No. 215 March 22, 2020
 
CREV System Instrumentation B 3.3.7.1 BASES (continued)
SURVEILLANCE      As noted (Note 1) at the beginning of the SRs, the SRs for REQUIREMENTS      each CREV System instrumentation Function are located in the SRs column of Table 3.3.7.1-1.
The Surveillances are modified by a Note (Note 2) 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 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 (Refs. 3 and 4) 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.
The Surveillances are modified by a third Note (Note 3) to indicate that for Functions 3 and 4, when a channel is placed in an inoperable status solely for performance of a CHANNEL CALIBRATION or maintenance, entry into associated Conditions and Required Actions may be delayed for up to 24 hours provided the downscale trip of the inoperable channel is placed in the tripped condition. Upon completion of the Surveillance or maintenance, or expiration of the 24 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
(continued)
BFN-UNIT 3                        B 3.3-252                    Amendment No. 213 September 03, 1998
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS    A.1 and A.2 (continued)
Condition C or D, as applicable, must be entered immediately.
The 15 day allowable out of service time is justified based on the two-out-of-three permissive logic scheme provided for these relays. If the inoperable relay channel cannot be restored to OPERABLE status within the allowable out of service time, the degraded voltage relay channel must be placed in the tripped condition per Required Action A.2. 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 F must be entered and its Required Action taken.
B.1 With two or more degraded voltage relay channels or one or more associated timers inoperable on one or more shutdown boards, the Function is not capable of performing the intended function. Required Action B.1 provides a 10 day allowable out of service time provided the loss of voltage relay channels on the affected shutdown board(s) are OPERABLE.
The 10 day allowable out of service time is justified since the loss of voltage relay channels on the same shutdown board are independent of the degraded voltage relay channel(s) and will continue to function and start the diesel generators on a complete loss of voltage. If the inoperable channel(s) cannot (continued)
BFN-UNIT 3                    B 3.3-262                            Revision 118 Amendment No. 213 December 12, 2019
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS    B.1 (continued) be restored to OPERABLE status within the allowable out of service time, the channel(s) must be placed in the tripped condition per Required Action B.1. Placing the inoperable channel(s) 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(s) in trip (e.g., as in the case where placing the channel(s) in trip would result in a DG initiation), Condition F must be entered and its Required Action taken.
C.1 With one or more loss of voltage relay channels inoperable on one or more shutdown boards, the Function is not capable of performing the intended function. Required Action C.1 provides a 10 day allowable out of service time provided two or more degraded voltage relay channels and associated timers on the affected shutdown board(s) are OPERABLE. The 10 day allowable out of service time is justified since the degraded voltage relay channels on the same shutdown board are independent of the loss of voltage relay channels and will continue to function and start the diesel generators on a complete loss of voltage. If the inoperable channels cannot be restored to OPERABLE status within the allowable out of service time, the channel(s) must be placed in the tripped condition per Required Action C.1. Placing the inoperable channel(s) 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(s) in trip (e.g., as in the case where placing the channel(s) in trip would result in a DG initiation), Condition F must be entered and its Required Action taken.
(continued)
BFN-UNIT 3                    B 3.3-263                            Revision 118 Amendment No. 213 December 12, 2019
 
Recirculation Loops Operating B 3.4.1 BASES (continued)
SURVEILLANCE      SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 70% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is
                  < 70% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.
The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered inoperable. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                          B 3.4-9                          Revision 0, 123 Amendment No. 221 March 22, 2020
 
Jet Pumps B 3.4.2 BASES SURVEILLANCE SR 3.4.2.1 (continued)
REQUIREMENTS Individual jet pumps in a recirculation loop normally do not have the same flow. The unequal flow is due to the drive flow manifold, which does not distribute flow equally to all risers.
The flow (or jet pump diffuser to lower plenum differential pressure) pattern or relationship of one jet pump to the loop average is repeatable. An appreciable change in this relationship is an indication that increased (or reduced) resistance has occurred in one of the jet pumps. This may be indicated by an increase in the relative flow for a jet pump that has experienced beam cracks.
The deviations from normal are considered indicative of a potential problem in the recirculation drive flow or jet pump system (Ref. 2). Normal flow ranges and established jet pump flow and differential pressure patterns are established by plotting historical data as discussed in Reference 2.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by two Notes. Note 1 allows this Surveillance not to be performed until 4 hours after the associated recirculation loop is in operation, since these checks can only be performed during jet pump operation. The 4 hours is an acceptable time to establish conditions appropriate for data collection and evaluation.
(continued)
BFN-UNIT 3                    B 3.4-15                          Revision 0, 123 March 22, 2020
 
S/RVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 4.4.6.
: 2. FSAR, Section 14.5.1.
: 3. ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code).
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                  B 3.4-22                    Revision 0, 81, 123 Amendment No. 215 March 22, 2020
 
RCS Operational LEAKAGE B 3.4.4 BASES ACTIONS      C.1 and C.2 (continued)
If any Required Action and associated Completion Time of Condition A or B is not met or if pressure boundary LEAKAGE exists, 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 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 safety systems.
SURVEILLANCE SR 3.4.4.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE. Leakage detection instrumentation is discussed in more detail in the Bases for LCO 3.4.5, "RCS Leakage Detection Instrumentation." Sump level and flow rate are typically monitored to determine actual LEAKAGE rates; however, other methods may be used to quantify LEAKAGE. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.4-28                        Revision 0, 123 March 22, 2020
 
RCS Operational LEAKAGE B 3.4.4 BASES (continued)
REFERENCES        1. 10 CFR 50.2.
: 2. 10 CFR 50.55a(c).
: 3. 10 CFR 50, Appendix A, GDC 55.
: 4. GEAP-5620, "Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws," April 1968.
: 5. NUREG-75/067, "Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping in Boiling Water Reactors,"
October 1975.
: 6. FSAR, Section 4.10.3.2.
: 7. Deleted.
: 8. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                        B 3.4-29                        Revision 0, 123 March 22, 2020
 
RCS Leakage Detection Instrumentation B 3.4.5 BASES (continued)
SURVEILLANCE      SR 3.4.5.1 REQUIREMENTS This SR is for the performance of a CHANNEL CHECK of the required primary containment atmospheric monitoring system instrumentation. The check gives reasonable confidence that the channel is operating properly. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.5.2 This SR is for the performance of a CHANNEL FUNCTIONAL TEST of the required primary containment atmospheric monitoring system instrumentation. The test ensures that the monitors can perform their function in the desired manner. The test also verifies the alarm setpoint and relative accuracy of the instrument string. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.5.3 This SR is for the performance of a CHANNEL CALIBRATION of required drywell floor drain sump flow integrator instrumentation channels. The calibration verifies the accuracy of the instrument string. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                          B 3.4-35                          Revision 0, 123 March 22, 2020
 
RCS Leakage Detection Instrumentation B 3.4.5 BASES SURVEILLANCE SR 3.4.5.4 REQUIREMENTS (continued) This SR is for the performance of a CHANNEL CALIBRATION of required leakage detection system instrumentation channels.
The calibration verifies the accuracy of the instrument string.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 30.
: 2. FSAR, Section 4.10.3.
: 3. GEAP-5620, "Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws," April 1968.
: 4. NUREG-75/067, "Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping in Boiling Water Reactors,"
October 1975.
: 5. FSAR, Section 4.10.3.2.
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                    B 3.4-36                            Revision 123 Amendment No. 215 March 22, 2020
 
RCS Specific Activity B 3.4.6 BASES (continued)
SURVEILLANCE      SR 3.4.6.1 REQUIREMENTS This Surveillance is performed to ensure iodine remains within limit during normal operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that requires this Surveillance to be performed only in MODE 1 because the level of fission products generated in other MODES is much less.
REFERENCES        1. 10 CFR 50.67.
: 2. FSAR, Section 14.6.5.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                        B 3.4-41                    Revision 0, 29, 123 March 22, 2020
 
RCS P/T Limits B 3.4.9 BASES BACKGROUND  The P-T curves incorporate a fluence calculated in accordance (continued) with GE Licensing Topical Report NEDC-32983P (Ref. 10),
which has been approved by the NRC and is in compliance with Regulatory Guide 1.190 (Ref. 11). The fluence represents an Extended Power Uprate (EPU) for the rated power of 3952 MWt, and is conservatively applied for the rated power of 3458 MWt.
The 1998 Edition of the ASME Section XI Boiler and Pressure Vessel Code including 2000 Addenda was used in accordance with 10 CFR 50.55a.
Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.
The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel.
(continued)
BFN-UNIT 3                      B 3.4-55a                          Revision 26, 99 February 5, 2016
 
RCS P/T Limits B 3.4.9 BASES BACKGROUND  Composite P-T curves were generated for each of the Pressure (continued) Test and Core Not Critical conditions by enveloping the most restrictive P-T limits from the separate beltline, upper vessel, and closure flange assembly P-T limits. A separate Bottom Head Limits (CRD Nozzle) curve is also individually included with the composite curve for the Pressure Test and Core Not Critical condition. This refinement can minimize heating requirements prior to pressure testing. A composite P-T curve was also generated for the Core Critical condition by enveloping the most restrictive P-T requirements from the separate beltline, upper vessel, bottom head, and closure flange assembly P-T limits. These curves, as previously noted, are contained in the Technical Specifications as Figures 3.4.9-1 and 3.4.9-2.
For the Core Not Critical and the Core Critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 100&deg;F/hour or less for which the curves are applicable.
However, the Core Not Critical and the Core Critical curves were also developed to bound transients defined on the RPV thermal cycle diagram and the nozzle thermal cycle diagrams.
For the Hydrostatic Pressure and Leak Test curve, a coolant heatup and cooldown temperature rate of 15&deg;F/hour or less must be maintained at all time. The lowest service temperature (LST) for the bolting material is 70&deg;F. The highest RTNDT in the closure flange region is 23.1&deg;F, for the vertical electroslag material in the upper shell. Thus, the higher of the LST and the RTNDT + 60&deg;F is 83.1&deg;F, the bolt-up limit in the closure flange region.
The P/T curves for hydrotesting conditions of the reactor vessels are controlled by the beltline region, which is subjected to radiation embrittlement due to higher fluence (neutron flux).
The bottom head region receives no significant fluence and therefore is not subjected to radiation embrittlement to the extent that the beltline region is.
(continued)
BFN-UNIT 3                    B 3.4-56b                            Revision 26 March 17, 2004
 
RCS P/T Limits B 3.4.9 BASES ACTIONS      B.1 and B.2 (continued)
If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress, or a sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. With the reduced pressure and temperature conditions, the possibility of propagation of undetected flaws is decreased.
Pressure and temperature are reduced by placing the plant in at least MODE 3 within 12 hours and in 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 Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.
Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed.
This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 212&deg;F. Several methods may be used, including comparison with pre-analyzed transients, new (continued)
BFN-UNIT 3                    B 3.4-61                              Revision 0
 
ECCS - Operating B 3.5.1 BASES ACTIONS      G.1 and G.2 (continued)
If any Required Action and associated Completion Time of Condition C, D, E, or F is not met, or if two or more ADS valves are inoperable, the plant must be brought to a condition 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 reactor steam dome pressure reduced to d 150 psig 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.
H.1 When multiple ECCS subsystems are inoperable, as stated in Condition H, the plant is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCI System, CS System, and LPCI subsystems full of water ensures that the ECCS will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.5-12                          Revision 0, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.2 REQUIREMENTS (continued) Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the HPCI System, this SR also includes the steam flow path for the turbine and the flow controller position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the RHR low pressure permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. This allows operation in the RHR shutdown cooling mode during MODE 3, if necessary.
(continued)
BFN-UNIT 3                    B 3.5-13                      Revision 0, 109, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.3 REQUIREMENTS (continued) Verification that ADS air supply header pressure is t 81 psig ensures adequate air pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least two valve actuations can occur with the drywell at 62.5% of design pressure plus three additional actuations at 0 psig drywell pressure (Ref. 10). The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of t 81 psig is provided by the Drywell Control Air System. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.1.4 Verification that the LPCI cross tie valve is closed and power to its operator is disconnected or that the manual shutoff valve in the cross tie between loops is closed ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI subsystem. Acceptable methods of removing power to the operator include de-energizing breaker control power, racking out or removing the breaker, or disconnecting the motor leads.
If the LPCI cross tie between loops is not isolated as described above, both LPCI subsystems must be considered inoperable.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.5-14                          Revision 0, 123 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.6, SR 3.5.1.7, and SR 3.5.1.8 (continued)
REQUIREMENTS pressure and flow are achieved to perform these tests. Reactor startup is allowed prior to performing the low pressure Surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance test is short. Alternately, the low pressure Surveillance test may be performed prior to startup using an auxiliary steam supply. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure test has been satisfactorily completed and there is no indication or reason to believe that HPCI is inoperable.
Therefore, SR 3.5.1.7 and SR 3.5.1.8 are modified by Notes that state the Surveillances are not required to be performed until 12 hours after the reactor steam pressure and flow are adequate to perform the test.
The Frequency for SR 3.5.1.6 is in accordance with the INSERVICE TESTING PROGRAM requirements. The Frequencies for SR 3.5.1.7 and SR 3.5.1.8 are controlled under the Surveillance Frequency Control.
(continued)
BFN-UNIT 3                    B 3.5-17                      Revision 109, 123 Amendment No. 215 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.9 REQUIREMENTS (continued) The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This SR also ensures that the HPCI System will automatically restart on an RPV low-low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool.
The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
(continued)
BFN-UNIT 3                    B 3.5-18                              Revision 123 Amendment No. 215 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.10 REQUIREMENTS (continued) The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e., solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that excludes valve actuation.
This prevents an RPV pressure blowdown.
(continued)
BFN-UNIT 3                    B 3.5-19                            Revision 123 Amendment No. 215 March 22, 2020
 
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.11 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.1.12 (Deleted)
(continued)
BFN-UNIT 3                  B 3.5-21                        Revision 83, 123 Amendment No. 215 March 22, 2020
 
RCIC System B 3.5.3 BASES (continued)
SURVEILLANCE      SR 3.5.3.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge line of the RCIC System full of water ensures that the system will perform properly, injecting its full capacity into the Reactor Coolant System upon demand. This will also prevent a water hammer following an initiation signal. One acceptable method of ensuring the line is full is to vent at the high points. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.3.2 Verifying the correct alignment for manual, power operated, and automatic valves in the RCIC flow path provides assurance that the proper flow path will exist for RCIC operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are (continued)
BFN-UNIT 3                          B 3.5-34                            Revision 0, 123 March 22, 2020
 
RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.2 (continued)
REQUIREMENTS in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the RCIC System, this SR also includes the steam flow path for the turbine and the flow controller position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.3.3 and SR 3.5.3.4 The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated. The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Reactor steam pressure must be t 950 psig to perform SR 3.5.3.3 and t 150 psig to perform SR 3.5.3.4. Adequate steam flow is represented by at least one turbine bypass valve full open for SR 3.5.3.3 and at least one turbine bypass valve > 50% open for SR 3.5.3.4. Therefore, sufficient time is allowed (continued)
BFN-UNIT 3                    B 3.5-35                    Revision 53, 109, 123 Amendment No. 214 March 22, 2020
 
RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.3 and SR 3.5.3.4 (continued)
REQUIREMENTS after adequate pressure and flow are achieved to perform these SRs. Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance is short. Alternately, the low pressure Surveillance test may be performed prior to startup using an auxiliary steam supply. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure Surveillance has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable. Therefore, these SRs are modified by Notes that state the Surveillances are not required to be performed until 12 hours after the reactor steam pressure and flow are adequate to perform the test.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                  B 3.5-36                        Revision 109, 123 Amendment No. 215 March 22, 2020
 
RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.5 REQUIREMENTS (continued) The RCIC System is required to actuate automatically in order to perform its design function satisfactorily. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of the RCIC System will cause the system to operate as designed, including actuation of the system throughout its emergency operating sequence; that is, automatic pump startup and actuation of all automatic valves to their required positions. This test also ensures the RCIC System will automatically restart on an RPV low-low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.2 overlaps this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that excludes vessel injection during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
(continued)
BFN-UNIT 3                    B 3.5-37                              Revision 123 Amendment No. 215 March 22, 2020
 
Primary Containment B 3.6.1.1 BASES SURVEILLANCE SR 3.6.1.1.2 (continued)
REQUIREMENTS Satisfactory performance of this SR can be achieved by establishing a known differential pressure between the drywell and the suppression chamber and verifying that the pressure in either the suppression chamber or the drywell does not change by more than 0.25 inch of water per minute over a 10 minute period. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. NEDC-33860P, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate, Section 2.6.
: 2. FSAR, Section 14.6.
: 3. 10 CFR 50, Appendix J, Option B.
: 4. NEI 94-01, Revision 3A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J."
: 5. ANSI/ANS-56.8-2002, "American National Standard for Containment System Leakage Testing Requirements."
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                    B 3.6-6                Revision 110, 114, 123 Amendment No. 215 March 22, 2020
 
Primary Containment Air Lock B 3.6.1.2 BASES SURVEILLANCE SR 3.6.1.2.1 (continued)
REQUIREMENTS The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 requires the results of airlock leakage tests be evaluated against the acceptance criteria of the Primary Containment Leakage Rate Testing Program, 5.5.12. This ensures that the airlock leakage is properly accounted for in determining the combined Type B and C primary containment leakage.
SR 3.6.1.2.2 The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment pressure, closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment.
Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.6-16                          Revision 0, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.1 (continued)
REQUIREMENTS following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.2 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured, and is required to be closed during accident conditions is closed.
The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.
This SR does not require any testing or valve manipulation.
Rather, it involves verification that those PCIVs outside primary containment, and capable of being mispositioned, are in the correct position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.6-30                          Revision 0, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.4 REQUIREMENTS (continued) The traversing incore probe (TIP) shear isolation valves are actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that TIP valves will actuate when required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.5 Verifying the isolation time of each power operated, automatic PCIV is within limits is required to demonstrate OPERABILITY.
MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.6. The isolation time test ensures that the valve will isolate in a time period less than or equal to that assumed in the safety analyses. The isolation time and Frequency of this SR are in accordance with the requirements of the INSERVICE TESTING PROGRAM.
SR 3.6.1.3.6 Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. The isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the DBA analyses. This ensures that the calculated radiological consequences of these events remain within 10 CFR 50.67 limits. The Frequency of this SR is in accordance with the requirements of the INSERVICE TESTING PROGRAM.
(continued)
BFN-UNIT 3                    B 3.6-33                  Revision 0, 29, 109, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.7 REQUIREMENTS (continued) Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.1 overlaps this SR to provide complete testing of the safety function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.8 This SR requires a demonstration that a representative sample of reactor instrumentation line excess flow check valves (EFCVs) are OPERABLE by verifying that the valves actuate to the isolation position on an actual or simulated instrument line break signal. This SR provides assurance that the instrumentation line EFCVs will perform so that the radiological consequences will not exceed the predicted radiological consequences during events evaluated in Reference 5. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.6-34                Amendment No. 215, 228 Revision 0, 123 March 22, 2020
 
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.9 REQUIREMENTS (continued) The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.1.3.10 The analyses in References 1 and 5 are based on leakage that is less than the specified leakage rate. Leakage through each MSIV must be d 100 scfh when tested at t Pt (25 psig). The combined leakage rate for all four main steam lines must be d 150 scfh when tested at t 25 psig in accordance with the Primary Containment Leakage Rate Testing Program. If the leakage rate through an individual MSIV exceeds 100 scfh, the leakage rate shall be restored below the alarm limit value as specified in the Containment Leakage Rate Testing Program referenced in TS 5.5.12. This ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate. The Frequency is specified in the Primary Containment Leakage Rate Testing Program.
(continued)
BFN-UNIT 3                    B 3.6-35                        Revision 62, 123 Amendment No. 212, 215, 223, 227 March 22, 2020
 
PCIVs B 3.6.1.3 BASES REFERENCES 1. FSAR, Section 14.6.
: 2. BFN Technical Instruction (TI), 0-TI-360.
: 3. 10 CFR 50, Appendix J, Option B.
: 4. FSAR, Section 5.2.
: 5. FSAR, Section 14.6.5.
: 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 7. FSAR Table 5.2-2.
: 8. Deleted.
: 9. MDQ0000012016000566 Revision 0, Main Steam Isolation Valve (MSIV) Loss of Coolant Accident (LOCA) Closure Analysis, dated September 2016.
BFN-UNIT 3                B 3.6-36                Revision 0, 62, 105, 123 Amendment No. 223, 228 March 22, 2020
 
Drywell Air Temperature B 3.6.1.4 BASES ACTIONS      B.1 and B.2 (continued)
If the drywell average air temperature cannot be restored to within limit 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.
SURVEILLANCE SR 3.6.1.4.1 REQUIREMENTS Verifying that the drywell average air temperature is within the LCO limit ensures that operation remains within the limits assumed for the primary containment analyses. Drywell air temperature is monitored in various quadrants and at various elevations (referenced to mean sea level). Due to the shape of the drywell, a volumetric average is used to determine an accurate representation of the actual average temperature.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.6-39                        Revision 0, 123 March 22, 2020
 
Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.5 BASES (continued)
SURVEILLANCE      SR 3.6.1.5.1 REQUIREMENTS Each vacuum breaker is verified to be closed to ensure that a potential breach in the primary containment boundary is not present. This Surveillance is performed by observing local or control room indications of vacuum breaker position or by verifying a differential pressure of 0.5 psid is maintained between the reactor building and suppression chamber. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Two Notes are added to this SR. The first Note allows reactor building-to-suppression chamber vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable breakers. A second Note is included to clarify that vacuum breakers open due to an actual differential pressure, are not considered as failing this SR.
SR 3.6.1.5.2 Each vacuum breaker must be cycled to ensure that it opens properly to perform its design function and returns to its fully closed position. This ensures that the safety analysis assumptions are valid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                          B 3.6-47                      Revision 0, 109, 123 March 22, 2020
 
Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.5 BASES SURVEILLANCE SR 3.6.1.5.3 REQUIREMENTS (continued) Demonstration of vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of d 0.5 psid is valid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. TVA Calculation ND-Q0064-900040.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                    B 3.6-48                          Revision 123 Amendment No. 215 March 22, 2020
 
Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.6 BASES (continued)
SURVEILLANCE      SR 3.6.1.6.1 REQUIREMENTS Each vacuum breaker is verified closed to ensure that this potential large bypass leakage path is not present. This Surveillance is performed by observing the vacuum breaker position indication or by verifying that the rate of increase in suppression chamber pressure is less than 0.25 inches of water per minute over a ten minute period at a differential pressure of at least 1.0 psi. Note 2 specifies that vacuum breaker may be nonfully closed provided it is not more than 3&deg; open as indicated by position indication lights. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Note 1 has been added to this SR which allows suppression chamber-to-drywell vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers.
SR 3.6.1.6.2 Each required (i.e., required to be OPERABLE for opening) vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position. This ensures that the safety analysis assumptions are valid. The INSERVICE TESTING PROGRAM Frequency is based on operating experience that has demonstrated that the Frequency is adequate to assure OPERABILITY.
(continued)
BFN-UNIT 3                          B 3.6-55                      Revision 0, 109, 123 March 22, 2020
 
Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.6 BASES SURVEILLANCE SR 3.6.1.6.3 REQUIREMENTS (continued) Verification of the differential pressure required to open the vacuum breaker is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of 0.5 psid is valid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 5.2.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 3. Technical Requirements Manual.
BFN-UNIT 3                    B 3.6-56                            Revision 123 Amendment No. 215 March 22, 2020
 
Suppression Pool Average Temperature B 3.6.2.1 BASES (continued)
SURVEILLANCE      SR 3.6.2.1.1 REQUIREMENTS The suppression pool average temperature is regularly monitored to ensure that the required limits are satisfied. The average temperature is determined by taking an arithmetic average of OPERABLE suppression pool water temperature channels. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The 5 minute Frequency during testing is justified by the rates at which tests will heat up the suppression pool, has been shown to be acceptable based on operating experience, and provides assurance that allowable pool temperatures are not exceeded.
The Frequency is further justified in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature condition.
REFERENCES        1. FSAR, Section 5.2.
: 2. FSAR, Section 14.6.
: 3. NUREG-0783, Suppression Pool Temperature Limits for BWR Containments, November 1981.
: 4. NUREG-0661, "Safety Evaluation Report Mark I Containment Long Term Program - Resolution of Generic Technical Activity A-7," July 1980.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 6. NEDC-22004-P, Browns Ferry Nuclear Plant Units 1, 2, and 3 Suppression Pool Temperature Response, October 1981.
BFN-UNIT 3                        B 3.6-63                  Revision 0, 66, 85, 123 March 22, 2020
 
Suppression Pool Water Level B 3.6.2.2 BASES ACTIONS      B.1 and B.2 (continued)
If suppression pool water level cannot be restored to within limits 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.
SURVEILLANCE SR 3.6.2.2.1 REQUIREMENTS Verification of the suppression pool water level is to ensure that the required limits are satisfied. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Sections 5.2 and 14.6.3.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                    B 3.6-67                        Revision 0, 123 March 22, 2020
 
RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS      D.1 and D.2 (continued)
If any Required Action and associated Completion Time cannot be met, 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.
SURVEILLANCE SR 3.6.2.3.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.6-72                          Revision 0, 123 Amendment No. 230 March 22, 2020
 
RHR Suppression Pool Spray B 3.6.2.4 BASES SURVEILLANCE SR 3.6.2.4.1 (continued)
REQUIREMENTS sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool spray mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.2.4.2 This Surveillance is performed using air or water to verify that the spray nozzles are not obstructed and that flow will be provided when required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Sections 5.2 and 14.6.3.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                    B 3.6-79                          Revision 0, 123 March 22, 2020
 
RHR Drywell Spray B 3.6.2.5 BASES ACTIONS      D.1 and D.2 (continued)
If any Required Action and the associated Completion Time cannot be met, 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 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.
SURVEILLANCE SR 3.6.2.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR drywell spray mode flow path provides assurance that the proper flow paths will exist for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR drywell cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.6-84                            Revision 0, 123 March 22, 2020
 
RHR Drywell Spray B 3.6.2.5 BASES SURVEILLANCE SR 3.6.2.5.2 REQUIREMENTS (continued) This Surveillance is performed using air to verify that the spray nozzles are not obstructed and that flow will be provided when required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Sections 5.2 and 14.6.3.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                  B 3.6-85                          Revision 3, 123 March 22, 2020
 
Drywell-to-Suppression Chamber Differential Pressure B 3.6.2.6 BASES (continued)
SURVEILLANCE      SR 3.6.2.6.1 REQUIREMENTS The drywell-to-suppression chamber differential pressure is regularly monitored to ensure that the required limits are satisfied. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 5.2.3.9.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                        B 3.6-89                        Revision 0, 123 March 22, 2020
 
CAD System B 3.6.3.1 BASES (continued)
SURVEILLANCE      SR 3.6.3.1.1 REQUIREMENTS Verifying that there is t 2615 gal of liquid nitrogen supply in each nitrogen storage tank will ensure at least 7 days of post-LOCA CAD operation. This minimum volume of liquid nitrogen represents the analytical limit assumed in the analysis of the primary containment atmosphere following a postulated LOCA and does not include allowance for potential nitrogen boiloff and tank level instrumentation inaccuracies. This minimum volume of liquid nitrogen allows sufficient time after an accident to replenish the nitrogen supply for long term inerting.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.3.1.2 Verifying the correct alignment for manual, power operated, and automatic valves in each of the CAD subsystem flow paths provides assurance that the proper flow paths exist for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing.
A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable because the CAD System is manually initiated. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.
(continued)
BFN-UNIT 3                        B 3.6-95                      Revision 0, 110, 123 Amendment No. 225 March 22, 2020
 
CAD System B 3.6.3.1 BASES SURVEILLANCE SR 3.6.3.1.2 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. AEC Safety Guide 7, Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident, March 10, 1971.
: 2. FSAR, Section 5.2.6.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 4. ANP-3403P, Fuel Uprate Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3, Section 2.6.4.
BFN-UNIT 3                  B 3.6-96                    Revision 0, 110, 123 Amendment No. 225 March 22, 2020
 
Primary Containment Oxygen Concentration B 3.6.3.2 BASES (continued)
SURVEILLANCE      SR 3.6.3.2.1 REQUIREMENTS The primary containment (drywell and suppression chamber) must be determined to be inert by verifying that oxygen concentration is < 4.0 v/o. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 5.2.6.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                        B 3.6-100                        Revision 0, 123 March 22, 2020
 
Secondary Containment B 3.6.4.1 BASES (continued)
SURVEILLANCE      SR 3.6.4.1.1 and SR 3.6.4.1.2 REQUIREMENTS Verifying that secondary containment equipment hatches and one access door in each access opening are closed ensures that the infiltration of outside air of such a 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 the access opening is closed. An access opening contains one inner and one outer door. In some cases, secondary containment access openings are shared such that a secondary containment barrier may have multiple inner doors. The main Equipment Access Lock (EAL) has a smaller sub-door on each of the large inner and outer main EAL doors. For the EAL, maintaining secondary containment OPERABILITY requires verifying that a large door and its integral sub-door are both closed. The intent is to not breach the secondary containment at any time when secondary containment is required. This is achieved by maintaining the inner or outer portion of the barrier closed at all times. However, all secondary containment access doors are normally kept closed, except when the access opening is being used for entry and exit or when maintenance is being performed on an access opening. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                          B 3.6-105                  Revision 0, 23, 29, 123 Amendment No. 224 March 22, 2020
 
Secondary Containment B 3.6.4.1 BASES SURVEILLANCE SR 3.6.4.1.3 and SR 3.6.4.1.4 REQUIREMENTS (continued) The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment. To ensure that all fission products are treated, SR 3.6.4.1.3 verifies that the SGT System will rapidly establish and maintain a pressure in the secondary containment that is less than the lowest postulated pressure external to the secondary containment boundary. This is confirmed by demonstrating that two SGT subsystems will draw down the secondary containment to t 0.25 inches of vacuum water gauge in d 120 seconds. This cannot be accomplished if the secondary containment boundary is not intact. SR 3.6.4.1.4 demonstrates that two SGT subsystems can maintain t 0.25 inches of vacuum water gauge at a stable flow rate d 12,000 cfm. Both of these SRs are performed under neutral
( 5 mph) wind conditions. Therefore, these two tests are used to ensure secondary containment boundary integrity. Since these SRs are secondary containment tests, they need not be performed with each combination of SGT subsystems. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 5.3.
: 2. FSAR, Section 14.6.3.
: 3. Deleted.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                  B 3.6-106                      Revision 29, 123 Amendment No. 215 March 22, 2020
 
SCIVs B 3.6.4.2 BASES (continued)
SURVEILLANCE      SR 3.6.4.2.1 REQUIREMENTS Verifying that the isolation time of each power operated, automatic SCIV is within limits is required to demonstrate OPERABILITY. The isolation time test ensures that the SCIV will isolate in a time period less than or equal to that assumed in the safety analyses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.4.2.2 Verifying that each automatic SCIV closes on a secondary containment isolation signal is required to prevent leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SCIV 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 14.6.3.
: 2. Deleted.
: 3. Technical Requirements Manual.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                          B 3.6-113                        Revision 29, 123 Amendment No. 215 March 22, 2020
 
SGT System B 3.6.4.3 BASES (continued)
SURVEILLANCE      SR 3.6.4.3.1 REQUIREMENTS Operating each SGT subsystem for t 15 continuous minutes with heaters on ensures that the 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. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.4.3.2 This SR verifies that the required SGT filter testing is performed in accordance with the 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). This SR will also include a chemical smoke test to check the sealing of gaskets for filter housing doors.
Specific test frequencies and additional information are discussed in detail in the VFTP.
(continued)
BFN-UNIT 3                        B 3.6-120                    Revision 0, 108, 123 March 22, 2020
 
SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.3 REQUIREMENTS (continued) This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.4.3.4 This SR verifies that the SGT decay heat discharge dampers are in the correct position. This ensures that the decay heat removal mode of SGT System operation is available. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 41.
: 2. FSAR, Section 5.3.3.7.
: 3. FSAR, Section 14.6.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                  B 3.6-121                            Revision 123 Amendment No. 215 March 22, 2020
 
RHRSW System B 3.7.1 BASES ACTIONS      G.1 and G.2 (continued)
If the RHRSW subsystem(s) or the RHRSW pump(s) cannot be restored to OPERABLE status within the associated Completion Times, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.1.1 REQUIREMENTS Verifying the correct alignment for each manual and power operated valve in each RHRSW subsystem flow path provides assurance that the proper flow paths will exist for RHRSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be realigned to its accident position. This is acceptable because the RHRSW System is a manually initiated system.
This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                      B 3.7-9                    Revision 73, 110, 123 Amendment No. 214 March 22, 2020
 
EECW System and UHS B 3.7.2 BASES ACTIONS      A.1 (continued)
The 7 day Completion Time is based on the redundant EECW System capabilities afforded by the remaining OPERABLE pumps, the low probability of an accident occurring during this time period and is consistent with the allowed Completion Time for restoring an inoperable DG.
B.1 and B.2 If the required EECW pump cannot be restored to OPERABLE status within the associated Completion Time, or two or more EECW pumps are inoperable or the UHS is determined inoperable, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.2.1 REQUIREMENTS Verification of the UHS temperature ensures that the heat removal capability of the EECW System is within the assumptions of the DBA analysis. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.7-14                Revision 0, 69, 110, 123 March 22, 2020
 
EECW System and UHS B 3.7.2 BASES SURVEILLANCE SR 3.7.2.2 REQUIREMENTS (continued) Verifying the correct alignment for each manual and power operated valve in the EECW System flow paths provide assurance that the proper flow paths will exist for EECW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be automatically realigned to its accident position within the required time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
This SR is modified by a Note indicating that isolation of the EECW System to components or systems may render those components or systems inoperable, but does not affect the OPERABILITY of the EECW System. As such, when required EECW pumps, valves, and piping are OPERABLE, but a branch connection off the main header is isolated, the EECW System is still OPERABLE.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.7-15                            Revision 0, 123 March 22, 2020
 
EECW System and UHS B 3.7.2 BASES SURVEILLANCE SR 3.7.2.3 REQUIREMENTS (continued) This SR verifies that the EECW System pumps will automatically start to provide cooling water to the required safety related equipment during an accident event. This is demonstrated by the use of an actual or simulated initiation signal. This SR includes a functional test of the initiation logic and a functional test and calibration of the EECW pump timers (both normal power and diesel power).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Chapter 5.
: 2. FSAR, Chapter 14.
: 3. FSAR, Section 10.10.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                  B 3.7-16                              Revision 123 Amendment No. 215 March 22, 2020
 
CREV System B 3.7.3 BASES (continued)
SURVEILLANCE      SR 3.7.3.1 REQUIREMENTS This SR verifies that a subsystem in a standby mode starts on demand and continues to operate. Standby systems should be checked periodically to ensure that they start and function properly. As the environmental and normal operating conditions of this system are not severe, testing each subsystem once every month provides an adequate check on this system.
Operation with the heaters on for  15 continuous minutes demonstrates OPERABILITY of the system. Periodic operation ensures that heater failure, blockage, fan or motor failure, or excessive vibration can be detected for corrective action. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.3.2 This SR verifies that the required CREV testing is performed in accordance with the 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).
Specific test frequencies and additional information are discussed in detail in the VFTP.
(continued)
BFN-UNIT 3                        B 3.7-24                Revision 0, 67, 108, 123 March 22, 2020
 
CREV System B 3.7.3 BASES SURVEILLANCE SR 3.7.3.3 REQUIREMENTS (continued) This SR verifies that on an actual or simulated initiation signal, each CREV subsystem starts and operates. This SR includes verification that dampers necessary for proper CREV operation function as required. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.7.1.4 and SR 3.3.7.1.6 overlaps this SR to provide complete testing of the safety function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.3.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.
The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than 5 REM TEDE and the CRE occupants are protected from hazardous chemicals and smoke.
There is no automatic CREV actuation for hazardous chemical releases or smoke and there are no Surveillance Requirements to verify the OPERABILITY in cases of hazardous chemicals or smoke. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analysis of DBA consequences. When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered. Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for occupants following an accident.
Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 6) which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 7).
These compensatory measures may also be used as mitigating actions as required by Required Action B.2. Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref. 8).
BFN-UNIT 3                      B 3.7-25                      Revision 90, 123 Amendment No. 215, 261 March 22, 2020
 
Control Room AC System B 3.7.4 BASES (continued)
SURVEILLANCE      SR 3.7.4.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the control room heat load assumed in the safety analyses. The SR consists of a combination of testing and calculation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 10.12.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                        B 3.7-31                            Revision 123 Amendment No. 215 March 22, 2020
 
Main Turbine Bypass System B 3.7.5 BASES ACTIONS      B.1 (continued)
Turbine Bypass System is not required to protect fuel integrity during abnormal operational transients. The 4 hour Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Cycling each main turbine bypass valve through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and will function when required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.5.2 The Main Turbine Bypass System is required to actuate automatically to perform its design function. This SR demonstrates that, with the required system initiation signals, the valves will actuate to their required position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.7-35                            Revision 123 Amendment No. 215 March 22, 2020
 
Main Turbine Bypass System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.3 REQUIREMENTS (continued) This SR ensures that the TURBINE BYPASS SYSTEM RESPONSE TIME is in compliance with the assumptions of the appropriate safety analysis. The response time limits are specified in the cycle specific transient analyses performed to support the preparation of FSAR, Appendix N, Supplemental Reload Licensing Report (Ref. 4). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 7.11.
: 2. FSAR, Section 14.5.1.1 and 14.5.1.2.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
: 4. FSAR, Appendix N.
BFN-UNIT 3                    B 3.7-36                    Revision 25, 103, 123 Amendment No. 215 March 22, 2020
 
Spent Fuel Storage Pool Water Level B 3.7.6 BASES (continued)
SURVEILLANCE      SR 3.7.6.1 REQUIREMENTS This SR verifies that sufficient water is available in the event of a fuel handling accident. The water level in the spent fuel storage pool must be checked periodically. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 10.3.
: 2. FSAR, Section 14.6.4.
: 3. NUREG-0800, Section 15.0.1.
: 4. 10 CFR 50.67.
: 5. Regulatory Guide 1.183.
: 6. FSAR, Section 14.6.4.5.
: 7. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                        B 3.7-39                      Revision 0, 29, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.1 and SR 3.8.1.4 (continued)
REQUIREMENTS SR 3.8.1.4 requires that the DG starts from standby conditions and achieves required voltage and frequency within 10 seconds.
The 10 second start requirement supports the assumptions in the design basis LOCA analysis of FSAR, Section 14.6.3 (Ref. 10). The 10 second start requirement is not applicable to SR 3.8.1.1 (see the Note for SR 3.8.1.1), when a modified start procedure as described above is used. If a modified start is not used, the 10 second start requirement of SR 3.8.1.4 applies.
Since SR 3.8.1.4 does require a 10 second start, it is more restrictive than SR 3.8.1.1, and it may be performed in lieu of SR 3.8.1.1. This procedure is the intent of Note 1 of SR 3.8.1.1.
In addition to the SR requirements, the time for the DG to reach steady state operation, unless the modified DG start method is employed, is periodically monitored and the trend evaluated to identify degradation of governor performance.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.8-29                          Revision 0, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 REQUIREMENTS (continued) This Surveillance demonstrates that the DGs are capable of synchronizing and accepting greater than 90 percent of the continuous rating. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source.
Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Note 1 modifies this Surveillance to indicate that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized.
Note 2 modifies this Surveillance by stating that momentary transients because of changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test.
Note 3 indicates that this Surveillance should be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations.
Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.
(continued)
BFN-UNIT 3                    B 3.8-30                            Revision 0, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.3 REQUIREMENTS (continued) This Surveillance demonstrates that each required fuel oil transfer pump operates and transfers fuel oil from its associated 7-day storage tank to its associated engine fuel oil tank. It is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE.
The design of fuel transfer systems is such that pumps that transfer the fuel oil operate automatically in order to maintain an adequate volume of fuel oil in the engine tank during or following DG operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.1.4 See SR 3.8.1.1.
(continued)
BFN-UNIT 3                    B 3.8-31                          Revision 0, 123 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.5 (continued)
REQUIREMENTS The voltage tolerances specified in this SR are based on the degraded voltage and overvoltage relay settings. The frequency tolerances specified in this SR are derived from Safety Guide 9 (Ref. 3) recommendations for response during load sequence intervals. The voltage and frequency specified are consistent with the design range of the equipment powered by the DG. SR 3.8.1.5.a corresponds to the maximum frequency excursion, while SR 3.8.1.5.b and 3.8.1.5.c are steady state voltage and frequency values to which the system must recover following load rejection. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, the Note requires that, if synchronized to offsite power, testing must be performed using a power factor d 0.9. This power factor is chosen to be representative of the actual design basis inductive loading that the DG would experience.
SR 3.8.1.6 This Surveillance demonstrates that the DG automatically starts from the design basis actuation signal (LOCA signal). This test will also verify the start of the Unit 1 and 2 DGs aligned to the SGT and CREV Systems on an accident signal from Unit 3. In order to minimize the number of DGs involved in testing, demonstration of automatic starts of the Unit 1 and 2 DGs on an accident signal from Unit 3 may be performed in conjunction (continued)
BFN-UNIT 3                      B 3.8-33                            Revision 123 Amendment No. 215 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.6 (continued)
REQUIREMENTS with testing to demonstrate automatic starts of the Unit 1 and 2 DGs on an accident signal from Unit 1 or 2. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
To minimize wear and tear on the DGs, this SR has been modified by a Note which permits DG starts to be preceded by an engine prelube period followed by a warmup period.
SR 3.8.1.7 Demonstration periodically that the DGs can start and run continuously at full load capability for an interval of not less than 24 hours - 22 hours of which is at a load equivalent to the continuous rating of the DG, and 2 hours of which is at a load equivalent to 105 percent to 110 percent of the continuous duty rating of the DG. The DG starts for this Surveillance can be performed either from standby or hot conditions. The provisions for prelube and warmup, discussed in SR 3.8.1.1, and for gradual loading, discussed in SR 3.8.1.2, are applicable to this SR.
In order to ensure that the DG is tested under load conditions that are as close to design conditions as possible, testing must be performed using a power factor d 0.9. This power factor is chosen to be representative of the actual design basis inductive loading that the DG could experience. A load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.
(continued)
BFN-UNIT 3                    B 3.8-34                              Revision 123 Amendment No. 215 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.7 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This Surveillance has been modified by a Note that states that momentary transients due to changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test.
SR 3.8.1.8 Under accident conditions (and loss of offsite power) loads are sequentially connected to the shutdown boards by automatic individual pump timers. The individual pump timers control the permissive and starting signals to motor breakers to prevent overloading of the DGs due to high motor starting currents. This SR is demonstrated by performance of SR 3.3.5.1.5 for the Core Spray and LPCI pump timers, SR 3.7.2.3 for the EECW pump timers, and SR 3.8.1.9.b for the 480 V load shed logic timers. The allowable values for these timers ensure that sufficient time exists for the DG to restore frequency and voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not violated. Reference 2 provides a summary of the automatic loading of ESF shutdown boards.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.8-35                            Revision 123 Amendment No. 215 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.9 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note. The reason for the Note is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil being continuously circulated and temperature maintained consistent with manufacturer recommendations.
SR 3.8.1.10 This Surveillance is provided to direct that the appropriate Surveillances for the required Unit 1 and 2 DGs are governed by the Unit 1 and 2 Technical Specifications. Performance of the applicable Unit 1 and 2 Surveillances will satisfy any Unit 1 and 2 requirements, as well as this Unit 3 Surveillance requirement.
The Frequency required by the applicable Unit 1 and 2 SR also governs performance of that SR for both Units.
(continued)
BFN-UNIT 3                    B 3.8-37                            Revision 123 Amendment No. 215 March 22, 2020
 
AC Sources - Operating B 3.8.1 BASES (continued)
REFERENCES        1. 10 CFR 50, Appendix A, GDC 17.
: 2. FSAR, Chapter 8.
: 3. Safety Guide 9.
: 4. FSAR, Chapter 6.
: 5. FSAR, Chapter 14.
: 6. Regulatory Guide 1.93.
: 7. Generic Letter 84-15.
: 8. Deleted.
: 9. ANSI C84.1, 1982.
: 10. FSAR, Section 14.6.3.
: 11. IEEE Standard 308.
: 12. FSAR, Section 8.5, Table 8.5-6.
: 13. FSAR, Section 8.5.2.
: 14. TVA Design Criteria BFN-50-7082.
: 15. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                        B 3.8-38                      Revision 0, 123 March 22, 2020
 
Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES (continued)
SURVEILLANCE      SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 7 days at full load. The fuel oil level equivalent to a 7-day supply is 35,280 gallons when calculated in accordance with References 2 and 6. The required fuel storage volume is determined using the most limiting energy content of the stored fuel. Using the known correlation of diesel fuel oil absolute specific gravity or API gravity to energy content, the required diesel generator output, and the corresponding fuel consumption rate, the onsite fuel storage volume required for 7 days of operation can be determined. SR 3.8.3.3 requires new fuel to be tested to verify that the absolute specific gravity or API gravity is within the range assumed in the diesel fuel oil consumption calculations. The 7-day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.3.2 This Surveillance ensures that sufficient lubricating oil inventory is available to support at least 7 days of full load operation for each DG. The lube oil inventory equivalent to a 7-day supply is 175 gallons and is based on the DG manufacturer's consumption values for the run time of the DG.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                          B 3.8-55                  Revision 0, 63, 95, 123 March 22, 2020
 
Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.4 REQUIREMENTS (continued) This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available.
The system design requirements provide for at least one start cycle from one of two redundant air start systems without recharging. A start cycle is defined by the DG vendor, but usually is measured in terms of time (seconds of cranking) or engine cranking speed. The pressure specified in this SR is the lowest pressure at which at least one start attempt can be accomplished using one of two redundant air start systems.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.3.5 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Periodic removal of water from the fuel storage tanks eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling.
In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
BFN-UNIT 3                    B 3.8-56b                      Revision 0, 95, 123 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES (continued)
SURVEILLANCE      SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or a battery cell) in a fully charged state, while supplying adequate power to the connected DC loads. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.4.2 and SR 3.8.4.5 Battery charger capability requirements are based on the design capacity of the chargers (Ref. 4). According to Regulatory Guide 1.32 (Ref. 8), the battery charger supply is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and verification of the charger's ability to recharge the battery ensures that these requirements can be satisfied.
(continued)
BFN-UNIT 3                          B 3.8-64                            Revision 0, 123 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.2 and SR 3.8.4.5 (continued)
REQUIREMENTS SR 3.8.4.2 verifies that the chargers are capable of charging the batteries after their designed duty cycle testing and ensures that the chargers will perform their design function. This SR is modified by a Note that allows the performance of SR 3.8.4.5 in lieu of this Surveillance requirement. SR 3.8.4.5 verifies that the chargers are capable of charging the batteries after each discharge test and ensures that the chargers are capable of performing at maximum output. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.4.5 is modified by a Note. The Note is added to this SR to acknowledge that credit may be taken for unplanned events that satisfy the Surveillance.
SR 3.8.4.3 A battery service test is a special test of the battery's capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length corresponds to the design duty cycle requirements as specified in Reference 4.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.8-65                              Revision 123 Amendment No. 215 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.3 (continued)
REQUIREMENTS This SR is modified by a Note that allows the performance of a modified performance discharge test in lieu of a service test.
The modified performance discharge test is a simulated duty cycle consisting of just two periods (the one minute rate, followed by the test rate employed for the performance test) or three periods (the one minute rate, followed by the second minute rate followed by the test rate employed for the performance test) both of which envelope the duty cycle of the service test. Since the ampere-hours removed by the rated one or two minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.
A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity.
Initial conditions for the modified performance discharge test should be identical to those specified for a service test.
(continued)
BFN-UNIT 3                      B 3.8-66                      Revision 0, 58, 123 March 22, 2020
 
DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.4 REQUIREMENTS (continued) A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.
A battery modified performance discharge test is described in the Bases for SR 3.8.4.3. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.4; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.4 while satisfying the requirements of SR 3.8.4.3 at the same time.
The acceptance criteria for this Surveillance is consistent with IEEE-450 (Ref. 7) and IEEE-485 (Ref. 10). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. If the battery shows degradation, or if the battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity t 100% of the manufacturer's rating. Degradation is indicated, according to IEEE-450 (Ref. 7), when the battery capacity drops by more than 10%
relative to its capacity on the previous performance test or when it is 10% below the manufacturer's rating. All these Frequencies are consistent with the recommendations in IEEE-450 (Ref. 7).
(continued)
BFN-UNIT 3                    B 3.8-67                          Revision 0, 123 March 22, 2020
 
Battery Cell Parameters B 3.8.6 BASES (continued)
SURVEILLANCE      SR 3.8.6.1 REQUIREMENTS This SR verifies that Category A battery cell parameters are consistent with IEEE-450 (Ref. 3), including voltage, specific gravity, and electrolyte temperature of pilot cells. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.6.2 The inspection of specific gravity and voltage is consistent with IEEE-450 (Ref. 3). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.8.6.3 This Surveillance verification that the average temperature of representative cells is within limits is consistent with a recommendation of IEEE-450 (Ref. 3) that states that the temperature of electrolytes in representative (10 percent of) cells should be determined. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Lower than normal temperatures act to inhibit or reduce battery capacity. This SR ensures that the operating temperatures remain within an acceptable operating range. This limit is based on manufacturer's recommendations.
(continued)
BFN-UNIT 3                        B 3.8-78                            Revision 0, 123 March 22, 2020
 
Distribution Systems - Operating B 3.8.7 BASES ACTIONS      H.1 and H.2 (continued)
If the inoperable distribution subsystem cannot be restored to OPERABLE status within the associated Completion Time, the unit 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.
I.1 Condition I corresponds to a level of degradation in the electrical distribution system that causes a required safety function to be lost. When more than one AC or DC electrical power distribution subsystem is lost, and this results in the loss of a required function, the plant is in a condition outside the accident analysis. Therefore, no additional time is justified for continued operation. LCO 3.0.3 must be entered immediately to commence a controlled shutdown.
SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the AC and DC electrical power distribution subsystem is functioning properly, with the buses energized. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                    B 3.8-101                          Revision 0, 123 March 22, 2020
 
Distribution Systems - Shutdown B 3.8.8 BASES (continued)
SURVEILLANCE      SR 3.8.8.1 REQUIREMENTS This Surveillance verifies that the AC and DC electrical power distribution subsystem is functioning properly, with the buses energized. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. This may be performed by verification of an absence of low voltage alarm. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Chapter 6.
: 2. FSAR, Chapter 14.
: 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                        B 3.8-109                          Revision 0, 123 March 22, 2020
 
Refueling Equipment Interlocks B 3.9.1 BASES ACTIONS      A.1, A.2.1, and A.2.2 (continued) remain inserted). Required Action A.2.2 is normally performed after placing the rod withdrawal block in effect, and provides a verification that all control rods are fully inserted. This verification that all control rods are fully inserted is in addition to the periodic verifications required by SR 3.9.3.1. Like Required Action A.1, Required Actions A.2.1 and A.2.2 ensure unacceptable operations are blocked (e.g., loading fuel into a cell with the control rod withdrawn). It is not the intent of Actions A.2 to eliminate the first performance of SR 3.9.1.1 prior to in-vessel fuel movement. It is expected that the refueling interlocks would be operable except for equipment failure or expiration of the required surveillance interval, and Actions A.2 would not be entered as a convenience for avoiding the first performance of SR 3.9.1.1.
SURVEILLANCE SR 3.9.1.1 REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST demonstrates each required refueling equipment interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested. This SR is only required for refueling equipment in use.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                      B 3.9-5                        Revision 0, 16, 123 Amendment No. 232 March 22, 2020
 
Refuel Position One-Rod-Out Interlock B 3.9.2 BASES (continued)
SURVEILLANCE      SR 3.9.2.1 REQUIREMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel. During control rod withdrawal in MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock will be OPERABLE when required. By "locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel),
an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested.
(continued)
BFN-UNIT 3                          B 3.9-9                          Revision 0, 123 March 22, 2020
 
Refuel Position One-Rod-Out Interlock B 3.9.2 BASES SURVEILLANCE SR 3.9.2.2 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. To perform the required testing, the applicable condition must be entered (i.e., a control rod must be withdrawn from its full-in position). Therefore, SR 3.9.2.2 has been modified by a Note that states the CHANNEL FUNCTIONAL TEST is not required to be performed until 1 hour after any control rod is withdrawn.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 26.
: 2. FSAR, Section 7.6.3.
: 3. FSAR, Section 14.5.4.3.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                    B 3.9-10                      Revision 0, 60, 123 March 22, 2020
 
Control Rod Position B 3.9.3 BASES (continued)
ACTIONS          A.1 With all control rods not fully inserted during the applicable conditions, an inadvertent criticality could occur that is not analyzed in the FSAR. All fuel loading operations must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE      SR 3.9.3.1 REQUIREMENTS During refueling, to ensure that the reactor remains subcritical, all control rods must be fully inserted prior to and during fuel loading. Periodic checks of the control rod position ensure this condition is maintained.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. 10 CFR 50, Appendix A, GDC 26.
: 2. FSAR, Section 14.5.4.3.
: 3. FSAR, Section 14.5.4.4.
: 4. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                          B 3.9-13                      Revision 0, 60, 123 March 22, 2020
 
Control Rod OPERABILITY - Refueling B 3.9.5 BASES (continued)
ACTIONS          A.1 With one or more withdrawn control rods inoperable, action must be immediately initiated to fully insert the inoperable control rod(s). Inserting the control rod(s) ensures the shutdown and scram capabilities are not adversely affected. Actions must continue until the inoperable control rod(s) is fully inserted.
SURVEILLANCE      SR 3.9.5.1 and SR 3.9.5.2 REQUIREMENTS During MODE 5, the OPERABILITY of control rods is primarily required to ensure a withdrawn control rod will automatically insert if a signal requiring a reactor shutdown occurs. Because no explicit analysis exists for automatic shutdown during refueling, the shutdown function is satisfied if the withdrawn control rod is capable of automatic insertion and the associated CRD scram accumulator pressure is t 940 psig.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. An automatic accumulator monitor may be used to continuously satisfy SR 3.9.5.2.
SR 3.9.5.1 is modified by a Note that allows 7 days after withdrawal of the control rod to perform the Surveillance. This acknowledges that the control rod must first be withdrawn before performance of the Surveillance, and therefore avoids potential conflicts with SR 3.0.3 and SR 3.0.4.
(continued)
BFN-UNIT 3                          B 3.9-21                          Revision 0, 123 March 22, 2020
 
RPV Water Level B 3.9.6 BASES (continued)
ACTIONS          A.1 If the water level is < 22 ft above the top of the RPV flange, all operations involving movement of fuel assemblies and handling of control rods within the RPV shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement and control rod handling shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE      SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 22 ft above the top of the RPV flange ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. 2).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. Regulatory Guide 1.183.
: 2. FSAR, Section 14.6.4.
: 3. NUREG-0800, Section 15.0.1.
: 4. 10 CFR 50.67.
: 5. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                          B 3.9-25                      Revision 0, 29, 123 March 22, 2020
 
RHR-High Water Level B 3.9.7 BASES ACTIONS      C.1 and C.2 (continued)
If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour. This alternative method may utilize forced or natural circulation. The Completion Time is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation.
During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR Shutdown Cooling System), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.
SURVEILLANCE SR 3.9.7.1 REQUIREMENTS This Surveillance demonstrates that the RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. 10 CFR 50, Appendix A, GDC 34.
: 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 3                    B 3.9-30                          Revision 0, 123 March 22, 2020
 
Reactor Mode Switch Interlock Testing B 3.10.2 BASES (continued)
SURVEILLANCE      SR 3.10.2.1 and SR 3.10.2.2 REQUIREMENTS Meeting the requirements of this Special Operations LCO maintains operation consistent with or conservative to operating with the reactor mode switch in the shutdown position (or the refuel position for MODE 5). The functions of the reactor mode switch interlocks that are not in effect, due to the testing in progress, are adequately compensated for by the Special Operations LCO requirements. The administrative controls are to be periodically verified to ensure that the operational requirements continue to be met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 7.2.3.7.
: 2. FSAR, Section 14.5.3.3.
: 3. FSAR, Section 14.5.3.4.
BFN-UNIT 3                        B 3.10-12                          Revision 0, 123 March 22, 2020
 
Single Control Rod Withdrawal - Hot Shutdown B 3.10.3 BASES (continued)
SURVEILLANCE      SR 3.10.3.1, SR 3.10.3.2, and SR 3.10.3.3 REQUIREMENTS The other LCOs made applicable in this Special Operations LCO are required to have their Surveillances met to establish that this Special Operations LCO is being met. If the local array of control rods is inserted and disarmed (electrically or hydraulically) while the scram function for the withdrawn rod is not available, periodic verification in accordance with SR 3.10.3.2 is required to preclude the possibility of criticality.
SR 3.10.3.2 has been modified by a Note, which clarifies that this SR is not required to be met if SR 3.10.3.1 is satisfied for LCO 3.10.3.d.1 requirements, since SR 3.10.3.2 demonstrates that the alternative LCO 3.10.3.d.2 requirements are satisfied.
Also, SR 3.10.3.3 verifies that all control rods other than the control rod being withdrawn are fully inserted. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. FSAR, Section 14.5.3.3.
BFN-UNIT 3                        B 3.10-18                          Revision 0, 123 March 22, 2020
 
Single Control Rod Withdrawal - Cold Shutdown B 3.10.4 BASES (continued)
SURVEILLANCE      SR 3.10.4.1, SR 3.10.4.2, SR 3.10.4.3, and SR 3.10.4.4 REQUIREMENTS The other LCOs made applicable by this Special Operations LCO are required to have their associated Surveillances met to establish that this Special Operations LCO is being met. If the local array of control rods is inserted and disarmed (electrically or hydraulically) while the scram function for the withdrawn rod is not available, periodic verification is required to ensure that the possibility of criticality remains precluded. Verification that all the other control rods are fully inserted is required to meet the SDM requirements. Verification that a control rod withdrawal block has been inserted ensures that no other control rods can be inadvertently withdrawn under conditions when position indication instrumentation is inoperable for the affected control rod. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.10.4.2 and SR 3.10.4.4 have been modified by Notes, which clarify that these SRs are not required to be met if the alternative requirements demonstrated by SR 3.10.4.1 are satisfied.
REFERENCES        1. FSAR, Section 14.5.3.3.
BFN-UNIT 3                        B 3.10-25                            Revision 0, 123 March 22, 2020
 
Single CRD Removal - Refueling B 3.10.5 BASES SURVEILLANCE SR 3.10.5.1, SR 3.10.5.2, SR 3.10.5.3, SR 3.10.5.4, REQUIREMENTS and SR 3.10.5.5 (continued) control rod. The Surveillance for LCO 3.1.1, which is made applicable by this Special Operations LCO, is required in order to establish that this Special Operations LCO is being met.
Verification that no other CORE ALTERATIONS are being made is required to ensure the assumptions of the safety analysis are satisfied.
Periodic verification of the administrative controls established by this Special Operations LCO is prudent to preclude the possibility of an inadvertent criticality. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. FSAR, Section 14.5.3.3.
BFN-UNIT 3                    B 3.10-31                          Revision 0, 123 March 22, 2020
 
Multiple Control Rod Withdrawal - Refueling B 3.10.6 BASES (continued)
ACTIONS          A.1, A.2, A.3.1, and A.3.2 If one or more of the requirements of this Special Operations LCO are not met, the immediate implementation of these Required Actions restores operation consistent with the normal requirements for refueling (i.e., all control rods inserted in core cells containing one or more fuel assemblies) or with the exceptions granted by this Special Operations LCO. The Completion Times for Required Action A.1, Required Action A.2, Required Action A.3.1, and Required Action A.3.2 are intended to require that these Required Actions be implemented in a very short time and carried through in an expeditious manner to either initiate action to restore the affected CRDs and insert their control rods, or initiate action to restore compliance with this Special Operations LCO.
SURVEILLANCE      SR 3.10.6.1, SR 3.10.6.2, and SR 3.10.6.3 REQUIREMENTS Periodic verification of the administrative controls established by this Special Operations LCO is prudent to preclude the possibility of an inadvertent criticality. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. SR 3.10.6.3 is modified by a Note stating that the SR is only required to be met during refueling.
REFERENCES        1. FSAR, Section 14.5.3.3.
BFN-UNIT 3                          B 3.10-35                          Revision 0, 123 March 22, 2020
 
SDM Test -Refueling B 3.10.8 BASES (continued)
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 this Special Operations LCO, are required to have applicable Surveillances met to establish that this Special Operations LCO is being met. However, the control rod withdrawal sequences during the SDM tests may be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2 requirements) or by a second licensed operator or other qualified member of the technical staff (i.e., personnel trained in accordance with an approved training program for this test). As noted, either the applicable SRs for the RWM (LCO 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.8.2), or the proper movement of control rods must be verified (SR 3.10.8.3).
This latter verification (i.e., SR 3.10.8.3) must be performed during control rod movement to prevent deviations from the specified sequence. These Surveillances provide adequate assurance that the specified test sequence is being followed.
SR 3.10.8.4 Periodic verification of the administrative controls established by this LCO will ensure that the reactor is operated within the bounds of the safety analysis. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
BFN-UNIT 3                          B 3.10-47                            Revision 123 Amendment No. 213 March 22, 2020
 
SDM Test -Refueling B 3.10.8 BASES SURVEILLANCE SR 3.10.8.5 REQUIREMENTS (continued) Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the "full out" notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events.
SR 3.10.8.6 CRD charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram signal. Since the reactor is depressurized in MODE 5, there is insufficient reactor pressure to scram the control rods. Verification of charging water pressure ensures that if a scram is required, capability for rapid control rod insertion would exist. The minimum pressure of 940 psig, which is well below the expected pressure of approximately 1100 psig, ensures sufficient pressure for rapid control rod insertion. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES  1. NEDE-24011-P-A-13, "General Electric Standard Application for Reactor Fuel," August 1996.
: 2. Letter from T. Pickens (BWROG) to G. C. Lainas, NRC, "Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," August 15, 1986.
BFN-UNIT 3                    B 3.10-48                            Revision 0, 123 March 22, 2020
 
ENCLOSURE 3 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Requirements Manual (TRM) Changes and Additions BFN Unit 1 TRM Requirement Changes TRMR Page No.                    Revision No. Effective Date App. B (TVA-COLR-BF1C13, Revision 1)            143            03/05/2020 App. B (TVA-COLR-BF1C14, Revision 0)            147            10/16/2020 BFN Unit 2 TRM Requirement Changes TRMR Page No.                    Revision No. Effective Date 3.3-1                                            142            03/05/2020 3.3-2                                            142            03/05/2020 3.3-35                                          142            03/05/2020 App. B (TVA-COLR-BF2C21, Revision 1)            145            04/27/2020 App. B (TVA-COLR-BF2C22, Revision 0)            149            03/02/2021 BFN Unit 3 TRM Requirement Changes TRMR Page No.                    Revision No. Effective Date App. B (TVA-COLR-BF3C20, Revision 0)            144            03/10/2020 App. B (TVA-COLR-BF3C20, Revision 1)            146            10/16/2020
 
ENCLOSURE 3 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Technical Requirements Manual (TRM) Changes and Additions BFN Unit 1 TRM Bases Changes TRMB Page No.                    Revision No. Effective Date 3.3-1                                            148            12/02/2020 3.3-2                                            148            12/02/2020 3.3-35                                          148            12/02/2020 BFN Unit 2 TRM Bases Changes TRMB Page No.                    Revision No. Effective Date 3.3-1                                            148            12/02/2020 3.3-2                                            148            12/02/2020 3.3-35                                          148            12/02/2020 BFN Unit 3 TRM Bases Changes TRMB Page No.                    Revision No. Effective Date 3.3-1                                            148            12/02/2020 3.3-2                                            148            12/02/2020 3.3-35                                          148            12/02/2020
 
EDMS L94 200113 800 QA Document BFE-4349 , Revision 1 Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga , TN 37402 Browns Ferry Unit 1 Cycle 13 Core Operating Limits Report, (120% OLTP, MELLLA+)
TVA-COLR-Bf 1C13                      Revision 1 (Final)
(Revision Log, Page v)
January 2020
                                                                                      ~
Date : ...._ _    11--+-
                                                                                                      , _cl_o..1..o Verified :                                                    Date :
                              ~;_6itfifchell, Engineer Approved :    /4) ~/ ~ " -
0 . M. Brown, General Manager Date :      1/t7
                                                                                          ,          /4020 Reviewed :      ~~                                            Date : _ _*_/ _, - ~-~0_"':_G-_*, _
D. D. Coffey;-M"anager, Reactor Engineering Approved:    //~
                            ~ffuiTr-man, PORC t-- U          ///Mc,//2,V (A:.,,f ll::-1-<4,\J)
Date : _l_,,--_2_2_..-_ u_:.J_Zu    _\
I    .'1 Approved :
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                                  ~  Manager Date : ___,_( &#xb5;/h=?:+/-==tcl'2,L. _ _
I    l
 
EDMS: L94 200113 800 Reactor Engineering and Fuels - BWRFE NPG Date: January 16, 2020 1101 Market Street, Chattanooga TN 37402 Table of Contents Total Number of Pages = 50 (including review cover sheet)
List of Tables ............................................................................................................................. iii List of Figures ............................................................................................................................ iv Revision Log ...............................................................................................................................v Nomenclature ............................................................................................................................ vi References .............................................................................................................................. viii 1      Introduction ........................................................................................................................1 1.1        Purpose .......................................................................................................................1 1.2        Scope ..........................................................................................................................1 1.3        Fuel Loading................................................................................................................1 1.4        Acceptability ................................................................................................................2 2      APLHGR Limits ..................................................................................................................3 2.1        Rated Power and Flow Limit: APLHGRRATED ...............................................................3 2.2        Off-Rated Power Dependent Limit: APLHGRP ............................................................3 2.2.1            Startup without Feedwater Heaters..................................................................... 3 2.3        Off-Rated Flow Dependent Limit: APLHGRF ...............................................................3 2.4        Single Loop Operation Limit: APLHGRSLO ...................................................................3 2.5        Equipment Out-Of-Service Corrections........................................................................6 3      LHGR Limits.......................................................................................................................7 3.1        Rated Power and Flow Limit: LHGRRATED....................................................................7 3.2        Off-Rated Power Dependent Limit: LHGRP .................................................................7 3.2.1            Startup without Feedwater Heaters..................................................................... 7 3.3        Off-Rated Flow Dependent Limit: LHGRF....................................................................8 3.4        Equipment Out-Of-Service Corrections........................................................................8 4      OLMCPR Limits................................................................................................................19 4.1        Flow Dependent MCPR Limit: MCPRF ......................................................................19 4.2        Power Dependent MCPR Limit: MCPRP ...................................................................19 4.2.1            Startup without Feedwater Heaters....................................................................19 4.2.2            Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS) ................................20 4.2.3            Exposure Dependent Limits ...............................................................................20 4.2.4            Equipment Out-Of-Service (EOOS) Options ......................................................21 4.2.5            Single-Loop-Operation (SLO) Limits ..................................................................21 4.2.6            Below Pbypass Limits ........................................................................................21 5      Thermal-Hydraulic Stability Protection..............................................................................33 6      APRM Flow Biased Rod Block Trip Settings.....................................................................35 7      Rod Block Monitor (RBM) Trip Setpoints and Operability .................................................36 8      Shutdown Margin Limit .....................................................................................................38 Appendix A:            MBSP Maps..................................................................................................... A-1 Browns Ferry Unit 1 Cycle 13                                                                                                          Page ii Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 mil NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 List of Tables Nuclear Fuel Types ................................................................................................................ 2 Startup Feedwater Temperature Basis....................................................................................... 7 Nominal Scram Time Basis .......................................................................................................20 MCPRP Limits for All Fuel Types: Optimum Scram Time Basis .............................................23 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis ...............................................24 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis ........................26 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Nominal Scram Time Basis ...................................................................................................28 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Nominal Scram Time Basis ...................................................................................................29 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Technical Specification Scram Time Basis ............................................................................30 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Technical Specification Scram Time Basis ............................................................................31 MCPRP Limits for All Fuel Types: Single Loop Operation for All Scram Times ......................32 ABSP Setpoints for the Scram Region ......................................................................................33 Analyzed MBSP Endpoints: Nominal Feedwater Temperature.................................................34 Analyzed MBSP Endpoints: Reduced Feedwater Temperature ...............................................34 Analytical RBM Trip Setpoints ...............................................................................................36 RBM Setpoint Applicability ........................................................................................................36 Control Rod Withdrawal Error Results.......................................................................................37 Browns Ferry Unit 1 Cycle 13                                                                                                Page iii Core Operating Limits Report, (120% OLTP, MELLLA+)                                            TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 mil NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 List of Figures APLHGRRATED for ATRIUM-10 Fuel............................................................................................ 4 APLHGRRATED for ATRIUM-10XM Fuel....................................................................................... 5 LHGRRATED for ATRIUM-10 Fuel................................................................................................. 9 LHGRRATED for ATRIUM-10XM Fuel ..........................................................................................10 Base Operation LHGRFACP for ATRIUM-10 Fuel .....................................................................11 Base Operation LHGRFACP for ATRIUM-10XM Fuel................................................................12 LHGRFACF for ATRIUM-10 Fuel...............................................................................................13 LHGRFACF for ATRIUM-10XM Fuel .........................................................................................14 Startup Operation LHGRFACP for ATRIUM-10 Fuel: Table 3.1 Temperature Range 1..............15 Startup Operation LHGRFACP for ATRIUM-10 Fuel: Table 3.1 Temperature Range 2..............16 Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 1 ........17 Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 2 ........18 MCPRF for All Fuel Types .........................................................................................................22 MBSP Boundaries For Nominal Feedwater Temperature........................................................ A-2 MBSP Boundaries For Reduced Feedwater Temperature ...................................................... A-3 Browns Ferry Unit 1 Cycle 13                                                                                                Page iv Core Operating Limits Report, (120% OLTP, MELLLA+)                                            TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 Reactor Engineering and Fuels - BWRFE NPG Date: January 16, 2020 1101 Market Street, Chattanooga TN 37402 Revision Log Number            Page                                      Description 1-R1            vi-vii  Updated nomenclature pages for new items.
2-R1            viii  Updated References 1 and 4 Added New Reference (26) for the DSS-CD topical report supporting 3-R1            ix-x    MELLLA+. Other References incremented by +1 for renumbering.
Updated References 27-30. Removed Reference 31.
In Section 1.2, the Oscillation Power Range Monitor (OPRM) Setpoint 4-R1              1    is no longer applicable for MELLLA+ operation. This part has been more generically renamed Hydraulic Stability Protection 5-R1          11-18    Updated Figures 3.3-3.10 as needed for new information 6-R1            20    In Section 4.2.3, exposure dependent limits updated as needed.
7-R1          23-32    Updated Tables 4.2-4.9 as needed.
Re-write of Section 5 for DSS-CD stability change. Tables 5.1, 5.2, 8-R1          33-34 and 5.3 are new information supporting MELLLA+
9-R1            35    Updated to include two loop settings, clarify definitions Added clarifying note regarding limit applicability for mode 3, 4, & 5 10-R1              38 activity.
11-R1          A1-A3    New material supporting MELLLA+
0-R0            All    New document.
Browns Ferry Unit 1 Cycle 13                                                                                      Page v Core Operating Limits Report, (120% OLTP, MELLLA+)                                      TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 mil NPG                          Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Nomenclature ABSP                  Automatic Backup Stability Protection APLHGR                Average Planar LHGR APRM                  Average Power Range Monitor AREVA NP              Vendor (Framatome, Siemens)
BOC                    Beginning of Cycle BSP                    Backup Stability Protection BWR                    Boiling Water Reactor CAVEX                  Core Average Exposure CD                    Coast Down CMSS                  Core Monitoring System Software COLR                  Core Operating Limits Report CPR                    Critical Power Ratio CRWE                  Control Rod Withdrawal Error CSDM                  Cold SDM DIVOM                  Delta CPR over Initial CPR vs. Oscillation Magnitude DSS-CD                Detect and Suppress Solution - Confirmation Density EOC                    End of Cycle EOCLB                  End-of-Cycle Licensing Basis EOOS                  Equipment OOS EPU                    Extended Power Uprate (120% OLTP)
FFTR                  Final Feedwater Temperature Reduction FFWTR                  Final Feedwater Temperature Reduction FHOOS                  Feedwater Heaters OOS ft                    Foot: English unit of measure for length GNF                    Vendor (General Electric, Global Nuclear Fuels)
GWd                    Giga Watt Day HTSP                  High TSP ICA                    Interim Corrective Action ICF                    Increased Core Flow (beyond rated)
IS                    In-Service kW                    kilo watt: SI unit of measure for power.
LCO                    License Condition of Operation LFWH                  Loss of Feedwater Heating LHGRFAC                LHGR Multiplier (Power or Flow dependent)
LPRM                  Low Power Range Monitor LRNB                  Generator Load Reject, No Bypass Browns Ferry Unit 1 Cycle 13                                                                                    Page vi Core Operating Limits Report, (120% OLTP, MELLLA+)                                    TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 mil NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 MAPFAC                MAPLHGR multiplier (Power or Flow dependent)
MBSP                  Manual Backup Stability Protection MCPR                  Minimum CPR MELLLA                Maximum Extended Load Line Limit Analysis MELLLA+                Maximum Extended Load Line Limit Analysis Plus MSRV                  Moisture Separator Reheater Valve MSRVOOS                MSRV OOS MTU                    Metric Ton Uranium MWd/MTU                Mega Watt Day per Metric Ton Uranium NEOC                  Near EOC NRC                    United States Nuclear Regulatory Commission NSS                    Nominal Scram Speed NTSP                  Nominal TSP OLMCPR                MCPR Operating Limit OLTP                  Original Licensed Thermal Power OOS                    Out-Of-Service OPRM                  Oscillation Power Range Monitor OSS                    Optimum Scram Speed PBDA                  Period Based Detection Algorithm Pbypass                Power, below which TSV Position and TCV Fast Closure Scrams are Bypassed PLU                    Power Load Unbalance PLUOOS                PLU OOS PRNM                  Power Range Neutron Monitor RBM                    Rod Block Monitor RCPOOS                Recirculation Pump OOS (SLO)
RDF                    Rated Drive Flow RPS                    Reactor Protection System RPT                    Recirculation Pump Trip RPTOOS                RPT OOS RTP                    Rated Thermal Power SDM                    Shutdown Margin SLMCPR                MCPR Safety Limit SLO                    Single Loop Operation TBV                    Turbine Bypass Valve TBVIS                  TBV IS TBVOOS                Turbine Bypass Valves OOS TIP                    Transversing In-core Probe TIPOOS                TIP OOS TLO                    Two Loop Operation TSP                    Trip Setpoint TSSS                  Technical Specification Scram Speed TVA                    Tennessee Valley Authority Browns Ferry Unit 1 Cycle 13                                                                                Page vii Core Operating Limits Report, (120% OLTP, MELLLA+)                                  TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 Reactor Engineering and Fuels - BWRFE NPG Date: January 16, 2020 1101 Market Street, Chattanooga TN 37402 References
: 1.        ANP-3802, Revision 1, Browns Ferry Unit 1 Cycle 13 MELLLA+ Reload Analysis, Framatome, Inc., January 2020.
: 2.        ANP-3031P, Revision 3, Mechanical Design Report for Browns Ferry Units 1, 2, and 3 ATRIUM-10 Fuel Assemblies, AREVA NP, Inc., May 2014.
: 3.        ANP-3150P, Revision 4, Mechanical Design Report for Browns Ferry ATRIUM 10XM Fuel Assemblies, AREVA Inc., November 2017.
: 4.        ANP-3648P, Revision 1, Browns Ferry Unit 1 Cycle 13 Plant Parameters Document, Framatome Inc., May 2019.
: 5.        BFE-4361, Revision 0, Browns Ferry Unit 1 Cycle 13 In Core Shuffle, Calculation File, Tennessee Valley Authority, September 2018.
Methodology References
: 6.        XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
: 7.        XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
: 8.        EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
: 9.        ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
: 10.      XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.
: 11.      XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.
: 12.      EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.
: 13.      XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.
: 14.      XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.
Browns Ferry Unit 1 Cycle 13                                                                                Page viii Core Operating Limits Report, (120% OLTP, MELLLA+)                                  TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 Reactor Engineering and Fuels - BWRFE NPG Date: January 16, 2020 1101 Market Street, Chattanooga TN 37402
: 15.      ANP-10307PA, Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP Inc., June 2011.
: 16.      ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.
: 17.      ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Advanced Nuclear Fuels Corporation, September 2005.
: 18.      EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP Inc.,
September 2009.
: 19.      EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001, as supplemented by the site specific approval in NRC safety evaluation dated April 27, 2012.
: 20.      EMF-2292(P)(A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
: 21.      EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
: 22.      BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP Inc., May 2008.
: 23.      BAW-10247PA, Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP Inc., April 2008.
: 24.      ANP-10298PA, Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc., March 2010.
: 25.      ANP-3140(P), Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc.,
August 2012.
: 26.      NEDC-33075P-A, Revision 8, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, GE Hitachi, November 2013.
Setpoint References
: 27.      EDQ2092900118, R35, Setpoint and Scaling Calculation for Neutron Monitoring &
Recirculation Flow Loops, Calculation File, Tennessee Valley Authority, August 9, 2019.
: 28.      Task T0500, Revision 0, Neutron Monitoring System w/RBM, Project Task Report, GE Hitachi Nuclear Energy, June 2017.
: 29.      Task T0506, Revision 0, TS Instrument Setpoints, Project Task Report, Tennessee Valley Authority, August, 2017.
: 30.      NEDC-33006P-A, Revision 3, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, GE Energy Nuclear, June 2009.
Browns Ferry Unit 1 Cycle 13                                                                                  Page ix Core Operating Limits Report, (120% OLTP, MELLLA+)                                  TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 mil NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 1 Introduction In anticipation of cycle startup, it is necessary to describe the expected limits of operation.
1.1    Purpose The primary purpose of this document is to satisfy requirements identified by unit technical specification section 5.6.5. This document may be provided, upon final approval, to the NRC..
1.2    Scope This document will discuss the following areas:
3/4 Average Planar Linear Heat Generation Rate (APLHGR) Limit (Technical Specifications 3.2.1 and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3.2.3, 3.3.4.1, and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Minimum Critical Power Ratio Operating Limit (OLMCPR)
(Technical Specifications 3.2.2, 3.3.4.1, 3.7.5 and Table 3.3.2.1-1)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Thermal-Hydraulic Stability Protection (Technical Specification Table 3.3.1.1)
Applicability: Mode 1,  (as specified in Technical Specifications Table 3.3.1.1-1) 3/4 Average Power Range Monitor (APRM) Flow Biased Rod Block Trip Setting (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
Applicability: Mode 1,  (as specified in Technical Requirements Manuals Table 3.3.4-1) 3/4 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
Applicability: Mode 1,  % RTP as specified in Table 3.3.2.1-1 (TS definition of RTP) 3/4 Shutdown Margin (SDM) Limit (Technical Specification 3.1.1)
Applicability: All Modes 1.3    Fuel Loading The core will contain previously exposed Framatome, Inc., ATRIUM-10, along with fresh and exposed ATRIUM-10XM. Nuclear fuel types used in the core loading are shown in Table 1.1.
The core shuffle and final loading were explicitly evaluated for BOC cold shutdown margin performance as documented per Reference 5.
Browns Ferry Unit 1 Cycle 13                                                                                  Page 1 Core Operating Limits Report, (120% OLTP, MELLLA+)                                  TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 1.1 Nuclear Fuel Types
* Nuclear Original        Number of          Fuel Type            Fuel Names Fuel Description                              Cycle          Assemblies            (NFT)                (Range)
ATRIUM-10 A10-3840B-14GV80-FAB                                      11                  49                23              FAB301-FAB500 ATRIUM-10 A10-4117B-13GV70-FAB                                      11                  72                24              FAB501-FAB572 ATRIUM-10 A10-4112B-15GV70-FAB                                      11                31                25              FAB573-FAB608 ATRIUM-10XM XMLC-4102B-11GV70-FAC-B                                  12                  40                26              FAC701-FAC740 ATRIUM-10XM XMLC-3969B-13GV80-FAC-C                                  12                128                27              FAC741-FAC868 ATRIUM-10XM XMLC-3948B-13GV70-FAC-B                                  12                112                28              FAC869-FAC980 ATRIUM-10XM XMLC-3967B-15GV80-FAD-B                                  13                164                29              FAD001-FAD164 ATRIUM-10XM XMLC-3945B-14GV80-FAD-C                                  13                72                30              FAD165-FAD236 ATRIUM-10XM XMLC-3951B-14GV80-FAD-C                                  13                40                31              FAD237-FAD276 ATRIUM-10XM XMLC-4091B-13GV80-FAD-B                                  13                  56                32              FAD277-FAD332 1.4      Acceptability Limits discussed in this document were generated based on NRC approved methodologies per References 6 through 25.
* The table identifies the expected fuel type breakdown in anticipation of final core loading. The final composition of the core depends upon uncertainties during the outage such as discovering a failed fuel bundle, or other bundle damage. Minor core loading changes, due to unforeseen events, will conform to the safety and monitoring requirements identified in this document.
Browns Ferry Unit 1 Cycle 13                                                                                                            Page 2 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                                        TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                          Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 2 APLHGR Limits (Technical Specifications 3.2.1 & 3.7.5)
The APLHGR limit is determined by adjusting the rated power APLHGR limit for off-rated power, off-rated flow, and SLO conditions. The most limiting of these is then used as follows:
APLHGR limit = MIN ( APLHGRP , APLHGRF, APLHGRSLO )
where:
APLHGRP              off-rated power APLHGR limit                [APLHGRRATED
* MAPFACP]
APLHGRF              off-rated flow APLHGR limit                  [APLHGRRATED
* MAPFACF]
APLHGRSLO            SLO APLHGR limit                            [APLHGRRATED
* SLO Multiplier]
2.1    Rated Power and Flow Limit: APLHGRRATED The rated conditions APLHGR for all fuel are identified per Reference 1. The rated conditions APLHGR for ATRIUM-10 fuel are shown in Figure 2.1. The rated conditions APLHGR for ATRIUM-10XM are shown in Figure 2.2.
2.2    Off-Rated Power Dependent Limit: APLHGRP Reference 1 does not specify a power dependent APLHGR. Therefore, MAPFACP is set to a value of 1.0.
2.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. No additional power dependent limitation is required.
2.3    Off-Rated Flow Dependent Limit: APLHGRF Reference 1 does not specify a flow dependent APLHGR. Therefore, MAPFACF is set to a value of 1.0.
2.4    Single Loop Operation Limit: APLHGRSLO The single loop operation multiplier for ATRIUM-10 and ATRIUM-10XM fuel is 0.85, per Reference 1.
Browns Ferry Unit 1 Cycle 13                                                                                      Page 3 Core Operating Limits Report, (120% OLTP, MELLLA+)                                      TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 15 12 APLHGR (kW/ft) 9 6
3 0
0              20                    40                    60                      80 Planar Average Exposure (GWd/MTU)
Planar Avg.      APLHGR Exposure          Limit (GWd/MTU)          (kW/ft) 0.0            12.5 15.0            12.5 67.0              7.3 Figure 2.1 APLHGRRATED for ATRIUM-10 Fuel Browns Ferry Unit 1 Cycle 13                                                                                                Page 4 Core Operating Limits Report,        (120% OLTP, MELLLA+)                                        TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 15 12 APLHGR (kW/ft) 9 6
3 0
0              20                    40                    60                      80 Planar Average Exposure (GWd/MTU)
Planar Avg.      APLHGR Exposure          Limit (GWd/MTU)          (kW/ft) 0.0            13.0 15.0            13.0 67.0              7.6 Figure 2.2 APLHGRRATED for ATRIUM-10XM Fuel Browns Ferry Unit 1 Cycle 13                                                                                                Page 5 Core Operating Limits Report,        (120% OLTP, MELLLA+)                                        TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                          Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 2.5      Equipment Out-Of-Service Corrections The limits shown in Figure 2.1 and Figure 2.2 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
In-Service                              All equipment In-Service
* RPTOOS                                  EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                  Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                  Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                        Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                  One Recirculation Pump Out-Of-Service
* All equipment service conditions assume 1 SRVOOS.
Browns Ferry Unit 1 Cycle 13                                                                                        Page 6 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                          Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 3 LHGR Limits (Technical Specification 3.2.3, 3.3.4.1, & 3.7.5)
The LHGR limit is determined by adjusting the rated power LHGR limit for off-rated power and off-rated flow conditions. The most limiting of these is then used as follows:
LHGR limit = MIN ( LHGRP, LHGRF )
where:
LHGRP                off-rated power LHGR limit                  [LHGRRATED
* LHGRFACP]
LHGRF                off-rated flow LHGR limit                    [LHGRRATED
* LHGRFACF]
3.1    Rated Power and Flow Limit: LHGRRATED The rated conditions LHGR for all fuel are identified per Reference 1. The rated conditions LHGR for ATRIUM-10 are shown in Figure 3.1. The rated conditions LHGR for ATRIUM-10XM fuel is shown in Figure 3.2. The LHGR limit is consistent with References 2 and 3.
3.2    Off-Rated Power Dependent Limit: LHGRP LHGR limits are adjusted for off-rated power conditions using the LHGRFACP multiplier provided in Reference 1. The multiplier is split into two sub cases: turbine bypass valves in and out-of-service. The base case multipliers are shown in Figure 3.3 and Figure 3.4.
3.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional limits are shown in Figure 3.7 through Figure 3.10, based on temperature conditions identified in Table 3.1.
Table 3.1 Startup Feedwater Temperature Basis Temperature Power            Range 1                Range 2
(% Rated)              (&deg;F)                  (&deg;F) 23                155.0                  150.0 30                162.0                  157.0 40                172.0                  167.0 50                182.0                  177.0 Browns Ferry Unit 1 Cycle 13                                                                                      Page 7 Core Operating Limits Report, (120% OLTP, MELLLA+)                                      TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                          Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 3.3      Off-Rated Flow Dependent Limit: LHGRF LHGR limits are adjusted for off-rated flow conditions using the LHGRFACF multiplier provided in Reference 1. Multipliers are shown in Figure 3.5 through Figure 3.6.
3.4      Equipment Out-Of-Service Corrections The limits shown in Figure 3.1 and Figure 3.2 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
* In-Service                            All equipment In-Service RPTOOS                                EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                      Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                One Recirculation Pump Out-Of-Service Off-rated power corrections shown in Figure 3.3 and Figure 3.4 are dependent on operation of the Turbine Bypass Valve system. For this reason, separate limits are to be applied for TBVIS or TBVOOS operation. The limits have no dependency on RPTOOS, PLUOOS, FHOOS/FFWTR, or SLO.
Off-rated flow corrections shown in Figure 3.5 and Figure 3.6 are bounding for all EOOS conditions.
Off-rated power corrections shown in Figure 3.7 through Figure 3.10 are also dependent on operation of the Turbine Bypass Valve system. In this case, limits support FHOOS operation during startup. These limits have no dependency on RPTOOS, PLUOOS, or SLO.
* All equipment service conditions assume 1 SRVOOS.
Browns Ferry Unit 1 Cycle 13                                                                                        Page 8 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 15 12 9
LHGR (kW/ft) 6 3
0 0              20                    40                    60                      80 Pellet Exposure (GWd/MTU)
Pellet              LHGR Exposure            Limit (GWd/MTU)          (kW/ft) 0.0            13.4 18.9              13.4 74.4              7.1 Figure 3.1 LHGRRATED for ATRIUM-10 Fuel Browns Ferry Unit 1 Cycle 13                                                                                              Page 9 Core Operating Limits Report,      (120% OLTP, MELLLA+)                                        TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 15 12 9
LHGR (kW/ft) 6 3
0 0              20                    40                    60                      80 Pellet Exposure (GWd/MTU)
Pellet              LHGR Exposure            Limit (GWd/MTU)          (kW/ft) 0.0            14.1 18.9              14.1 74.4              7.4 Figure 3.2 LHGRRATED for ATRIUM-10XM Fuel Browns Ferry Unit 1 Cycle 13                                                                                              Page 10 Core Operating Limits Report,      (120% OLTP, MELLLA+)                                        TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 1.10 1.00 0.90 0.80 Turbine Bypass Valve In-Service, TBVIS LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 7%9,6 &RUH)ORZ TBVIS, > 50% Core Flow 7%9226 &RUH)ORZ 0.50 TBVOOS, > 50% Core Flow 0.40 0.30 0.20 20      30          40            50        60        70        80      90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power          LHGRFACP                    Power          LHGRFACP
(% Rated)                                    (% Rated) 100.0        1.00                          100.0        0.95 77.6        0.77                            77.6        0.75 26.0        0.61                            26.0        0.60 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0        0.53                            26.0        0.45 23.0        0.50                            23.0        0.42
                                &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0        0.57                            26.0        0.52 23.0        0.54                            23.0        0.47 Figure 3.3 Base Operation LHGRFACP for ATRIUM-10 Fuel (Independent of other EOOS conditions)
Browns Ferry Unit 1 Cycle 13                                                                                                Page 11 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                              TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 Reactor Engineering and Fuels - BWRFE lID:J                                                                                      Date: January 16, 2020 1101 Market Street, Chattanooga TN 37402 NPG 1.10 1.00 0.90 Turbine Bypass Valve In-Service, TBVIS 0.80 Turbine Bypass Valve Out-of-Service, TBVOOS 0.70 0.60 LHGRFACP 0.50            7%9,6 &RUH)ORZ 7%9226 &RUH)ORZ
                            ~    TBVIS, > 50% Core Flow 0.40 TBVOOS, > 50% Core Flow
                              /
0.30 0.20 20        30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                          Core Power            LHGRFACP                    Power            LHGRFACP
(% Rated)                                    (% Rated) 100.0        1.00                            100.0        1.00 26.0        0.63                            26.0        0.62 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0        0.45                            26.0        0.38 23.0        0.42                            23.0        0.35
                                  &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0        0.49                            26.0        0.48 23.0        0.46                            23.0        0.44 Figure 3.4 Base Operation LHGRFACP for ATRIUM-10XM Fuel (Independent of other EOOS conditions)
Browns Ferry Unit 1 Cycle 13                                                                                                  Page 12 Core Operating Limits Report,    (120% OLTP, MELLLA+)                                              TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                    Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 1.10 1.05 1.00 0.95 0.90 LHGRFACF 0.85 0.80 0.75 0.70 0.65 0.60 0.55 20      30      40        50        60        70        80        90        100        110 Core Flow (% Rated)
Core Flow            LHGRFACF
(% Rated) 0.0              1.00 30.0            1.00 107.0            1.00 Figure 3.5 LHGRFACF for ATRIUM-10 Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 1 Cycle 13                                                                                          Page 13 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                        TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 Reactor Engineering and Fuels - BWRFE lID:J                                                                          Date: January 16, 2020 1101 Market Street, Chattanooga TN 37402 NPG 1.10 1.05 1.00 0.95 0.90 0.85 LHGRFACF    0.80 0.75 0.70 0.65 0.60 0.55 20      30      40        50        60        70        80        90        100        110 Core Flow (% Rated)
Core Flow            LHGRFACF
(% Rated) 0.0              0.61 30.0            0.61 78.1            1.00 107.0            1.00 Figure 3.6 LHGRFACF for ATRIUM-10XM Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 1 Cycle 13                                                                                        Page 14 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                      TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 1.10 1.00 0.90 0.80 Turbine Bypass Valve In-Service, TBVIS LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 7%9,6 &RUH)ORZ 0.50          TBVIS, > 50% Core Flow 7%9226 &RUH)ORZ TBVOOS, > 50% Core Flow 0.40 0.30 0.20 20      30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power          LHGRFACP                    Power            LHGRFACP
(% Rated)                                  (% Rated) 100.0        1.00                          100.0        0.95 77.6        0.77                            77.6        0.75 26.0        0.54                            26.0        0.54 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0        0.49                            26.0        0.42 23.0        0.46                            23.0        0.39
                                &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0        0.52                            26.0        0.48 23.0        0.49                            23.0        0.44 Figure 3.7 Startup Operation LHGRFACP for ATRIUM-10 Fuel:
Table 3.1 Temperature Range 1 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 1 Cycle 13                                                                                                Page 15 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                              TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                          Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 1.10 1.00 0.90 0.80 Turbine Bypass Valve In-Service, TBVIS LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 7%9,6 &RUH)ORZ 0.50          TBVIS, > 50% Core Flow 7%9226 &RUH)ORZ TBVOOS, > 50% Core Flow 0.40 0.30 0.20 20      30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                  Power            LHGRFACP
(% Rated)                                  (% Rated) 100.0        1.00                          100.0        0.95 77.6        0.77                            77.6        0.75 26.0        0.54                            26.0        0.54 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0        0.49                            26.0        0.42 23.0        0.46                            23.0        0.39
                                &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0        0.52                            26.0        0.48 23.0        0.49                            23.0        0.40 Figure 3.8 Startup Operation LHGRFACP for ATRIUM-10 Fuel:
Table 3.1 Temperature Range 2 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 1 Cycle 13                                                                                                Page 16 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                              TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                          Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 1.10 1.00 0.90 0.80          Turbine Bypass Valve In-Service, TBVIS Turbine Bypass Valve Out-of-Service, TBVOOS LHGRFACP 0.70 0.60 0.50 7%9,6 &RUH)ORZ 7%9226 &RUH)ORZ 0.40 TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20      30          40          50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                    Power          LHGRFACP
(% Rated)                                    (% Rated) 100.0        1.00                          100.0        1.00 26.0        0.53                            26.0        0.52 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0        0.39                            26.0        0.35 23.0        0.37                            23.0        0.32
                                &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0        0.43                            26.0        0.43 23.0        0.40                            23.0        0.38 Figure 3.9 Startup Operation LHGRFACP for ATRIUM-10XM Fuel:
Table 3.1 Temperature Range 1 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 1 Cycle 13                                                                                                Page 17 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                              TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 1.10 1.00 0.90 0.80          Turbine Bypass Valve In-Service, TBVIS Turbine Bypass Valve Out-of-Service, TBVOOS LHGRFACP 0.70 0.60 0.50 7%9,6 &RUH)ORZ 7%9226 &RUH)ORZ 0.40 TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20      30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power          LHGRFACP                    Power          LHGRFACP
(% Rated)                                  (% Rated) 100.0        1.00                          100.0        1.00 26.0        0.53                          26.0        0.52 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0        0.39                          26.0        0.35 23.0        0.36                          23.0        0.32
                                &RUH)ORZ5DWHG                      &RUH)ORZ5DWHG 26.0        0.42                          26.0        0.42 23.0        0.38                          23.0        0.38 Figure 3.10 Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 2 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 1 Cycle 13                                                                                              Page 18 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                            TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                      Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 4 OLMCPR Limits (Technical Specification 3.2.2, 3.3.4.1, & 3.7.5)
OLMCPR is calculated to be the most limiting of the flow or power dependent values OLMCPR limit = MAX ( MCPRF , MCPRP )
where:
MCPRF          core flow-dependent MCPR limit MCPRP          power-dependent MCPR limit 4.1    Flow Dependent MCPR Limit: MCPRF MCPRF limits are dependent upon core flow (% of Rated), and the max core flow limit, (Rated or Increased Core Flow, ICF). MCPRF limits are shown in Figure 4.1, per Reference 1. Limits are valid for all EOOS combinations. No adjustment is required for SLO conditions.
4.2    Power Dependent MCPR Limit: MCPRP MCPRP limits are dependent upon:
Core Power Level (% of Rated)
Technical Specification Scram Speed (TSSS), Nominal Scram Speed (NSS), or Optimum Scram Speed (OSS)
Cycle Operating Exposure (NEOC, EOC, and CD - as defined in this section)
Equipment Out-Of-Service Options Two or Single recirculation Loop Operation (TLO vs. SLO)
The MCPRP limits are provided in Table 4.2 through Table 4.9, where each table contains the limits for all fuel types and EOOS options (for a specified scram speed and exposure range).
The CMSS determines MCPRP limits, from these tables, based on linear interpolation between the specified powers.
4.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional power dependent limits are shown in Table 4.5 through Table 4.8 based on temperature conditions identified in Table 3.1.
Browns Ferry Unit 1 Cycle 13                                                                                  Page 19 Core Operating Limits Report, (120% OLTP, MELLLA+)                                    TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 4.2.2    Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS)
MCPRP limits are provided for three different sets of assumed scram speeds. The Technical Specification Scram Speed (TSSS) MCPRP limits are applicable at all times, as long as the scram time surveillance demonstrates the times in Technical Specification Table 3.1.4-1 are met. Both Nominal Scram Speeds (NSS) and/or Optimum Scram Speeds (OSS) may be used, as long as the scram time surveillance demonstrates Table 4.1 times are applicable.
* Table 4.1 Nominal Scram Time Basis Notch                    Nominal              Optimum Position            Scram Timing          Scram Timing (index)                (seconds)              (seconds) 46                        0.421                  0.392 36                        0.991                  0.887 26                        1.620                  1.487 6                      3.040                  3.040 In demonstrating compliance with the NSS and/or OSS scram time basis, surveillance requirements from Technical Specification 3.1.4 apply; accepting the definition of SLOW rods should conform to scram speeds shown in Table 4.1. If conformance is not demonstrated, TSSS based MCPRP limits are applied.
On initial cycle startup, TSSS limits are used until the successful completion of scram timing confirms NSS and/or OSS based limits are applicable.
4.2.3    Exposure Dependent Limits Exposures are tracked on a Core Average Exposure basis (CAVEX, not Cycle Exposure).
Higher exposure MCPRP limits are always more limiting and may be used for any Core Average Exposure up to the ending exposure. Per Reference 1, MCPRP limits are provided for the following exposure ranges:
BOC to NEOC                                NEOC corresponds to                        29,058.4 MWd / MTU BOC to EOCLB                                EOCLB corresponds to                        33,485.8 MWd / MTU BOC to End of Coast                        End of Coast                                35,008.8 MWd / MTU NEOC refers to a Near EOC exposure point.
* Reference 1 analysis results are based on information identified in Reference 4.
Drop out times consistent with method used to perform actual timing measurements (i.e., including pickup/dropout effects).
Browns Ferry Unit 1 Cycle 13                                                                                                    Page 20 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                      TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 The EOCLB exposure point is not the true End-Of-Cycle exposure. Instead it corresponds to a licensing exposure window exceeding expected end-of-full-power-life.
The End of Coast exposure point represents a licensing exposure point exceeding the expected end-of-cycle exposure including cycle extension options.
4.2.4      Equipment Out-Of-Service (EOOS) Options EOOS options
* covered by MCPRP limits are given by the following:
In-Service                                          All equipment In-Service RPTOOS                                              EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                              Turbine Bypass Valve(s) Out-Of-Service RPTOOS+TBVOOS                                      Combined RPTOOS and TBVOOS PLUOOS                                              Power Load Unbalance Out-Of-Service PLUOOS+RPTOOS                                      Combined PLUOOS and RPTOOS PLUOOS+TBVOOS                                      Combined PLUOOS and TBVOOS PLUOOS+TBVOOS+RPTOOS                                Combined PLUOOS, RPTOOS, and TBVOOS FHOOS (or FFWTR)                                    Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)
RCPOOS                                              One Recirculation Pump Out-Of-Service For exposure ranges up to NEOC and EOCLB, additional combinations of MCPRP limits are also provided including FHOOS. The coast down exposure range assumes application of FFWTR. FHOOS based MCPRP limits for the coast down exposure are redundant because the temperature setdown assumption is identical with FFWTR.
4.2.5      Single-Loop-Operation (SLO) Limits When operating in RCPOOS conditions, MCPRp limits are constructed differently from the normal operating RCP conditions. The limiting event for RCPOOS is a pump seizure scenario, which sets the upper bound for allowed core power and flow . This event is not impacted by scram time assumptions. Specific MCPRP limits are shown in Table 4.9.
4.2.6      Below Pbypass Limits Below Pbypass (26% rated power), MCPRP limits depend upon core flow. One set of MCPRP limits applies for core flow above 50% of rated; a second set applies if the core flow is less than or equal to 50% rated.
* All equipment service conditions assume 1 SRVOOS.
RCPOOS limits are only valid up to 43.75% rated core power, 50% rated core flow, and an active recirculation drive flow of 17.73 Mlbm/hr.
Browns Ferry Unit 1 Cycle 13                                                                                                        Page 21 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 2.00 1.80 1.60 MCPRF 1.40 1.20 1.00 30        40      50          60          70          80          90        100          110 Core Flow (% Rated)
Core Flow            MCPRF
(% Rated) 30.0              1.58 84.0              1.34 107.0              1.34 Figure 4.1 MCPRF for All Fuel Types (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 1 Cycle 13                                                                                          Page 22 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                        TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 4.2 MCPRP Limits for All Fuel Types: Optimum Scram Time Basis
* ATRIUM-10                              ATRIUM-10XM BOC        BOC          BOC          BOC          BOC          BOC Pow er            to          to      to End of        to            to      to End of Operating Condition        (% of rated)      NEOC      EOCLB        Coast        NEOC        EOCLB        Coast 100            1.47        1.50        1.54        1.39          1.40        1.44 90            1.53        1.56        1.59        1.44          1.45        1.49 65            1.71        1.72        1.76        1.59          1.60        1.64 50            1.79        1.79        1.85        1.70          1.70        1.76 50            2.03        2.03        2.03        1.79          1.80        1.80 Base Case                  40            2.10        2.10        2.10        1.87          1.87        1.93 26            2.34        2.34        2.46        2.29          2.29        2.41 26 at > 50%F        2.65        2.65        2.75        2.60          2.60        2.70 23 at > 50%F        2.81        2.81        2.92        2.76          2.76        2.87
                                  DW)        2.57        2.57        2.67        2.52          2.52        2.62
                                  DW)        2.73        2.73        2.85        2.68          2.68        2.80 100            1.51        1.54          ---        1.43          1.44          ---
90            1.57        1.59          ---        1.48          1.49          ---
65            1.74        1.76          ---        1.64          1.64          ---
50            1.85        1.85          ---        1.76          1.76          ---
50            2.03        2.03          ---        1.80          1.80          ---
FHOOS                      40            2.10        2.10          ---        1.93          1.93          ---
26            2.46        2.46          ---        2.41          2.41          ---
26 at > 50%F        2.75        2.75          ---        2.70          2.70          ---
23 at > 50%F        2.92        2.92          ---        2.87          2.87          ---
                                  DW)        2.67        2.67          ---        2.62          2.62          ---
                                  DW)        2.85        2.85          ---        2.80          2.80          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR/FHOOS is supported for the BOC to End of Coast limits.
Browns Ferry Unit 1 Cycle 13                                                                                                    Page 23 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                                    TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10                      ATRIUM-10XM BOC      BOC        BOC        BOC        BOC          BOC Pow er          to        to    to End of      to          to      to End of Operating Condition      (% of rated)    NEOC    EOCLB      Coast      NEOC      EOCLB          Coast 100          1.50    1.53        1.56      1.42      1.43          1.46 90          1.57    1.59        1.62      1.47      1.48          1.51 65          1.73    1.74        1.79      1.62      1.62          1.67 50          1.82    1.82        1.88      1.72      1.72          1.79 50          2.04    2.04        2.04      1.80      1.82          1.82 Base Case              40          2.11    2.11        2.11      1.87      1.88          1.96 26          2.37    2.37        2.49      2.32      2.32          2.44 26 at > 50%F      2.65    2.65        2.75      2.60      2.60          2.70 23 at > 50%F      2.81    2.81        2.92      2.76      2.76          2.87
                                        DW)      2.57    2.57        2.67      2.52      2.52          2.62
                                        DW)      2.73    2.73        2.85      2.68      2.68          2.80 100          1.56    1.59        1.62      1.45      1.46          1.49 90          1.62    1.65        1.67      1.50      1.51          1.54 65          1.79    1.81        1.84      1.65      1.65          1.69 50          1.86    1.86        1.92      1.74      1.74          1.80 50          2.04    2.04        2.04      1.80      1.82          1.82 TBVOOS                40          2.11    2.11        2.12      1.88      1.88          1.96 26          2.37    2.37        2.49      2.32      2.32          2.44 26 at > 50%F      3.17    3.17        3.29      3.12      3.12          3.24 23 at > 50%F      3.44    3.44        3.56      3.39      3.39          3.51
                                        DW)      2.91    2.91        3.05      2.86      2.86          3.00
                                        DW)      3.20    3.20        3.35      3.15      3.15          3.30 100          1.54    1.56        ---      1.46      1.46          ---
90          1.60    1.62        ---      1.51      1.51          ---
65          1.79    1.79        ---      1.67      1.67          ---
50          1.88    1.88        ---      1.79      1.79          ---
50          2.04    2.04        ---      1.80      1.82          ---
FHOOS                  40          2.11    2.11        ---      1.96      1.96          ---
26          2.49    2.49        ---      2.44      2.44          ---
26 at > 50%F      2.75    2.75        ---      2.70      2.70          ---
23 at > 50%F      2.92    2.92        ---      2.87      2.87          ---
                                        DW)      2.67    2.67        ---      2.62      2.62          ---
                                        DW)      2.85    2.85        ---      2.80      2.80          ---
100          1.51    1.53        1.56      1.42      1.43          1.46 90          1.57    1.59        1.62      1.47      1.48          1.51 65          1.96    1.96        1.96      1.73      1.75          1.75 50          ---      ---        ---        ---        ---          ---
50          2.04    2.04        2.04      1.81      1.82          1.82 PLUOOS                40          2.11    2.11        2.11      1.87      1.88          1.96 26          2.37    2.37        2.49      2.32      2.32          2.44 26 at > 50%F      2.65    2.65        2.75      2.60      2.60          2.70 23 at > 50%F      2.81    2.81        2.92      2.76      2.76          2.87
                                        DW)      2.57    2.57        2.67      2.52      2.52          2.62
                                        DW)      2.73    2.73        2.85      2.68      2.68          2.80
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 1 Cycle 13                                                                                                    Page 24 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                                    TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis (continued)
* ATRIUM-10                        ATRIUM-10XM BOC      BOC        BOC        BOC        BOC          BOC Pow er          to      to      to End of      to          to      to End of Operating Condition      (% of rated)    NEOC    EOCLB      Coast      NEOC      EOCLB        Coast 100          1.59    1.62        ---        1.48        1.49          ---
90          1.65    1.67        ---        1.53        1.54          ---
65          1.84    1.84        ---        1.68        1.69          ---
50          1.92    1.92        ---        ---        1.80          ---
50          2.04    2.04        ---        1.80        1.82          ---
TBVOOS 40          2.12    2.12        ---        1.96        1.96          ---
FHOOS 26          2.49    2.49        ---        2.44        2.44          ---
26 at > 50%F      3.29    3.29        ---        3.24        3.24          ---
23 at > 50%F      3.56    3.56        ---        3.51        3.51          ---
                                      DW)      3.05    3.05        ---        3.00        3.00          ---
                                      DW)      3.35    3.35        ---        3.30        3.30          ---
100          1.56    1.59        1.62        1.45        1.46        1.49 90          1.62    1.65        1.67        1.50        1.51        1.54 65          1.96    1.96        1.96        1.73        1.75        1.75 50            ---      ---        ---        ---        ---          ---
50          2.04    2.04        2.04        1.81        1.82        1.82 TBVOOS 40          2.11    2.11        2.12        1.88        1.88        1.96 PLUOOS 26          2.37    2.37        2.49        2.32        2.32        2.44 26 at > 50%F      3.17    3.17        3.29        3.12        3.12        3.24 23 at > 50%F      3.44    3.44        3.56        3.39        3.39        3.51
                                      DW)      2.91    2.91        3.05        2.86        2.86        3.00
                                      DW)      3.20    3.20        3.35        3.15        3.15        3.30 100          1.54    1.56        ---        1.46        1.46          ---
90          1.60    1.62        ---        1.51        1.51          ---
65          1.96    1.96        ---        1.73        1.75          ---
50            ---      ---        ---        ---        ---          ---
50          2.04    2.04        ---        1.81        1.82          ---
FHOOS 40          2.11    2.11        ---        1.96        1.96          ---
PLUOOS 26          2.49    2.49        ---        2.44        2.44          ---
26 at > 50%F      2.75    2.75        ---        2.70        2.70          ---
23 at > 50%F      2.92    2.92        ---        2.87        2.87          ---
                                      DW)      2.67    2.67        ---        2.62        2.62          ---
                                      DW)      2.85    2.85        ---        2.80        2.80          ---
100          1.59    1.62        ---        1.48        1.49          ---
90          1.65    1.67        ---        1.53        1.54          ---
65          1.96    1.96        ---        1.73        1.75          ---
50            ---      ---        ---        ---        ---          ---
TBVOOS                  50          2.04    2.04        ---        1.81        1.82          ---
FHOOS                    40          2.12    2.12        ---        1.96        1.96          ---
PLUOOS                  26          2.49    2.49        ---        2.44        2.44          ---
26 at > 50%F      3.29    3.29        ---        3.24        3.24          ---
23 at > 50%F      3.56    3.56        ---        3.51        3.51          ---
                                      DW)      3.05    3.05        ---        3.00        3.00          ---
                                      DW)      3.35    3.35        ---        3.30        3.30          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 1 Cycle 13                                                                                                    Page 25 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                                  TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10                        ATRIUM-10XM BOC      BOC        BOC        BOC        BOC          BOC Pow er          to      to      to End of      to          to      to End of Operating Condition      (% of rated)    NEOC    EOCLB      Coast      NEOC      EOCLB        Coast 100          1.55    1.55        1.59        1.45        1.45        1.49 90          1.62    1.62        1.64        1.50        1.50        1.53 65          1.77    1.77        1.83        1.64        1.64        1.70 50          1.85    1.85        1.91        1.75        1.75        1.82 50          2.05    2.05        2.05        1.81        1.83        1.83 Base Case                40          2.11    2.11        2.12        1.89        1.89        1.98 26          2.39    2.39        2.52        2.34        2.34        2.47 26 at > 50%F      2.65    2.65        2.75        2.60        2.60        2.70 23 at > 50%F      2.81    2.81        2.92        2.76        2.76        2.87
                                      DW)      2.57    2.57        2.67        2.52        2.52        2.62
                                      DW)      2.73    2.73        2.85        2.68        2.68        2.80 100          1.60    1.67        1.69        1.48        1.49        1.52 90          1.67    1.67        1.69        1.53        1.54        1.56 65          1.83    1.83        1.88        1.67        1.67        1.71 50          1.90    1.90        1.96        1.76        1.76          ---
50          2.05    2.05        2.05        1.81        1.83        1.83 TBVOOS                  40          2.11    2.11        2.15        1.90        1.90        1.99 26          2.39    2.39        2.52        2.34        2.34        2.47 26 at > 50%F      3.17    3.17        3.29        3.12        3.12        3.24 23 at > 50%F      3.44    3.44        3.56        3.39        3.39        3.51
                                      DW)      2.91    2.91        3.05        2.86        2.86        3.00
                                      DW)      3.20    3.20        3.35        3.15        3.15        3.30 100          1.57    1.59        ---        1.48        1.49          ---
90          1.64    1.64        ---        1.53        1.53          ---
65          1.83    1.83        ---        1.70        1.70          ---
50          1.91    1.91        ---        ---        1.82          ---
50          2.05    2.05        ---        1.82        1.83          ---
FHOOS                    40          2.12    2.12        ---        1.98        1.98          ---
26          2.52    2.52        ---        2.47        2.47          ---
26 at > 50%F      2.75    2.75        ---        2.70        2.70          ---
23 at > 50%F      2.92    2.92        ---        2.87        2.87          ---
                                      DW)      2.67    2.67        ---        2.62        2.62          ---
                                      DW)      2.85    2.85        ---        2.80        2.80          ---
100          1.55    1.63        1.68        1.45        1.45        1.49 90          1.62    1.63        1.68        1.50        1.50        1.53 65          1.97    1.97        1.97        1.74        1.76        1.76 50            ---      ---        ---        ---        ---          ---
50          2.05    2.05        2.05        1.82        1.84        1.84 PLUOOS                  40          2.11    2.11        2.12        1.89        1.89        1.98 26          2.39    2.39        2.52        2.34        2.34        2.47 26 at > 50%F      2.65    2.65        2.75        2.60        2.60        2.70 23 at > 50%F      2.81    2.81        2.92        2.76        2.76        2.87
                                      DW)      2.57    2.57        2.67        2.52        2.52        2.62
                                      DW)      2.73    2.73        2.85        2.68        2.68        2.80
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 1 Cycle 13                                                                                                    Page 26 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                                  TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis (continued)
* ATRIUM-10                        ATRIUM-10XM BOC      BOC        BOC        BOC        BOC          BOC Pow er          to      to      to End of      to          to      to End of Operating Condition      (% of rated)    NEOC    EOCLB      Coast      NEOC      EOCLB        Coast 100          1.62    1.67        ---        1.51        1.52          ---
90          1.69    1.69        ---        1.56        1.56          ---
65          1.88    1.88        ---        1.71        1.71          ---
50          1.96    1.96        ---        ---        ---          ---
50          2.05    2.05        ---        1.83        1.83          ---
TBVOOS 40          2.15    2.15        ---        1.99        1.99          ---
FHOOS 26          2.52    2.52        ---        2.47        2.47          ---
26 at > 50%F      3.29    3.29        ---        3.24        3.24          ---
23 at > 50%F      3.56    3.56        ---        3.51        3.51          ---
                                        DW)      3.05    3.05        ---        3.00        3.00          ---
                                        DW)      3.35    3.35        ---        3.30        3.30          ---
100          1.60    1.67        1.69        1.48        1.49        1.52 90          1.67    1.67        1.69        1.53        1.54        1.56 65          1.97    1.97        1.97        1.74        1.76        1.76 50            ---      ---        ---        ---        ---          ---
50          2.05    2.05        2.05        1.82        1.84        1.84 TBVOOS 40          2.11    2.11        2.15        1.90        1.90        1.99 PLUOOS 26          2.39    2.39        2.52        2.34        2.34        2.47 26 at > 50%F      3.17    3.17        3.29        3.12        3.12        3.24 23 at > 50%F      3.44    3.44        3.56        3.39        3.39        3.51
                                        DW)      2.91    2.91        3.05        2.86        2.86        3.00
                                        DW)      3.20    3.20        3.35        3.15        3.15        3.30 100          1.57    1.63        ---        1.48        1.49          ---
90          1.64    1.64        ---        1.53        1.53          ---
65          1.97    1.97        ---        1.74        1.76          ---
50            ---      ---        ---        ---        ---          ---
50          2.05    2.05        ---        1.82        1.84          ---
FHOOS 40          2.12    2.12        ---        1.98        1.98          ---
PLUOOS 26          2.52    2.52        ---        2.47        2.47          ---
26 at > 50%F      2.75    2.75        ---        2.70        2.70          ---
23 at > 50%F      2.92    2.92        ---        2.87        2.87          ---
                                        DW)      2.67    2.67        ---        2.62        2.62          ---
                                        DW)      2.85    2.85        ---        2.80        2.80          ---
100          1.62    1.67        ---        1.51        1.52          ---
90          1.69    1.69        ---        1.56        1.56          ---
65          1.97    1.97        ---        1.74        1.76          ---
50            ---      ---        ---        ---        ---          ---
TBVOOS                  50          2.05    2.05        ---        1.83        1.84          ---
FHOOS                    40          2.15    2.15        ---        1.99        1.99          ---
PLUOOS                  26          2.52    2.52        ---        2.47        2.47          ---
26 at > 50%F      3.29    3.29        ---        3.24        3.24          ---
23 at > 50%F      3.56    3.56        ---        3.51        3.51          ---
                                        DW)      3.05    3.05        ---        3.00        3.00          ---
                                        DW)      3.35    3.35        ---        3.30        3.30          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 1 Cycle 13                                                                                                    Page 27 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                                  TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 4.5 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10                      ATRIUM-10XM BOC        BOC        BOC      BOC        BOC        BOC Pow er            to          to    to End of    to          to      to End of Operating Condition          (% of rated)      NEOC        EOCLB      Coast    NEOC        EOCLB      Coast 100            1.54        1.56      1.56    1.46        1.46        1.46 90            1.60        1.62      1.62    1.51        1.51        1.51 65            1.96        1.96      1.96    1.73        1.75        1.75 50              ---          ---        ---    ---        ---        ---
50            2.07        2.07      2.07    1.95        1.95        1.95 TBVIS                      40            2.32        2.32      2.32    2.16        2.16        2.16 26            2.82        2.82      2.82    2.77        2.77        2.77 26 at > 50%F        3.03        3.03      3.03    2.98        2.98        2.98 23 at > 50%F        3.28        3.28      3.28    3.23        3.23        3.23
                                    DW)        2.96        2.96      2.96    2.91        2.91        2.91
                                    DW)        3.19        3.19      3.19    3.14        3.14        3.14 100            1.59        1.62      1.62    1.48        1.49        1.49 90            1.65        1.67      1.67    1.53        1.54        1.54 65            1.96        1.96      1.96    1.73        1.75        1.75 50              ---          ---        ---    ---        ---        ---
50            2.09        2.09      2.09    1.95        1.95        1.95 TBVOOS                      40            2.33        2.33      2.33    2.16        2.16        2.16 26            2.82        2.82      2.82    2.77        2.77        2.77 26 at > 50%F        3.53        3.53      3.53    3.48        3.48        3.48 23 at > 50%F        3.83        3.83      3.83    3.78        3.78        3.78
                                    DW)        3.29        3.29      3.29    3.24        3.24        3.24
                                    DW)        3.62        3.62      3.62    3.57        3.57        3.57
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 1 Cycle 13                                                                                                  Page 28 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                              TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 4.6 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10                      ATRIUM-10XM BOC        BOC        BOC      BOC        BOC        BOC Pow er            to          to    to End of    to          to      to End of Operating Condition          (% of rated)      NEOC        EOCLB      Coast    NEOC        EOCLB      Coast 100            1.54        1.56      1.56    1.46        1.46        1.46 90            1.60        1.62      1.62    1.51        1.51        1.51 65            1.96        1.96      1.96    1.73        1.75        1.75 50              ---          ---        ---    ---        ---        ---
50            2.07        2.07      2.07    1.96        1.96        1.96 TBVIS                      40            2.33        2.33      2.33    2.17        2.17        2.17 26            2.83        2.83      2.83    2.78        2.78        2.78 26 at > 50%F        3.05        3.05      3.05    3.00        3.00        3.00 23 at > 50%F        3.30        3.30      3.30    3.25        3.25        3.25
                                    DW)        2.98        2.98      2.98    2.93        2.93        2.93
                                    DW)        3.27        3.27      3.27    3.22        3.22        3.22 100            1.59        1.62      1.62    1.48        1.49        1.49 90            1.65        1.67      1.67    1.53        1.54        1.54 65            1.96        1.96      1.96    1.73        1.75        1.75 50              ---          ---        ---    ---        ---        ---
50            2.10        2.10      2.10    1.96        1.96        1.96 TBVOOS                      40            2.34        2.34      2.34    2.17        2.17        2.17 26            2.83        2.83      2.83    2.78        2.78        2.78 26 at > 50%F        3.54        3.54      3.54    3.49        3.49        3.49 23 at > 50%F        3.84        3.84      3.84    3.79        3.79        3.79
                                    DW)        3.30        3.30      3.30    3.25        3.25        3.25
                                    DW)        3.73        3.73      3.73    3.68        3.68        3.68
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 1 Cycle 13                                                                                                  Page 29 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                              TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 4.7 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10                      ATRIUM-10XM BOC        BOC        BOC      BOC        BOC        BOC Pow er            to          to    to End of    to          to      to End of Operating Condition          (% of rated)      NEOC        EOCLB      Coast    NEOC        EOCLB      Coast 100            1.57        1.63      1.68    1.48        1.49        1.49 90            1.64        1.64      1.68    1.53        1.53        1.53 65            1.97        1.97      1.97    1.74        1.76        1.76 50              ---          ---        ---    ---        ---        ---
50            2.09        2.09      2.09    1.97        1.97        1.97 TBVIS                      40            2.35        2.35      2.35    2.19        2.19        2.19 26            2.84        2.84      2.84    2.79        2.79        2.79 26 at > 50%F        3.03        3.03      3.03    2.98        2.98        2.98 23 at > 50%F        3.28        3.28      3.28    3.23        3.23        3.23
                                    DW)        2.96        2.96      2.96    2.91        2.91        2.91
                                    DW)        3.19        3.19      3.19    3.14        3.14        3.14 100            1.62        1.67      1.69    1.51        1.52        1.52 90            1.69        1.69      1.69    1.56        1.56        1.56 65            1.97        1.97      1.97    1.74        1.76        1.76 50              ---          ---        ---    ---        ---        ---
50            2.13        2.13      2.13    1.97        1.97        1.97 TBVOOS                      40            2.37        2.37      2.37    2.19        2.19        2.19 26            2.84        2.84      2.84    2.79        2.79        2.79 26 at > 50%F        3.53        3.53      3.53    3.48        3.48        3.48 23 at > 50%F        3.83        3.83      3.83    3.78        3.78        3.78
                                    DW)        3.29        3.29      3.29    3.24        3.24        3.24
                                    DW)        3.62        3.62      3.62    3.57        3.57        3.57
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 1 Cycle 13                                                                                                  Page 30 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                              TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 4.8 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10                      ATRIUM-10XM BOC        BOC        BOC      BOC        BOC        BOC Pow er            to          to    to End of    to          to      to End of Operating Condition          (% of rated)      NEOC        EOCLB      Coast    NEOC        EOCLB      Coast 100            1.57        1.63      1.68    1.48        1.49        1.49 90            1.64        1.64      1.68    1.53        1.53        1.53 65            1.97        1.97      1.97    1.74        1.76        1.76 50              ---          ---        ---    ---        ---        ---
50            2.10        2.10      2.10    1.98        1.98        1.98 TBVIS                      40            2.36        2.36      2.36    2.19        2.19        2.19 26            2.86        2.86      2.86    2.81        2.81        2.81 26 at > 50%F        3.05        3.05      3.05    3.00        3.00        3.00 23 at > 50%F        3.30        3.30      3.30    3.25        3.25        3.25
                                    DW)        2.98        2.98      2.98    2.93        2.93        2.93
                                    DW)        3.27        3.27      3.27    3.22        3.22        3.22 100            1.62        1.67      1.69    1.51        1.52        1.52 90            1.69        1.69      1.69    1.56        1.56        1.56 65            1.97        1.97      1.97    1.74        1.76        1.76 50              ---          ---        ---    ---        ---        ---
50            2.14        2.14      2.14    1.98        1.98        1.98 TBVOOS                      40            2.38        2.38      2.38    2.19        2.19        2.19 26            2.86        2.86      2.86    2.81        2.81        2.81 26 at > 50%F        3.54        3.54      3.54    3.49        3.49        3.49 23 at > 50%F        3.84        3.84      3.84    3.79        3.79        3.79
                                    DW)        3.30        3.30      3.30    3.25        3.25        3.25
                                    DW)        3.73        3.73      3.73    3.68        3.68        3.68
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 1 Cycle 13                                                                                                  Page 31 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                              TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 4.9 MCPRP Limits for All Fuel Types: Single Loop Operation for All Scram Times
* Pow er            BOC to End of COAST Operating Condition      (% of rated)      ATRIUM-10      ATRIUM-10XM 100        2.12            2.05 43.75        2.12            2.05 40        2.14            2.05 RCPOOS                        26        2.54            2.49 FHOOS              26 at > 50%F          2.77            2.72 23 at > 50%F          2.94            2.89
                                                        DW)          2.69            2.64
                                                        DW)          2.87            2.82 100        2.14            2.05 43.75        2.14            2.05 RCPOOS                        40        2.17            2.05 TBVOOS                        26        2.54            2.49 PLUOOS            26 at > 50%F          3.31            3.26 FHOOS              23 at > 50%F          3.58            3.53
                                                        DW)          3.07            3.02
                                                        DW)          3.37            3.32 100        2.30            2.13 43.75        2.30            2.13 40        2.39            2.21 RCPOOS 26        2.86            2.81 TBVOOS 26 at > 50%F          3.55            3.50 FHOOS1 23 at > 50%F          3.85            3.80
                                                        DW)          3.31            3.26
                                                        DW)          3.64            3.59 100        2.31            2.14 43.75        2.31            2.14 40        2.40            2.21 RCPOOS 26        2.88            2.83 TBVOOS 26 at > 50%F          3.56            3.51 FHOOS2 23 at > 50%F          3.86            3.81
                                                        DW)          3.32            3.27
                                                        DW)          3.75            3.70
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop.
RCPOOS limits are only valid up to 50% rated core power, 50% rated core flow, and an active recirculation drive flow of 17.73 Mlbm/hr.
Browns Ferry Unit 1 Cycle 13                                                                                                      Page 32 Core Operating Limits Report,  (120% OLTP, MELLLA+)                                                    TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 5      Thermal-Hydraulic Stability Protection (Technical Specification 3.3.1.1)
Technical Specification Table 3.3.1.1-1, Function 2f, identifies the function.
Instrument setpoints are established, such that the reactor will be tripped before an oscillation can grow to the point where the SLMCPR is exceeded. With application of Reference 30, the DSS-CD stability solution will be used per Reference 26. The DSS-CD SAD setpoint is 1.10 for TLO and SLO.
New analyses have been developed based on Reference 26. With the implementation of the MELLLA+ operating domain expansion, an ABSP trip is required when the OPRM is out-of-service. The ABSP trip settings define a region of the power to flow map within which an automatic reactor scram occurs. The ABSP trip settings are provided in Table 5.1. If both the OPRM and ABSP are out-of-service, operation within the MEL/LA+ domain is not allowed and the MBSP Regions provide stability protection. Table 5.2 and Table 5.3 provide the endpoints for the MBSP regions for nominal and reduced feedwater temperature conditions.
Table 5.1 ABSP Setpoints for the Scram Region Parameter            Symbol          Setting Value (unit)                        Comments Slope of ABSP APRM low Flow Slope for Trip          m TRIP        2.00 (% RTP/% RDF)
Biased Trip Linear Segment ABSP APRM Flow Biased Trip Constant Power                                                      Setpoint Power Intercept. Constant PBSP-TRIP          35.0 (% RTP)
Line for Trip                                                      Power Line for Trip from Zero Drive Flow to Flow Breakpoint Value ABSP APRM Flow Biased Trip Constant Flow                                                      Setpoint Drive Flow Intercept.
WBSP-TRIP            49 (% RDF)
Line for Trip                                                      Constant Flow Line for Trip (see Note 1 below)
Flow Breakpoint        WBSP-BREAK          30.0 (% RDF)              Flow Breakpoint Value Note 1: WBSP-TRIP can be set to 49.0 % RDF or any higher value up to the intersection of the ABSP sloped line w ith the APRM Flow Biased STP scram line.
Browns Ferry Unit 1 Cycle 13                                                                                                  Page 33 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                      Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 5.2 Analyzed MBSP Endpoints: Nominal Feedwater Temperature Power      Core Flow Endpoint                                                Definition
(% Rated)    (% Rated)
Scram Region (Region I)
A1        75.9          52.7          Boundary Intercept on MELLLA+ Line Scram Region (Region I)
B1        35.5          29            Boundary Intercept on Natural Circulation Line (NCL)
Controlled Entry Region (Region A2        66.1          52            II) Boundary Intercept on MELLLA Line Controlled Entry Region (Region B2        25.5          29            II) Boundary Intercept on Natural Circulation Line (NCL)
Table 5.3 Analyzed MBSP Endpoints: Reduced Feedwater Temperature Power      Core Flow Endpoint                                                Definition
(% Rated)    (% Rated)
Scram Region (Region I)
A1        64.9          50.5          Boundary Intercept on MELLLA Line Scram Region (Region I)
B1        29.4          29            Boundary Intercept on Natural Circulation Line (NCL)
Controlled Entry Region (Region A2        68.3          54.9          II) Boundary Intercept on MELLLA Line Controlled Entry Region (Region B2        24.5          29            II) Boundary Intercept on Natural Circulation Line (NCL)
Browns Ferry Unit 1 Cycle 13                                                                                  Page 34 Core Operating Limits Report, (120% OLTP, MELLLA+)                                    TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                          Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 6 APRM Flow Biased Rod Block Trip Settings (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
The APRM rod block trip setting is based upon References 27 & 29, and is defined by the following:
for two loop operation:
SRB d (0.61Wd + 63.3)                            Allowable Value SRB d (0.61Wd + 62.0)                            Nominal Trip Setpoint (NTSP) where:
SRB          =        Rod Block setting in percent of rated thermal power (3952 MWt)
Wd          =        Recirculation drive flow rate in percent of rated (100% drive flow required to achieve 100% core power and flow) and for single loop operation:
SRB d (0.55(Wd-'W) + 60.5)                                Allowable Value SRB d (0.55(Wd-'W) + 58.5)                                Nominal Trip Setpoint (NTSP) where:
SRB          =        Rod Block setting in percent of rated thermal power (3952 MWt)
Wd          =        Recirculation drive flow rate in percent of rated (100% drive flow required to achieve 100% core power and flow)
          'W          =        Difference between two-loop and single-loop effective recirculation flow at the same core flow ('W=0.0 for two-loop operation)
The APRM rod block trip setting is clamped at a maximum allowable value of 115%
(corresponding to a NTSP of 113%).
Browns Ferry Unit 1 Cycle 13                                                                                    Page 35 Core Operating Limits Report, (120% OLTP, MELLLA+)                                      TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 7 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
The RBM trip setpoints and applicable power ranges, based on References 27 & 28, are shown in Table 7.1. Setpoints are based on an HTSP, unfiltered analytical limit of 114%. Unfiltered setpoints are consistent with a nominal RBM filter setting of 0.0 seconds; filtered setpoints are consistent with a nominal RBM filter setting less than 0.5 seconds. Cycle specific CRWE analyses of OLMCPR are documented in Reference 1, superseding values reported in References 27, 28, and 29.
Table 7.1 Analytical RBM Trip Setpoints
* Allowable              Nominal Trip RBM                              Value                Setpoint Trip Setpoint                            (AV)                  (NTSP)
LPSP                                      27%                    25%
IPSP                                      62%                    60%
HPSP                                      82%                    80%
LTSP - unfiltered                      121.7%                  120.0%
                                          - filtered                    120.7%                  119.0%
ITSP      - unfiltered                  116.7%                  115.0%
                                          - filtered                    115.7%                  114.0%
HTSP - unfiltered                      111.7%                  110.0%
                                          - filtered                    110.9%                  109.2%
DTSP                                      90%                    92%
As a result of cycle specific CRWE analyses, RBM setpoints in Technical Specification Table 3.3.2.1-1 are applicable as shown in Table 7.2. Cycle specific setpoint analysis results are shown in Table 7.3, per Reference 1.
Table 7.2 RBM Setpoint Applicability Thermal Power                      Applicable                Notes from
(% Rated)                        MCPR                Table 3.3.2.1-1              Comment
                                                        < 1.71              (a), (b), (f), (h)    two loop operation
            > 27% and < 90%
                                                        < 1.75              (a), (b), (f), (h)    single loop operation 90%                                      < 1.40                      (g)            two loop operation
* Values are considered maximums. Using lower values, due to RBM system hardware/software limitations, is conservative, and acceptable.
MCPR values shown correspond with, (support), SLMPCR values identified in Reference 1.
Greater than 90% rated power is not attainable in single loop operation.
Browns Ferry Unit 1 Cycle 13                                                                                                  Page 36 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                  TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Table 7.3 Control Rod Withdrawal Error Results RBM                          CRWE HTSP Analytical Limit                OLMCPR Unfiltered 107                            1.26 111                            1.29 114                            1.33 117                            1.37 Results, compared against the base case OLMCPR results of Table 4.2, indicate SLMCPR remains protected for RBM inoperable conditions (i.e., 114% unblocked).
Browns Ferry Unit 1 Cycle 13                                                                                  Page 37 Core Operating Limits Report, (120% OLTP, MELLLA+)                                    TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 mil NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 8 Shutdown Margin Limit (Technical Specification 3.1.1)
Assuming the strongest OPERABLE control blade is fully withdrawn, and all other OPERABLE control blades are fully inserted, the core shall be sub-critical and meet the following minimum shutdown margin:
SDM      > 0.38% dk/k Note: The basis for the SDM value is tied to manufacturing tolerance and uncertainty stack-up. This tie only has the potential to change with new fuel type introduction. Since no new fuel type is being introduced, the value is good for all Mode 3, 4, and 5 operation prior to initial cycle startup, during cycle operation, and after final cycle shutdown.
Browns Ferry Unit 1 Cycle 13                                                                                      Page 38 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                      Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Appendix A: MBSP Maps Browns Ferry Unit 1 Cycle 13                                                                                Page A-1 Core Operating Limits Report, (120% OLTP, MELLLA+)                                  TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                                Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Core Power (% Rated: 100% = 3952MWt) 110 100 MELLLA+ Region 90 80                                                                                  BSP Boundary Manual 70                                    Scram Region I 60 MELLLA Region    ICF Region MELLLA Upper Boundary 50 87.5% Rod Line 40                                                      Controlled Entry Region II 30                                                              Min. Flow Control 20 Natural                                Min. Pow er Line Circulation 10                                          20% Pump Speed Line 0
0          10        20        30        40        50        60      70    80      90    100    110        120 Core Flow (% Rated: 100% =102.5 MLbm/hr)
Figure A.1 MBSP Boundaries For Nominal Feedwater Temperature (Operation in the MELLLA+ Region Prohibited for Feedwater Temperature greater than 10 degrees F below the Nominal Feedwater Temperature)
Browns Ferry Unit 1 Cycle 13                                                                                            Page A-2 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF1C13, Revision 1 (Final)
 
EDMS: L94 200113 800 ii!i] NPG                                Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: January 16, 2020 Core Power (% Rated: 100% = 3952MWt) 110 100 MELLLA+ Region 90 80 BSP Boundary 70 Manual Scram 60                                  Region I MELLLA Region MELLLA Upper Boundary                                            ICF Region 50 87.5% Rod Line Controlled Entry 40                                                                  Region II 30                                                            Min. Flow Control 20 Natural                              Min. Pow er Line Circulation 10                                          20% Pump Speed Line 0
0          10        20        30        40        50      60        70      80      90    100      110        120 Core Flow (% Rated: 100% =102.5 MLbm/hr)
Figure A.2 MBSP Boundaries For Reduced Feedwater Temperature (Operation in the MELLLA+ Region Prohibited for a Reduced Feedwater Temperature greater than 10 degrees F below the Nominal Feedwater Temperature)
Browns Ferry Unit 1 Cycle 13                                                                                          Page A-3 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF1C13, Revision 1 (Final)
 
ECM L32 200806 800 QA Record II!fil Reactor Engineering and Fuels - BWR FE 11 01 Market Street. Chattano oga. TN 37402 BFE-4537. Revision 0 Bro wn s Fer ry Un it 1 Cy cle 14 Core Ope ratin g Limi ts Repo rt, (120% OL TP, MEL LLA+ )
TVA-CO LR-B f 1C14                      Revisio n O (Final)
(Revision Log , Page v)
Verner-Dingl e, Whitne y                      Dig itally signed by Verner-Ding le, Septe mber 2020                                                                  Wh itney Kayla Kayla                                        Da te: 2020.09.15 09:24:30 -0tl'00' Roberts Claude C Jr Digil allysigned byHoberl s, ClaudcC Jr Prepared:      ______              , ______              Date: 20 20.09.fj~                                :s_s _-0_4*0_0_* _ _ __
W. K. Verner-Dingle/C. C. Roberts . Enginee r M *, tc he 11 Brye C I
Digitally signed by Mitchell, llrye C.
* Dal e: 20 20.09.15 l 0:34:0 2 -04'00' Verified :                                                                              Date: _ __ _ __ _
8 . C. Mitchell, Enginee r o"l 'l,111,' ,.g,,"1.1    ll) SMl"1 C111w1n(" f'I ON d ( gov d( ,,.. U( "''"' uu h~tn ou-(f)!J)QtJtfl P
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en ~r,1..,1 ChrtU,nt'A, ..,r.14 c,1'-t"'ol~ I ,tv*u11v fl* .nnn    I ,tr'II d l l(]IO't"*l1tJ t h!\ nocwrrlll Approved:
                                  - - - -- -- - - - -- -                        D.!I" ]OJ(H)Qloate:04""
C. A. Setter, Manage r. BWRFE                                                                - - - - - --
Reviewed:      ~---=  _5____:.,,_.?                                                  Date:                    Cf /t.,    / lU D. D. Coffey, Manager, Reactor Engineering Approv ed:      ~c) ..J?~                                                              Date:                  9/:J.."J/ ZO :2-i?
Chairman. POR Approved      ~  ant Manage r
                                                      &#xa5;          ~
GGPS$ISJT%VOO GP S$ISJT%VO Date: _           _ _ _ _ __
BFN - Unit 1                                        Page 1 of 44                                                                TRM Revision 147 September 28, 2020
 
Table of Contents Total Number of Pages = 44 (including review cover sheet)
List of Tables.................................................................................................................................iii List of Figures .............................................................................................................................. iv Revision Log ................................................................................................................................. v Nomenclature............................................................................................................................... vi References..................................................................................................................................viii 1      Introduction ..........................................................................................................................1 1.1      Purpose .........................................................................................................................1 1.2      Scope ............................................................................................................................1 1.3      Fuel Loading..................................................................................................................1 1.4      Acceptability ..................................................................................................................2 2      APLHGR Limits ....................................................................................................................3 2.1      Rated Power and Flow Limit: APLHGRRATED ................................................................ 3 2.2      Off-Rated Power Dependent Limit: APLHGRP ............................................................. 3 2.2.1      Startup without Feedwater Heaters .......................................................................3 2.3      Off-Rated Flow Dependent Limit: APLHGRF ................................................................ 3 2.4      Single Loop Operation Limit: APLHGRSLO.................................................................... 3 2.5      Equipment Out-Of-Service Corrections......................................................................... 5 3      LHGR Limits.........................................................................................................................6 3.1      Rated Power and Flow Limit: LHGRRATED..................................................................... 6 3.2      Off-Rated Power Dependent Limit: LHGRP .................................................................. 6 3.2.1      Startup without Feedwater Heaters .......................................................................6 3.3      Off-Rated Flow Dependent Limit: LHGRF..................................................................... 7 3.4      Equipment Out-Of-Service Corrections......................................................................... 7 4      OLMCPR Limits .................................................................................................................13 4.1      Flow Dependent MCPR Limit: MCPRF ....................................................................... 13 4.2      Power Dependent MCPR Limit: MCPRP .................................................................... 13 4.2.1      Startup without Feedwater Heaters .....................................................................13 4.2.2      Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS)................................. 14 4.2.3      Exposure Dependent Limits ................................................................................14 4.2.4      Equipment Out-Of-Service (EOOS) Options .......................................................15 4.2.5      Single-Loop-Operation (SLO) Limits ...................................................................15 4.2.6      Below Pbypass Limits .........................................................................................15 5      Thermal-Hydraulic Stability Protection ............................................................................... 27 6      APRM Flow Biased Rod Block Trip Settings...................................................................... 29 7      Rod Block Monitor (RBM) Trip Setpoints and Operability .................................................. 30 8      Shutdown Margin Limit....................................................................................................... 32 Appendix A:      MBSP Maps ...................................................................................................... A-1 BFN - Unit 1                                              Page 2 of 44                                          TRM Revision 147 September 28, 2020
 
List of Tables Nuclear Fuel Types      ................................................................................................................. 2 Startup Feedwater Temperature Basis ......................................................................................... 6 Nominal Scram Time Basis ......................................................................................................... 14 MCPRP Limits for All Fuel Types: Optimum Scram Time Basis                                  ............................................ 17 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis                                .............................................. 18 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis                                        ...................... 20 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Nominal Scram Time Basis            ................................................................................................... 22 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Nominal Scram Time Basis            ................................................................................................... 23 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Technical Specification Scram Time Basis                  ............................................................................ 24 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Technical Specification Scram Time Basis                  ............................................................................ 25 MCPRP Limits for All Fuel Types: Single Loop Operation for All Scram Times                                        .................... 26 ABSP Setpoints for the Scram Region ....................................................................................... 27 Analyzed MBSP Endpoints: Nominal Feedwater Temperature ................................................. 28 Analyzed MBSP Endpoints: Reduced Feedwater Temperature ................................................ 28 Analytical RBM Trip Setpoints          ............................................................................................... 30 RBM Setpoint Applicability .......................................................................................................... 30 Control Rod Withdrawal Error Results ........................................................................................ 31 BFN - Unit 1                                        Page 3 of 44                                          TRM Revision 147 September 28, 2020
 
List of Figures APLHGRRATED for ATRIUM-10XM Fuel ......................................................................................... 4 LHGRRATED for ATRIUM-10XM Fuel .............................................................................................. 8 Base Operation LHGRFACP for ATRIUM-10XM Fuel ................................................................... 9 LHGRFACF for ATRIUM-10XM Fuel ........................................................................................... 10 Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 1 ........ 11 Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 2 ........ 12 MCPRF for All Fuel Types ........................................................................................................... 16 MBSP Boundaries For Nominal Feedwater Temperature ........................................................ A-2 MBSP Boundaries For Reduced Feedwater Temperature ....................................................... A-3 BFN - Unit 1                                    Page 4 of 44                                          TRM Revision 147 September 28, 2020
 
Revision Log Number    Page                        Description 0-R0    All New document.
BFN - Unit 1                    Page 5 of 44          TRM Revision 147 September 28, 2020
 
Nomenclature ABSP        Automatic Backup Stability Protection APLHGR      Average Planar LHGR APRM        Average Power Range Monitor AREVA NP    Vendor (Framatome, Siemens)
BOC          Beginning of Cycle BSP          Backup Stability Protection BWR          Boiling Water Reactor CAVEX        Core Average Exposure CD          Coast Down CMSS        Core Monitoring System Software COLR        Core Operating Limits Report CPR          Critical Power Ratio CRWE        Control Rod Withdrawal Error CSDM        Cold SDM DIVOM        Delta CPR over Initial CPR vs. Oscillation Magnitude DSS-CD      Detect and Suppress Solution - Confirmation Density EOC          End of Cycle EOCLB        End-of-Cycle Licensing Basis EOOS        Equipment OOS EPU          Extended Power Uprate (120% OLTP)
FFTR        Final Feedwater Temperature Reduction FFWTR        Final Feedwater Temperature Reduction FHOOS        Feedwater Heaters OOS ft          Foot: English unit of measure for length GNF          Vendor (General Electric, Global Nuclear Fuels)
GWd          Giga Watt Day HTSP        High TSP ICA          Interim Corrective Action ICF          Increased Core Flow (beyond rated)
IS          In-Service kW          kilo watt: SI unit of measure for power.
LCO          License Condition of Operation LFWH        Loss of Feedwater Heating LHGRFAC      LHGR Multiplier (Power or Flow dependent)
LPRM        Low Power Range Monitor LRNB        Generator Load Reject, No Bypass MAPFAC      MAPLHGR multiplier (Power or Flow dependent)
BFN - Unit 1                            Page 6 of 44              TRM Revision 147 September 28, 2020
 
MBSP        Manual Backup Stability Protection MCPR        Minimum CPR MELLLA      Maximum Extended Load Line Limit Analysis MELLLA+      Maximum Extended Load Line Limit Analysis Plus MSRV        Moisture Separator Reheater Valve MSRVOOS      MSRV OOS MTU          Metric Ton Uranium MWd/MTU      Mega Watt Day per Metric Ton Uranium NEOC        Near EOC NRC          United States Nuclear Regulatory Commission NSS          Nominal Scram Speed NTSP        Nominal TSP OLMCPR      MCPR Operating Limit OLTP        Original Licensed Thermal Power OOS          Out-Of-Service OPRM        Oscillation Power Range Monitor OSS          Optimum Scram Speed PBDA        Period Based Detection Algorithm Pbypass      Power, below which TSV Position and TCV Fast Closure Scrams are Bypassed PLU          Power Load Unbalance PLUOOS      PLU OOS PRNM        Power Range Neutron Monitor RBM          Rod Block Monitor RCPOOS      Recirculation Pump OOS (SLO)
RDF          Rated Drive Flow RPS          Reactor Protection System RPT          Recirculation Pump Trip RPTOOS      RPT OOS RTP          Rated Thermal Power SDM          Shutdown Margin SLMCPR      MCPR Safety Limit SLO          Single Loop Operation TBV          Turbine Bypass Valve TBVIS        TBV IS TBVOOS      Turbine Bypass Valves OOS TIP          Transversing In-core Probe TIPOOS      TIP OOS TLO          Two Loop Operation TSP          Trip Setpoint TSSS        Technical Specification Scram Speed TVA          Tennessee Valley Authority BFN - Unit 1                          Page 7 of 44                  TRM Revision 147 September 28, 2020
 
References
: 1.      ANP-3856, Revision 0, Browns Ferry Unit 1 Cycle 14 Reload Analysis, Framatome Inc., July 2020. [L94 200810 802]
: 2.      Not Used.
: 3.      ANP-3150P, Revision 4, Mechanical Design Report for Browns Ferry ATRIUM 10XM Fuel Assemblies, AREVA Inc., November 2017. [L86 171205 001]
: 4.      ANP-3830P Revision 0, Browns Ferry Unit 1 Cycle 14 Plant Parameters Document, Framatome Inc., February 2020. [L94 200706 801]
: 5.      BFE-4534, Revision 0, Browns Ferry Unit 1 Cycle 14 In-Core Shuffle, Tennessee Valley Authority, September 2020 [L32 200723 800].
Methodology References
: 6.      XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
: 7.      XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
: 8.      EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
: 9.      ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
: 10. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.
: 11. XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.
: 12. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.
: 13. XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.
: 14. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.
: 15. ANP-10307PA, Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP Inc., June 2011.
BFN - Unit 1                            Page 8 of 44                  TRM Revision 147 September 28, 2020
: 16.      ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.
: 17.      ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Advanced Nuclear Fuels Corporation, September 2005.
: 18.      EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP Inc.,
September 2009.
: 19.      EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001, as supplemented by the site specific approval in NRC safety evaluation dated April 27, 2012.
: 20.      EMF-2292(P)(A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
: 21.      EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
: 22.      BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP Inc., May 2008.
: 23.      BAW-10247PA, Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP Inc., April 2008.
: 24.      ANP-10298PA, Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc., March 2010.
: 25.      ANP-3140(P), Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc.,
August 2012.
: 26.      NEDC-33075P-A, Revision 8, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, GE Hitachi, November 2013.
Setpoint References
: 27.      EDQ2092900118, R36, Setpoint and Scaling Calculation for Neutron Monitoring &
Recirculation Flow Loops, Calculation File, Tennessee Valley Authority, August 9, 2019.
: 28.      Task T0500, Revision 0, Neutron Monitoring System w/RBM, Project Task Report, GE Hitachi Nuclear Energy, June 2017.
: 29.      Task T0506, Revision 0, TS Instrument Setpoints, Project Task Report, Tennessee Valley Authority, August, 2017.
: 30.      NEDC-33006P-A, Revision 3, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, GE Energy Nuclear, June 2009.
BFN - Unit 1                              Page 9 of 44                      TRM Revision 147 September 28, 2020
 
1 Introduction In anticipation of cycle startup, it is necessary to describe the expected limits of operation.
1.1    Purpose The primary purpose of this document is to satisfy requirements identified by unit technical specification section 5.6.5. This document may be provided, upon final approval, to the NRC.
1.2    Scope This document will discuss the following areas:
3/4 Average Planar Linear Heat Generation Rate (APLHGR) Limit (Technical Specifications 3.2.1 and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3.2.3, 3.3.4.1, and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Minimum Critical Power Ratio Operating Limit (OLMCPR)
(Technical Specifications 3.2.2, 3.3.4.1, 3.7.5 and Table 3.3.2.1-1)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Thermal-Hydraulic Stability Protection (Technical Specification Table 3.3.1.1)
Applicability: Mode 1,  (as specified in Technical Specifications Table 3.3.1.1-1) 3/4 Average Power Range Monitor (APRM) Flow Biased Rod Block Trip Setting (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
Applicability: Mode 1,  (as specified in Technical Requirements Manuals Table 3.3.4-1) 3/4 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
Applicability: Mode 1,  % RTP as specified in Table 3.3.2.1-1 (TS definition of RTP) 3/4 Shutdown Margin (SDM) Limit (Technical Specification 3.1.1)
Applicability: All Modes 1.3    Fuel Loading The core will contain fresh, and previously exposed ATRIUM-10XM. Nuclear fuel types used in the core loading are shown in Table 1.1. The core shuffle and final loading were explicitly evaluated for BOC cold shutdown margin performance as documented per Reference 5.
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Table 1.1 Nuclear Fuel Types
* Original        Number of        Nuclear Fuel              Fuel Names Fuel Description                            Cycle          Assemblies          Type (NFT)                (Range)
ATRIUM 10XM XMLC-4102B-11GV70-FAC-B                                12                35                  26              FAC701-FAC740 ATRIUM 10XM XMLC-3969B-13GV80-FAC-C                                12                33                  27              FAC754-FAC868 ATRIUM 10XM XMLC-3948B-13GV70-FAC-B                                12                48                  28              FAC869-FAC980 ATRIUM 10XM XMLC-3967B-15GV80-FAD-B                                13                164                  29              FAD001-FAD164 ATRIUM 10XM XMLC-3945B-14GV80-FAD-C                                13                72                  30              FAD165-FAD236 ATRIUM 10XM XMLC-3951B-14GV80-FAD-C                                13                40                  31              FAD237-FAD276 ATRIUM 10XM XMLC-4091B-13GV80-FAD-B                                13                56                  32              FAD277-FAD332 ATRIUM 10XM XMLC-3943B-15GV80-FAE                                  14                96                  33              FAE333-FAE428 ATRIUM 10XM XMLC-3944B-14GV80-FAE                                  14                140                  34              FAE429-FAE568 ATRIUM 10XM XMLC-4001B-12GV80-FAE                                  14                80                  35              FAE569-FAE648 1.4      Acceptability Limits discussed in this document were generated based on NRC approved methodologies per References 6 through 26.
* The table identifies the expected fuel type breakdown in anticipation of final core loading. The final composition of the core depends upon uncertainties during the outage such as discovering a failed fuel bundle, or other bundle damage. Minor core loading changes, due to unforeseen events, will conform to the safety and monitoring requirements identified in this document.
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2 APLHGR Limits (Technical Specifications 3.2.1 & 3.7.5)
The APLHGR limit is determined by adjusting the rated power APLHGR limit for off-rated power, off-rated flow, and SLO conditions. The most limiting of these is then used as follows:
APLHGR limit = MIN ( APLHGRP , APLHGRF, APLHGRSLO )
where:
APLHGRP          off-rated power APLHGR limit          [APLHGRRATED
* MAPFACP]
APLHGRF          off-rated flow APLHGR limit          [APLHGRRATED
* MAPFACF]
APLHGRSLO        SLO APLHGR limit                      [APLHGRRATED
* SLO Multiplier]
2.1    Rated Power and Flow Limit: APLHGRRATED The rated conditions APLHGR for all fuel are identified per Reference 1. The rated conditions APLHGR for ATRIUM-10XM are shown in Figure 2.1.
2.2    Off-Rated Power Dependent Limit: APLHGRP Reference 1 does not specify a power dependent APLHGR. Therefore, MAPFACP is set to a value of 1.0.
2.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. No additional power dependent limitation is required.
2.3    Off-Rated Flow Dependent Limit: APLHGRF Reference 1 does not specify a flow dependent APLHGR. Therefore, MAPFACF is set to a value of 1.0.
2.4    Single Loop Operation Limit: APLHGRSLO The single loop operation multiplier for ATRIUM-10XM fuel is 0.85, per Reference 1.
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15 12 APLHGR (kW/ft) 9 6
3 0
0          20                40              60            80 Planar Average Exposure (GWd/MTU)
Planar Avg. APLHGR Exposure      Limit (GWd/MTU)          (kW/ft) 0.0            13.0 15.0            13.0 67.0              7.6 Figure 2.1 APLHGRRATED for ATRIUM-10XM Fuel BFN - Unit 1                                    Page 13 of 44                  TRM Revision 147 September 28, 2020
 
2.5      Equipment Out-Of-Service Corrections The limits shown in Figure 2.1 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
In-Service                              All equipment In-Service
* RPTOOS                                  EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                  Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                  Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                        Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                  One Recirculation Pump Out-Of-Service
* All equipment service conditions assume 1 SRVOOS.
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3 LHGR Limits (Technical Specification 3.2.3, 3.3.4.1, & 3.7.5)
The LHGR limit is determined by adjusting the rated power LHGR limit for off-rated power and off-rated flow conditions. The most limiting of these is then used as follows:
LHGR limit = MIN ( LHGRP, LHGRF )
where:
LHGRP            off-rated power LHGR limit            [LHGRRATED
* LHGRFACP]
LHGRF            off-rated flow LHGR limit            [LHGRRATED
* LHGRFACF]
3.1    Rated Power and Flow Limit: LHGRRATED The rated conditions LHGR for all fuel are identified per Reference 1. The rated conditions LHGR for ATRIUM-10XM fuel is shown in Figure 3.1. The LHGR limit is consistent with Reference 3.
3.2    Off-Rated Power Dependent Limit: LHGRP LHGR limits are adjusted for off-rated power conditions using the LHGRFACP multiplier provided in Reference 1. The multiplier is split into two sub cases: turbine bypass valves in and out-of-service. The base case multipliers are shown in Figure 3.2.
3.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional limits are shown in Figure 3.4 and Figure 3.5, based on temperature conditions identified in Table 3.1.
Table 3.1 Startup Feedwater Temperature Basis Temperature Power          Range 1            Range 2
(% Rated)          (&deg;F)              (&deg;F) 23              155.0            150.0 30              162.0            157.0 40              172.0            167.0 50              182.0            177.0 BFN - Unit 1                              Page 15 of 44                      TRM Revision 147 September 28, 2020
 
3.3      Off-Rated Flow Dependent Limit: LHGRF LHGR limits are adjusted for off-rated flow conditions using the LHGRFACF multiplier provided in Reference 1. Multipliers are shown in Figure 3.3.
3.4      Equipment Out-Of-Service Corrections The limits shown in Figure 3.1 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
* In-Service                            All equipment In-Service RPTOOS                                EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                      Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                One Recirculation Pump Out-Of-Service Off-rated power corrections shown in Figure 3.2 are dependent on operation of the Turbine Bypass Valve system. For this reason, separate limits are to be applied for TBVIS or TBVOOS operation. The limits have no dependency on RPTOOS, PLUOOS, FHOOS/FFWTR, or SLO.
Off-rated flow corrections shown in Figure 3.3 are bounding for all EOOS conditions.
Off-rated power corrections shown in Figure 3.4 and Figure 3.5 are also dependent on operation of the Turbine Bypass Valve system. In this case, limits support FHOOS operation during startup. These limits have no dependency on RPTOOS, PLUOOS, or SLO.
* All equipment service conditions assume 1 SRVOOS.
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15 12 9
LHGR (kW/ft) 6 3
0 0        20              40              60            80 Pellet Exposure (GWd/MTU)
Pellet        LHGR Exposure        Limit (GWd/MTU)        (kW/ft) 0.0            14.1 18.9            14.1 74.4            7.4 Figure 3.1 LHGRRATED for ATRIUM-10XM Fuel BFN - Unit 1                                Page 17 of 44                TRM Revision 147 September 28, 2020
 
1.10 1.00 0.90 Turbine Bypass Valve In-Service, TBVIS 0.80 Turbine Bypass Valve Out-of-Service, TBVOOS LHGRFACP 0.70 0.60 0.50          7%9,6&RUH)ORZ 7%9226&RUH)ORZ TBVIS, > 50% Core Flow 0.40 TBVOOS, > 50% Core Flow 0.30 0.20 20      30          40          50        60          70        80        90        100    110 Core Power (% Rated)
Turbine Bypass In-Service                  Turbine Bypass Out-of-Service Core                                          Core Power          LHGRFACP                      Power            LHGRFACP
(% Rated)                                      (% Rated) 100.0          1.00                          100.0          0.99 26.0          0.61                            26.0          0.59 Core Flow > 50% Rated                          Core Flow > 50% Rated 26.0          0.43                            26.0          0.37 23.0          0.41                            23.0          0.34
                          &RUH)ORZ5DWHG                          &RUH)ORZ5DWHG 26.0          0.48                            26.0          0.48 23.0          0.46                            23.0          0.43 Figure 3.2 Base Operation LHGRFACP for ATRIUM-10XM Fuel (Independent of other EOOS conditions)
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1.10 1.05 1.00 0.95 0.90 LHGRFACF 0.85 0.80 0.75 0.70 0.65 0.60 0.55 20    30        40      50        60      70      80        90      100      110 Core Flow (% Rated)
Core Flow            LHGRFACF
(% Rated) 0.0            0.61 30.0            0.61 78.1            1.00 107.0            1.00 Figure 3.3 LHGRFACF for ATRIUM-10XM Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
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1.10 1.00 0.90 0.80 Turbine Bypass Valve In-Service, TBVIS 0.70                                                    Turbine Bypass Valve Out-of-Service, TBVOOS LHGRFACP  0.60 0.50 7%9,6&RUH)ORZ 7%9226&RUH)ORZ 0.40 TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20        30          40          50        60          70        80        90        100    110 Core Power (% Rated)
Turbine Bypass In-Service                  Turbine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                    Power            LHGRFACP
(% Rated)                                    (% Rated) 100.0          1.00                        100.0          0.99 26.0          0.51                          26.0          0.50 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0          0.39                          26.0          0.34 23.0          0.35                          23.0          0.30
                            &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0          0.43                          26.0          0.43 23.0          0.40                          23.0          0.38 Figure 3.4 Startup Operation LHGRFACP for ATRIUM-10XM Fuel:
Table 3.1 Temperature Range 1 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
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1.10 1.00 0.90 0.80              Turbine Bypass Valve In-Service, TBVIS Turbine Bypass Valve Out-of-Service, TBVOOS LHGRFACP 0.70 0.60 0.50 7%9,6&RUH)ORZ 0.40 1//
7%9226&RUH)ORZ TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20          30        40          50        60        70          80        90        100      110 Core Power (% Rated)
Turbine Bypass In-Service                Turbine Bypass Out-of-Service Core                                      Core Power          LHGRFACP                  Power            LHGRFACP
(% Rated)                                  (% Rated) 100.0          1.00                      100.0          0.99 26.0          0.50                        26.0          0.49 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0          0.39                        26.0          0.34 23.0          0.35                        23.0          0.30
                              &RUH)ORZ5DWHG                      &RUH)ORZ5DWHG 26.0          0.42                        26.0          0.42 23.0          0.40                        23.0          0.38 Figure 3.5 Startup Operation LHGRFACP for ATRIUM-10XM Fuel:
Table 3.1 Temperature Range 2 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
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4 OLMCPR Limits (Technical Specification 3.2.2, 3.3.4.1, & 3.7.5)
OLMCPR is calculated to be the most limiting of the flow or power dependent values OLMCPR limit = MAX ( MCPRF , MCPRP )
where:
MCPRF        core flow-dependent MCPR limit MCPRP        power-dependent MCPR limit 4.1    Flow Dependent MCPR Limit: MCPRF MCPRF limits are dependent upon core flow (% of Rated), and the max core flow limit, (Rated or Increased Core Flow, ICF). MCPRF limits are shown in Figure 4.1, per Reference 1. Limits are valid for all EOOS combinations. No adjustment is required for SLO conditions.
4.2    Power Dependent MCPR Limit: MCPRP MCPRP limits are dependent upon:
Core Power Level (% of Rated)
Technical Specification Scram Speed (TSSS), Nominal Scram Speed (NSS), or Optimum Scram Speed (OSS)
Cycle Operating Exposure (NEOC, EOC, and CD - as defined in this section)
Equipment Out-Of-Service Options Two or Single recirculation Loop Operation (TLO vs. SLO)
The MCPRP limits are provided in Table 4.2 through Table 4.9, where each table contains the limits for all fuel types and EOOS options (for a specified scram speed and exposure range).
The CMSS determines MCPRP limits, from these tables, based on linear interpolation between the specified powers.
4.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional power dependent limits are shown in Table 4.5 through Table 4.8 based on temperature conditions identified in Table 3.1.
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4.2.2    Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS)
MCPRP limits are provided for three different sets of assumed scram speeds. The Technical Specification Scram Speed (TSSS) MCPRP limits are applicable at all times, as long as the scram time surveillance demonstrates the times in Technical Specification Table 3.1.4-1 are met. Both Nominal Scram Speeds (NSS) and/or Optimum Scram Speeds (OSS) may be used, as long as the scram time surveillance demonstrates Table 4.1 times are applicable.
* Table 4.1 Nominal Scram Time Basis Notch                    Nominal                Optimum Position              Scram Timing            Scram Timing (index)                  (seconds)                (seconds) 46                        0.420                  0.380 36                        0.980                  0.875 26                        1.600                  1.465 6                        2.900                  2.900 In demonstrating compliance with the NSS and/or OSS scram time basis, surveillance requirements from Technical Specification 3.1.4 apply; accepting the definition of SLOW rods should conform to scram speeds shown in Table 4.1. If conformance is not demonstrated, TSSS based MCPRP limits are applied.
On initial cycle startup, TSSS limits are used until the successful completion of scram timing confirms NSS and/or OSS based limits are applicable.
4.2.3    Exposure Dependent Limits Exposures are tracked on a Core Average Exposure basis (CAVEX, not Cycle Exposure).
Higher exposure MCPRP limits are always more limiting and may be used for any Core Average Exposure up to the ending exposure. Per Reference 1, MCPRP limits are provided for the following exposure ranges:
BOC to NEOC                                NEOC corresponds to                        30,758.8 MWd / MTU BOC to EOCLB                                EOCLB corresponds to                        34,078.5 MWd / MTU BOC to End of Coast                        End of Coast                                35,767.8 MWd / MTU NEOC refers to a Near EOC exposure point.
* Reference 1 analysis results are based on information identified in Reference 4.
Drop out times consistent with method used to perform actual timing measurements (i.e., including pickup/dropout effects).
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The EOCLB exposure point is not the true End-Of-Cycle exposure. Instead it corresponds to a licensing exposure window exceeding expected end-of-full-power-life.
The End of Coast exposure point represents a licensing exposure point exceeding the expected end-of-cycle exposure including cycle extension options.
4.2.4      Equipment Out-Of-Service (EOOS) Options EOOS options
* covered by MCPRP limits are given by the following:
In-Service                                          All equipment In-Service RPTOOS                                              EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                              Turbine Bypass Valve(s) Out-Of-Service RPTOOS+TBVOOS                                      Combined RPTOOS and TBVOOS PLUOOS                                              Power Load Unbalance Out-Of-Service PLUOOS+RPTOOS                                      Combined PLUOOS and RPTOOS PLUOOS+TBVOOS                                      Combined PLUOOS and TBVOOS PLUOOS+TBVOOS+RPTOOS                                Combined PLUOOS, RPTOOS, and TBVOOS FHOOS (or FFWTR)                                    Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)
RCPOOS                                              One Recirculation Pump Out-Of-Service For exposure ranges up to NEOC and EOCLB, additional combinations of MCPRP limits are also provided including FHOOS. The coast down exposure range assumes application of FFWTR. FHOOS based MCPRP limits for the coast down exposure are redundant because the temperature setdown assumption is identical with FFWTR.
4.2.5      Single-Loop-Operation (SLO) Limits When operating in RCPOOS conditions, MCPRp limits are constructed differently from the normal operating RCP conditions. The limiting event for RCPOOS is a pump seizure scenario, which sets the upper bound for allowed core power and flow . This event is not impacted by scram time assumptions. Specific MCPRP limits are shown in Table 4.9.
4.2.6      Below Pbypass Limits Below Pbypass (26% rated power), MCPRP limits depend upon core flow. One set of MCPRP limits applies for core flow above 50% of rated; a second set applies if the core flow is less than or equal to 50% rated.
* All equipment service conditions assume 1 SRVOOS.
RCPOOS limits are only valid up to 43.75% rated core power, 50% rated core flow, and an active recirculation drive flow of 17.73 Mlbm/hr.
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2.00 1.80 1.60 MCPRF 1.40 1.20 1.00 30    40        50        60        70          80      90        100        110 Core Flow (% Rated)
Core Flow          MCPRF
(% Rated) 30.0            1.58 84.0            1.34 107.0            1.34 Figure 4.1 MCPRF for All Fuel Types (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
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Table 4.2 MCPRP Limits for All Fuel Types: Optimum Scram Time Basis
* ATRIUM-10XM BOC            BOC            BOC Power                to              to        to End of Operating
(% of rated)          NEOC          EOCLB            Coast Condition 100                1.40            1.41            1.42 90              1.44            1.44            1.47 77.6              1.49            1.50            1.52 65              1.55            1.55            1.59
                                                        >50              1.65            1.65            1.72 Base Case                  50              1.81            1.81            1.81 40              1.89            1.89            1.92 26              2.36            2.36            2.49 26 at > 50%F            2.66            2.66            2.78 23 at > 50%F            2.84            2.84            2.97
                                                  DW)            2.53            2.53            2.64
                                                  DW)            2.70            2.70            2.83 100                1.42            1.42            ---
90              1.46            1.47            ---
77.6              1.52            1.52            ---
65              1.58            1.59            ---
                                                        >50              1.72            1.72            ---
50              1.81            1.81            ---
FHOOS 40              1.92            1.92            ---
26              2.49            2.49            ---
26 at > 50%F            2.78            2.78            ---
23 at > 50%F            2.97            2.97            ---
                                                  DW)            2.64            2.64            ---
                                                  DW)            2.83            2.83            ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR/FHOOS is supported for the BOC to End of Coast limits.
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Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM BOC          BOC        BOC Power            to            to      to End of Operating
(% of rated)      NEOC        EOCLB        Coast Condition 100            1.41        1.42        1.44 90            1.45        1.45        1.48 77.6            1.50        1.50        1.53 65            1.56        1.56        1.61
                                                            >50            1.68        1.68        1.74 50            1.81        1.81        1.81 Base Case              40            1.89        1.89        1.95 26            2.39        2.39        2.52 26 at > 50%F        2.66        2.66        2.78 23 at > 50%F        2.84        2.84        2.97
                                                      DW)        2.53        2.53        2.64
                                                      DW)        2.70        2.70        2.83 100            1.45        1.46        1.48 90            1.48        1.49        1.52 77.6            1.53        1.54        1.57 65            1.59        1.60        1.64
                                                            >50            1.68        1.68        1.76 50            1.81        1.81        1.82 TBVOOS                  40            1.89        1.89        1.96 26            2.39        2.39        2.53 26 at > 50%F        3.21        3.21        3.33 23 at > 50%F        3.46        3.46        3.62
                                                      DW)        2.88        2.88        3.04
                                                      DW)        3.18        3.18        3.35 100            1.43        1.44          ---
90            1.47        1.48          ---
77.6            1.52        1.53          ---
65            1.61        1.61          ---
                                                            >50            1.74        1.74          ---
50            1.81        1.81          ---
FHOOS                  40            1.95        1.95          ---
26            2.52        2.52          ---
26 at > 50%F        2.78        2.78          ---
23 at > 50%F        2.97        2.97          ---
                                                      DW)        2.64        2.64          ---
                                                      DW)        2.83        2.83          ---
100            1.41        1.42        1.44 90            1.45        1.45        1.48 77.6            1.50        1.50        1.53 65            1.74        1.74        1.74
                                                            >50            ---          ---        ---
50            1.82        1.82        1.82 PLUOOS                  40            1.89        1.89        1.95 26            2.39        2.39        2.52 26 at > 50%F        2.66        2.66        2.78 23 at > 50%F        2.84        2.84        2.97
                                                      DW)        2.53        2.53        2.64
                                                      DW)        2.70        2.70        2.83
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
BFN - Unit 1                                                Page 27 of 44                                      TRM Revision 147 September 28, 2020
 
Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis (continued)
* ATRIUM-10XM BOC          BOC          BOC Power              to            to        to End of Operating
(% of rated)      NEOC          EOCLB          Coast Condition 100            1.47          1.48            ---
90            1.51          1.52            ---
77.6            1.57          1.57            ---
65            1.64          1.64            ---
                                                          >50            1.76          1.76            ---
TBVOOS                    50            1.82          1.82            ---
FHOOS                      40            1.96          1.96            ---
26            2.53          2.53            ---
26 at > 50%F        3.33          3.33            ---
23 at > 50%F        3.62          3.62            ---
                                                    DW)        3.04          3.04            ---
                                                    DW)        3.35          3.35            ---
100            1.45          1.46          1.48 90            1.48          1.49          1.52 77.6            1.53          1.54          1.57 65            1.74          1.74          1.75
                                                          >50              ---            ---          ---
TBVOOS                    50            1.82          1.82          1.82 PLUOOS                    40            1.89          1.89          1.96 26            2.39          2.39          2.53 26 at > 50%F        3.21          3.21          3.33 23 at > 50%F        3.46          3.46          3.62
                                                    DW)        2.88          2.88          3.04
                                                    DW)        3.18          3.18          3.35 100            1.43          1.44            ---
90            1.47          1.48            ---
77.6            1.52          1.53            ---
65            1.74          1.74            ---
                                                          >50              ---            ---          ---
FHOOS                    50            1.82          1.82            ---
PLUOOS                    40            1.95          1.95            ---
26            2.52          2.52            ---
26 at > 50%F        2.78          2.78            ---
23 at > 50%F        2.97          2.97            ---
                                                    DW)        2.64          2.64            ---
                                                    DW)        2.83          2.83            ---
100            1.47          1.48            ---
90            1.51          1.52            ---
77.6            1.57          1.57            ---
65            1.75          1.75            ---
                                                          >50              ---            ---          ---
TBVOOS 50            1.82          1.82            ---
FHOOS 40            1.96          1.96            ---
PLUOOS 26            2.53          2.53            ---
26 at > 50%F        3.33          3.33            ---
23 at > 50%F        3.62          3.62            ---
                                                    DW)        3.04          3.04            ---
                                                    DW)        3.35          3.35            ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
BFN - Unit 1                                                Page 28 of 44                                      TRM Revision 147 September 28, 2020
 
Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM BOC          BOC          BOC Power              to            to        to End of Operating
(% of rated)      NEOC          EOCLB          Coast Condition 100            1.42          1.43          1.46 90            1.46          1.47          1.50 77.6            1.51          1.51          1.57 65            1.59          1.59          1.65
                                                          >50            1.70          1.70          1.78 50            1.82          1.82          1.83 Base Case 40            1.90          1.90          1.98 26            2.41          2.41          2.55 26 at > 50%F        2.66          2.66          2.79 23 at > 50%F        2.84          2.84          2.98
                                                    DW)        2.53          2.53          2.65
                                                    DW)        2.70          2.70          2.84 100            1.46          1.47          1.50 90            1.49          1.50          1.54 77.6            1.54          1.55          1.60 65            1.61          1.61          1.68
                                                          >50            1.71          1.71          1.80 50            1.82          1.82          1.84 TBVOOS 40            1.90          1.90          2.00 26            2.41          2.41          2.56 26 at > 50%F        3.21          3.21          3.34 23 at > 50%F        3.46          3.46          3.63
                                                    DW)        2.88          2.88          3.05
                                                    DW)        3.18          3.18          3.36 100            1.46          1.46            ---
90            1.50          1.50            ---
77.6            1.57          1.57            ---
65            1.65          1.65            ---
                                                          >50            1.78          1.78            ---
50            1.83          1.83            ---
FHOOS 40            1.98          1.98            ---
26            2.55          2.55            ---
26 at > 50%F        2.79          2.79            ---
23 at > 50%F        2.98          2.98            ---
                                                    DW)        2.65          2.65            ---
                                                    DW)        2.84          2.84            ---
100            1.42          1.43          1.46 90            1.46          1.47          1.50 77.6            1.51          1.51          1.57 65            1.74          1.74          1.75
                                                          >50              ---            ---          ---
50            1.83          1.83          1.83 PLUOOS 40            1.90          1.90          1.90 26            2.41          2.41          2.55 26 at > 50%F        2.66          2.66          2.79 23 at > 50%F        2.84          2.84          2.98
                                                    DW)        2.53          2.53          2.65
                                                    DW)        2.70          2.70          2.84
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
BFN - Unit 1                                                Page 29 of 44                                      TRM Revision 147 September 28, 2020
 
Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis (continued)
* ATRIUM-10XM BOC          BOC          BOC Power              to            to      to End of Operating
(% of rated)      NEOC        EOCLB          Coast Condition 100            1.50          1.50          ---
90            1.53          1.54          ---
77.6            1.60          1.60          ---
65            1.68          1.68          ---
                                                          >50            1.80          1.80          ---
TBVOOS                  50            1.84          1.84          ---
FHOOS                    40            2.00          2.00          ---
26            2.56          2.56          ---
26 at > 50%F        3.34          3.34          ---
23 at > 50%F        3.63          3.63          ---
                                                      DW)        3.05          3.05          ---
                                                      DW)        3.36          3.36          ---
100            1.46          1.47          1.50 90            1.49          1.50          1.54 77.6            1.54          1.55          1.60 65            1.74          1.74          1.76
                                                          >50              ---          ---          ---
TBVOOS                  50            1.83          1.83          1.84 PLUOOS                    40            1.90          1.90          2.00 26            2.41          2.41          2.56 26 at > 50%F        3.21          3.21          3.34 23 at > 50%F        3.46          3.46          3.63
                                                      DW)        2.88          2.88          3.05
                                                      DW)        3.18          3.18          3.36 100            1.46          1.46          ---
90            1.50          1.50          ---
77.6            1.57          1.57          ---
65            1.75          1.75          ---
                                                          >50              ---          ---          ---
FHOOS                    50            1.83          1.83          ---
PLUOOS                    40            1.98          1.98          ---
26            2.55          2.55          ---
26 at > 50%F        2.79          2.79          ---
23 at > 50%F        2.98          2.98          ---
                                                      DW)        2.65          2.65          ---
                                                      DW)        2.84          2.84          ---
100            1.50          1.50          ---
90            1.53          1.54          ---
77.6            1.60          1.60          ---
65            1.76          1.76          ---
                                                          >50              ---          ---          ---
TBVOOS 50            1.84          1.84          ---
FHOOS 40            2.00          2.00          ---
PLUOOS 26            2.56          2.56          ---
26 at > 50%F        3.34          3.34          ---
23 at > 50%F        3.63          3.63          ---
                                                      DW)        3.05          3.05          ---
                                                      DW)        3.36          3.36          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
BFN - Unit 1                                                Page 30 of 44                                      TRM Revision 147 September 28, 2020
 
Table 4.5 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM BOC          BOC          BOC Power            to          to        to End of Operating
(% of rated)      NEOC          EOCLB        Coast Condition 100            1.43          1.44          1.44 90            1.47          1.48          1.48 77.6          1.52          1.53          1.53 65            1.74          1.74          1.74
                                                        >50              ---          ---          ---
50            1.89          1.89          1.89 TBVIS 40            2.14          2.14          2.14 26            2.82          2.82          2.82 26 at > 50%F        3.06          3.06          3.06 23 at > 50%F        3.31          3.31          3.31
                                                    DW)        2.91          2.91          2.91
                                                    DW)        3.14          3.14          3.14 100            1.47          1.48          1.48 90            1.51          1.52          1.52 77.6          1.57          1.57          1.57 65            1.75          1.75          1.75
                                                        >50              ---          ---          ---
50            1.90          1.90          1.90 TBVOOS 40            2.15          2.15          2.15 26            2.83          2.83          2.83 26 at > 50%F        3.57          3.57          3.57 23 at > 50%F        3.87          3.87          3.87
                                                    DW)        3.27          3.27          3.27
                                                    DW)        3.61          3.61          3.61
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
BFN - Unit 1                                                Page 31 of 44                                TRM Revision 147 September 28, 2020
 
Table 4.6 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM BOC          BOC          BOC Power            to          to        to End of Operating
(% of rated)      NEOC          EOCLB        Coast Condition 100            1.43          1.44          1.44 90            1.47          1.48          1.48 77.6          1.52          1.53          1.53 65            1.74          1.74          1.74
                                                        >50              ---          ---          ---
50            1.90          1.90          1.90 TBVIS 40            2.16          2.16          2.16 26            2.85          2.85          2.85 26 at > 50%F        3.08          3.08          3.08 23 at > 50%F        3.32          3.32          3.32
                                                    DW)        2.92          2.92          2.92
                                                    DW)        3.17          3.17          3.17 100            1.47          1.48          1.48 90            1.51          1.52          1.52 77.6          1.57          1.57          1.57 65            1.75          1.75          1.75
                                                        >50              ---          ---          ---
50            1.91          1.91          1.91 TBVOOS 40            2.17          2.17          2.17 26            2.86          2.86          2.86 26 at > 50%F        3.58          3.58          3.58 23 at > 50%F        3.89          3.89          3.89
                                                    DW)        3.28          3.28          3.28
                                                    DW)        3.63          3.63          3.63
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
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Table 4.7 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM BOC          BOC          BOC Power            to          to        to End of Operating
(% of rated)      NEOC          EOCLB        Coast Condition 100            1.46          1.46          1.46 90            1.50          1.50          1.50 77.6          1.57          1.57          1.57 65            1.75          1.75          1.75
                                                        >50              ---          ---          ---
50            1.93          1.93          1.93 TBVIS 40            2.18          2.18          2.18 26            2.86          2.86          2.86 26 at > 50%F        3.07          3.07          3.07 23 at > 50%F        3.32          3.32          3.32
                                                    DW)        2.92          2.92          2.92
                                                    DW)        3.15          3.15          3.15 100            1.50          1.50          1.50 90            1.53          1.54          1.54 77.6          1.60          1.60          1.60 65            1.76          1.76          1.76
                                                        >50              ---          ---          ---
50            1.94          1.94          1.94 TBVOOS 40            2.19          2.19          2.19 26            2.87          2.87          2.87 26 at > 50%F        3.58          3.58          3.58 23 at > 50%F        3.88          3.88          3.88
                                                    DW)        3.28          3.28          3.28
                                                    DW)        3.62          3.62          3.62
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
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Table 4.8 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM BOC          BOC          BOC Power            to          to        to End of Operating
(% of rated)      NEOC          EOCLB        Coast Condition 100            1.46          1.46          1.46 90            1.50          1.50          1.50 77.6          1.57          1.57          1.57 65            1.75          1.75          1.75
                                                        >50              ---          ---          ---
50            1.94          1.94          1.94 TBVIS 40            2.20          2.20          2.20 26            2.89          2.89          2.89 26 at > 50%F        3.09          3.09          3.09 23 at > 50%F        3.33          3.33          3.33
                                                    DW)        2.93          2.93          2.93
                                                    DW)        3.18          3.18          3.18 100            1.50          1.50          1.50 90            1.53          1.54          1.54 77.6          1.60          1.60          1.60 65            1.76          1.76          1.76
                                                        >50              ---          ---          ---
50            1.95          1.95          1.95 TBVOOS 40            2.21          2.21          2.21 26            2.90          2.90          2.90 26 at > 50%F        3.59          3.59          3.59 23 at > 50%F        3.90          3.90          3.90
                                                    DW)        3.29          3.29          3.29
                                                    DW)        3.64          3.64          3.64
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
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Table 4.9 MCPRP Limits for All Fuel Types: Single Loop Operation for All Scram Times
* Power              BOC to End of COAST Operating Condition        (% of rated)              ATRIUM-10XM 100                  2.09 43.75                  2.09 40                  2.09 RCPOOS                            26                  2.57 FHOOS                26 at > 50%F                    2.81 23 at > 50%F                    3.00
                                                          DW)                    2.67
                                                          DW)                    2.86 100                  2.09 43.75                  2.09 RCPOOS                            40                  2.09 TBVOOS                            26                  2.58 PLUOOS                26 at > 50%F                    3.36 FHOOS                23 at > 50%F                    3.65
                                                          DW)                    3.07
                                                          DW)                    3.38 100                  2.12 43.75                  2.12 40                  2.21 RCPOOS 26                  2.89 TBVOOS 26 at > 50%F                    3.60 FHOOS1 23 at > 50%F                    3.90
                                                          DW)                    3.30
                                                          DW)                    3.64 100                  2.14 43.75                  2.14 40                  2.23 RCPOOS 26                  2.92 TBVOOS 26 at > 50%F                    3.61 FHOOS2 23 at > 50%F                    3.92
                                                          DW)                    3.31
                                                          DW)                    3.66
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop.
RCPOOS limits are only valid up to 50% rated core power, 50% rated core flow, and an active recirculation drive flow of 17.73 Mlbm/hr.
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5      Thermal-Hydraulic Stability Protection (Technical Specification 3.3.1.1)
Technical Specification Table 3.3.1.1-1, Function 2f, identifies the function.
Instrument setpoints are established, such that the reactor will be tripped before an oscillation can grow to the point where the SLMCPR is exceeded. With application of Reference 30, the DSS-CD stability solution will be used per Reference 26. The DSS-CD SAD setpoint is 1.10 for TLO and SLO.
New analyses have been developed based on Reference 26. With the implementation of the MELLLA+ operating domain expansion, an ABSP trip is required when the OPRM is out-of-service. The ABSP trip settings define a region of the power to flow map within which an automatic reactor scram occurs. The ABSP trip settings are provided in Table 5.1. If both the OPRM and ABSP are out-of-service, operation within the MELLLA+ domain is not allowed and the MBSP Regions provide stability protection. Table 5.2 and Table 5.3 provide the endpoints for the MBSP regions for nominal and reduced feedwater temperature conditions.
Table 5.1 ABSP Setpoints for the Scram Region Parameter                Symbol            Setting Value (unit)                            Comments Slope of ABSP APRM low Flow Biased Slope for Trip              mTRIP          2.00 (% RTP/% RDF)
Trip Linear Segment ABSP APRM Flow Biased Trip Constant Power                                                                  Setpoint Power Intercept. Constant PBSP-TRIP              35.0 (% RTP)
Line for Trip                                                                  Power Line for Trip from Zero Drive Flow to Flow Breakpoint Value ABSP APRM Flow Biased Trip Constant Flow Line                                                                Setpoint Drive Flow Intercept.
WBSP-TRIP              49 (% RDF) for Trip                                                                    Constant Flow Line for Trip          (see Note 1 below)
Flow Breakpoint            WBSP-BREAK              30.0 (% RDF)                  Flow Breakpoint Value Note 1: WBSP-TRIP can be set to 49.0 % RDF or any higher value up to the intersection of the ABSP sloped line with the APRM Flow Biased STP scram line.
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Table 5.2 Analyzed MBSP Endpoints: Nominal Feedwater Temperature Power  Core Flow Endpoint                                    Definition
(% Rated) (% Rated)
Scram Region (Region I)
A1        75.9      52.7    Boundary Intercept on MELLLA+
Line Scram Region (Region I)
B1        35.5      29.0    Boundary Intercept on Natural Circulation Line (NCL)
Controlled Entry Region (Region A2        66.1      52.0    II) Boundary Intercept on MELLLA Line Controlled Entry Region (Region B2        25.5      29.0    II) Boundary Intercept on Natural Circulation Line (NCL)
Table 5.3 Analyzed MBSP Endpoints: Reduced Feedwater Temperature Power  Core Flow Endpoint                                    Definition
(% Rated) (% Rated)
Scram Region (Region I)
A1        64.9      50.5    Boundary Intercept on MELLLA Line Scram Region (Region I)
B1        29.4      29.0    Boundary Intercept on Natural Circulation Line (NCL)
Controlled Entry Region (Region A2        68.3      54.9    II) Boundary Intercept on MELLLA Line Controlled Entry Region (Region B2        24.5      29.0    II) Boundary Intercept on Natural Circulation Line (NCL)
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6 APRM Flow Biased Rod Block Trip Settings (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
The APRM rod block trip setting is based upon References 27 & 29, and is defined by the following:
for two loop operation:
SRB d (0.61Wd + 63.3)                    Allowable Value SRB d (0.61Wd + 62.0)                    Nominal Trip Setpoint (NTSP) where:
SRB    =      Rod Block setting in percent of rated thermal power (3952 MWt)
Wd      =      Recirculation drive flow rate in percent of rated (100% drive flow required to achieve 100% core power and flow) and for single loop operation:
SRB d (0.55(Wd-'W) + 60.5)                        Allowable Value SRB d (0.55(Wd-'W) + 58.5)                        Nominal Trip Setpoint (NTSP) where:
SRB    =      Rod Block setting in percent of rated thermal power (3952 MWt)
Wd      =      Recirculation drive flow rate in percent of rated (100% drive flow required to achieve 100% core power and flow)
        'W      =      Difference between two-loop and single-loop effective recirculation flow at the same core flow ('W=0.0 for two-loop operation)
The APRM rod block trip setting is clamped at a maximum allowable value of 115%
(corresponding to a NTSP of 113%).
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7 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
The RBM trip setpoints and applicable power ranges, based on References 27 & 28, are shown in Table 7.1. Setpoints are based on an HTSP, unfiltered analytical limit of 114%. Unfiltered setpoints are consistent with a nominal RBM filter setting of 0.0 seconds; filtered setpoints are consistent with a nominal RBM filter setting less than 0.5 seconds. Cycle specific CRWE analyses of OLMCPR are documented in Reference 1, superseding values reported in References 27, 28, and 29.
Table 7.1 Analytical RBM Trip Setpoints
* Allowable              Nominal Trip RBM                              Value                Setpoint Trip Setpoint                            (AV)                (NTSP)
LPSP                                        27%                    25%
IPSP                                        62%                    60%
HPSP                                        82%                    80%
LTSP - unfiltered                        121.7%                  120.0%
                                          - filtered                    120.7%                  119.0%
ITSP      - unfiltered                  116.7%                  115.0%
                                          - filtered                    115.7%                  114.0%
HTSP - unfiltered                        111.7%                  110.0%
                                          - filtered                    110.9%                  109.2%
DTSP                                        90%                    92%
As a result of cycle specific CRWE analyses, RBM setpoints in Technical Specification Table 3.3.2.1-1 are applicable as shown in Table 7.2. Cycle specific setpoint analysis results are shown in Table 7.3, per Reference 1.
Table 7.2 RBM Setpoint Applicability Thermal Power                      Applicable                Notes from
(% Rated)                          MCPR                Table 3.3.2.1-1              Comment
                                                        < 1.65              (a), (b), (f), (h)    two loop operation 27% and < 90%
                                                        < 1.68              (a), (b), (f), (h)    single loop operation 90%                                      < 1.36                      (g)            two loop operation
* Values are considered maximums. Using lower values, due to RBM system hardware/software limitations, is conservative, and acceptable.
MCPR values shown correspond with, (support), SLMPCR values identified in Reference 1.
Greater than 90% rated power is not attainable in single loop operation.
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Table 7.3 Control Rod Withdrawal Error Results RBM                    CRWE HTSP Analytical Limit          OLMCPR Unfiltered 107                    1.26 111                    1.30 114                    1.31 117                    1.33 Results, compared against the base case OLMCPR results of Table 4.2, indicate SLMCPR remains protected for RBM inoperable conditions (i.e., 114% unblocked).
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8 Shutdown Margin Limit (Technical Specification 3.1.1)
Assuming the strongest OPERABLE control blade is fully withdrawn, and all other OPERABLE control blades are fully inserted, the core shall be sub-critical and meet the following minimum shutdown margin:
SDM    0.38% dk/k BFN - Unit 1                              Page 41 of 44                        TRM Revision 147 September 28, 2020
 
Appendix A: MBSP Maps BFN - Unit 1    Page 42 of 44      TRM Revision 147 September 28, 2020
 
Core Power  (% Rated: 100% = 3952MWt) 110 100 MELLLA+ Region 90 80 BSP Boundary Manual 70                              Scram Region I 60 MELLLA Region      ICF Region MELLLA Upper Boundary 50 87.5% Rod Line 40                                                  Controlled Entry Region II 30                                                          Min. Flow Control 20 Natural                                  Min. Power Line Circulation 10                                    20% Pump Speed Line 0
0      10      20          30        40      50        60        70        80      90    100  110  120 Core Flow    (% Rated: 100% =102.5 MLbm /hr)
Figure A.1 MBSP Boundaries For Nominal Feedwater Temperature (Operation in the MELLLA+ Region Prohibited for Feedwater Temperature greater than 10 degrees F below the Nominal Feedwater Temperature)
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Core Power    (% Rated: 100% = 3952MWt) 110 100 MELLLA+ Region 90 80 BSP Boundary 70 Manual Scram 60                              Region I MELLLA Region MELLLA Upper Boundary                                              ICF Region 50 87.5% Rod Line Controlled Entry 40                                                          Region II 30                                                      Min. Flow Control 20 Natural                              Min. Power Line Circulation 10                                    20% Pump Speed Line 0
0      10      20          30        40      50      60        70        80      90    100  110    120 Core Flow  (% Rated: 100% =102.5 MLbm /hr)
Figure A.2 MBSP Boundaries For Reduced Feedwater Temperature (Operation in the MELLLA+ Region Prohibited for a Reduced Feedwater Temperature greater than 10 degrees F below the Nominal Feedwater Temperature)
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Reactor Protection System Instrumentation TR 3.3.1 TR 3.3      INSTRUMENTATION TR 3.3.1          Reactor Protection System (RPS) Instrumentation LCO 3.3.1                  There shall be two OPERABLE or tripped trip systems with a minimum of two OPERABLE instrument channels per trip system for the Turbine First Stage Pressure Permissive. The pressure switch allowable values shall be <116.7 psig.
APPLICABILITY:            ! 26% RTP (Turbine First Stage Pressure ! 116.7 psig)
-------------------------------------------------------NOTE-------------------------------------------------------
Required Actions shall be taken only if the permissive fails in such a manner to prevent the affected RPS logic from performing its intended function. Otherwise, no action is required.
ACTIONS CONDITION                              REQUIRED ACTION                        COMPLETION TIME A. One or more required                    ------------------NOTE-----------------
channels are inoperable.              Inoperable Turbine First Stage Pressure Permissive channel(s) or subsystem(s) may also affect Technical Specifications LCO 3.3.1.1 and LCO 3.3.4.1.
A.1        Trip the inoperable                1 hour channel(s) or entire trip system(s).
OR A.2.1      Initiate insertion of              Immediately OPERABLE rods.
AND A.2.2      Complete insertion of all 4 hours OPERABLE rods.
OR (continued)
BFN-UNIT 2                                            3.3-1                          TRM Revision 0, 12, 142 December 6, 2018
 
Reactor Protection System Instrumentation TR 3.3.1 ACTIONS CONDITION                              REQUIRED ACTION                        COMPLETION TIME A. (continued)                            A.3      Reduce power to less                4 hours than 26 percent of rated.
-------------------------------------------------------NOTE-------------------------------------------------------
A channel may be placed in an INOPERABLE status for up to 6 hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY TSR 3.3.1.1                Perform CHANNEL FUNCTIONAL TEST.                                  92 days TSR 3.3.1.2                Perform CHANNEL CALIBRATION.                                      24 months BFN-UNIT 2                                            3.3-2                          TRM Revision 8, 12, 142 December 6, 2018
 
Control Rod Block Instrumentation TR 3.3.4 Table 3.3.4-1 (page 1 of 3)
Control Rod Block Instrumentation APPLICABLE        REQUIRED      CONDITIONS        TECHNICAL        ALLOWABLE MODES OR          CHANNELS    REFERENCED        SURVEILLANCE          VALUE OTHER            PER TRIP        FROM          REQUIREMENTS SPECIFIED        FUNCTION      REQUIRED FUNCTION            CONDITIONS            (a)        ACTION A.1
: 1. Average Power Range Monitors
: a. APRM Upscale              1              3            B              TSR 3.3.4.1    (b)
(Flow Bias)                                                            TSR 3.3.4.2 TSR 3.3.4.7
: b. APRM Upscale              2              3            B              TSR 3.3.4.1    < 10%
(Startup) (c)                                                          TSR 3.3.4.2 TSR 3.3.4.7
: c. APRM                      1              3            B              TSR 3.3.4.1    > 3%
Downscale (d)                                                          TSR 3.3.4.2 TSR 3.3.4.7
: d. APRM                    1, 2              3            B              TSR 3.3.4.1    (e)
Inoperative                                                            TSR 3.3.4.2
: 2. Intermediate Range Monitors
: a. IRM Upscale (c)          2              6            B              TSR 3.3.4.1     108/125 of TSR 3.3.4.3    full scale TSR 3.3.4.6
: b. IRM Downscale            2              6            B              TSR 3.3.4.1    ! 5/125 of full (c) (f)                                                                TSR 3.3.4.3    scale TSR 3.3.4.6 (continued)
(a) During repair or calibration of equipment, not more than one SRM or APRM channel nor more than two IRM channels may be bypassed. Bypassed channels are not counted as OPERABLE channels to meet the minimum OPERABLE channel requirements.
(b) The APRM Rod Block Allowable Value shall be less than or equal to the limit specified in the CORE OPERATING LIMITS REPORT.
(c) This function is bypassed when the MODE switch is placed in the "Run" position.
(d) This function is only active when the MODE switch is in the "Run" position.
(e) The inoperative trips for the APRMs are produced by the following functions:
: 1. Local APRM chassis MODE switch not in operate.
: 2. Less than the required minimum number of LPRM inputs, both total and per axial level.
: 3. APRM module unplugged.
: 4. Self-test detected critical fault.
(f)  IRM downscale is bypassed when it is on its lowest range.
BFN-UNIT 2                                          3.3-35                        TRM Revision 0, 142 December 6, 2018
 
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EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table of Contents Total Number of Pages = 50 (including review cover sheet)
List of Tables ............................................................................................................................. iii List of Figures ............................................................................................................................ iv Revision Log ...............................................................................................................................v Nomenclature ............................................................................................................................ vi References .............................................................................................................................. viii 1      Introduction ........................................................................................................................1 1.1        Purpose .......................................................................................................................1 1.2        Scope ..........................................................................................................................1 1.3        Fuel Loading................................................................................................................1 1.4        Acceptability ................................................................................................................2 2      APLHGR Limits ..................................................................................................................3 2.1        Rated Power and Flow Limit: APLHGRRATED ...............................................................3 2.2        Off-Rated Power Dependent Limit: APLHGRP ............................................................3 2.2.1            Startup without Feedwater Heaters..................................................................... 3 2.3        Off-Rated Flow Dependent Limit: APLHGRF ...............................................................3 2.4        Single Loop Operation Limit: APLHGRSLO ...................................................................3 2.5        Equipment Out-Of-Service Corrections........................................................................6 3      LHGR Limits.......................................................................................................................7 3.1        Rated Power and Flow Limit: LHGRRATED....................................................................7 3.2        Off-Rated Power Dependent Limit: LHGRP .................................................................7 3.2.1            Startup without Feedwater Heaters..................................................................... 7 3.3        Off-Rated Flow Dependent Limit: LHGRF....................................................................8 3.4        Equipment Out-Of-Service Corrections........................................................................8 4      OLMCPR Limits................................................................................................................19 4.1        Flow Dependent MCPR Limit: MCPRF ......................................................................19 4.2        Power Dependent MCPR Limit: MCPRP ...................................................................19 4.2.1            Startup without Feedwater Heaters....................................................................19 4.2.2            Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS) ................................20 4.2.3            Exposure Dependent Limits ...............................................................................20 4.2.4            Equipment Out-Of-Service (EOOS) Options ......................................................21 4.2.5            Single-Loop-Operation (SLO) Limits ..................................................................21 4.2.6            Below Pbypass Limits ........................................................................................21 5      Thermal-Hydraulic Stability Protection..............................................................................33 6      APRM Flow Biased Rod Block Trip Settings.....................................................................35 7      Rod Block Monitor (RBM) Trip Setpoints and Operability .................................................36 8      Shutdown Margin Limit .....................................................................................................38 Appendix A:            MBSP Maps..................................................................................................... A-1 Browns Ferry Unit 2 Cycle 21                                                                                                          Page ii Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 List of Tables Nuclear Fuel Types ................................................................................................................ 2 Startup Feedwater Temperature Basis....................................................................................... 7 Nominal Scram Time Basis .......................................................................................................20 MCPRP Limits for All Fuel Types: Optimum Scram Time Basis .............................................23 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis ...............................................24 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis ........................26 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Nominal Scram Time Basis ...................................................................................................28 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Nominal Scram Time Basis ...................................................................................................29 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Technical Specification Scram Time Basis ............................................................................30 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Technical Specification Scram Time Basis ............................................................................31 MCPRP Limits for All Fuel Types: Single Loop Operation for All Scram Times ......................32 ABSP Setpoints for the Scram Region ......................................................................................33 Analyzed MBSP Endpoints: Nominal Feedwater Temperature.................................................34 Analyzed MBSP Endpoints: Reduced Feedwater Temperature ...............................................34 Analytical RBM Trip Setpoints ...............................................................................................36 RBM Setpoint Applicability ........................................................................................................36 Control Rod Withdrawal Error Results.......................................................................................37 Browns Ferry Unit 2 Cycle 21                                                                                                Page iii Core Operating Limits Report, (120% OLTP, MELLLA+)                                            TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 List of Figures APLHGRRATED for ATRIUM-10XM Fuel....................................................................................... 4 APLHGRRATED for ATRIUM-11 Fuel ............................................................................................ 5 LHGRRATED for ATRIUM-10XM Fuel ........................................................................................... 9 LHGRRATED for ATRIUM-11 Fuel................................................................................................10 Base Operation LHGRFACP for ATRIUM-10XM Fuel................................................................11 Base Operation LHGRFACP for ATRIUM-11 Fuel .....................................................................12 LHGRFACF for ATRIUM-10XM Fuel .........................................................................................13 LHGRFACF for ATRIUM-11 Fuel...............................................................................................14 Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 1 ........15 Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 2 ........16 Startup Operation LHGRFACP for ATRIUM-11 Fuel: Table 3.1 Temperature Range 1..............17 Startup Operation LHGRFACP for ATRIUM-11 Fuel: Table 3.1 Temperature Range 2..............18 MCPRF for All Fuel Types .........................................................................................................22 MBSP Boundaries For Nominal Feedwater Temperature........................................................ A-2 MBSP Boundaries For Reduced Feedwater Temperature ...................................................... A-3 Browns Ferry Unit 2 Cycle 21                                                                                                Page iv Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Revision Log Number            Page                                          Description 1-R1            vi-vii      Updated nomenclature pages for new items.
2-R1            viii      Updated References 1 and 4 Added New Reference (26) for the DSS-CD topical report supporting 3-R1            ix-x        MELLLA+. Other References incremented by +1 for renumbering.
Updated References 27-30. Removed Reference 31.
In Section 1.2, the Oscillation Power Range Monitor (OPRM) Setpoint 4-R1              1        is no longer applicable for MELLLA+ operation. This part has been more generically renamed Hydraulic Stability Protection 5-R1          11-18        Updated Figures 3.3-3.10 as needed for new information 6-R1            20        In Section 4.2.3, exposure dependent limits updated as needed.
7-R1          23-32        Updated Tables 4.2-4.9 as needed.
Re-write of Section 5 for DSS-CD stability change. Tables 5.1, 5.2, 8-R1          33-34 and 5.3 are new information supporting MELLLA+
9-R1            35        Updated to include two loop settings, clarify definitions 10-R1              38        Updated  symbol per CR 1596417 identification.
11-R1          A1-A3        New material supporting MELLLA+
0-R0            All        New document.
Browns Ferry Unit 2 Cycle 21                                                                                          Page v Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Nomenclature ABSP                  Automatic Backup Stability Protection APLHGR                Average Planar LHGR APRM                  Average Power Range Monitor AREVA NP              Vendor (Framatome, Siemens)
BOC                    Beginning of Cycle BSP                    Backup Stability Protection BWR                    Boiling Water Reactor CAVEX                  Core Average Exposure CD                    Coast Down CMSS                  Core Monitoring System Software COLR                  Core Operating Limits Report CPR                    Critical Power Ratio CRWE                  Control Rod Withdrawal Error CSDM                  Cold SDM DIVOM                  Delta CPR over Initial CPR vs. Oscillation Magnitude DSS-CD                Detect and Suppress Solution - Confirmation Density EOC                    End of Cycle EOCLB                  End-of-Cycle Licensing Basis EOOS                  Equipment OOS EPU                    Extended Power Uprate (120% OLTP)
FFTR                  Final Feedwater Temperature Reduction FFWTR                  Final Feedwater Temperature Reduction FHOOS                  Feedwater Heaters OOS ft                    Foot: English unit of measure for length GNF                    Vendor (General Electric, Global Nuclear Fuels)
GWd                    Giga Watt Day HTSP                  High TSP ICA                    Interim Corrective Action ICF                    Increased Core Flow (beyond rated)
IS                    In-Service kW                    kilo watt: SI unit of measure for power.
LCO                    License Condition of Operation LFWH                  Loss of Feedwater Heating LHGRFAC                LHGR Multiplier (Power or Flow dependent)
LPRM                  Low Power Range Monitor LRNB                  Generator Load Reject, No Bypass MAPFAC                MAPLHGR multiplier (Power or Flow dependent)
Browns Ferry Unit 2 Cycle 21                                                                                        Page vi Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
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[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 MBSP                  Manual Backup Stability Protection MCPR                  Minimum CPR MELLLA                Maximum Extended Load Line Limit Analysis MELLLA+                Maximum Extended Load Line Limit Analysis Plus MSRV                  Moisture Separator Reheater Valve MSRVOOS                MSRV OOS MTU                    Metric Ton Uranium MWd/MTU                Mega Watt Day per Metric Ton Uranium NEOC                  Near EOC NRC                    United States Nuclear Regulatory Commission NSS                    Nominal Scram Speed NTSP                  Nominal TSP OLMCPR                MCPR Operating Limit OLTP                  Original Licensed Thermal Power OOS                    Out-Of-Service OPRM                  Oscillation Power Range Monitor OSS                    Optimum Scram Speed PBDA                  Period Based Detection Algorithm Pbypass                Power, below which TSV Position and TCV Fast Closure Scrams are Bypassed PLU                    Power Load Unbalance PLUOOS                PLU OOS PRNM                  Power Range Neutron Monitor RBM                    Rod Block Monitor RCPOOS                Recirculation Pump OOS (SLO)
RDF                    Rated Drive Flow RPS                    Reactor Protection System RPT                    Recirculation Pump Trip RPTOOS                RPT OOS RTP                    Rated Thermal Power SDM                    Shutdown Margin SLMCPR                MCPR Safety Limit SLO                    Single Loop Operation TBV                    Turbine Bypass Valve TBVIS                  TBV IS TBVOOS                Turbine Bypass Valves OOS TIP                    Transversing In-core Probe TIPOOS                TIP OOS TLO                    Two Loop Operation TSP                    Trip Setpoint TSSS                  Technical Specification Scram Speed TVA                    Tennessee Valley Authority Browns Ferry Unit 2 Cycle 21                                                                                        Page vii Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 References
        ANP-3824, Revision , Browns Ferry Unit 2 Cycle 21 MELLLA+ Reload Analysis, Framatome, Inc., March 2020.
        ANP-3031P, Revision 3, Mechanical Design Report for Browns Ferry Units 1, 2, and 3 ATRIUM-10 Fuel Assemblies, AREVA NP, Inc., May 2014.
        ANP-3150P, Revision 4, Mechanical Design Report for Browns Ferry ATRIUM 10XM Fuel Assemblies, AREVA Inc., November 2017.
        ANP-3706P, Revision 1, Browns Ferry Unit 2 Cycle 21 Plant Parameters Document, Framatome Inc., June 2019.
        BFE-4410, Revision 0, Browns Ferry Unit 2 Cycle 21 In Core Shuffle, Calculation File, Tennessee Valley Authority, February 2019.
Methodology References
: 6.        XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
: 7.        XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
: 8.        EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
: 9.        ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
: 10.      XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.
: 11.      XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.
: 12.      EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.
: 13.      XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.
: 14.      XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.
: 15.      ANP-10307PA, Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP Inc., June 2011.
Browns Ferry Unit 2 Cycle 21                                                                                        Page viii Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
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[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020
: 16.      ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.
: 17.      ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Advanced Nuclear Fuels Corporation, September 2005.
: 18.      EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP Inc.,
September 2009.
: 19.      EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001, as supplemented by the site specific approval in NRC safety evaluation dated February 15, 2013 and July 31, 2014.
: 20.      EMF-2292(P)(A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
: 21.      EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
: 22.      BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP Inc., May 2008.
: 23.      BAW-10247PA, Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP Inc., April 2008.
: 24.      ANP-10298PA, Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc., March 2010.
: 25.      ANP-3140(P), Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc.,
August 2012.
: 26.      NEDC-33075P-A, Revision 8, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, GE Hitachi, November 2013.
Setpoint References
: 27.      EDQ2092900118, R35, Setpoint and Scaling Calculation for Neutron Monitoring &
Recirculation Flow Loops, Calculation File, Tennessee Valley Authority, August 9, 2019.
: 28.      Task T0500, Revision 0, Neutron Monitoring System w/RBM, Project Task Report, GE Hitachi Nuclear Energy, June 2017.
: 29.      Task T0506, Revision 0, TS Instrument Setpoints, Project Task Report, Tennessee Valley Authority, August, 2017.
: 30.      NEDC-33006P-A, Revision 3, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, GE Energy Nuclear, June 2009.
Browns Ferry Unit 2 Cycle 21                                                                                        Page ix Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
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[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 1 Introduction In anticipation of cycle startup, it is necessary to describe the expected limits of operation.
1.1    Purpose The primary purpose of this document is to satisfy requirements identified by unit technical specification section 5.6.5. This document may be provided, upon final approval, to the NRC.
1.2    Scope This document will discuss the following areas:
3/4 Average Planar Linear Heat Generation Rate (APLHGR) Limit (Technical Specifications 3.2.1 and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3.2.3, 3.3.4.1, and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Minimum Critical Power Ratio Operating Limit (OLMCPR)
(Technical Specifications 3.2.2, 3.3.4.1, 3.7.5 and Table 3.3.2.1-1)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Thermal-Hydraulic Stability Protection (Technical Specification Table 3.3.1.1)
Applicability: Mode 1,  (as specified in Technical Specifications Table 3.3.1.1-1) 3/4 Average Power Range Monitor (APRM) Flow Biased Rod Block Trip Setting (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
Applicability: Mode 1,  (as specified in Technical Requirements Manuals Table 3.3.4-1) 3/4 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
Applicability: Mode 1,  % RTP as specified in Table 3.3.2.1-1 (TS definition of RTP) 3/4 Shutdown Margin (SDM) Limit (Technical Specification 3.1.1)
Applicability: All Modes 1.3    Fuel Loading The core will contain fresh and exposed ATRIUM-10XM, as well as a limited number of ATRIUM-11 lead fuel assemblies. Nuclear fuel types used in the core loading are shown in Table 1.1. The core shuffle and final loading were explicitly evaluated for BOC cold shutdown margin performance as documented per Reference 5.
Browns Ferry Unit 2 Cycle 21                                                                                          Page 1 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 1.1 Nuclear Fuel Types
* Nuclear Original          Number of                              Fuel Names Fuel Description                                                              Fuel Type Cycle          Assemblies                                  (Range)
(NFT)
ATRIUM 10XM XMLC-3904B-15GV80-FBF                                19                  93                15          FBF401-FBF572 ATRIUM 10XM XMLC-4035B-13GV80-FBF                                19                  58                16          FBF573-FBF652 ATRIUM 11 A11-3693B-13GV80-FBF                                    19                  6                17          FBF653-FBF660 ATRIUM 10XM XMLC-4102B-11GV70-FBG-B                              20                  48                18          FBG701-FBG748 ATRIUM 10XM XMLC-4062B-13GV80-FBG-C                              20                151                19          FBG749-FBG900 ATRIUM 10XM XMLC-3948B-13GV70-FBG-B                              20                  88                20          FBG901-FBG988 ATRIUM-10XM XMLC-4087B-15GV80-FBH                                21                176                21          FBH001-FBH176 ATRIUM-10XM XMLC-4036B-15GV80-FBH                                21                  88                22          FBH177-FBH264 ATRIUM-10XM XMLC-4093B-10GV80-FBH                                21                  56                23          FBH265-FBH320 1.4      Acceptability Limits discussed in this document were generated based on NRC approved methodologies per References 6 through 26.
* The table identifies the expected fuel type breakdown in anticipation of final core loading. The final composition of the core depends upon uncertainties during the outage such as discovering a failed fuel bundle, or other bundle damage. Minor core loading changes, due to unforeseen events, will conform to the safety and monitoring requirements identified in this document.
Browns Ferry Unit 2 Cycle 21                                                                                                            Page 2 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 2 APLHGR Limits (Technical Specifications 3.2.1 & 3.7.5)
The APLHGR limit is determined by adjusting the rated power APLHGR limit for off-rated power, off-rated flow, and SLO conditions. The most limiting of these is then used as follows:
APLHGR limit = MIN ( APLHGRP , APLHGRF, APLHGRSLO )
where:
APLHGRP                off-rated power APLHGR limit                  [APLHGRRATED
* MAPFACP]
APLHGRF                off-rated flow APLHGR limit                    [APLHGRRATED
* MAPFACF]
APLHGRSLO              SLO APLHGR limit                              [APLHGRRATED
* SLO Multiplier]
2.1    Rated Power and Flow Limit: APLHGRRATED The rated conditions APLHGR for all fuel are identified per Reference 1. The rated conditions APLHGR for ATRIUM-10XM fuel are shown in Figure 2.1. The rated conditions APLHGR for ATRIUM-11 are shown in Figure 2.2.
2.2    Off-Rated Power Dependent Limit: APLHGRP Reference 1 does not specify a power dependent APLHGR. Therefore, MAPFACP is set to a value of 1.0.
2.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. No additional power dependent limitation is required.
2.3    Off-Rated Flow Dependent Limit: APLHGRF Reference 1 does not specify a flow dependent APLHGR. Therefore, MAPFACF is set to a value of 1.0.
2.4    Single Loop Operation Limit: APLHGRSLO The single loop operation multiplier for ATRIUM-10XM and ATRIUM-11 fuel is 0.85, per Reference 1.
Browns Ferry Unit 2 Cycle 21                                                                                          Page 3 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                    1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 15 12 APLHGR (kW/ft) 9 6
3 0
0          20                      40                    60                      80 Planar Average Exposure (GWd/MTU)
Planar Avg.        APLHGR Exposure            Limit (GWd/MTU)          (kW/ft) 0.0            13.0 15.0            13.0 67.0              7.6 Figure 2.1 APLHGRRATED for ATRIUM-10XM Fuel Browns Ferry Unit 2 Cycle 21                                                                                              Page 4 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                    1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 15 12 APLHGR (kW/ft) 9 6
3 0
0          20                      40                    60                      80 Planar Average Exposure (GWd/MTU)
Planar Avg.          APLHGR Exposure            Limit (GWd/MTU)          (kW/ft) 0.0              10 10.0              10 67.0              5.9 Figure 2.2 APLHGRRATED for ATRIUM-11 Fuel Browns Ferry Unit 2 Cycle 21                                                                                              Page 5 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 2.5      Equipment Out-Of-Service Corrections The limits shown in Figure 2.1 and Figure 2.2 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
In-Service                              All equipment In-Service
* RPTOOS                                  EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                  Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                  Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                        Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                  One Recirculation Pump Out-Of-Service
* All equipment service conditions assume 1 SRVOOS.
Browns Ferry Unit 2 Cycle 21                                                                                          Page 6 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 3 LHGR Limits (Technical Specification 3.2.3, 3.3.4.1, & 3.7.5)
The LHGR limit is determined by adjusting the rated power LHGR limit for off-rated power and off-rated flow conditions. The most limiting of these is then used as follows:
LHGR limit = MIN ( LHGRP, LHGRF )
where:
LHGRP                  off-rated power LHGR limit                    [LHGRRATED
* LHGRFACP]
LHGRF                  off-rated flow LHGR limit                      [LHGRRATED
* LHGRFACF]
3.1    Rated Power and Flow Limit: LHGRRATED The rated conditions LHGR for all fuel are identified per Reference 1. The rated conditions LHGR for ATRIUM-10XM are shown in Figure 3.1. The rated conditions LHGR for ATRIUM-11 fuel is shown in Figure 3.2. The LHGR limit is consistent with References 2 and 3.
3.2    Off-Rated Power Dependent Limit: LHGRP LHGR limits are adjusted for off-rated power conditions using the LHGRFACP multiplier provided in Reference 1. The multiplier is split into two sub cases: turbine bypass valves in and out-of-service. The base case multipliers are shown in Figure 3.3 and Figure 3.4.
3.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional limits are shown in Figure 3.7 through Figure 3.10, based on temperature conditions identified in Table 3.1.
Table 3.1 Startup Feedwater Temperature Basis Temperature Power              Range 1                Range 2
(% Rated)                (&deg;F)                  (&deg;F) 23                  155.0                150.0 30                  162.0                157.0 40                  172.0                167.0 50                  182.0                177.0 Browns Ferry Unit 2 Cycle 21                                                                                          Page 7 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 3.3      Off-Rated Flow Dependent Limit: LHGRF LHGR limits are adjusted for off-rated flow conditions using the LHGRFACF multiplier provided in Reference 1. Multipliers are shown in Figure 3.5 through Figure 3.6.
3.4      Equipment Out-Of-Service Corrections The limits shown in Figure 3.1 and Figure 3.2 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
* In-Service                              All equipment In-Service RPTOOS                                    EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                    Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                    Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                        Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                    One Recirculation Pump Out-Of-Service Off-rated power corrections shown in Figure 3.3 and Figure 3.4 are dependent on operation of the Turbine Bypass Valve system. For this reason, separate limits are to be applied for TBVIS or TBVOOS operation. The limits have no dependency on RPTOOS, PLUOOS, FHOOS/FFWTR, or SLO.
Off-rated flow corrections shown in Figure 3.5 and Figure 3.6 are bounding for all EOOS conditions.
Off-rated power corrections shown in Figure 3.7 through Figure 3.10 are also dependent on operation of the Turbine Bypass Valve system. In this case, limits support FHOOS operation during startup. These limits have no dependency on RPTOOS, PLUOOS, or SLO.
* All equipment service conditions assume 1 SRVOOS.
Browns Ferry Unit 2 Cycle 21                                                                                          Page 8 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                      1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 15 12 9
LHGR (kW/ft) 6 3
0 0            20                      40                    60                      80 Pellet Exposure (GWd/MTU)
Pellet              LHGR Exposure            Limit (GWd/MTU)          (kW/ft) 0.0              14.1 18.9              14.1 74.4              7.4 Figure 3.1 LHGRRATED for ATRIUM-10XM Fuel Browns Ferry Unit 2 Cycle 21                                                                                              Page 9 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                      1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 15 12 9
LHGR (kW/ft) 6 3
0 0            20                      40                    60                      80 Pellet Exposure (GWd/MTU)
Pellet              LHGR Exposure            Limit (GWd/MTU)          (kW/ft) 0.0              12.2 18.9              12.2 74.4              6.4 Figure 3.2 LHGRRATED for ATRIUM-11 Fuel Browns Ferry Unit 2 Cycle 21                                                                                            Page 10 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                          1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 1.10 1.00 0.90 Turbine Bypass Valve In-Service, TBVIS 0.80 LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 7%9,6 &RUH)ORZ 0.50 7%9226 &RUH)ORZ TBVIS, > 50% Core Flow 0.40          TBVOOS, > 50% Core Flow 0.30 0.20 20        30        40            50        60      70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service              Turb ine Bypass Out-of-Service Core                                      Core Power          LHGRFACP                  Power          LHGRFACP
(% Rated)                                  (% Rated) 100.0        1.00                        100.0        1.00 26.0        0.61                          26.0        0.60 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0        0.44                          26.0        0.38 23.0        0.42                          23.0        0.34
                                &RUH)ORZ5DWHG                      &RUH)ORZ5DWHG 26.0        0.48                          26.0        0.48 23.0        0.45                          23.0        0.43 Figure 3.3 Base Operation LHGRFACP for ATRIUM-10XM Fuel (Independent of other EOOS conditions)
Browns Ferry Unit 2 Cycle 21                                                                                            Page 11 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                          1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 1.10 1.00 0.90 Turbine Bypass Valve In-Service, TBVIS 0.80 LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 0.50          7%9,6 &RUH)ORZ 7%9226 &RUH)ORZ TBVIS, > 50% Core Flow 0.40          TBVOOS, > 50% Core Flow 0.30 0.20 20      30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                    Power          LHGRFACP
(% Rated)                                    (% Rated) 100.0        1.00                          100.0        1.00 26.0        0.61                            26.0        0.60 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0        0.44                            26.0        0.38 23.0        0.42                            23.0        0.34
                              &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0        0.48                            26.0        0.48 23.0        0.45                            23.0        0.43 Figure 3.4 Base Operation LHGRFACP for ATRIUM-11 Fuel (Independent of other EOOS conditions)
Browns Ferry Unit 2 Cycle 21                                                                                            Page 12 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                              1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 1.10 1.05 1.00 0.95 0.90 LHGRFACF 0.85 0.80 0.75 0.70 0.65 0.60 25      35          45            55        65          75      85        95          105 Core Flow (% Rated)
Core Flow          LHGRFACF
(% Rated) 0.0              0.62 30.0            0.62 76.7            1.00 107.0            1.00 Figure 3.5 LHGRFACF for ATRIUM-10XM Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 2 Cycle 21                                                                                            Page 13 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                              1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 1.10 1.05 1.00 0.95 0.90 LHGRFACF 0.85 0.80 0.75 0.70 0.65 0.60 25      35          45            55        65          75      85        95          105 Core Flow (% Rated)
Core Flow          LHGRFACF
(% Rated) 0.0              0.66 30.0            0.66 73.7            1.00 107.0            1.00 Figure 3.6 LHGRFACF for ATRIUM-11 Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 2 Cycle 21                                                                                            Page 14 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                          1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 1.10 1.00 0.90 0.80          Turbine Bypass Valve In-Service, TBVIS LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 0.50 7%9,6 &RUH)ORZ 0.40          7%9226 &RUH)ORZ TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20    30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                    Power            LHGRFACP
(% Rated)                                    (% Rated) 100.0        1.00                          100.0        1.00 26.0        0.51                          26.0        0.51 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0        0.39                          26.0        0.34 23.0        0.36                          23.0        0.31
                              &RUH)ORZ5DWHG                      &RUH)ORZ5DWHG 26.0        0.41                          26.0        0.41 23.0        0.41                          23.0        0.38 Figure 3.7 Startup Operation LHGRFACP for ATRIUM-10XM Fuel:
Table 3.1 Temperature Range 1 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 2 Cycle 21                                                                                            Page 15 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                          1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 1.10 1.00 0.90 0.80          Turbine Bypass Valve In-Service, TBVIS LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 0.50 7%9,6 &RUH)ORZ 0.40          7%9226 &RUH)ORZ TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20    30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                    Power            LHGRFACP
(% Rated)                                    (% Rated) 100.0        1.00                          100.0        1.00 26.0        0.51                          26.0        0.51 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0        0.39                          26.0        0.34 23.0        0.35                          23.0        0.30
                              &RUH)ORZ5DWHG                      &RUH)ORZ5DWHG 26.0        0.41                          26.0        0.41 23.0        0.40                          23.0        0.38 Figure 3.8 Startup Operation LHGRFACP for ATRIUM-10XM Fuel:
Table 3.1 Temperature Range 2 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 2 Cycle 21                                                                                            Page 16 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                          1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 1.10 1.00 0.90 0.80            Turbine Bypass Valve In-Service, TBVIS LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 0.50 7%9,6 &RUH)ORZ 0.40          7%9226 &RUH)ORZ TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20      30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                    Power          LHGRFACP
(% Rated)                                    (% Rated) 100.0        1.00                            100.0        1.00 26.0        0.51                            26.0        0.51 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0        0.39                            26.0        0.34 23.0        0.36                            23.0        0.31
                              &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0        0.41                            26.0        0.41 23.0        0.41                            23.0        0.38 Figure 3.9 Startup Operation LHGRFACP for ATRIUM-11 Fuel:
Table 3.1 Temperature Range 1 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 2 Cycle 21                                                                                              Page 17 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                          1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 1.10 1.00 0.90 0.80          Turbine Bypass Valve In-Service, TBVIS LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 0.50 7%9,6 &RUH)ORZ 0.40          7%9226 &RUH)ORZ TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20      30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service              Turb ine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                  Power            LHGRFACP
(% Rated)                                  (% Rated) 100.0        1.00                          100.0        1.00 26.0        0.51                          26.0        0.51 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0        0.39                          26.0        0.34 23.0        0.35                          23.0        0.30
                                &RUH)ORZ5DWHG                      &RUH)ORZ5DWHG 26.0        0.41                          26.0        0.41 23.0        0.40                          23.0        0.38 Figure 3.10 Startup Operation LHGRFACP for ATRIUM-11 Fuel:
Table 3.1 Temperature Range 2 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 2 Cycle 21                                                                                            Page 18 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
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[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 4 OLMCPR Limits (Technical Specification 3.2.2, 3.3.4.1, & 3.7.5)
OLMCPR is calculated to be the most limiting of the flow or power dependent values OLMCPR limit = MAX ( MCPRF , MCPRP )
where:
MCPRF              core flow-dependent MCPR limit MCPRP              power-dependent MCPR limit 4.1    Flow Dependent MCPR Limit: MCPRF MCPRF limits are dependent upon core flow (% of Rated), and the max core flow limit, (Rated or Increased Core Flow, ICF). MCPRF limits are shown in Figure 4.1, per Reference 1. Limits are valid for all EOOS combinations. No adjustment is required for SLO conditions.
4.2    Power Dependent MCPR Limit: MCPRP MCPRP limits are dependent upon:
Core Power Level (% of Rated)
Technical Specification Scram Speed (TSSS), Nominal Scram Speed (NSS), or Optimum Scram Speed (OSS)
Cycle Operating Exposure (NEOC, EOC, and CD - as defined in this section)
Equipment Out-Of-Service Options Two or Single recirculation Loop Operation (TLO vs. SLO)
The MCPRP limits are provided in Table 4.2 through Table 4.9, where each table contains the limits for all fuel types and EOOS options (for a specified scram speed and exposure range).
The CMSS determines MCPRP limits, from these tables, based on linear interpolation between the specified powers.
4.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional power dependent limits are shown in Table 4.5 through Table 4.8 based on temperature conditions identified in Table 3.1.
Browns Ferry Unit 2 Cycle 21                                                                                        Page 19 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 4.2.2    Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS)
MCPRP limits are provided for three different sets of assumed scram speeds. The Technical Specification Scram Speed (TSSS) MCPRP limits are applicable at all times, as long as the scram time surveillance demonstrates the times in Technical Specification Table 3.1.4-1 are met. Both Nominal Scram Speeds (NSS) and/or Optimum Scram Speeds (OSS) may be used, as long as the scram time surveillance demonstrates Table 4.1 times are applicable.
* Table 4.1 Nominal Scram Time Basis Notch                  Nominal              Optimum Position            Scram Timing          Scram Timing (index)                (seconds)              (seconds) 46                        0.420                  0.380 36                        0.980                  0.875 26                        1.600                  1.465 6                      2.900                  2.900 In demonstrating compliance with the NSS and/or OSS scram time basis, surveillance requirements from Technical Specification 3.1.4 apply; accepting the definition of SLOW rods should conform to scram speeds shown in Table 4.1. If conformance is not demonstrated, TSSS based MCPRP limits are applied.
On initial cycle startup, TSSS limits are used until the successful completion of scram timing confirms NSS and/or OSS based limits are applicable.
4.2.3    Exposure Dependent Limits Exposures are tracked on a Core Average Exposure basis (CAVEX, not Cycle Exposure).
Higher exposure MCPRP limits are always more limiting and may be used for any Core Average Exposure up to the ending exposure. Per Reference 1, MCPRP limits are provided for the following exposure ranges:
BOC to NEOC                                NEOC corresponds to                        29,301.9 MWd / MTU BOC to EOCLB                                EOCLB corresponds to                        34,230.7 MWd / MTU BOC to End of Coast                        End of Coast                                35,897.6 MWd / MTU NEOC refers to a Near EOC exposure point.
* Reference 1 analysis results are based on information identified in Reference 4.
Drop out times consistent with method used to perform actual timing measurements (i.e., including pickup/dropout effects).
Browns Ferry Unit 2 Cycle 21                                                                                                    Page 20 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                      TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 The EOCLB exposure point is not the true End-Of-Cycle exposure. Instead it corresponds to a licensing exposure window exceeding expected end-of-full-power-life.
The End of Coast exposure point represents a licensing exposure point exceeding the expected end-of-cycle exposure including cycle extension options.
4.2.4      Equipment Out-Of-Service (EOOS) Options EOOS options
* covered by MCPRP limits are given by the following:
In-Service                                          All equipment In-Service RPTOOS                                              EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                              Turbine Bypass Valve(s) Out-Of-Service RPTOOS+TBVOOS                                      Combined RPTOOS and TBVOOS PLUOOS                                              Power Load Unbalance Out-Of-Service PLUOOS+RPTOOS                                      Combined PLUOOS and RPTOOS PLUOOS+TBVOOS                                      Combined PLUOOS and TBVOOS PLUOOS+TBVOOS+RPTOOS                                Combined PLUOOS, RPTOOS, and TBVOOS FHOOS (or FFWTR)                                    Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)
RCPOOS                                              One Recirculation Pump Out-Of-Service For exposure ranges up to NEOC and EOCLB, additional combinations of MCPRP limits are also provided including FHOOS. The coast down exposure range assumes application of FFWTR. FHOOS based MCPRP limits for the coast down exposure are redundant because the temperature setdown assumption is identical with FFWTR.
4.2.5      Single-Loop-Operation (SLO) Limits When operating in RCPOOS conditions, MCPRp limits are constructed differently from the normal operating RCP conditions. The limiting event for RCPOOS is a pump seizure scenario, which sets the upper bound for allowed core power and flow . This event is not impacted by scram time assumptions. Specific MCPRP limits are shown in Table 4.9.
4.2.6      Below Pbypass Limits Below Pbypass (26% rated power), MCPRP limits depend upon core flow. One set of MCPRP limits applies for core flow above 50% of rated; a second set applies if the core flow is less than or equal to 50% rated.
* All equipment service conditions assume 1 SRVOOS.
RCPOOS limits are only valid up to 43.75% rated core power, 50% rated core flow, and an active recirculation drive flow of 17.73 Mlbm/hr.
Browns Ferry Unit 2 Cycle 21                                                                                                        Page 21 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                              1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 2.00 1.80 1.60 MCPRF 1.40 1.20 1.00 30        40          50            60          70          80          90        100          110 Core Flow (% Rated)
Core Flow            MCPRF
(% Rated) 30.0              1.58 84.0              1.34 107.0              1.34 Figure 4.1 MCPRF for All Fuel Types (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 2 Cycle 21                                                                                            Page 22 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                          1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 4.2 MCPRP Limits for All Fuel Types: Optimum Scram Time Basis
* ATRIUM-10XM                                  ATRIUM-11 BOC          BOC            BOC            BOC          BOC        BOC Pow er              to            to        to End of        to            to    to End of Operating Condition        (% of rated)      NEOC          EOCLB          Coast        NEOC          EOCLB      Coast 100            1.37          1.41          1.43          1.37          1.43      1.43 90            1.42          1.45          1.48          1.42          1.46      1.48 65            1.59          1.59          1.64          1.59          1.59      1.64
                                    > 50            1.70          1.70          1.77          1.73          1.73      1.81 50            1.81          1.81          1.81          1.91          1.91      1.91 Base Case                    40            1.88          1.88          1.94          1.99          1.99      2.03 26            2.30          2.30          2.42          2.51          2.51      2.66 26 at > 50%F        2.54          2.54          2.65          2.68          2.68      2.81 23 at > 50%F        2.69          2.69          2.81          2.93          2.93      3.05
                              DW)        2.42          2.42          2.51          2.63          2.63      2.72
                              DW)        2.55          2.55          2.67          2.93          2.93      3.05 100            1.42          1.43            ---          1.42          1.43        ---
90            1.48          1.48            ---          1.48          1.48        ---
65            1.64          1.64            ---          1.64          1.64        ---
                                    > 50            1.77          1.77            ---          1.81          1.81        ---
50            1.81          1.81            ---          1.91          1.91        ---
FHOOS                        40            1.94          1.94            ---          2.03          2.03        ---
26            2.42          2.42            ---          2.66          2.66        ---
26 at > 50%F        2.65          2.65            ---          2.81          2.81        ---
23 at > 50%F        2.81          2.81            ---          3.05          3.05        ---
                              DW)        2.51          2.51            ---          2.72          2.72        ---
                              DW)        2.67          2.67            ---          3.05          3.05        ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR/FHOOS is supported for the BOC to End of Coast limits.
Browns Ferry Unit 2 Cycle 21                                                                                                    Page 23 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM                              ATRIUM-11 BOC        BOC          BOC          BOC        BOC        BOC Pow er            to          to        to End of        to          to    to End of Operating Condition      (% of rated)      NEOC        EOCLB        Coast        NEOC        EOCLB      Coast 100            1.41        1.42          1.50        1.41        1.44        1.50 90          1.46        1.46          1.55        1.46        1.47        1.55 65          1.61        1.61          1.72        1.61        1.61        1.72
                                        > 50            1.73        1.73          1.84        1.76        1.76        1.84 50            1.81        1.81          1.85        1.92        1.92        1.92 Base Case                40          1.88        1.88          2.00        2.00        2.00        2.06 26          2.33        2.33          2.45        2.55        2.55        2.69 26 at > 50%F        2.54        2.54          2.65        2.68        2.68        2.81 23 at > 50%F        2.69        2.69          2.81        2.93        2.93        3.05
                                    DW)        2.42        2.42          2.51        2.63        2.63        2.72
                                    DW)        2.55        2.55          2.67        2.93        2.93        3.05 100            1.45        1.48          1.54        1.46        1.49        1.54 90          1.50        1.51          1.59        1.50        1.52        1.59 65          1.65        1.65          1.75        1.65        1.65        1.75
                                        > 50            1.76        1.76          ---        1.78        1.78        1.87 50            1.82        1.82          1.87        1.92        1.92        1.92 TBVOOS                  40          1.90        1.90          2.03        2.00        2.00        2.06 26          2.33        2.33          2.45        2.55        2.55        2.69 26 at > 50%F        3.11        3.11          3.23        3.28        3.28        3.38 23 at > 50%F        3.35        3.35          3.47        3.57        3.57        3.69
                                    DW)        2.78        2.78          2.90        3.04        3.04        3.16
                                    DW)        3.04        3.04          3.18        3.32        3.32        3.56 100            1.50        1.50          ---        1.50        1.50        ---
90          1.55        1.55          ---        1.55        1.55        ---
65          1.72        1.72          ---        1.72        1.72        ---
                                        > 50            1.84        1.84          ---        1.84        1.84        ---
50            1.85        1.85          ---        1.92        1.92        ---
FHOOS                    40          2.00        2.00          ---        2.06        2.06        ---
26          2.45        2.45          ---        2.69        2.69        ---
26 at > 50%F        2.65        2.65          ---        2.81        2.81        ---
23 at > 50%F        2.81        2.81          ---        3.05        3.05        ---
                                    DW)        2.51        2.51          ---        2.72        2.72        ---
                                    DW)        2.67        2.67          ---        3.05        3.05        ---
100            1.41        1.42          1.50        1.41        1.44        1.50 90          1.46        1.46          1.55        1.46        1.47        1.55 65          1.73        1.73          1.77        1.83        1.83        1.83
                                        > 50            ---        ---          ---          ---        ---        ---
50            1.81        1.81          1.85        1.92        1.92        1.92 PLUOOS                  40          1.88        1.88          2.00        2.00        2.00        2.06 26          2.33        2.33          2.45        2.55        2.55        2.69 26 at > 50%F        2.54        2.54          2.65        2.68        2.68        2.81 23 at > 50%F        2.69        2.69          2.81        2.93        2.93        3.05
                                    DW)        2.42        2.42          2.51        2.63        2.63        2.72
                                    DW)        2.55        2.55          2.67        2.93        2.93        3.05
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 2 Cycle 21                                                                                                    Page 24 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis (continued)
* ATRIUM-10XM                              ATRIUM-11 BOC        BOC          BOC          BOC          BOC        BOC Pow er            to        to        to End of        to          to      to End of Operating Condition      (% of rated)      NEOC      EOCLB          Coast        NEOC      EOCLB        Coast 100            1.54      1.54          ---        1.54        1.54        ---
90            1.59      1.59          ---        1.59        1.59        ---
65            1.75      1.75          ---        1.75        1.75        ---
                                        > 50            ---        ---          ---        1.87        1.87        ---
50            1.87      1.87          ---        1.92        1.92        ---
TBVOOS 40            2.03      2.03          ---        2.06        2.06        ---
FHOOS 26            2.45      2.45          ---        2.69        2.69        ---
26 at > 50%F        3.23      3.23          ---        3.38        3.38        ---
23 at > 50%F        3.47      3.47          ---        3.69        3.69        ---
                                  DW)        2.90      2.90          ---        3.16        3.16        ---
                                  DW)        3.18      3.18          ---        3.56        3.56        ---
100            1.45      1.48          1.54        1.46        1.49        1.54 90            1.50      1.51          1.59        1.50        1.52        1.59 65            1.74      1.74          1.79        1.83        1.83        1.83
                                        > 50            ---        ---          ---          ---        ---        ---
50            1.82      1.82          1.87        1.92        1.92        1.92 TBVOOS 40            1.90      1.90          2.03        2.00        2.00        2.06 PLUOOS 26            2.33      2.33          2.45        2.55        2.55        2.69 26 at > 50%F        3.11      3.11          3.23        3.28        3.28        3.38 23 at > 50%F        3.35      3.35          3.47        3.57        3.57        3.69
                                  DW)        2.78      2.78          2.90        3.04        3.04        3.16
                                  DW)        3.04      3.04          3.18        3.32        3.32        3.56 100            1.50      1.50          ---        1.50        1.50        ---
90            1.55      1.55          ---        1.55        1.55        ---
65            1.77      1.77          ---        1.83        1.83        ---
                                        > 50            ---        ---          ---          ---        ---        ---
50            1.85      1.85          ---        1.92        1.92        ---
FHOOS 40            2.00      2.00          ---        2.06        2.06        ---
PLUOOS 26            2.45      2.45          ---        2.69        2.69        ---
26 at > 50%F        2.65      2.65          ---        2.81        2.81        ---
23 at > 50%F        2.81      2.81          ---        3.05        3.05        ---
                                  DW)        2.51      2.51          ---        2.72        2.72        ---
                                  DW)        2.67      2.67          ---        3.05        3.05        ---
100            1.54      1.54          ---        1.54        1.54        ---
90            1.59      1.59          ---        1.59        1.59        ---
65            1.79      1.79          ---        1.83        1.83        ---
                                        > 50            ---        ---          ---          ---        ---        ---
TBVOOS                  50            1.87      1.87          ---        1.92        1.92        ---
FHOOS                    40            2.03      2.03          ---        2.06        2.06        ---
PLUOOS                  26            2.45      2.45          ---        2.69        2.69        ---
26 at > 50%F        3.23      3.23          ---        3.38        3.38        ---
23 at > 50%F        3.47      3.47          ---        3.69        3.69        ---
                                  DW)        2.90      2.90          ---        3.16        3.16        ---
                                  DW)        3.18      3.18          ---        3.56        3.56        ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 2 Cycle 21                                                                                                    Page 25 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM                              ATRIUM-11 BOC        BOC          BOC          BOC          BOC        BOC Pow er            to        to        to End of        to          to      to End of Operating Condition      (% of rated)      NEOC      EOCLB          Coast        NEOC      EOCLB        Coast 100            1.47      1.47          1.56        1.47        1.47        1.56 90            1.52      1.52          1.61        1.52        1.52        1.61 65            1.67      1.67          1.78        1.67        1.67        1.78
                                        > 50            1.78      1.78          ---        1.79        1.79        1.90 50            1.84      1.84          1.90        1.93        1.93        1.93 Base Case                40            1.93      1.93          2.06        2.01        2.01        2.08 26            2.35      2.35          2.47        2.57        2.57        2.72 26 at > 50%F        2.54      2.54          2.65        2.68        2.68        2.81 23 at > 50%F        2.69      2.69          2.81        2.93        2.93        3.05
                                  DW)        2.42      2.42          2.51        2.63        2.63        2.72
                                  DW)        2.55      2.55          2.67        2.93        2.93        3.05 100            1.53      1.53          1.60        1.53        1.53        1.60 90            1.57      1.57          1.65        1.57        1.57        1.65 65            1.74      1.74          1.82        1.74        1.74        1.82
                                        > 50            1.83      1.83          ---        1.83        1.83        ---
50            1.87      1.87          1.93        1.93        1.93        1.93 TBVOOS                  40            1.97      1.97          2.08        2.01        2.01        2.09 26            2.36      2.36          2.49        2.57        2.57        2.72 26 at > 50%F        3.11      3.11          3.23        3.28        3.28        3.38 23 at > 50%F        3.35      3.35          3.47        3.57        3.57        3.69
                                  DW)        2.78      2.78          2.90        3.04        3.04        3.16
                                  DW)        3.04      3.04          3.18        3.32        3.32        3.56 100            1.56      1.56          ---        1.56        1.56        ---
90            1.61      1.61          ---        1.61        1.61        ---
65            1.78      1.78          ---        1.78        1.78        ---
                                        > 50            ---        ---          ---        1.90        1.90        ---
50            1.90      1.90          ---        1.93        1.93        ---
FHOOS                    40            2.06      2.06          ---        2.08        2.08        ---
26            2.47      2.47          ---        2.72        2.72        ---
26 at > 50%F        2.65      2.65          ---        2.81        2.81        ---
23 at > 50%F        2.81      2.81          ---        3.05        3.05        ---
                                  DW)        2.51      2.51          ---        2.72        2.72        ---
                                  DW)        2.67      2.67          ---        3.05        3.05        ---
100            1.47      1.47          1.56        1.47        1.47        1.56 90            1.52      1.52          1.61        1.52        1.52        1.61 65            1.76      1.76          1.81        1.84        1.84        1.84
                                        > 50            ---        ---          ---          ---        ---        ---
50            1.84      1.84          1.90        1.93        1.93        1.93 PLUOOS                  40            1.93      1.93          2.06        2.01        2.01        2.08 26            2.35      2.35          2.47        2.57        2.57        2.72 26 at > 50%F        2.54      2.54          2.65        2.68        2.68        2.81 23 at > 50%F        2.69      2.69          2.81        2.93        2.93        3.05
                                  DW)        2.42      2.42          2.51        2.63        2.63        2.72
                                  DW)        2.55      2.55          2.67        2.93        2.93        3.05
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 2 Cycle 21                                                                                                    Page 26 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis (continued)
* ATRIUM-10XM                              ATRIUM-11 BOC        BOC          BOC          BOC          BOC        BOC Pow er            to        to      to End of        to          to    to End of Operating Condition      (% of rated)      NEOC      EOCLB          Coast        NEOC      EOCLB        Coast 100            1.60      1.60          ---        1.60        1.60        ---
90            1.65      1.65          ---        1.65        1.65        ---
65            1.82      1.82          ---        1.82        1.82        ---
                                        > 50            ---        ---          ---          ---        ---        ---
50            1.93      1.93          ---        1.93        1.93        ---
TBVOOS 40            2.08      2.08          ---        2.09        2.09        ---
FHOOS 26            2.49      2.49          ---        2.72        2.72        ---
26 at > 50%F        3.23      3.23          ---        3.38        3.38        ---
23 at > 50%F        3.47      3.47          ---        3.69        3.69        ---
                                    DW)        2.90      2.90          ---        3.16        3.16        ---
                                    DW)        3.18      3.18          ---        3.56        3.56        ---
100            1.53      1.53          1.60        1.53        1.53        1.60 90            1.57      1.57          1.65        1.57        1.57        1.65 65            1.79      1.79          1.83        1.84        1.84        1.84
                                        > 50            ---        ---          ---          ---        ---        ---
50            1.87      1.87          1.93        1.93        1.93        1.93 TBVOOS 40            1.97      1.97          2.08        2.01        2.01        2.09 PLUOOS 26            2.36      2.36          2.49        2.57        2.57        2.72 26 at > 50%F        3.11      3.11          3.23        3.28        3.28        3.38 23 at > 50%F        3.35      3.35          3.47        3.57        3.57        3.69
                                    DW)        2.78      2.78          2.90        3.04        3.04        3.16
                                    DW)        3.04      3.04          3.18        3.32        3.32        3.56 100            1.56      1.56          ---        1.56        1.56        ---
90            1.61      1.61          ---        1.61        1.61        ---
65            1.81      1.81          ---        1.84        1.84        ---
                                        > 50            ---        ---          ---          ---        ---        ---
50            1.90      1.90          ---        1.93        1.93        ---
FHOOS 40            2.06      2.06          ---        2.08        2.08        ---
PLUOOS 26            2.47      2.47          ---        2.72        2.72        ---
26 at > 50%F        2.65      2.65          ---        2.81        2.81        ---
23 at > 50%F        2.81      2.81          ---        3.05        3.05        ---
                                    DW)        2.51      2.51          ---        2.72        2.72        ---
                                    DW)        2.67      2.67          ---        3.05        3.05        ---
100            1.60      1.60          ---        1.60        1.60        ---
90            1.65      1.65          ---        1.65        1.65        ---
65            1.83      1.83          ---        1.84        1.84        ---
                                        > 50            ---        ---          ---          ---        ---        ---
TBVOOS                  50            1.93      1.93          ---        1.93        1.93        ---
FHOOS                    40            2.08      2.08          ---        2.09        2.09        ---
PLUOOS                  26            2.49      2.49          ---        2.72        2.72        ---
26 at > 50%F        3.23      3.23          ---        3.38        3.38        ---
23 at > 50%F        3.47      3.47          ---        3.69        3.69        ---
                                    DW)        2.90      2.90          ---        3.16        3.16        ---
                                    DW)        3.18      3.18          ---        3.56        3.56        ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 2 Cycle 21                                                                                                    Page 27 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                              1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 4.5 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM                        ATRIUM-11 BOC        BOC        BOC      BOC          BOC          BOC Pow er            to          to    to End of    to          to      to End of Operating Condition          (% of rated)      NEOC        EOCLB      Coast    NEOC        EOCLB        Coast 100            1.50        1.50      1.50      1.50        1.50        1.50 90          1.55        1.55      1.55      1.55        1.55        1.55 65          1.77        1.77      1.77      1.83        1.83        1.83
                                        > 50            ---          ---      ---      ---          ---          ---
50            1.98        1.98      1.98      2.02        2.02        2.02 TBVIS                        40          2.18        2.18      2.18      2.29        2.29        2.29 26          2.77        2.77      2.77      3.07        3.07        3.07 26 at > 50%F        3.06        3.06      3.06      3.12        3.12        3.12 23 at > 50%F        3.15        3.15      3.15      3.42        3.42        3.42
                                    DW)        2.79        2.79      2.79      3.07        3.07        3.07
                                    DW)        3.01        3.01      3.01      3.37        3.37        3.37 100            1.54        1.54      1.54      1.54        1.54        1.54 90          1.59        1.59      1.59      1.59        1.59        1.59 65          1.79        1.79      1.79      1.83        1.83        1.83
                                        > 50            ---          ---      ---      ---          ---          ---
50            2.00        2.00      2.00      2.02        2.02        2.02 TBVOOS                        40          2.20        2.20      2.20      2.29        2.29        2.29 26          2.77        2.77      2.77      3.07        3.07        3.07 26 at > 50%F        3.55        3.55      3.55      3.60        3.60        3.60 23 at > 50%F        3.71        3.71      3.71      4.00        4.00        4.00
                                    DW)        3.14        3.14      3.14      3.39        3.39        3.39
                                    DW)        3.45        3.45      3.45      3.89        3.89        3.89
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 2 Cycle 21                                                                                                  Page 28 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                              1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 4.6 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM                        ATRIUM-11 BOC        BOC        BOC      BOC          BOC          BOC Pow er            to          to    to End of    to          to      to End of Operating Condition          (% of rated)      NEOC        EOCLB      Coast    NEOC        EOCLB        Coast 100            1.50        1.50      1.50      1.50        1.50        1.50 90            1.55        1.55      1.55      1.55        1.55        1.55 65            1.77        1.77      1.77      1.83        1.83        1.83
                                        > 50            ---          ---      ---      ---          ---          ---
50            1.99        1.99      1.99      2.02        2.02        2.02 TBVIS                        40          2.19        2.19      2.19      2.30        2.30        2.30 26          2.79        2.79      2.79      3.09        3.09        3.09 26 at > 50%F        3.09        3.09      3.09      3.14        3.14        3.14 23 at > 50%F        3.17        3.17      3.17      3.44        3.44        3.44
                                    DW)        2.81        2.81      2.81      3.09        3.09        3.09
                                    DW)        3.03        3.03      3.03      3.39        3.39        3.39 100            1.54        1.54      1.54      1.54        1.54        1.54 90          1.59        1.59      1.59      1.59        1.59        1.59 65          1.79        1.79      1.79      1.83        1.83        1.83
                                        > 50            ---          ---      ---      ---          ---          ---
50            2.01        2.01      2.01      2.02        2.02        2.02 TBVOOS                        40          2.21        2.21      2.21      2.30        2.30        2.30 26          2.79        2.79      2.79      3.09        3.09        3.09 26 at > 50%F        3.58        3.58      3.58      3.62        3.62        3.62 23 at > 50%F        3.73        3.73      3.73      4.02        4.02        4.02
                                    DW)        3.15        3.15      3.15      3.40        3.40        3.40
                                    DW)        3.46        3.46      3.46      3.90        3.90        3.90
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 2 Cycle 21                                                                                                  Page 29 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                              1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 4.7 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM                        ATRIUM-11 BOC        BOC        BOC      BOC          BOC          BOC Pow er            to          to    to End of    to          to      to End of Operating Condition          (% of rated)      NEOC        EOCLB      Coast    NEOC        EOCLB        Coast 100            1.56        1.56      1.56      1.56        1.56        1.56 90            1.61        1.61      1.61      1.61        1.61        1.61 65            1.81        1.81      1.81      1.84        1.84        1.84
                                        > 50            ---          ---      ---      ---          ---          ---
50            2.03        2.03      2.03      2.04        2.04        2.04 TBVIS                        40          2.23        2.23      2.23      2.32        2.32        2.32 26          2.79        2.79      2.79      3.10        3.10        3.10 26 at > 50%F        3.06        3.06      3.06      3.12        3.12        3.12 23 at > 50%F        3.15        3.15      3.15      3.42        3.42        3.42
                                    DW)        2.79        2.79      2.79      3.10        3.10        3.10
                                    DW)        3.01        3.01      3.01      3.40        3.40        3.40 100            1.60        1.60      1.60      1.60        1.60        1.60 90          1.65        1.65      1.65      1.65        1.65        1.65 65          1.83        1.83      1.83      1.84        1.84        1.84
                                        > 50            ---          ---      ---      ---          ---          ---
50            2.05        2.05      2.05      2.05        2.05        2.05 TBVOOS                        40          2.25        2.25      2.25      2.32        2.32        2.32 26          2.79        2.79      2.79      3.10        3.10        3.10 26 at > 50%F        3.55        3.55      3.55      3.60        3.60        3.60 23 at > 50%F        3.71        3.71      3.71      4.00        4.00        4.00
                                    DW)        3.14        3.14      3.14      3.39        3.39        3.39
                                    DW)        3.45        3.45      3.45      3.89        3.89        3.89
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 2 Cycle 21                                                                                                  Page 30 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                              1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 4.8 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM                        ATRIUM-11 BOC        BOC        BOC      BOC          BOC          BOC Pow er            to          to    to End of    to          to      to End of Operating Condition          (% of rated)      NEOC        EOCLB      Coast    NEOC        EOCLB        Coast 100            1.56        1.56      1.56      1.56        1.56        1.56 90            1.61        1.61      1.61      1.61        1.61        1.61 65            1.81        1.81      1.81      1.84        1.84        1.84
                                        > 50            ---          ---      ---      ---          ---          ---
50            2.04        2.04      2.04      2.05        2.05        2.05 TBVIS                        40          2.24        2.24      2.24      2.33        2.33        2.33 26          2.81        2.81      2.81      3.12        3.12        3.12 26 at > 50%F        3.09        3.09      3.09      3.14        3.14        3.14 23 at > 50%F        3.17        3.17      3.17      3.44        3.44        3.44
                                    DW)        2.81        2.81      2.81      3.12        3.12        3.12
                                    DW)        3.03        3.03      3.03      3.42        3.42        3.42 100            1.60        1.60      1.60      1.60        1.60        1.60 90          1.65        1.65      1.65      1.65        1.65        1.65 65          1.83        1.83      1.83      1.84        1.84        1.84
                                        > 50            ---          ---      ---      ---          ---          ---
50            2.06        2.06      2.06      2.06        2.06        2.06 TBVOOS                        40          2.26        2.26      2.26      2.33        2.33        2.33 26          2.81        2.81      2.81      3.12        3.12        3.12 26 at > 50%F        3.58        3.58      3.58      3.62        3.62        3.62 23 at > 50%F        3.73        3.73      3.73      4.02        4.02        4.02
                                    DW)        3.15        3.15      3.15      3.40        3.40        3.40
                                    DW)        3.46        3.46      3.46      3.90        3.90        3.90
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 2 Cycle 21                                                                                                  Page 31 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF2C21, Revision 1 (Final)
 
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[I!D                          1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 4.9 MCPRP Limits for All Fuel Types: Single Loop Operation for All Scram Times
* Pow er                All Cycle Exposures Operating Condition      (% of rated)      ATRIUM-10XM        ATRIUM-11 100                2.02            2.05 43.75              2.02            2.05 40                2.08            2.10 RCPOOS                  26                2.49            2.74 FHOOS            26 at > 50%F            2.67            2.83 23 at > 50%F            2.83            3.07
                                                      DW)            2.53            2.74
                                                      DW)            2.69            3.07 100                2.05            2.05 43.75              2.05            2.05 RCPOOS                  40                2.10            2.11 TBVOOS                  26                2.51            2.74 PLUOOS            26 at > 50%F            3.25            3.40 FHOOS            23 at > 50%F            3.49            3.71
                                                      DW)            2.92            3.18
                                                      DW)            3.20            3.58 100                2.20            2.24 43.75              2.20            2.24 40                2.27            2.34 RCPOOS 26                2.81            3.12 TBVOOS 26 at > 50%F            3.57            3.62 FHOOS1 23 at > 50%F            3.73            4.02
                                                      DW)            3.16            3.41
                                                      DW)            3.47            3.91 100                2.21            2.25 43.75              2.21            2.25 40                2.28            2.35 RCPOOS 26                2.83            3.14 TBVOOS 26 at > 50%F            3.60            3.64 FHOOS2 23 at > 50%F            3.75            4.04
                                                      DW)            3.17            3.42
                                                      DW)            3.48            3.92
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop.
RCPOOS limits are only valid up to 50% rated core power, 50% rated core flow, and an active recirculation drive flow of 17.73 Mlbm/hr.
Browns Ferry Unit 2 Cycle 21                                                                                                      Page 32 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 5      Thermal-Hydraulic Stability Protection (Technical Specification 3.3.1.1)
Technical Specification Table 3.3.1.1-1, Function 2f, identifies the function.
Instrument setpoints are established, such that the reactor will be tripped before an oscillation can grow to the point where the SLMCPR is exceeded. With application of Reference 30, the DSS-CD stability solution will be used per Reference 26. The DSS-CD SAD setpoint is 1.10 for TLO and SLO.
New analyses have been developed based on Reference 26. With the implementation of the MELLLA+ operating domain expansion, an ABSP trip is required when the OPRM is out-of-service. The ABSP trip settings define a region of the power to flow map within which an automatic reactor scram occurs. The ABSP trip settings are provided in Table 5.1. If both the OPRM and ABSP are out-of-service, operation within the MELLLA+ domain is not allowed and the MBSP Regions provide stability protection. Table 5.2 and Table 5.3 provide the endpoints for the MBSP regions for nominal and reduced feedwater temperature conditions.
Table 5.1 ABSP Setpoints for the Scram Region Parameter                Symbol          Setting Value (unit)                      Comments Slope of ABSP APRM low Flow Biased Slope for Trip              mTRIP          2.00 (% RTP/% RDF)
Trip Linear Segment ABSP APRM Flow Biased Trip Setpoint Constant Power                                                            Power Intercept. Constant Power Line PBSP-TRIP            35.0 (% RTP)
Line for Trip                                                          for Trip from Zero Drive Flow to Flow Breakpoint Value ABSP APRM Flow Biased Trip Setpoint Constant Flow Line W BSP-TRIP            49 (% RDF)            Drive Flow Intercept. Constant Flow for Trip Line for Trip      (see Note 1 below)
Flow Breakpoint              W BSP-BREAK          30.0 (% RDF)            Flow Breakpoint Value Note 1: WBSP-TRIP can be set to 49.0 % RDF or any higher value up to the intersection of the ABSP sloped line with the APRM Flow Biased STP scram line.
Browns Ferry Unit 2 Cycle 21                                                                                              Page 33 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C21, Revision 1 (Final)
 
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[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 5.2 Analyzed MBSP Endpoints: Nominal Feedwater Temperature Power      Core Flow Endpoint                                                Definition
(% Rated) (% Rated)
Scram Region (Region I)
A1          75.9        52.7          Boundary Intercept on MELLLA+ Line Scram Region (Region I)
B1          35.5          29          Boundary Intercept on Natural Circulation Line (NCL)
Controlled Entry Region (Region A2          66.1          52          II) Boundary Intercept on MELLLA Line Controlled Entry Region (Region B2          25.5          29          II) Boundary Intercept on Natural Circulation Line (NCL)
Table 5.3 Analyzed MBSP Endpoints: Reduced Feedwater Temperature Power      Core Flow Endpoint                                                Definition
(% Rated) (% Rated)
Scram Region (Region I)
A1          64.9        50.5          Boundary Intercept on MELLLA Line Scram Region (Region I)
B1          29.4          29          Boundary Intercept on Natural Circulation Line (NCL)
Controlled Entry Region (Region A2          68.3        54.9          II) Boundary Intercept on MELLLA Line Controlled Entry Region (Region B2          24.5          29          II) Boundary Intercept on Natural Circulation Line (NCL)
Browns Ferry Unit 2 Cycle 21                                                                                          Page 34 Core Operating Limits Report, (120% OLTP, MELLLA+)                                            TVA-COLR-BF2C21, Revision 1 (Final)
 
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[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 6 APRM Flow Biased Rod Block Trip Settings (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
The APRM rod block trip setting is based upon References 27 & 29, and is defined by the following:
for two loop operation:
SRB d (0.61Wd + 63.3)                              Allowable Value SRB d (0.61Wd + 62.0)                              Nominal Trip Setpoint (NTSP) where:
SRB          =          Rod Block setting in percent of rated thermal power (3952 MWt)
Wd          =          Recirculation drive flow rate in percent of rated (100% drive flow required to achieve 100% core power and flow) and for single loop operation:
SRB d (0.55(Wd-'W) + 60.5)                                    Allowable Value SRB d (0.55(Wd-'W) + 58.5)                                    Nominal Trip Setpoint (NTSP) where:
SRB          =          Rod Block setting in percent of rated thermal power (3952 MWt)
Wd          =          Recirculation drive flow rate in percent of rated (100% drive flow required to achieve 100% core power and flow)
          'W          =          Difference between two-loop and single-loop effective recirculation flow at the same core flow ('W=0.0 for two-loop operation)
The APRM rod block trip setting is clamped at a maximum allowable value of 115%
(corresponding to a NTSP of 113%).
Browns Ferry Unit 2 Cycle 21                                                                                        Page 35 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
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[I!D                              1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 7 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
The RBM trip setpoints and applicable power ranges, based on References 27 & 28, are shown in Table 7.1. Setpoints are based on an HTSP, unfiltered analytical limit of 114%. Unfiltered setpoints are consistent with a nominal RBM filter setting of 0.0 seconds; filtered setpoints are consistent with a nominal RBM filter setting less than 0.5 seconds. Cycle specific CRWE analyses of OLMCPR are documented in Reference 1, superseding values reported in References 27, 28, and 29.
Table 7.1 Analytical RBM Trip Setpoints
* Allowable              Nominal Trip RBM                              Value                Setpoint Trip Setpoint                          (AV)                  (NTSP)
LPSP                                    27%                    25%
IPSP                                    62%                    60%
HPSP                                    82%                    80%
LTSP - unfiltered                    121.7%                  120.0%
                                          - filtered                    120.7%                  119.0%
ITSP    - unfiltered                  116.7%                  115.0%
                                          - filtered                    115.7%                  114.0%
HTSP - unfiltered                    111.7%                  110.0%
                                          - filtered                    110.9%                  109.2%
DTSP                                    90%                    92%
As a result of cycle specific CRWE analyses, RBM setpoints in Technical Specification Table 3.3.2.1-1 are applicable as shown in Table 7.2. Cycle specific setpoint analysis results are shown in Table 7.3, per Reference 1.
Table 7.2 RBM Setpoint Applicability Thermal Power                      Applicable                Notes from
(% Rated)                        MCPR                Table 3.3.2.1-1              Comment
                                                        < 1.66              (a), (b), (f), (h)    two loop operation 27% and < 90%
                                                        < 1.70              (a), (b), (f), (h)    single loop operation 90%                                      < 1.36                      (g)            two loop operation
* Values are considered maximums. Using lower values, due to RBM system hardware/software limitations, is conservative, and acceptable.
MCPR values shown correspond with, (support), SLMPCR values identified in Reference 1.
Greater than 90% rated power is not attainable in single loop operation.
Browns Ferry Unit 2 Cycle 21                                                                                                  Page 36 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                  TVA-COLR-BF2C21, Revision 1 (Final)
 
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[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Table 7.3 Control Rod Withdrawal Error Results RBM                          CRWE HTSP Analytical Limit                OLMCPR Unfiltered 107                          1.29 111                          1.32 114                          1.33 117                          1.34 Results, compared against the base case OLMCPR results of Table 4.2, indicate SLMCPR remains protected for RBM inoperable conditions (i.e., 114% unblocked).
Browns Ferry Unit 2 Cycle 21                                                                                        Page 37 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
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[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 8 Shutdown Margin Limit (Technical Specification 3.1.1)
Assuming the strongest OPERABLE control blade is fully withdrawn, and all other OPERABLE control blades are fully inserted, the core shall be sub-critical and meet the following minimum shutdown margin:
SDM        0.38% dk/k Browns Ferry Unit 2 Cycle 21                                                                                        Page 38 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                            1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Appendix A: MBSP Maps Browns Ferry Unit 2 Cycle 21                                                                                      Page A-1 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800
                ~
Reactor Engineering and Fuels - BWRFE Date: April 7, 2020 1101 Market Street, Chattanooga TN 37402 Core Power (% Rated: 100% = 3952MWt) 110 100 MELLLA+ Region 90 80                                                                                    BSP Boundary Manual 70                                  Scram Region I 60 MELLLA Region        ICF Region MELLLA Upper Boundary 50 87.5% Rod Line 40                                                        Controlled Entry Region II 30                                                              Min. Flow Control 20 Natural                                  Min. Pow er Line Circulation 10                                          20% Pump Speed Line 0
0      10        20          30        40        50      60        70      80        90      100      110      120 Core Flow (% Rated: 100% =102.5 MLbm/hr)
Figure A.1 MBSP Boundaries For Nominal Feedwater Temperature (Operation in the MELLLA+ Region Prohibited for Feedwater Temperature greater than 10 degrees F below the Nominal Feedwater Temperature)
Browns Ferry Unit 2 Cycle 21                                                                                                Page A-2 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C21, Revision 1 (Final)
 
EDMS: L94 200405 800 Reactor Engineering and Fuels - BWRFE
[I!D                              1101 Market Street, Chattanooga TN 37402 Date: April 7, 2020 Core Power (% Rated: 100% = 3952MWt) 110 100 MELLLA+ Region 90 80 BSP Boundary 70 Manual Scram 60                                    Region I MELLLA Region MELLLA Upper Boundary                                            ICF Region 50 87.5% Rod Line Controlled Entry 40                                                                  Region II 30                                                              Min. Flow Control 20 Natural                              Min. Pow er Line Circulation 10                                            20% Pump Speed Line 0
0          10          20        30        40        50      60        70      80      90      100    110        120 Core Flow (% Rated: 100% =102.5 MLbm/hr)
Figure A.2 MBSP Boundaries For Reduced Feedwater Temperature (Operation in the MELLLA+ Region Prohibited for a Reduced Feedwater Temperature greater than 10 degrees F below the Nominal Feedwater Temperature)
Browns Ferry Unit 2 Cycle 21                                                                                            Page A-3 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF2C21, Revision 1 (Final)
 
ECM L94 210104 800 QA Record nm BFE-4581, Revision 0 Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga, TN 37402 Browns Ferry Unit 2 Cycle 22
  . Core Operating Limits Report, (120% OLTP, MELLLA+)
TVA-COLR-BF2C22                        Revision O (Flnal)
(Revision Log, Page v)
February 2021 Weidman, Adam~i              *DlgltaUyslgnedbyWeldman, dam John-Earl Prepared:
John-Earl              _,/ --  ti2021.02.os10:11:14-os*oo*
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                                                                                    * -Bate: 2021,02.05    11:49:14-05'00' Verified:
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                                                                              ,.,f                      Date:        _ _ _ _ __
W. K. Verner-Dingle, Engineer Dlghally signed by Seifert. Christine A ON: ck=gov, dcatva, dc=maln, ou=Maln. ou.:::Corpoflte, ou:at.lsffl, cnaSelfert. Christine A. emallzzcaselrmotve.gov 119son: I 11m approving this document Date: 2021.02.0512'44:21 -OS'OO' Approved:                                                                            Date: _ _ _ _ __
C. A. Setter, Manager, BWRFE Reviewed:      ~-                                                                    Date:            '2 IS/ 2. (
D.D.offey:Manager, Reactor Engineering Approved:      a., V-:,hr?
Chairman, PORC d&"'o/-                                          Date:          11>1-/1'/I.I Approved: .5YMRR          *~ ~ : _                                                                    
Plant Ma.niger BFN - UNIT 2                                      Page 1 of 50                                                          February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table of Contents Total Number of Pages = 50 (including review cover sheet)
List of Tables ............................................................................................................................. iii List of Figures ............................................................................................................................ iv Revision Log ...............................................................................................................................v Nomenclature ............................................................................................................................ vi References .............................................................................................................................. viii 1      Introduction ........................................................................................................................1 1.1        Purpose .......................................................................................................................1 1.2        Scope ..........................................................................................................................1 1.3        Fuel Loading................................................................................................................1 1.4        Acceptability ................................................................................................................2 2      APLHGR Limits ..................................................................................................................3 2.1        Rated Power and Flow Limit: APLHGRRATED ...............................................................3 2.2        Off-Rated Power Dependent Limit: APLHGRP ............................................................3 2.2.1            Startup without Feedwater Heaters..................................................................... 3 2.3        Off-Rated Flow Dependent Limit: APLHGRF ...............................................................3 2.4        Single Loop Operation Limit: APLHGRSLO ...................................................................3 2.5        Equipment Out-Of-Service Corrections........................................................................6 3      LHGR Limits.......................................................................................................................7 3.1        Rated Power and Flow Limit: LHGRRATED....................................................................7 3.2        Off-Rated Power Dependent Limit: LHGRP .................................................................7 3.2.1            Startup without Feedwater Heaters..................................................................... 7 3.3        Off-Rated Flow Dependent Limit: LHGRF....................................................................8 3.4        Equipment Out-Of-Service Corrections........................................................................8 4      OLMCPR Limits................................................................................................................19 4.1        Flow Dependent MCPR Limit: MCPRF ......................................................................19 4.2        Power Dependent MCPR Limit: MCPRP ...................................................................19 4.2.1            Startup without Feedwater Heaters....................................................................19 4.2.2            Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS) ................................20 4.2.3            Exposure Dependent Limits ...............................................................................20 4.2.4            Equipment Out-Of-Service (EOOS) Options ......................................................21 4.2.5            Single-Loop-Operation (SLO) Limits ..................................................................21 4.2.6            Below Pbypass Limits ........................................................................................21 5      Thermal-Hydraulic Stability Protection..............................................................................33 6      APRM Flow Biased Rod Block Trip Settings.....................................................................35 7      Rod Block Monitor (RBM) Trip Setpoints and Operability .................................................36 8      Shutdown Margin Limit .....................................................................................................38 Appendix A:            MBSP Maps..................................................................................................... A-1 Browns Ferry Unit 2 Cycle 22                                                                                                          Page ii Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE
[I!lil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 List of Tables Nuclear Fuel Types ................................................................................................................ 2 Startup Feedwater Temperature Basis....................................................................................... 7 Nominal Scram Time Basis .......................................................................................................20 MCPRP Limits for All Fuel Types: Optimum Scram Time Basis .............................................23 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis ...............................................24 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis ........................26 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Nominal Scram Time Basis ...................................................................................................28 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Nominal Scram Time Basis ...................................................................................................29 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Technical Specification Scram Time Basis ............................................................................30 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Technical Specification Scram Time Basis ............................................................................31 MCPRP Limits for All Fuel Types: Single Loop Operation for All Scram Times ......................32 ABSP Setpoints for the Scram Region ......................................................................................33 Analyzed MBSP Endpoints: Nominal Feedwater Temperature.................................................34 Analyzed MBSP Endpoints: Reduced Feedwater Temperature ...............................................34 Analytical RBM Trip Setpoints ...............................................................................................36 RBM Setpoint Applicability ........................................................................................................36 Control Rod Withdrawal Error Results.......................................................................................37 Browns Ferry Unit 2 Cycle 22                                                                                                Page iii Core Operating Limits Report, (120% OLTP, MELLLA+)                                            TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE
[I!lil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 List of Figures APLHGRRATED for ATRIUM-10XM Fuel....................................................................................... 4 APLHGRRATED for ATRIUM-11 Fuel ............................................................................................ 5 LHGRRATED for ATRIUM-10XM Fuel ........................................................................................... 9 LHGRRATED for ATRIUM-11 Fuel................................................................................................10 Base Operation LHGRFACP for ATRIUM-10XM Fuel................................................................11 Base Operation LHGRFACP for ATRIUM-11 Fuel .....................................................................12 LHGRFACF for ATRIUM-10XM Fuel .........................................................................................13 LHGRFACF for ATRIUM-11 Fuel...............................................................................................14 Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 1 ........15 Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 2 ........16 Startup Operation LHGRFACP for ATRIUM-11 Fuel: Table 3.1 Temperature Range 1..............17 Startup Operation LHGRFACP for ATRIUM-11 Fuel: Table 3.1 Temperature Range 2..............18 MCPRF for All Fuel Types .........................................................................................................22 MBSP Boundaries For Nominal Feedwater Temperature........................................................ A-2 MBSP Boundaries For Reduced Feedwater Temperature ...................................................... A-3 Browns Ferry Unit 2 Cycle 22                                                                                                Page iv Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Revision Log Number            Page                                          Description 0-R0            All        New document.
Browns Ferry Unit 2 Cycle 22                                                                                        Page v Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE
[I!lil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Nomenclature ABSP                  Automatic Backup Stability Protection APLHGR                Average Planar LHGR APRM                  Average Power Range Monitor AREVA NP              Vendor (Framatome, Siemens)
BOC                    Beginning of Cycle BSP                    Backup Stability Protection BWR                    Boiling Water Reactor CAVEX                  Core Average Exposure CD                    Coast Down CMSS                  Core Monitoring System Software COLR                  Core Operating Limits Report CPR                    Critical Power Ratio CRWE                  Control Rod Withdrawal Error CSDM                  Cold SDM DIVOM                  Delta CPR over Initial CPR vs. Oscillation Magnitude DSS-CD                Detect and Suppress Solution - Confirmation Density EOC                    End of Cycle EOCLB                  End-of-Cycle Licensing Basis EOOS                  Equipment OOS EPU                    Extended Power Uprate (120% OLTP)
FFTR                  Final Feedwater Temperature Reduction FFWTR                  Final Feedwater Temperature Reduction FHOOS                  Feedwater Heaters OOS ft                    Foot: English unit of measure for length GNF                    Vendor (General Electric, Global Nuclear Fuels)
GWd                    Giga Watt Day HTSP                  High TSP ICA                    Interim Corrective Action ICF                    Increased Core Flow (beyond rated)
IS                    In-Service kW                    kilo watt: SI unit of measure for power.
LCO                    License Condition of Operation LFWH                  Loss of Feedwater Heating LHGRFAC                LHGR Multiplier (Power or Flow dependent)
LPRM                  Low Power Range Monitor LRNB                  Generator Load Reject, No Bypass MAPFAC                MAPLHGR multiplier (Power or Flow dependent)
Browns Ferry Unit 2 Cycle 22                                                                                      Page vi Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE
[I!lil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 MBSP                  Manual Backup Stability Protection MCPR                  Minimum CPR MELLLA                Maximum Extended Load Line Limit Analysis MELLLA+                Maximum Extended Load Line Limit Analysis Plus MSRV                  Moisture Separator Reheater Valve MSRVOOS                MSRV OOS MTU                    Metric Ton Uranium MWd/MTU                Mega Watt Day per Metric Ton Uranium NEOC                  Near EOC NRC                    United States Nuclear Regulatory Commission NSS                    Nominal Scram Speed NTSP                  Nominal TSP OLMCPR                MCPR Operating Limit OLTP                  Original Licensed Thermal Power OOS                    Out-Of-Service OPRM                  Oscillation Power Range Monitor OSS                    Optimum Scram Speed PBDA                  Period Based Detection Algorithm Pbypass                Power, below which TSV Position and TCV Fast Closure Scrams are Bypassed PLU                    Power Load Unbalance PLUOOS                PLU OOS PRNM                  Power Range Neutron Monitor RBM                    Rod Block Monitor RCPOOS                Recirculation Pump OOS (SLO)
RDF                    Rated Drive Flow RPS                    Reactor Protection System RPT                    Recirculation Pump Trip RPTOOS                RPT OOS RTP                    Rated Thermal Power SDM                    Shutdown Margin SLMCPR                MCPR Safety Limit SLO                    Single Loop Operation TBV                    Turbine Bypass Valve TBVIS                  TBV IS TBVOOS                Turbine Bypass Valves OOS TIP                    Transversing In-core Probe TIPOOS                TIP OOS TLO                    Two Loop Operation TSP                    Trip Setpoint TSSS                  Technical Specification Scram Speed TVA                    Tennessee Valley Authority Browns Ferry Unit 2 Cycle 22                                                                                      Page vii Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 References
: 1.        ANP-3880, Revision 0, Browns Ferry Unit 2 Cycle 22 Reload Analysis, Framatome, Inc., November 2020.
: 2.        Not Used.
: 3.        ANP-3150P, Revision 4, Mechanical Design Report for Browns Ferry ATRIUM 10XM Fuel Assemblies, AREVA Inc., November 2017.
: 4.        ANP-3855P, Revision 0, Browns Ferry Unit 2 Cycle 22 Plant Parameters Document, Framatome Inc., May 2020.
: 5.        BFE-4587, Revision 0, Browns Ferry Unit 2 Cycle 22 In Core Shuffle, Calculation File, Tennessee Valley Authority, February 2021.
Methodology References
: 6.        XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
: 7.        XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
: 8.        EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
: 9.        ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
: 10.      XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.
: 11.      XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.
: 12.      EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.
: 13.      XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.
: 14.      XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.
: 15.      ANP-10307PA, Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP Inc., June 2011.
Browns Ferry Unit 2 Cycle 22                                                                                      Page viii Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021
: 16.      ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.
: 17.      ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Advanced Nuclear Fuels Corporation, September 2005.
: 18.      EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP Inc.,
September 2009.
: 19.      EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001, as supplemented by the site specific approval in NRC safety evaluation dated February 15, 2013 and July 31, 2014.
: 20.      EMF-2292(P)(A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
: 21.      EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
: 22.      BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP Inc., May 2008.
: 23.      BAW-10247PA, Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP Inc., April 2008.
: 24.      ANP-10298PA, Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc., March 2010.
: 25.      ANP-3140(P), Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc.,
August 2012.
: 26.      NEDC-33075P-A, Revision 8, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, GE Hitachi, November 2013.
Setpoint References
: 27.      EDQ2092900118, R37, Setpoint and Scaling Calculation for Neutron Monitoring &
Recirculation Flow Loops, Calculation File, Tennessee Valley Authority, December 4, 2020.
: 28.      Task T0500, Revision 0, Neutron Monitoring System w/RBM, Project Task Report, GE Hitachi Nuclear Energy, June 2017.
: 29.      Task T0506, Revision 0, TS Instrument Setpoints, Project Task Report, Tennessee Valley Authority, August, 2017.
: 30.      NEDC-33006P-A, Revision 3, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, GE Energy Nuclear, June 2009.
Browns Ferry Unit 2 Cycle 22                                                                                      Page ix Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE
[I!lil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 1 Introduction In anticipation of cycle startup, it is necessary to describe the expected limits of operation.
1.1    Purpose The primary purpose of this document is to satisfy requirements identified by unit technical specification section 5.6.5. This document may be provided, upon final approval, to the NRC.
1.2    Scope This document will discuss the following areas:
3/4 Average Planar Linear Heat Generation Rate (APLHGR) Limit (Technical Specifications 3.2.1 and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3.2.3, 3.3.4.1, and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Minimum Critical Power Ratio Operating Limit (OLMCPR)
(Technical Specifications 3.2.2, 3.3.4.1, 3.7.5 and Table 3.3.2.1-1)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Thermal-Hydraulic Stability Protection (Technical Specification Table 3.3.1.1)
Applicability: Mode 1,  (as specified in Technical Specifications Table 3.3.1.1-1) 3/4 Average Power Range Monitor (APRM) Flow Biased Rod Block Trip Setting (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
Applicability: Mode 1,  (as specified in Technical Requirements Manuals Table 3.3.4-1) 3/4 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
Applicability: Mode 1,  % RTP as specified in Table 3.3.2.1-1 (TS definition of RTP) 3/4 Shutdown Margin (SDM) Limit (Technical Specification 3.1.1)
Applicability: All Modes 1.3    Fuel Loading The core will contain fresh and exposed ATRIUM-10XM, as well as a limited number of ATRIUM-11 lead fuel assemblies. Nuclear fuel types used in the core loading are shown in Table 1.1. The core shuffle and final loading were explicitly evaluated for BOC cold shutdown margin performance as documented per Reference 5.
Browns Ferry Unit 2 Cycle 22                                                                                        Page 1 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                              1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 1.1 Nuclear Fuel Types
* Original          Number of Nuclear Fuel                    Fuel Names Fuel Description Cycle          Assemblies Type (NFT)                        (Range)
ATRIUM 11 A11-3693B-13GV80-FBF                                        19                  4                17              FBF653-FBF660 ATRIUM 10XM XMLC-4102B-11GV70-FBG-B                                  20                  47                18              FBG701-FBG748 ATRIUM 10XM XMLC-4062B-13GV80-FBG-C                                  20                  77                19              FBG749-FBG900 ATRIUM 10XM XMLC-3948B-13GV70-FBG-B                                  20                  8                20              FBG901-FBG988 ATRIUM-10XM XMLC-4087B-15GV80-FBH                                    21                176                21              FBH001-FBH176 ATRIUM-10XM XMLC-4036B-15GV80-FBH                                    21                  88                22              FBH177-FBH264 ATRIUM-10XM XMLC-4093B-10GV80-FBH                                    21                  56                23              FBH265-FBH320 ATRIUM-10XM XMLC-4058B-15GV80-FBJ                                    22                216                24              FBJ331-FBJ546 ATRIUM-10XM XMLC-4015B-15GV80-FBJ                                    22                  92                25              FBJ547-FBJ638 1.4      Acceptability Limits discussed in this document were generated based on NRC approved methodologies per References 6 through 26.
* The table identifies the expected fuel type breakdown in anticipation of final core loading. The final composition of the core depends upon uncertainties during the outage such as discovering a failed fuel bundle, or other bundle damage. Minor core loading changes, due to unforeseen events, will conform to the safety and monitoring requirements identified in this document.
Browns Ferry Unit 2 Cycle 22                                                                                                            Page 2 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                          TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 2 APLHGR Limits (Technical Specifications 3.2.1 & 3.7.5)
The APLHGR limit is determined by adjusting the rated power APLHGR limit for off-rated power, off-rated flow, and SLO conditions. The most limiting of these is then used as follows:
APLHGR limit = MIN ( APLHGRP , APLHGRF, APLHGRSLO )
where:
APLHGRP                off-rated power APLHGR limit                  [APLHGRRATED
* MAPFACP]
APLHGRF                off-rated flow APLHGR limit                  [APLHGRRATED
* MAPFACF]
APLHGRSLO              SLO APLHGR limit                              [APLHGRRATED
* SLO Multiplier]
2.1    Rated Power and Flow Limit: APLHGRRATED The rated conditions APLHGR for all fuel are identified per Reference 1. The rated conditions APLHGR for ATRIUM-10XM fuel are shown in Figure 2.1. The rated conditions APLHGR for ATRIUM-11 are shown in Figure 2.2.
2.2    Off-Rated Power Dependent Limit: APLHGRP Reference 1 does not specify a power dependent APLHGR. Therefore, MAPFACP is set to a value of 1.0.
2.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. No additional power dependent limitation is required.
2.3    Off-Rated Flow Dependent Limit: APLHGRF Reference 1 does not specify a flow dependent APLHGR. Therefore, MAPFACF is set to a value of 1.0.
2.4    Single Loop Operation Limit: APLHGRSLO The single loop operation multiplier for ATRIUM-10XM and ATRIUM-11 fuel is 0.85, per Reference 1.
Browns Ferry Unit 2 Cycle 22                                                                                        Page 3 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                  1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 15 12 APLHGR (kW/ft) 9 6
3 0
0          20                    40                      60                    80 Planar Average Exposure (GWd/MTU)
Planar Avg.        APLHGR Exposure            Limit (GWd/MTU)          (kW/ft) 0.0              13.0 15.0              13.0 67.0              7.6 Figure 2.1 APLHGRRATED for ATRIUM-10XM Fuel Browns Ferry Unit 2 Cycle 22                                                                                          Page 4 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                  1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 15 12 APLHGR (kW/ft) 9 6
3 0
0          20                    40                      60                    80 Planar Average Exposure (GWd/MTU)
Planar Avg.        APLHGR Exposure            Limit (GWd/MTU)          (kW/ft) 0.0                10 10.0                10 67.0              5.9 Figure 2.2 APLHGRRATED for ATRIUM-11 Fuel Browns Ferry Unit 2 Cycle 22                                                                                          Page 5 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 2.5      Equipment Out-Of-Service Corrections The limits shown in Figure 2.1 and Figure 2.2 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
In-Service                              All equipment In-Service
* RPTOOS                                  EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                  Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                  Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                        Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                  One Recirculation Pump Out-Of-Service
* All equipment service conditions assume 1 SRVOOS.
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 3 LHGR Limits (Technical Specification 3.2.3, 3.3.4.1, & 3.7.5)
The LHGR limit is determined by adjusting the rated power LHGR limit for off-rated power and off-rated flow conditions. The most limiting of these is then used as follows:
LHGR limit = MIN ( LHGRP, LHGRF )
where:
LHGRP                  off-rated power LHGR limit                    [LHGRRATED
* LHGRFACP]
LHGRF                  off-rated flow LHGR limit                    [LHGRRATED
* LHGRFACF]
3.1    Rated Power and Flow Limit: LHGRRATED The rated conditions LHGR for all fuel are identified per Reference 1. The rated conditions LHGR for ATRIUM-10XM are shown in Figure 3.1. The rated conditions LHGR for ATRIUM-11 fuel is shown in Figure 3.2. The LHGR limit is consistent with Reference 3.
3.2    Off-Rated Power Dependent Limit: LHGRP LHGR limits are adjusted for off-rated power conditions using the LHGRFACP multiplier provided in Reference 1. The multiplier is split into two sub cases: turbine bypass valves in and out-of-service. The base case multipliers are shown in Figure 3.3 and Figure 3.4.
3.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional limits are shown in Figure 3.7 through Figure 3.10, based on temperature conditions identified in Table 3.1.
Table 3.1 Startup Feedwater Temperature Basis Temperature Power              Range 1                Range 2
(% Rated)                (&deg;F)                  (&deg;F) 23                155.0                  150.0 30                162.0                  157.0 40                172.0                  167.0 50                182.0                  177.0 Browns Ferry Unit 2 Cycle 22                                                                                        Page 7 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 3.3      Off-Rated Flow Dependent Limit: LHGRF LHGR limits are adjusted for off-rated flow conditions using the LHGRFACF multiplier provided in Reference 1. Multipliers are shown in Figure 3.5 through Figure 3.6.
3.4      Equipment Out-Of-Service Corrections The limits shown in Figure 3.1 and Figure 3.2 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
* In-Service                              All equipment In-Service RPTOOS                                  EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                  Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                  Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                        Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                  One Recirculation Pump Out-Of-Service Off-rated power corrections shown in Figure 3.3 and Figure 3.4 are dependent on operation of the Turbine Bypass Valve system. For this reason, separate limits are to be applied for TBVIS or TBVOOS operation. The limits have no dependency on RPTOOS, PLUOOS, FHOOS/FFWTR, or SLO.
Off-rated flow corrections shown in Figure 3.5 and Figure 3.6 are bounding for all EOOS conditions.
Off-rated power corrections shown in Figure 3.7 through Figure 3.10 are also dependent on operation of the Turbine Bypass Valve system. In this case, limits support FHOOS operation during startup. These limits have no dependency on RPTOOS, PLUOOS, or SLO.
* All equipment service conditions assume 1 SRVOOS.
Browns Ferry Unit 2 Cycle 22                                                                                        Page 8 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
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LHGR (kW/ft) 6 3
0 0            20                    40                      60                    80 Pellet Exposure (GWd/MTU)
Pellet          LHGR Exposure          Limit (GWd/MTU)          (kW/ft) 0.0              14.1 18.9              14.1 74.4              7.4 Figure 3.1 LHGRRATED for ATRIUM-10XM Fuel Browns Ferry Unit 2 Cycle 22                                                                                          Page 9 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                    1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 15 12 9
LHGR (kW/ft) 6 3
0 0            20                    40                      60                    80 Pellet Exposure (GWd/MTU)
Pellet          LHGR Exposure          Limit (GWd/MTU)          (kW/ft) 0.0              12.2 18.9              12.2 74.4              6.4 Figure 3.2 LHGRRATED for ATRIUM-11 Fuel Browns Ferry Unit 2 Cycle 22                                                                                        Page 10 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 1.10 1.00 0.90 Turbine Bypass Valve In-Service, TBVIS 0.80 LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 7%9,6&RUH)ORZ 0.50 7%9226&RUH)ORZ TBVIS, > 50% Core Flow 0.40          TBVOOS, > 50% Core Flow 0.30 0.20 20      30          40          50        60        70        80            90        100          110 Core Power (% Rated)
Turbine Bypass In-Service                Turbine Bypass Out-of-Service Core                                      Core Power            LHGRFACP                  Power            LHGRFACP
(% Rated)                                  (% Rated) 100.0          1.00                        100.0          1.00 26.0          0.63                          26.0          0.62 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0          0.47                          26.0          0.39 23.0          0.43                          23.0          0.36
                            &RUH)ORZ5DWHG                      &RUH)ORZ5DWHG 26.0          0.51                          26.0          0.50 23.0          0.49                          23.0          0.44 Figure 3.3 Base Operation LHGRFACP for ATRIUM-10XM Fuel (Independent of other EOOS conditions)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 1.10 1.00 0.90 Turbine Bypass Valve In-Service, TBVIS 0.80 LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 0.50          7%9,6&RUH)ORZ 7%9226&RUH)ORZ TBVIS, > 50% Core Flow 0.40          TBVOOS, > 50% Core Flow 0.30 0.20 20      30          40          50        60        70        80          90          100        110 Core Power (% Rated)
Turbine Bypass In-Service                Turbine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                  Power            LHGRFACP
(% Rated)                                    (% Rated) 100.0          1.00                          100.0          1.00 26.0          0.61                          26.0          0.60 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0          0.44                          26.0          0.38 23.0          0.42                          23.0          0.34
                          &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0          0.48                          26.0          0.48 23.0          0.45                          23.0          0.43 Figure 3.4 Base Operation LHGRFACP for ATRIUM-11 Fuel (Independent of other EOOS conditions)
Browns Ferry Unit 2 Cycle 22                                                                                        Page 12 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 1.10 1.05 1.00 0.95 0.90 LHGRFACF 0.85 0.80 0.75 0.70 0.65 0.60 25        35          45          55        65          75            85          95          105 Core Flow (% Rated)
Core Flow        LHGRFACF
(% Rated) 0.0            0.63 30.0            0.63 76.4            1.00 107.0            1.00 Figure 3.5 LHGRFACF for ATRIUM-10XM Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 2 Cycle 22                                                                                            Page 13 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 1.10 1.05 1.00 0.95 0.90 LHGRFACF 0.85 0.80 0.75 0.70 0.65 0.60 25        35          45          55        65          75            85          95          105 Core Flow (% Rated)
Core Flow        LHGRFACF
(% Rated) 0.0            0.63 30.0            0.63 76.4            1.00 107.0            1.00 Figure 3.6 LHGRFACF for ATRIUM-11 Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 1.10 1.00 0.90 0.80            Turbine Bypass Valve In-Service, TBVIS LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 0.50 7%9,6&RUH)ORZ 7%9226&RUH)ORZ 0.40            TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20        30          40          50        60        70        80        90          100        110 Core Power (% Rated)
Turbine Bypass In-Service                Turbine Bypass Out-of-Service Core                                        Core Power          LHGRFACP                    Power            LHGRFACP
(% Rated)                                  (% Rated) 100.0          1.00                        100.0          1.00 26.0          0.54                        26.0          0.53 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0          0.41                        26.0          0.36 23.0          0.38                        23.0          0.32
                              &RUH)ORZ5DWHG                      &RUH)ORZ5DWHG 26.0          0.45                        26.0          0.42 23.0          0.44                        23.0          0.40 Figure 3.7 Startup Operation LHGRFACP for ATRIUM-10XM Fuel:
Table 3.1 Temperature Range 1 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 1.10 1.00 0.90 0.80            Turbine Bypass Valve In-Service, TBVIS LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 0.50 7%9,6&RUH)ORZ 7%9226&RUH)ORZ 0.40          TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20      30          40          50        60        70        80        90        100          110 Core Power (% Rated)
Turbine Bypass In-Service                  Turbine Bypass Out-of-Service Core                                        Core Power          LHGRFACP                    Power            LHGRFACP
(% Rated)                                  (% Rated) 100.0          1.00                          100.0          1.00 26.0          0.54                          26.0          0.53 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0          0.40                          26.0          0.35 23.0          0.38                          23.0          0.32
                            &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0          0.45                          26.0          0.42 23.0          0.43                          23.0          0.40 Figure 3.8 Startup Operation LHGRFACP for ATRIUM-10XM Fuel:
Table 3.1 Temperature Range 2 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 1.10 1.00 0.90 0.80            Turbine Bypass Valve In-Service, TBVIS LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 0.50 7%9,6&RUH)ORZ 0.40        7%9226&RUH)ORZ TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20      30          40          50        60        70        80        90        100          110 Core Power (% Rated)
Turbine Bypass In-Service                  Turbine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                    Power            LHGRFACP
(% Rated)                                    (% Rated) 100.0          1.00                          100.0          1.00 26.0          0.51                            26.0          0.51 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0          0.39                            26.0          0.34 23.0          0.36                            23.0          0.31
                          &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0          0.41                            26.0          0.41 23.0          0.41                            23.0          0.38 Figure 3.9 Startup Operation LHGRFACP for ATRIUM-11 Fuel:
Table 3.1 Temperature Range 1 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 1.10 1.00 0.90 0.80          Turbine Bypass Valve In-Service, TBVIS LHGRFACP 0.70 Turbine Bypass Valve Out-of-Service, TBVOOS 0.60 0.50 7%9,6&RUH)ORZ 0.40          7%9226&RUH)ORZ TBVIS, > 50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20      30        40          50        60        70        80        90          100        110 Core Power (% Rated)
Turbine Bypass In-Service                Turbine Bypass Out-of-Service Core                                      Core Power          LHGRFACP                  Power            LHGRFACP
(% Rated)                                  (% Rated) 100.0          1.00                        100.0          1.00 26.0          0.51                          26.0          0.51 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0          0.39                          26.0          0.34 23.0          0.35                          23.0          0.30
                            &RUH)ORZ5DWHG                      &RUH)ORZ5DWHG 26.0          0.41                          26.0          0.41 23.0          0.40                          23.0          0.38 Figure 3.10 Startup Operation LHGRFACP for ATRIUM-11 Fuel:
Table 3.1 Temperature Range 2 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 4 OLMCPR Limits (Technical Specification 3.2.2, 3.3.4.1, & 3.7.5)
OLMCPR is calculated to be the most limiting of the flow or power dependent values OLMCPR limit = MAX ( MCPRF , MCPRP )
where:
MCPRF              core flow-dependent MCPR limit MCPRP              power-dependent MCPR limit 4.1    Flow Dependent MCPR Limit: MCPRF MCPRF limits are dependent upon core flow (% of Rated), and the max core flow limit, (Rated or Increased Core Flow, ICF). MCPRF limits are shown in Figure 4.1, per Reference 1. Limits are valid for all EOOS combinations. No adjustment is required for SLO conditions.
4.2    Power Dependent MCPR Limit: MCPRP MCPRP limits are dependent upon:
Core Power Level (% of Rated)
Technical Specification Scram Speed (TSSS), Nominal Scram Speed (NSS), or Optimum Scram Speed (OSS)
Cycle Operating Exposure (NEOC, EOC, and CD - as defined in this section)
Equipment Out-Of-Service Options Two or Single recirculation Loop Operation (TLO vs. SLO)
The MCPRP limits are provided in Table 4.2 through Table 4.9, where each table contains the limits for all fuel types and EOOS options (for a specified scram speed and exposure range).
The CMSS determines MCPRP limits, from these tables, based on linear interpolation between the specified powers.
4.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional power dependent limits are shown in Table 4.5 through Table 4.8 based on temperature conditions identified in Table 3.1.
Browns Ferry Unit 2 Cycle 22                                                                                      Page 19 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                            1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 4.2.2    Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS)
MCPRP limits are provided for three different sets of assumed scram speeds. The Technical Specification Scram Speed (TSSS) MCPRP limits are applicable at all times, as long as the scram time surveillance demonstrates the times in Technical Specification Table 3.1.4-1 are met. Both Nominal Scram Speeds (NSS) and/or Optimum Scram Speeds (OSS) may be used, as long as the scram time surveillance demonstrates Table 4.1 times are applicable.
* Table 4.1 Nominal Scram Time Basis Notch                    Nominal                Optimum Position              Scram Timing            Scram Timing (index)                  (seconds)                (seconds) 46                        0.420                    0.380 36                        0.980                    0.875 26                        1.600                    1.465 6                        2.900                    2.900 In demonstrating compliance with the NSS and/or OSS scram time basis, surveillance requirements from Technical Specification 3.1.4 apply; accepting the definition of SLOW rods should conform to scram speeds shown in Table 4.1. If conformance is not demonstrated, TSSS based MCPRP limits are applied.
On initial cycle startup, TSSS limits are used until the successful completion of scram timing confirms NSS and/or OSS based limits are applicable.
4.2.3    Exposure Dependent Limits Exposures are tracked on a Core Average Exposure basis (CAVEX, not Cycle Exposure).
Higher exposure MCPRP limits are always more limiting and may be used for any Core Average Exposure up to the ending exposure. Per Reference 1, MCPRP limits are provided for the following exposure ranges:
BOC to NEOC                                NEOC corresponds to                        31,055.6 MWd / MTU BOC to EOCLB                                EOCLB corresponds to                        34,233.1 MWd / MTU BOC to End of Coast                        End of Coast                                35,926.7 MWd / MTU NEOC refers to a Near EOC exposure point.
* Reference 1 analysis results are based on information identified in Reference 4.
Drop out times consistent with method used to perform actual timing measurements (i.e., including pickup/dropout effects).
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                            1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 The EOCLB exposure point is not the true End-Of-Cycle exposure. Instead it corresponds to a licensing exposure window exceeding expected end-of-full-power-life.
The End of Coast exposure point represents a licensing exposure point exceeding the expected end-of-cycle exposure including cycle extension options.
4.2.4      Equipment Out-Of-Service (EOOS) Options EOOS options
* covered by MCPRP limits are given by the following:
In-Service                                          All equipment In-Service RPTOOS                                              EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                              Turbine Bypass Valve(s) Out-Of-Service RPTOOS+TBVOOS                                      Combined RPTOOS and TBVOOS PLUOOS                                              Power Load Unbalance Out-Of-Service PLUOOS+RPTOOS                                      Combined PLUOOS and RPTOOS PLUOOS+TBVOOS                                      Combined PLUOOS and TBVOOS PLUOOS+TBVOOS+RPTOOS                                Combined PLUOOS, RPTOOS, and TBVOOS FHOOS (or FFWTR)                                    Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)
RCPOOS                                              One Recirculation Pump Out-Of-Service For exposure ranges up to NEOC and EOCLB, additional combinations of MCPRP limits are also provided including FHOOS. The coast down exposure range assumes application of FFWTR. FHOOS based MCPRP limits for the coast down exposure are redundant because the temperature setdown assumption is identical with FFWTR.
4.2.5      Single-Loop-Operation (SLO) Limits When operating in RCPOOS conditions, MCPRp limits are constructed differently from the normal operating RCP conditions. The limiting event for RCPOOS is a pump seizure scenario, which sets the upper bound for allowed core power and flow . This event is not impacted by scram time assumptions. Specific MCPRP limits are shown in Table 4.9.
4.2.6      Below Pbypass Limits Below Pbypass (26% rated power), MCPRP limits depend upon core flow. One set of MCPRP limits applies for core flow above 50% of rated; a second set applies if the core flow is less than or equal to 50% rated.
* All equipment service conditions assume 1 SRVOOS.
RCPOOS limits are only valid up to 43.75% rated core power, 50% rated core flow, and an active recirculation drive flow of 17.73 Mlbm/hr.
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                              1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 2.00 1.80 1.60 MCPRF 1.40 1.20 1.00 30        40            50          60          70          80        90          100          110 Core Flow (% Rated)
Core Flow            MCPRF
(% Rated) 30.0              1.58 84.0              1.34 107.0              1.34 Figure 4.1 MCPRF for All Fuel Types (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 4.2 MCPRP Limits for All Fuel Types: Optimum Scram Time Basis
* ATRIUM-10XM                                      ATRIUM-11 BOC            BOC            BOC            BOC            BOC        BOC Power              to              to        to End of          to            to      to End of Operating
(% of rated)      NEOC          EOCLB          Coast          NEOC          EOCLB        Coast Condition 100            1.38          1.42            1.46          1.38          1.43        1.46 90            1.44          1.45            1.51          1.44          1.46        1.51 77.6            1.52          1.53            1.58          1.52          1.53        1.59 65            1.58          1.60            1.66          1.59          1.60        1.66
                                  > 50            1.67          1.68            1.76          1.73          1.73        1.81 50            1.82          1.82            1.84          1.91          1.91        1.91 Base Case 40            1.92          1.92            1.94          1.99          1.99        2.03 26            2.26          2.26            2.39          2.51          2.51        2.66 26 at > 50%F        2.47          2.47            2.59          2.68          2.68        2.81 23 at > 50%F        2.62          2.62            2.76          2.93          2.93        3.05
                              DW)        2.38          2.38            2.49          2.63          2.63        2.72
                              DW)        2.52          2.52            2.65          2.93          2.93        3.05 100            1.43          1.46            ---          1.43          1.46          ---
90            1.49          1.51            ---          1.49          1.51          ---
77.6            1.55          1.58            ---          1.57          1.59          ---
65            1.64          1.66            ---          1.64          1.66          ---
                                  > 50            1.74          1.76            ---          1.81          1.81          ---
50            1.83          1.84            ---          1.91          1.91          ---
FHOOS 40            1.93          1.94            ---          2.03          2.03          ---
26            2.38          2.39            ---          2.66          2.66          ---
26 at > 50%F        2.58          2.59            ---          2.81          2.81          ---
23 at > 50%F        2.75          2.76            ---          3.05          3.05          ---
                              DW)        2.48          2.49            ---          2.72          2.72          ---
                              DW)        2.64          2.65            ---          3.05          3.05          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR/FHOOS is supported for the BOC to End of Coast limits.
Browns Ferry Unit 2 Cycle 22                                                                                                      Page 23 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                              Page 32 of 50                                        February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                            1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM                                ATRIUM-11 BOC          BOC          BOC          BOC          BOC          BOC Power            to          to        to End of        to            to        to End of Operating
(% of rated)      NEOC        EOCLB          Coast        NEOC        EOCLB          Coast Condition 100            1.43        1.45          1.51        1.43          1.45          1.51 90            1.48        1.49          1.56        1.48          1.49          1.56 77.6            1.56        1.56          1.62        1.56          1.56          1.64 65            1.62        1.63          1.69        1.62          1.63          1.72
                                      > 50            1.70        1.71          1.79        1.76          1.76          1.84 50            1.83        1.83          1.86        1.92          1.92          1.92 Base Case 40            1.93        1.93          1.96        2.00          2.00          2.06 26            2.29        2.29          2.43        2.55          2.55          2.69 26 at > 50%F        2.48        2.48          2.61        2.68          2.68          2.81 23 at > 50%F        2.63        2.63          2.78        2.93          2.93          3.05
                                  DW)        2.39        2.39          2.51        2.63          2.63          2.72
                                  DW)        2.53        2.53          2.67        2.93          2.93          3.05 100            1.48        1.52          1.56        1.48          1.52          1.56 90            1.53        1.55          1.61        1.53          1.55          1.61 77.6            1.60        1.62          1.68        1.60          1.62          1.68 65            1.67        1.69          1.75        1.67          1.69          1.75
                                      > 50            1.75        1.77          1.85        1.78          1.78          1.87 50            1.85        1.86          1.89        1.92          1.92          1.92 TBVOOS 40            1.95        1.96          1.99        2.00          2.00          2.06 26            2.32        2.33          2.47        2.55          2.55          2.69 26 at > 50%F        3.04        3.05          3.19        3.28          3.28          3.38 23 at > 50%F        3.28        3.29          3.42        3.57          3.57          3.69
                                  DW)        2.79        2.80          2.94        3.04          3.04          3.16
                                  DW)        3.04        3.05          3.22        3.32          3.32          3.56 100            1.48        1.51            ---        1.50          1.51          ---
90            1.54        1.56            ---        1.55          1.56          ---
77.6            1.60        1.62            ---        1.64          1.64          ---
65            1.68        1.69            ---        1.72          1.72          ---
                                      > 50            1.78        1.79            ---        1.84          1.84          ---
50            1.85        1.86            ---        1.92          1.92          ---
FHOOS 40            1.95        1.96            ---        2.06          2.06          ---
26            2.42        2.43            ---        2.69          2.69          ---
26 at > 50%F        2.60        2.61            ---        2.81          2.81          ---
23 at > 50%F        2.77        2.78            ---        3.05          3.05          ---
                                  DW)        2.50        2.51            ---        2.72          2.72          ---
                                  DW)        2.66        2.67            ---        3.05          3.05          ---
100            1.43        1.45          1.51        1.43          1.45          1.51 90            1.48        1.49          1.56        1.48          1.49          1.56 77.6            1.56        1.56          1.62        1.66          1.66          1.70 65            1.75        1.75          1.78        1.83          1.83          1.83
                                      > 50            ---          ---            ---          ---          ---          ---
50            1.83        1.83          1.86        1.92          1.92          1.92 PLUOOS 40            1.93        1.93          1.96        2.00          2.00          2.06 26            2.29        2.29          2.43        2.55          2.55          2.69 26 at > 50%F        2.48        2.48          2.61        2.68          2.68          2.81 23 at > 50%F        2.63        2.63          2.78        2.93          2.93          3.05
                                  DW)        2.39        2.39          2.51        2.63          2.63          2.72
                                  DW)        2.53        2.53          2.67        2.93          2.93          3.05
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 2 Cycle 22                                                                                                      Page 24 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                                Page 33 of 50                                      February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                            1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis (continued)
* ATRIUM-10XM                                ATRIUM-11 BOC        BOC            BOC          BOC          BOC        BOC Power            to          to        to End of        to            to      to End of Operating
(% of rated)      NEOC        EOCLB          Coast        NEOC        EOCLB        Coast Condition 100            1.54        1.56            ---        1.54          1.56          ---
90            1.59        1.61            ---        1.59          1.61          ---
77.6            1.66        1.68            ---        1.67          1.68          ---
65            1.74        1.75            ---        1.75          1.75          ---
                                      > 50            1.83        1.85            ---        1.87          1.87          ---
TBVOOS                      50            1.88        1.89            ---        1.92          1.92          ---
FHOOS                        40            1.98        1.99            ---        2.06          2.06          ---
26            2.46        2.47            ---        2.69          2.69          ---
26 at > 50%F        3.18        3.19            ---        3.38          3.38          ---
23 at > 50%F        3.41        3.42            ---        3.69          3.69          ---
                                DW)        2.93        2.94            ---        3.16          3.16          ---
                                DW)        3.21        3.22            ---        3.56          3.56          ---
100            1.48        1.52          1.56        1.48          1.52        1.56 90            1.53        1.55          1.61        1.53          1.55        1.61 77.6            1.60        1.62          1.68        1.68          1.69        1.72 65            1.77        1.78          1.81        1.83          1.83        1.83
                                      > 50            ---          ---            ---          ---          ---          ---
TBVOOS                      50            1.85        1.86          1.89        1.92          1.92        1.92 PLUOOS                      40            1.95        1.96          1.99        2.00          2.00        2.06 26            2.32        2.33          2.47        2.55          2.55        2.69 26 at > 50%F        3.04        3.05          3.19        3.28          3.28        3.38 23 at > 50%F        3.28        3.29          3.42        3.57          3.57        3.69
                                DW)        2.79        2.80          2.94        3.04          3.04        3.16
                                DW)        3.04        3.05          3.22        3.32          3.32        3.56 100            1.48        1.51            ---        1.50          1.51          ---
90            1.54        1.56            ---        1.55          1.56          ---
77.6            1.60        1.62            ---        1.69          1.70          ---
65            1.77        1.78            ---        1.83          1.83          ---
                                      > 50            ---          ---            ---          ---          ---          ---
FHOOS                      50            1.85        1.86            ---        1.92          1.92          ---
PLUOOS                      40            1.95        1.96            ---        2.06          2.06          ---
26            2.42        2.43            ---        2.69          2.69          ---
26 at > 50%F        2.60        2.61            ---        2.81          2.81          ---
23 at > 50%F        2.77        2.78            ---        3.05          3.05          ---
                                DW)        2.50        2.51            ---        2.72          2.72          ---
                                DW)        2.66        2.67            ---        3.05          3.05          ---
100            1.54        1.56            ---        1.54          1.56          ---
90            1.59        1.61            ---        1.59          1.61          ---
77.6            1.66        1.68            ---        1.71          1.72          ---
65            1.80        1.81            ---        1.83          1.83          ---
                                      > 50            ---          ---            ---          ---          ---          ---
TBVOOS 50            1.88        1.89            ---        1.92          1.92          ---
FHOOS 40            1.98        1.99            ---        2.06          2.06          ---
PLUOOS 26            2.46        2.47            ---        2.69          2.69          ---
26 at > 50%F        3.18        3.19            ---        3.38          3.38          ---
23 at > 50%F        3.41        3.42            ---        3.69          3.69          ---
                                DW)        2.93        2.94            ---        3.16          3.16          ---
                                DW)        3.21        3.22            ---        3.56          3.56          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 2 Cycle 22                                                                                                    Page 25 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                              Page 34 of 50                                        February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                            1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM                                ATRIUM-11 BOC        BOC            BOC          BOC          BOC        BOC Power            to          to        to End of        to            to      to End of Operating
(% of rated)      NEOC        EOCLB          Coast        NEOC        EOCLB        Coast Condition 100            1.48        1.50          1.55        1.48          1.50        1.56 90            1.53        1.53          1.60        1.53          1.53        1.61 77.6            1.60        1.60          1.66        1.60          1.60        1.70 65            1.66        1.66          1.73        1.67          1.67        1.78
                                      > 50            1.74        1.74          1.83        1.79          1.79        1.90 50            1.86        1.86          1.89        1.93          1.93        1.93 Base Case 40            1.96        1.96          1.99        2.01          2.01        2.08 26            2.33        2.33          2.48        2.57          2.57        2.72 26 at > 50%F        2.50        2.50          2.63        2.68          2.68        2.81 23 at > 50%F        2.65        2.65          2.80        2.93          2.93        3.05
                                DW)        2.41        2.41          2.53        2.63          2.63        2.72
                                DW)        2.55        2.55          2.69        2.93          2.93        3.05 100            1.54        1.57          1.61        1.54          1.57        1.61 90            1.59        1.60          1.65        1.59          1.60        1.65 77.6            1.66        1.66          1.72        1.67          1.67        1.74 65            1.72        1.72          1.79        1.74          1.74        1.82
                                      > 50            1.79        1.80          1.88        1.83          1.83          ---
50            1.89        1.89          1.92        1.93          1.93        1.93 TBVOOS 40            1.99        1.99          2.03        2.01          2.01        2.09 26            2.38        2.38          2.51        2.57          2.57        2.72 26 at > 50%F        3.07        3.07          3.21        3.28          3.28        3.38 23 at > 50%F        3.31        3.31          3.44        3.57          3.57        3.69
                                DW)        2.82        2.82          2.96        3.04          3.04        3.16
                                DW)        3.07        3.07          3.24        3.32          3.32        3.56 100            1.53        1.55            ---        1.56          1.56          ---
90            1.59        1.60            ---        1.61          1.61          ---
77.6            1.64        1.66            ---        1.70          1.70          ---
65            1.72        1.73            ---        1.78          1.78          ---
                                      > 50            1.82        1.83            ---        1.90          1.90          ---
50            1.88        1.89            ---        1.93          1.93          ---
FHOOS 40            1.98        1.99            ---        2.08          2.08          ---
26            2.47        2.48            ---        2.72          2.72          ---
26 at > 50%F        2.62        2.63            ---        2.81          2.81          ---
23 at > 50%F        2.79        2.80            ---        3.05          3.05          ---
                                DW)        2.52        2.53            ---        2.72          2.72          ---
                                DW)        2.68        2.69            ---        3.05          3.05          ---
100            1.48        1.50          1.55        1.48          1.50        1.56 90            1.53        1.53          1.60        1.53          1.53        1.61 77.6            1.60        1.60          1.66        1.69          1.69        1.73 65            1.78        1.78          1.81        1.84          1.84        1.84
                                      > 50            ---          ---            ---          ---          ---          ---
50            1.86        1.86          1.89        1.93          1.93        1.93 PLUOOS 40            1.96        1.96          1.99        2.01          2.01        2.08 26            2.33        2.33          2.48        2.57          2.57        2.72 26 at > 50%F        2.50        2.50          2.63        2.68          2.68        2.81 23 at > 50%F        2.65        2.65          2.80        2.93          2.93        3.05
                                DW)        2.41        2.41          2.53        2.63          2.63        2.72
                                DW)        2.55        2.55          2.69        2.93          2.93        3.05
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 2 Cycle 22                                                                                                    Page 26 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                              Page 35 of 50                                        February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                            1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis (continued)
* ATRIUM-10XM                                ATRIUM-11 BOC        BOC            BOC          BOC          BOC        BOC Power            to          to        to End of        to            to      to End of Operating
(% of rated)      NEOC        EOCLB          Coast        NEOC        EOCLB        Coast Condition 100            1.59        1.61            ---        1.60          1.61          ---
90            1.64        1.65            ---        1.65          1.65          ---
77.6            1.70        1.72            ---        1.74          1.74          ---
65            1.78        1.79            ---        1.82          1.82          ---
                                      > 50            1.87        1.88            ---          ---          ---          ---
TBVOOS                      50            1.91        1.92            ---        1.93          1.93          ---
FHOOS                        40            2.02        2.03            ---        2.09          2.09          ---
26            2.50        2.51            ---        2.72          2.72          ---
26 at > 50%F        3.20        3.21            ---        3.38          3.38          ---
23 at > 50%F        3.43        3.44            ---        3.69          3.69          ---
                                  DW)        2.95        2.96            ---        3.16          3.16          ---
                                  DW)        3.23        3.24            ---        3.56          3.56          ---
100            1.54        1.57          1.61        1.54          1.57        1.61 90            1.59        1.60          1.65        1.59          1.60        1.65 77.6            1.66        1.66          1.72        1.72          1.72        1.75 65            1.81        1.81          1.84        1.84          1.84        1.84
                                      > 50            ---          ---            ---          ---          ---          ---
TBVOOS                      50            1.89        1.89          1.92        1.93          1.93        1.93 PLUOOS                      40            1.99        1.99          2.03        2.01          2.01        2.09 26            2.38        2.38          2.51        2.57          2.57        2.72 26 at > 50%F        3.07        3.07          3.21        3.28          3.28        3.38 23 at > 50%F        3.31        3.31          3.44        3.57          3.57        3.69
                                  DW)        2.82        2.82          2.96        3.04          3.04        3.16
                                  DW)        3.07        3.07          3.24        3.32          3.32        3.56 100            1.53        1.55            ---        1.56          1.56          ---
90            1.59        1.60            ---        1.61          1.61          ---
77.6            1.64        1.66            ---        1.73          1.73          ---
65            1.80        1.81            ---        1.84          1.84          ---
                                      > 50            ---          ---            ---          ---          ---          ---
FHOOS                      50            1.88        1.89            ---        1.93          1.93          ---
PLUOOS                      40            1.98        1.99            ---        2.08          2.08          ---
26            2.47        2.48            ---        2.72          2.72          ---
26 at > 50%F        2.62        2.63            ---        2.81          2.81          ---
23 at > 50%F        2.79        2.80            ---        3.05          3.05          ---
                                  DW)        2.52        2.53            ---        2.72          2.72          ---
                                  DW)        2.68        2.69            ---        3.05          3.05          ---
100            1.59        1.61            ---        1.60          1.61          ---
90            1.64        1.65            ---        1.65          1.65          ---
77.6            1.70        1.72            ---        1.75          1.75          ---
65            1.83        1.84            ---        1.84          1.84          ---
                                      > 50            ---          ---            ---          ---          ---          ---
TBVOOS 50            1.91        1.92            ---        1.93          1.93          ---
FHOOS 40            2.02        2.03            ---        2.09          2.09          ---
PLUOOS 26            2.50        2.51            ---        2.72          2.72          ---
26 at > 50%F        3.20        3.21            ---        3.38          3.38          ---
23 at > 50%F        3.43        3.44            ---        3.69          3.69          ---
                                  DW)        2.95        2.96            ---        3.16          3.16          ---
                                  DW)        3.23        3.24            ---        3.56          3.56          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 2 Cycle 22                                                                                                    Page 27 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                              Page 36 of 50                                        February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                              1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 4.5 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM                            ATRIUM-11 BOC          BOC        BOC        BOC          BOC          BOC Power              to            to      to End of      to            to        to End of Operating
(% of rated)      NEOC          EOCLB        Coast    NEOC          EOCLB          Coast Condition 100            1.48          1.51        1.51      1.50          1.51          1.51 90            1.54          1.56        1.56      1.55          1.56          1.56 77.6            1.60          1.62        1.62      1.69          1.70          1.70 65            1.77          1.78        1.78      1.83          1.83          1.83
                                      > 50            ---            ---        ---        ---          ---          ---
50            1.94          1.95        1.95      2.02          2.02          2.02 TBVIS 40            2.11          2.12        2.12      2.29          2.29          2.29 26            2.68          2.69        2.69      3.07          3.07          3.07 26 at > 50%F        2.85          2.86        2.86      3.12          3.12          3.12 23 at > 50%F        3.06          3.07        3.07      3.42          3.42          3.42
                                DW)        2.74          2.75        2.75      3.07          3.07          3.07
                                DW)        2.93          2.94        2.94      3.37          3.37          3.37 100            1.54          1.56        1.56      1.54          1.56          1.56 90            1.59          1.61        1.61      1.59          1.61          1.61 77.6            1.66          1.68        1.68      1.71          1.72          1.72 65            1.80          1.81        1.81      1.83          1.83          1.83
                                      > 50            ---            ---        ---        ---          ---          ---
50            1.97          1.98        1.98      2.02          2.02          2.02 TBVOOS 40            2.14          2.15        2.15      2.29          2.29          2.29 26            2.71          2.72        2.72      3.07          3.07          3.07 26 at > 50%F        3.36          3.37        3.37      3.60          3.60          3.60 23 at > 50%F        3.62          3.63        3.63      4.00          4.00          4.00
                                DW)        3.12          3.13        3.13      3.39          3.39          3.39
                                DW)        3.42          3.43        3.43      3.89          3.89          3.89
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 2 Cycle 22                                                                                                  Page 28 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                                Page 37 of 50                                February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                              1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 4.6 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM                            ATRIUM-11 BOC          BOC        BOC        BOC          BOC          BOC Power              to            to    to End of      to            to        to End of Operating
(% of rated)      NEOC          EOCLB      Coast      NEOC          EOCLB          Coast Condition 100            1.48          1.51        1.51      1.50          1.51          1.51 90            1.54          1.56        1.56      1.55          1.56          1.56 77.6            1.60          1.62        1.62      1.69          1.70          1.70 65            1.77          1.78        1.78      1.83          1.83          1.83
                                      > 50            ---            ---        ---        ---          ---          ---
50            1.95          1.96        1.96      2.02          2.02          2.02 TBVIS 40            2.12          2.13        2.13      2.30          2.30          2.30 26            2.70          2.71        2.71      3.09          3.09          3.09 26 at > 50%F        2.86          2.87        2.87      3.14          3.14          3.14 23 at > 50%F        3.08          3.09        3.09      3.44          3.44          3.44
                                DW)        2.75          2.76        2.76      3.09          3.09          3.09
                                DW)        2.95          2.96        2.96      3.39          3.39          3.39 100            1.54          1.56        1.56      1.54          1.56          1.56 90            1.59          1.61        1.61      1.59          1.61          1.61 77.6            1.66          1.68        1.68      1.71          1.72          1.72 65            1.80          1.81        1.81      1.83          1.83          1.83
                                      > 50            ---            ---        ---        ---          ---          ---
50            1.98          1.99        1.99      2.02          2.02          2.02 TBVOOS 40            2.15          2.16        2.16      2.30          2.30          2.30 26            2.73          2.74        2.74      3.09          3.09          3.09 26 at > 50%F        3.38          3.39        3.39      3.62          3.62          3.62 23 at > 50%F        3.63          3.64        3.64      4.02          4.02          4.02
                                DW)        3.13          3.14        3.14      3.40          3.40          3.40
                                DW)        3.43          3.44        3.44      3.90          3.90          3.90
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 2 Cycle 22                                                                                                  Page 29 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                                Page 38 of 50                                February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                              1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 4.7 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM                            ATRIUM-11 BOC          BOC        BOC        BOC          BOC          BOC Power              to            to    to End of      to            to        to End of Operating
(% of rated)      NEOC          EOCLB      Coast      NEOC          EOCLB          Coast Condition 100            1.53          1.55        1.55      1.56          1.56          1.56 90            1.59          1.60        1.60      1.61          1.61          1.61 77.6            1.64          1.66        1.66      1.73          1.73          1.73 65            1.80          1.81        1.81      1.84          1.84          1.84
                                      > 50            ---            ---        ---        ---          ---          ---
50            1.98          1.99        1.99      2.04          2.04          2.04 TBVIS 40            2.15          2.16        2.16      2.32          2.32          2.32 26            2.73          2.74        2.74      3.10          3.10          3.10 26 at > 50%F        2.87          2.88        2.88      3.12          3.12          3.12 23 at > 50%F        3.08          3.09        3.09      3.42          3.42          3.42
                                DW)        2.76          2.77        2.77      3.10          3.10          3.10
                                DW)        2.95          2.96        2.96      3.40          3.40          3.40 100            1.59          1.61        1.61      1.60          1.61          1.61 90            1.64          1.65        1.65      1.65          1.65          1.65 77.6            1.70          1.72        1.72      1.75          1.75          1.75 65            1.83          1.84        1.84      1.84          1.84          1.84
                                      > 50            ---            ---        ---        ---          ---          ---
50            2.01          2.02        2.02      2.05          2.05          2.05 TBVOOS 40            2.19          2.20        2.20      2.32          2.32          2.32 26            2.76          2.77        2.77      3.10          3.10          3.10 26 at > 50%F        3.38          3.39        3.39      3.60          3.60          3.60 23 at > 50%F        3.64          3.65        3.65      4.00          4.00          4.00
                                DW)        3.14          3.15        3.15      3.39          3.39          3.39
                                DW)        3.44          3.45        3.45      3.89          3.89          3.89
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 2 Cycle 22                                                                                                  Page 30 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                              1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 4.8 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM                            ATRIUM-11 BOC          BOC        BOC        BOC          BOC          BOC Power              to            to    to End of      to            to        to End of Operating
(% of rated)      NEOC          EOCLB      Coast      NEOC          EOCLB          Coast Condition 100            1.53          1.55        1.55      1.56          1.56          1.56 90            1.59          1.60        1.60      1.61          1.61          1.61 77.6            1.64          1.66        1.66      1.73          1.73          1.73 65            1.80          1.81        1.81      1.84          1.84          1.84
                                      > 50            ---            ---        ---        ---          ---          ---
50            1.99          2.00        2.00      2.05          2.05          2.05 TBVIS 40            2.16          2.17        2.17      2.33          2.33          2.33 26            2.75          2.76        2.76      3.12          3.12          3.12 26 at > 50%F        2.88          2.89        2.89      3.14          3.14          3.14 23 at > 50%F        3.10          3.11        3.11      3.44          3.44          3.44
                                DW)        2.77          2.78        2.78      3.12          3.12          3.12
                                DW)        2.97          2.98        2.98      3.42          3.42          3.42 100            1.59          1.61        1.61      1.60          1.61          1.61 90            1.64          1.65        1.65      1.65          1.65          1.65 77.6            1.70          1.72        1.72      1.75          1.75          1.75 65            1.83          1.84        1.84      1.84          1.84          1.84
                                      > 50            ---            ---        ---        ---          ---          ---
50            2.02          2.03        2.03      2.06          2.06          2.06 TBVOOS 40            2.20          2.21        2.21      2.33          2.33          2.33 26            2.78          2.79        2.79      3.12          3.12          3.12 26 at > 50%F        3.40          3.41        3.41      3.62          3.62          3.62 23 at > 50%F        3.65          3.66        3.66      4.02          4.02          4.02
                                DW)        3.15          3.16        3.16      3.40          3.40          3.40
                                DW)        3.45          3.46        3.46      3.90          3.90          3.90
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 4.9 MCPRP Limits for All Fuel Types: Single Loop Operation for All Scram Times
* Power                  All Cycle Exposures Operating Condition        (% of rated)      ATRIUM-10XM          ATRIUM-11 100                2.03              2.05 43.75                2.03              2.05 40                2.03              2.10 RCPOOS                    26                2.50              2.74 FHOOS                26 at > 50%F            2.65              2.83 23 at > 50%F            2.82              3.07
                                                      DW)            2.55              2.74
                                                      DW)            2.71              3.07 100                2.03              2.05 43.75                2.03              2.05 RCPOOS                    40                2.05              2.11 TBVOOS                    26                2.53              2.74 PLUOOS              26 at > 50%F            3.23              3.40 FHOOS                23 at > 50%F            3.46              3.71
                                                      DW)            2.98              3.18
                                                      DW)            3.26              3.58 100                2.16              2.24 43.75                2.16              2.24 40                2.22              2.34 RCPOOS 26                2.79              3.12 TBVOOS 26 at > 50%F            3.41              3.62 FHOOS1 23 at > 50%F            3.67              4.02
                                                      DW)            3.17              3.41
                                                      DW)            3.47              3.91 100                2.17              2.25 43.75                2.17              2.25 40                2.23              2.35 RCPOOS 26                2.81              3.14 TBVOOS 26 at > 50%F            3.43              3.64 FHOOS2 23 at > 50%F            3.68              4.04
                                                      DW)            3.18              3.42
                                                      DW)            3.48              3.92
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop.
RCPOOS limits are only valid up to 50% rated core power, 50% rated core flow, and an active recirculation drive flow of 17.73 Mlbm/hr.
Browns Ferry Unit 2 Cycle 22                                                                                                      Page 32 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 5      Thermal-Hydraulic Stability Protection (Technical Specification 3.3.1.1)
Technical Specification Table 3.3.1.1-1, Function 2f, identifies the function.
Instrument setpoints are established, such that the reactor will be tripped before an oscillation can grow to the point where the SLMCPR is exceeded. With application of Reference 30, the DSS-CD stability solution will be used per Reference 26. The DSS-CD SAD setpoint is 1.10 for TLO and SLO.
New analyses have been developed based on Reference 26. With the implementation of the MELLLA+ operating domain expansion, an ABSP trip is required when the OPRM is out-of-service. The ABSP trip settings define a region of the power to flow map within which an automatic reactor scram occurs. The ABSP trip settings are provided in Table 5.1. If both the OPRM and ABSP are out-of-service, operation within the MELLLA+ domain is not allowed and the MBSP Regions provide stability protection. Table 5.2 and Table 5.3 provide the endpoints for the MBSP regions for nominal and reduced feedwater temperature conditions.
Table 5.1 ABSP Setpoints for the Scram Region Parameter                Symbol        Setting Value (unit)                          Comments Slope of ABSP APRM low Flow Biased Slope for Trip              mTRIP          2.00 (% RTP/% RDF)
Trip Linear Segment ABSP APRM Flow Biased Trip Constant Power                                                              Setpoint Power Intercept. Constant PBSP-TRIP          35.0 (% RTP)
Line for Trip                                                            Power Line for Trip from Zero Drive Flow to Flow Breakpoint Value ABSP APRM Flow Biased Trip Constant Flow Line                                                            Setpoint Drive Flow Intercept.
WBSP-TRIP            49 (% RDF) for Trip                                                              Constant Flow Line for Trip (see Note 1 below)
Flow Breakpoint            WBSP-BREAK          30.0 (% RDF)              Flow Breakpoint Value Note 1: WBSP-TRIP can be set to 49.0 % RDF or any higher value up to the intersection of the ABSP sloped line with the APRM Flow Biased STP scram line.
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 5.2 Analyzed MBSP Endpoints: Nominal Feedwater Temperature Power      Core Flow Endpoint                                                Definition
(% Rated)    (% Rated)
Scram Region (Region I)
A1          75.9        52.7        Boundary Intercept on MELLLA+
Line Scram Region (Region I)
B1          35.5        29.0        Boundary Intercept on Natural Circulation Line (NCL)
Controlled Entry Region (Region A2          66.1        52.0        II) Boundary Intercept on MELLLA Line Controlled Entry Region (Region B2          25.5        29.0        II) Boundary Intercept on Natural Circulation Line (NCL)
Table 5.3 Analyzed MBSP Endpoints: Reduced Feedwater Temperature Power      Core Flow Endpoint                                                Definition
(% Rated)    (% Rated)
Scram Region (Region I)
A1          64.9        50.5        Boundary Intercept on MELLLA Line Scram Region (Region I)
B1          29.4        29.0        Boundary Intercept on Natural Circulation Line (NCL)
Controlled Entry Region (Region A2          68.3        54.9        II) Boundary Intercept on MELLLA Line Controlled Entry Region (Region B2          24.5        29.0        II) Boundary Intercept on Natural Circulation Line (NCL)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 6 APRM Flow Biased Rod Block Trip Settings (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
The APRM rod block trip setting is based upon References 27 & 29, and is defined by the following:
for two loop operation:
SRB d (0.61Wd + 63.3)                              Allowable Value SRB d (0.61Wd + 62.0)                              Nominal Trip Setpoint (NTSP) where:
SRB          =          Rod Block setting in percent of rated thermal power (3952 MWt)
Wd          =          Recirculation drive flow rate in percent of rated (100% drive flow required to achieve 100% core power and flow) and for single loop operation:
SRB d (0.55(Wd-'W) + 60.5)                                  Allowable Value SRB d (0.55(Wd-'W) + 58.5)                                  Nominal Trip Setpoint (NTSP) where:
SRB          =          Rod Block setting in percent of rated thermal power (3952 MWt)
Wd          =          Recirculation drive flow rate in percent of rated (100% drive flow required to achieve 100% core power and flow)
          'W          =          Difference between two-loop and single-loop effective recirculation flow at the same core flow ('W=0.0 for two-loop operation)
The APRM rod block trip setting is clamped at a maximum allowable value of 115%
(corresponding to a NTSP of 113%).
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                              1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 7 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
The RBM trip setpoints and applicable power ranges, based on References 27 & 28, are shown in Table 7.1. Setpoints are based on an HTSP, unfiltered analytical limit of 114%. Unfiltered setpoints are consistent with a nominal RBM filter setting of 0.0 seconds; filtered setpoints are consistent with a nominal RBM filter setting less than 0.5 seconds. Cycle specific CRWE analyses of OLMCPR are documented in Reference 1, superseding values reported in References 27, 28, and 29.
Table 7.1 Analytical RBM Trip Setpoints
* Allowable              Nominal Trip RBM                              Value                Setpoint Trip Setpoint                          (AV)                  (NTSP)
LPSP                                    27%                    25%
IPSP                                    62%                    60%
HPSP                                    82%                    80%
LTSP - unfiltered                    121.7%                  120.0%
                                          - filtered                    120.7%                  119.0%
ITSP    - unfiltered                  116.7%                  115.0%
                                          - filtered                    115.7%                  114.0%
HTSP - unfiltered                    111.7%                  110.0%
                                          - filtered                    110.9%                  109.2%
DTSP                                    90%                    92%
As a result of cycle specific CRWE analyses, RBM setpoints in Technical Specification Table 3.3.2.1-1 are applicable as shown in Table 7.2. Cycle specific setpoint analysis results are shown in Table 7.3, per Reference 1.
Table 7.2 RBM Setpoint Applicability Thermal Power                      Applicable                Notes from
(% Rated)                        MCPR                Table 3.3.2.1-1              Comment
                                                        < 1.72              (a), (b), (f), (h)    two loop operation 27% and < 90%
                                                        < 1.76              (a), (b), (f), (h)    single loop operation 90%                                      < 1.41                      (g)            two loop operation
* Values are considered maximums. Using lower values, due to RBM system hardware/software limitations, is conservative, and acceptable.
MCPR values shown correspond with, (support), SLMPCR values identified in Reference 1.
Greater than 90% rated power is not attainable in single loop operation.
Browns Ferry Unit 2 Cycle 22                                                                                                  Page 36 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                  TVA-COLR-BF2C22, Revision 0 (Final)
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ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Table 7.3 Control Rod Withdrawal Error Results RBM                          CRWE HTSP Analytical Limit                OLMCPR Unfiltered 107                          1.28 111                          1.34 114                          1.37 117                          1.39 Results, compared against the base case OLMCPR results of Table 4.2, indicate SLMCPR remains protected for RBM inoperable conditions (i.e., 114% unblocked).
Browns Ferry Unit 2 Cycle 22                                                                                      Page 37 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                            Page 46 of 50                          February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE
[I!lil                        1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 8 Shutdown Margin Limit (Technical Specification 3.1.1)
Assuming the strongest OPERABLE control blade is fully withdrawn, and all other OPERABLE control blades are fully inserted, the core shall be sub-critical and meet the following minimum shutdown margin:
SDM      0.38% dk/k Browns Ferry Unit 2 Cycle 22                                                                                      Page 38 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                          Page 47 of 50                            February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                          1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Appendix A: MBSP Maps Browns Ferry Unit 2 Cycle 22                                                                                    Page A-1 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                          Page 48 of 50                            February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                              1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Core Power (% Rated: 100% = 3952MWt) 110 100 MELLLA+ Region 90 80 BSP Boundary Manual 70                                      Scram Region I 60 MELLLA Region      ICF Region MELLLA Upper Boundary 50 87.5% Rod Line 40                                                        Controlled Entry Region II 30                                                                Min. Flow Control 20 Natural                                Min. Power Line Circulation 10                                            20% Pump Speed Line 0
0          10        20        30        40      50        60        70        80      90      100        110          120 Core Flow (% Rated: 100% =102.5 MLbm /hr)
Figure A.1 MBSP Boundaries For Nominal Feedwater Temperature (Operation in the MELLLA+ Region Prohibited for Feedwater Temperature greater than 10 degrees F below the Nominal Feedwater Temperature)
Browns Ferry Unit 2 Cycle 22                                                                                              Page A-2 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                              Page 49 of 50                                  February 11, 2021 Revision 149
 
ECM: L94 210104 800 Reactor Engineering and Fuels - BWRFE mil                              1101 Market Street, Chattanooga TN 37402 Date: January 4, 2021 Core Power (% Rated: 100% = 3952MWt) 110 100 MELLLA+ Region 90 80 BSP Boundary 70 Manual Scram 60                                      Region I MELLLA Region MELLLA Upper Boundary                                                ICF Region 50 87.5% Rod Line Controlled Entry 40                                                                    Region II 30                                                                Min. Flow Control 20 Natural                                Min. Power Line Circulation 10                                            20% Pump Speed Line 0
0          10        20        30        40      50        60        70      80      90      100        110          120 Core Flow (% Rated: 100% =102.5 MLbm /hr)
Figure A.2 MBSP Boundaries For Reduced Feedwater Temperature (Operation in the MELLLA+ Region Prohibited for a Reduced Feedwater Temperature greater than 10 degrees F below the Nominal Feedwater Temperature)
Browns Ferry Unit 2 Cycle 22                                                                                            Page A-3 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF2C22, Revision 0 (Final)
BFN - UNIT 2                                              Page 50 of 50                                February 11, 2021 Revision 149
 
EDMS L94 200213 800 QA Document BFE-4485, Revision 0 imNPG Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga , TN 37402 Browns Ferry Unit 3 Cycle 20 Core Operating Limits Report, (120% OL TP, MELLLA+)
TVA-COLR-BF3C20                            Revision O (Final)
(Revision Log , Page v)
February 2020 Prepared ~                      &#xa3;,:/4_L /
                                  ..Eichenberg , Sr. Spe<lattst Date:  .:ft    Ir I -2 0-10 Verified :                                      Date: _ _  2_/_V_l/2_2_0_
                            .....g:'"""c_Mitchell, Engineer Approved :      ill~!~
C. A. Setter, Manager, BWRFE Date :  z..{ ZOil.DZ D
                                                                          ;2./21(2.0 Approved :
c~~~                            Date:      o;z.../2-7fao Approved :
Pltb./IL                        Date :  3[7-( Zo I  1
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table of Contents Total Number of Pages = 44 (including review cover sheet)
List of Tables ............................................................................................................................. iii List of Figures ............................................................................................................................ iv Revision Log ...............................................................................................................................v Nomenclature ............................................................................................................................ vi References .............................................................................................................................. viii 1      Introduction ........................................................................................................................1 1.1        Purpose .......................................................................................................................1 1.2        Scope ..........................................................................................................................1 1.3        Fuel Loading................................................................................................................1 1.4        Acceptability ................................................................................................................2 2      APLHGR Limits ..................................................................................................................3 2.1        Rated Power and Flow Limit: APLHGRRATED ...............................................................3 2.2        Off-Rated Power Dependent Limit: APLHGRP ............................................................3 2.2.1            Startup without Feedwater Heaters..................................................................... 3 2.3        Off-Rated Flow Dependent Limit: APLHGRF ...............................................................3 2.4        Single Loop Operation Limit: APLHGRSLO ...................................................................3 2.5        Equipment Out-Of-Service Corrections........................................................................5 3      LHGR Limits.......................................................................................................................6 3.1        Rated Power and Flow Limit: LHGRRATED....................................................................6 3.2        Off-Rated Power Dependent Limit: LHGRP .................................................................6 3.2.1            Startup without Feedwater Heaters..................................................................... 6 3.3        Off-Rated Flow Dependent Limit: LHGRF....................................................................7 3.4        Equipment Out-Of-Service Corrections........................................................................7 4      OLMCPR Limits................................................................................................................13 4.1        Flow Dependent MCPR Limit: MCPRF ......................................................................13 4.2        Power Dependent MCPR Limit: MCPRP ...................................................................13 4.2.1            Startup without Feedwater Heaters....................................................................13 4.2.2            Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS) ................................14 4.2.3            Exposure Dependent Limits ...............................................................................14 4.2.4            Equipment Out-Of-Service (EOOS) Options ......................................................15 4.2.5            Single-Loop-Operation (SLO) Limits ..................................................................15 4.2.6            Below Pbypass Limits ........................................................................................15 5      Thermal-Hydraulic Stability Protection..............................................................................27 6      APRM Flow Biased Rod Block Trip Settings.....................................................................29 7      Rod Block Monitor (RBM) Trip Setpoints and Operability .................................................30 8      Shutdown Margin Limit .....................................................................................................32 Appendix A:            MBSP Maps..................................................................................................... A-1 Browns Ferry Unit 3 Cycle 20                                                                                                          Page ii Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800 mil NPG                                Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 List of Tables Nuclear Fuel Types                .............................................................................................................. 2 Startup Feedwater Temperature Basis....................................................................................... 6 Nominal Scram Time Basis .......................................................................................................14 MCPRP Limits for All Fuel Types: Optimum Scram Time Basis                                          ............................................17 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis                                          .............................................18 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis                                                ......................20 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Nominal Scram Time Basis                      ..................................................................................................22 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Nominal Scram Time Basis                      ..................................................................................................23 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Technical Specification Scram Time Basis                              ..........................................................................24 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Technical Specification Scram Time Basis                              ..........................................................................25 MCPRP Limits for All Fuel Types: Single Loop Operation for All Scram Times                                                  ....................26 ABSP Setpoints for the Scram Region ......................................................................................27 Analyzed MBSP Endpoints: Nominal Feedwater Temperature.................................................28 Analyzed MBSP Endpoints: Reduced Feedwater Temperature ...............................................28 Analytical RBM Trip Setpoints                      .............................................................................................30 RBM Setpoint Applicability ........................................................................................................30 Control Rod Withdrawal Error Results.......................................................................................31 Browns Ferry Unit 3 Cycle 20                                                                                                              Page iii Core Operating Limits Report, (120% OLTP, MELLLA+)                                                          TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800 mil NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 List of Figures APLHGRRATED for ATRIUM-10XM Fuel....................................................................................... 4 LHGRRATED for ATRIUM-10XM Fuel ........................................................................................... 8 Base Operation LHGRFACP for ATRIUM-10XM Fuel................................................................. 9 LHGRFACF for ATRIUM-10XM Fuel .........................................................................................10 Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 1 ........11 Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 2 ........12 MCPRF for All Fuel Types .........................................................................................................16 MBSP Boundaries For Nominal Feedwater Temperature........................................................ A-2 MBSP Boundaries For Reduced Feedwater Temperature ...................................................... A-3 Browns Ferry Unit 3 Cycle 20                                                                                                Page iv Core Operating Limits Report, (120% OLTP, MELLLA+)                                            TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Revision Log Number            Page                                          Description 0-R0            All        New document.
Browns Ferry Unit 3 Cycle 20                                                                                        Page v Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800 mil NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Nomenclature ABSP                  Automatic Backup Stability Protection APLHGR                Average Planar LHGR APRM                  Average Power Range Monitor AREVA NP              Vendor (Framatome, Siemens)
BOC                    Beginning of Cycle BSP                    Backup Stability Protection BWR                    Boiling Water Reactor CAVEX                  Core Average Exposure CD                    Coast Down CMSS                  Core Monitoring System Software COLR                  Core Operating Limits Report CPR                    Critical Power Ratio CRWE                  Control Rod Withdrawal Error CSDM                  Cold SDM DIVOM                  Delta CPR over Initial CPR vs. Oscillation Magnitude DSS-CD                Detect and Suppress Solution - Confirmation Density EOC                    End of Cycle EOCLB                  End-of-Cycle Licensing Basis EOOS                  Equipment OOS EPU                    Extended Power Uprate (120% OLTP)
FFTR                  Final Feedwater Temperature Reduction FFWTR                  Final Feedwater Temperature Reduction FHOOS                  Feedwater Heaters OOS ft                    Foot: English unit of measure for length GNF                    Vendor (General Electric, Global Nuclear Fuels)
GWd                    Giga Watt Day HTSP                  High TSP ICA                    Interim Corrective Action ICF                    Increased Core Flow (beyond rated)
IS                    In-Service kW                    kilo watt: SI unit of measure for power.
LCO                    License Condition of Operation LFWH                  Loss of Feedwater Heating LHGRFAC                LHGR Multiplier (Power or Flow dependent)
LPRM                  Low Power Range Monitor LRNB                  Generator Load Reject, No Bypass Browns Ferry Unit 3 Cycle 20                                                                                        Page vi Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800 mil NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 MAPFAC                MAPLHGR multiplier (Power or Flow dependent)
MBSP                  Manual Backup Stability Protection MCPR                  Minimum CPR MELLLA                Maximum Extended Load Line Limit Analysis MELLLA+                Maximum Extended Load Line Limit Analysis Plus MSRV                  Moisture Separator Reheater Valve MSRVOOS                MSRV OOS MTU                    Metric Ton Uranium MWd/MTU                Mega Watt Day per Metric Ton Uranium NEOC                  Near EOC NRC                    United States Nuclear Regulatory Commission NSS                    Nominal Scram Speed NTSP                  Nominal TSP OLMCPR                MCPR Operating Limit OLTP                  Original Licensed Thermal Power OOS                    Out-Of-Service OPRM                  Oscillation Power Range Monitor OSS                    Optimum Scram Speed PBDA                  Period Based Detection Algorithm Pbypass                Power, below which TSV Position and TCV Fast Closure Scrams are Bypassed PLU                    Power Load Unbalance PLUOOS                PLU OOS PRNM                  Power Range Neutron Monitor RBM                    Rod Block Monitor RCPOOS                Recirculation Pump OOS (SLO)
RDF                    Rated Drive Flow RPS                    Reactor Protection System RPT                    Recirculation Pump Trip RPTOOS                RPT OOS RTP                    Rated Thermal Power SDM                    Shutdown Margin SLMCPR                MCPR Safety Limit SLO                    Single Loop Operation TBV                    Turbine Bypass Valve TBVIS                  TBV IS TBVOOS                Turbine Bypass Valves OOS TIP                    Transversing In-core Probe TIPOOS                TIP OOS TLO                    Two Loop Operation TSP                    Trip Setpoint TSSS                  Technical Specification Scram Speed TVA                    Tennessee Valley Authority Browns Ferry Unit 3 Cycle 20                                                                                      Page vii Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 References
: 1.        ANP-3813, Revision 0, Browns Ferry Unit 3 Cycle 20 Reload Analysis, Framatome Inc., January 2020.
: 2.        Not Used.
: 3.        ANP-3150P, Revision 4, Mechanical Design Report for Browns Ferry ATRIUM 10XM Fuel Assemblies, AREVA Inc., November 2017.
: 4.        ANP-3793P Revision 0, Browns Ferry Unit 3 Cycle 20 Plant Parameters Document, Framatome Inc., June 2019.
: 5.        BFE-4468, Revision 0, Browns Ferry Unit 3 Cycle 20 In-Core Shuffle, Tennessee Valley Authority, January 31, 2020.
Methodology References
: 6.        XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
: 7.        XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
: 8.        EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
: 9.        ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
: 10.      XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.
: 11.      XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.
: 12.      EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.
: 13.      XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.
: 14.      XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.
: 15.      ANP-10307PA, Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP Inc., June 2011.
Browns Ferry Unit 3 Cycle 20                                                                                      Page viii Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
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[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020
: 16.      ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.
: 17.      ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Advanced Nuclear Fuels Corporation, September 2005.
: 18.      EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP Inc.,
September 2009.
: 19.      EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001, as supplemented by the site specific approval in NRC safety evaluations dated February 15, 2013 and July 31, 2014.
: 20.      EMF-2292(P)(A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
: 21.      EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
: 22.      BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP Inc., May 2008.
: 23.      BAW-10247PA, Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP Inc., April 2008.
: 24.      ANP-10298PA, Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc., March 2010.
: 25.      ANP-3140(P), Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc.,
August 2012.
: 26.      NEDC-33075P-A, Revision 8, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, GE Hitachi, November 2013.
Setpoint References
: 27.      EDQ2092900118, R35, Setpoint and Scaling Calculation for Neutron Monitoring &
Recirculation Flow Loops, Calculation File, Tennessee Valley Authority, August 9, 2019.
: 28.      Task T0500, Revision 0, Neutron Monitoring System w/RBM, Project Task Report, GE Hitachi Nuclear Energy, June 2017.
: 29.      Task T0506, Revision 0, TS Instrument Setpoints, Project Task Report, Tennessee Valley Authority, August, 2017.
: 30.      NEDC-33006P-A, Revision 3, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, GE Energy Nuclear, June 2009.
Browns Ferry Unit 3 Cycle 20                                                                                        Page ix Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800 mil NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 1 Introduction In anticipation of cycle startup, it is necessary to describe the expected limits of operation.
1.1    Purpose The primary purpose of this document is to satisfy requirements identified by unit technical specification section 5.6.5. This document may be provided, upon final approval, to the NRC.
1.2    Scope This document will discuss the following areas:
3/4 Average Planar Linear Heat Generation Rate (APLHGR) Limit (Technical Specifications 3.2.1 and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3.2.3, 3.3.4.1, and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Minimum Critical Power Ratio Operating Limit (OLMCPR)
(Technical Specifications 3.2.2, 3.3.4.1, 3.7.5 and Table 3.3.2.1-1)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Thermal-Hydraulic Stability Protection (Technical Specification Table 3.3.1.1)
Applicability: Mode 1,  (as specified in Technical Specifications Table 3.3.1.1-1) 3/4 Average Power Range Monitor (APRM) Flow Biased Rod Block Trip Setting (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
Applicability: Mode 1,  (as specified in Technical Requirements Manuals Table 3.3.4-1) 3/4 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
Applicability: Mode 1,  % RTP as specified in Table 3.3.2.1-1 (TS definition of RTP) 3/4 Shutdown Margin (SDM) Limit (Technical Specification 3.1.1)
Applicability: All Modes 1.3    Fuel Loading The core will contain fresh, and previously exposed ATRIUM-10XM. Nuclear fuel types used in the core loading are shown in Table 1.1. The core shuffle and final loading were explicitly evaluated for BOC cold shutdown margin performance as documented per Reference 5.
Browns Ferry Unit 3 Cycle 20                                                                                        Page 1 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                                Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 1.1 Nuclear Fuel Types
* Nuclear Original        Number of          Fuel Type            Fuel Names Fuel Description                              Cycle          Assemblies            (NFT)                (Range)
ATRIUM-10XM XMLC-4105B-11GV70-FCG                                    18                72                19            FCG601-FCG672 ATRIUM-10XM XMLC-4096B-12GV80-FCG                                    18                22                20            FCG673-FCG808 ATRIUM-10XM XMLC-4055B-13GV70-FCG                                    18                  14                21            FCG809-FCG904 ATRIUM-10XM XMLC-3911B-13GV80-FCH                                    19                240                22            FCH001-FCH240 ATRIUM-10XM XMLC-4053B-12GV80-FCH                                    19                104                23            FCH241-FCH344 ATRIUM-10XM XMLC-3920B-14GV80-FCJ                                    20                224                24              FCJ345-FCJ568 ATRIUM-10XM XMLC-3957B-12GV80-FCJ                                    20                  88                25              FCJ569-FCJ656 1.4      Acceptability Limits discussed in this document were generated based on NRC approved methodologies per References 6 through 25.
* The table identifies the expected fuel type breakdown in anticipation of final core loading. The final composition of the core depends upon uncertainties during the outage such as discovering a failed fuel bundle, or other bundle damage. Minor core loading changes, due to unforeseen events, will conform to the safety and monitoring requirements identified in this document.
Browns Ferry Unit 3 Cycle 20                                                                                                            Page 2 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                          TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 2 APLHGR Limits (Technical Specifications 3.2.1 & 3.7.5)
The APLHGR limit is determined by adjusting the rated power APLHGR limit for off-rated power, off-rated flow, and SLO conditions. The most limiting of these is then used as follows:
APLHGR limit = MIN ( APLHGRP , APLHGRF, APLHGRSLO )
where:
APLHGRP                off-rated power APLHGR limit                  [APLHGRRATED
* MAPFACP]
APLHGRF                off-rated flow APLHGR limit                  [APLHGRRATED
* MAPFACF]
APLHGRSLO              SLO APLHGR limit                              [APLHGRRATED
* SLO Multiplier]
2.1    Rated Power and Flow Limit: APLHGRRATED The rated conditions APLHGR for all fuel are identified per Reference 1. The rated conditions APLHGR for ATRIUM-10XM are shown in Figure 2.1.
2.2    Off-Rated Power Dependent Limit: APLHGRP Reference 1 does not specify a power dependent APLHGR. Therefore, MAPFACP is set to a value of 1.0.
2.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. No additional power dependent limitation is required.
2.3    Off-Rated Flow Dependent Limit: APLHGRF Reference 1 does not specify a flow dependent APLHGR. Therefore, MAPFACF is set to a value of 1.0.
2.4    Single Loop Operation Limit: APLHGRSLO The single loop operation multiplier for ATRIUM-10XM fuel is 0.85, per Reference 1.
Browns Ferry Unit 3 Cycle 20                                                                                        Page 3 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 15 12 APLHGR (kW/ft) 9 6
3 0
0          20                    40                      60                      80 Planar Average Exposure (GWd/MTU)
Planar Avg. APLHGR Exposure        Limit (GWd/MTU)        (kW/ft) 0.0            13.0 15.0            13.0 67.0            7.6 Figure 2.1 APLHGRRATED for ATRIUM-10XM Fuel Browns Ferry Unit 3 Cycle 20                                                                                                Page 4 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 2.5      Equipment Out-Of-Service Corrections The limits shown in Figure 2.1 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
In-Service                              All equipment In-Service
* RPTOOS                                  EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                  Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                  Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                        Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                  One Recirculation Pump Out-Of-Service
* All equipment service conditions assume 1 SRVOOS.
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EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 3 LHGR Limits (Technical Specification 3.2.3, 3.3.4.1, & 3.7.5)
The LHGR limit is determined by adjusting the rated power LHGR limit for off-rated power and off-rated flow conditions. The most limiting of these is then used as follows:
LHGR limit = MIN ( LHGRP, LHGRF )
where:
LHGRP                  off-rated power LHGR limit                    [LHGRRATED
* LHGRFACP]
LHGRF                  off-rated flow LHGR limit                    [LHGRRATED
* LHGRFACF]
3.1    Rated Power and Flow Limit: LHGRRATED The rated conditions LHGR for all fuel are identified per Reference 1. The rated conditions LHGR for ATRIUM-10XM fuel is shown in Figure 3.1. The LHGR limit is consistent with Reference 3.
3.2    Off-Rated Power Dependent Limit: LHGRP LHGR limits are adjusted for off-rated power conditions using the LHGRFACP multiplier provided in Reference 1. The multiplier is split into two sub cases: turbine bypass valves in and out-of-service. The base case multipliers are shown in Figure 3.2.
3.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional limits are shown in Figure 3.4 and Figure 3.5, based on temperature conditions identified in Table 3.1.
Table 3.1 Startup Feedwater Temperature Basis Temperature Power              Range 1                Range 2
(% Rated)                (&deg;F)                  (&deg;F) 23                  160.0                  155.0 30                  167.0                  162.0 40                  177.0                  172.0 50                  187.0                  182.0 Browns Ferry Unit 3 Cycle 20                                                                                          Page 6 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 3.3      Off-Rated Flow Dependent Limit: LHGRF LHGR limits are adjusted for off-rated flow conditions using the LHGRFACF multiplier provided in Reference 1. Multipliers are shown in Figure 3.3.
3.4      Equipment Out-Of-Service Corrections The limits shown in Figure 3.1 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
* In-Service                              All equipment In-Service RPTOOS                                  EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                  Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                  Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                        Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                  One Recirculation Pump Out-Of-Service Off-rated power corrections shown in Figure 3.2 are dependent on operation of the Turbine Bypass Valve system. For this reason, separate limits are to be applied for TBVIS or TBVOOS operation. The limits have no dependency on RPTOOS, PLUOOS, FHOOS/FFWTR, or SLO.
Off-rated flow corrections shown in Figure 3.3 are bounding for all EOOS conditions.
Off-rated power corrections shown in Figure 3.4 and Figure 3.5 are also dependent on operation of the Turbine Bypass Valve system. In this case, limits support FHOOS operation during startup. These limits have no dependency on RPTOOS, PLUOOS, or SLO.
* All equipment service conditions assume 1 SRVOOS.
Browns Ferry Unit 3 Cycle 20                                                                                        Page 7 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 15 12 9
LHGR (kW/ft) 6 3
0 0            20                    40                      60                      80 Pellet Exposure (GWd/MTU)
Pellet          LHGR Exposure        Limit (GWd/MTU)        (kW/ft) 0.0            14.1 18.9            14.1 74.4              7.4 Figure 3.1 LHGRRATED for ATRIUM-10XM Fuel Browns Ferry Unit 3 Cycle 20                                                                                                Page 8 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800 Reactor Engineering and Fuels - BWRFE Date: February 18, 2020
                ~                                    1101 Market Street, Chattanooga TN 37402 NPG 1.10 1.00 0.90 Turbine Bypass Valve In-Service, TBVIS 0.80 Turbine Bypass Valve Out-of-Service, TBVOOS 0.70 0.60 LHGRFACP 0.50            7%9,6 &RUH)ORZ 7%9226 &RUH)ORZ
                            ~    TBVIS, > 50% Core Flow 0.40 TBVOOS, > 50% Core Flow
                            /
0.30 0.20 20        30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                    Power        LHGRFACP
(% Rated)                                    (% Rated) 100.0        1.00                            100.0        1.00 26.0        0.64                            26.0        0.62 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0        0.45                            26.0        0.38 23.0        0.41                            23.0        0.35
                                &RUH)ORZ5DWHG                        &RUH)ORZ5DWHG 26.0        0.49                            26.0        0.49 23.0        0.46                            23.0        0.42 Figure 3.2 Base Operation LHGRFACP for ATRIUM-10XM Fuel (Independent of other EOOS conditions)
Browns Ferry Unit 3 Cycle 20                                                                                                  Page 9 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                  TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800 Reactor Engineering and Fuels - BWRFE Date: February 18, 2020
                ~                              1101 Market Street, Chattanooga TN 37402 NPG 1.10 1.05 1.00 0.95 0.90 0.85 LHGRFACF  0.80 0.75 0.70 0.65 0.60 0.55 20    30          40          50        60      70        80        90        100        110 Core Flow (% Rated)
Core Flow          LHGRFACF
(% Rated) 0.0            0.63 30.0            0.63 75.6            1.00 107.0            1.00 Figure 3.3 LHGRFACF for ATRIUM-10XM Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 3 Cycle 20                                                                                        Page 10 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
                ~      NPG Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 1.10 1.00 0.90 0.80        Turbine Bypass Valve In-Service, TBVIS Turbine Bypass Valve Out-of-Service, TBVOOS LHGRFACP 0.70 0.60 0.50 7%9,6 &RUH)ORZ TBVIS, > 50% Core Flow 0.40    ''  7%9226 &RUH)ORZ TBVOOS, > 50% Core Flow
                            \
0.30 0.20 20    30          40          50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                  Power        LHGRFACP
(% Rated)                                  (% Rated) 100.0        1.00                          100.0        1.00 26.0        0.52                          26.0        0.51 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0        0.41                          26.0        0.34 23.0        0.38                          23.0        0.32
                            &RUH)ORZ5DWHG                      &RUH)ORZ5DWHG 26.0        0.44                          26.0        0.40 23.0        0.40                          23.0        0.39 Figure 3.4 Startup Operation LHGRFACP for ATRIUM-10XM Fuel:
Table 3.1 Temperature Range 1 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 3 Cycle 20                                                                                          Page 11 Core Operating Limits Report, (120% OLTP, MELLLA+)                                            TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
                ~      NPG Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 1.10 1.00 0.90 0.80        Turbine Bypass Valve In-Service, TBVIS Turbine Bypass Valve Out-of-Service, TBVOOS LHGRFACP 0.70 0.60 0.50 7%9,6 &RUH)ORZ TBVIS, > 50% Core Flow 0.40    ''  7%9226 &RUH)ORZ TBVOOS, > 50% Core Flow
                            \
0.30 0.20 20    30          40          50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service              Turb ine Bypass Out-of-Service Core                                        Core Power            LHGRFACP                  Power        LHGRFACP
(% Rated)                                  (% Rated) 100.0        1.00                        100.0        1.00 26.0        0.52                          26.0        0.51 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0        0.41                          26.0        0.34 23.0        0.38                          23.0        0.32
                            &RUH)ORZ5DWHG                      &RUH)ORZ5DWHG 26.0        0.44                          26.0        0.40 23.0        0.40                          23.0        0.39 Figure 3.5 Startup Operation LHGRFACP for ATRIUM-10XM Fuel:
Table 3.1 Temperature Range 2 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 3 Cycle 20                                                                                        Page 12 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 4 OLMCPR Limits (Technical Specification 3.2.2, 3.3.4.1, & 3.7.5)
OLMCPR is calculated to be the most limiting of the flow or power dependent values OLMCPR limit = MAX ( MCPRF , MCPRP )
where:
MCPRF              core flow-dependent MCPR limit MCPRP              power-dependent MCPR limit 4.1    Flow Dependent MCPR Limit: MCPRF MCPRF limits are dependent upon core flow (% of Rated), and the max core flow limit, (Rated or Increased Core Flow, ICF). MCPRF limits are shown in Figure 4.1, per Reference 1. Limits are valid for all EOOS combinations. No adjustment is required for SLO conditions.
4.2    Power Dependent MCPR Limit: MCPRP MCPRP limits are dependent upon:
Core Power Level (% of Rated)
Technical Specification Scram Speed (TSSS), Nominal Scram Speed (NSS), or Optimum Scram Speed (OSS)
Cycle Operating Exposure (NEOC, EOC, and CD - as defined in this section)
Equipment Out-Of-Service Options Two or Single recirculation Loop Operation (TLO vs. SLO)
The MCPRP limits are provided in Table 4.2 through Table 4.9, where each table contains the limits for all fuel types and EOOS options (for a specified scram speed and exposure range).
The CMSS determines MCPRP limits, from these tables, based on linear interpolation between the specified powers.
4.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional power dependent limits are shown in Table 4.5 through Table 4.8 based on temperature conditions identified in Table 3.1.
Browns Ferry Unit 3 Cycle 20                                                                                      Page 13 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 4.2.2    Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS)
MCPRP limits are provided for three different sets of assumed scram speeds. The Technical Specification Scram Speed (TSSS) MCPRP limits are applicable at all times, as long as the scram time surveillance demonstrates the times in Technical Specification Table 3.1.4-1 are met. Both Nominal Scram Speeds (NSS) and/or Optimum Scram Speeds (OSS) may be used, as long as the scram time surveillance demonstrates Table 4.1 times are applicable.
* Table 4.1 Nominal Scram Time Basis Notch                  Nominal              Optimum Position            Scram Timing          Scram Timing (index)                (seconds)              (seconds) 46                        0.421                  0.392 36                        0.991                  0.887 26                        1.620                  1.487 6                      3.040                  3.040 In demonstrating compliance with the NSS and/or OSS scram time basis, surveillance requirements from Technical Specification 3.1.4 apply; accepting the definition of SLOW rods should conform to scram speeds shown in Table 4.1. If conformance is not demonstrated, TSSS based MCPRP limits are applied.
On initial cycle startup, TSSS limits are used until the successful completion of scram timing confirms NSS and/or OSS based limits are applicable.
4.2.3    Exposure Dependent Limits Exposures are tracked on a Core Average Exposure basis (CAVEX, not Cycle Exposure).
Higher exposure MCPRP limits are always more limiting and may be used for any Core Average Exposure up to the ending exposure. Per Reference 1, MCPRP limits are provided for the following exposure ranges:
BOC to NEOC                                NEOC corresponds to                        27,972.7 MWd / MTU BOC to EOCLB                                EOCLB corresponds to                        33,104.7 MWd / MTU BOC to End of Coast                        End of Coast                                34,799.5 MWd / MTU NEOC refers to a Near EOC exposure point.
* Reference 1 analysis results are based on information identified in Reference 4.
Drop out times consistent with method used to perform actual timing measurements (i.e., including pickup/dropout effects).
Browns Ferry Unit 3 Cycle 20                                                                                                    Page 14 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                      TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 The EOCLB exposure point is not the true End-Of-Cycle exposure. Instead it corresponds to a licensing exposure window exceeding expected end-of-full-power-life.
The End of Coast exposure point represents a licensing exposure point exceeding the expected end-of-cycle exposure including cycle extension options.
4.2.4      Equipment Out-Of-Service (EOOS) Options EOOS options
* covered by MCPRP limits are given by the following:
In-Service                                          All equipment In-Service RPTOOS                                              EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                              Turbine Bypass Valve(s) Out-Of-Service RPTOOS+TBVOOS                                      Combined RPTOOS and TBVOOS PLUOOS                                              Power Load Unbalance Out-Of-Service PLUOOS+RPTOOS                                      Combined PLUOOS and RPTOOS PLUOOS+TBVOOS                                      Combined PLUOOS and TBVOOS PLUOOS+TBVOOS+RPTOOS                                Combined PLUOOS, RPTOOS, and TBVOOS FHOOS (or FFWTR)                                    Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)
RCPOOS                                              One Recirculation Pump Out-Of-Service For exposure ranges up to NEOC and EOCLB, additional combinations of MCPRP limits are also provided including FHOOS. The coast down exposure range assumes application of FFWTR. FHOOS based MCPRP limits for the coast down exposure are redundant because the temperature setdown assumption is identical with FFWTR.
4.2.5      Single-Loop-Operation (SLO) Limits When operating in RCPOOS conditions, MCPRp limits are constructed differently from the normal operating RCP conditions. The limiting event for RCPOOS is a pump seizure scenario, which sets the upper bound for allowed core power and flow . This event is not impacted by scram time assumptions. Specific MCPRP limits are shown in Table 4.9.
4.2.6      Below Pbypass Limits Below Pbypass (26% rated power), MCPRP limits depend upon core flow. One set of MCPRP limits applies for core flow above 50% of rated; a second set applies if the core flow is less than or equal to 50% rated.
* All equipment service conditions assume 1 SRVOOS.
RCPOOS limits are only valid up to 43.75% rated core power, 50% rated core flow, and an active recirculation drive flow of 17.73 Mlbm/hr.
Browns Ferry Unit 3 Cycle 20                                                                                                        Page 15 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 2.00 1.80 1.60 MCPRF 1.40 1.20 1.00 30        40          50          60        70          80          90        100          110 Core Flow (% Rated)
Core Flow            MCPRF
(% Rated) 30.0              1.58 84.0              1.34 107.0              1.34 Figure 4.1 MCPRF for All Fuel Types (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 3 Cycle 20                                                                                          Page 16 Core Operating Limits Report, (120% OLTP, MELLLA+)                                            TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 4.2 MCPRP Limits for All Fuel Types: Optimum Scram Time Basis
* ATRIUM-10XM BOC          BOC        BOC Pow er            to            to      to End of Operating Condition      (% of rated)        NEOC        EOCLB        Coast 100              1.39          1.41        1.44 90              1.45          1.46        1.48 77.6            1.50          1.51        1.54 65              1.57          1.57        1.61
                                                        >50              1.65          1.65        1.70 50              1.79          1.79        1.79 Base Case 40              1.87          1.87        1.88 26              2.27          2.27        2.38 26 at > 50%F          2.60          2.60        2.70 23 at > 50%F          2.76          2.76        2.88
                                                    DW)          2.49          2.49        2.60
                                                    DW)          2.64          2.64        2.77 100              1.42          1.44          ---
90              1.48          1.48          ---
77.6            1.54          1.54          ---
65              1.61          1.61          ---
                                                        >50              1.70          1.70          ---
50              1.79          1.79          ---
FHOOS 40              1.88          1.88          ---
26              2.38          2.38          ---
26 at > 50%F          2.70          2.70          ---
23 at > 50%F          2.88          2.88          ---
                                                    DW)          2.60          2.60          ---
                                                    DW)          2.77          2.77          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR/FHOOS is supported for the BOC to End of Coast limits.
Browns Ferry Unit 3 Cycle 20                                                                                                    Page 17 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                                Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM BOC        BOC        BOC Pow er          to        to      to End of Operating Condition    (% of rated)    NEOC      EOCLB        Coast 100          1.42      1.44        1.46 90          1.48      1.48        1.51 77.6          1.53      1.53        1.56 65          1.59      1.59        1.63
                                                              >50          1.67      1.67        1.72 50          1.80      1.80        1.80 Base Case 40          1.88      1.88        1.90 26          2.30      2.30        2.41 26 at > 50%F      2.60      2.60        2.70 23 at > 50%F      2.76      2.76        2.88
                                                        DW)      2.49      2.49        2.60
                                                        DW)      2.64      2.64        2.77 100          1.46      1.47        1.50 90          1.51      1.51        1.54 77.6          1.56      1.56        1.60 65          1.62      1.62        1.66
                                                              >50          1.70      1.70        1.75 50          1.80      1.80        1.81 TBVOOS 40          1.88      1.88        1.91 26          2.30      2.30        2.42 26 at > 50%F      3.11      3.11        3.25 23 at > 50%F      3.36      3.36        3.50
                                                        DW)      2.83      2.83        2.99
                                                        DW)      3.11      3.11        3.28 100          1.46      1.46          ---
90          1.51      1.51          ---
77.6          1.56      1.56          ---
65          1.63      1.63          ---
                                                              >50          1.72      1.72          ---
50          1.80      1.80          ---
FHOOS 40          1.90      1.90          ---
26          2.41      2.41          ---
26 at > 50%F      2.70      2.70          ---
23 at > 50%F      2.88      2.88          ---
                                                        DW)      2.60      2.60          ---
                                                        DW)      2.77      2.77          ---
100          1.42      1.44        1.46 90          1.48      1.48        1.51 77.6          1.53      1.53        1.56 65          1.72      1.73        1.73
                                                              >50          ---        ---          ---
50          1.80      1.80        1.80 PLUOOS 40          1.88      1.88        1.90 26          2.30      2.30        2.41 26 at > 50%F      2.60      2.60        2.70 23 at > 50%F      2.76      2.76        2.88
                                                        DW)      2.49      2.49        2.60
                                                        DW)      2.64      2.64        2.77
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 3 Cycle 20                                                                                                      Page 18 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis (continued)
* ATRIUM-10XM BOC        BOC          BOC Pow er          to          to      to End of Operating Condition    (% of rated)    NEOC        EOCLB        Coast 100          1.49        1.50          ---
90          1.54        1.54          ---
77.6          1.60        1.60          ---
65          1.66        1.66          ---
                                                            >50          1.75        1.75          ---
TBVOOS              50          1.81        1.81          ---
FHOOS                40          1.91        1.91          ---
26          2.42        2.42          ---
26 at > 50%F      3.25        3.25          ---
23 at > 50%F      3.50        3.50          ---
                                                      DW)      2.99        2.99          ---
                                                      DW)      3.28        3.28          ---
100          1.46        1.47          1.50 90          1.51        1.51          1.54 77.6          1.56        1.56          1.60 65          1.72        1.73          1.74
                                                            >50          ---        ---          ---
TBVOOS              50          1.80        1.80          1.81 PLUOOS                40          1.88        1.88          1.91 26          2.30        2.30          2.42 26 at > 50%F      3.11        3.11          3.25 23 at > 50%F      3.36        3.36          3.50
                                                      DW)      2.83        2.83          2.99
                                                      DW)      3.11        3.11          3.28 100          1.46        1.46          ---
90          1.51        1.51          ---
77.6          1.56        1.56          ---
65          1.72        1.73          ---
                                                            >50          ---        ---          ---
FHOOS                50          1.80        1.80          ---
PLUOOS                40          1.90        1.90          ---
26          2.41        2.41          ---
26 at > 50%F      2.70        2.70          ---
23 at > 50%F      2.88        2.88          ---
                                                      DW)      2.60        2.60          ---
                                                      DW)      2.77        2.77          ---
100          1.49        1.50          ---
90          1.54        1.54          ---
77.6          1.60        1.60          ---
65          1.73        1.74          ---
                                                            >50          ---        ---          ---
TBVOOS 50          1.81        1.81          ---
FHOOS 40          1.91        1.91          ---
PLUOOS 26          2.42        2.42          ---
26 at > 50%F      3.25        3.25          ---
23 at > 50%F      3.50        3.50          ---
                                                      DW)      2.99        2.99          ---
                                                      DW)      3.28        3.28          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 3 Cycle 20                                                                                                      Page 19 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM BOC        BOC          BOC Pow er          to          to      to End of Operating Condition    (% of rated)    NEOC        EOCLB        Coast 100          1.46        1.46          1.50 90          1.50        1.50          1.55 77.6          1.55        1.55          1.59 65          1.61        1.61          1.66
                                                            >50          1.68        1.68          1.75 50          1.80        1.80          1.82 Base Case 40          1.88        1.88          1.94 26          2.32        2.32          2.44 26 at > 50%F      2.60        2.60          2.71 23 at > 50%F      2.76        2.76          2.89
                                                      DW)      2.49        2.49          2.61
                                                      DW)      2.64        2.64          2.78 100          1.50        1.50          1.54 90          1.55        1.55          1.59 77.6          1.59        1.59          1.64 65          1.65        1.65          1.70
                                                            >50          1.73        1.73          1.79 50          1.81        1.81          1.84 TBVOOS 40          1.89        1.89          1.96 26          2.34        2.34          2.47 26 at > 50%F      3.12        3.12          3.27 23 at > 50%F      3.37        3.37          3.52
                                                      DW)      2.84        2.84          3.01
                                                      DW)      3.12        3.12          3.30 100          1.50        1.50          ---
90          1.55        1.55          ---
77.6          1.59        1.59          ---
65          1.66        1.66          ---
                                                            >50          1.75        1.75          ---
50          1.81        1.81          ---
FHOOS 40          1.94        1.94          ---
26          2.44        2.44          ---
26 at > 50%F      2.71        2.71          ---
23 at > 50%F      2.89        2.89          ---
                                                      DW)      2.61        2.61          ---
                                                      DW)      2.78        2.78          ---
100          1.46        1.46          1.50 90          1.50        1.50          1.55 77.6          1.55        1.55          1.59 65          1.73        1.74          1.75
                                                            >50          ---        ---          ---
50          1.80        1.80          1.82 PLUOOS 40          1.88        1.88          1.94 26          2.32        2.32          2.44 26 at > 50%F      2.60        2.60          2.71 23 at > 50%F      2.76        2.76          2.89
                                                      DW)      2.49        2.49          2.61
                                                      DW)      2.64        2.64          2.78
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 3 Cycle 20                                                                                                      Page 20 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis (continued)
* ATRIUM-10XM BOC        BOC          BOC Pow er          to          to      to End of Operating Condition    (% of rated)    NEOC        EOCLB        Coast 100          1.54        1.54          ---
90          1.59        1.59          ---
77.6          1.64        1.64          ---
65          1.70        1.70          ---
                                                            >50          1.79        1.79          ---
TBVOOS              50          1.83        1.83          ---
FHOOS                40          1.96        1.96          ---
26          2.47        2.47          ---
26 at > 50%F      3.27        3.27          ---
23 at > 50%F      3.52        3.52          ---
                                                        DW)      3.01        3.01          ---
                                                        DW)      3.30        3.30          ---
100          1.50        1.50          1.54 90          1.55        1.55          1.59 77.6          1.59        1.59          1.64 65          1.74        1.75          1.77
                                                            >50          ---        ---          ---
TBVOOS              50          1.81        1.81          1.84 PLUOOS                40          1.89        1.89          1.96 26          2.34        2.34          2.47 26 at > 50%F      3.12        3.12          3.27 23 at > 50%F      3.37        3.37          3.52
                                                        DW)      2.84        2.84          3.01
                                                        DW)      3.12        3.12          3.30 100          1.50        1.50          ---
90          1.55        1.55          ---
77.6          1.59        1.59          ---
65          1.74        1.75          ---
                                                            >50          ---        ---          ---
FHOOS                50          1.81        1.81          ---
PLUOOS                40          1.94        1.94          ---
26          2.44        2.44          ---
26 at > 50%F      2.71        2.71          ---
23 at > 50%F      2.89        2.89          ---
                                                        DW)      2.61        2.61          ---
                                                        DW)      2.78        2.78          ---
100          1.54        1.54          ---
90          1.59        1.59          ---
77.6          1.64        1.64          ---
65          1.76        1.77          ---
                                                            >50          ---        ---          ---
TBVOOS 50          1.83        1.83          ---
FHOOS 40          1.96        1.96          ---
PLUOOS 26          2.47        2.47          ---
26 at > 50%F      3.27        3.27          ---
23 at > 50%F      3.52        3.52          ---
                                                        DW)      3.01        3.01          ---
                                                        DW)      3.30        3.30          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 3 Cycle 20                                                                                                      Page 21 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 4.5 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM BOC        BOC        BOC Pow er          to          to    to End of Operating Condition      (% of rated)    NEOC        EOCLB      Coast 100          1.46        1.46      1.46 90          1.51        1.51      1.51 77.6          1.56        1.56      1.56 65          1.72        1.73      1.73
                                                          >50            ---        ---        ---
50          1.84        1.84      1.84 TBVIS 40          2.07        2.07      2.07 26          2.66        2.66      2.66 26 at > 50%F      2.92        2.92      2.92 23 at > 50%F      3.14        3.14      3.14
                                                      DW)      2.82        2.82      2.82
                                                      DW)      3.04        3.04      3.04 100          1.49        1.50      1.50 90          1.54        1.54      1.54 77.6          1.60        1.60      1.60 65          1.73        1.74      1.74
                                                          >50            ---        ---        ---
50          1.85        1.85      1.85 TBVOOS 40          2.08        2.08      2.08 26          2.67        2.67      2.67 26 at > 50%F      3.44        3.44      3.44 23 at > 50%F      3.69        3.69      3.69
                                                      DW)      3.18        3.18      3.18
                                                      DW)      3.51        3.51      3.51
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 3 Cycle 20                                                                                                  Page 22 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 4.6 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types: Nominal Scram Time Basis
* ATRIUM-10XM BOC        BOC        BOC Pow er          to          to    to End of Operating Condition      (% of rated)    NEOC        EOCLB      Coast 100          1.46        1.46      1.46 90          1.51        1.51      1.51 77.6          1.56        1.56      1.56 65          1.72        1.73      1.73
                                                          >50            ---        ---        ---
50          1.85        1.85      1.85 TBVIS 40          2.08        2.08      2.08 26          2.68        2.68      2.68 26 at > 50%F      2.94        2.94      2.94 23 at > 50%F      3.15        3.15      3.15
                                                      DW)      2.84        2.84      2.84
                                                      DW)      3.06        3.06      3.06 100          1.49        1.50      1.50 90          1.54        1.54      1.54 77.6          1.60        1.60      1.60 65          1.73        1.74      1.74
                                                          >50            ---        ---        ---
50          1.86        1.86      1.86 TBVOOS 40          2.09        2.09      2.09 26          2.69        2.69      2.69 26 at > 50%F      3.45        3.45      3.45 23 at > 50%F      3.71        3.71      3.71
                                                      DW)      3.20        3.20      3.20
                                                      DW)      3.52        3.52      3.52
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 3 Cycle 20                                                                                                  Page 23 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 4.7 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM BOC        BOC        BOC Pow er          to          to    to End of Operating Condition      (% of rated)    NEOC        EOCLB      Coast 100          1.50        1.50      1.50 90          1.55        1.55      1.55 77.6          1.59        1.59      1.59 65          1.74        1.75      1.75
                                                          >50            ---        ---        ---
50          1.88        1.88      1.88 TBVIS 40          2.11        2.11      2.11 26          2.70        2.70      2.70 26 at > 50%F      2.93        2.93      2.93 23 at > 50%F      3.15        3.15      3.15
                                                      DW)      2.83        2.83      2.83
                                                      DW)      3.05        3.05      3.05 100          1.54        1.54      1.54 90          1.59        1.59      1.59 77.6          1.64        1.64      1.64 65          1.76        1.77      1.77
                                                          >50            ---        ---        ---
50          1.90        1.90      1.90 TBVOOS 40          2.13        2.13      2.13 26          2.72        2.72      2.72 26 at > 50%F      3.46        3.46      3.46 23 at > 50%F      3.71        3.71      3.71
                                                      DW)      3.20        3.20      3.20
                                                      DW)      3.53        3.53      3.53
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 3 Cycle 20                                                                                                  Page 24 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 4.8 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types: Technical Specification Scram Time Basis
* ATRIUM-10XM BOC        BOC        BOC Pow er          to          to    to End of Operating Condition      (% of rated)    NEOC        EOCLB      Coast 100          1.50        1.50      1.50 90          1.55        1.55      1.55 77.6          1.59        1.59      1.59 65          1.74        1.75      1.75
                                                          >50            ---        ---        ---
50          1.89        1.89      1.89 TBVIS 40          2.12        2.12      2.12 26          2.72        2.72      2.72 26 at > 50%F      2.95        2.95      2.95 23 at > 50%F      3.16        3.16      3.16
                                                      DW)      2.85        2.85      2.85
                                                      DW)      3.07        3.07      3.07 100          1.54        1.54      1.54 90          1.59        1.59      1.59 77.6          1.64        1.64      1.64 65          1.76        1.77      1.77
                                                          >50            ---        ---        ---
50          1.91        1.91      1.91 TBVOOS 40          2.14        2.14      2.14 26          2.74        2.74      2.74 26 at > 50%F      3.47        3.47      3.47 23 at > 50%F      3.73        3.73      3.73
                                                      DW)      3.22        3.22      3.22
                                                      DW)      3.54        3.54      3.54
* Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
Limits are applicable for all other EOOS scenarios, apart from TBV.
Limits are only valid up to 50% rated core power.
Browns Ferry Unit 3 Cycle 20                                                                                                  Page 25 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 4.9 MCPRP Limits for All Fuel Types: Single Loop Operation for All Scram Times
* Pow er          BOC to End of COAST Operating Condition      (% of rated)            ATRIUM-10XM 100                2.10 43.75                2.10 40                2.10 RCPOOS                        26                2.46 FHOOS            26 at > 50%F                  2.73 23 at > 50%F                  2.91
                                                          DW)                  2.63
                                                          DW)                  2.80 100                2.10 43.75                2.10 RCPOOS                        40                2.10 TBVOOS                        26                2.49 PLUOOS            26 at > 50%F                  3.29 FHOOS            23 at > 50%F                  3.54
                                                          DW)                  3.03
                                                          DW)                  3.32 100                2.15 43.75                2.15 40                2.15 RCPOOS 26                2.74 TBVOOS 26 at > 50%F                  3.48 FHOOS1 23 at > 50%F                  3.73
                                                          DW)                  3.22
                                                          DW)                  3.55 100                2.16 43.75                2.16 40                2.16 RCPOOS 26                2.76 TBVOOS 26 at > 50%F                  3.49 FHOOS2 23 at > 50%F                  3.75
                                                          DW)                  3.24
                                                          DW)                  3.56
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop.
RCPOOS limits are only valid up to 50% rated core power, 50% rated core flow, and an active recirculation drive flow of 17.73 Mlbm/hr.
Browns Ferry Unit 3 Cycle 20                                                                                                      Page 26 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 5      Thermal-Hydraulic Stability Protection (Technical Specification 3.3.1.1)
Technical Specification Table 3.3.1.1-1, Function 2f, identifies the function.
Instrument setpoints are established, such that the reactor will be tripped before an oscillation can grow to the point where the SLMCPR is exceeded. With application of Reference 30, the DSS-CD stability solution will be used per Reference 26. The DSS-CD SAD setpoint is 1.10 for TLO and SLO.
New analyses have been developed based on Reference 26. With the implementation of the MELLLA+ operating domain expansion, an ABSP trip is required when the OPRM is out-of-service. The ABSP trip settings define a region of the power to flow map within which an automatic reactor scram occurs. The ABSP trip settings are provided in Table 5.1. If both the OPRM and ABSP are out-of-service, operation within the MELLLA+ domain is not allowed and the MBSP Regions provide stability protection. Table 5.2 and Table 5.3 provide the endpoints for the MBSP regions for nominal and reduced feedwater temperature conditions.
Table 5.1 ABSP Setpoints for the Scram Region Parameter                Symbol          Setting Value (unit)                      Comments Slope of ABSP APRM low Flow Slope for Trip              m TRIP        2.00 (% RTP/% RDF)
Biased Trip Linear Segment ABSP APRM Flow Biased Trip Constant Power                                                            Setpoint Power Intercept. Constant PBSP-TRIP          35.0 (% RTP)
Line for Trip                                                          Power Line for Trip from Zero Drive Flow to Flow Breakpoint Value ABSP APRM Flow Biased Trip Constant Flow                                                            Setpoint Drive Flow Intercept.
WBSP-TRIP            49 (% RDF)
Line for Trip                                                          Constant Flow Line for Trip (see Note 1 below)
Flow Breakpoint              WBSP-BREAK          30.0 (% RDF)              Flow Breakpoint Value Note 1: WBSP-T RIP can be set to 49.0 % RDF or any higher value up to the intersection of the ABSP sloped line w ith the APRM Flow Biased STP scram line.
Browns Ferry Unit 3 Cycle 20                                                                                              Page 27 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 5.2 Analyzed MBSP Endpoints: Nominal Feedwater Temperature Power      Core Flow Endpoint                                                Definition
(% Rated)    (% Rated)
Scram Region (Region I)
A1            75.9        52.7          Boundary Intercept on MELLLA+ Line Scram Region (Region I)
B1            35.5        29.0          Boundary Intercept on Natural Circulation Line (NCL)
Controlled Entry Region (Region A2            66.1        52.0          II) Boundary Intercept on MELLLA Line Controlled Entry Region (Region B2            25.5        29.0          II) Boundary Intercept on Natural Circulation Line (NCL)
Table 5.3 Analyzed MBSP Endpoints: Reduced Feedwater Temperature Power      Core Flow Endpoint                                                Definition
(% Rated)    (% Rated)
Scram Region (Region I)
A1            64.9        50.5          Boundary Intercept on MELLLA Line Scram Region (Region I)
B1            29.4        29.0          Boundary Intercept on Natural Circulation Line (NCL)
Controlled Entry Region (Region A2            68.3        54.9          II) Boundary Intercept on MELLLA Line Controlled Entry Region (Region B2            24.5        29.0          II) Boundary Intercept on Natural Circulation Line (NCL)
Browns Ferry Unit 3 Cycle 20                                                                                        Page 28 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 6 APRM Flow Biased Rod Block Trip Settings (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
The APRM rod block trip setting is based upon References 27 & 29, and is defined by the following:
for two loop operation:
SRB d (0.61Wd + 63.3)                              Allowable Value SRB d (0.61Wd + 62.0)                              Nominal Trip Setpoint (NTSP) where:
SRB          =          Rod Block setting in percent of rated thermal power (3952 MWt)
Wd          =          Recirculation drive flow rate in percent of rated (100% drive flow required to achieve 100% core power and flow) and for single loop operation:
SRB d (0.55(Wd-'W) + 60.5)                                  Allowable Value SRB d (0.55(Wd-'W) + 58.5)                                  Nominal Trip Setpoint (NTSP) where:
SRB          =          Rod Block setting in percent of rated thermal power (3952 MWt)
Wd          =          Recirculation drive flow rate in percent of rated (100% drive flow required to achieve 100% core power and flow)
          'W          =          Difference between two-loop and single-loop effective recirculation flow at the same core flow ('W=0.0 for two-loop operation)
The APRM rod block trip setting is clamped at a maximum allowable value of 115%
(corresponding to a NTSP of 113%).
Browns Ferry Unit 3 Cycle 20                                                                                      Page 29 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
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[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 7 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
The RBM trip setpoints and applicable power ranges, based on References 27 & 28, are shown in Table 7.1. Setpoints are based on an HTSP, unfiltered analytical limit of 114%. Unfiltered setpoints are consistent with a nominal RBM filter setting of 0.0 seconds; filtered setpoints are consistent with a nominal RBM filter setting less than 0.5 seconds. Cycle specific CRWE analyses of OLMCPR are documented in Reference 1, superseding values reported in References 27, 28, and 29.
Table 7.1 Analytical RBM Trip Setpoints
* Allowable              Nominal Trip RBM                              Value                Setpoint Trip Setpoint                          (AV)                  (NTSP)
LPSP                                    27%                    25%
IPSP                                    62%                    60%
HPSP                                    82%                    80%
LTSP - unfiltered                    121.7%                  120.0%
                                          - filtered                    120.7%                  119.0%
ITSP    - unfiltered                  116.7%                  115.0%
                                          - filtered                    115.7%                  114.0%
HTSP - unfiltered                    111.7%                  110.0%
                                          - filtered                    110.9%                  109.2%
DTSP                                    90%                    92%
As a result of cycle specific CRWE analyses, RBM setpoints in Technical Specification Table 3.3.2.1-1 are applicable as shown in Table 7.2. Cycle specific setpoint analysis results are shown in Table 7.3, per Reference 1.
Table 7.2 RBM Setpoint Applicability Thermal Power                      Applicable                Notes from
(% Rated)                        MCPR                Table 3.3.2.1-1              Comment
                                                        < 1.74              (a), (b), (f), (h)    two loop operation
            > 27% and < 90%
                                                        < 1.78              (a), (b), (f), (h)    single loop operation 90%                                      < 1.38                      (g)            two loop operation
* Values are considered maximums. Using lower values, due to RBM system hardware/software limitations, is conservative, and acceptable.
MCPR values shown correspond with, (support), SLMPCR values identified in Reference 1.
Greater than 90% rated power is not attainable in single loop operation.
Browns Ferry Unit 3 Cycle 20                                                                                                  Page 30 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                  TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Table 7.3 Control Rod Withdrawal Error Results RBM                        CRWE HTSP Analytical Limit                OLMCPR Unfiltered 107                          1.26 111                          1.28 114                          1.30 117                          1.36 Results, compared against the base case OLMCPR results of Table 4.2, indicate SLMCPR remains protected for RBM inoperable conditions (i.e., 114% unblocked).
Browns Ferry Unit 3 Cycle 20                                                                                      Page 31 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800 mil NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 8 Shutdown Margin Limit (Technical Specification 3.1.1)
Assuming the strongest OPERABLE control blade is fully withdrawn, and all other OPERABLE control blades are fully inserted, the core shall be sub-critical and meet the following minimum shutdown margin:
SDM      > 0.38% dk/k Note: The basis for the SDM value is tied to manufacturing tolerance and uncertainty stack-up. This tie only has the potential to change with new fuel type introduction. Since no new fuel type is being introduced, the value is good for all Mode 3, 4, and 5 operation prior to initial cycle startup, during cycle operation, and after final cycle shutdown.
Browns Ferry Unit 3 Cycle 20                                                                                      Page 32 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Appendix A: MBSP Maps Browns Ferry Unit 3 Cycle 20                                                                                      Page A-1 Core Operating Limits Report, (120% OLTP, MELLLA+)                                        TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                                  Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Core Power (% Rated: 100% = 3952MWt) 110 100 MELLLA+ Region 90 80                                                                                  BSP Boundary Manual 70                                    Scram Region I 60 MELLLA Region    ICF Region MELLLA Upper Boundary 50 87.5% Rod Line 40                                                        Controlled Entry Region II 30                                                              Min. Flow Control 20 Natural                                Min. Pow er Line Circulation 10                                            20% Pump Speed Line 0
0          10          20        30        40        50        60      70    80      90    100    110        120 Core Flow (% Rated: 100% =102.5 MLbm/hr)
Figure A.1 MBSP Boundaries For Nominal Feedwater Temperature (Operation in the MELLLA+ Region Prohibited for Feedwater Temperature greater than 10 degrees F below the Nominal Feedwater Temperature)
Browns Ferry Unit 3 Cycle 20                                                                                            Page A-2 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF3C20, Revision 0 (Final)
 
EDMS: L94 200213 800
[lI1 NPG                                  Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: February 18, 2020 Core Power (% Rated: 100% = 3952MWt) 110 100 MELLLA+ Region 90 80 BSP Boundary 70 Manual Scram 60                                    Region I MELLLA Region MELLLA Upper Boundary                                            ICF Region 50 87.5% Rod Line Controlled Entry 40                                                                  Region II 30                                                              Min. Flow Control 20 Natural                              Min. Pow er Line Circulation 10                                            20% Pump Speed Line 0
0          10          20        30        40        50      60        70      80      90      100    110        120 Core Flow (% Rated: 100% =102.5 MLbm/hr)
Figure A.2 MBSP Boundaries For Reduced Feedwater Temperature (Operation in the MELLLA+ Region Prohibited for a Reduced Feedwater Temperature greater than 10 degrees F below the Nominal Feedwater Temperature)
Browns Ferry Unit 3 Cycle 20                                                                                            Page A-3 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF3C20, Revision 0 (Final)
 
ECM L32 200817 800 QA RECORD BFE-4485 . Revision 1
  &#xa3;mlNPG Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga, TN 37402 Bro wns Ferr y Unit 3 Cyc le 20 Core Opera ting Limits Report, (120% OLTP, MELL LA+)
TVA-COL R-BF3 C20                      Revision 1 (Final)
(Revision Log. Page v)
August2020
                                      . h II                    Digitally si gned by M t<hell. Brye Mite e , Brye C. c.Date: 2020.08 25 14:24:le>-Sft?.l' Prepared:                                                                                        ------
8 . C. Mitchell, Engineer Digitally signed by Roberts, Claude Roberts, Claude C Jr                        c1 r Date:          2019cfll26 09:33:36 -04'00' Verified:
                              ---        ---Engineer C. C. Roberts,        --- ----
Oi9 ,, fy \i'Jnt'db)' ~,.1tt1n , ttviu 1nt,.
UN: dl ~QOV, O<
* t~. d ( " rTl.llf'I, ou ;iM.fn. ov-(N p.,1.#tt, o..,..,.Uu*r\
o, .- ~eilfe,,. ,,,,,    /\ ... n ..lu>,H~,.. ,t ~ QOV r<HM>n I m p o,CN11t 9 ' ""' d<!<ur,, nt Approved: _
_ _ _ _ _ _ _ _ __ _o.._,.                        )Q)008]~ t i j :D'_ _ _ _ __
C. A. Setter, Manager, BWRFE Reviewed:    --<~                                                              Date:              q/      /1, / 20 D. D. Coffey, Manager, Reactor Engineering Approved:
Date: _ _9/_2_8/_2_0_20                      __
Approved :
BFN - Unit 3                                      Page 1 of 44                                                      TRM Revision 146 September 23, 2020
 
ECM: L32 200817 800
[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 Table of Contents Total Number of Pages = 44 (including review cover sheet)
List of Tables.................................................................................................................................iii List of Figures .............................................................................................................................. iv Revision Log ................................................................................................................................. v Nomenclature ............................................................................................................................... vi References.................................................................................................................................. viii 1    Introduction .......................................................................................................................... 1
1.1      Purpose ......................................................................................................................... 1
1.2      Scope ............................................................................................................................ 1
1.3      Fuel Loading.................................................................................................................. 1
1.4      Acceptability .................................................................................................................. 2
2    APLHGR Limits .................................................................................................................... 3
2.1      Rated Power and Flow Limit: APLHGRRATED ................................................................ 3
2.2      Off-Rated Power Dependent Limit: APLHGRP ............................................................. 3
2.2.1          Startup without Feedwater Heaters ....................................................................... 3
2.3      Off-Rated Flow Dependent Limit: APLHGRF ................................................................ 3
2.4      Single Loop Operation Limit: APLHGRSLO.................................................................... 3
2.5      Equipment Out-Of-Service Corrections ......................................................................... 5
3    LHGR Limits ......................................................................................................................... 6
3.1      Rated Power and Flow Limit: LHGRRATED..................................................................... 6
3.2      Off-Rated Power Dependent Limit: LHGRP .................................................................. 6
3.2.1          Startup without Feedwater Heaters ....................................................................... 6
3.3      Off-Rated Flow Dependent Limit: LHGRF..................................................................... 7
3.4      Equipment Out-Of-Service Corrections ......................................................................... 7
4    OLMCPR Limits ................................................................................................................. 13
4.1      Flow Dependent MCPR Limit: MCPRF ....................................................................... 13
4.2      Power Dependent MCPR Limit: MCPRP .................................................................... 13
4.2.1          Startup without Feedwater Heaters ..................................................................... 13
4.2.2          Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS) ................................. 14
4.2.3          Exposure Dependent Limits ................................................................................ 14
4.2.4          Equipment Out-Of-Service (EOOS) Options ....................................................... 15
4.2.5          Single-Loop-Operation (SLO) Limits ................................................................... 15
4.2.6          Below Pbypass Limits ......................................................................................... 15
5    Thermal-Hydraulic Stability Protection ............................................................................... 27
6    APRM Flow Biased Rod Block Trip Settings ...................................................................... 29
7    Rod Block Monitor (RBM) Trip Setpoints and Operability .................................................. 30
8    Shutdown Margin Limit....................................................................................................... 32
Appendix A:            MBSP Maps ...................................................................................................... A-1
Browns Ferry Unit 3 Cycle 20                                                                                                            Page ii Core Operating Limits Report, (120% OLTP, MELLLA+)                                                      TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 List of Tables Nuclear Fuel Types0F0F0F ................................................................................................................. 2
Startup Feedwater Temperature Basis ......................................................................................... 6
Nominal Scram Time Basis ......................................................................................................... 14
MCPRP Limits for All Fuel Types: Optimum Scram Time Basis7F7F7F ............................................ 17
MCPRP Limits for All Fuel Types: Nominal Scram Time Basis8F8F8F .............................................. 18
MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis10F10F10F ...................... 20
Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Nominal Scram Time Basis12F12F12F ................................................................................................... 22
Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Nominal Scram Time Basis13F13F13F ................................................................................................... 23
Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types:
Technical Specification Scram Time Basis14F14F14F............................................................................ 24
Startup Operation MCPRP Limits for Table 3.1 Temperature Range 2 for All Fuel Types:
Technical Specification Scram Time Basis15F15F15F............................................................................ 25
MCPRP Limits for All Fuel Types: Single Loop Operation for All Scram Times16F16F16F .................... 26
ABSP Setpoints for the Scram Region ....................................................................................... 27
Analyzed MBSP Endpoints: Nominal Feedwater Temperature ................................................. 28
Analyzed MBSP Endpoints: Reduced Feedwater Temperature ................................................ 28
Analytical RBM Trip Setpoints18F18F17F ............................................................................................... 30
RBM Setpoint Applicability .......................................................................................................... 30
Control Rod Withdrawal Error Results ........................................................................................ 31
Browns Ferry Unit 3 Cycle 20                                                                                                      Page iii Core Operating Limits Report, (120% OLTP, MELLLA+)                                                TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 List of Figures APLHGRRATED for ATRIUM-10XM Fuel ......................................................................................... 4
LHGRRATED for ATRIUM-10XM Fuel .............................................................................................. 8
Base Operation LHGRFACP for ATRIUM-10XM Fuel ................................................................... 9
LHGRFACF for ATRIUM-10XM Fuel ........................................................................................... 10
Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 1 ........ 11
Startup Operation LHGRFACP for ATRIUM-10XM Fuel: Table 3.1 Temperature Range 2 ........ 12
MCPRF for All Fuel Types ........................................................................................................... 16
MBSP Boundaries For Nominal Feedwater Temperature ........................................................ A-2
MBSP Boundaries For Reduced Feedwater Temperature ....................................................... A-3
Browns Ferry Unit 3 Cycle 20                                                                                                  Page iv Core Operating Limits Report, (120% OLTP, MELLLA+)                                            TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 Revision Log Number            Page                                          Description 1-R1            14        Updated Table 4.1 Nominal Scram Time Basis per CR 1604385 2-R1            32        Updated  symbol per CR 1596417 0-R0            All        New document.
Browns Ferry Unit 3 Cycle 20                                                                                          Page v Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 Nomenclature ABSP                  Automatic Backup Stability Protection APLHGR                Average Planar LHGR APRM                  Average Power Range Monitor AREVA NP              Vendor (Framatome, Siemens)
BOC                    Beginning of Cycle BSP                    Backup Stability Protection BWR                    Boiling Water Reactor CAVEX                  Core Average Exposure CD                    Coast Down CMSS                  Core Monitoring System Software COLR                  Core Operating Limits Report CPR                    Critical Power Ratio CRWE                  Control Rod Withdrawal Error CSDM                  Cold SDM DIVOM                  Delta CPR over Initial CPR vs. Oscillation Magnitude DSS-CD                Detect and Suppress Solution - Confirmation Density EOC                    End of Cycle EOCLB                  End-of-Cycle Licensing Basis EOOS                  Equipment OOS EPU                    Extended Power Uprate (120% OLTP)
FFTR                  Final Feedwater Temperature Reduction FFWTR                  Final Feedwater Temperature Reduction FHOOS                  Feedwater Heaters OOS ft                    Foot: English unit of measure for length GNF                    Vendor (General Electric, Global Nuclear Fuels)
GWd                    Giga Watt Day HTSP                  High TSP ICA                    Interim Corrective Action ICF                    Increased Core Flow (beyond rated)
IS                    In-Service kW                    kilo watt: SI unit of measure for power.
LCO                    License Condition of Operation LFWH                  Loss of Feedwater Heating LHGRFAC                LHGR Multiplier (Power or Flow dependent)
LPRM                  Low Power Range Monitor LRNB                  Generator Load Reject, No Bypass Browns Ferry Unit 3 Cycle 20                                                                                        Page vi Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 MAPFAC                MAPLHGR multiplier (Power or Flow dependent)
MBSP                  Manual Backup Stability Protection MCPR                  Minimum CPR MELLLA                Maximum Extended Load Line Limit Analysis MELLLA+                Maximum Extended Load Line Limit Analysis Plus MSRV                  Moisture Separator Reheater Valve MSRVOOS                MSRV OOS MTU                    Metric Ton Uranium MWd/MTU                Mega Watt Day per Metric Ton Uranium NEOC                  Near EOC NRC                    United States Nuclear Regulatory Commission NSS                    Nominal Scram Speed NTSP                  Nominal TSP OLMCPR                MCPR Operating Limit OLTP                  Original Licensed Thermal Power OOS                    Out-Of-Service OPRM                  Oscillation Power Range Monitor OSS                    Optimum Scram Speed PBDA                  Period Based Detection Algorithm Pbypass                Power, below which TSV Position and TCV Fast Closure Scrams are Bypassed PLU                    Power Load Unbalance PLUOOS                PLU OOS PRNM                  Power Range Neutron Monitor RBM                    Rod Block Monitor RCPOOS                Recirculation Pump OOS (SLO)
RDF                    Rated Drive Flow RPS                    Reactor Protection System RPT                    Recirculation Pump Trip RPTOOS                RPT OOS RTP                    Rated Thermal Power SDM                    Shutdown Margin SLMCPR                MCPR Safety Limit SLO                    Single Loop Operation TBV                    Turbine Bypass Valve TBVIS                  TBV IS TBVOOS                Turbine Bypass Valves OOS TIP                    Transversing In-core Probe TIPOOS                TIP OOS TLO                    Two Loop Operation TSP                    Trip Setpoint TSSS                  Technical Specification Scram Speed TVA                    Tennessee Valley Authority Browns Ferry Unit 3 Cycle 20                                                                                        Page vii Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 References
: 1.        ANP-3813, Revision 0, Browns Ferry Unit 3 Cycle 20 Reload Analysis, Framatome Inc., January 2020.
: 2.        Not Used.
: 3.        ANP-3150P, Revision 4, Mechanical Design Report for Browns Ferry ATRIUM 10XM Fuel Assemblies, AREVA Inc., November 2017.
: 4.        ANP-3793P Revision 0, Browns Ferry Unit 3 Cycle 20 Plant Parameters Document, Framatome Inc., June 2019.
: 5.        BFE-4468, Revision 0, Browns Ferry Unit 3 Cycle 20 In-Core Shuffle, Tennessee Valley Authority, January 31, 2020.
Methodology References
: 6.        XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
: 7.        XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
: 8.        EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
: 9.        ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
: 10.      XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.
: 11.      XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.
: 12.      EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.
: 13.      XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.
: 14.      XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.
: 15.      ANP-10307PA, Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP Inc., June 2011.
Browns Ferry Unit 3 Cycle 20                                                                                        Page viii Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020
: 16.      ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.
: 17.      ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Advanced Nuclear Fuels Corporation, September 2005.
: 18.      EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP Inc.,
September 2009.
: 19.      EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001, as supplemented by the site specific approval in NRC safety evaluations dated February 15, 2013 and July 31, 2014.
: 20.      EMF-2292(P)(A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
: 21.      EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
: 22.      BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP Inc., May 2008.
: 23.      BAW-10247PA, Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP Inc., April 2008.
: 24.      ANP-10298PA, Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc., March 2010.
: 25.      ANP-3140(P), Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc.,
August 2012.
: 26.      NEDC-33075P-A, Revision 8, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, GE Hitachi, November 2013.
Setpoint References
: 27.      EDQ2092900118, R35, Setpoint and Scaling Calculation for Neutron Monitoring &
Recirculation Flow Loops, Calculation File, Tennessee Valley Authority, August 9, 2019.
: 28.      Task T0500, Revision 0, Neutron Monitoring System w/RBM, Project Task Report, GE Hitachi Nuclear Energy, June 2017.
: 29.      Task T0506, Revision 0, TS Instrument Setpoints, Project Task Report, Tennessee Valley Authority, August, 2017.
: 30.      NEDC-33006P-A, Revision 3, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, GE Energy Nuclear, June 2009.
Browns Ferry Unit 3 Cycle 20                                                                                        Page ix Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 1 Introduction In anticipation of cycle startup, it is necessary to describe the expected limits of operation.
1.1    Purpose The primary purpose of this document is to satisfy requirements identified by unit technical specification section 5.6.5. This document may be provided, upon final approval, to the NRC.
1.2    Scope This document will discuss the following areas:
3/4 Average Planar Linear Heat Generation Rate (APLHGR) Limit (Technical Specifications 3.2.1 and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3.2.3, 3.3.4.1, and 3.7.5)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Minimum Critical Power Ratio Operating Limit (OLMCPR)
(Technical Specifications 3.2.2, 3.3.4.1, 3.7.5 and Table 3.3.2.1-1)
Applicability: Mode 1,  23% RTP (Technical Specifications definition of RTP) 3/4 Thermal-Hydraulic Stability Protection (Technical Specification Table 3.3.1.1)
Applicability: Mode 1,  (as specified in Technical Specifications Table 3.3.1.1-1) 3/4 Average Power Range Monitor (APRM) Flow Biased Rod Block Trip Setting (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
Applicability: Mode 1,  (as specified in Technical Requirements Manuals Table 3.3.4-1) 3/4 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
Applicability: Mode 1,  % RTP as specified in Table 3.3.2.1-1 (TS definition of RTP) 3/4 Shutdown Margin (SDM) Limit (Technical Specification 3.1.1)
Applicability: All Modes 1.3    Fuel Loading The core will contain fresh, and previously exposed ATRIUM-10XM. Nuclear fuel types used in the core loading are shown in Table 1.1. The core shuffle and final loading were explicitly evaluated for BOC cold shutdown margin performance as documented per Reference 5.
Browns Ferry Unit 3 Cycle 20                                                                                          Page 1 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                                Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 Table 1.1 Nuclear Fuel Types
* 0F0F0F Nuclear Original        Number of          Fuel Type            Fuel Names Fuel Description                              Cycle          Assemblies            (NFT)                (Range)
ATRIUM-10XM XMLC-4105B-11GV70-FCG                                    18                  72                19            FCG601-FCG672 ATRIUM-10XM XMLC-4096B-12GV80-FCG                                    18                  22                20            FCG673-FCG808 ATRIUM-10XM XMLC-4055B-13GV70-FCG                                    18                  17                21            FCG809-FCG904 ATRIUM-10XM XMLC-3911B-13GV80-FCH                                    19                238                22            FCH001-FCH240 ATRIUM-10XM XMLC-4053B-12GV80-FCH                                    19                103                23            FCH241-FCH344 ATRIUM-10XM XMLC-3920B-14GV80-FCJ                                    20                224                24              FCJ345-FCJ568 ATRIUM-10XM XMLC-3957B-12GV80-FCJ                                    20                  88                25              FCJ569-FCJ656 1.4      Acceptability Limits discussed in this document were generated based on NRC approved methodologies per References 6 through 25.
* The table identifies the expected fuel type breakdown in anticipation of final core loading. The final composition of the core depends upon uncertainties during the outage such as discovering a failed fuel bundle, or other bundle damage. Minor core loading changes, due to unforeseen events, will conform to the safety and monitoring requirements identified in this document.
Browns Ferry Unit 3 Cycle 20                                                                                                            Page 2 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                          TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 2 APLHGR Limits (Technical Specifications 3.2.1 & 3.7.5)
The APLHGR limit is determined by adjusting the rated power APLHGR limit for off-rated power, off-rated flow, and SLO conditions. The most limiting of these is then used as follows:
APLHGR limit = MIN ( APLHGRP , APLHGRF, APLHGRSLO )
where:
APLHGRP                off-rated power APLHGR limit                  [APLHGRRATED
* MAPFACP]
APLHGRF                off-rated flow APLHGR limit                    [APLHGRRATED
* MAPFACF]
APLHGRSLO              SLO APLHGR limit                              [APLHGRRATED
* SLO Multiplier]
2.1    Rated Power and Flow Limit: APLHGRRATED The rated conditions APLHGR for all fuel are identified per Reference 1. The rated conditions APLHGR for ATRIUM-10XM are shown in Figure 2.1.
2.2    Off-Rated Power Dependent Limit: APLHGRP Reference 1 does not specify a power dependent APLHGR. Therefore, MAPFACP is set to a value of 1.0.
2.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. No additional power dependent limitation is required.
2.3    Off-Rated Flow Dependent Limit: APLHGRF Reference 1 does not specify a flow dependent APLHGR. Therefore, MAPFACF is set to a value of 1.0.
2.4    Single Loop Operation Limit: APLHGRSLO The single loop operation multiplier for ATRIUM-10XM fuel is 0.85, per Reference 1.
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[i!D NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 15 12 APLHGR (kW/ft) 9 6
3 0
0          20                      40                    60                      80 Planar Average Exposure (GWd/MTU)
Planar Avg.      APLHGR Exposure          Limit (GWd/MTU)          (kW/ft) 0.0            13.0 15.0            13.0 67.0              7.6 Figure 2.1 APLHGRRATED for ATRIUM-10XM Fuel Browns Ferry Unit 3 Cycle 20                                                                                              Page 4 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 2.5      Equipment Out-Of-Service Corrections The limits shown in Figure 2.1 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
In-Service                              All equipment In-Service
* 1F1F1F RPTOOS                                  EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                  Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                  Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                        Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                  One Recirculation Pump Out-Of-Service
* All equipment service conditions assume 1 SRVOOS.
Browns Ferry Unit 3 Cycle 20                                                                                          Page 5 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 3 LHGR Limits (Technical Specification 3.2.3, 3.3.4.1, & 3.7.5)
The LHGR limit is determined by adjusting the rated power LHGR limit for off-rated power and off-rated flow conditions. The most limiting of these is then used as follows:
LHGR limit = MIN ( LHGRP, LHGRF )
where:
LHGRP                  off-rated power LHGR limit                    [LHGRRATED
* LHGRFACP]
LHGRF                  off-rated flow LHGR limit                      [LHGRRATED
* LHGRFACF]
3.1    Rated Power and Flow Limit: LHGRRATED The rated conditions LHGR for all fuel are identified per Reference 1. The rated conditions LHGR for ATRIUM-10XM fuel is shown in Figure 3.1. The LHGR limit is consistent with Reference 3.
3.2    Off-Rated Power Dependent Limit: LHGRP LHGR limits are adjusted for off-rated power conditions using the LHGRFACP multiplier provided in Reference 1. The multiplier is split into two sub cases: turbine bypass valves in and out-of-service. The base case multipliers are shown in Figure 3.2.
3.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional limits are shown in Figure 3.4 and Figure 3.5, based on temperature conditions identified in Table 3.1.
Table 3.1 Startup Feedwater Temperature Basis Temperature Power              Range 1                Range 2
(% Rated)                (&deg;F)                  (&deg;F) 23                  160.0                  155.0 30                  167.0                  162.0 40                  177.0                  172.0 50                  187.0                  182.0 Browns Ferry Unit 3 Cycle 20                                                                                          Page 6 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 3.3      Off-Rated Flow Dependent Limit: LHGRF LHGR limits are adjusted for off-rated flow conditions using the LHGRFACF multiplier provided in Reference 1. Multipliers are shown in Figure 3.3.
3.4      Equipment Out-Of-Service Corrections The limits shown in Figure 3.1 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
* 2F2F2F In-Service                              All equipment In-Service RPTOOS                                  EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                  Turbine Bypass Valve(s) Out-Of-Service PLUOOS                                  Power Load Unbalance Out-Of-Service FHOOS (or FFWTR)                        Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction RCPOOS                                  One Recirculation Pump Out-Of-Service Off-rated power corrections shown in Figure 3.2 are dependent on operation of the Turbine Bypass Valve system. For this reason, separate limits are to be applied for TBVIS or TBVOOS operation. The limits have no dependency on RPTOOS, PLUOOS, FHOOS/FFWTR, or SLO.
Off-rated flow corrections shown in Figure 3.3 are bounding for all EOOS conditions.
Off-rated power corrections shown in Figure 3.4 and Figure 3.5 are also dependent on operation of the Turbine Bypass Valve system. In this case, limits support FHOOS operation during startup. These limits have no dependency on RPTOOS, PLUOOS, or SLO.
* All equipment service conditions assume 1 SRVOOS.
Browns Ferry Unit 3 Cycle 20                                                                                          Page 7 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 15 12 9
LHGR (kW/ft) 6 3
0 0            20                      40                    60                      80 Pellet Exposure (GWd/MTU)
Pellet              LHGR Exposure            Limit (GWd/MTU)          (kW/ft) 0.0              14.1 18.9              14.1 74.4              7.4 Figure 3.1 LHGRRATED for ATRIUM-10XM Fuel Browns Ferry Unit 3 Cycle 20                                                                                              Page 8 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 1.10 1.00 0.90 Turbine Bypass Valve In-Service, TBVIS 0.80                        I            I            Turbine Bypass Valve Out-of-Service, TBVOOS LHGRFACP 0.70 0.60 0.50          TBVIS,  50% Core Flow TBVOOS,  50% Core Flow TBVIS, > 50% Core Flow 0.40 TBVOOS, > 50% Core Flow 0.30 0.20 20      30          40          50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service              Turb ine Bypass Out-of-Service Core                                      Core Power          LHGRFACP                  Power        LHGRFACP
(% Rated)                                  (% Rated) 100.0        1.00                        100.0        1.00 26.0        0.64                          26.0        0.62 Core Flow > 50% Rated                      Core Flow > 50% Rated 26.0        0.45                          26.0        0.38 23.0        0.41                          23.0        0.35 Core Flow  50% Rated                      Core Flow  50% Rated 26.0        0.49                          26.0        0.49 23.0        0.46                          23.0        0.42 Figure 3.2 Base Operation LHGRFACP for ATRIUM-10XM Fuel (Independent of other EOOS conditions)
Browns Ferry Unit 3 Cycle 20                                                                                              Page 9 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                      Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 1.10 1.05 1.00 0.95 0.90 LHGRFACF 0.85 0.80 0.75 0.70 0.65 0.60 0.55 20    30          40          50        60        70        80        90        100        110 Core Flow (% Rated)
Core Flow            LHGRFACF
(% Rated) 0.0              0.63 30.0            0.63 75.6            1.00 107.0              1.00 Figure 3.3 LHGRFACF for ATRIUM-10XM Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 3 Cycle 20                                                                                            Page 10 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 1.10 1.00 0.90 0.80 I
Turbine Bypass Valve In-Service, TBVIS I
I        I Turbine Bypass Valve Out-of-Service, TBVOOS LHGRFACP 0.70 0.60 0.50 TBVIS,  50% Core Flow TBVIS, > 50% Core Flow 0.40        TBVOOS,  50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20    30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power          LHGRFACP                    Power        LHGRFACP
(% Rated)                                    (% Rated) 100.0        1.00                          100.0        1.00 26.0        0.52                            26.0        0.51 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0        0.41                            26.0        0.34 23.0        0.38                            23.0        0.32 Core Flow  50% Rated                        Core Flow  50% Rated 26.0        0.44                            26.0        0.40 23.0        0.40                            23.0        0.39 Figure 3.4 Startup Operation LHGRFACP for ATRIUM-10XM Fuel:
Table 3.1 Temperature Range 1 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 3 Cycle 20                                                                                            Page 11 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                        Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 1.10 1.00 0.90 0.80 I
Turbine Bypass Valve In-Service, TBVIS I              Turbine Bypass Valve Out-of-Service, TBVOOS LHGRFACP 0.70 0.60 0.50 TBVIS,  50% Core Flow TBVIS, > 50% Core Flow 0.40        TBVOOS,  50% Core Flow TBVOOS, > 50% Core Flow 0.30 0.20 20    30          40            50        60        70        80        90        100        110 Core Power (% Rated)
Turb ine Bypass In-Service                Turb ine Bypass Out-of-Service Core                                        Core Power          LHGRFACP                    Power        LHGRFACP
(% Rated)                                    (% Rated) 100.0        1.00                          100.0        1.00 26.0        0.52                            26.0        0.51 Core Flow > 50% Rated                        Core Flow > 50% Rated 26.0        0.41                            26.0        0.34 23.0        0.38                            23.0        0.32 Core Flow  50% Rated                        Core Flow  50% Rated 26.0        0.44                            26.0        0.40 23.0        0.40                            23.0        0.39 Figure 3.5 Startup Operation LHGRFACP for ATRIUM-10XM Fuel:
Table 3.1 Temperature Range 2 (no Feedwater heating during startup)
(Limits valid at and below 50% power)
Browns Ferry Unit 3 Cycle 20                                                                                            Page 12 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 4 OLMCPR Limits (Technical Specification 3.2.2, 3.3.4.1, & 3.7.5)
OLMCPR is calculated to be the most limiting of the flow or power dependent values OLMCPR limit = MAX ( MCPRF , MCPRP )
where:
MCPRF              core flow-dependent MCPR limit MCPRP              power-dependent MCPR limit 4.1    Flow Dependent MCPR Limit: MCPRF MCPRF limits are dependent upon core flow (% of Rated), and the max core flow limit, (Rated or Increased Core Flow, ICF). MCPRF limits are shown in Figure 4.1, per Reference 1. Limits are valid for all EOOS combinations. No adjustment is required for SLO conditions.
4.2    Power Dependent MCPR Limit: MCPRP MCPRP limits are dependent upon:
Core Power Level (% of Rated)
Technical Specification Scram Speed (TSSS), Nominal Scram Speed (NSS), or Optimum Scram Speed (OSS)
Cycle Operating Exposure (NEOC, EOC, and CD - as defined in this section)
Equipment Out-Of-Service Options Two or Single recirculation Loop Operation (TLO vs. SLO)
The MCPRP limits are provided in Table 4.2 through Table 4.9, where each table contains the limits for all fuel types and EOOS options (for a specified scram speed and exposure range).
The CMSS determines MCPRP limits, from these tables, based on linear interpolation between the specified powers.
4.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reached a significant operating power level. Additional power dependent limits are shown in Table 4.5 through Table 4.8 based on temperature conditions identified in Table 3.1.
Browns Ferry Unit 3 Cycle 20                                                                                        Page 13 Core Operating Limits Report, (120% OLTP, MELLLA+)                                          TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 4.2.2    Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS)
MCPRP limits are provided for three different sets of assumed scram speeds. The Technical Specification Scram Speed (TSSS) MCPRP limits are applicable at all times, as long as the scram time surveillance demonstrates the times in Technical Specification Table 3.1.4-1 are met. Both Nominal Scram Speeds (NSS) and/or Optimum Scram Speeds (OSS) may be used, as long as the scram time surveillance demonstrates Table 4.1 times are applicable.
* 3F3F3F 4F4F4F Table 4.1 Nominal Scram Time Basis Notch                  Nominal              Optimum Position            Scram Timing          Scram Timing (index)                (seconds)              (seconds) 46                        0.420                  0.380 36                        0.980                  0.875 26                        1.600                  1.465 6                      2.900                  2.900 In demonstrating compliance with the NSS and/or OSS scram time basis, surveillance requirements from Technical Specification 3.1.4 apply; accepting the definition of SLOW rods should conform to scram speeds shown in Table 4.1. If conformance is not demonstrated, TSSS based MCPRP limits are applied.
On initial cycle startup, TSSS limits are used until the successful completion of scram timing confirms NSS and/or OSS based limits are applicable.
4.2.3    Exposure Dependent Limits Exposures are tracked on a Core Average Exposure basis (CAVEX, not Cycle Exposure).
Higher exposure MCPRP limits are always more limiting and may be used for any Core Average Exposure up to the ending exposure. Per Reference 1, MCPRP limits are provided for the following exposure ranges:
BOC to NEOC                                NEOC corresponds to                        27,972.7 MWd / MTU BOC to EOCLB                                EOCLB corresponds to                        33,104.7 MWd / MTU BOC to End of Coast                        End of Coast                                34,799.5 MWd / MTU NEOC refers to a Near EOC exposure point.
* Reference 1 analysis results are based on information identified in Reference 4.
Drop out times consistent with method used to perform actual timing measurements (i.e., including pickup/dropout effects).
Browns Ferry Unit 3 Cycle 20                                                                                                              Page 14 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                      TVA-COLR-BF3C20, Revision 1 (Final)
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[i!D NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 The EOCLB exposure point is not the true End-Of-Cycle exposure. Instead it corresponds to a licensing exposure window exceeding expected end-of-full-power-life.
The End of Coast exposure point represents a licensing exposure point exceeding the expected end-of-cycle exposure including cycle extension options.
4.2.4      Equipment Out-Of-Service (EOOS) Options EOOS options
* covered by MCPRP limits are given by the following:
5F5F5F In-Service                                          All equipment In-Service RPTOOS                                              EOC-Recirculation Pump Trip Out-Of-Service TBVOOS                                              Turbine Bypass Valve(s) Out-Of-Service RPTOOS+TBVOOS                                      Combined RPTOOS and TBVOOS PLUOOS                                              Power Load Unbalance Out-Of-Service PLUOOS+RPTOOS                                      Combined PLUOOS and RPTOOS PLUOOS+TBVOOS                                      Combined PLUOOS and TBVOOS PLUOOS+TBVOOS+RPTOOS                                Combined PLUOOS, RPTOOS, and TBVOOS FHOOS (or FFWTR)                                    Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)
RCPOOS                                              One Recirculation Pump Out-Of-Service For exposure ranges up to NEOC and EOCLB, additional combinations of MCPRP limits are also provided including FHOOS. The coast down exposure range assumes application of FFWTR. FHOOS based MCPRP limits for the coast down exposure are redundant because the temperature setdown assumption is identical with FFWTR.
4.2.5      Single-Loop-Operation (SLO) Limits When operating in RCPOOS conditions, MCPRp limits are constructed differently from the normal operating RCP conditions. The limiting event for RCPOOS is a pump seizure scenario, which sets the upper bound for allowed core power and flow . This event is not impacted by 6F6F6F scram time assumptions. Specific MCPRP limits are shown in Table 4.9.
4.2.6      Below Pbypass Limits Below Pbypass (26% rated power), MCPRP limits depend upon core flow. One set of MCPRP limits applies for core flow above 50% of rated; a second set applies if the core flow is less than or equal to 50% rated.
* All equipment service conditions assume 1 SRVOOS.
RCPOOS limits are only valid up to 43.75% rated core power, 50% rated core flow, and an active recirculation drive flow of 17.73 Mlbm/hr.
Browns Ferry Unit 3 Cycle 20                                                                                                        Page 15 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF3C20, Revision 1 (Final)
BFN - Unit 3                                            Page 24 of 44                                        TRM Revision 146 September 23, 2020
 
ECM: L32 200817 800
[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 2.00 1.80 1.60 MCPRF 1.40 1.20 1.00 30        40          50            60        70          80          90        100        110 Core Flow (% Rated)
Core Flow            MCPRF
(% Rated) 30.0              1.58 84.0              1.34 107.0              1.34 Figure 4.1 MCPRF for All Fuel Types (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 3 Cycle 20                                                                                            Page 16 Core Operating Limits Report, (120% OLTP, MELLLA+)                                              TVA-COLR-BF3C20, Revision 1 (Final)
BFN - Unit 3                                              Page 25 of 44                                TRM Revision 146 September 23, 2020
 
ECM: L32 200817 800
[i!D NPG                            Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 Table 4.2 MCPRP Limits for All Fuel Types: Optimum Scram Time Basis
* 7F7F7F ATRIUM-10XM BOC          BOC        BOC Pow er              to            to      to End of Operating Condition      (% of rated)        NEOC        EOCLB        Coast 100              1.39          1.41        1.44 90              1.45          1.46        1.48 77.6            1.50          1.51        1.54 65              1.57          1.57        1.61
                                                        >50              1.65          1.65        1.70 Base Case              50              1.79          1.79        1.79 40              1.87          1.87        1.88 26              2.27          2.27        2.38 26 at > 50%F          2.60          2.60        2.70 23 at > 50%F          2.76          2.76        2.88 26 at  50%F          2.49          2.49        2.60 23 at  50%F          2.64          2.64        2.77 100              1.42          1.44          ---
90              1.48          1.48          ---
77.6            1.54          1.54          ---
65              1.61          1.61          ---
                                                        >50              1.70          1.70          ---
FHOOS                  50              1.79          1.79          ---
40              1.88          1.88          ---
26              2.38          2.38          ---
26 at > 50%F          2.70          2.70          ---
23 at > 50%F          2.88          2.88          ---
26 at  50%F          2.60          2.60          ---
23 at  50%F          2.77          2.77          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR/FHOOS is supported for the BOC to End of Coast limits.
Browns Ferry Unit 3 Cycle 20                                                                                                    Page 17 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                    TVA-COLR-BF3C20, Revision 1 (Final)
BFN - Unit 3                                                Page 26 of 44                                      TRM Revision 146 September 23, 2020
 
ECM: L32 200817 800
[i!D NPG                                Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis
* 8F8F8F ATRIUM-10XM BOC        BOC        BOC Pow er          to        to      to End of Operating Condition    (% of rated)    NEOC      EOCLB        Coast 100          1.42      1.44        1.46 90          1.48      1.48        1.51 77.6          1.53      1.53        1.56 65          1.59      1.59        1.63
                                                              >50          1.67      1.67        1.72 50          1.80      1.80        1.80 Base Case 40          1.88      1.88        1.90 26          2.30      2.30        2.41 26 at > 50%F      2.60      2.60        2.70 23 at > 50%F      2.76      2.76        2.88 26 at  50%F      2.49      2.49        2.60 23 at  50%F      2.64      2.64        2.77 100          1.46      1.47        1.50 90          1.51      1.51        1.54 77.6          1.56      1.56        1.60 65          1.62      1.62        1.66
                                                              >50          1.70      1.70        1.75 50          1.80      1.80        1.81 TBVOOS 40          1.88      1.88        1.91 26          2.30      2.30        2.42 26 at > 50%F      3.11      3.11        3.25 23 at > 50%F      3.36      3.36        3.50 26 at  50%F      2.83      2.83        2.99 23 at  50%F      3.11      3.11        3.28 100          1.46      1.46          ---
90          1.51      1.51          ---
77.6          1.56      1.56          ---
65          1.63      1.63          ---
                                                              >50          1.72      1.72          ---
50          1.80      1.80          ---
FHOOS 40          1.90      1.90          ---
26          2.41      2.41          ---
26 at > 50%F      2.70      2.70          ---
23 at > 50%F      2.88      2.88          ---
26 at  50%F      2.60      2.60          ---
23 at  50%F      2.77      2.77          ---
100          1.42      1.44        1.46 90          1.48      1.48        1.51 77.6          1.53      1.53        1.56 65          1.72      1.73        1.73
                                                              >50          ---        ---          ---
50          1.80      1.80        1.80 PLUOOS 40          1.88      1.88        1.90 26          2.30      2.30        2.41 26 at > 50%F      2.60      2.60        2.70 23 at > 50%F      2.76      2.76        2.88 26 at  50%F      2.49      2.49        2.60 23 at  50%F      2.64      2.64        2.77
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 3 Cycle 20                                                                                                    Page 18 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                      TVA-COLR-BF3C20, Revision 1 (Final)
BFN - Unit 3                                                Page 27 of 44                                      TRM Revision 146 September 23, 2020
 
ECM: L32 200817 800
[i!D NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 Table 4.3 MCPRP Limits for All Fuel Types: Nominal Scram Time Basis (continued)
* 9F9F9F ATRIUM-10XM BOC        BOC          BOC Pow er          to          to      to End of Operating Condition    (% of rated)    NEOC        EOCLB        Coast 100          1.49      1.50          ---
90          1.54      1.54          ---
77.6          1.60      1.60          ---
65          1.66      1.66          ---
                                                            >50          1.75      1.75          ---
TBVOOS              50          1.81      1.81          ---
FHOOS                40          1.91      1.91          ---
26          2.42      2.42          ---
26 at > 50%F      3.25      3.25          ---
23 at > 50%F      3.50      3.50          ---
26 at  50%F      2.99      2.99          ---
23 at  50%F      3.28      3.28          ---
100          1.46      1.47          1.50 90          1.51      1.51          1.54 77.6          1.56      1.56          1.60 65          1.72      1.73          1.74
                                                            >50            ---        ---          ---
TBVOOS              50          1.80      1.80          1.81 PLUOOS                40          1.88      1.88          1.91 26          2.30      2.30          2.42 26 at > 50%F      3.11      3.11          3.25 23 at > 50%F      3.36      3.36          3.50 26 at  50%F      2.83      2.83          2.99 23 at  50%F      3.11      3.11          3.28 100          1.46      1.46          ---
90          1.51      1.51          ---
77.6          1.56      1.56          ---
65          1.72      1.73          ---
                                                            >50            ---        ---          ---
FHOOS                50          1.80      1.80          ---
PLUOOS                40          1.90      1.90          ---
26          2.41      2.41          ---
26 at > 50%F      2.70      2.70          ---
23 at > 50%F      2.88      2.88          ---
26 at  50%F      2.60      2.60          ---
23 at  50%F      2.77      2.77          ---
100          1.49      1.50          ---
90          1.54      1.54          ---
77.6          1.60      1.60          ---
65          1.73      1.74          ---
TBVOOS              >50            ---        ---          ---
FHOOS                50          1.81      1.81          ---
40          1.91      1.91          ---
PLUOOS 26          2.42      2.42          ---
26 at > 50%F      3.25      3.25          ---
23 at > 50%F      3.50      3.50          ---
26 at  50%F      2.99      2.99          ---
23 at  50%F      3.28      3.28          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 3 Cycle 20                                                                                                      Page 19 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                        TVA-COLR-BF3C20, Revision 1 (Final)
BFN - Unit 3                                                Page 28 of 44                                      TRM Revision 146 September 23, 2020
 
ECM: L32 200817 800
[i!D NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis
* 10F10F10F ATRIUM-10XM BOC        BOC          BOC Pow er          to          to      to End of Operating Condition    (% of rated)    NEOC        EOCLB        Coast 100          1.46      1.46          1.50 90          1.50      1.50          1.55 77.6          1.55      1.55          1.59 65          1.61      1.61          1.66
                                                            >50          1.68      1.68          1.75 Base Case            50          1.80      1.80          1.82 40          1.88      1.88          1.94 26          2.32      2.32          2.44 26 at > 50%F      2.60      2.60          2.71 23 at > 50%F      2.76      2.76          2.89 26 at  50%F      2.49      2.49          2.61 23 at  50%F      2.64      2.64          2.78 100          1.50      1.50          1.54 90          1.55      1.55          1.59 77.6          1.59      1.59          1.64 65          1.65      1.65          1.70
                                                            >50          1.73      1.73          1.79 TBVOOS              50          1.81      1.81          1.84 40          1.89      1.89          1.96 26          2.34      2.34          2.47 26 at > 50%F      3.12      3.12          3.27 23 at > 50%F      3.37      3.37          3.52 26 at  50%F      2.84      2.84          3.01 23 at  50%F      3.12      3.12          3.30 100          1.50      1.50          ---
90          1.55      1.55          ---
77.6          1.59      1.59          ---
65          1.66      1.66          ---
                                                            >50          1.75      1.75          ---
FHOOS                50          1.81      1.81          ---
40          1.94      1.94          ---
26          2.44      2.44          ---
26 at > 50%F      2.71      2.71          ---
23 at > 50%F      2.89      2.89          ---
26 at  50%F      2.61      2.61          ---
23 at  50%F      2.78      2.78          ---
100          1.46      1.46          1.50 90          1.50      1.50          1.55 77.6          1.55      1.55          1.59 65          1.73      1.74          1.75
                                                            >50            ---        ---          ---
PLUOOS              50          1.80      1.80          1.82 40          1.88      1.88          1.94 26          2.32      2.32          2.44 26 at > 50%F      2.60      2.60          2.71 23 at > 50%F      2.76      2.76          2.89 26 at  50%F      2.49      2.49          2.61 23 at  50%F      2.64      2.64          2.78
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 3 Cycle 20                                                                                                      Page 20 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                        TVA-COLR-BF3C20, Revision 1 (Final)
BFN - Unit 3                                                Page 29 of 44                                      TRM Revision 146 September 23, 2020
 
ECM: L32 200817 800
[i!D NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 Table 4.4 MCPRP Limits for All Fuel Types: Technical Specification Scram Time Basis (continued)
* 1F1F1F ATRIUM-10XM BOC        BOC          BOC Pow er          to          to      to End of Operating Condition    (% of rated)    NEOC        EOCLB        Coast 100          1.54      1.54          ---
90          1.59      1.59          ---
77.6          1.64      1.64          ---
65          1.70      1.70          ---
                                                            >50          1.79      1.79          ---
TBVOOS              50          1.83      1.83          ---
FHOOS                40          1.96      1.96          ---
26          2.47      2.47          ---
26 at > 50%F      3.27      3.27          ---
23 at > 50%F      3.52      3.52          ---
26 at  50%F      3.01      3.01          ---
23 at  50%F      3.30      3.30          ---
100          1.50      1.50          1.54 90          1.55      1.55          1.59 77.6          1.59      1.59          1.64 65          1.74      1.75          1.77
                                                            >50            ---        ---          ---
TBVOOS              50          1.81      1.81          1.84 PLUOOS                40          1.89      1.89          1.96 26          2.34      2.34          2.47 26 at > 50%F      3.12      3.12          3.27 23 at > 50%F      3.37      3.37          3.52 26 at  50%F      2.84      2.84          3.01 23 at  50%F      3.12      3.12          3.30 100          1.50      1.50          ---
90          1.55      1.55          ---
77.6          1.59      1.59          ---
65          1.74      1.75          ---
                                                            >50            ---        ---          ---
FHOOS                50          1.81      1.81          ---
PLUOOS                40          1.94      1.94          ---
26          2.44      2.44          ---
26 at > 50%F      2.71      2.71          ---
23 at > 50%F      2.89      2.89          ---
26 at  50%F      2.61      2.61          ---
23 at  50%F      2.78      2.78          ---
100          1.54      1.54          ---
90          1.59      1.59          ---
77.6          1.64      1.64          ---
65          1.76      1.77          ---
TBVOOS              >50            ---        ---          ---
FHOOS                50          1.83      1.83          ---
40          1.96      1.96          ---
PLUOOS 26          2.47      2.47          ---
26 at > 50%F      3.27      3.27          ---
23 at > 50%F      3.52      3.52          ---
26 at  50%F      3.01      3.01          ---
23 at  50%F      3.30      3.30          ---
* All limits, including Base Case, support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the BOC to End of COAST exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
Browns Ferry Unit 3 Cycle 20                                                                                                      Page 21 Core Operating Limits Report, (120% OLTP, MELLLA+)                                                        TVA-COLR-BF3C20, Revision 1 (Final)
BFN - Unit 3                                                Page 30 of 44                                      TRM Revision 146 September 23, 2020
 
ECM: L32 200817 800
[i!D NPG                              Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: August 17, 2020 Table 4.5 Startup Operation MCPRP Limits for Table 3.1 Temperature Range 1 for All Fuel Types: Nominal Scram Time Basis
* 12F12F12F ATRIUM-10XM BOC        BOC        BOC Pow er            to        to    to End of Operating Condition      (% of rated)    NEOC        EOCLB      Coast 100          1.46        1.46      1.46 90          1.51        1.51      1.51 77.6          1.56        1.56      1.56 65          1.72        1.73      1.73
                                                          >50            ---        ---        ---
TBVIS                  50          1.84        1.84      1.84 40          2.07        2.07      2.07 26          2.66        2.66      2.66 26 at > 50%F      2.92        2.92      2.92 23 at > 50%F      3.14        3.14      3.14 26 at  50%F      2.82        2.82      2.82 23 at  50%F      3.04        3.04      3.04 100          1.49        1.50      1.50 90          1.54        1.54      1.54 77.6          1.60        1.60      1.60 65          1.73        1.74      1.74
                                                          >50            ---        ---        ---
TBVOOS                  50          1.85        1.85      1.85}}

Latest revision as of 03:55, 19 November 2024

Summary Report for 10 CFR 50.59 Evaluations, Technical Specifications Bases Changes, Technical Requirements Manual Changes, and NRC Commitment Revisions
ML21335A067
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 11/26/2021
From: Rasmussen M
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML21335A067 (927)
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