Text

Improved Tech Specs Pages Re Section 5, Administrative Controls.
	ML18039A198
Person / Time
Site:	Browns Ferry
Issue date:	12/03/1997
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML18039A197	List: ML18039A196; ML18039A198;
References
	NUDOCS 9712080173
	Download: ML18039A198 (464)
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Improved Technical Specifications All Pages Unit 1 97i2084i73 97i243 PDR ADOCK OSOOOaS9 P PDR

Responsibility 5.1

5. 0 ADMINISTRATIVE CONTROLS

5. 1 Responsibility 5.1.1 The Site Vice-President shall be responsible for overall activities at the site, while the Plant Manager shall be responsible for overall unit operation. The Site Vice-President and the Plant Hanager shall delegate in writing the succession to this responsibility during their absence.

The Plant Hanager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

5.1.2 The Shift Manager shall be responsible for the control room command function. During any absence of the Shift Manager from the control room while the unit is in'ODE 1, 2, or 3, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function.

During any absence of the Shift Manager from the control room.

while the unit is in MODE 4 or 5, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

BFN-UNIT 1 Amendment

Organization 5.2

5. 0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Or anizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities

'ffecting safety of the nuclear power plant.

'a ~ Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A);

b. The Plant Manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant; c~ The Chief Nuclear Officer and Executive Vice President, TVA Nuclear, shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and

d. The individuals who train the operating staff, carry out radiological controls, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2 Unit Staff The unit staff organization shall include the following:

(continued)

BFN-UNIT I 5.0-2 Amendment *Rl

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) a ~ A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in MODES 1, 2, or 3.

When all three units are shutdown or defueled, a total of three non-licensed operators shall be assigned for all three units.

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

C. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and Specifications 5.2.2.a and 5.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

A radiological controls technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hour s of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, radiological controls technicians, auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an 8, 10, or 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months day, nominal 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant (continued)

BFN-UNIT 1 5.0-3 Amendment *Rl

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) modification, on a temporary basis the following guidelines shall be followed:

l. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time;

2. An individual should not be permitted to work more than 16 hours in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized in advance by the Plant Manager or his designee, in accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or his designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.

f. The Operations Superintendent shall hold a current SRO license on a Browns Ferry unit.

9. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Manager in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

BFN-UNIT I 5.0-4 Amendment *Rl

Unit Staff gualifications 5.3

5. 0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff qualifications 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications for comparable positions as specified in the TVA Nuclear guality Assurance Plan (TVA-N(A-PLN89-A).

BFN-UNIT I 5.0-5 Amendment

e c,

Procedures 5.4

5. 0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Mritten procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;

b. The emergency operating instructions required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;

c. equality assurance for effluent and environmental monitoring;

d. Fire Protection Program implementation; and

e. All programs specified in Specification 5.5.

BFN-UNIT 1 5.0-6 Amendment

6 Programs and Manuals 5.5

5. 0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented and maintained.

5.5.1 Offsite Dose Calculation Manual ODCM a~ The ODCH shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the Radiological Environmental Honitoring Program; and

b. The ODCH shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release, reports required by Specification 5.6.2 and Specification 5.6.3.

Licensee initiated changes to the ODCH:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2. a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20. 1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b. Shall become effective after review and acceptance by the process described in TVA-NgA-PLN89-A; and

c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCH was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page (continued)

BFN-UNIT 1 5.0-7 Amendment

Programs and Manuals 5.5 5.5 Programs and manuals 5.5.1 Offsite Dose Calculation Manual ODCH (continued) that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.5.2 Primar Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include the Core Spray, High Pressure Coolant Injection, Residual Heat Removal, and Reactor Core Isolation Cooling. The program shall include the following preventive maintenance:

a. Periodic visual inspection requirements; and

b. System leak test requirements for each system, to the extent permitted by system design and radiological conditions, at refueling cycle intervals or less. II Post Accident Sam lin This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following:

a. Training of personnel;

b. Procedures for sampling and analysis; and

c. Provisions for maintenance of sampling and analysis equipment.

5.5.4 Radioactive Effluent Controls Pro ram This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably (continued)

BFN-UNIT 1 5.0-8 Amendment

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Pro ram (continued) achievable. The program shall be contained in the ODCH, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a~ Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCH;

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 times the concentration values in 10 CFR 20, Appendix B, Table 2, Column 2;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I;

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days;

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed PA of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary shall be limited to the following;

1. For noble gases: a dose rate of x 500 mrem/yr to the total body and ( 3000 mrem/yr to the skin, and (continued)

BFN-UNIT 1 5.0-9 Amendment *Rl

Programs and Manuals 5.5 5.5 Programs and Hanuals 5.5.4 Radioactive Effluent Controls Pro ram (continued)

2. For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half lives > 8 days: a dose rate of x 1500 mrem/yr to any organ;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

5.5.5 Com onent C clic or Transient Limit This program provides controls to track the FSAR Section 4.2.5, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 Inservice Testin Pro ram This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies specified in Section XI of the ASNE Boiler and Pressure Vessel Code and applicable Addenda are as follows:

(continued)

BFN-UNIT 1 5.0-10 Amendment *Rl

Programs and Hanuals 5.5 5.5 Programs and Hanuals 5.5.6 Inservice Testin Pro ram (continued)

ASHE Boiler and Pressure Vessel Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testin activities Weekly At least once per 7 days Honthly At least once per 31 days quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASHE Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

5.5.7 Ventilation Filter Testin Pro ram VFTP The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

a. Demonstrate for each of the ESF systems (Standby Gas Treatment (SGT) System and Control Room Emergency Ventilation (CREV) System) that an inplace test (continued)

BFN-UNIT 1 5.0-11 Amendment *Rl

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Programs and Manuals 5.5 Ventilation Filter Testin Pro ram VFTP (continued) of the HEPA filters shows a penetration and system bypass

< 1.% when tested in accordance with ANSI N510-1975 at the system flowrate specified below, i 10K.

ESF Ventilation System Flowrate (cfm)

SGT System 9000 CREV System 3000 This testing shall be performed 1) every 18 months, 2) after partial or complete replacement of HEPA filters, 3) after any structural maintenance on the system housing, or 4) following significant painting, 'fire, or chemical release in any ventilation zone communicating with the system.

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 1.0X when tested in accordance with ANSI N510-1975 at the system flowrate specified below, i 1(C.

ESF Ventilation System Flowrate (cfm)

SGT System 9000 CREV System 3000 This testing shall be performed 1) every 18 months, 2) after partial or complete replacement of the charcoal adsorber bank, 3) after any structural maintenance on the system housing, or 4) following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

c ~ Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, shows a methyl iodide efficiency ) 90K when tested in accordance with ASTM D3803-1989.

This testing shall be performed 1) every 18 months, 2) after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, or 3) following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

(continued) 5.0-12 Amendment *Rl

Programs and Manuals 5.5 5.5 Programs ind Manuals S.S.7 t T ro (continued)

d. Once every 18 months demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the pr efilters, and the charcoal adsorber s is less than the value specified below at the system flowrate specified below, i 1%:

ESF Ventilation System Delta P Flowrate (inches water) (cfm)

SGT System 9000 CREV System 3000

e. Once every 18 months demonstrate that the heaters for the SGT System dissipate 40 kW k 1PA when tested in accordance with ANSI N510-1975.

5.5.8 l v G s S o io i it M it rin P o r m This program provides controls for potentially explosive gas mixtures contained downstream of the offgas recombiner s, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

a~ The limits for concentrations of hydrogen downstream of the offgas recombiners and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and

b. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tanks'ontents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in (continued)

BFN-UNIT 1 5.0-13 Amendment *Rl

Programs and Hanuals 5.5 Ex losive Gas and Stora e Tank Radioactivit Honitorin Pro ram (continued) an unrestricted area, in the event of an uncontrolled release of the tanks'ontents.

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

Diesel Fuel Oil Testin Pro ram A diesel fuel oil testing program to implement required testing of the fuel oil in each 7-day fuel oil tank shall be established.

The purpose of the program is to establish the following:

a. The quality of the fuel oil in each 7-day fuel oil tank. is within the acceptable limits specified in Table I of ASTH D-975-1989 when tested every 92 days; and

b. Total particulate concentration of the fuel oil in each 7-day fuel oil tank is w 10 mg/1 when tested every 92 days in accordance with ASTH D-2276, Hethod A-2 or A-3.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program testing frequencies.

Technical S ecifications TS Bases Control Pro ram This program provides a means for processing changes to the Bases of these Technical Specifications.

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

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

I. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

(continued) 5.0-14 Amendment *Rl

Programs and Manuals 5.5 Technical S ecifications TS Bases Control Pro ram (continued)

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

d. Proposed changes that meet the criteria of Specification 5.5. 10b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

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

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

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

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

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

(continued) 5.0-15 Amendment *Rl

I Programs and Manuals 5.5 Safet Function Determination Pro ram SFDP (continued)

a. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or

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

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

Primar Containment Leaka e Rate Testin Pro ram A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1. 163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 49.6 psig. The maximum allowable primary containment leakage rate, L., shall be 2X of primary containment air weight per day at P,.

Leakage Rate acceptance criteria are:

a. The primary containment leakage rate acceptance criteria is w 1.0 L.. During the first unit startup following the testing performed in accordance with this program, the leakage rate acceptance criteria are a 0.60 L, for the Type B and Type C tests, and z 0.75 L. for the Type A test; and (continued) 5.0-16 Amendment *Rl

Programs and Manuals 5.5 Primar Containment Leaka e Rate Testin Pro ram (continued)

b. Air lock testing acceptance criteria are:

1) Overal.l air lock leakage rate z 0.05 L. when tested at a p..

2) Air lock door seals leakage rate is c 0.02 L, when the overall air lock is pressurized to a 2.5 psig for at least 15 minutes.

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

The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

5.0-17 Amendment *Rl

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Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Occu ational Radiation Ex osure Re ort NOTE A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors), for whom monitoring was performed, receiving an annual deep dose equivalent > 100 mrems and the associated collective deep dose equivalent (reported in person - rem) according to work and job functions (e.g., reator operations and surveillance, inservice inspection, routine maintenance, special maintenance[describe maintenance), waste processing, and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, thermoluminescence dosimeter {TLD), electronic dosimeter, or film badge measurements. Small exposures totaling ( 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total deep dose equivalent received from external sources should be assigned to specific major work functions. The report covering the previous calendar year shall be submitted by April 30 of each year.

5.6.2 Annual Radiolo ical Environmental 0 eratin Re ort NOTE A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with

{continued)

BFN-UNIT I 5.0-18 Amendment *Rl

Reporting Requirements 5.6

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5.6

~ Reporting Requirements 5.6.2 Annual Radiolo ical Environmental 0 eratin Re ort (continued) the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

5.6.3 Radioactive Effluent Release Re ort NOTE A single submittal may be made for a multiple unit station. The submittal shall combine sections common to all units at the station; however for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Radioactive Effluent Release Report covering the operation of the unit during the .previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

5.6.4 Monthl 0 eratin Re orts Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.6.5 CORE OPERATING LIMITS REPORT COLR

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

(1) The APLHGRs for Specification 3.2.1; (2) The LHGR for Specification 3.2.3; (continued)

BFN-UNIT 1 5.0-19 Amendment *Rl

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIHITS REPORT COLR (continued)

(3) The HCPR Operating Limits for Specification 3.2.2; and (4) The RBH setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in NEOE-24011-P-A, "General Electric Standard Application for Reactor approved version for BFN).

Fuel,'latest C. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDH, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 ~PAM Re ort When a report is required by Condition B or G of LCO 3.3.3.1, "Post Accident Honitoring (PAH) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

BFN-UNIT 1 5.0-20 Amendment *Rl

High Radiation Area 5.7

5. 0 ADMINISTRATIVE CONTROLS High Radiation Area

~ ~ ~

5.7 As provided in IO CFR 20, paragraph 20.1601(c), the following controls may be applied to high radiation areas as an alternative to the controls required by 10 CFR 20.1601(a) and (b):

5.7.1 Each high radiation area, as defined in 10 CFR 20, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Mork Permit RWP). Individuals qualified in radiation protection procedures (e.g., a radiological controls technician) or personnel escorted by such individuals, shall be exempt from the RMP requirements during the performance of their assigned duties in high radiation areas where radiation doses could be received that are < I rem in one hour as measured at 30 centimeters from the radiation source or from the surface which the radiation penetrates, provided they otherwise comply with approved radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

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b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them.

C. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the RNP.

5.7.2 In addition to the requirements of Specification 5.7.1, areas that are accessible to personnel and that have radiation levels ) I rem in one hour as measured at 30 centimeters, but < 500 rads in one (continued)

BFN-UNIT I 5.0-21 Amendment *Rl

High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 (continued) hour at one meter from the radiation source or from the surface which the radiation penetrates, shall be provided with locked or continuously guarded doors to prevent unauthorized entry. The keys shall be under the administrative control of the duty Shift Manager, Radiological Controls Manager, or their respective designees. Doors shall remain locked except during periods of access by personnel under an approved RWP which specifies the dose rates in the immediate work areas and the maximum allowable stay times for individuals in that area. In lieu of the stay time requirement of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by individuals qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.7.3 Individual high radiation areas that are accessible to personnel, have radiation levels > I rem in one hour as measured at 30 centimeters, but < 500 rads in one hour at one meter from the radiation source, are located within large areas where no enclosure exists for purposes of locking and where no enclosure can be reasonably constructed around the individual area, shall be barricaded, conspicuously posted, and a flashing light shall be activated as 'a warning device whenever the dose rate in the area exceeds or will shortly exceed I rem in one hour as measured at 30 centimeters from the radiation source or from the surface which the radiation penetrates.

BFN-ONIT I 5.0-22 Amendment *Rl

0 Improved Technical Specifications AllPages Unit 2

Respons ibi1 i ty 5.1

5. 0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The Site Vice-President shall be responsible for overall activities at the site, while the Plant Manager shall be responsible for overall unit operation. The Site Vice-President and the Plant Manager shall delegate in writing the succession to this responsibility during their absence.

The Plant Manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

5.1.2 The Shift Hanager shall be responsible for the control room command function. During any absence of the Shift Manager from the control room while the unit is in MODE 1, 2, or 3, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function.

During any absence of the Shift Manager from the control room while the unit is in MODE 4 or 5, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

BFN-UNIT 2 5.0-1 Amendment

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Or anizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. 'he onsite and offsite organizations shall include the positions for activities

'affecting safety of the nuclear power plant.

a0 Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel, positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A);

b. The Plant Hanager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant;

c. The Chief Nuclear Officer and Executive Vice President, TVA Nuclear, shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and

d. The individuals who train the operating staff, carry out radiological controls, or perform quality assurance functions.may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2 Unit Staff The unit staff organization shall include the following:

(continued)

BFN-UNIT 2 5.0-2 Amendment *Rl

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued)

'a ~ A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in NODES 1, 2, or 3.

When all three units are shutdown or defueled, a total of three non-licensed operators shall be assigned for all three units.

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in NODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

C. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and Specifications 5.2".2.a and 5.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

d. A radiological controls technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, radiological controls technicians; auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an 8, 10, or 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months day, nominal 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant (continued)

BFN-UNIT 2 5.0-3 Amendment *Rl

0 0

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) modification, on a temporary basis the following guidelines shall be followed:

l. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time;

2. An individual should not be permitted to work more than 16 hours in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized in advance by the Plant Manager or his designee, in accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or his designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.

f. The Operations Superintendent shall hold a current SRO license on a Browns Ferry unit.

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Manager in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

BFN-UNIT 2 5.0-4 Amendment *Rl

Uni t Sta ff Qual i fi cat i ons 5.3 5.0 AOHINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications for comparable positions as specified in the TVA Nuclear Quality Assurance Plan (TVA-NQA-PLN89-A).

BFN-UNIT 2 5.0-5 Amendment

Procedures t

5.4

5. 0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Mritten procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;

b. The emergency operating instructions required to implement the requirements of NUREG-0737 and to NUREG-0737,

'Supplement 1, as stated in Generic Letter 82-33;

c. guality assurance for effluent and environmental monitoring;

d. Fire Protection Program implementation; and

e. All programs specified in Specification 5.5.

BFN-UNIT 2 5.0-6 Amendment

Programs and Hanuals 5.5 e 5. 0 5.5 ADMINISTRATIVE CONTROLS Programs and Manuals The following programs shall be established, implemented and maintained.

5.5.1 Offsite Dose Calculation Manual ODCM

'a ~ The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the Radiological Environmental Honitoring Program; and

b. The ODCH shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release, reports required by Specification 5.6.2 and Specification 5.6.3.

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2. a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b. Shall become effective after review and acceptance by the process described in TVA-NgA-PLN89-A; and c~ Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page (continued)

BFN-UNIT 2 5.0-7 Amendment

Programs and Manuals 5.5 5.5 Programs and Hanuals 5.5.1 Offsite Dose Calculation Manual ODCN (continued) that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.5.2 Primar Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include the Core Spray, High Pressure Coolant Injection, Residual Heat Removal, and Reactor Core Isolation Cooling. The program shall include the following preventive maintenance:

a. Periodic visual inspection requirements; and

b. System leak test requirements for each system, to the extent permitted by system design and radiological conditions, at refueling cycle intervals or less.

Post Accident Sam lin This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following:

a. Training of personnel;

b. Procedures for sampling and analysis; and

c. Provisions for maintenance of sampling and analysis equipment.

5.5.4 Radioactive Effluent Controls Pro ram This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably (continued)

BFN-UNIT 2 5.0-8 Amendment

Programs and Manuals 5.5 5.5 Programs and Hanuals 5.5.4 Radioactive Effluent Controls Pro ram (continued) achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 times the concentration values in 10 CFR 20, Appendix B, Table 2, Column 2;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I;

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days;

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2X of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary shall be limited to the following;

1. For noble gases: a dose rate of c 500 mrem/yr to the total body and < 3000 mrem/yr to the skin, and (continued)

BFN-UNIT 2 5.0-9 Amendment *Rl

N Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Pro ram (continued)

2. For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half lives ) 8 days: a dose rate of a 1500 mrem/yr to any organ;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

Com onent C clic or Transient Limit This program provides controls to track the FSAR Section 4.2.5, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 Inservice Testin Pro ram This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda are as follows:

(continued)

BFN-UNIT 2 5.0-10 Amendment *Rl

Programs and Manuals 5.5 Inservice Testin Pro ram (continued)

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testin activities Meekly At least once per 7 days Monthly At least once per 31 days quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months A't least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

Ventilation Filter Testin Pro ram VFTP The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

a. Oemonstrate for each of the ESF systems (Standby Gas Treatment (SGT) System and Control Room Emergency Ventilation (CREV) System) that an inplace test (continued) 5.0-11 Amendment *Rl

Programs and Manuals t

5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testin Pro ram VFTP (continued) of the HEPA filters shows a penetration and system bypass

< 1.% when tested in accordance with ANSI N510-1975 at the system flowrate specified below, a l(C.

ESF Ventilation System Flowrate (cfm)

SGT System 9000 CREV System 3000 This testing shall be performed 1) every 18 months, 2) after partial or complete replacement of HEPA filters, 3) after any structural maintenance on the system housing, or 4) following significant painting,'ire, or chemical release in any ventilation zone communicating with the system.

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 1.% when tested in accordance with ANSI N510-1975 at the system flowrate specified below, i Il4.

ESF Ventilation System Flowrate (cfm)

SGT System 9000 CREV System 3000 This testing shall be performed I) every 18 months, 2) after partial or complete replacement of the charcoal adsorber bank, 3) after any structural maintenance on the system housing, or 4) following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

C. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, shows a methyl iodide efficiency > 9K when tested in accordance with ASTM D3803-1989.

This testing shall be performed 1) every 18 months, 2) after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, or 3) following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

(continued)

BFN-UNIT 2 5.0-12 Amendment *Rl

Programs and Kanuals 5.5 5.5 Programs and Hanuals Ventil tion Filter Testin Pro r VFTP (continued)

d. Once every 18 months demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsor bers is less than the value specified below at the system flowrate specified below, i 1PA,:

ESF Ventilation System Delta P Flowrate (inches water) (cfm)

SGT System 9000 CREV System 3000

e. Once every 18 months demonstrate that the heaters for the SGT System dissipate 40 kM 2 10K when tested in accordance with ANSI N510-1975.

5.5.8 losiv as an t ra e n io ct v t Honitorin Pro ram This program provides controls for potentially explosive gas mixtures contained downstream of the offgas recombiners, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

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

b. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tanks'ontents,and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in (continued)

BFN-UNIT 2 5.0-13 Amendment 'Rl

Programs and Manuals t

5.5 5.5 Programs and Manuals 5.5.8 Ex losive Gas and Stora e Tank Radioactivit Honitorin Pro ram (continued) an unrestricted area, in the event of an uncontrolled release of the tanks'ontents.

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

5.5.9 Diesel Fuel Oil Testin Pro ram A diesel fuel oil testing program to implement required testing of the fuel oil in each 7-day fuel oil tank shall be established.

The purpose of the program is to establish the following:

a. The quality of the fuel oil in each 7-day fuel oil tank is within the acceptable limits specified in Table,l of ASTH D-975-1989 when tested every 92 days; and

b. Total particulate concentration of the fuel oil in each 7-day fuel oil tank is x 10 mg/1 when tested every 92 days in accordance with ASTH D-2276, Method A-2 or A-3.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program testing frequencies.

5.5.10 Technical S ecifications TS Bases Control Pro ram This program provides a means for processing changes to the Bases of these Technical Specifications.

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

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

l. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

(continued)

BFN-UNIT 2 5.0-14 Amendment *Rl

Programs and Manuals 5.5 5.5.10 Technical S ecifications TS Bases Control Pro ram (continued)

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

d. Proposed changes that meet the criteria of Specification 5.5. 10b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

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

a. Provisions for cross division checks to ensure a loss of the capability to perform the safety function assumed in the

'ccident analysis does not go undetected;

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

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

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

(continued)

BFN-UNIT 2 5.0-15 Amendment *Rl

k Programs and Manuals 5.5 5.5.ll Safet Function Determination Pro ram SFDP (continued)

a. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or

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

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

5.5.12 Primar Containment Leaka e Rate Testin Pro ram A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 49.6 psig. The maximum allowable primary containment leakage rate, L., shall be 2X of primary containment air weight per day at P..

Leakage Rate acceptance criteria are:

a. The primary containment leakage rate acceptance criteria is c 1.0 L.. During the first unit startup following the testing performed in accordance with this program, the leakage rate acceptance criteria are c 0.60 L. for the Type B and Type C tests, and x 0.75 L, for the Type A test; and (continued)

BFN-UNIT 2 5.0-16 Amendment *Rl

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primar Containment Leaka e Rate Testin Pro ram (continued)

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate z 0.05 L, when tested at a P..

2) Air lock door seals leakage rate is x 0.02 L, when the overall air lock is pressurized to a 2.5 psig for at least 15 minutes.

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

The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

BFN-UNIT 2 5.0-17 Amendment *Rl

Reporting Requirements t

5.6

5. 0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Occu ational Radiation Ex osure Re ort NOTE A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors), for whom monitoring ,

was performed, receiving an annual deep dose equivalent > 100 mrems and the associated collective deep dose equivalent (reported in person - rem) according to work and job functions (e.g., reator operations and surveillance, inservice inspection, routine maintenance, special maintenance[describe maintenance], waste processing, and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, thermoluminescence dosimeter (TLD), electronic dosimeter, or film badge measurements. Small exposures totaling < 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total deep dose equivalent received from external sources should be assigned to specific major work functions. The report covering the previous calendar year shall be submitted by April 30 of each year.

5.6.2 Annual Radiolo ical Environmental 0 eratin Re ort NOTE A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with (continued)

BFN-UNIT 2 5.0-18 Amendment *RI

Reporting Requirements 5.6 5.6.2 Annual Radiolo ical Environmental 0 eratin Re ort (continued) the objectives outlined in the Offsite Dose Calculation Hanual (ODCH), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

5.6.3 Radioactive Effluent Release Re ort NOTE A single submittal may be made for a multiple unit station. The submittal shall combine sections common to all units at the station; however for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to Hay 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCH and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B. 1.

5.6.4 Monthl 0 eratin Re orts Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.6.5 CORE OPERATING LIMITS REPORT COLR

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

(1) The APLHGRs for Specification 3.2.1; (2) The LHGR for Specification 3.2.3; (continued)

BFN-UNIT 2 5.0-19 Amendment *Rl

Reporting Requirements 5.6 5.6.5 CORE OPERATING LIHITS REPORT COLR (continued)

(3) The HCPR Operating Limits for Specification 3.2.2; and (4) The RBH setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel,"

(latest approved version for BFN).

c ~ The core operating limits shall .be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDH, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 PAH Re ort Mhen a report is required by Condition B or G of LCO 3.3.3.1, "Post Accident Honitoring (PAH) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

BFN-UNIT 2 5.0-20 Amendment *Rl

High Radiation Area I

5.7

5. 0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area As provided in 10 CFR 20, paragraph 20. 1601(c), the following controls may be applied to high radiation areas as an alternative to the controls .required by 10 CFR 20.1601(a) and (b):

5.7.1 Each high radiation area, as defined in 10 CFR 20, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., a radiological controls technician) or personnel escorted by such individuals, shall be exempt from the RWP requirements during the performance of their assigned duties in high radiation areas where radiation doses could be received that are < 1 rem in one hour, as measured at 30 centimeters from the radiation source or from the surface which the radiation penetrates, provided they otherwise comply with approved radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the'ose rate levels in the area have been established and personnel have been made knowledgeable of them.

c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the RWP.

5.7.2 In addition to the requirements of Specification 5.7.1, areas that are accessible to personnel and that have radiation levels > 1 rem in one hour as measured at 30 centimeters, but < 500 rads in one (continued)

BFN-UNIT 2 5.0-21 Amendment *Rl

High Radiation Area 5.7 5.7.2 (continued) hour at one meter from the radiation source or from the surface which the radiation penetrates, shall be provided with locked or continuously guarded doors to prevent unauthorized entry. The keys shall be under the administrative control of the duty Shift Hanager, Radiological Controls Hanager, or their respective designees. Doors shall remain locked except during periods of access by personnel under an approved RWP which specifies the dose rates in the immediate work areas and the maximum allowable stay times for individuals in that area. In lieu of the stay time requirement of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by individuals qualified in radiation protection procedures to provide positive exposure control over the activities .being performed within the area.

5.7.3 Individual high radiation areas that are accessible to personnel, have radiation levels ) I rem in one hour as measured at 30 centimeters, but ( 500 rads in one hour at one meter from the radiation source, are located within large areas where no enclosure exists for purposes of locking and where no enclosure can be reasonably constructed around the individual area, shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device whenever the dose rate in the area exceeds or will shortly exceed I rem in one hour as measured at 30 centimeters from the radiation source or from the surface which the radiation penetrates.

BFN-UNIT 2 5.0-22 Amendment *Rl

Improved Technical Specifications All Pages Unit 3

Responsibil i ty 5.1

5. 0 ADMINISTRATIVE CONTROLS

5. 1 Responsibility 5.1.1 The Site Vice-President shall be responsible for overall activities at the site, while the Plant Manager shall be responsible for overall unit operation. The Site Vice-President and the Plant Manager shall delegate in writing the succession to this responsibility during their absence.

The Plant Manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

The Shift Hanager shall be responsible for the control room command function. During any absence of the Shift Manager from the control room while the unit is in NODE 1, 2, or 3, an individual with an active Senior Reactor Operator {SRO) license shall be designated to assume the control room command function.

During any absence of the Shift Manager from the control room while the unit is in MODE 4 or 5, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

BFN-UNIT 3 Amendment

t Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Or anizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a~ Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as'ppropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A);

b. The Plant Manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant;

c. The Chief Nuclear Officer and Executive Vice President, TVA Nuclear, shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and

d. The individuals who train the operating staff, carry out radiological controls, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2 Unit Staff The unit staff organization shall include the following:

(continued)

BFN-UNIT 3 5.0-2 Amendment *Rl

Organization 5.2 5.2.2 Unit Staff (continued) a a A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in NODES 1, 2, or 3.

When all three units are shutdown or defueled, a total of three non-licensed operators shall be assigned for all three units.

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in NODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

c ~ Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and Specifications 5.2.2.a and 5.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore .the shift crew composition to within the minimum requirements.

d. A radiological controls technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e., Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, radiological controls technicians, auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an 8, 10, or 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months day, nominal 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant (continued)

BFN-UNIT 3 5.0-3 Amendment *Rl

Organization 5.2 5.2 Organization 5.2.2 Unit Staff {continued) modification, on a temporary basis the following guidelines shall be followed:

1. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time;

2. An individual should not be permitted to work more than 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months s in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized in advance by the Plant Manager or his designee, in accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the P1ant Manager or his designee to ensure that excessive hours have not been assigned. Routine deviation from the 'above guidelines is not authorized.

f. The Operations Superintendent shall hold a current SRO license on a Browns Ferry unit.

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Manager in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

BFN-UNIT 3 5.0-4 Amendment "Rl

Unit Staff gualifications 5.3

5. 0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff gualifications 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications for comparable positions as specified in the TVA Nuclear guality Assurance Plan (TYA-NgA-PLN89-A).

BFN-UNIT 3 5.0-5 Amendment

Procedures 5.4

5. 0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. . The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;

b. The emergency operating instructions required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;

c. guality assurance for effluent and environmental monitoring;

d. Fire Protection Program implementation; and

e. All programs specified in Specification 5.5.

BFN-UNIT 3 5.0-6 Amendment

t Programs and Manuals 5.5

5. 0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented and maintained.

5.5.1 Offsite Dose Calculation Manual ODCH

'a ~ The ODCH shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program; and

b. The ODCH shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release, reports required by Specification 5.6.2 and Specification 5.6.3.

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2. a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20. 1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b. Shall become effective after review and acceptance by the process described in TYA-NgA-PLN89-A; and c ~ Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCH as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page (continued)

BFN-UNIT 3 5.0-7 Amendment

Programs and Manuals 5.5 5.5.1 Offsite Dose Calculation Manual ODCM (continued) that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.5.2 Primar Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include the Core Spray, High Pressure Coolant Injection, Residual Heat Removal, and Reactor Core Isolation Cooling. The program shall include the following preventive maintenance:

a. Periodic visual inspection requirements; and

b. System leak test requirements for each system, to the extent permitted by system design and radiological conditions, at refueling cycle intervals or less.

Post Accident Sam lin This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following:

a., Training of personnel;

b. Procedures for sampling and analysis; and

c. Provisions for maintenance of sampling and analysis equipment.

5.5.4 Radioactive Effluent Controls Pro ram This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably (continued)

BFN-UNIT 3 5.0-8 Amendment

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Pro ram (continued) achievable. The program shall be contained in the ODCH, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 times the concentration values in 10 CFR 20, Appendix B, Table 2, Column 2;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I;

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days; Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2X of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary shall be limited to the following; 1., For noble gases: a dose rate of x 500 mrem/yr to the total body and < 3000 mrem/yr to the skin, and (continued)

BFN-UNIT 3 5.0-9 Amendment *Rl

Programs and Manuals 5.5 5.5.4 Radioacti've Effluent Controls Pro ram {continued)

2. For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half lives > 8 days: a dose rate of w 1500 mrem/yr to any organ;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

Com onent C clic or Transient Limit This program provides controls to track the FSAR Section 4.2.5, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 Inservice Testin Pro ram This program provides controls fot inservice testing of ASHE Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda are as follows:

(continued)

BFN-UNIT 3 5.0-10 Amendment *Rl

0

'I

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.6 Inservice Testin Pro ram (continued)

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testin activities Meekly At least once per 7 days Monthly At least once per 31 days quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

5.5.7 Ventilation Filter Testin Pro ram VFTP The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

a. Demonstrate for each of the ESF systems (Standby Gas Treatment (SGT) System and Control Room Emergency Ventilation (CREV) System) that an inplace test (continued)

BFN-UNIT 3 5.0-11 Amendment *Rl

Programs and manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testin Pro ram VFTP (continued) of the HEPA filters shows a penetration and system bypass

( 1.(C when tested in accordance with ANSI N510-1975 at the system flowrate specified below, a l(C.

ESF Ventilation System Flowrate (cfm)

SGT System 9000 I

CREV System 3000 This testing shall be performed 1) every 18 months, 2) after partial or complete replacement of HEPA filters, 3) after any structural maintenance on the system housing, or 4) following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass ( 1.% when tested in accordance with ANSI N510-1975 at the system flowrate specified below, a IlC.

ESF Ventilation System Flowrate (cfm)

SGT System 9000 CREV System 3000 This testing shall be performed 1) every 18 months, 2) after partial or complete replacement of the charcoal adsorber bank, 3) after any structural maintenance on the system housing, or 4) following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

c~ Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, shows a methyl iodide efficiency > 901. when tested in accordance with ASTH D3803-1989.

This testing shall be performed 1) every 18 months, 2) after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, or 3) following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

(continued)

BFN-UNIT 3 5.0-12 Amendment *Rl

0 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testin Pro ram VFTP (continued)

d. Once every 18 months demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below at the system flowrate specified below, a 1%:

ESF Ventilation System Delta P Flowrate (inches water) (cfm)

SGT System 9000 CREV System 3000

e. Once every 18 months demonstrate that the heaters for the SGT System dissipate 40 kW f l(C when tested in accordance with ANSI N510-1975.

5.5.8 Ex losive Gas and Stora e Tank Radioactivit Monitorin Pro ram This program provides controls for potentially explosive gas mixtures contained downstream of the offgas recombiners, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

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

b. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tanks'ontents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in (continued)

~ BFN-UNIT 3 5.0-13 Amendment *Rl

Programs and Manuals 5.5 Ex losive Gas and Stora e Tank Radioactivit Monitorin Pro ram (continued) an unrestricted area, in the event of an uncontrolled release of the tanks'ontents.

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

Diesel Fuel Oil Testin Pro ram A diesel fuel oil testing program to implement required testing of the fuel oil in each 7-day fuel oil tank shall be established.

The purpose of the program is to establish the following:

a. The quality of the fuel oil in each 7-day fuel oil tank is within the acceptable limits specified in Table 1 of ASTH 0-975-1989 when tested every 92 days; and

b. Total particulate concentration of the fuel oil in each 7-day fuel oil tank is w 10 mg/1 when tested every 92 days in accordance with ASTH 0-2276, Method A-2 or A-3.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program testing frequencies.

Technical S ecifications TS Bases Control Pro ram This program provides a means for processing changes to the Bases of these Technical Specifications.

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

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

l. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

(continued) 5.0-14 Amendment *Rl

Programs and Manuals 5.5 5.5.10 Technical S ecifications TS Bases Control Pro ram (continued)

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

d. Proposed changes that meet the criteria of Specification 5.5.10b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

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

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

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

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

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

(continued)

BFN-UNIT 3 5.0-15 Amendment *Rl

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safet Function Determination Pro ram SFDP (continued)

a. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or

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

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

5.5.12 Primar Containment Leaka e Rate Testin Pro ram A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 49.6 psig. The maximum allowable primary containment leakage rate, L shall be 2X of primary containment air weight per day at P..

Leakage Rate acceptance criteria are:

a. The primary containment leakage rate acceptance criteria is w 1.0 L,. During the first unit startup following the testing performed in accordance with this program, the leakage rate acceptance criteria are x 0.60 L. for the Type B and Type C tests, and x 0.75 L, for the Type A test; and (continued)

BFN-UNIT 3 5.0-16 Amendment *Rl

Programs and Manuals 5.5 5.5.12 Primar Containment Leaka e Rate Testin Pro ram (continued)

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage r ate a 0.05 L, when tested at w p..

2) Air lock door seals leakage rate is a 0.02 L. when the overall air lock is pressurized to a 2.5 psig for at least 15 minutes.

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

The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

BFN-UNIT 3 5.0-17 Amendment *Rl

Reporting Requirements 5.6

5. 0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6. 1 Occu ational Radiation Ex osure Re ort NOTE A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors), for whom monitoring was performed, receiving an annual deep dose equivalent > 100 mrems and the associated collective deep dose equivalent (reported in person - rem) according to work and job functions (e.g., reator operations and surveillance, inservice inspection, routine maintenance, special maintenance[describe maintenance], waste processing, and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, thermoluminescence dosimeter (TLD), electronic dosimeter, or film badge measurements. Small exposures totaling < 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total deep dose equivalent received from external sources should be assigned to specific major work functions. The report covering the previous calendar year shall be submitted by April 30 of each year.

5.6.2 Annual Radiolo ical Environmental 0 eratin Re ort

-NOTE-A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by Hay 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Honitoring Program for the reporting period. The material provided shall be consistent with (continued)

BFN-UNIT 3 5.0-18 Amendment *Rl

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiolo ical Environmental 0 eratin Re ort (continued) the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

5.6.3 Radioactive Effluent Release Re ort NOTE A single submittal may be made for a multiple unit station. The submittal shall combine sections common to all units at the station; however for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B. 1.

5.6.4 Monthl 0 eratin Re orts Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.6.5 CORE OPERATING LIMITS REPORT COLR a ~ Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

(1) The APLHGRs for Specification 3.2.1; (2) The LHGR for Specification 3.2.3; (continued)

BFN-UNIT 3 5.0-'19 Amendment *Rl

Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT COLR (continued)

(3) The MCPR Operating Limits for Specification 3.2.2; and (4) The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel,"

(latest approved version for BFN).

co The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 PAM Re ort Mhen a report is required by Condition B or G of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

BFN-UNIT 3 5.0-20 Amendment *Rl

High Radiation Area 5.7

5. 0 ADMINISTRATIVE CONTROLS

~ ~

High Radiation Area

~

5.7 As provided in 10 CFR 20, paragraph 20. 1601(c), the following controls may be applied to high radiation areas as an alternative to the controls required by 10 CFR 20.1601(a) and (b):

5.7.1 Each high radiation area, as defined in 10 CFR 20, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., a radiological controls technician) or personnel escorted by such individuals, shall be exempt from the RWP requirements during the performance of their assigned duties in high radiation areas where radiation doses could be received that are < 1 rem in one hour. as measured at 30 centimeters from the radiation source or from the surface which the radiation penetrates, provided they otherwise comply with approved radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them.

C. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the RWP.

5.7.2 In addition to the requirements of Specification 5.7.1, areas that are accessible to personnel and that have radiation levels > 1 rem in one hour as measured at 30 centimeters, but < 500 rads in one (continued)

BFN-UNIT 3 5.0-21 Amendment *Rl

0 High Radiation Area 5.7 High Radiation Area

~

5.7

~

5.7.2 (continued) hour at one meter from the radiation source or from the surface which the radiation penetrates, shall be provided with locked or continuously guarded doors to prevent unauthorized entry. The keys shall be under the administrative control of the duty Shift Hanager, Radiological Controls Manager, or their respective designees. Doors shall remain locked except during periods of access by personnel under an approved RWP which specifies the dose rates in the immediate work areas and the maximum allowable stay times for individuals in that area. In lieu of the stay time requirement of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by individuals qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.7.3 Individual high radiation areas that are accessible to personnel, have radiation levels ) I rem in one hour as measured at 30 centimeters, but ( 500 rads in one hour at one meter from the radiation source, are located within large areas where no enclosure exists for purposes of locking and where no enclosure can be reasonably constructed around the individual area, shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device whenever the dose rate in the area exceeds or will shortly exceed I rem in one hour as measured at 30 centimeters from the radiation source or from the surface which the radiation penetrates.

BFN-UNIT 3 5.0-22 Amendment *Rl

BROWNS FERRY NUCLEAR PLANT - IMPROVED TECHNICAL SPECIFICATIONS SECTION 5.0 LIST OF REVISED PAGES UNIT 1 CURRENT TECHNICAL SPECIFICATIONS MARKUP NOTE: Allpages are provided.

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BROWNS FERRY NUCLEAR PLANT - IMPROVED TECHNICAL SPECIFICATIONS SECTION 5.0 LIST OF REVISED PAGES UNIT 2 CURRENT TECHNICAL SPECIFICATIONS ~UP NOTE: Allpages are provided.

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BROWNS FERRY NUCLEAR PLANT - IMPROVED TECHNICAL SPECIFICATIONS SECTION 5.0 LIST OF REVISED PAGES UNIT 3 CURRENT TECHNICAL SPECIFICATIONS MARKUP NOTE: Allpages are provided.

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UNIT 1 CURRENT TECHNICAL SPECIFICATION MARKUP

TS S fl'ci'Ac 'o O g/ ~0 BFN TECHNXCAL SPECXFXCATXONS ADMXNXSTRATXVE CONTROLS S, )

(EKrPP as ~~y S f ~,W-P~c'dvQ sbn)) ~ resp ns,'etc&C r u~tralt Ocgv;b4>> <f f hc s.Q, a 4 l~

+1.1 The Plant Manager shall be responsible for overall unit operation

~0-lg CJLT'

~ shall delegate in writing the succession to this responsibility during his absence. Thc 6;~ Vsce tres)dcnten< fj2.

~

~ Ctg dna~

5g,'gg panamal.'r

~1. 2 The r ee ring A'bsence from the S control ~om, shall be responsible for the Control Room command function.

pf p MSZa7 b.o-6.2 6.2.1 An onsite and offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A)

Stc yN%4i'egg'o~ c p (/agcS 4'ww x'sr'.~

BFN 6. 0-1 Unit 1

INSERT 6.0-1A The Plant Manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

INSERT 6.0-1B while the unit is in MODE 1, 2, or 3, an intlividual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the Shift Manager from the control room while the unit is in MODE 4 or 5, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

(Ql Zexcep+ ns mantel) Ts 37+

c;girth'on s.~

5 ~

BFN TECHNZCAL SPECZ:"ZCATZONS ADMZNZSTRATZVE CONTROLS Spt Sgc'F445%4iALhb/l pe~ Q,$ angg5 6.1 as~ )SyS S.i r

6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Shift Operations Supervisor (or during his ahsence from the Control Room, a designated individual) shall he responsible for the Control Room command function.

g

~ugly intermediate level and operating organization positions. These relationships shall be documented and updated, as appropriate, i organizatioqg9 charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A)

~~BFN 6.0 "1 Unit 1~~

~~0 ~~~

~~0 ~ 0 0 0 0 ~~~

~~0 ~~~

~~~ 0 0 0 ~~~

~~~ 0 0 0 0 0 0 0 '~~

~~0 ~~~

~~0 0~~

~~~ 0 0 ~ e ~~~

~~0 0~~

~~~ ~ 0 0 ~ 0 ~~~

~~~ 0 0 0 0 0 0 ~ 0 0 0 0 0 4h~~

0'

~LA I A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in NODES 1, 2, or 3.

~~TS 372~~

~~~Xc + aS~(~ Spec)4'~ahern 5. Z 0 f Cc'rnH~A 0~~

~~dur ng 1ce ed~~

/

reactor op rat rs shall 5e i the 'control room y c ld s artups, hil shu ing do th re ct , and d rin e rom uni trip. Zn 'tion, a genio license shall be in the cont ol room ~-chat 5'erato gCccar

<',M4~can'f/esf cdh'k <'s'n iOdc'r (6'C'o) >r @~ c t~g7lOgIMI 505I~ y when Awe~ fuel in the reactor.

~~Nn 5)fc. (5 5;2,g,~ ~ ~T ~ o-3A R~d CPS (r.8,l,],5 y5l P (P 0 4 5 2.2.4 4~ ~SEXT Cc.g-ZB 5'.2.g ~SWAT (r,o -g~~~

/oj fjon ~lip ~ Vocont QC nOQ plO5C, ~n 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided immediate action is taken to fill the required position.

~~gin p.W t ((Ou5de W An~~ aksoncc>~~

~~6.0-3 BFN Unit 1 PAGE~OF~~

0 Administrative procedures shall be developed and implemented to limit the wor king hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, radiological controls technicians, auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an 8, 10, or 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months day, nominal 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

l. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time;
2. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;
~~3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time;~~
~~4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.~~
Any deviation from the above guidelines shall be authorized in advance by the Plant Manager or his designee, in accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.
Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or,his designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.
PAVE OF~
INSERT 6.0-3B @pl The Operations Superintendent shall hold a current SRO license on a Browns Ferry unit.
INSERT 6.0-3C
g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Manager in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.
PAGE
1 2 Senior Operators 1 1 1 1 SRO Senior Operator 0 1 2 SRO Licensed Opera rs 3 3 3 RO or SRO Additional censed Operatorsc 0 1 2 2 RO or 0 t Unit 0 era ors AUO) lC J4 3 8'0 5 ~e hift 'fechx Mal AdviaW (STA) 0 1 1 1 Health Physics Technician 1 1 .1 1 None
~~a. A s ior oper or vill be assi ed responsibi ty fo lant 0 cra't a is fu in an )p egg +o~sv Q)5)(P p~~~
So sPecs6'tabes s.q,t, he shif t crew for a period of time not to exceed tvo hours in order to accoaanodatc unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to vithin the minimum requirements of Table .A. This provi on docs n perm any shift cree position t be ed upon shift hange due an omin shift aAwman being ate or sen't o
~~c. One of the Additiona Lxcen cd Operators be assigned to eac control room th an operating uni .~~
~~d. The umber of required censed personn , <<hen the op ating units arc cont oiled from a cotmn control room, are tvo senio operators and four operators.~~
BPN 6.0-4 AMENOMERTHU. T4 9 Unit 1
+$ 3~g
.7.1 (Cont'd) Spec,(:cuban 5.2.
e d. Critical operation of the authorized by the Commission unit shall not be resumed until See Se4:4'o k.~ 0 C4 8/hl IS~ g,2.
6.8 6.8.1 PROCEDURES 6.8.1.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
~~a. The applicable procedures recommended in Appendix A of e latory Guide 1.33, Revision 2, February 1978.~~
Limitations on the amount of overtime wdrked by individuals 5'2.g.e.
performing safety-related functions in accordance with NRC A3 Policy statement on working hours (Generic Letter No. 82-12)
~~c. Surveillance and test activities of safety-related equipment.~~
~~d. (Deleted)~~
~~e. (Deleted) f." Fire Protection Program implementation.~~
~~g. (Deleted)~~
~~h. (Deleted)~~
~~i. Offsite Dose Calculation Manual.~~
,j. Administrative procedures which control technical and cross-disciplinary review.
~Qcggon Q BFN 4'>>
S+
)5Ts 5,g g( ~~)
~~6. -6 Unit 1~~
~$ E7P SD. I O,p 5 cificah'on 5.3 (ense y~e)
Each member of the unit's staff shall meet or exceed the minimum qualifications for comparable positions as specified in the TvA Nuclear Quality Assurance Plan (TVA-NQA-PLN89-A) .
SC'e PNSM 'A~ ~~
6.7 0r 8Phl lDTS.'2 ?.
6.7.1 The following acti'ons shall be taken in the event a Safety Limit is violated:
a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The President, TVA Nuclear and Chief Nuclear Officer, and the NSRB I shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
~~b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PORC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects~~
~~'of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence.~~
~~c. The Safety Limit Violation Report shall be submitted to the Commission, the NSRB, and the President, TVA Nuclear and Chief Nuclear Officer, within 14 days of the violation.~~
~~6. 0-5 r q'gp~9.- p g0 BFN Unit 1~~
0 0
TS 87p
.7.1 (Cont'd) f ~+AM+ 4.
r Bpu (Srs i~ c~ SPcc.i4mhon S.R
~~d. Critical operation of the unit shall not be resumed until authorized by the Commission.~~
5:q PROCEDURES Se9
.~ Written procedures el shall be established, implemented and maintained covering the activities go I(au;~
~~a. The applicable procedures recommended in A endix A ~~~~
~~Regulatory Guide 1.33, Revision 2, February 1978.~~
b. Limitations on the amount of overtime worked by individuals performing safety-related functions in accordance with NRC Policy statement on workin s Generic Le -12 Scc %aftiAcaliM 4 chaniCt) h 8'srs S.~
~~c. Surve lance an yvqae moment.~~
~~d. (Deleted)~~
~~e. (Deleted) 5Q )Q~ Fire Protection Program implementation.~~
~~g. (Deleted)~~
~~h. (Deleted) ffsit Dose Calcul~ion Manbel.~~
inistra 've proce res which con ol tec zeal nd cross-
/y dis iplinary eview.
('Q s.e.[,1,~
BFN ~~ 5',q 6.0-6 Unit 1
~ S.g, i,c~
pAGE~1OF
~Ill ERE
~~b. The emergency operating instructions required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; R~~
c.
~ guality assurance for effluent
~
and environmental monitoring; PAGE~GP~
DRZLLS gP>
.8.2 Drills on actions to be taken under emergency conditions invol 'ng release of radioactivity are specified in the Radiolog'l Emergency Pl and shall be co ucted annually. ual drills sha also be conducte on the actions to be taken following failures of safety- related systems or components.
RADZATZON CONTROL PROCEDURES 6 ~ 8.3 Radiation Control Procedures shall be maintained and made available to all station personnel. These procedures shall contain radiation dose limits and shall be consistent with the requirements of 10 CFR 20. This radiation protection program shall be organized to meet the requirements of 10 CFR 20 except for the "control device" or "alarm signal" required by 20.1601(a).
6.8.3.1 Each high radiation area as defined in 10 CFR 20 shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiological Work Permit. Zndividuals qualified in radiation protection procedures (e.g., a radiological control technician) or personnel escorted by such individuals, shall be exempt from RWP requirements during the performance of their assigned duties in high radiation areas with radiation dose rates equal to or less
/pe 5~g Ji Ci$ '~ At'~g~
4r EP<<sr< $ .7 i~
PAGE OF 6.0-7
TS 372
~~6. 9. 2 (Cont' SCC Bp<~~
5~~:Cg~~
isis s.s gO~ (~ Qm Sgf.c,l.~h'on S, ~
further procedural, hardware or operational changes to be incorporated into the site diesel generator improvement program and the schedule for implementation of those changes.
~~8. (Deleted)~~
9. High Range Primary 3.2.F Within 7 days Containment Radiation after 7 days of Monitors and Recorders inoperability.
10. Wide Range Gaseous 3.2.F Within 7 days Ef fluent Radiation after 7 days of Monitor and Recorder inoperability.
Cert@A <<t <<ar ~~)
s.s.)
Li (cffmc'ntMKd hanges to the ODCM:
T~~eRT p~gc
~ CTs l,o-/D mrs l,O.BS f g tnt~c4(
Shall be documented and records shall be This documentation shall contain:
~s>
Puffioient information to support the ohangePtogether with the appropriate analyses or evaluations justifying the chang > ~
BFN 6.0-20 PAGE~LoF~
Unit 1
@+i (s)
Jy. g deteteinaticn that the change will maintain the leva+sf radioactive effluent control pursuant to 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I oe 10 CFR Bast 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculation >
Shall become effective after review and acceptance by the process described in TVA-NQA-PLN89- ) ar ol I 5,5,1m Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCN as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the QDCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.gea month/year) the change was implemented.
BFN 6. 0-21 PAGE~OF~
Unit 1
0 TS 3'7 ISTIC +CC s 4 Cg fied 5, 5 Sss >~>Rssiisr 4 Shenys) 1.0 (Cont'd) ~s'PH /,o A functional test is the manual operation or initiation of a system, subsystem, or component to verify that it functions within design tolerances (e.g., the manual start of a core spray pump to verify that it runs and that it pumps the required volume of water).
X. ~~gwg - The reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alterations
~
are being performed.
Y. - An engineered safeguard is a safety system the actions of which are essential to a safety action required in response to accidents.
- A reportable event shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
AA ~ ~~JJ
&C Qn S,5,f - @hall contain the metho ology and parameters use xn t e calculation of offsi doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring+a ip etpoints, and in the conduct o
'l nvironmental Radiological Nonitoring Progra ~ The ODCM shall a so contain ~the 74dioactiv
+fluent controls d vzronmental lgbnitoring descriptions of the information that should be included in the Annual Radiological Environmental
~~
Operating and Aaawh Radioactive Effluent Release eports required by Specificationg d'a.~ and SfCLo i OA gsgrs Q CC. - T e con zo e process o discharging air or gas from the primary containment to maintain temperature, pressure, humidity, concentration, or other operating condition in such a manner that replacement air or gas is required to purify the containment.
DD. ~~SU.
EE. MelrJ~
FF. ~a~ - The controlled process of discharging air or gas from the primary containment to maintain temperature, pressure, humidity, concentration, or other operating condition in such a manner that replacement air or gas is not provided or required. Vent, used in s stem names, does no v ntin rocess 0 PA(iE Vi'FN
~~1. 0-10 Unit 1~~
~p o ultogg MAY1g >ggg l%t. Survc llance equir ts fo ASM' c on XZ V 5 4 o I rvi T ting of de~ Code Close l, 2> ssd 3 eosoeeeets~
~%te- 7lepio ~a~ i'ncl&<
e tes o ESNE Code
~4'nserv Class 1, 2, and 3 p s and valves the ll be erfozmed Boiler and Press e Code accordan vith Sects XZ of applicablc ddenda as re red by 10 CFR SO, S ction SO.S (g), except ere spe ific vrit cn relief been g tcd by the Commission pursuant to gad 'i PCCO&nCJ CS specified in Section XE of thc ESNE Boiler and Pressure Vessel Code and applicable Addenda for crv e test ac v t cs rcqa rc e Boiler and Pres are Vessel Code and a licable Addenda 1 b applic lc as follovs in these technical specif atio sc 4$ le< s ASME Boiler and Prcssare Vessel Reqaired frequenc es Code and applicable Addenda for performing inservicc terminology for inservice v Meekly At least once per 7 days Nonthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Every Semiannually or every 6 months At least once per 184 days
/4 I ~ 9 months Yearly or annually bwania)ly oe evcdq ~gmdc At At k+
least least ICod.S +
once once ass CC per per err 276 days 366 days The provisions of are applicable to the .
f above required rcqucncics for performing inservice testing actiViticas 5'a ,P,2
~~4. erfo e~~
e of in aihlition to a vc r ecif ce test ll activit+are ir entb s 1 d ~
~ Kothing in thc kSNE Boiler and Pressure Vessel Code shall be construed to supckscdc thc rcquircmcn'ts of any tcchnical speci fication.
6. e inse ce inspect program for iping idea ified in Qcncri Letter 88< shall be perf ed in ac rdance vi thc sta positions schedule, me ods, pers e, and sample ezp ion include this ene letter.
P3 I
~g (god y.ons 2 SR Z.o. 3 u ~rP >'~~~
~En'CHVo H
~~BHf 1 0-12 emomsL ~66 Unit 1 PAGE e o. ~~ <-'~~
~~TS Z7Z A<ifi~bon S.g~~
~~6. 8. 3 (Cont'd) Sce gwwQ~fj~ go~ g~~~~~~
Qr BAN lSTS S:T shall be un er t e administrative control of the duty Shift Operations Supervisor, Radiological Control Manager or their respective designees. Doors shall remain locked except during periods of access by personnel under an approved RWP which specifies the dose rates in the immediate work areas and maximum allowable stay time for individuals in that area. Zn lieu of the stay time requirement of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by individuals qualified in radiation protection procedures to provide positive exposure control over the activities being performed in the area.
6.8.3.3 Individual radiation areas that are accessible to personnel, have radiation levels greater than 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months as measured at 30 centimeters, but less than 500 rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 1 meter from the radiation source, are located within large areas where no enclosure exists for the purpose of locking and where no enclosure can be reasonably constructed around the individual area, shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device whenever the dose rate in the area exceeds or will shortly exceed 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
8( (cmfr4 45 ~~ "~)
C.S Pnqi~< ~ ~<~~<
The following programs shall be established, implemented, and maintained.
BFN 6.0-9 Unit 1
S, s. g 6
~;g ~program conform~
for the control of radioactive effluents J
~ fc and 10 CFR for 50.36a maintaining the doses to MEMBERS OF THE PUBLZC from radioactive effluents as low as reasonably achievable. The program~shall be contained in the ODCM, ~shall be implemented by epaee~g procedures, and ~shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following Ll elements:
~~~;~l Gm~:I ~~~~
~~and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the OD~~
b. Limitations on the concentrations of radioactive material released in liquid effluents to 404 conforming to 10 times the concentration values stated in 0 CFR 2 4 Anpendix B, Table Column 2> +c
~~c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the DDclgj~~
d. Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLZC from radioactive PAGE ~ OF BFN 6.0-10 Unit 1
0 37).
5
.s;q'Cont'd) <~ep4 es ~~8'S S+cif ccchh< Q 5 materials in liquid effluents released from each unit to UNREsTRzcTED AR~> conforming eo A endix g to 10 CFR Part 50>
e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current year in accordance with the methodology and parameters in the ODCM at least every 31 days.
ace~l Co b: log f Limitations
~ on the and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioa tivity when the projected doses in a 31 da perio ould exceed g~~t of the guidelines for the annual dose or dose commitment conforming te Appendix ~10 CFR Part 50~ /0
~~g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas ~~+beyond the SITE BOUNDARY shall be limited to the following:~~
~~1. For noble ga a dose rate of 500 mrem/yr to the total body and 3000 mrem/yr to the skin, and~~
~~2. For odine-131, odine-133, tritium, and for all radionuclides in particulate form with half-lives days:~~
~~ 8 1500 mrem/yr to any organj BFN 6 ~ 0-11 ~AG~ >~ Oe 5~
Unit 1
s, s.g (Cont')
on
~~h. Limitations the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the Sggg B~PPg conforming to Appendix D 10 cpR part 30<~~
~~i. Limitations on the annual and quarterly doses to a~~
@@+ PP Q jfggg from )odine 13-3, iodine 133,-
tritium, and all radionuclides in particulate form with half -lives 8 days in gaseous effluents released from each unit to areas beyond the B~Q conforming to Appendix 3 10 CPR Part 50>
j. Limitations on the annual dose or dose commitment to any ~Pg P~ /~kg dne to releases of radioactivity and to radiation from uranium fuel cycle sourc conforming to 40 CFR Part 190.
ftec 384+ftca,g on *d'A4geS 8vN Isrs s:s's-6.8.4.3 PRZMARY CONTAZNMENT LEAKAGE RATE TESTZNG PROGRAM A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test program, dated September 1995".
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 49.6 psig.
BFN 6. 0-12 Unit 1
m g72
g

rake+ 5.5 (Cont'd) e SeS ~ 'o~ fog Changer P s The maximum allowable primary containment leakage rate, La, at Pa, shall be 2t of primary containment air weight per day.
~~Leakage Rate acceptance criteria are:~~
a. Primary Containment leakage rate acceptance criterion is S 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are S 0.60 La for the Type B and Type C tests and g 0.75 La for Type A tests;
~~b. Air lock testing acceptance criteria are:~~
(1) Overall air lock leakage rate is g 0.05 La when tested at ? Pa, (2) Air lock door seals leakage rate is s 0.02 La when the overall air lock is pressurized to p 2.5 psig for at least 15 minutes.
'9+ ~~ prew'ICE Con&ik the&
ensure the capability faseS.
to obtain and analyze reactor coolant, radioactive media~
and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. Agree
/~ t+ffp s shall include the following:
Training of personneQ
~) procedures for sanpling and anaiysi~eed g ~i) Provisions for maintenance of sampling and analysisqeqc'qflic~q, BFN 6.0-13 Unit 1
4 H
d jt c'Sat cn 5'. 5" 4 CO S S S (os oe agree go)
NAP 3 0 IQQP G CO OHS 0 0 HTS 3 '.B. ta b Gas atme t S ste Except as specified in e7 t 1 st on e p r yphr, Specification 3.7.B.3 belov, th fo ov ng onditio all three trains of the s ll e em tpate standby gas treatment system shall be OPERABLZ at all g,g,7,g ~4Pressuke
<a ~~ de,~~~~
drop across times vhen secondary the o bined HEPh containment integrity is filters, and charcoal required. adso er banks is less th " inches of vater at a flov of 9000 cfm eC bc 5'uStlficdfion Qc Qsn5/g Bye lsTS ~.~ + 3 (g 10K)~
The inlet eetets dasu~cn4r on
~
each circuit are tested in accordance vith ASSI H510-1975, and are capable of an output of at least 40 kM.
~~c. hir istr ution is orm v thin X~~
-L a oss Ph fi ters coal dsorbers.
BFS 3.7/4.7-13 AMENDMENT NO. T 74 Unit 1 PAGE~V- OF gg
~H cold results Thc DOP of the in-place and halogenate (4
. The tests and sample analysis of hydrocarbon tests at g 10 Specification 3.7.B.2 design flovonHEPh filters shall be performed~
5 7 and charcoal adsorber banks shall shov 299K DOP removal -eyekc-emonce every and 299K halogenated 18 months ~ichever-hydrocarbon removal vhen tested in accordance vith or after every AHSZ H510-1975. 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of s stem operation and folloving significant painting, fire, or chemical release in any.
ventilation rone communicating vith the system.
b. The results of laboratory Af. Cold DOP testing shall carbon sampl analysis be performed after 8 5 7 < shall shov 0 radioactive each complete or partial methyl iodide removal vhen replacement of the HEPh tested in accordance vith filter bank or after any hSTN D3803. structural maintenance on the system housing.
0'+8
~ System operate shall be shown vithin ply to design
+ Haloeenated testing shall hpdrooarnon be flov. performed after each corn P lete or P artial
~S 7.b replacement of the charcoal adsorbcr bank
%c frost'c:o~s 4 SR 5.o 2 d or after any structural 5R30g~~pj;mlle 4 M maintenance on thc gF$ c ucncicS. system housing.
NENSMmr e.
BFI 3.7/4.7-14 Oncet 1 PAGE X~ GF~> @"~
I C4444os ~ 5 W~C<~go&
thai)cs ~ 8C'A I Jf5 3.7.3 FE81s i~s5 3.7.E. 4o7.E
1. Except as specified in l. At least once every 1S months, Specification 3.7.E.3 belov, the pressure drop across thc both control room emergency combined HEPh filters anl prc'ssurisation systems <k7l charcoal adsorber banks shall shall be OPERABLE at all bc demonstrated to bc less than times vhen any reactor 6 inches of vater at system vessel contains irradiated design flov rate + 10Z).
~~fuel.~~
2. a. The results of thc inplace 2. a. The tests and sample cold DOP and halogenatcd .analysis of Specification hydrocarbon tests at design 3.7.E.2 shall be performed S.S. 7.w flove on HEPT filters and
$ . 5. 1,g charcoal adsorber banks once every shall shov g99Z DOP removal 1S months, and 299Z halogenatcd hydrocarbon removal vhen S.57 r after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of tested in accordance vith system operation fUll kESI 5510-1975 ~ ollovtng, significant 5.5:1. painting, fire, or chemical release in any ventilation xone ccaaanmicating vith the systems
b. The r ts of laboratory b. Cold DOP testing shall be carbon ample analysis shall performed after each shov radioactive methyl complete or partial iodide removal at a velocity replacement of the HEPT vhen teated in accordance filter bank or after any vith ASTN D3803, structural maintenance on the system hyusing.
f<>ccrc as mrhd)
IFI aait l
~ ~ 3.7/4 7-19 NENDMKNTNO.
PA<<>~ OF~Og~v.a 2 I5
S <cd~cahon S.S 4 CO S S S APR 0 9 ]993 G CO 0 S 0 OP 0 3.7.E. C o o 0 a o
c. System flov rate shall bc . Halogcnatcd hydrocarbon shovn to be vithin glOX testing shall be performed design flov vhen tested in after each complete or
, accordance vith ASSI partial replacement of the H510-1975. 5.s.s.k. chircoal adsorber bank or after any structural maintenance on the system housing.
~~d. Ea circuit shall be operated at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month.~~
3. From and after the date that 3. ht least once every 18 months, one of thc control room automatic initiation of the emergency pressurization control room emergency systems is made or found to pressurization system shall be be inoperable for aay reason, demonstrated.
REACTOR POWER OPERATXOSS or refueling operatioas arc permissible only during the succeeding 7 days unless such circuit is sooner made OPERABLE.
. If these conditions cannot be 4. Duriag the simulated automatic actuation teat of this system met, reactor shutdovn shall bc initiated and all reactors sha11 be in COLD SHUTDOWN (aee Table 4.2.C), it shall be verified that the necessary vithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for REACTOR dampera operate as required.
POWER OPERATIORS and refucliag operations shall be terminated vithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
Sf@ 3us+S >~eB paar ghgqgg g, Bp~ t~r5 z,q,y 3.7/4.7-20 AMENDMENrgg. Z93 BFH Unit 1 '~n~c '7~ ~ g~ A@V'3
C'Cist Cd9 d4 S.
SEP 2 2 S'tttteet<<<<ted) 1993
~~4. (Deleted) 4. (Deleted)~~
~~We dt~~
~~5. The maximum activity to be 5. (Dele ted )~~
~ P contained in one liquid
'radwaste tank or temporary storage tank that can be discharged directly to th environs shall not exceed 10 curies excluding tritium and dissolved/entrained noble gas.
~~6. With rad oactive liquid ~~~
G.s., Th quan y of waste ceeding 3 .h.5 adio ive +rial limits without lay con ined any sus d all add iona of o side quid radi active ma erial to the radw e stor e tan and with 48 ho s, t shal e re uce the t k cont ts to ete to b w thin the imit. eats wit the e ding to this con ition 1 t by lysin t be r ported the nex pres tive hnnual ioactiv Effluent saap of th tank 's Release Report (S ction 5.2 eats least of the ODCN). ce pe 7 days radi ctive rials re be Lpi the t g'ded P+uiStOnS'f SC 340.2 unA
~l 3.d9 3 eC, pgltCabiC Ergbc've i'cs a~a Sa~~ T~k Cllfipgg fgVe b~ /gad /9d'g ~Aha Sue'veil[ance ~uencics BFN 3.8/4.8<<2 m@D~So. 199 Unit 1 PAGE ~~ OP~ Bee.L
S Cgf 0 5 ~ 5
~PI csc~ptrs ~art) gEp g p sM
~~1. (Deleted) l. (Deleted)~~
~~2. (Deleted) 2. (Deleted)~~
~~3. (Deleted) 3. (Deleted)~~
4. (Deleted) 4. (Deleted) 5'. 5'.g.a . The concentration of 5~ (Deleted) hydrogen downstream of the recombiners shall be determined to be within the limits of 3.8.B.9 by continuously monitoring the off-gas whenever the SJAE is in servic sing rument desc bed in Table 3. K. I t t surveil ce requir ents are syeci ied Tabl 4.2.K.
~~6. (Deleted)~~
~~7. (Deleted)~~
8. (Deleted) s.s.g Whenever the SJAE is in service, 'the concentration of hydrogen in the offgas downstream of the recombiners shall be limited pl 10 'ith the concentration of hydrogen exceeding the limit of 3.8.B.9 above, restore the concentration to within the imit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .
BPH 3.8/4 '-3 Neon'O. 19 9 Unit 1 0
5 op~@
0 SEP 2 2 1993 S 0
.2.K. 4,2.K.
c t t 0 The czplosivc gas l. Each o the explosive axmLtortug qua ts gas toriag listed in Table .2.K shall inst ents shall. be be OPZRILELZ vi the applica- trated OPERLRIZ bility as in Tables by erfoammce of 3.2.K/4.2.K. Llaza t ts in accordance setpoints be set to th Table 4 2 K ensure tha the 1haits of Specifics on 3.8.B.9 are no't czc The act on required vhen number of OPEIUQKZ channe is le s than the Ninianan els OPEBhBLE requir ent
.is ecified in the not s for Ta e 3.2.K. Ezert be t Lgf ef orts to return the truments to OPE atua vithin 30 da uccessful, prep e and and, if submit a special r port to the commission purs t to Specification 6..1.4 to mplain vhy the inoperabi ity vas not correc d in a t ely manner.
~~3. (Deleted)~~
~~4. (Deleted)~~
~~5. The provisions of Specification 1.0.C are not applicable.~~
NENDMENTNO; Z g g BFS 3.2/4.2-6 Unit 1 PAGE~b'F gD
0 1ABLE 3.2.K lllni Channelsl Intrm&
~~1. (Oeleted)~~
~~2. (Oeleted)~~
~~3. (Oeleted)~~
~~4. (Oat 'ted)~~
~~5. OFF GAS OROGEH ANALYZER (H2A, H2~~
~~6. (Oeleted) c hJ I~~
4J C)
~
' ~
~ ~
.~
D.,
~ ', ~ P
~ ~
~"
P SFtc s C>cQ *<7~
SEP 22 1993
+(Deleted)
+*(Deleted)
~~~During main condens offgas treatment system operation LQXIQKk (Del~d) dRGQKJk (Delete'd) hQXZQ~~~
(QaLae.ed)
(Delet d)
Vi the number o channels OPERABLE les than required by e Kinsman Channels OPERABLE reqair ent, operation of main condenser offgas t eatment system may continue proved that a temporary mo tor is installed 'o grab samples are collected at 1 t once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months analyzed vithin e folloving 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
(Deleted)
Zer 3.2/4.2-39 <ENGMENT N6.
Unit 1 1gg PAGE~OF $ 0
Chare 7 f'un tlonal hmfzument
~~1. {Oeleted)~~
~~2. (Oeleted)~~
~~3. (Oeleted)~~
~~4. (Oeleted)~~
~~5. OFF GAS ROGEH ANALYZER (H2A, H ) R(Nl~~
~~6. {Oel ~ ed)~~
5+c4ic~ 5.S 0 0 SEP 2 2 1993 (1)'Deleted)
(2) (Deleted)
(3) The charm calibration sha include the use of tandard ga ples conta a nclhiDkl:
~~a. Zer votuue percent owen (campressed ) and, f~~
b.. mime percent men, balance trodden.
(4) ( lated)
(S) (Deleted)
(6) (Deleted)
BFH Unit 1 3.2/4.2-63 NENDMENT ND. I9 9 PAGE 55 pF~C
OCT 2 5 1993 3-9.A. 4.9.A.
3-9.A.l.c- (Cont'd) 4.9.A.l.b (Cont'd)
4) The Athens 161-kV line is available to the units 1 load sequencing, and and 2 shutdown boards operates for greater through a conmon than or equal to five station-service minutes while its gener-transformer when unit 1 is ator is loaded with the in Cold Shutdown and unit 3 emergency loads.
is not claimiag the Athens line as an offsite source. (3) On diesel generator breaker trip, the loads NOTE FOR (3) AND (4): are shed from the emer-gency buses and the diesel With no cooling tower pumps output breaker recloses or fans running, a cooling on the auto-start signal, tower transformer may be the emergency buses are substituted for a common energized with permanently station-service transformer. connected loads, the auto-connected emergency loads are energized through load sequencing, and the diesel operates for greater than or equal to five minutes while its
~k> 40;~h'an generator is loaded wi:th 4'< 9~m tSTS 3.g. i the emergency loads.
c~ Once a month the quantity of diesel fuel available shall be logged.
IWPisg S,S.q dO Each diesel generator shall be inspected in accordance with instructions based on the manufacturer's 4 d',5. f.b pos recommendations once every months.
Quarterly the quality of each diesel generator's 5.S.9~ (A, B, C, and D) seven-day fuel supply shall be checked. The fuel oil quality shall be within the acceptable limits specified in Table 1 of AS'-D975-89.
BFN 3 9/4.9-3 AMENDMENTNO. 200 Unit 1 PAGE St QF~>
TS gag Spec'F e~ SS' 6.8.4 (Cont'd) ~>~~W'lie is g~.*r r~~ g,~
h. Limitations of the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50.
i. Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Zodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50.
~~j. Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.~~
0M A program P~~~@P P+~
shall be established to implement the leakage rate testing of the containment as required by 10 CFR ~
50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1. 163, "Performance-Based Containment Leak-Test program, dated September 1995". ~
I'he peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 49.6 psig.
BFN 6.0-12 Unit 1 PAGG~XQF
gw The maximum allowable primary containment leakage rate, La, at P shall be 2% of primary containment air weight per day o Leakage Rate acceptance criteria are:
a. Primary Containment leakage rate acceptance criterion is S 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are S 0.60 La for the Type B and Type C tests and S 0.75 La for Type A tests;
~~b. Air lock testing acceptance criteria are:~~
(1) Overall air lock leakage rate is s 0.05 La when tested at p Pa, (2) Air lock door seals leakage rate is Z 0.02 La when the overall air lock is pressurized to ~ 2.5 psig ZA5FRT for at least 15 minutes.
C,o <<4+
PROGRAMS
~~st,-C:eahi A<<he~ k~
8PS l~TS 5',S.q Postaccident sampling activities will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. These activities shall include the following:
(i) Training of personnel, (ii) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis.
BFN 6. 0-13 Unit 1 PAGE~S OF
The provisions of SR 3.0.2 do not apply to the test frequencies, specified in the Primary Containment Leakage Rate Testing Program.
The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.
PAGE~3pF 5D
INSERT SPECIFICATION 5.5.2 INSERT SPECIFICATION 5.5.5 INSERT SPECIFICATION 5.5.10 INSERT SPECIFICATION 5.5.11 ~q PlLCii=~DoF~
c'cKcrpk as ~g) n addition to the applicable reportin re irements of Title 10 de o deral Re lations e following reports shall be submitted al
,'n 4Lccolc4ncr ag,'~ Iy ~Q 5',g S. b.J A t u ion on an annual basis of the number of stat', util y and her per nnel (i eluding co tractors for w m monito ing was equired r ceiving annual d ep dose eguivale t expos es greate than 10 mrem d their associa d man r m exposur accord' to and jo functi s, , reactor operat's and i
eill ce, ins ice pection, utine intenanc special maint ce (de cribe mai enanc, waste proc sing, d refu in Th dose a signment C
v ious d y funct'ons may b esti based n ea e ents self ading do meter; or f '
badg Small osure totalin 0%
of the ind idual tot l dos need not e accoun d for.
n the gregate, leas 80t of e total ep dose a)d equiv ent expos re re ived fr external sources be assigned t spec'c major ork func on
'Wx rc~W 1l M ~~ (Reine ca f sub'Oifeoh r Ce otf kv o H ~.6.I
~ ~
y
)(L single submittal may be made for a multiple unit station.
Th'a lat' su lem nts e re irem ts o 20. 06 o 10 CFR art 0 BFN 6.0-14 Unit 1
RPR-22-S7 TUE 10:06 I ~ ~a 7 /TED-/sz A tabulation on an annual basis of the number of station, utiTity, and other personnel (including contractors), for whom monitoring was performed, receiving an annual deep dose equivalent > 100 mrems and the associated collective deep dose equivalent (reported in person - rem) according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [describe maintenance], waste processing, and ref'ueling). This tabulation supplements the requirements of 10 CPR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, thermoluminescence dosimeter (TLD),
electronic dosimeter, or film badge measurements. Small exposures totaling < 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total deep dose equivalent received from external sources should be assigned to specific major work fiinctions. The report covering the previous calendar year shall be submitted by April 30 of each year.
~ ~ ~ ~
~~b. An mainste m relief va ve that opens in response to reac ing its etpoint or due to op ator a tion to cont l reactor ressure s all be re orted.~~
S,g,g MONTHLY OPERATZNG REPORT Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the S. Nuc ear Regu a o o ssxon, ATTN: Document ntrol sk, Wa hington, D..
~ ~ 20, ~ with a opy to the egional 0 ice, t be submitted no later than the fifteenth of each month following the calendar month covered by the report. narrat e summa of op rating expe'rie e shall e submitted i the abo e schedul
6. 9. l. 4 REPORTABLE EVENTS 0 Reportabl events, prevent re- currence, s i luding corr ive actions ll be reporte to the d measu s to NRC n acco nce with Section 50.73 to 10 CFR 50.
The Annual Radiological Environmental Operating Report covering the operation he unit during the previous calendar year shall be submitted May 15 f each year. A single submitta may
+~' (~ e made for a multi-unit station The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections ZV.B.2, ZV.B.3, and ZV.C of Appendix Z to 10 CFR Part 50.
BFN 6.0-15 Unit 1
'T5 Z7 g spec;4 eagan 5; h
+I C'"c~(< os ~wc)
~~6. 9 .6 OUR TESTS R suits f re ared l k tes perfo ed on s urces if the ests reve the esen of 0. 5 micro rie or ore o remov le co tamin ion.~~
5'.~.s
~~7 CORE OPE TZNG LZMZTS REPORT
((OGRE'.
Core operating limits shall be established T prior to each sapoaatiap cycle, or prioi to any remaining portion of an apez~g cycle, for the following:
(1) The APLHGR for Specification~~ 3,~f ~
(2) The LHGR for Specification ~ 7.2.3 >
3i 2J (3) The MCPR Operating Limit for Specification
~~4) The AP Flow Biased d Block Trip Setting f ecifica 'on 2.1.A.l.c, Table 3.2.C, and Speci 'cation~~
~~3. .L CSf The RBM Upscale (Flow Bias) Trip Setting and clipped value for this setting for Table ~~ 'Z.3'.2,l-l 5 Ce<ACaHan 3 X 2o] p~~
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric I
Standard Application for Reactor Fuel" (latest approved version) .
AG~ ~g 'D~~O BFN 6.0-16 Unit 1
0
'T 5 3C2 Spec;$ ecca5oo 5.A, 8'I (estrse ss ssmysd) g g+ c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, cere the -hydraulic limits, cCS) limits, uucl ear limits Son such as transient analysis limits, and accident analysis limits) bf the safety analysis are met. metgc~cy ~ ~);
~~d. The including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.~~
ms~mush ~g>PqdpP z~ ~f ajygg K+V iq 4cco la~ ddlicat pcFw ace>w The 4jaasal Radioactive Effluent Release Report covering the of the unit> yaw'peration eyeraeien shall be submitted+ The ph 4~g report shall include summaries of tMe quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. single submitta may e ma e for a multx-unxt s a o The submittal should combine those sections that are common to all units at the stationi however, for units with separate radwaste systems, the submittal shall s ecify the releases of radioactive material from each unit The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50e36a and Section ZV.Bel of Appendix I to 10 CFR Part 50.
BFN 6.0-17 Unit 1
~S 3U 2.
+Pec,l: c<<pen 5. ~
6.9.2 SPECZAL REPORTS Repo s on the ollowing eas shall e submitte in wri 'ng to the Di ctor of Re onal Offi , Divisio Reactor ojects.
~~1. Fat' Usag 6. 10 .. q Annual Ope ting ort~~
~~2. Re 'ef Valve Tail '~~
3.2.F Wi in 30 days after inoper-ability f ermocoup e and coustic monit on one valve.
Seism'c Znstrumentation 3.2.J.3 Within 10 days Zno rabilit after days of inop rability.
4. Me eorologi 1 Monitoring 3.2.Q2 Within 10 days R2 Znst entati n afte 7 da of Znoper ility inoperabxlity.
~~6. D as be r trieved from a seismg inst ments actua ed dur' a se'ic ev nt and lyz to de rmine the magnit e of t e vibr ory gro d mo ion. A pecial~~~
R ort s ll be ubmitte withi 10 da s after the ev nt escrib' th magnitu e, fre ency pectrum and esultant effec upon plant fe ures import t to sa ety.
0 PAGE~S BFN 6.0-18 Unit 1
7537K CCI ~iCQg~ S.Je 6.9.2 (Cont'd)
7. Diesel Generator Reliability Improvement Program Report shall be submitted within 30 days of meeting failuze criteria in Table 4.9.A. As a minimum, the Reliability Improvement Program report for NRC audit shall include:
~~a. A summary o all tests (valid and invalid) that occurred within t time period over which the last 20/100 valid tests ze performed.~~
~~b. lysis of failures and determin tion of root causes of failures.~~
c. Evaluation of each of the r ommendations of NUREG/CR-0660, "Enhancement of Ons e Emergency Diesel Generator Reliability in Operatin Reactors," with respect to their application to e plant.
~~d. Identification o all actions taken or to be taken to (1) Correct th zoot causes of failures defined in b above and (2 Achieve a general improvement of iesel generator liability.~~
e. A supplemental report shall be prepared or an NRC audit within 30 days after each subsequent ailure during a valid demand, for so long as the a fected diesel generator unit continues to vio te the criteria (3/20 or 6/100) for the reliabilit improvement program remedial action. The sup cmental report need only update the failure/dema d history for the affected diesel generator uni since the last report for that diesel generator. The supplemental report shall also present an analysis of the failure(s) with a root cause determination, if possible, and shall delineate any BFN 6. 0-19 Unit 1
further procedural, hardware or operational changes to be inco rated into the si diesel gener or improvement rogram and the sc dule for implementation of those changes.
EC,C 9. High Range Primary 3.2.F Within days FMd n~ Containment Radiation after 7 days of ppg Monitors and Recorders inoperability.
f~ IMQ ICg~ CC a~
Wx e ange Gaseous 3.2.F
~ ~ Within 7 days Ef nt Radiation afte 7 days of Monitor an Recorder inoperability.
54stjfsttafioa 4r (4lngcc W 8Ft4 l5ys 5 6.12 Changes to the ODCM:
~~1. Shall be documented and records shall be kept .in a manner convenient for review. This documentation shall contain:~~
~~a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change.~~
~~BFN 6.0-20 Unit 1 PASS~0"~~~I~~
~~6. 8 1 (Cont')~~
0
~
~~6. 8. 1. 2 (Deleted)~~
~~6. 8. l. 3 (Deleted)~~
St@ ~ti4cafion @< Ches)
DRILLS
>~ >>~ tsrs g.y 6.8.2 Drills on actions to be taken under emergency conditions involving release of radioactivity are specified in the Radiological Emergency Plan and shall be conducted annually. Annual drills shall also be conducted on the actions to be taken following failures of safety-related systems or components.
RADIATION CONTROL PROCEDURES 6.8.3 Ra ation Control P cedures shall be mai tained and made avail le to all stati personnel. These rocedures shall contain adiation dose li its and shall be co istent with the requir ents of 10 CFR . This radiation protection program shal be organised to et the requirem ts of 10 CFR 20 except f the "control dev e" or "alarm signal" required by 20.1601(a) .
g g,( ~~ 5,7 Ny), ~laHo~ +~c'a Each high radiation area as defined in 10 CFR 20 shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiological Work Permit. Individuals qualified in radiation protection procedures (e.g., a radiological control technician) or personnel escorted by such individuals, shall be exempt from RWP requirements during the performance of their assigned duties in high radiation areas
&~Rois ~~ ~ ~
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BFN 6.0-7 Unit 1
5.'7. I (Cont'd> ~pc gieaAon 5,7, e~ cnhim~ 805m at 30 centimeters, provided they otherwise comply with approved radiation protecticn procedures for entry into high radiation areas. Any 05 Me@5~ individual or group of individuals permitted to enter such aieas shall be providid with or accompanied by onc or more of the following:
~~a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.~~
~~b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.~~
~~Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have bccn made knowledgeable of them.~~
c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in thc Radiological Work Permit.
4l addition, areas that are accessible to personnel and that have radiation lcvcls greater than 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months as measured at 30 centimeters, but less than 500 rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 1 meter from the radiation source or from the surface which the radiation penctrates shall bc provided with locked doors to prevent unauthorized entry. The keys
" ~Mnucuslt guagdcd BFN 6.0-8 AG E~OF~ORop. 2 Unit 1
Tb 37 2-5'7, m +p< I'f ccagdg
~ 5~ 7 Cont'd) shall be under the administrative control of the duty -Shift Nano. C'r Radiological Control Manager or their respective designees. Doors shall remain locked except during periods of access by personnel under an approved RWP which specifies the dose rates in the immediate work areas and maximum allowable stay time for individuals in that area. Zn lieu of the stay time requirement of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by individuals qualified in radiation protection procedures to provide positive exposure control over the activities being performed in the area.
~i~
<1.>
Zndividual radiation areas that are accessible to personnel, have radiation levels greater than 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months as measured at 30 centimeters, but less than 500 rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 1 meter from the radiation source, are located within large areas where no enclosure exists for the purpose of locking and where no enclosure can be reasonably a5 ~~ ab jo constructed around the individual area, shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device whenever the dose
>4AsMcu.
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l
~
5e,~
4A~ ~c~s cu-
~g rate in the area exceeds or will shortly exceed 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
6.8.4 RADZOACTZVE EFFLUENT CONTROLS/RADZOLOGZCAL ENVZRONMENTAL MONZTORZNG PROGRAMS The following programs shall be established, implemented, and maintained.
SuS44i'Wh'o~ Qr Qg~g Bee (5yL g.q R~ ~
pAGE~WOF BFN 6.0-9 Unit. 1
UNIT 2 CURRENT TECHNICAL SPECIFICATION MARKUP
BFN TECHNICAL SPECZFICATIONS A< QA'. 0 ADMZNIS~TZVE CONTROLS CpVCcp~ CS mal+
LX, 5'. 1 Se c Vice.-Pres Jt-k $ <~l< bc ~i@" @+
4t OvCMI'l CaCAv ~ 4 pg ~ F 4W S 'Q I ~ l: IC. f"C.
Xl.l The Plant Manager shall be responsible for overall unit operation shall delegate in writing the succession to this responsibility NNS CcT during his absence. Wc. s'r v'cc Prrs.'cA~F ~ p,~g
~~4. d-/A C. Pl+i f ga<4)gp~~
.4, aha~
~w s)',>
t S'h'F4 R<etger
%.1.2 The AFring hats>absence from the 5
control gfoom ) shall be responsible for the Control Room command function.
me Sd'cT C.o-/g 6.2 6.2.1 An onsite and offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.
a ~ Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the'orm of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A).
Sw. Nazi 44~ 4r C4~~~~
pf hl )sly Z 6
6.0-1 p~GE~OF~I BFN Unit 2
INSERT 6.0-1A gl The Plant Hanager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.
INSERT 6.0-1B while the unit is in HODE 1, 2, or 3, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the Shift Hanager from the control room while the unit is in HODE 4 or 5, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.
BFN TECHNZCAL SPECZFZCATZONS 0 .0 5A ~ 0 ADMZNZSTRATZVE CONTROLS Ra msC;Ci o:ku~<~C4o-pcs Qjw IS TS 5'. I 6.1 6.1.1 Thc Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his ahsence.
6.1.2 The Shift Operations Supervisor (or during his absence from the Control Room, a designated individual) shall be responsible for the Control Room command function
~gpggAH4 4e/nsite and offsite organization shall be cstablishcd for unit operation and corporate managemen The onsite and offsite organizatioqgshall include thc positions for activities affecting the safety of the nuclear power plant.
~~a. Lines of authority, responsibility, and communication shall be established and defined from the highest management level~~~~
intermediate level d all operating organization positions. These relationships shall be documented and updated, as appropriate, in organizatiorg) charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions. or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Rcport (TVA-NPOD89-A).
0 BFN 6. 0-1 Unit 2
TS 3,/~
S Ce lt CWgjOamt 5, 2 54.2. (Cont'd) A~
1 C.'CX'CCP4 45
~ &eCakue YsCC. ~~ <A~C45 Cg. s~, u e Chief Nuclear Office , shall I
have co orate responsibility for overall plant nuclear safety~
.4~$
shall take any measures needed to ensuze acceptable performance of the staff in operating, maintaining, and providing technical suppdrt in the plant nuclear safety j 4 ra/
,g 45e b>. The Plant Manager shall be responsible for overall(~~safe operation and shall have control over those onsite necessary for safe operation and maintenance of the plan~~
I ad'Ok)'Cg C S" ml>
d. The individuals who train the operating staf carry out s and quality assuzance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from opezating pressures.
~~y A.2.2 WMnt, Staff~~
~~gg +; $~~
~~l~ + 4/4c~~
~~a. itt mannxng r~iramants, a~as a minimgm, a escribed in Tab e 6.2.A and below~~
+At
<IV5C4T t 0-zA A license sensor r actor operato shall be pres at the site at a times when there is fuel in the reactor.
g les4.r~ (go rt'8m b8. licensed~ctor rdgera 5
or shall be in the control room wh fuel xn the reactor.
PAGE OF BFR 6.0-2 Unit 2
0 A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in NODES 1, 2, or 3.
0 ~ 0 ~
0 ~
4k~ 0 0 0 0 ~
. ~
0
~IN KRT 6.-
e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, radiological controls technicians, auxiliary operators, and key maintenance personnel).
Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an 8, 10, or 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months day, nominal 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months week while the unit is operating. However, in the event that unforeseen problems requite substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:
An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time;
2. An individual should not be permitted to work more than 16 hours in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;
3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time;
~~4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.~~
Any deviation from the above guidelines shall be authorized in advance by the Plant Hanager or his designee, in accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.
Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or his designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.
INSERT 6.0-3B
~~f. The Operations Superintendent shall hold a current SRO license on a Browns Ferry unit.~~
~~INSERT 6.0<<3C~~
g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Manager in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.
S ec 4~.4. S:~
JUN ~0 ~q88
~o~to t Oeato e o ce se 1
Senior Operator 1 1 1 1 SRO Senior Operator 0 1 2 2 SRO Licensed Operat rs 3 3 3 3 RO or SRO Additional L censed Operatorsc 0 1 2 RO or S Assistant,AJnit Operators g3 py p'g 5 Honp.
pe~ (AUO)
Shift Wchnical Advisor (STA) 0 1 1 1 H'one Health Physics Technician 1 1 1 1 Hone t g Qp a.
g3" 0 K
A senior o operation at for a tor all vill be LE ed responsibi~ for overall plant times there is fuel in any unit.
, ACe period of time not to exceed tvo hours in order to accommodate t~ ro i4r~~ ygyt)
SpCC aC+vfag ee5 S.2.2.a. o~cl 5:2.2.
ft crev unexpected absence of on-duty shift crev members provided immediate action is taken to restore the shift crev composition to vithin the minimum requirements of Table 6.2. . s prov s on does not permit any i s ift tres p sition to hs ed upon shift change due to/an oncoming shift cre being late r absent.
~~c. One of the Ad itional Licensed Operators must be ass gned to each control room vith operating unit.~~
~~d. The num r of required licensed ersonnel, vhen the ope ting units are contr led from a common cont 1 room, are tvo senior operators and four oper tors..~~
0 BFH Unit 2 6.0-4 AMENDNHTRO, 7gg PAGE
6.8
~~6. 8. 1 PROCEDURES 6.8.1.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:~~
~~a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978.~~
~~52 2..M Limitations on the amount of overtime worked by individuals performing safety-related functions in accordance with NRC Policy statement on working hours (Generic Letter No. 82-~~
~~12) .~~
~~c. Surveillance and test activities of safety-related equipment.~~
~~d. (Deleted)~~
~~e. (Deleted)~~
~~Fire Protection Program implementation.~~
~~g. (Deleted)~~
~~h. (Deleted)~~
~~i. Offsite Dose Calculation Manual.~~
~~j. Administrative procedures which control technical and cross-disciplinary review.~~
See, a~s4Ci caAo 4o C~er k~ EM Isis s:g BFN 6.0-6 Unit 2
T'5 Z7Q 5 cc'ig(c+)-;o+ 5 $
s.3.l QAl ICe 4c p~~)
Each member of the unit's staff shall meet or exmed the minimum qualifications for comparable positions as specified in the TVA Nuclear Quality Assurance Plan (TVA-NQA-PLN89-A) .
544 3Mgl<CiChgo~
4<<GFd IsTJ 2.Z 6.7
~ 7.1 The following actions shall be taken in the event a Safety Limit is violated:
a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The
~~'resident, TVA.Nuclear and Chief Nuclear Officer, and the NSRB shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .~~
b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PORC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence.
~~c. The Safety Limit Violation Report, shall be submitted to the Commission, the NSRB, and the President, TVA Nuclear and Chief Nuclear Officer, within 14 days of the violation.~~
~~d. Critical operation of the unit shall not be resumed until authorized by the Commission.~~
PAGE BFN 6.0-5 Unit 2
0 I
WS 272 5 Poc s fs caQy ~ 5 . 4f 5 f PROCEDURES
~At Cemep& es ms ~)
5: V.)
Written procedures shall be established, implemented and maintained covering the activities
/I r'~
~~a. The applicable procedures recommended in Appendix Regulatory Guide 1.33, Revision 2, February 1978.~~
~~Limitations on the amount of overtime worked by individuals performing safety-related functions in accordance with NRC Policy statement on working hours (Generic Letter No. 82-~~
~~12) .~~
ee W~shCm~4~~ 4o~ C4o g~
8/Al Is~ g.'
~~c. Surveillance nd test act vxtxes o sa ety-re a e A2, e ipment~~
~~d. (Deleted)~~
~~e. (Deleted)~~
~~5. g.j. gW. Fire Protection Program implementation.~~
~~g. (Deleted)~~
~~h. (Deleted) p,3 i. i~fsite D~Calculhcgon Mahsal.~~
' 'nistrative proce es which contro technical an cross-discs linary review.
BFN 6.0-6 Unit 2
QMI The emergency operating instructions required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; c.~ guality
~
as'surance for effluent and environmental monitoring;
~~e. All programs specified in Specification 5.5.~~
V e
DRZLLS 6.8.2 rills.. on actions,.to b taken.under emergency conditions in ving release of radar ctivity are specified 'he Radiolo cal Emergency Plan shall be conducted ually.
Annual drills shall also be condu ed on the actions to be taken following failures of safety-related systems or co onents.
RADZATZON CONTROL PROCEDURES 6.8.3 Radiation Control Procedures shall be maintained and made available to all station personnel. These procedures shall contain radiation dose limits and shall, be consistent with the requirements of 10 CFR 20. This radiation protection program shall be organized to meet the requirements of 10 CFR 20 except for the "control device" or "alarm signal" required by
~~20. 1601 (a) .~~
6.8.3.1 Each high radiation area as defined in 10 CFR 20 shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiological Work Permit. Zndividuals qualified in radiation protection procedures (e.g., a radiological control technician) or personnel escorted by such individuals, shall be exempt rom RWP requirements during the performance of their assigned duties in high radiation areas with radiation dose rates equal to or less than 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 30 centimeters, provided they otherwise comply with approved radiation protection procedures
~8~ J~$ 4l4Crk)gw 4y~ C41 gg f BFN Unit 2 4r g~n/ Is'rz 5: 7 6.0-7 ebs '- / VF~
'TS Z7 2 sp <c.4 QAI (e rqket ..nW)
~~8. (Deleted)~~
~~9. High-Range Primary 3.2.F Within '7 days Containment Radiation after 7 days of Monitors and Recorders inoperability.~~
~~0. Wide-Range Gaseous Effluent 3.2.F Within 7 days Radiation Monitor and after 7 days of Recorder inoperability.~~
>e 'sk:li'c~k(o~.g~ 0 ~ ~~
BPu (sr'h 5:s. I L(c c~$ m ( .0 iM WS(g:87 f~m CT5 ges to the ODCM: P"a< I.o-la, cTS /0, $ 8 Shall be documented and records shall be cg:~
This documentation shall contain:
Ls) jkfficient information to support the chang together with the appropriate a.nalyses or evaloations jnstigying the ohang~zeQ
~s) g dsteatLination that the change will maintain the lovepot Part 190. 10 CFR 50.36a, and ~
radioactive effluent control pursuant to 10 CFR 20.1302, 40 CFR en~ I to-10 CFR not adversely impact the accuracy or reliability of effluent, 50 and dose, or setpoint calculations Shall become effective after review and acceptance by the process described in TVA-NQA-PLN89-PAGE OF BFN 6.0-20 Unit 2
Shall be submitted to the Commission.in the form of a complete, legible copy of the entire ODCM as a part cf or concurrent with the k
Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings'n the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.
BPN 6.0-21 Unit 2
5 h7>
5 cc:Cica,4.io~ 5:6 S'ec. Tggkl Si(~flow Cej- C ~~~J4J 1.0 (Cont'd) for SF< iS~S /.0 W.
- A functional test is the manual operation or initiation of a system, subsystem, or component to verify that it functions within design tolerances (e.g., the manual start of i" it a core spray pump to verify that runs and that pumps the required volume of water).
X. The reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alterations are being performed.
Y. - An engineered safeguard is a safety system the actions of which are essential to a safety action required in response to accidents.
- A reportable event shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
4 7Xc 5:5: I methodology and parameters used
<~ll contain in the calculation of offsite the doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring a rip Petpoints, and in the condu th var nmenta
~
radiological nxtorxng rogra J the radioactive +fluent controls and+a The ODCM shall xo also cal contain Oa
~
environmental monitoring descriptions of the information that should the Annual Radiological Environmental Operating~ and ~saL be included and in Radioactive Effluent Release geports required by Specification<
5 C.Z,~~and ge ig d+0 S CC. - The controlled process of xsc rgb.ng air or gas from the primary containment to maintain temperature, pressure, humidity, concentration, or other operating condition in such a manner that replacement air or gas is required to purify the containment.
DD.
EE.
FF. - The controlled process of discharging air or gas from the primary containment to maintain temperature, pressure, humidity, concentration, or other operating condition in such a manner that replacement air or gas is not provided or requirec.
Vent, used in system names, does not imply a ventin rocess.
C'~c See X~c4 L'c~Ay Co +J 8FN isn I.a PAQE N BFN 1. 0-10 Unit 2

~~+li (rX~~

$ ~y yr<<~ 'PrOVoA'S C'C +mt/ MAY 19 iBRg a v t or Inservice Testing of ASME Code Class 1, 2, and 3 components . Wc p~gyy:~ gLnll a'~~ ~ AIL, skasPPetk nserv ce test ng o ME Code Class pumps and valves s the AS ll bc performed in cordance vith Sect n XI of Boiler and Pressure ode and applicable Addenda as re fred by 10 CFR 50, ection 50.55a(g), ept vhere spec ic vrfttcn relief s been granted by e Commission pursuant to 10 CFR 50, Section 50.55(g)(6)(i). 7<drn Cre'Cia spccfffcd fn Section XI of the hSME Boiler and Pressurt Vessel Code and applicable hddenda op zlscrv ct Boiler d Prtssurc Vc el Code and applicable hdd da shall c applicable a these techni ci ca'tio ' ~g gl owg ASME Boiler and Pressure Vessel Required frequencies Code and applicable hddenda for performing fnservfce terminology for inservice t a v tt 't tv 0 llesly Monthly Quarterly or every 3 months Semiannually or every 6 months ht ht ht ht least once per least once per least once per least onct pcr 7 days 31 days 92 days 184 days I~Every months 9 ht least once per 276 days l~oarly or annnally ht least once per l~+ 366 days Gre<yl<all~ ayo ~~ 2 alta. c Quad cillance re ufrehcnt Hothfng in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any technical specification.
6. Th inserv ct nspection program for pfpfng idcntiffed in HRC neric Letter 88-01 shal be performed fn a ordanct vith th taff positions on ache e, methods, pers el, and sample expansion included fn this generic letter.

p Q C. t l'<<Stra a~ io<$ 4 SR 3.03 ~ ~fP I'c~ + ]pg~a~d +tf, BFR 1.0-12 AhtENOMENT go. ygg Unit 2 <c "~l-o= 'Iq'g ~, ~$ S7+ Cilia(Og 5 g

6. 8. 3 (Cont') 5'gg Jggfe giCC4104 4Q C i+'%f g~ @pe/ I$ 7"S S 7 that area. In lieu of the stay time requirement, of the RWP, direct, or remote (such as closed circuit TV cameras) continuous surveillance may be made by individuals qualified in radiation protection procedures to provide positive exposure control over the activities being performed in'he area.

6.8.3.3 Individual radiation areas that are accessible to personnel, have radiation levels greater than 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months as measured at 30 centimeters, but less than 500 rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 1 meter from the radiation source, are located within large areas where no enclosure exists for the purpose of locking and where no enclosure can be reasonably constructed around the individual area, shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device whenever the dose rate in the area exceeds or will shortly exceed 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Al 8 except ei ~~) The following programs shall be established, implemented, and maintained. g4'S A program conform'L-10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program ~shall be contained in the shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements: 6.0-9 BFN Unit 2 PAGE~QF~ 0 5.5;4 Al de)ccepf- ar ~ (Cont') g.l cubi Q ca a ~ Limitations on the of raQioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the OD

~~b. Limitations on the concentrations of radioactive material released in liquid effluents to ~R6g~ ~~~

conforming to 10 times the concentration values stated i 10 CFR 20. 0 .2 ~ Appendix B, Table 2. Column 2A t C. Monitozing, sampling, and analysis of zadioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM'. Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming A endix I~ o 10 CFR Part 50> Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current year in accordance with the methodology and parameters in the ODCM at least every 31 days> ,$ ~ c'~~C,/'Q gaseous effluent tzeatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a~31'a perio would exceed 2 of the guidelines for the annual dose or dose commitment conforming Appendix to 10 CFR Part 50> BFH 6. 0-10 Unit 2 ~S Z7Q S eci lm4n~ SW (Cont') C I Ctree/ e~ ~ ~)

~~g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas ~~+ beyond the SZTE BOUNDARY shall be limited to the foilowing:~~

c5

~~1. For noble gas: a dose rate of 500 mrem/yr to the total body and 3000 mrem/yr to the skin, and C~~
~~2. For odine-131, odine-133, tritium, and for all radionuclides in particulate form with half-lives gsoaece~~~

~~~>stats S days: 1500 mrem/yr to any organ> ~o~~~

h. Limitations~the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each toit to areas heyocd the ~IOg@lflCcooformiog to Appendix Zz 0 CFR Part 50>

Limitations on the annual and quarterly doses to a gQ5f8 tk g Qg from Qdine-131, +dine-133, tritium, and all radionuclides in particulate form with half-lives ~~see 8 days in gaseous effluents released from each unit to areas beyond the ~+[$6+ conforming to Appendix Z to 10 CFR Part 50 o~g Limitations on the annual dose or dose commitment to any due to releases of radioactivity and to radiation from uranium fuel cycle source ~ onforming to 40 CFR Part 190. BPN 6. 0-11 PA<<>> oF I'I R4~ Unit 2 '4 75 37K S ccegiccgfio+ Q5 (Q (c A' pc~) ~;da w~/g AA ensure the capability to obtain and analyze reactor coolant, radioactive ~~~ 84 SCAN and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. Theee shall include the following: 4.~ Training of personnel b ~ r +H~ Procedures for sampling and analysi Provisions for maintenance of sampling and analysis /PM P~ 4, 6.9 ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10. Code of Pederal Regulations, the following identified reports shall be submitted to the Director of the Regional Office of NRC, unless otherwise noted.

~~6. 9.1. 1 (Deleted) c <ask CircCs~~~

g~ BP'nl (Syg gP ~ C~~~ BPN 6. 0-13 Unit 2 SptciCicc~lon 5:~ ) MAR 3 0 tggo G 0 0 S 0 3.7.B. S st

1. Except as specified in 5 . 3A- At ea t dc pe ye Specification 3.7.8.3 belov, e 1 vi c ndi io all three trains of the tr betimes standby gas treatment system shall hc OPERABLE at all 5S'. 7.+ P sha e IS Preh'sure

~s, o ~~~~c-ed. drop across vhen secondary thc combined HEPA containment integrity is lterq and charcoal required. adsorbcr banks is less L f than at a nches flov of of vater 9000 cfm C ~~, >~~~ T~'4fichf >on (g 10K). Sce gol C~+ J~ Og~ Ep'N IS7S 3.C.43 5.'s.7. ~ O'he inlet heaters on each circuit are tested in accordance vith ASSI 8510<<1975, and are capable of an output of at least 40 kV. c Air gistribution is uni rm vi in 2 acy'oss HE A fil and charcoal adsorbcrs rs BEE 3.7/4.7-13 AMENDMENT No. 17 Unit 2 PAGE~OF grq. 5 0' p,l ~~ cold results The DOP of thc in-place and halogenatcd 2 ~The tests analysis of and sample hydrocarbon tests at g lOZ Specification 3.7.B.2 design flov on HEPh filters shall be performed-at-and charcoal adsorbcr banks +Shst~ncc~cr~pcr Ltkhg 5:51,.$ -.. shall shov g99Z DOP removal wyeke-or- once every Il and g99Z halogensted hydrocarbon removal shen tested in accordance vith scxvkee or after every AHSI H510-1975. 720 ours of stem operation and folloving' ficant painting, fire, or chemical rcleasc in any ventilation zone communicating vith the system. The results of laboratory Cold DOP testing shall

5. S'1.c carbon sample analysis be performed after shall shov 90 a oactive 55 l,~ each complete or partial methyl iodide removal vhen replacement of the HEPh tested in accordance vith filter bank or after ag hSTH D3803 ~ structural maintenance oa the system housing.

c4 System shall be shorn to alogenated hydrocarbon 5'$.7 operate within glOX design testing, shall be 4, 6 lcm'lo~, performed after each complete or partial s.s.q replacemeat of the charcoal adsorber bank or after any structural

~~74. p~vssto~g C,C Jg 3'.o.Z maintenance oa the systek housings~~

~~~ '3 ~re app/;~<~~

~~~ l Fmkcp-c ., s~~

~~SC'C 35$ 4~1C4~1O~ 44~ 5 C1 fle1' s.' C~~s*r gFrv ($ % >.7+ I FEB S 595 3.7 E1 4.7.E~~

l. Except as specified in l. At least once every 18 months, Specification 3.7.E.3 belov, the pressure Crop across thc both control room emergency ~.~ 1 + combined HEPA filters and pressurization systems charcoal adsorber banks Cull shall be OPERA1KZ at all be demonstrated to to be less times vhcn any reactor than 6 inches oi vater at vessel contains irraCiated system design flov rate fuel+ + 10K).
2. a. The results of the inplace 2. a. The teats and sample cold DOP and halogenated analysis of Specification hydrocarbon teats at design 3.7.E.2 shall be performed flova on HEPA filters anC charcoal adsorber baaks ~c-er-once every

%%7.$ shall ahov g99X DOP removal 18 months, and 2,99K halogenatcd hydrocarbon removal vhen 55.$ or after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of teated in accordance vith tea o eration and ASSI 5510-1975 ~ fo s ficant painting, fire, or chemical release in 4$$y ventilation xone coae>mfcating vith the system+

b. The results o ratorg 3f Cold DOP testing sluL11 be carbon le analysis shall performed atter each ahov ioactive methyl complete or partial

~.~ ~C iodide removil at a velocity replacement of the HRPA vhen teated in accordance filter b4ZQc or'fter any vith LSD D3803. structural maintenance on the system housing. (<L<gf i1 w<< -J) %X AMtNMSrNL pi i Unit 2 PAGE ~~ GF~ ~.~ S,,h..4o ~5 APR 0 9 1993 3.7.E.

c. System flov rate shall be . Halogenatcd hydrocarbon shown to be vithin glOX testing shall be performed design flov vhen tested in after each complete or accordance vith ANSI partial replacement of the H510-1975. charcoal adsorber bank or after any structaral maintenance on thc system housing.
~~d. Each circuit shall bc opcratcd at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month.~~
3. From and after the date that 3. At least once every 18 months, one of thc control room automatic initiation of the emergency pressurization control room emcrgcncy systems is made or found to pressurization system shall bc bc inoperablc for any reason, demonstrated.

~~REACTOR POWER OPERATIOHS or refueling operations arc permissible only during thc succeeding 7 days unless such circuit is sooner made OPERABLE.~~

4. If these conditions cannot be 4. During the simulated aatomatic met, reactor shutdown shall be actaation test of this system initiated shall and all reactors be in COLD (see Table 4.2.C), it shall be rerificd that the necessary 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for REACTOR SHUTDOWNS'ithin dampers operate as required.

~~POWER OPERATIOES and refueling operations shall bc terminated vithin 2 hours. IjtEN~~ NL2 0 8 BFK tMt 2 3.7/4.7-2O 4>o-~ @,i ll S cc C 4'~ Sg ('aft+ ~ SEP 2219~~

~~4. (Deleted) 4- (Deleted)~~
5. The max imum activity to be 5. (Deleted) contained in one liquid ra&taste tank or temporary storage tank that can be discharged directly to the environs shall not excee 10 curies excluding tritium an dissolved/entrained noble gas.
6. Pith radioactive liquid waste 5.R exceeding 3 .h.5 limits, 6 Th ity of vithout de ay suspend all dio tiv ma rial con in in y additi materia of radioactiv to the tank d o sid or e li t ks d te

~ith 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , red e the de ed t be th tan contents to v in the e a 1 t. Events le ding to this ana c dition must e reported in the next hnn res tat e s 1e Effluent Rele Radioactive Report

~~f th 's t s e~~

(Section 5.2 of the ODCN). at t onc per s te als en r dio ti La e ad d t the P~aA fdI~ ~K SA 3.0.Z. Sa a,o.3 o ~ op) L'c~ E~los,'vc. ~i o~ J~~ 7~~ 4g QI<r 4v'rre) PreglT ~ ~roag/4 tee Stel Ilc I l~ ~ +jet BFN 3.8/4.8-2 Unit 2 hMENOMEgr gL 2g6

~~l. (Deleted) l. (Deleted)~~
~~2. (Deleted) 2. (Deleted) 3~ (Deleted) 3. (Deleted)~~
~~4. (Deleted) 4. (Dclcted)~~
5. (Deleted) 5.5,$ ,~ ~ The concentration of hydrogen dovnstream of the rccombiners shall be determined to bc vithin the limits of 3.8.B.9 by continuously monitoring the off-gas vhcnevcr thc SJAE is in service nstrumcnts described in Tabl 3.2.X.

~~Inst ent surv llance reg rcments are specified in Table 4.2.X. 6~ (Dcletcd)~~

~~7. (Dclcted) 8~ (Delctcd)~~
~~9. Whenever the SJAE is in scrvicc, the concentration of hydrogen in the offgas dovnstream of the recombiners shall be limited~~
~~10. arith concentration of hydro cxceedi the lim of 3.8.B above r tore the nccntra on t vithin t limit ithin 48 hours.~~

~~BFH 3 8/4 8 3 PA~ &(END 2 6 Unit 2 SEP 2 2 1993 4 2.X~~

1. The explosive aonltoring 1. Each of the explosive inst~ents sted in Cas aonitoring Table 3.2. Shall be OPERAS instnuaents shall be

.vith the pplicabllity as deaonstrated OPEMKZ ahovn Tables 3.2.K/4.2.K. by performance of test Al setpoints vill be in accordallce vith se to ensure that the 11aits Table 4.2.K. Syeciflcation 3.8.5.9 are not exceeded.

2. The actlcm repaired vh the member of OPERAS's ls less than the Channels 0 reqaireaent ls specified the notes for Table 3.2.I. ert beat efforts to etnrn the lnstnaea to OPEMKZ statLLS vithln days and, lf ansnc fnl, prepare and ta special report to the comisslcm pursuant to Syeclflcaticm 6.9.1 4 to explain vs the ~

~~inoyerability vas not corrected ln a tiaely er.~~

~~3. (Deleted)~~
~~4. (Deleted)~~

~~S. The provlslo. of Specificatio:. 1.0.C are not ayylicable. blÃ 3.2/4.2-6 AMENOMENTgP p i6 Unit 2 PAGE TABLE 3.2.K Hlnleam Channels/~~

~~l. (Deleted)~~
~~2. (Deleted)~~
~~3. (Del ed)~~
~~4. (Deleted~~
~~5. OFF GAS HY IHALYZER (H2A, H2b)~~
~~6. (Deleted)~~

~~A n P O SEP2 p~~~

~~(Delctcd)~~

~~~+(Deleted)~~

~~+*During main condenser offga treatment system operation (Deleted) hQLEKR (Deleted)~~

MXZQ~ ~ ~ ~ i ~ (Dcletcd) (Delete Mi thc number of channels OPE LE less than re ired by the Niniimm Channels OPEMLE requirement, operation of mai condenser offgas treatment system may continue provided t a temporary. onitor is installed or grab samples are collected at lc t once pcr 4 h s and analyzed within the follovtng 4 hours. EH'~ (Deleted) BfÃ 3.2/4.2-39 Uait 2 C W Chinnel functional ~rank Q19mthn

~~l. (Deleted)~~
~~2. (Delet~~
~~3. (Deleted)~~
~~4. (Deleted)~~
~~5. OFF GAS HYOROGEN ANAL g(3)~~

~~(H2A, H2b)~~

~~6. (Deleted)~~

~~M b4 CD SEP 22 1993 (l) (Deleted) (2) (Deleted) (3) The charm calibration shall include the use of standard gas samples containi a nominal:~~

~~a. ro volume percent hydrog (compressed air) b One volume percent hgd gen, balance nitrogen.~~

(4) (Deleted) ) (Deleted) (6) (Deleted) ~gal BFH 3.2/4.2-63 AHENDMEgr No. 2 16 Unit 2 ~aaH'A! 4(hP;4'cwsweso;ex..sw> P eve~~ ~ ~ ~rtetial%a>>aittetL<iih84ili%~J1>:<;%0 ~ i ~ ~ \l I ~'I ~ I IV%~ tI '\ ~ l% ~ a ~ II II ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ l~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ I ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 'I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ II ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ , ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~~~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~: ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~ I~ ~ I ~ ~ ~ ~ ~ ~ II ~ ~ ~ ~ ~ ~ ~ ~ I~ ~ ~ A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test program, dated September 1995". The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 49.6 psig. The maximum allowable Primary containment leakage rate, La, at Pa, shall be 2% of primary containment air wei ht er da Leakage Rate acceptance criteria are:

a. Primary Containment leakage rate acceptance criterion is s 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are g 0.60 La for the Type B and Type C tests and S 0.75 La for Type A tests;
~~b. Air lock testing acceptance criteria are:~~

(1) Overall air lock leakage rate is s 0.05 La when tested at p Pa, (2) Air lock door seals leakage rate is s 0.02 La when the overall air lock is pressurized to p 2.5 psig for at least 15 minutes ~ BFB 6.0-12 Unit 2 The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program. SPeeilicc4 S. S'NSERT SPECIFICATION 5.5.2 gl INSERT SPECIFICATION 5.5.5 Z INSERT SPECIFICATION 5.5.10 g3 SPECIFICATION 5. 5. 11 'NSERT 0 7S 37K Spz Ce~ko~ 5.C

6. 8. 5 PROKVNS Sec Z P.4'.~how 4~ ears 4 ~ DFN ised 5.S Postaccident sampling activities will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. These activities shall include the following:

(i) Training of personnel, (ii) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis s& pi de~g re pape') /he following reports shall be submitted n acct<<m w44 I~ CFA.Zo.g. BFN 6.0-13 Unit 2 r 0 7 597z ~pcc<ficggkiy ~ on A~ (csr~k os ~~) S;C.I ce ~,A,~ R~4.<<~ ~x ~~~i <<L ~ ulatxo on an ual b is of e n r of atio , utility d othe personn (inc ding c tract s) 'fo whom onitori was 'r ired r ceivin annual eep d e equi alent osure gr ter t 100 mr and t eir a ociate man r expos e ccordin to wor and go funct'ons, g., actor o rations l4~4t VsN- K2. and s eill e, ins ice 'ecti, rou axe main enance, sp ial ma'enanc (desc ma enance, waste rocessi g, and efueli The se a ignment o var us duty unctio may be esti t bas d on m asurcmen 9 elf re g do meter, TLD) o film ba . S 1 expos es tot ing 0% of e in 'dual to al dose need not acco ted fo . In the ega , at lea 80t o the tota dee se equi alent osure eceived om exte 1 sour s s be assi ed to speci c major rk func ons. W epos- s4 sue, ~'F-W; ~ e c 4 'iso ~ . Any mauxsteam re x ve chat opens ~response to t ching its'etp or due to opera action to control reactor pressur shall be rted. Pu4

~~5.4. ~ ~ I PA single submittal may be made for a multiple unit station.~~

T s u ati n ppl men s t e re ire ents of .220 of 0 CFR~ Pa 20 BFN 6.0-14 Unit 2 PAGE~OF HPR-22-97 TUE 10:Ob r rI Uv 7 571-isz prism A'abulation on an annual basis ofthe number of station, utility, and other personnel (including contractors), for whom monitoring was performed, receiving an annual deep dose equivalent > 100 mrems and the associated collective deep dose equivalent (reported in person - rem) according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [describe maintenance], waste processing, and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, tbermoluminescence dosimeter (TLD), electronic dosimeter, or film badge measurements. Small exposures totaling < 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total deep dose equivalent received from external sources should be assigned to speci6c major work &actions. The report covering the previous calendar year shall be submitted by April 30 of each year. V5 Z7W CCiQksen 5: C TZ Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. N c ear Regula Commission, Document ntrol Desk, Was 'on, D.C. 20555 w c fice to be submitted no later than the fifteenth of each month following the calendar month covered by the report. A narrati summary of o erato expertvutoe shall b~ubtdtted tat the above so edule.

~~6. 9. l. 4 REPORTABLE EVENTS Rep table events, including correct've actions and meas es to prevent -occurrence hall be reported the NRC in accordance with Section 50e73 to 10 CFR 50.~~

S'.C.2. ZRO The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall po4 be submitted y f each year. A single submittal ma KC,.2, ~ Vo be made for a multi-unit station. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the ob)ectives outlined in (1) the ODCM and (2) Sections ZV.B.2, ZV.B.3, and ZV.C of Appendix Z to 10 CFR Part 50. BFN 6. 0-15 Unit 2 rs '87Z. cciltchlirn 5 QP,l terc+p4 ar ~~>M) 6.9.1.6 SOURCE TESTS Result f required lea tests per ormed sources x the tests reveal the esence of 0.00 icrocurie. or. more of removable contamination S'.CS CORE OPERATING LIMITS REPORT (COUR

~~a. Core operating limits shall be established prior to each rctead opoeeeeeg cycle, or prior to any remaining portion of an cycle, for the following:~~

(1) The APLHGR for Specification ~~3.2.I. (2) The LHGR for Specification ~~3.23> g.2. 2 j (4) e APRM Flow Bi ed Rod Block ip Setting fo p,g Spec ication 2.1.A..c, Table 3.2. and Specifi tion 3.5.L ) The RBM Upscale (Flow Bias) Trip Setting and clipped value for this setting for Table 4 ~ 73.2..)-t 5 cificgkie~ 3.3.a(

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (latest approved version) .

BFN 6 ~ 0-16 Unit 2 PAGE~20F 0 0, c ~ The core operating limits shall be determined such that, all applicable limits (e.g., fuel thermal-mechanical limits, core thexmal-hydraulic limits, (ECCS) limits, nuclear limits 5DW such as transient analysis limits, and accident analysis limits) of the safety analysis are met. ('~l'-p S>s4 6~gg cg eorc. s C.oa-R do The including any midcycle revisions or supplements, shall bc provided upon issuance for each reload cycle to the NRC. ~~5.4 5 T8~%9cKa ~ p~ TO i~ 4ceordw~ ~.& IeWR 50.34'he %assai Radioacti Effluent Release Report covering the opexation of the unit opea'a+ion shall be submitted report shall include summaries of the quantities of radioactive liquid and gaseous cffluents and solid waste released from the unit. A single submittal may be made for multi-unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from eac it Thc material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section ZU.B.1 of Appendix I to 10 CFR Part 50. 6.9.2 SPECIAL REPORTS Repo on the fol ing areas shall be submitted in iting to the Direct of Regional Office, Division of Reactor Pro'ect BFN 6. 0-17 Qnit 2 PACi1~~QF~~gO/. ~

~~1. Fats Usage . 10. 1. q Annual Ope ing Report~~
~~2. Relic alve Tailpipe Within 30 days a er inoper-abi 'ty of thermo uple and acoustl.~~

~~monitor on one valve.~~

~~3. Sei ic Znstrumentatio 3.2.J.3 W'thin 10 days Znoper 'ty afte 30 day f inoperability.~~
~~4. teorological Monitoring 3 .2.Z.2 Within 10 days after 7 da of Inoperability inoperability.~~

Data s retrieved from all seismic instruments actuate ing a seismic event analyzed to determine e ma 'tude of the vibratory ound motion. A Special eport shall b submitted within 10 day after the event describing the magnitu frequency spectrum, and resultant effect upon plant features important to safety.

7. Diesel Generator Reliability Improvement Progr Report shall be itted within 30 s of meeting failure c 'teria in Table 9.A. As a minimum, e Reliability Improvement Program re ort for NRC audit shall include:

~~BFH 6.0-18 Unit 2 7 SZ7Z spQc~Ci(~kq~ 5 Q~~

~~a. A summary of all tests (valid and invalid) that occurred within the time pe iod over which the last 20/100 valid tests were perfo ed.~~
~~b. Analysis of ilures and determination of root causes of L,A5 failures.~~

C. Evalua ion of each of the recommendations of NUREG/CR-0660 "Enhancement of Onsite Eme ency Diesel Generator R iability in Operating Reacto s," with respect to eir application to the pl d Identification of all ac ons taken or to be taken to (1) Correct the root c es of failures defined in b above and (2) Achiev a general improvement of diesel generator reliabi ty. A supplementa report shall be prepared for an NRC audit within 30 s after each subsequent failure during a valid de d, for so long as the affected dies generato unit continues to violate the cri ria (3/20 or 6/1 ) for the reliability improvement rogram remedial action. The supplemental repo need only update the failure/demand history for he affected diesel generator unit since the las report for that diesel generator. The supplemen l report shall also present an analysis of the fai e(s) with'a root cause deterammtion, if possible, d shall delineate any further procedural, hardw e or operational changes to be incorporated into th site diesel generator improvement program and the schedule for implementation of those changes. BFH 6.0-19 Vnit 2 Ms 372 Speciale A~ Cexcep> 4s ~~ M)

~~g. I~~
5. 6. g~ High-Range Primary 3.2.F Withi days P d oker Containment Radiation after 7 days of PPM Monitors and Recorders inoperability.

I fpW' p fPbsÃc~++ Wid -Range Gaseous Event 3.2.F Within 7 days Radiati Monitor and afte days o Recorder inoperabili Sac. ~~)$ li~4o~ 4~ e4+~ 81M Is~ 5~

~~6. 12 Changes to the ODCM:~~
~~1. Shall be documented and records shall be kept in a manner convenient for review. This documentation shall contain:~~
~~a. Sufficient'information to support the change together with the appropriate analyses or evaluations justifying the change.~~
b. A deteanination that the change will maintain the level of radioactive effluent control pursuant to 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR. Part 50 and not adversely impact the accuracy or reliabili.ty of effluent, dose, or setpoint calculations.
~~2. Shall become effective after review and acceptance by the process described in TVA-NQA-PLN89-A.~~

~~BFN C.0-20 Unit 2 e 0 7S37Z S cr t'c<4 cm <7 .6.8.1 (Cont'd)~~

~~6. 8.1.2 (Deleted)~~

QPl (rtccpk as ~M) 6.8.1.3 (Deleted) sec <~44~~<;~ 4i o DRZLLS Q~ @hl Isr5 S:g .8.2 Drills on actions to be taken under emergency conditions involving release of radioactivity are specified in the Radiological Emergency Plan and shall be conducted annually. Annual drills shall also be conducted on the actions to be take following failures of safety-related systems or components. ZATZON CONTROL PROCED 6.8.3 diation Control Procedures shall be maintained and made ava'ble to all station pe sonnel. These procedures shall contai radiation dose limits d shall be consiste t with the requiremen of 10 CFR 20. This r ation protection ogram shall be organ ed to meet the requirements of 10 CFR 20 excep for the "control device" or "alarm signal" required by ~.~ "~a4

~~20. 1601 (a) 8~ its Arec~~

~We. We. ~m ~~red'~ &e. 5'.gf ~.-3-.% Each high radiation area as defined in 10 CFR 20 shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiological Work Permit. Zndividuals qualified in radiation a JaSag protection procedures (e.g., a radiological control technician) or personnel escorted by such individuals, shall be exempt from iw 4~ RWP requirements during the performance of their assigned duties 4S Waa~ in high radiation areas at 30 centimeters, provided they otherwise comply with approved radiation protection procedures e4 ~~ 9~4 W i'~ la cf=w zd P"~S~P+ 2o IC4( ( ), pg f // ~~)~ Sn la CFR ze,jg,s~~ BFN > e~ Unit 2 QP eg.L for entry into high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

~~a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.~~
b. .A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

C ~ An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation s urveillance at the frequency specified in the Radiological Work Permit. S.7. X Zn \ br ~~ay ~~ addition, areas that are accessible to personnel and that have radiation levels greater than 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months as measured at 30 centimeters, but, less than 500 rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 1 meter from the radiation source or from the surface which the radiation penetrates shall be provided with locked doors to prevent unauthorized entry. The keys shall be under the administrative SI gq NanA er control of the duty Radiological Control Manager or their respective designees. Doors shall remain locked except during periods of access by personnel under an approved RWP which specifies the dose rates in the immediate work areas and maximum allowable stay time for individuals in PAGE~OF~ 6 + BFN 6.0-8 Unit 2 p't (Cont') area. Zn lieu of the stay time requirement of the RWP. oz'emote (such as closed circuit TV cameras) continuous surveillance may be made by individuals qualified in radiation protection pzocedures to provide positive exposure control over t the activities being performed in the area. $ .~.> Zndividual<radiation areas that are accessible to personnel, have radiation levels greater than 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months as measured at 30 centimeters, but less than 500 rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 1 meter from the radiation source, are located within large areas where no enclosure exists for the purpose of locking and where no aS ~6~ m ~ enclosure can be reasonably constructed around the individual ~gohse area, shall be barricaded, conspicuously posted, and a flashing ~~ ~e light shall be activated as a warning device whenever the dose ~4k~ S~~-Re. zate in. the area exceeds or will shortly exceed 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . 6.8.4 RADZOACTZVE EPPLUENT CONTROLS/RADZOLOGZCAL ENVZRONMENTAL MONZTORZNQ PRO%VMS The following programs shall be established, implemented, and maintained. .8.4.1 RADZOACTZVE EFFLUENT CONTROLS PROGRAM A program shall be provided confozming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS Ot THE PUBLZC from radioactive effluents as low as reasonably achievable. The progzam (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements: 5og JsegtCichg~ r kr gpji/ Is7$ g, 5'FN 6.0-9 Unit 2 V 0 UNIT 3 CURRENT TECHNICAL SPECIFICATION MARKUP 1 S 3'72. ~ Ccsfi on BFN TECHNZCAL SPECZFZCATZONS 8I 5~0 ADMZNZSTRATZVE CONTROLS (.~~+ Rbmavgccf) si~ vice'res gt4$ slrrll Je rc~rgib]e < CVC<a aC ~ FAC','FC, u/Ai'IC +be W.1..1 The Plant Manager shall be responsible for overall unit operation aacL shall delegate in writing the succession to this XivSe'Rt 4.0-I+ responsibility during his absence. Pt';0 I/ice ~i@'nf ~ ~ A.1.2 The Qg~ M er f ltd P1ynvl <r r (~ urging 'bsence from the 5 ontrol oom shall be respon ible for the Control Room command function. Any g~g 5'h;4 j P<~)ey ~rFg,7 6.2 b.o- (8 6.2.1 An onsite and offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant. a Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all h operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report ~A-NPOD89-A) . 1g'on Ar (Q~~ BFN 6.0-1 "'~~ lsrs s,~ Unit 3 INSERT 6.0-1A The Plant Manager or his designee shall approve, prior .to implementation, each proposed test, experiment or modification to systems or equipment that affect nucl r safety. INSERT 6.0-1B while the unit is in MODE 1, 2, or 3, an individual with an active Senior Reactor Operator (SRO)'license shall be designated to assume the con'trol room command function. During any absence of the Shift Manager from the control room while the unit is in MODE 4 or 5, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function. PAGE~OF 5 T5 pl gcwc~pr es ~ )grC 37'('~cia;cahen S 2. BFN TECHNZCAL SPECZFZCATZONS ~0 ADMZNZSTRATZVE CONTROLS 6.1 sec a'~shA'~h~ W 8nl js7s, 4r CAu ~ 6'.g.l The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. 6.1.2 The Shift Operations Supervisor (or during his absence from the Control Room, a designated individual) shall be responsible for he Control Room command function. 5 5'2.1 g~ig<Rlclg nsite and offsite organization shall established for unit operation and corporate managemen . The onsite and offsite organizatio shall include the positions for activities affecting&aC safety of the nuclear power plant.

a. Linea af authority, responsibility, and coneunication shall be established and defined from the highest management level z intermediate leve and 4aekedHg all operating organization pasitions. These relationships shall be documented and u dated, as appropriate, 4A organizationgj) charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A)

T's'7M J.2. (Cont'd) 5 1 4l<A 6x~uc. Viic /~~ ~p ~~ cP c~~a Chief Nuclear Officer, shall I have corporate responsibility for overall plant nuclear safety ' 4q 'hall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant Cn$ urc The e o eration 4~42geg f fbi /kgb)+ Plant Manager shall be responsible for overall and shall have control over those onsite ~ necessary for safe operation and maintenance of the plan > ntdtoJ tCgg Cc~olC>Or P~if I~

d. The individuals who train the operating staf carry out quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

~~sJ" ~~ Qn,' Staf f TllC Kpt t >AS PpgaSle~~ SAg)) it)C,lid~'< ~~~~~

~~a. hift ing r iremegs, shall a$ a mani, be s d cribe 'n Table .2.A Rd below.~~

~~TASer R (e-gg~~

~~b. A license senior eactor ope tor shall presenb at the si at all imes wh there '~~

in the actor 4 /cosh ~no (go> Prison t g licensedgeactor gper js r shall be in the control room whe uel in the reactor. BFN 6.0-2 PAGE 5 GF~5 Unit 3 A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in MODES 1, 2, or 3. kl (<<ac~pt es mci~q) TS zqg +Ccohocago~ 5, sX.. (Cont') Q (.(m Yanu&) o 1 cen d r ac r op rat rs s ll e i the co tro ro d ing ny old tar ps, hil shu6ti g d ~./ he act r, a d d rin rec e rom it tr'. In ad Clan enio perato cens shall be in the ontro room CsRc ) s44i r~ JC ~ egg/'S;n llloPE 2, or 3 Ot'tOSC one A 4gtoloeicel Cntsr]S s J. when Were~ fuel in the reactor. on 5';~ IS 3 Al ~sage k.~k ~ CTS G 8.~.l.b ~ t>> b.e-6 5.22,f 'fAsggp g,o 5;22.g Zan O,O-Z C ~ ~ P&:Ron Wy ~ g~ no+ 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided immediate action is taken to fill the required position. g fpp~ t pfov pr <<n CW+ICA PbSCnl.C, BFN Unit 3 6.0-3 t"Al) F~Qp~p Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, radiological controls technicians, auxiliary operators, and key maintenance personnel). Adequate shift coverage shall be maintained without routine heavy use of overtime. The'bjective shall be to have operating personnel work an 8, 10, or 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months day, nominal 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

~~l. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time;~~
2. An individual should not be permitted to work more than 16 hours in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;
~~3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time;~~
~~4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.~~

Any deviation from the above guidelines shall be authorized in advance by the Plant Hanager or his designee, in accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or his designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized. PAGE OF I INSER 6.0-3B

~~f. The Operations Superintendent shall hold a current SRO license on a Browns Ferry unit.~~

~~INSERT 6.0-3C~~

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift'anager in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

~++:Gca+'oq 5: z J~N 80 588 Eud~u 1 Senior 0 ratora 1 1 1 SRO Senior perator 0 1 2 2 Lic ed Operators 3 3 3 RO or SRO Additional Licensed Operatorsc 0 1 2 2 RO or S 0 >..Z.a hesistant Unit Oyerators (AUO) 3 A'9 S Ione i S.2.z. Shift Technical Ldvisor (STA) 0 1 1 1 Health Physics Technician 1 1 1 1 /otal Ml a. A senio operator wi 1 be assign esponsibility overall y ' o P eration all tiaes ere is fuel unit. R go .so w)U)Ci3 JWAi&c~tioss S.g,.'2,~ l.3 ~ xrtes ahi t coiposition aalu be ~ less than the aiatam requireaents o for a period of tiae not to exceed two hours in order to accoaaolate e crew unexpected absence of on-duty shift crew aeabers provided hmediate action is tLken to restore the shift crew coEposition to within the Iln1eaa requireaents of T ble 6.2.A. Thi yrovision do not pe any shift crpw position to e mmsnned uycm ft change e to sn oming shift ch'an being 1 e or 'absent.

~~c. One o the klditional Lic ed Operators aust be. as gned to each ontrol r~ th'n operating to~~
~~d. Th natcher of require licensed personnel, wh the operating ts are c trolled troa a c control rooa, are tw senior operators and four I~~

0 erators ~ BFK ill . NEemrr~, gag Unit 3 PAGE~&pF ~ f, Qg ~ 6.8.1 PROCEDURES 6.8.1.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

~~a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978.~~

~~S2,g,a P Limitations on the amount of overtime worked by individuals performing sa fet y -related functions in accordance with NRC Policy statement on working hours (Generic Letter No. 82-12).~~

~~c. Surveillance and test activities of safety-related equipment.~~
~~d. (Deleted)~~
~~e. (Deleted)~~
~~f. Fire Protection Program implementation.~~
~~g. (Deleted)~~
~~h. (Deleted)~~
~~i. Offsite Dose Calculation Manual.~~
~~j. Administrative procedures which control technical and cross-disciplinary review.~~

a4P:rghe 4 Boo Is~~ > < ., r"sUI-~~~/ BFN 6.0-6 Unit 3 /fl (enkicc Page) Each member of the unit's staff shall meet or exceed the minimum qualifications for comparable positions as specified in the TVA Nuclear Quality Assurance Plan (TVA-NQA-PLN89-A). SCe 0'onf LftlCItncflOn *r C/glgngcg,g Istic q.~ 6.7 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The President, TVA Nuclear and Chief Nuclear Officer, and the NSRB shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PORC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence.

~~The Safety Limit Violation Report shall be submitted to F the Commission, the NSRB, and the President, TVA Nuclear and Chief Nuclear Officer, within 14 days of the f' violation.~~

~~d. Critical operation of the unit shall not be resumed until authorized by the Commission.~~

BFN 6.0-5 PAGE~2gF $ Unit 3 T5 z 7g s.q ~A> (cxccPP 4s Nqrird) PROCZDmum S'yo ( ~written procedures shall be established,.implemented and maintained covering the activities '

~~a. The~~

~~)/~and applicable procedures reconaended Regulatory Guide 1.33, Revision 2, February 1978. endix A. ~~~

b. Limitations on the amount of overtime worked by individuals performing safety-.related functions in accordance With NRC Policy statement on wrking hours (Generic Letter No. 82-12) . ~c+ >4>+~ichton Ar Qangog For SC'W f~g 6.~
~~c. Surveil ce and test ctivities of afety-rela ed e ipment.~~
~~d. (Deleted)~~
~~e. (Deleted) 5.9.l.g g Pire Protac:@ion paograa ieplsaeataeioa~~
~~g. (Deleted)~~
~~h. (Deleted)~~

~~Of fait se Calculaaaon Man~. Adahnist cross-discip 'y tive procedure review. which control te ical and BFH 6. 0-6 Vnit 3~~

~~b. The emergency operating instructions required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; c.~ guality~~

~~~ assurance for effluent and environmental monitoring;~~

~~e. All programs specified in Specification 5.5.~~

P,l e e 6.8.2 rills on action to.be taken under e rgency conditions in olving"release radioactivity are s ecified in the Radi logical Emergen Plan and shall be onducted annually. Annual ills shall als be conducted on t actions to be taken following failures of safety-related systems or RADIATION CONTROL PROCEDURES 6.8.3 Radiation Control Procedures shall be maintained and made available to all station personnel. These procedures shall contain radiation dose limits and shall be consistent with the requirements of 10 CFR 20. This radiation protection program shall be organized to meet the requirements of 10 CFR 20 except for the "control device" or "alarm signal" required by

~~20. 1601 (a) .~~

6.8.3.1 Each high radiation area as defined in 10 CFR 20 shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiological Work Permit. Individuals qualified in radiation protection procedures (e.g., a radiological control technician) or personnel escorted by such individuals, shall be exempt from RWP requirements during the performance of their assigned duties in high radiation areas with radiation dose rates equal to or less than 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 30 centimeters. / provided they otherwise comply with approved radiation ~<~ S~~C;~~. g..o < BFN 4 ISYS g 7 6 0 P i4(,gQ QP Unit 3 T'S > 7X $ /C C ice Cat~ S,Q I i<acefres ptnfttf)

~~8. (Deleted)~~
~~9. High Range Primary 3.2.F Within 7 days Containment Radiation after 7 days of Monitors and Recorders inoperability.~~
~~10. Wide Range Gaseous 3.2.F Within 7 days Effluent Radiation after 7 days of Monitor and Recorder inoperability.~~

Ce pgSHQratio~ Ar gp~~~ A Bg~ )STS S-.k c Ai'ktokcd Panges to the ODCM: '~<%~7 f~~< f, cps I,O-<o CrS t,o,~g Shall be documented and records shall be ~~eQ I I 4) )officiant information to support the change together with the appropriate analyses or evaluations justifying the chang end (s) determination that the change will maintain the.'eve f radioactive effluent control pursuant'o 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and ppendix CFR Dent 50 and not adversely impact the accuracy or reliability of ef fluent, dose, or setpoint calculation P Shall become effective after review and acceptance by the process described in TVA-NQA-PLN89-si BFN 6.0-20 Unit 3 ~, TS g 7g. Speci( +6'on S.s S.S.l. c. p Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM,was made. Each change shall ~ be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented. BFR 6.0-21 TS Sea. S u)%$ K (s Cc( JOIE fglf CJsi5gg +Pc g (Cont'd) +r BRA fsTS l.o W. - A functional test is the manual operation or initiation of a system, subsystem, or components to verify that it functions within design tolerances (e.g., the manual start of a core spray pump to verify that it runs and that it pumps the required volume of water) . X. ~~~ reactor mode The reactor is in a shutdown condition when the switch is in the shutdown mode position and no core alterations are being performed. - An engineered safeguard is a safety system the actions of which are essential to a safety action required in response to accidents' A reportable event shall be any of those conditions specified in section 50.73 to'0 CFR Part 50.

g. 4e hall contain the 5i5e( met o o ogy an parameters used in t e ca culation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitorin a (rip j/et tints, and in the eonduet of v'ment

~ adio ogical monitoring Qogra The OD the gadioaotive Pf fluent gontrols and Padiologiea P" vironmental )monitoring 1 also con should be included in tLnd ~descriptions of the information that the Annual Radiological Environmental Operating and kaseak Radioactive Effluent Release~jfeports required by Specification+ 5g,g~~ and 6v&e~ 5pc~.,gg ~~ g g g CC. - The controlled process of discharging air or gas from the primary containment to maintain temperature, pressure, humidity, concentration, or other operating condition in such a manner that replacement air or gas is required to purify the containment. DD. EE. FF. ga~lg - The controlled process of 4ischarging air or gas from the primary containment to maintain temperature, pressure, humidity, concentration, or other operating condition in such a manner that replacement air or gas is not provided or required. Vent, used in s stem names does not i 1 a ventin rocess. sec +~he ~on b~ I~T~ l.o g g4 cs BFN 1.0-10 Unit 3 A/ (m r+ ger:igi caruso~ S. 5 MAY 19 1989 S. s.g lR. urve llsnc Requir ts for andM illance R or Laservice Teat~ of LSD Code Class 1, 2> and 3.coaycments %< program ShaM inc4dc 44Mo~;a

1. Inservice testee f ASS Class 1> 2> and 3 and valves bc kfox%et accordanc If%th Secti ZZ f the Boiler Pr Vessel e and app cabl Add aa re red by 10 vhere apecifi vritten r ief has 50, Se icm SO.SS ranted i), ept asian t to CPR 50 Section 6 i Ti-S+W Ar'CanC i CS cified in Secticm XK~ the JAR Boiler and Pressure Vessel Code and applicable JSd for the ce rsd iy Boiler and sure Vessel end applicabl applicab e aa foDoccs in these technical cificati ice foJ~g AS'oiler and Pressure Vessel Required frequcmcies qS Code and applicable Addenda for perforate inservice tezalnology for inscrvice Wetly Ncmth1y it it least least once once per per 7 days 31 days Quarterly or every 3 mmtha it least once per

~~ 92 days SaaLannnally or every 6 aaths Scary 9 oaths it least it least once cmce per per 18'ays 276 days I Tearly or annually At least cmce per 366 days ~ Blcnnially or ~y 2 )canA }CO S+ Inc.e POl 73 L &o 'ho provisions of are applicable to the above required requencies for perforalne inservice testhg activities. Sod. 2

~~4. P ozaaac o ~ e ce test ac't tgca be additi to other spec ed survei ance re r~~

~ Rothfag in the LQR Boiler and Pressure Vessel Code ahall construed to supersede the requireaeata of any technical bc. <<pecificaticm. C. inservice inspecti prograa for pip identif in IRC Cene ic Letter 8$ -01 1 be perforaet acco vith the s ff positioxLa on s adulate sothoday raouncli aaaple icm.included in this generic le er. %r ihau'bio~ pr' l2 3.b.3 eu. <<SC<<CM fCSfi~ CCC.t'v~'4 Cr amdt BE5 jlSERSENf NL 13 7 Unit 3 FAG E~Gi ~J R~ ~<<>~st 8'p'Al Fi~ I57g g,g g r Ch TS P 72. >f'CCs 4t cc,~ 6.8.3 (Cont'd) individuals in that area. Zn lieu of the stay time requirement of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by individuals qualified in radiation protection procedures to provide positive exposure control over the'activities being performed in the area. 6.8.3:3 Zndividual radiation areas that are accessible to personnel, have radiation levels greater than 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months as measured at 30 centimeters, but less than 500 rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 1 meter from the radiation source, are located within large areas where no enclosure exists for the purpose of locking and where no enclosure can be reasonably constructed around the individual area, shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device whenever the dose rate in the area exceeds or will shortly exceed 1 rem in 1 I (@metre ss maie~g ) OLOG 5.S P~r~ a ~ Nenu<ls The following programs shall be established, implemented, and maintained. 5.5.I ~~ zing~ co~/ v~ ~$ program conform~ S ~ 10 CFR 50.36a the control of radioactive effluents and for maintaining the for doses to MEMBERS OF THE PUBLZC from radioactive effluents as low as reasonably achievable. The program ~ shall be BFN 6. 0-9 LA+~ I . S( Unit 3 T'S Z7P SS;9 AI (~gp

~ J) SfcciC 'ca hon (cont') ~ procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements: ncHoncl Ca~b~ Li+0 a.. Limitations on the of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the OD
b. Limitations on the concentrations of radioactive material released in liquid effluents t.o conforming to 10 times the concentration values stated in 10 CFR 20 - . 4 Appendix B T e 2 Column 2+e
~~c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM.~~
d. Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED MQM conforming ~ Appendix IJ to 10 CFR Part 50~
e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current year in accordance with the methodology and parameters in the ODCM at least every 31 days ~

~~Ll 4 ~-~e ~b:i,'eq~~

f. Limitations on the d use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce BFN 6 '-10 pAGp Pl OF~~

Unit 3 Q age 5 5tSs Q (Cont') p)( (e~P dy Py) CCige Cggoh a '5 releases of ra~oactivity the projected doses in a ~~ whe 31)(da 'periodpould exceed of the guidelines for the annual dose or dose commitment confozmin ~ Appen xx I to 10 CFR Part 50

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SXTE BOUNDARY shall be limited to the following:

CS

~~1. For noble gas a dose rate of 500 mrem/yr to the I~~

~~total body and 3000 mrem/yr to the skin, and~~

~~2. For odine-131, odine-133, tritium, and for all radionuclides in particulate form with half-lives 8 days:~~

~~~~4 1500 mrem/yr to any organ.~~

h. LimitationsM ~btt the annual and quarterly air doses resulting from noble gases released in gaseous effluents from eath unit to areas beyond the Pgg nglpg6gg conforming to Appendix I> ee 10 CFR Part 50>
i. Limitations on the annual and quarterly doses to a PP @ gj@g from jf di oe 1n31, g-odine-133, tritium, and all radionuclides in particulate form with half-lives each unit to areas beyond the sPQ B~g conforming to Appendix I to 10 CFR Part 50~ ~

~~BFN 6 ~ O-ll Unit 3 (Cont'd)~~

~~j. Limitations on the annual dose or dose commitment to any Pg Qg TPg PP@p'oe to releases of radioaeeivity aad eo radiation from uranium fuel cycle source conforming to 40 CFR Part 190.~~

.8.4.2 (Deleted) See r~reW:sad d dao saayo) Pd'.s<5 S'.C.l< 6.8.4.3 PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the containment as requ'ired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test program, dated September 1995". The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 49.6 psig. The maximum allowable primary containment leakage rate, La, at Pa, shall be 2t of primary containment air weight per day. Leakage Rate acceptance criteria are:

a. Primary Containment leakage rate acceptance criterion is S 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are g 0.60 La for the Type B and Type C tests and g 0.75 La for Type A tests;
~~b. Air lock testing acceptance criteria are:~~

0 (1) Overall air lock leakage rate is s 0. 05 at 2. Pa, La when tested BFN 6. 0-12 Unit 3 5'e5'eQ (Cont ') A'( (ennre +e)

~~2) Air lock door seals leakage rate is s, 0.02 La when the overall air lock is pressurized to ~ 2.5 psig for at least 15 minutes SCC Jog Pjl(ygp den 4r C(i~)~~

$ ~ gF~ iS m S S.I2. ThiS fio9tam /~A~ cootrols Nerf ensure the capability to obtain 8R$ cS 'and analyze reactor coolant, radioactive 4ec&aoe>and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. Therm fngga~ '. shall include the following: lj ~ Traininp ef p r creen@a! Procedures for sampling and analysi; crnP C ~i Provisions for maintenance of sampling and analysis. eau>e~cat, 6.9 ROUTINE REPORTS 6.9.1 Zn addition to the applicable reporting requirements of Title 10,, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the Regional Office of NRC, unless otherwise noted.

~~6. 9. 1. 1 (Deleted)~~

~~Sec S'iS hV'oih'ron go~ g+y cs Ar bAJ 1sT';g BFN 6.0-13 Unit 3 3.7 B. lg~~

~~l. Ercept as specified in Specification 3.7.B.3 belov, e inc lov co it ns all three trains of the standby gas treatment syst shall be OPERhBLE at all s~~

~~reste ~ 1 b d ig ~~ tr ~~~ ed. drop'cross the combined HEPT times vhen secondary containment integrity is filters, and charcoal ~ required. adsorber banks is less P than 1 nches of vater at a flov of 9000 cfm ~ xox). l7~~ ~~~~ ~ee kr 3u564'cg+< g Qy~ The iaXet heaters on bPN [5T5 p Q y ~ 8.8. 7. a each circuit are tested in accordance vith hHSI 5510-1975, and are capable of an output of at least 40 )N.~ c~ Sir istri tion s oza thin 2 a oss A fil ers cha coal ad orb rs. NENDMENT Rm. TC g BHf 3.7/4.7-13 anent 3 ~A<<>> oF~/k~.V 3~ 4,7 ~ ~W The results of the in-place ~M The tests and sample analysis of cold and haloaenated ~ SOP hpdrocarbozL tests at g Specification 3.7.5.2 design flov on HEPA filters shall b>> performed <5" 4 and charcoal adsorber banks shall shov g99X DOP removal mpc4a-ey-once every S %'l. b and g99X haloaenated bydrocarbon removal vhen 5.g 7 tested in accordance vith or after every kKSI 5510-1975. 720 hcnars of stem o tion and follovtng $ .51. sigRificant paQltixlg fire, or chemical release in any ventilation cone cemnanicatiag vith the <S '7 C ~ The results of laboratory carbon saaple shag. shw i radioactive / system o ColC DOt Castings be perforned after shall each cosplete or partial aethyl iodide resoval vhen 5 ~g ~ replacescnt of the HEPA tested in accordance vith filter bank or after any L5ZN D3I03. structural Iiaintenance on the system hoash~. . Systea shaD be ahovn to /o ogenated hydrocarbon operate vithin g10X design test@~ shall be flov, performed after each coaylete or partial 5;S gb replacement of the charcoal adsorber bank Thc fjhow's'ms' f 5 g g, y. g or after any structural aaintenance on the ~+ SlS.0.3 Sic +plscg+c system honsh~. 8 ~c f F1P Qcecc~ipc SiN/4+7 14 SCC S~ggfl(4,fi+n 4r QCcf g cAA+~ ~ ~Qc< Qr BRA IsT5 r.7~ EB-1 3 gg 3.7.E. c c Vc at o 4.7.E

1. Ezccpt as specified in 1. ht least once cvcry 18 months, Specification 3.7.E.3 below, thc pressure drop across the both control room emergency combined HEPT filters and pressurization systems charcoal adsorber banks shall shall be OPERhBLE at all be demonstrated to to be less times vhen any reactor than 6 inches of vater at vessel contains irradiated system design flov rate fuel. (g 10%).
2. a. The results of thc inplacc 2. a. Thc tests and sample cold DOP and halogenated analysis of Specification hydrocarbon tests at design 3.7.E.2 shall bc pcrformcd flovs on HEPT filters and charcoal adsorber banks ke- once every

< 5. 7.b shall shov g99X DOP removal 5. l 18 months, and g99X halogenated hydrocarbon removal vhen or after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of tested in accordance vith system operation and ANSI R510-1975 ~ o ov significant painting; fire, or chemical 5:5.1.c. release in any ventilation zone cosmxaicating vith the system+

b. The resul s of laboratory b. Coll DOP testing shall be carbon le analysis ahall perforaed after each shov radioactive methyl complete or partial 5;5.$ .< iodide cmval at a velocity S.S.7.~ replaceaent of the HEPh vhcn tested in accordance filter baal or after'ny vith ASTI D3803. structural maintenance on the systea housing.

~~(~ic~l t cs nute~cd) 185 3.7/4.7-19 NIlENNBffNO. I8 8 Unit 3 0 Wi+c'admen 5 S ~% 0 9 l993 3.7.E.~~

c. System flov rate shall bc p'. Halogenated hydrocarbon ahovn to be vithin glOZ testing shall be performed design flov vhen teated in after each complete or accordance vith ANSI partial replacement of the 5510-1975. 5;5. g after ~ etructural charcoal adaorbcr bank or aainten11mce on the system housing,
~~d. Each circuit shall be operated at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month.~~
3. ?rom and after the date that 3. At least once every 18 months, one of the control room autmatic initiation of the cacrency pressurization control room emergency systems 'is made or found to pressurization system shall bc bc inoperable for any reason, dcaonstrated.

~~REACTOR POMER OPERATIONS or refueling operations are pcriiaaible only durham the auccaadlag 7 days unless such circuit is sooner made OPERABLE.~~

4. If these conditions cannot be 4. Durtz~ the simulated automatic aet, reactor shutdovn shall bc actuation 'teat of this system initiated and all reactors shall be in COLD SHDTDOWK (aee Table 4.2.C), it ahall verified that the necessary bc vithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for REACTOR daspera operate aa reqai.red.

PINER OPERATIONS and refuelixe operations shall bc terainatcd vithin 2 hours. ~'caRo~ 4i Chnayrg 4i bPal tsvs z,7.3 3.7/4.7-20 MBeeee. Z 65 PAGE 2$ pp g( p, S crgmMn 5EP (crew+ 4o ~raced) 2 2 1993

~~4. (Delseed) 4. (Deleted)~~
~~5. The aaxianm activity to be 5. (Deleted) contained in oae liquid~~

.radrute tank or temporary storage tank that caa be discharged directly to the irons shall aot exceed 10 curiae excluding trftf~ aad dissolved/ entrained noble gas. tbsp s ; .g,g N.th dioactfv liquid raste 6~ ty exce fag 3.8 ~ .5 1fsii adi tive ter 1 sit t delay suspend 11 coa ed an outside ad tfons of radioac ve 1 uid adzes sto age terial t the tan aad shal e ithfa 48 ours, r uce the de rsifn to b vith teak coa eats to thin the ab e 1 lfILft. Events 1 diag to aaal fag 'this c dition t be re essa tive le of repor fn t next hnaual s c ont at Radioactive lueat Release leu oace r 7 ays Report (Sectf 5.2 of the rad ct oem). mter s be g add to ~ . l %C psouiS.onJ Q SR 3.D.2 Cn4 SR 7.o.3'e qg[I'c k ~c Q ye EpgfuVue 6w anr( shi~~c T~~ fad oacHui'by /Mini 4i:g ~A+~+~ Snivel llance O'ce~e~c >ex e. BlÃ 3.8/4.8-2 Uait 3 ~Zlt a (g'l (~cptas martcct) $ EP 2 2 i>>~

~~l. (Dcletcd) I ~ (Deleted) 2~ (Deleted) 2. (Deleted)~~
~~3. (Deleted) 3~ (Dclcted)~~
~~4. (Deleted) 4. (Deleted)~~
5. (Deleted) S $.Ba ~ The concentration of hF4rocen dovnstream of the recombiners shall be detezmined to be vithin the limits of 3.8.B.9 b7 continuously monitoring the off-gas vhencvcr the SJAE is in scrrice ng rum ts scribe in Table 3 2.K. Inst cnt survci lane rcqui ements are s eeif ed in able 4 2. ~

~~6~ (Deleted)~~

~~7. (Deleted)~~
~~8. (Deleted)~~
~~9. Whenever the STATE is in serTice, the concentration of hy4rogcn in the ofideas dovnstreaa of the recoabiners shall bc limited to g4% volume.~~

liked.- M c oneentrat on of o8 ezcc ing t limit 3.8. .9 a vc, r tore hc co ccntr tion vithin the l it v thin 4 hours. PAGE 3 ~ oF~t. 3.8/4.8-3 IMENOMEHT RO. I7 " >ft c ificofjon < 5 SEP 22 t993 3.2.K 42K 1, The exp sfve gas aonitorhe 1. Bach of the explosive inst ts listed in Table gas aonitorfnS 3.2.K be OPERA1KZ vith fnstrueents shall be the plicability as shovn in deaonatrated OPBRiBLE Tabl 3+2+K/4~2~Ko Alara by yerfozaance of tests set inta vill be set in accordance vith to ensure that the liaits f Table 4.2.K. S ecification 3.8.B.9 ar not ceded

2. e actfon reguir en the maher of 0 channels fs less than Madam Channels 0 re@of reRent is specf fied the notes fo Table 3.2.. Exert best efforts return the fnst ts to OPBRiILE tatus vi 30 dna and, i unsu essful, preyar and ta special r rt to c~ssfon suant to Specification 6 1.4 to explain very inoyerabili vas not corr ted in a t sMsoro
~~3. (Deleted)~~
~~4. (Deist )~~

~~S. The provisions of Specification 1.0.C a licable BFI 3.2/4.2W Unit 3 I TABLE 3.2.K Ninl~ h anne 1 el~~

~~1. ( 1eted)~~
~~2. (Oel ed)~~

~~(Oelete~~

~~4. (Oeleted)~~
~~5. OFF GAS HYOR AINLYZE1 (H2A, H2b) ('1)~~
~~6. (Oeleted)~~

~~~2Sm +(Deleted)~~

~~(belated)~~
~~a*Dnring gag c cnser offgas rcatment system operation dQXEKh (Deleted)~~

(Deleted) (Dclc ed) (D lcted) Vith the n er of els OPERhBLZ 1 s than required thc Channels OPERMKZ r ircacnt, peration of ma condenser offg treatment system may continue orided tha a temporary m tor is installe or grab s les ara collected at least ~ per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed vi the follovtag 4 boars. (Dclcted) BZK '3;2/4.2-38 that 3 PAG~ I ARABLE 4.2.K V n I. {Deleted)

~~2. (Deleted)~~
~~3. (Deleted)~~
~~4. {Deleted~~
~~5. OFF GA HYDROGN ANALYZER (H2A H28)~~
~~6. (Deleted)~~

c(.,&capon >. < SEP 22 1933 (l) (Delete (2) (Dele d) (3) The el calibrat on shall include the u of standard as samples co Lining a nomina ! Zero volume p rcent hydrogen (compre aed air) and,

~~b. One volume ercent hydrogen, bal 0 nitrogen.~~

(4 (Deleted) (5) (Deleted) lg/ (6) (Deleted) BPH 3.2/4.2-62 NENDMENT HO 1 7R Unit 3 5 ctefic~k>g~ S5 OCT 2 5 1993 3.9.A. 4.9.A. 3.9.A.l.c.(3) (Cont'd) 4.9.A.l.b. (Cont'd) through load sequencing, and operates for greater than or equal to five minutes while i,ts gener-ator i.s loaded with NOTES FOR (2) AND (3)- the emergency loads. If both Athens and (3) On diesel generator Trini.ty lines arc breaker trip, thc claimed as thc two loads are shed from offsite sources for the emergency busee unit 3, no credit may and the diesel output bc taken for the breaker closes on the Athens-Trinity line auto<<start signal, the tie breaker. emergeacy buses are Specifically, the energised with Athens linc supplies permanently connected unit 3 through coamon loads, the auto-station-service conaccted emergency transformer A or loads are energized cooling tower through load transformer 1, aad sequencing, and the the Trinity line diesel operates for must supply uni.t 3 greater than or equal through coaMaon etation- to five minutes while scrvice transformer B or its geaerator is cooling tower loaded wi.th the transformer Z. emergency loads.

~~c. Once a month the quantity of diesel fuel available shall be loggede~~

~~s~<;~~4 C<~ d. Each diesel generator shall be 8F~ lsrs s,.r.l inspected in accordance with instructions based on the manufacturer'e recoamcndations 24 months. P~ps 5: s.q Quarterly the quality of each S.S:9.~ diesel generator'e (3A, 3B, 3C, and 3D) seven-day fuel load I'.s~ 5.s.q. g supply shall be checked. fuel oil quality shall be within The the acceptable limits specified in Table 1 of ASSAM-D975-89. BFN 3.9/4.9-3 Unit 3 hMBBMBP Mlt. ~ ~M PAGE ~ ' Qccif ca+ a 5; q go< 5'ustiAek n g, (p~~ Ai BRv rCVS S.p Limitations on. the annual dose or dose commitment to any MENBER OP THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190 ' s.lg. ~P~g CQ$ A program ~ ~$ shall be Rpp ~$ PROGRAM established to, implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test program, dated September 1995". The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 49.6 psig. The maximum allowable primary containment leakage rate, La, at Pa, shall be 2% of primary containment air weight per da Leakage Rate acceptance criteria are:

a. Primary Containment leakage rate acceptance criterion is S 1.0 La. During the first unit startup following testing, in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the Type B and Type C tests and S 0.75 La for Type A tests;
~~b. Air lock testing acceptance criteria are:~~

(1) Overall air lock leakage rate is g 0.05 La when tested at g P<< BFN Unit 3 6.0-12 PAGE~7 O'a j (axcrPf (2) Air lock door seals leakage rate is g 0.02 La when the overall air lock is pressuiized to ~ 2.5 psig for at . least 15 minutes. ~H5Eg 7 6.o -/S+

6. 8. 5 PROGRAMS Postaccident sampling activities will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. These activities shall include the following:

(i) Training of personnel, (ii) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis. 6.9 ROUTZNB RBPORTS 6.9.1 Zn addition to the applicable reporting requirements of'itle 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the Regional Office of NRC, unless otherwise noted.

~~6. 9.1. 1 (Deleted)~~

++>~Ci~CHoe '~+ g ( ~g+ X57 S PcC BFN 6. 0-13 Unit 3 1 0 The provisions of SR 3.0.2 do not apply to the test frequencies ,specified in the Primary Containment Leakage Rate Testing Program. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program. INSERT SPECIFICATION 5.5.2 INS ERT SPEC IFI CATION 5.5. 5 ~ z. t INSERT SPECIFICATION 5.5.10 p3 INSERT SP EC IF I CATION 5.5. 11 TS 572 Spec(fi AI+joes $ , Q 6.e.i (Cont'd) (g) ~r lock dooz seals leakage race is ~ 0.02 La when the overall air lock is pressurized to g 2.5 psig for ac. t 15 minutes. Srg a'~sw4;ca5on Q. chop rs 6.8.5 PROGRAMS 8PN Z.sa 5.S 57, 5Ar Chen) <> Src 7ksbCiash'on B c'hl J ) Poataccident sampling activities will ensure the capability to obtain and analyze reactor coolant, radioactive iodinea and particulates in plant. gaseous effluenta, and containment atmosphere samples under accident conditions. These activities shall include the following: (i) Training of personnel, (ii) Procedurea for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis. s.b 10, he folloving report,s shall be submitted In pry),ga~c.g mi'hh IP CFC SA f BFN 6.0-13 Unit 3 U 31'0 P l Cggcqik s M) ~PCci4icrc~n 5:( 5'b; I A t at n on n basi of t e number of station, ut'ty d o er pe onnel inclu ing con ractors) for whom nito ing w s re red re eivin annual deep dos equi alent expos es gre ter t 100 em and eir as ociat man em exp ure cordin to work and job ld5ePx <s~>-) A. uncti s, g., re ctor peratio s and s eillance inse ice i spectio, rou e mai enance, special Qs mai tenan (descr be ma'ntenanc ), wast processi , and r fueli . The ose a ignment to vari us duty f ctions y be estimat base on mea rements sel rea n dosim ter, , or f lm badg . Small osure to alin 20% of t e indiv'dual tot 1 dose eed not e accoun ed for In the aggregat , at le t 80% f the total d ep dos equivale t expos e recei ed fro extern source 4 be assi ed to sp ific 'or wor functi ns. it 4e ticaigcCc n+ decimcte rcp lib< af ~4 chic'l ~

~~b. y mainst m reise va ve xn response re ching its s tpoint or due t operator act'on to control reactor ressure ll be re orte 5,S,I~~

'lQ single submittal may be made for a multiple unit station. h at' up lem nts he e ~erne ts o 20. 06 f C ar 20 BFN 6.0-14 Unit 3 A tabulation on an annual basis of the number of station, utility, and other personnel (iiicluding contractors), for whom monitoring was performed, receiving an annual deep dose equivalent > 100 mrems and the associated collective deep dose equivalent (reported in person - rem) according to work and job Sections (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [describe maintenance), waste processing, and refueling), This tabulation supplements the requirements of 10 CPR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, thermoluminescence dosimeter (TLD), electronic dosimeter, or film badge measurements. Small exposures totaling < 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 hdlh* b 'byAp'IM f hy .~e ~H percent of the total deep dose equivalent received from external sources should be assigned to speciQc major work functions. The report covering the previous calendar year 37> S <<cc$ ooeWon 5. g 5o &i CS ggfg 0~+G Qg'pgg k[ pumps a ~ca~) Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the .S. Nuc ear Regulat Commiss n, ATTN: Doc ent Control esk, Wa ington, D.C. 2055~, with a c he Re 'anal Office be s mitte no later than the fifteenth of each month following the calendar month covered. by the report. A arrative ummary of crating experi ce shall b submitted in he above s hedule.

~~6. 9. l. 4 REPORTABLE EVENTS Re rtable event, including orrective acti ns and mea es to preve t re-occurr ce, shall be reported to th NRC in~~

~ accordance with Section 50.73 to CFR 50. ~W~gf$ Pj@K~~ The Annual Radiological Environmental Operating Report covering he operation of th un't during the previous calendar year A~ shall be submitted May f each year. A sang +~ +~ submittal may be made for a -unxt station. he report G~Z~ shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlines in (1) the ODCM and (2) Sections ZV.B.2, ZV.B.3, and ZV.C of Appendix Z to 10 CFR Part 50. BFN Unit 3 6.0-15 PAGE 80F~ TS gyp CI'CCff Og ~cr ~(Q~P 6.9.1.6 SOURCE TESTS Resu reveal ts of requi contaminat n. e presence d leak f tests performed 0.005 microcu 'r on sour es more of if the vable tests S.f.5 CORE OPERATING LZ RT CMLR)

~~a. Core operating limits shall be established prior to each opeeat~ cycle, or prior to any remaining portion of an re(~/~~

epee ycle, for the following: (1) The APLHGR for Specification~~ g. g.f (2) The LHGR for Specification ~~ 3,2.g /~ 3.Z.2, p (3) The MCPR Operating Limit for Specification (4) he APRM Flo Biased Rod B ock Trip Set 'ng for S cification 2 .A.l.c, Tab 3.2.C, and cific tion 3 '.L The RBM Upscale (Flow Bias) Trip Setting and clipped value for this setting for Table ~~ cog'm'en 7,3.2,]'P O',Z.z,( -(

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (latest approved version) .

~~6.0-16 T~ ~7m S ccif:c n 5;r. (except aS ~gg)~~

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as s transient analysis limits, d accident analysis limits) of the safety analysis are met. ~~N~Y <~<<(~fi~g Sqs~>

~~Cod@~~

~~d. 'The including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.~~

Kc a %E~)iQAR ~4LcPOI~.gelr SW f~9gN sn RccerrlanCC u4+h The ~vagal Radioacti e Effluent Release Report covering the operation of the unit epesa4ioa shall be submitted report shall include summaries of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit A single submitta may be made for a multi-unit station. The submittal should combine those ~ sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section ZV.B.1 of Appendix I to 10 CFR Part 50. 6.9.2 SPECZAL REPORT Reports on the followi g areas shall e submitted in writing to the D'ctor of R gional Office, ivision of Re tor ro'ects: BFN 6. 0-17 Unit 3 "TS ping S~i $ cghp~ 5

1. Fa i e Usage 6..1.q ual Ope ting Report 2 Relief ve Tailpipe 3.M Wi hin 30 days aft r inoper-abilx y of t ermo ouple an acou ic monit on one val Sei ic Instrument ion 3.2.J'.3 'thin 10 days RJ Inoper ility aft 30 days of inoperability.
~~4. M eorological M itoring 3 ~2 .. 2 Withi 10 days Inst entatj.on after days of Inoperability inoperability.~~
6. Data shall be retrieved from all seismic instruments tuated during a seismic event an analyzed determine th magnitude of e vibratory groun motion. Special Repo shall be sub 'tted within 10 day after the vent descri 'ng the magnitude, frequency spect and resultant effect u on lant features i ort safet
7. Diesel Generator Reliability Improvement Program Report shall e submitted with'0 days of meetin failure criteria 'able 4.9.A. a minimum, the Re 'ability Improvement Program report for NRC audit shall include:
~~6. 0-18 BFN Unit 3 vAaE~4oF~6~~

~~7S ave W i~~~40~ 5',c~~

~~a. A summary of a ests valid and invalid) that occurred within he time period over which the last 20/100 valid t sts were performed.~~
~~b. Analysis o failures and determination of root causes of failur s.~~

C. Evalu ion of each of the recommendati ns of NUREG/CR-0660 "Enhancement of Onsite Emergen Diesel Ge rator Reliability in Operating eactors," with r spect to their application to e plant.

d. Identification of all action taken or to be taken to (1) Correct the root cause of failures defined in b above and (2) Achieve a neral improvement of diesel generator reliability.
e. A supplemental rep t shall be prepared fo an NRC audit within 30 ys after each subsequ t failure during a valid demand, for so long as he affected diesel gener tor unit continues to v olate the criteria ( /20 or 6/100) for the r liability improvem t program remedial act'on. The supplemental report eed only update the fa're/demand history f r the affected diesel generato unit since the last report for that diesel gen ator. The supplement report shall also presen an analysis of the failure(s) with a root ause determination, if possible, and shall lineate any further pro edural, hardware or operat'al changes to be incorporated into the site die el generator improvement program and the schedule for implementation of those changes.

BFN 6. 0-19 Unit 3 ~, Speci llcgI,~ g~ (taxi)C cs ~~++) High Range Primary 3~2. F Within days jLdk ~M Containment Radiation after 7 days of PM Monitors and Recorders inoperability. r(piA~

10. Wi Range Gaseous .2.F Within 7 days Effluen diation after 7 ys of Monitor and Recorder inoperabilx sic~.s/Ar,l.~ I~ c~y gghl Is ID XQ
~~6. 12 Changes to the ODCM:~~
~~1. Shall be documented and records shall be kept in a manner convenient for review. This documentation shall contain:~~
~~a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change.~~
b. A determination that the change will maintain the level of radioactive effluent control pursuant to 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
~~2. Shall become effective after review and acceptance by the process escribed in TVA-NQA-PLN89-A.~~

~~BFN 6.0-20 tTni t ~, 6.8. 1 (Cont ') HI C <Ã<<Pb qs mo~a<>) 6.8.1.2 (Deleted)~~

~~6. 8. 1.3 (Deleted) $44$ ~4 cygne~~~

W 8Pfll gyp g g DRZLLS 6.8.2 Drills on actions to be taken under emergency conditions involving release of radioactivity are specified in the Radiological Emergency Plan and shall be conducted annually. Annual drills shall also be conducted on the actions to by taken following failures of safety-related systems or components . RADZATZON CONTROL PROCEDURES C,A) 0 6.8.3 Radiat on Control Procedures shall be maintained and made availabl to all station pers el. These proced es shall contain ra tion dose limits shall be consist t with theI requirements o 10 CPR 20. This r ation protectio program shall be organiz d to meet the requir ts of 10 CFR except for the "control vice~ or "alarm signa ~ required by S.v H'5h

~~20. 1601 (4)~~

Guh'a]Van +co 4~We. ra%~ gwdAw ~ ~ a-~ rArkt roke 5.g,] ~~Each high radiation area as defined in 10 CFR 20 shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance 4 of a Radiological Mork Permit. Zndividuals qualified in radiation protection procedures (e.g., a radiological control technician) or personnel escorted by such individuals, shall be exempt from RWP requirements during the perfozmance of their assigned duties in high radiation areas at 30 centimeters, provided they otherwise comply with a radiation + pop;g(d in /0 cP'g g0, g@ ~ ~y~4LGQ'S ~])+ g g>+ fqgga Qon 70 /Pl(c) +c /II~'~g 8 )&~Is 4n glkt4014~ BPN Unit 3 (o.~(S ~~ P,h QP protection procedures for entry into high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or morc of the following:

~~a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.~~
b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the Radiological Work Permit. cd.

Al Zn addition, areas that are accessible to personnel and that have radiation levels greater than 1 rcm in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months as measured at 30 centimeters, but less than 500 rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 1 meter from the radiation source or from the surface which t radiation penetrates shall be provided with locked doors to prevent unauthorized entry. The keys shall be under the administrative control of the duty Shift Na <PCe Radiological Control Manager or their respective designees. Doors shall remain locked except during periods of access by personnel under an approved RWP which specifies the dose rates in thc immediate work areas and maximum allowable stay time for PAGE~DOF ~Q Pe J. L-BFN 6.0-8 Unit 3 T's 97g +tiki'+4+n 5. $ ;7a 7 (Cont ') individuals in that area. Zn lieu of the stay time requirement of the RWP, direct o remote (such as closed circuit TV cameras) continuous surveillance may be made by individuals qualified in radiation protection procedures to provide positive exposure control over the activities being performed in the ar a. i... C~O '+ ~~Individual radiation areas that are accessible to personnel, have radiation levels greater than 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months as measured at 30 centimeters, but less than 500 rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 1 meter from the radiation source, are located within large areas where QS vacagutcd 4P 3o no enclosure exists for the purpose of locking and where no + gal~ enclosure can be reasonably constructed around the individual vs ~ pQ~~~~ area, shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device whenever the dose rate in the area exceeds or will shortly exceed 1 rem in 1 6.8. i RADIOACTIVE EFFLUENT CONTROLS/RADIOLOGICAL ENVIRONMENTAL MONZTORZNQ PROG1NMS The following programs shall be established, implemented, and maintained.

6. 8. 4. 1 RADIOACTIVE EFFLUENT CONTROLS PROGRAM A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating 4't 5Ct. 54$ H ff~'yg IsT$ 5 g BFN 6. 0-9 Unit 3 PAGE~GDYSf .Pm >

BROWNS FERRY NUCLEAR PLANT - IMPROVED TECHNICAL SPECIFICATIONS SECTION 5.0 LIST OF REVISED PAGES CURRENT TECHNICAL SPECIFICATIONS JUSTIFICATION FOR CHANGES (Revised pages marked Revision I) NOTE: All pages are provided. Replaced Section 5.2 page 1 with Section 5.2 page I, Revision 1 Replaced Section 5.2 page 2 with Section 5.2 page 2, Revision 1 Added new Section 5.2 page 3, Revision 1 Replaced Section 5.4 page I with Section 5.4 page I, Revision 1 Replaced Section 5.4 page 2 with Section 5.4 page 2, Revision 1 Replaced Section 5.5 page I with Section 5.5 page I, Revision I Replaced Section 5.5 page 2 with Section 5.5 page 2, Revision I Replaced Section 5.5 page 3 with Section 5.5 page 3, Revision 1 Replaced Section 5.5 page 4 with Section 5.5 page 4, Revision 1 Added new Section 5.5 page 5, Revision 1 Replaced Section 5.6 page 1 with Section 5.6 page 1, Revision 1 Replaced Section 5.6 page 2 with Section 5.6 page 2, Revision 1 Replaced Section 5.6 page 3 with Section 5.6 page 3, Revision 1 Replaced Section 5.6 page 4 with Section 5.6 page 4, Revision 1 Replaced Section 5.7@age 1 with Section 5.7 page 1, Revision 1 Added new Section 5.7 page 2, Revision 1 JUSTIFICATION FOR CHANGES SECTION 5.1 - RESPONSIBILITY ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG. 1433. As a result the. Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or. Engl-ish language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change. A2 Site Vice President (SVP) added to BFN ISTS 5.1.1 to clarify that the SVP is responsible for the overall activities of the site and to require him to delegate that responsibility when absent. Changed the position title "Site Director" to "Site Vice President." This is not new information in terms of present site organization but does provide detail not described in existing Technical Specifications. TECHNICAL CHANGES - NORE RESTRICTIVE Hl This change proposes to add a requirement in Technical Specifications for the Plant Hanager, or his designee, to approve prior to implementation, each proposed test, or experiment to systems or equipment that affect nuclear safety. This change ensures the Plant Manager is aware of all changes with the potential to affect nuclear safety. This change adds requirements to the Technical Specifications which constitute a more restrictive change. This change is consistent with NUREG-1433, Rev. l. H2 This change proposed to specify the qualifications of the individual, who is designated to be responsible for the control room command function in the absence of the Shift Manager. In the CTS, no qualifications are specified for the designated individual. The change will require one of the two licensed operators to have an SRO during Modes 1, 2, and 3. Since this requirement will require one of the licensed operators to have an SRO (whereas currently both could have an RO) this is considered a more restrictive change. This change is consistent with NUREG-1433, Rev. l. BFN-UNITS 1, 2, 5 3 Revision 0 R JUSTIFICATION FOR CHANGES SECTION 5.2 - ORGANIZATION ADMINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BMR Standard Technical Specifications, NUREG 1433. As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BQR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change. A2 The term "Health Physics" in BFN TS 6.2.1 and 6.2.2 has been changed to "Radiological Controls" in proposed BFN ISTS 5.2.1.d 5.2.2.d to match plant-specific organization and function name. A3 Proposed BFN ISTS 5.2.2.e addresses administrative controls to limit working hours of unit staff who perform safety related functions. These requirements are currently addressed by CTS 6.8.1.1.b, which specifies that the overtime will be controlled in accordance with Generic Letter 82-12. Since the overtime controls proposed by ISTS 5.2.2.e are equivalent to those provided by Generic Letter 82-12, the proposed change is considered administrative. TECHNICAL CHANGES - NORE RESTRICTIVE Proposed Specification 5.2.2.f requires the Operations Superintendent to hold a current SRO license a BFN unit. Although BFN currently requires this qualification requirement to be met, it is currently not specified in Technical Specifications. Therefore, the proposed addition is considered more restrictive. M2 'ew requirements are being'dded Advisor. The proposed to specify the function of the Shift TS will require the STA to meet the Technical requirements of the NRC Policy Statement and will require the STA to provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to safe operation of the unit. This change is consistent with NUREG-1433, Rev. 1. BFN-UNITS 1, 2, K 3 Revision 1 IL JUSTIFICATION FOR CHANGES SECTION 5.2 - ORGANIZATION TECHNICAL CHANGES - LESS RESTRICTIV "Generic" LAl 1'VA proposed the Hinimum Shift Crew Composition Table not be retained in Technical Specifications. 10 CFR 50.54(k), (1), and (m) provide the requirements for the shift complement regarding licensed operators. The regulations describe the minimum shift composition for operating modes, as well as cold shutdown and refueling. Additionally, Specifications 5.1.2 and 5.2.2.b of the Improved Standard Technical Specifications specify the conditions when the licensed operator is required to be in the control room. Non-licensed operator requirements will be maintained in Specification 5.2.2.a. Removing the Table from Technical Specifications will not jeopardize plant safety nor is it necessary to be duplicated in order to assure safe operation of the facility. These requirements will also be included in plant procedures. LA2 CTS 6.2,2.b requires a licensed senior reactor operator to be present at the site at all times when there is fuel in the reactor. This is also required by 10 CFR 50.54(m)(2)(ii). Therefore, the removal of this specific requirement is a change in the presentation of requirements only and, therefore, is considered an administrative. Therefore, this unnecessary duplication of requirement has been eliminated. This change is made for consistency with NUREG-1433, Rev. l. "Specific'l CTS Table 6.2.A requires four non-licensed operators when all three units are shutdown or when one unit is in operation, and five non-licensed operators when two or three units are in operation. NUREG-1433, Rev. 1 requires three non-licensed operators for a two unit plant when both units are shutdown or defueled. Three non-licensed operators will be required if all three units are shutdown or defueled because Units 1 and 2 share a coaeon control room and because with one unit operating only three non-licensed operators are required by Improved Technical Specifications. In addition, with two units operating (different control rooms), four non-licensed operators will be required. Based on experience, TVA believes that the NUREG-1433, Rev. 1 provides adequate staffing levels for non-licensed operators to support safely controlling the plants, whether shutdown or operating. This change is consistent with NUREG-1433, Rev. l. BFN-UNITS 1, 2, < 3 Revision 1 JUSTIFICATION FOR CHANGES SECTION 5.2 - ORGANIZATION L2 CTS 6.2.2.d requires two licensed reactor operators and a licensed senior reactor operator during cold startups, plant shutdowns, and recovery from trips. Because of experience gained since Unit 2 restart in Hay 1991, TVA believes that this requirement is no longer'ecessary and that compliance with the manning requirements of ITS and 10 CFR 50.54 (k), (1), and (m) will assure adequate staffing of licensed positions. Therefore, this CTS requirement is deleted. This change is consistent with NUREG-1433, Rev. l. L3 CTS Table 6.2.A Note b, which allows the operating shift complement to be one less than the minimum requirement for up to two hours, is revised in ITS 5.2.2.c to delete the word 'one.'TS 5.2.2.c requires iaeediate action to fill the vacant position, and maintains the CTS requirement that such a condition shall not exceed two hours. Additionally, Footnote 1 to the Hinimum Staffing Table in 10 CFR 50.54(m) permits temporal'y deviations from the required staffing numbers as established in the unit's Technical Specifications. This change is consistent with NUREG-1433, Rev. l. L4 CTS Table 6.2.A Note b, which does not permit any shift crew position to be unmanned upon shift change, is not included in ITS. ITS 5.2.2.c requires immediate action to requirement that fill any vacant position, and does not condition shall not exceed two hours. negate the such a Additionally, Footnote 1 to the Hinimum Staffing Table in 10 CFR 50.54(m) permits temporary deviations from the required staffing numbers as established in the unit's technical specifications. This change is consistent with NUREG-1433, Rev. l. BFN-UNITS 1, 2, 5 3 Revision 1 JUSTIFICATION FOR CHANGES e SECTION ADMINISTRATIVE CHANGES Al 5.3 - UNIT STAFF QUALIFICATIONS Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433. .As' re'suit the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications; Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is .already approved, adding more detail does not result in a technical change. Revision 0 JUSTIFICATiON FOR CHANGES SECTION 5.4 - PROCEDURES ADMINISTRATIVE CHANGES A1 Reformatting and renumberi'ng are in'accordance with the BWR Standard Technical Specifications,'NUREG"'1433. "As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators-as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. -'Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change. A2 These types of procedures (CTS 6.8. l. l.c) are required by CTS 6.8. 1. l.a which requires BFN to establish procedures recommended by Regulatory Guide 1.33. The CTS requirement (6.8.l.l.a).is retained as proposed 0- ISTS (5.4.l.a), therefore, this deletion is considered a change in presentation \ The requirement and is therefore administrative. for ODCH implementation to be addressed by written procedures is covered 'by a more generic item, ISTS 5.4.l.e, which requires this activity for all Programs and Hanuals. Therefore, unnecessary to specifically identify each program. Since the it is requirements remain, this is considered a change in the presentation only and, therefor e, is considered an administrative change. TECHNICAL CHANGES - MORE RESTRICTIVE Hl This change proposes to add requirements for emergency operating instructions (EOIs) in the TS. The current TS do not specifically require the current form of EOPs (although BFN is committed to have them per NUREG-0737 and GL 82-33). The proposed TS will require EOIs which implement the requirements of the NUREG and GL. This change adds new requirements to the TS which constitutes a more restrictive change. This change is consistent with NUREG-1433. Revision 1 JUSTIFICATION FOR CHANGES.- SECTION 5.4 - PROCEDURES This change proposes to add the requirement that procedures be established, implemented, and -maintained for effluent and environmental monitoring. These procedures are not addressed by Regulatory Guide 1.33 and have been added to ensure that effluent and environmental monitoring functions are properly controlled. The addition of requirements in the TS constitutes a more restrictive change. This change is consistent with NUREG-1433. N3 This change proposes to add the requirement that procedures be established, implemented, and maintained for all programs identified in Specification 5-.5, ".Programs and Manuals." The addition of- the requirement that prdcedures be established, implemented, and maintained for the programs of Section'.5 is consistent with the requirement for these programs. The addition of requirements in the TS constitutes a more restrictive change. This change is consistent with NUREG-1433. TECHNIC L CHANGES - LESS RESTRICTIVE "Generic" LAl Requirements for procedures related to the review and audit 'function are proposed to be covered by the gA program, which will ensure these 0 requirements will be appropriately maintained. The change control process of 50.54(a) for the gA program will provide equivalent change control. eliminated. Therefore, this duplication of requirement has been LA2 10 CFR 50, Appendix E, section F establishes the requirements for conducting emergency drills. The onsite emergency plan is required to be exercised every two years. In addition, it is required that adequate emergency response capabilities be maintained between the biennial exercises by conducting drills. These requirements are implemented in TVA's Radiological Emergency Plan. Browns Ferry's experience is that adequate implementation of these requirements can only be achieved by running drills at least annually. In these drills, as an integral part of the scenario, failures of safety-related systems and components are required to achieve the scenario objectives. Therefore, TVA has determined that relocating the requirements contained in CTS 6.8.2 is acceptable and appropriate. This relocation is consistent with STS. It will not reduce the effectiveness of the emergency response organization because the Radiological Emergency Plan implements the applicable regulatory requirements. Changes to the Radiological Emergency Plan are controlled in accordance with 10 CFR 50.54(q). Revision 1 iTUSTZPZCATZON FOR CHANGES SECTZON 5 ~ 5 - PROGRAMS AND MANUALS ADMINISTRATIVE CHANGES A1 Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433. As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change. A2 This requirement merely restates that all applicable requirements must be met. Repeating this overall requirement as a specific detail is redundant and unnecessary; therefore, this detail can be omitted without ~- any technical change in the requirements. A statement of applicability of SR 3.0.3 is needed to maintain allowances for surveillance frequency extensions contained in the proposed Technical Specifications since these SRs are not normally applied to frequencies identified in the Administrative Controls section of the Technical Specifications. The addition of SR 3.0.3 is discussed in the proposed changes for BFN ISTS 3.0. Since this change is a clarification needed to maintain provisions that would be allowed in the LCO sections of the Technical Specifications, it is considered administrative in nature. A4 CTS 3.7.B.2.a states the results of the in-place cold DOP and halogenated hydrocarbon test at a 10% design flow on HEPA filters and char coal adsorber banks.... However, the Specification should state i "a 10K." The proposed Specification has corrected this error. A5 Instead of specifying the curie limit in the TS, proposed ISTS 5.5.8 contains a reference to 10 CFR 20, Appendix B, Table 2, Column 2, in order to determine allowable quantities of radioactivity in liquid holdup tanks. This change is considered a presentation preference only and, therefore, is an administrative change. Revision 1 JUSTIPZCATXOM FOR CHAHQES SECTION 5 ~ 5 PROGRAMS AND MANUALS A6 A statement of applicability of SR 3.0.2 and SR 3.0.3 is needed to clarify that the allowances for surveillance frequency extensions do apply, since these SRs are not normally applied to frequencies identified in the Administrative Controls section of the Technical Specifications. Since this change is a clarification needed to maintain provisions that would be allowed in the LCO sections of the Technical Specifications, it is considered administrative in nature. A7 Deleted (See response to NRC Coaeent 5.5-9). I AB Deleted (Not used in original submittal). A9 This statement has been added since the 1.25 extension allowed by SR 3.0.2 cannot be applied to Appendix J required test frequencies. A positive statement was added to specifically allow the SR 3.0.3 provision. These changes are consistent with the Appendix J, Option B Model for BNR/4 plants dated 10/31/95. A10 The 10 CFR 20 reference in CTS was based on a previous version of 10 CFR. The revised reference is correct for the current version of 10 CFR. TECHNICAL CHANGES - NORE RESTRICTI E NI This change proposes to add a requirement in the TS for the Primary Coolant Sources Outside Containment. This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. At BFN, this program is currently required by an NRC coaeitment and implemented by procedure and the normal preventive maintenance program for ECCS systems. Adding a requirement for this program to the TS constitutes a more restrictive change. BFN-UNITS 1, 2, L 3 Revision 1 e JUSTZPZCATZON POR CHANGES SECTZOM 5 ~ 5 - PROGRAMS AND MANUALS This. change proposes to add a requirement in TS for a Component Cyclic or Transient Limit Program. This program provides controls to track the cycl,ic and transient occurrences to ensure, that components are . maintained within the design limits. The addition of programs to the TS,.constitutes a more restrictive change. This change is consistent with NUREG-1433. H3 This change proposes to add a requirement in the TS for a Technical Specifications Bases Control Program. This program is provided to specifically delineate the appropriate methods and reviews necessary for a change to the Bases of...Technical. Specifications. H4 This change proposes to add a requirement in the TS for the Safety Function Determination Program. This program is included to support implementation of the support system Operability characteristics of the improved Standard Technical Specifications.'he addition of new requirements to the TS constitutes a more restrictive change. H5 A new requirement, proposed ISTS 5.5.9.b is added to the TS to perform a test for total particulate concentration of the fuel oil (w 10 mg/1) every 92 days utilizing ASTH D-2276, Hethod A-2 or A-3. TECHNICAL CHANGES - LESS RESTRICTIVE "Generic" LAl The following CTS requirements have been relocated to the Technical Requirements Hanual (TRH) for standardization and consistency with NUREG-1433, Rev. 1: CTS 3.8.A.6, actions to be taken with radioactive liquid waste exceeding the limits of CTS 3.8.A.5; CTS 4.8.A.6,determining the quantity of radioactive material in outside liquid radwaste storage tanks; CTS 3.8.B.9, limit for hydrogen concentration in the offgas system downstream of the recombiners; CTS 3.8.8.10, action to be taken when hydrogen concentration in the offgas system downstream of the recombiners exceeds the limits of CTS 3.8.B.9; CTS 4.8.8.5, reference to CTS 3.2.K and CTS 4.2.K; and CTS 3.2.K/4.2.K, explosive gas monitoring instrumentation. These, types of requirements are not required to be in technical specifications under 10 CFR 50.36. Changes to the TRH are controlled in accordance with 10 CFR 50.59. LA2 Details of the Inservice Testing (IST) program in the CTS are proposed to be relocated to the plant controlled IST program. The relocated requirements are duplicated in 10 CFR 50.55a, which requires the implementation of ASHE,Section XI and applicable addenda, for IST of ASHE Code Class 1, 2, and 3 pump and valves. Since the relocated provisions are required by regulations outside of TS, it is not Revision 1 0 I' JUSTXPXCATXOM POR CHANGES SECTXOM 5 ~ 5 PROGMHS AND MANUALS necessary to retain them in the ISTS. Changes to the IST program are controlled by the licensee controlled program requirements. i (V. LA3 Details of the Inse}vice Inspection Program (ISI) in the CTS are .proposed to be relocated to the plant controlled ISI program. The ISI Program is required by 10 CFR 50.55a to be performed in accordance with ASME Section .XI. The BFN ISI program implements the applicable provisions of ASHE Section XI. Generic Letter 88-01 provides an ISI program for piping in accordance with the NRC staff positions on schedule, methods, personnel, and:sample expansion or in accordance with alternate measures approved by the NRC .staff. " BFN-commitments to GL 88-01 do not need to be repeated in the proposed ISTS. Regulations and BFN commitments to the NRC .contain the necessary programmatic requirements for ISI without repeating them in the ISTS. Changes to the ISI program are controlled by the licensee controlled programs requirements. "Specific" Ll CTS 6.8.4.1.a 8 f uses the term "OPERABILITY" when referring to radioactive and gaseous monitoring instrumentation and treatment systems. Proposed ISTS 5.5.4 uses the term "functional capability." . The proposed change is necessary because the Radioactive Effluent Controls Program is located outside the Technical Specifications in the ODCH. Use of the term "OPERABILITY" can be confusing when used in programs which are not in the Technical Specifications. The term functional capability means that the component or system is capable of performing its design function. Since it is not a Technical Specification defined term, the use of "functional capability" is considered less restrictive than the use of the term "OPERABILITY." L2 The air distribution testing specified in CTS 4.7.B.1.c is not required by ANSI N510-1975 to be performed routinely. It is intended to be done only on initial installation. Therefore, this requirement is being deleted. L3 The frequency of the pressure drop test specified in CTS 4.7.B.1 for standby gas treatment system is being changed from once per year to once every 18 months. This change makes the testing frequency consistent with the same test that is done on the control room emergency ventilation system. Plant operating experience supports the extension of the surveillance interval. This change is consistent with NUREG-1433, Rev. l. The frequency of testing the inlet heaters for the standby gas treatment system specified in CTS 4.7.B.l.b is changed from once per year to once every 18 months. ANSI N510-1975 requires testing every two years, BFN-UNITS 1, 2, 5 3 Revision 1 JUSTIPXCATZOM ROR CHANGES SECTZOH 5 ~ 5 PROGRAMS AND MANUALS however, proposed ITS 5.5.7.e specifies once every 18 months .because this test is typically performed with the pressure drop test which ITS 5.5.7.d requires. at that frequency. Plant operating experience supports the extension of the surveillance interval. This change is consistent with NUREG-1433, Rev. 1. L4 The required value for pressure drop specified in CTS 4.7.B.l.ais changed from 6 inches of water to 7 inches of water. This change is being made because the .testing specified in ITS includes the prefilter and the testing that has been-performed as specified in CTS did not include the .prefilter, :The"additional one inch of water margin is necessary to account for the inclusion of this additional component. This-change is consistent with NUREG-1433, Rev. 1. L5 ,The frequency of testing specified in CTS 4.7.B.2.a is being changed as follows. The requirement to perform in-place testing specified by CTS 3.7.B.2.a after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation has been deleted. This requirement is intended to verify continued system effectiveness, after prolonged operation resulting from a reactor accident. The CTS 4.7.B.2.a requirement for a test every 18 months is carried forward in ITS 5.5.7. Also, the laboratory testing specified in CTS 3.7.B.2.b will continue to be required after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation and will adequately verify continued system effectiveness. Revision 1 JUSTIFICATION FOR CHANGES SECTION 5.6 - REPORTING':REgUIRENENTS ADNINISTRATIVE CHANGES Al Reformatting and renumbering are in accordance with the BWR Standard Technical Specifications, NUREG 1433. As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is done to make consistent with NUREG-1433. During ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change. A2 CTS 6.9.1.8 specifies a submittal date of "by April 1 of each calendar year" for the Radiological Effluent Release Report. Proposed BFN ISTS 5.6.3 states the submittal shall be in accordance with 10 CFR 50.36a. Since requirements for submitting this information are contained in 10 CFR 50.36a, specifying a specific submittal date is not necessary. Therefore, the proposed changes that eliminates the specific submittal date is considered administrative. A3 Proposed Specification 5.6.1 requires the Occupation Radiation Exposure Report to be submitted by April 30 of every year. Since 10 CFR 20.2206 specifies this date and CTS 6.9.1.2.a is in place to ensure these requirements are met, the proposed addition of the date is considered administrative. Current BFN procedures require the Annual Report, which contains this information, to be submitted within 45 days after the end of the calendar year. A4 Current Technical Specification 6.9.1.5 requires the Annual Radiological Environmental Operating Report to be submitted before Hay 1 of each year . Proposed Specification 5.6.2 requires the report to be submitted by Hay 15 of each year consistent with NUREG 1433, Revision 1. The proposed change still imposes the same requirement. As such, the minor adjustment in the submittal date is considered administrative. Revision 1 JUSTIFICATION FOR CHANGES SECTION 5.6>> REPORTING REgUIRENENTS TECHNICAL CHANGES - NORE RESTRICTIVE Hl CTS Table 3.2.F, Note 7, and CTS 6.9.2.9 requires a Special Report be submitted within 7 days after 7 days of inoperability of the High Range Primary Containment Radiation Honitor s and Recorders. Proposed BFN ISTS 5.6.6 requires a Special Report be submitted within 14 days, as required by Condition B or G of proposed BFN ISTS 3.3.3.1, when other PAH Instrumentation is inoperable. Since a Special Report is currently not required for other PAH instrumentation, the addition of this required is more restrictive. TECHNICAL CHANGES - LESS RESTRICTIVE "Generic" LA1 Deleted (Replaced by DOC L2. See response to NRC Cotangent 5 6-1 ) This change relocates the requirements for Reportable Events. These requirements are duplicated in 10 CFR 50.73. These requirements are relocated to plant procedures. The NRC and Industry have agreed to remove requirements from the Administrative Controls Section which are duplicated by other regulatory requirements. LA3 The requirements for reporting source test results that exceed allowable limits have been relocated to the Technical Requirements Hanual (TRH) for standardization and consistency with NUREG-1433, Rev. 1. This type of requirement is not required to be in technical specifications under 10 CFR 50.36. Changes to the TRH are controlled in accordance with 10 CFR 50.59. LA4 Deleted (Replaced by DOC L3,). The diesel generator (DG) special reporting requirements are relocated in their current licensing bases form to the Technical Requirements Hanual (TRH). A plant procedure implements the requirements and responsibilities for tracking emergency DG failures for the determination and reporting of reaching trigger values specified in NUHARC 87-00. These requirements are more restrictive than those specified in NUREG-1433, Rev. 1. In addition, Generic Letter 94-01, Removal of Accelerated Testing and Special Reporting Requirements for Diesel Generators, allows licensees to request removal from technical specifications of provisions for accelerated testing and special repol ting requirements for DGs. Browns Ferry proposes relocation only with no relaxation in the ITS conversion. The allowances of GL 94-01 BFN-UNITS 1, 2, K 3 Revision 1 JUSTIFICATION FOR CHANGES SECTION 5.6 - REPORTING REgVIRENENTS will be addressed separately, post-ITS implementation. Changes to the TRN are controlled in accordance with 10 CFR 50.59. This change is consistent with NUREG-1433, Rev. 1. "Specific" CTS Table 3.2.F, Note 7, and CTS 6.9.2.9 requires a Special Report be submitted within 7 days after 7 days of inoperability of the High Range Primary Containment Radiation Monitors and Recorders. Proposed BFN ISTS 5.6.6 will require a Special Report be submitted within 14 days of the allowed period of inoperability. The proposed change is less restrictive since 7 additional days are provided to prepare and submit the Special Report. Since the additional time has no affect on the safe operation of the plant and is consistent with NUREG-1433, the proposed change is considered acceptable. L2 CTS 6.9.1.2.b requires that any main steam relief valve that opens in response to reaching its setpoint or due to operator action to control reactor pressure to be reported to the NRC on an annual basis. The report provides a mechanism for the NRC to obtain information regarding challenges to safety relief valves after-the-fact, but provides no regulatory authority once the report is submitted {i.e., no requirement for NRC approval). Given that the report is only required annually and is not required to be approved by the NRC, it is clearly not necessary to assure operation of the facility in a safe manner. Therefore, this requirement is being deleted. This change is consistent with CTS. L3 CTS 6,9.2.1 requires that a Special Report of fatigue usage be included in the annual operating report. However, there is no other specific regulatory requirement to report this information which appears to be of limited value for reporting purposes. Therefore, the requirements specified by CTS 6.9.2.1 are being deleted to reduce administrative burden. BFN-UNITS 1, 2, L 3 Revision 1 JUSTIFICATION FOR CHANGES SECTION 5.6 - REPORTING REl}UIREHENTS RELOCATED SPECIFICATIONS Rl The requirements contained in CTS 6.9.2.3 regarding the reporting of Seismic Instrumentation inoperability have been relocated to the Technical Requirements Hanual (TRH) for standardization and consistency with NUREG-1433, Rev. 1. This type of requirement is not required to be in technical specifications under 10 CFR 50.36. Changes to the TRH are controlled in accordance with 10 CFR 50.59. R2 The requirements contained in CTS 6.9.2.4 regarding the reporting of Meteorological Monitoring Instrumentation inoperability have been relocated to the TRM for standardization and consistency with NUREG-1433, Rev. 1. This type of requirement is not required to be in technical specifications under 10 CFR 50.36. Changes to the TRM are controlled in accordance with 10 CFR 50.59. R3 The requirements contained in CTS 6.9.2.2 regarding Relief Valve Tailpipe Instrumentation and those contained in CTS 6.9.2.10 regarding the Wide Range Gaseous Effluent Monitor and Recorder have been relocated to the TRH for standardization and consistency with NUREG-1433, Rev. 1. This type of requirement is not required to be in technical specifications under 10 CFR 50.36. Changes to the TRM are controlled in accordance with 10 CFR 50.59. R4 The requirement contained in CTS 6.9.1.7.a{4) to establish core operating limits for the APRH flow biased rod block trip setting have been relocated to the TRM since the LCO and SR requirements related to this control rod block function have been relocated to the TRH. This type of requirement is not required to be in technical specifications under 10 CFR 50.36. Changes to the TRM are controlled in accordance with 10 CFR 50.59. BFN-UNITS 1, 2, K 3 Revision 1 JUSTIFICATION FOR CHANGES SECTION 5.7 - HIGH RADIATION AREA ADMINISTRATIVE CHANGES Al Reformatting and renumbe}ing are in accordance with the BWR Standard Technical Specifications, NUREG 1433. As a result the Technical Specifications should be more readily readable, and therefore, understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or-deleting) is done to make consistent.wi.th WUREG-1433. During .ISTS development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. Additional information has also been added to more fully describe each subsection. This wording is consistent with the BWR Standard Technical Specifications, NUREG-1433. Since the design is already approved, adding more detail does not result in a technical change. A2 These changes modify the requirements contained in CTS 6.8.3 for high radiation area control. Where dose rates are discussed, they are expressed in terms of the amount of radiation that an individual could receive in one hour, rather than instantaneous dose rates. These changes were made to make the specifications more consistent with the language used in 10 CFR 20 and to ensure that the specifications are not unnecessarily applied to situations where the radiation level is present for very short periods of time. In addition, distances for measurement were inserted where necessary since radiation levels vary with distance from the source. In CTS 6.8.2, the phrase "or continuously guarded" was placed preceding the word "doors" to permit the use of direct surveillance to control access to areas with dose rates greater that 1 rem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Both 10 CFR 20.1601(a) and (b) permit the use of an individual to continuously guard the entrance to a high radiation area for the purpose of access control. The use of this phrase is consistent with NUREG-1433, Rev. 1. TECHNICAL CHANGES - LESS RESTRICTIVE "Generic" LA1 The details contained in CTS 6.8.3 are proposed to be relocated to the procedures. This relocated program requires procedures to be t FSAR and prepared for personnel radiation protection consistent with 10 CFR 20. These procedures are for nuclear plant personnel and have no impact on BFN-UNITS 1, 2, 8E 3 1 Revision 1 ~, JUSTIFICATION FOR CHANGES SECTION 5.7 - HIGH RADIATION AREA nuclear safety or the heath and safety of the public. Requirements to have procedures to-.implement 10 CFR 20 are, contained in 10 CFR

~~20. 1101('c);~ Since. the*CTS'.requirements..a} e contained in the regulations; there, is no need .to repeat them in the BFN ISTS.~~

H'Q f I Revision 1 BROWNS FERRY NUCLEAR PLANT - IMPROVED TECHNICALSPECIFICATIONS SECTION 5.0 LIST OF REVISED PAGES NUREG-1433 BWR/4 STANDARD TECHNICAL SPECIFICATIONS MARKUP NOTE: Allpages are provided. Replaced page 449 of 478 (STS page 5.0-2) with page 449 of 478 Revl Replaced page 450 of 478 (STS page 5.0-3) with page 450 of 478 Revl Replaced page 457 of 478 (insert 5.0-9A) with page 457 of 478 Revl Replaced page 459 of 478 (STS page 5.0-11) with page 459 of 478 Revl Replaced page 460 of 478 (insert 5.0-11A) with page 460 of 478 Revl Replaced page 461 of 478 (STS page 5.0-12) with page 461 of 478 Revl Added inserts I and II after page 461 of 478 Replaced page 462 of 478 (STS page 5.0-13) with page 462 of 478 Revl Replaced page 464 of 478 (STS page 5.0-15) with page 464 of 478 Revl Added insert TSTF-118 after page 464 of 478 Replaced page 468 of 478 (STS page 5.0-18) with page 468 of 478 Revl Added insert TSTF-152 aAer page 468 of 478 Replaced page 469 of 478 (STS page 5.0-19) with page 469 of 478 Revl Replaced page 474 of 478 (STS page 5.0-23) with page 474 of 478 Revl Replaced page 475 of 478 (STS page 5.0-24) with page 475 of 478 Revl Responsibi 1 i ty 5el

~~5. 0 ADMINISTRATIVE CONTROLS M responds. ~4~~

~~'r AC 5I ~ pf cg pI $ 5 Pl hV r lo,ll~~

~~5. 1 Responsibility /Of- Cuba lt ae4'v 4l<f af W SIC > ~4 ~C~~

'hall ~ Ql 5.1.1 alfie Plant be responsible for overall unit w4sC C s e: C s ll Prt fse4% espons>> sty durin Hoe4 ~ operation.end. hall delegate in writing the succession to this absence. o yt F4c, ,The +lant s his designee shall approve, prior to PI~,F tv~> implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. ~ gh,A Wn~c< Sl C< Sl 5.1.2 The Shift Super~sar ~shall be responsible for the control room command function. During any abSence of the ~ g from the control room while the unit is in NODE 1, 2, or 3, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the rom the control room while the unit is in NODE 4 or 5, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function. 8( 5<;A g~g<z 5.0-1 Organization 5.2

5. 0 ADHINI STRATI VE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Or anizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.
~~a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management~~

~ levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentatio requirements shall be documented in the pfa~~r t7 &kH-a mp -g) P.QTP 5PT t 1 11 b p fb1 safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant; The shall have corporate responsibility for overall plant nuclear safety pp 'dc 4, vent and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and 44;46 N providing technical support to the plant to ensure nuclear cr CPc safety; and ~ y opycP coo~is P @f m eri'i,ick&cka

d. The individuals who train the operating staff, carry out E,gee.ukiac. Vi~ P f I 1tf 1 f 8m&~ <vl Act~ report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

~~5.2.2 Unit Staff The unit staff organization shall include the following:~~

~~a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator PAGE~OF~1~~~

(continued) 5.0-2 Organization 5.2 5.2 Organization 5.2.2 ~U t ~Sta f (continued) shall be assigned for each control room from which a reactor .is operatin in MODES 1, 2, or 3. Nl~~ unit hutdown or fueled, a total of three n licensed operato fpr the units. s

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, or 3, at least one licensed Senior Reactor Operator'SRO) shall be present in the control room.

~~S Ct /A lO'es~~

c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and .2.2.a and 5.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew Lt composition to within the minimum requirements.

~~P'A r ed lo Icag ct d o>~~

d. A gechniciangshall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in, order to provide for unexpected absence, provided immediate action is taken to fill the required position.
e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety functions (e.g., licensed SROs, licensed ROs, 4aakth 'elated

, auxiliary operators, and key maintenance perSOnnel ) ~ radial ~ g et% 5 cc sc/m 5 ]53 P$ ]Q Adequate sh>ft coverage shall be maintained without routine heavy use of overtime. The gbjective s a be to have operating personnel work an +~ r 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months day, nominal 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

~~1. An individual should not be permitted to wor k more than 16 hours straight, excluding shift turnover time; PpQE (continued)~~

~~> OF q7p + 5.0-3 i 0 Organization 5.2 5.2 Organization 5.2.2 Uni Staff (continued)~~

2. An individual should not be permitted to wor k more than 16 hours in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;
~~3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time;~~

~j

~~4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.~~

Any deviation from the above guidelines shall be authorized d h th Sgt ht d tg accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. +AN E shall be included in the procedures such that> JCt'ontrols individual overtime shall be reviewed monthly by the /&lant Y ht d tg t tht hours have not een assigned. Routine deviation from the above guidelines is not authorized. OR e amou .of ov rtime orked by unit s aff me ers perform'ng safe y rela d funct ons sh 1 be mited d contro led in ccorda e with he NRC Policy tateme on worki g hours (Gener'etter 82-12 . Sll I'aaa+acggF '

~~f. The gbperatioos (Q Q<~~~

flfgga) C" g. )shall hold aP Th dd SRO 'nse Curve at% h of thermal hydraulics, reactor engineering, analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by ~ry t On a 8p'ONES Rrrr hdtv Shift Technical Advisor STA) shall provide advisory and u< '~ uX th plant the Commission Policy Statement on Engineering Expertise on Shift. 5.0-4 0 Unit Staff qualifications 5.3

5. 0 ADHINISTRATI VE CONTROLS 5.3 Unit Staff gualifications Reviewer's Pote: Minimum qualifications f members of the unit staff shall be specifi/d by use of an overall quali 'cation statement refer cing an ANSI Standard 'cceptable to the NRC staff by specifying individ position qualifi tions. Generally, the firs method is preferable; owever, the second ethod is adaptable to those unit staffs requiring ecial qualification statements because of unique organizational structures.

5.3.1 Each me f the unit staff shall meet or exceed the minimum a ifications egu a ory u> e . , evisi n , , or more ecen revsssons, or ANSI tandard acceptable to he NRC staff]. The st f not covered by [Re latory Guide 1.8] sh meet or exceed t minimum qualificati s of [Regulations, Regulatory Guides, or SI Standards acceptable to NRC staff] 5.0-5 Procedures 5.4 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities: 'a ~ The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, F bruary 1978; g,g c~soKg

b. The emergency operating required to implement the requirements of NUREG-0737 and to NUREG-0737 Supplement 1, as stated in /Generic Letter 82-33/, 8 C. guality assurance for effluent and environmental monitoring;
~~d. Fire Protection Program implementation; and~~
~~e. All programs specified in Specification 5.5.~~

5.0-6 Programs and Manuals 5.5 5.5 Programs and Manuals The following programs shall be established, implemented and maintained. 5.5.1 Offsite Dose Calculation Manual ODCM

a. The ODCH shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip sgtpoints, and in the cond~et of theQdiological Pilvironmental Anitoring pHogram; and
b. The ODCH shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release, reports required by aq Specification .6.2+and Specification j5.6.~

~~Licensee initiated changes to the ODCH:~~

~~a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:~~
~~1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and~~
2. a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20. 1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;
~~b. Shall become effective after review and acceptance by the~~
~~and pro~si Wax . c4 ~ %vs"<Q4" P~N 5'-A~~
c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCH as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCH was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page PAG E~OF~~S (continued) 5.0-7 Programs and Manuals 5.5 5.5 Programs and Hanuals 5.5.1 1 ti Hn (continued) that was changed, and shall indicate the date (i.e., month and year) the change was implemented. 5.5a2 imar oo nt Sources Out id ontainm This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or acqident to s low as practicable. The systems include he 4aw-W ~~I-Core Spray, High Pressure Coolant Injection, Residual Heat Removal, Reactor Core Isolation Cooling The program shall include the followin : va

~~a. Ereven ive i te and+eriodic visual inspection requirements; and Sys leak test requirements for each system at refueling cycle intervals or less.~~

Pt a~~ Ly rye 5.5.3 Ac i n mli dn~ Ce Oakum, This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following:

~~a. Training of personnel;~~
~~b. Procedures for sampling and analysis; and~~
~~c. Provisions for maintenance of sampling and analysis equipment.~~

5.5a4 Radioactive ffluent Controls Pro ram This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effloents as low as reasonably PAGE~SOF (continued) Programs and Manuals 5.5 5.5g4 dioactive ffluent Controls Pro ra (continued) achievable. The program shall be contained in the ODCH, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements: a~ Limitations .on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCH; q'1

~~b. Limitations on the concentrations of radioactive material lo )g ~cs F~~~

co~('g <nomic l oPvcs iA n released 1 fgt ~ggggf in liquid effluents to unrestricted areas, t,fgl g,tl g: /0 cFgt 2' ~ Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20. 1302 and with the methodology and parameters in the ODCM;

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I;
e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCN at least every 31 days; Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;
~~g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary~~

~~.4eS r..g + ~ 4'oil 5:o- fA (continued) AGE ~~~ oFm<< 5.0-9 0, 0 ~TR RT~~

~~1. For noble gases: a dose rate of < 500 mrem/yr to the total body and 3000 mrem/yr to the skin, and A~~
~~2. For iodine- 31, iodine-133, tritium, and for all radionuclides in particulate form with half lives > 8 days:~~

~~.a dose rate of < 1500 mrem/yr to any organ; Programs and Manuals 5.5 5.5.4 Radioactive ffluent Controls Pro ra (continued)~~

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; 3 ~ Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation 4gom uranium fuel cycle sources, conforming to 40 CFR 190> an
k. Limit tions on venting purging of t Mark II cont inment through t Standby Gas Tr tment Syste to mai ain releases a ow as reasonab achievable in BWR/4s w' Mark II contai ments) 5.5.5 om onent C lic or T a sient t program provides controls CP to track the FSAR Section cyclic and transient occurrences to ensure that components are

~, d.z.s'his maintained within the design limits. .5.6 Pre-Stressed Concrete Containment Tendon Surveillance ro a This program provides con ols for monitoring any tend degradation in pre-stressed oncrete containments, inclu ing effectiveness of its corrosio protection medium, to ensu containment structural integri . The program shall inclu baseline measurements prior to i tial operations. The Tend veillance Program; inspection f quencies, and acceptance cri ria shall be in accordance with Regulatory Guide 1.35, Revis n 3, 1989]. The provi 'ons of SR 3.0.2 and SR. 3.0.3 ar applicable to the Tendon Sur 'llance Program. inspection frequ cies. PAQE~5~OF~1 (continued) 5.0-10 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) s.sk" ns vice Testin Pro ra This program provides controls Class I, 2, and 3 components program shall include the fol owing: a. 'f'c 'rvice testin of A Testing frequencies specified in Section XI of the ASME u ASME Code . The Boiler and Pressure Vessel Code and applicable Addenda are as follows: ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice ctiv't'e test'n a tivitie Meekly At least once per 7 days Monthly At least once per 31 days quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months. At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

~~b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;~~
~~c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and~~
~~d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.~~

5.5 P +8 p p Ve ti ation h F'er Testin 11~ t bli h~ Pro ram testing of Engineered Safety Feature systems. (ESF) the 4eqqe~ filter ventilationrequirede ieu X'N E'e A T':4-II OPs ZONE T FRoH S.O /3 (continued) 5.0-11 Section Sa, and maintenance ventilatio hours on following signi at 't the t o painti per 18 months or 1) after any structural ter or ne communicating w system operation. charcoal adsorber housings, 2) fire or chemical release in any e system, or 3) after every 720 PAGE~+Cl OF fed. 4 Programs and Manuals 5.5 ( s4~hp ~s 7iM~r~Cfog.v-) s~g a~et go~frwl Rw~ C'~+p~g yt f. g dt., CCP4V) S

~~5.5 Programs and Manuals ps.~~

Q~ Ventilation Filter Testin Pro ram V TP (continued) Demonstrate for each of the ESF systems that an inplace test of the HEPA filters shows a penetration an system byp s < ~45/% when tested in accordance with at the sys em flowrate specified below'++ 10g+~ga AN SX (Js(Q>>/979 8 ESF Ventilation System Flowrate CcF )(R) 5&w Sys*~ good Sys*~ 3040 i C$ 'GV t 4$ %r ~g~

b. Demons ate for each of the )SF systems that an inplace test of the arcoal adsorber shows a penetration and system b ass < . % when tested in accordance with 5 J.o at the system flowrate specified belowrtl: 10/~(5g ESF Ventilation System Flowrate pc3' ~Pa srr s~ $ 0Oo C RCV Syr*~ 3eoc Id':Lr ~ 0"
c. Demonstr ate for each of the ESF systems that a laboratory test of a sample of the char coal adsorber, yg , shows %he ~

methyl iodide e4~ -9>l b~w when tested in ac ordance with STM D3803-1989 133 ESF e S stem Penetration(g Hffc) ~Pc Sc < Sys*~ 95 CIMv'7s*~ +o +cswq cWo 4 i3'/ ~~i)~is q ~ ~ 4'~ 53+43~ ~q,4rO, ~ster-C>>3 aA sr QB vsa 4 ~ A. VCf~ (continued) PAGE~N~oF +~K 5.0-12 ~i+ [ l lg5~ 74g5 Wcg ~ ~~Wg r) pcu+d m a ~ ~~ ~~ >'L mcnab~ x) a>ca- + Acp/ $gvd4 %Secs ,, 3),~ ~~ /au.~ ~M Q Qua~ .~ca~ ~ ,I ~ I I ~ '.... Gr ~e ~1~>> ..~"~., Ce~w~~ ~ relMse iu, cLuu ~~ I 4 'tI ...SkuN.~ I M or eke. I C4c~ al. ~~r4cC (~ ) 'gJ Ct$ tcv gQ~'~ ~45f u ~ 4)~&icw4:...',.~.m d ~igULcel~ i~ 1 ~l nhac ~e eeywuu,ii,~ sl-c~ I ~ 4 ~ 'I 4~I

~~lp ~~~

,~ I Programs and Manuals 5.5 5.5 Programs and Manuals tpZ 5.5 7 atio ilt r Testin o am VFTP (continued) 't Review 's Note: Allowable p etration [10(C - me hyl iodide effic'ency for charcoal cred ed in staff safety e luation]/ (sa ty factor). fety factor [5] for ystems with heaters. [7] for systems without heaters.

d. ~onstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested the system flowrate specified as-fellows fs IN (3 P5 pe<~ ESF Ventilation System Delta Pli'~Q Flowr ate C<C ~Pa.

~~$4 < <yJh~ 'fQOO CgCV Syg+~ S&1~~

~~e. Jkmonstrate that the heaters dissipate for~I-of the~ grstem~~

@ lÃgwhen tested in accordance with 40k B3 Hf HOuL kca he provisions of SR 3.0e2 and SR 3.0.3 are applicable to the V Peed ~ test fre uencies. %0 II JO~~Jg gg g p< fJP esp ~Cease L 4gf S g S.S losive Gas and Stora e Tank Radioactivit Monitorin Pro ra his program rovides controls for potentially explosive as mixtures containe n e as e as o up quan sty ra y c tained i as ff a tre an the quantity of radioactivit contained in unprotected outdoor liquid storage tanks+ g<o (continued) PAGE A bZ OF /'7F N~TS 5.0-13 [ Programs and Manuals 5.5 5.5 Programs and Manuals ,.,Qg losive Gas and Stora e Tank Radioactivit Monitorin Pro ra (continued) gaseous r soac >vs y quantities shall be determined following the methodol y in [Branch Tech cal Position (BTP) TSB 11-5, "Postul ed Radioactive Re ase due to Waste G System Lea or Failur ]. The liquid r waste quantities s 11 be dete ned in accor ance with [Standar Review Plan, Section 15.7.3, " ostulated Radioactive Release d 'lures" . The program shall include: dy~eg4r .5 rem to any ndividual in an unr tricted area, in the event of fan uncontrolled release of t e tanks'on ~ A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwasfe tanks ~ ~ that are not surrounded by liners, dikes, or wallsg capable of holding the tanks'ontents and that do not have tank verflows and surrounding area drains connected to the lag iquiddfadwaste geatment siiystem~is less than the amount t at would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks'ontents. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies. (continued) PAGE~&3 OF <~~ 5.0-14 P Programs and Manuals 5.5 5.5 Programs and Manuals (continued) pA 5.5.48 Diesel Fuel Oil Testin Pro ra ;geCtI 7-/P C~~/O. t +.K A diesel fuel oil testing rogram to implement required testin of fuel oil s all be established ram s a ~ in e samp ~ ing an testing requirem n s, and acce nce cr'a The purpose of the program is to establish the ollowing: ~ ~ a ~ a i ity o new ue oi or use prior o a i ion o storage tanks by determining that the fuel oil has:

~~l. an API gravity or an absolute specific gravity within limits,~~
~~2. a sh point and kinemat viscosity wit 'n limits for ASTM fuel oil, and~~
3. a clear a bright appearance wit oper color; Other properties fo STM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks; and Total particulate concentration of the fuel when tested every days in accordance wit o's D-2276, ASTM

< 10 mg/1 Method A-2 or A-3. lwstM.< VsW-n C ~ P<'://~44eg7M>p fv Joe Technical S ecifications TS Bases Control Pro ra This program provides a means for processing changes to the Bases of these Technical Specifications.

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

~~Aflak~~

~~b. Licensees may make changes to Bases without prior NRC approval provided the changes do not 'involve either of the following:~~

~~I, a change in the TS incorporated in the license; or~~

~~2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.~~

P>e '~ gv~l+ pf $ 4c Qvcp p I w Pot( 7 cd' / h fpe~'~> '- ><4 I' cl ds / ~~/ if K7f-/5p(Q k(gg4/tuQ continued) PAGE~// OF ~l 5.0-15 INSERT TSTF-I I 8 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program testing frequencies. Programs and Hanuals 5.5 5.5 Programs and Manuals o ~~ 5.5. T chnical S ecifications TS ases Control P o ra (continued)

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

~

d. Proposed changes that meet the criteria of Specification 5.5. 11b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

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

~~a. Provisions for cross division checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;~~
~~b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;~~
~~c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and~~
~~d. Other appropriate limitations and remedial or compensatory actions.~~

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

~~a. A required system redundant. to/system(s) supported by the inoperable support system is also inoperable; or (continued)~~

~~PAGE~46 OF~78 5.0-16 Programs and Manuals 5.5 mC~3 5.5. I+ Safet Func ion etermination Pro ram SFDP (continued)~~

~~b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or~~
~~c. A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.~~

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. 5.0-17 i ar Containment eaka e Rate T stin Pro ram A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidel'ines contained in Regulatory Guide 1.163, '"Performance-Based Containment Leak-Test Program," dated September 1995. The peak calculated containment internal pressure for the design basis loss of coolant accident, P , is 49.6 psig. The maximum allowable primary containment leakage rate, L , shall be 2X of primary containment air weight per day at P . Leakage Rate acceptance criteria are:

~~a. The primary containment leakage rate acceptance criteria is~~

< 1.0 L~. During the first unit startup following the testing performed in accordance with this program, the leakage rate acceptance criteria are < 0.60 L~ for the Type B and Type C tests, and < 0.75 L~ for the Type A test; and

~~b. Air lock testing acceptance criteria are:~~
~~1) Overall air lock leakage rate < 0.05 L~ when tested at p~.~~
~~2) Air lock door seals leakage rate is < 0.02 L when the overall air lock is pressurized to 2 2.5 psig for at least 15 minutes.~~

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program. r AQE~b7 op V7f Reporting Requirements 5.6

~~5. 0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.~~

~~5.6.1 Oc 'ationa adia io osure e t dog v'~~

~~-NOTE A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the'tation.~~

tabul ion an annual bas'f the mber o tation, ut'lity, JAG' and o er p sonnel (includ g contra ors) rec 'ng res
> I mrem r and heir as ociated n rem ex sure acc ding to wo and ob fu tions ( .g., rea or operat'ons and s rveillan i servi insp ction, r utine m 'ntenance, pecial >ntenanc descr' mai tenance , waste ocessing and refu ing). T is abul tion ppleme s the r uirements of 10 CF 20.2206. The dos assi ments t various uty func ons may e estima d base on dosimet r, the uminesc t dosime r (TLD), r fil dge m asureme s. Sma exposur totalli g ( 20K o the ndivi ual tot dose n d not be accounte for. In he aggr ate, a least 8 of the tal recei ed from ext nal so ces sho d be ass'gned to ecific m or wor fu tions. The rep t shall e submit ed by Apr'0 o eac y ar.
5.6.2 nnu adi i a v'me t 0 i e o

~~NOTE A single submittal may be made for a multiple unit station. The submittal should combine sections commn to all units at the station.~~

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted bv Hay R of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Honitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Hanual PAG E~~OF~1'8 (continued)
BMR/4 STS 5.0-18 Rev I, 04/07/95
APR-22-S7 TUE 10:06 P. 09 7 $ 7$ -/sz
~SEILT, A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors), for whom monitoring was performed, receiving an annual deep dose equivalent > 100 mrems and the associated collective deep dose equivalent (reported in person - rem) according to work and job fimctions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [describe maintenance], waste processing, and refueling). This tabulation supplements the requirements of 10 CPR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, thermoluminescence dosimeter (TLD),
electronic dosimeter, or film badge measurements. Small exposures totaling < 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total deep dose equivalent received from external sources should be assigned to specific major work functions. The report covering the previous calendar year shall be submitted by April 30 of each year.
~ ~
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 nual Radiolo ical nvironmental 0 eratin Re ort (continued}
(ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.
he Annual Radiological Environmental Operating Report shall include the results of analyses of all radiologica1 environmental samples and of all environmental radiation measurements taken du ing the period pursuant the locations speci ' in the table and igures in the ODCM, as 11 as summarized and ulated resul of these analyses and asurements ~n the fo t of the table i the Radiological Assess nt Branch Technical Po ition, Revision , November 1979~ ~e port shall identify t TLD results th t represent collocated do imeters in relation to he NRC TLD pro am and the exposure perio associated with each result.~ In e event that some indivi al results are not available for i lusion with the report, he report shall be submitted noting nd explaining the reasons for the missing results. The miss g data shall be submitted in a supplementar report as soon as possible.
Radioac ive E fluent Rele se Re ort 54nll NOTE-A single ubmittal may be made for a multiple unit station. The submittal combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
dL .~4 ~ The Radioactive Effluent Release Report covering the operation of p vM q~ the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summ ry of the quantities of radioactive liquid and gaseous effluent and solid waste released from the unit. The material provide shall be consistent with the objectives outlined in the 0 CH and Process Control Program and in conformance with 10 CFR 50.3 a and 10 CFR 50, Appendix I, Section IV.B.1.
Of Cnc4.~
5.6.4 Monthl 0 eratin Re orts BMR/4 STS
$ 1.
Routine reports of operating PAGE~
5.0-19
~
statistics and shutdown experienceP,'-
'continued)
Rev l. 04/07/95
0 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6e4 Monthl 0 eratin Re orts (continued) eml~,+sha11 be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.
5.6.5 COR OP RAT NG M TS R PORT CO
~~a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:~~
The i 'dual specifi'cCctions that ad~ss core op~+ing limits mus be referenced h r .
he ana y >ca method used to determi the core operating limi shall be those pr 'sly reviewed approved by the NR , ecifically those scribed in the ollowing documents:
I tify the Topical Repo (s) by number, title, date, an NRC s ff approval document, r identify the st f Safety Evaluate Report for a plant cific methodology NRC letter and date.
c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
~~d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.~~
~~5.6.6 eactor Coolant S stem RCS PRESSURE AND TEMPERATURE LIMITS R ORT PTLR~~
~~a. R pressure and tempera e limits for heatup, c ldown, itically, and hydrosta 'c Q low mperature operation,~~
~ testin s well as heatup and establish ldown rates shall be and documented in the PTLR for the following:
PAGE~OOF QK- (continued) 5.0-20
Oi INSERT 5.0-20A (1) The APLHGRs for Specification 3.2.1; (2) The LHGR for Specification 3.2.3; (3) The MCPR Operating Limits for Specification 3.2.2; and (4) The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2. 1, Table 3.3.2.1-1.
INSERT 5.0-20B
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel,'" (latest approved version for BFN).
Reporting Requirements 5.6 5.6 .
Reporting Requirements 5.6.6 eactor Coola t S stem R PR SSURE AND T MP RATUR M TS
[The individual s cifications that address RCS pressure and temperature limi s must be referenced here.]
b. The analytical methods used to determine the RCS pressure and temperat e limits shall be those previously reviewed and approve by the NRC, specifically those described in the following ocuments. [Identify the NRC staff approval document y date.]
c ~ The PT shall be provided to the NRC upon suance for each react vessel fluence period and for any evision or supp ment thereto.
Reviewe s'otes: The methodology for the alculation of the P-T limit for NRC approval should include t following provisions:
The methodology shall describe h the neutron fluence is calculated (reference new Regul tory Guide when issued).
The Reactor Vessel Material urveillance Program shall comply with Appendix H to CFR 50. The reactor vessel material irradiation surv illance specimen removal schedule shall be provided, alon with how the specimen examinations shall be used to updat the PTLR curves.
~~3. Low Temperature Ove setting limits for the ressure Protection'(LTOP) System Power Operated Relief Valves (PO s),~~
~~lif developed using C-approved methodologies may be incl ded in the PTLR.~~
4. The adjusted eference temperature (ART) for eac eactot beltline m erial shall be calculated, accounti for radiation mbrittlement, in accordance with R ulatory Guide 1.99, Re sion 2.
The limiting ART shall be incorporated i o the calculation of the pressure and temperature limit rves in accordance with NUREG-0800 Standard Review Plan 5.3.2, Pressure-Temperature Limits.
~~6. The minimum temperature require nts of Appendix G to 10 CFR Part 50 shall be incorporated i to the pressure and temperature limit curves.~~
(continued) 5.0-21
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Rea tor Coola t S stem RCS PR SSUR AN T HP RATU PORT T (continued)
7. Licensees who have remov two or more capsules s uld compare for each surveilla e material the measure increase reference temperature (RT ) to the predicted in ease in RTNDT, where the predicte crease in RT~> is bas on the ean shift in RT > plus the wo standard Beviation value 2oa) specifieF7n Regulator Guide 1.99, Revision If the easured value exceeds the pr icted value (increas in RT~~ 2crz), the licensee should p vide a supplement to the PTER t demonstrate how the results feet the approved methodology.
5.6 G Failu es R ort If an individua emergency diesel generator ( G) experience four r more valid fai es in the last 25 demands, ese failures d a nonvalid failure xperienced by that EDG in t time perio sha be reported withi 0 days. Reports on EDG fa ures shall inclu the information re ended in Regulatory Guide .9, Revisio 3, Regulatory Positi .5, or existing Regulatory Guide 1.1 re o 6 t'+
5.6 ~PM II I uben a report is required by Condition 9 or G of LCO 3.3+3.1f, "Post Accident Honitoring (PAN) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
Wyiewer's Note: These eports may be required covering ins tion, test, and mai nance activities. These ports are determ d on an individual sis for each unit and the preparati and submittal are ignated in the Technical Specification .
PAGE~lOP~7 5.0-22
I (PPI)(mggd dgg 044) bligh Radiation Arear
$ 5.7+~
~~5. 0~~
~
ADMINISTRATIVE CONTROLS ~o 'i
..a(>>')
gc loll ~+ J; /; rNF k5.7 High Radiation Area/ ~~ al ow +i X~ kr 5.7.1 ~ pgoujJgg 10 0F0 00, 0 g 0 0 .10010 10 0F0 00:1601 h high di 11
'rk defined in 10 CFR 20 shall be barricaded and conspicuously posted as a high radiation area and entrance thereto Cubit shall be controlled by requiring issuance of a Permit (RWP). Individuals qualified in radiation protection rocedures (e.g., ) or personnel escorted by such individualg +ay be exem t from the
~ Q'pic RWP 4sseaaec requirements during the performance of their assigned duties in hi h radiation area S+
provided t ey are- otherwis ra ia io protection co,4~
procedures for entry into such high radiation areas. -Kc.
~
tV herc Wh~ doScb ce~ply i'd@- o~p~~+
~
Coul+ Any individual or group of individuals permitted to enter such A5
~~sech i<i.4owr areas shall be provided with or accompanied by one or more of the following:
Acc. ~
4Cuc.
pcalnah 6
~~a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.~~
b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such a) eas with this monitoring device may be made after the dose rate levels in the area have been established and personnel mare of them. pad esp Crt
~
c. An individual qualified in radiation rotection procedures s with a radiation dose rate monitot ing device, g responsible for providing positive ontrol over activities within the area and shall perform periodic radiation surveillance at the frequency specified k~he-in the RWP.
>i~inS 15 ~r < ip c~,Mr~, iD c agcss lsJ4,
~
C po PacCr~i+~c ~gag~~~
5.7.2 In additio o o eci ication . .1, areas with ra a io eve s hall be provided with locked
~ors to prevent unauthorized entry. e keys
,4A~W s a e u'nUer the administrative control of the~Shat J-Q
~esa Doors shall remain locked except during periods of access by personnel under an approved RWP specifg the dose rat~~ in the immediate work
~gp QILJ4J lg i Or P4'q'r rrf+iAOpggeerr ~
(continued)
PAGE BWR/4 STS 5.0-23 Rev 1, 04/07/95
l'(3 Cpecefl as ~) High Radiation Area gS.7P ~1
.7 High Radiation Are+
5.7.2 (continued) <pleo <C'~
arear'nd the maximum allowable stay time for' individuals in thdd area/. In lieu of the stay time of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made b qualified in radiation protection procedures to rovide positive exposure control over the activities being erformed within the area.,
5.7.3 V.
~Adividnal a + ~ ar ~ aa g $ + dr gaKCDahk high ra ia ion areasjvith radtatton levels ef Xre t'.h~ are located within arge areas where no enclosure exists for purposes and where no of locking
'hall enclosure ' can be reasonably constructed around the d~
individual area, be barricadedi and.
conspicuously .posted, and a flashing light shall be activated as a warning devic .
M <~c~ f4c AaLc ~le
~I g goo I-M + go t&~hgg
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~~ I~rAr g~~
4g 4 i.~s..u P
~ I "s 3 4O~ AS ~C.a.S~Cd
<~lake
>o ~ We,ej,,~ S ~ er ~~e PAG E~SQF~K BWR/4 STS 5.0-24 Rev I, 04/07/95
BROWNS FERRY NUCLEAR PLANT - IMPROVED TECHNICALSPECIFICATIONS SECTION 5.0 LIST OF REVISED PAGES JUSTIFICATION FOR CHANGES TO NUREG-1433 NOTE: Allpages are provided.
Replaced page 1 through 3 with page 1 through 3, Revision I.
Added new page 4, Revision l.
alUSTIFICATION FOR CHANGES TO NUREG-1433 SECTION 5.0 - ADMINISTRATIVE CONTROLS BRACKETED PLANT SPECIFIC INFORMATION Bl Incorporates BFN plant specific information regarding position titles, organization names, FSAR chapters, and ISTS section numbers. In 5.2.l.a the bracket [FSAR] is replaced with Nuclear Power Organization Topical Report (TVA-NPOD89-A).
B2 Brackets removed and optional wording deleted since it does not apply to BFN or the program has been relocated to plant procedures.
I ~
B3 Brackets removed and values revised to reflect BFN plant specific requirements.
B4 Brackets removed and wording revised as necessary to reflect appropriate plant specific requirements. Reviewers notes deleted.
B5 The change to the information contained within the brackets reflects BFN licensing bases requirements.
B6 These changes modify the requirements for high radiation area control.
0 In the opening paragraph, the word shall" was changed to "may" to make clear that all of the alternatives given in 10 CFR 20. 1601 are acceptable.
Where dose rates are discussed, they are expressed in terms of the amount of radiation that an individual could receive in one hour, rather than instantaneous dose rates. These changes were made to make the specifications more consistent with the language used in 10 CFR 20 and to ensure that the specifications are not unnecessarily applied to s'ituations where the radiation level is present for very short periods of time. In addition, distances for measurement were inserted where necessary since radiation levels vary with distance from the source.
Specification 5.7.2 applies to areas which exhibit radiation levels greater than, but not equal to, I rem in I hour. This is consistent with CTS and Regulatory Guide 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants."
Revision I
0 JUSTIFiCATION FOR CHANGES TO NUREG-1433 SECTION 5.0 - ADNINISTRATIVE CONTROLS NON-BRACKETED PLANT SPECIFIC CHANGES Pl Incorporates plant specific information regarding Site Vice President responsibilities. Added Site Vice President to TS to specify his responsibilities relative to the Plant Hanager.
P2 ISTS renumbering, or minor rewording (TVA preference).
P3 This change incorporates plant specific terminology and changes or deletes NUREG requirements to make the proposed BFN ISTS consistent with the BFN licensing bases.
P4 Incorporates plant specific information regarding position titles, organization names, and functions.
P5 Editorial, grammatical, typographical corrections.
p6 This change modifies requirements for leak testing primary coolant sources outside containment to limit tests to the extent permitted by system design and radiological controls. Rewording to clarify that preventive maintenance is fulfilled by periodic visual inspections and system leak tests. BFN does not have a comparable Technical Specification requirement. However, BFN's current programs for minimizing leakage are consistent with the proposed Technical Specification requirement.
P7 Proposed changes incorporate current licensing bases requirements that were recently changed (BFN TS 335, Amendment No. 201, 220, and 174 for Units 1, 2, and 3 respectively issued December 2, 1993) to incorporate the corresponding revised 10 CFR 20 terminology or to retain operational flexibility consistent with Appendix I to 10 CFR 50, concurrent with the implementation of the revised 10 CFR 20.
PB The requirement of Specification 5.5.4.k for the Radioactive Effluent Controls program to include the limitations on venting and purging of the Hark II containment was deleted based on the fact that BFN has a Hark I containment.
Revision 1
JUSTIFICATION FOR CHANGES TO NUREG-1433 SECTION 5.0 - ADNINISTRATIVE CONTROLS P9 The statement that SR 3.0.2 and SR 3.0.3 are applicable to the VFTP Frequencies has been moved to right after the paragraph stating the Frequencies. This is to ensure the allowances are not inadvertently missed and for user friendliness; the allowances should be after the Frequencies, not three pages later.
P10 The 'change deletes program requirements 5.5.9.b that are not applicable to the BFN plant specific design. There are not any gas storage tanks (other than holdup pipes) and no method to limit curie content in the holdup pipe except reactor isolation and the fuel integrity itself. The provisions in the NUREG for Waste Gas Systems are for PWRs and not applicable to BFN. guantities of radioactivity contained in all outdoor liquid radwaste tanks meeting the conditions of NUREG 5.5.9.c are determined in accordance with the specified surveillance pr ogr am. The sentence in the introductory paragraph is not needed to specify a method to determine liquid radwaste quantities.
P11 NUREG Specification 5.5.9 was modified to reflect the plant specific requirements of CTS 4.9.A.l.e. The requirements for testing new fuel oil priot to addition to the 7-day tanks has not been included. At BFN, new fuel oil is not added to the 7 day tanks until it is determined the fuel oil meets applicable ASTH standards. This provides assurance that the addition will have no effect on the fuel quality in the 7-day tanks.
DG operability is based on surveillance requirements imposed on the 7-day tanks. Also, wording changes were made to clarify that the sampling and testing, of a and b apply to 7-day fuel oil storage tanks only.
P12 The PTLR concept will not be used at BFN since an NRC approved methodology does not exist for BFN.
P13 The High Radiation Area Specification has been changed to be consistent with and reflect revisions to 10 CFR 20. These changes are consistent with the guidance of Regulatory Guide 8.38, "Control of Access to High and Very High Radiation Areas in Nuclear Power Plants." Proposed changes incorporate current licensing bases requirements that were recently changed (BFN TS 335, Amendment No. 201, 220, and 174 for Units 1, 2, and 3 respectively issued December 2, 1993) to incorporate the corresponding revised 10 CFR 20 terminology Revision 1
JUSTIFICATION FOR CHANGES TO NUREG-1433 SECTION 5.0 - ADMINISTRATIVE CONTROLS P14 BFN TS 364 incorporates the new 10 CFR 50 Appendix J, Option B requirements for containment leakage testing requirements. TS 364 was approved by Amendment 228, 243, and 203 to Unit 1, 2, and 3 Technical Specifications, respectively. Therefore, a Primary Containment Leakage Rate Testing Program (5.5.12) has been added to Programs and Hanuals (5.5) section of the proposed BFN ISTS.
P15 Current Technical Specifications for testing of fuel oil in storage tanks is conducted on a 92 day interval. This current interval has been found to be acceptable for use at BFN and is being retained in the proposed ISTS.
P16 The proposed change will revise the requirement for the Operations Hanager to hold a Senior Reactor Operator (SRO) license. The change will instead require the Operations Superintendent to hold a current SRO license on a BFN unit. At BFN, shift personnel report to the Shift Hanagers, who are required to be licensed as SROs for BFN in accordance with 10 CFR50.54(m)(2), and who in turn report directly to the Operations Superintendent.
The last paragraph of NUREG Specification 5.6.2, Annual Radiological Environmental Operating Report, has been omitted. The omitted paragraph requires additional information be submitted beyond what is currently required by the BFN licensing basis. BFN provides material consistent with the material outlined in the ODCH and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C as required by CTS 6.9.1.5.
P18 Supports do not lie within the scope of the inser vice testing program required by 10 CFR 50.55a(f). Therefore, the proposed change, deleting
...including applicable supports..." is appropriate.
GENERIC CHANGES Gl This change implements TSTF-118, Administr ative Contr ols Program I
I Exceptions.
I I
G2 This change implements TSTF-152, Revise Reporting Requirements to be I
I Consistent With 10 CFR 20.
Revision 1
0 BROWNS FERRY NUCLEAR PLANT - IMPROVED TECHNICAL SPECIFICATIONS SECTION 5.0 LIST OF REVISED PAGES NO SIGNIFICANT HAZARDS CONSIDERATIONS (Revised pages marked Revision I)
Added the following No Significant Hazards Considerations:
ITS 5.2, Organization, Ll ITS 5.2, Organization, L2 ITS 5.2, Organization, L3 ITS 5.2, Organization, L4 ITS 5.5, Programs and Manuals, L2 ITS 5.5, Programs and Manuals, L3 ITS 5.5, Programs and Manuals, L4 ITS 5.5, Programs and Manuals, LS ITS 5.6, Reporting Requirements, L2 ITS 5.6, Reporting Requirements, L3
NO SIGNIFICANT HAZiNDS CONSIDERATIONS BFN ISTS 5.2 - ORGANIZATION TECHNICAL CHANGES - LESS RESTRICTIVE Ll TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92.
s
~~1. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.~~
This change proposes to reduce the minimum required staffing level for non-licensed operators by one for the following conditions: when all three units are shutdown or defueled, when one unit is operating, and when two units are operating. This change will not significantly alter assumptions relative to the mitigation of an accident or transient event. The level of manning does not affect the probability of an accident. Therefore, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
20 The ro osed amendment does not create the ossibilit of a new or different kind of accident from an reviousl evaluated.
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing plant operation. Therefore, the proposed change does not create the possibility of a new or, different kind of accident from any previously evaluated.
~~3. The ro osed amendment does not involve a si nificant reduction in a ttiar in of safet .~~
This change does not involve a significant reduction in a margin of safety since the proposed change will continue to ensure that non-licensed operator manning levels will be adequate.
Page 1 of 12 Revision 1
0 TECHNIC GES -
NO SIGNIFICANT HAZARDS CONSIDERATIONS BFN ISTS 5.2 << ORGANIZATION ESS REST IC VE L2
~ g TVA has concluded that operation of Br owns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(l), of the three standards set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.~~
This change proposes to delete the CTS requirement for increased licensed operator staffing during cold startups, plant shutdowns, and recovery from trips. This change will not significantly alter assumptions relative to the mitigation of an'ccident or transient event. The level of manning does not affect the probability of an accident. Therefore, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
2~ The ro osed amendment does not create the ossibilit of a new or different kind of accident from an reviousl evaluated.
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3~ The ro osed amendment does not involve a si nificant reduction in a mar in of safet .
This change does not involve a significant reduction in a margin of safety since the proposed change will continue to ensure that the licensed operator manning requirements set forth in 10 CFR 50.54 (k),
(1), and (m) are met.
BFN-UNITS 1, 2, 5 3 Page 2 of 12 Revision 1
0 NO SIGNIFICANT HAZARDS CONSIDERATIONS BFN ISTS 5.2 - ORGANIZATION TECHNICAL CHANGES - LESS RESTRICTIVE L3 TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92.:...,
E ~ ~ kI
~~1. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated..~~
This change proposes to allow the shift crew composition to be less than the minimum requirement of 10 CFR 50.54(m)(2)(i), ITS 5.2.2.a and ITS 5.2.2.g for a period of time not to exceed two hours. This two hours is allowed to accommodate unexpected absence of on-duty shift crew members.
The proposed change does not affect the probability of an accident. The temporary reduction in the shift crew composition is for a short defined timeframe and does not affect the staffing requirements for licensed personnel in the control room contained in ITS 5.2.2.b (1 licensed RO when fuel is in the reactor and 1 licensed SRO in Nodes 1, 2, and 3).
The probability is small of an accident occur ring during the time the on shift crew composition is, reduced. This change will not significantly alter assumptions relative to the mitigation of an accident or transient event. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
20 The ro osed amendment does not create the ossibilit of a new or different kind of accident from an reviousl evaluated.
The proposed change does not involve a physical alteration of the plant
('no new or different type of equipment will be installed) or changes in methods governing plant operation. Therefore, the proposed change does create the possibility of a new or different kind of accident from 'ot any previously evaluated.
3~ The ro osed amendment does not involve a si nificant reduction in a mar in of safet .
This change does not involve a significant reduction in a margin of safety since the proposed change will continue to ensure that the licensed operator manning requirements set forth in 10 CFR 50.54 (k),
(1), and (m) are met.
BFN-UNITS 1, 2, 8L 3 Page 3 of 12 Revision 1
NO SIGNIFICANT HAZARDS CONSIDERATIONS BFN ISTS 5.2 - ORGANIZATION TECHNICAL CHANGES - LESS RESTRICTIVE L4 TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.~~
This change proposes to delete the requirement that does not permit any shift crew position to be unmanned upon shift change. This change will not significantly alter assumptions relative to the mitigation of an accident or transient event. The level of manning does not affect the probability of an accident. Therefore, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
2~ The ro osed amendment does not create the ossibilit of a new or different kind of accident from an reviousl evaluated.
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
~~3. The ro osed amendment does not involve a si nificant reduction in a mar in of safet .~~
This change does not involve a significant reduction in a margin of
. safety since the proposed change will continue to ensure that the licensed operator manning requirements set forth in 10 CFR 50.54 (k),
(1), and (m) are met.
Page 4 of 12 Revision 1
NO SIGNIFICANT HAZARDS CONSIDERATIONS BFN ISTS 5.5 - PROGRANS AND NNUALS TECHNICAL CHANGES - LESS RESTRICTIVE
~L1 TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a significant hazards consider ation. TVA's conclusion is based on its in accordance with 10 CFR 50.91 (a)(1), of the three standards set
'valuation, forth in 10 CFR 50.92.
~~1. The ro osed amendment= does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.~~
The Radioactive Effluents Control Program is contained in the ODCH and is implemented by plant procedures. The term "operability" is a Technical Specifications defined term and can be confusing when used for programs located outside the Technical Specifications. The use of the term "functional capability" is a more accUrate term. Functional capability means that the equipment can perform its intended function in the manner called for by the plant procedure. The proposed change will maintain the function of necessary equipment in order to implement the Radioactive Effluents Control Program. Therefore, the proposed change will not increase the probability or consequences of any accident previously evaluated.
~~2. The ro osed amendment does not create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.~~
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change will not impose or eliminate any new or different requirements. Thus, this change does not create the possibility of a new or different kind of accident from any previously evaluated.
~~3. The ro osed amendment does not involve a si nificant reduction in a mar in of safet .~~
This change does not involve a significant reduction in a margin of safety since the proposed change will continue to ensure that instrumentation and systems are functionally capable of performing radioactive environmental monitoring in accordance with Technical Specifications requirements. Technical Specifications programmatic requirements on these instruments and systems ensure that surveillance tests and setpoint determinations are performed.
Page 5 of 12 Revision 1
NO SIGNIFICANT HAZlNDS CONSIDERATIONS BFN ISTS 5.5 - PROGRNS AND MANUALS TECHNICAL CHANGES - ESS ESTRICTI V L2 TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a signi'ficant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.~~
This change proposes to delete the .requirement for air distribution testing specified in CTS 4.7.B:1.c. Eliminating testing is not an initiator of any analyzed accident. Therefore, the proposed change does not affect the probability of an accident previously evaluated. The proposed change follows the recommendations of ANSI N510-1975, which only requires this testing to be done on ini'tial installation.
Therefore, this change does not involve a significant increase in the consequences of an accident previously evaluated.
20 The ro osed amendment does not create the ossibilit of a new or different kind of accident from an reviousl evaluated.
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing plant'operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
The ro osed amendment does not involve a si nificant reduction in a mar in of safet .
This change does not involve a significant reduction in a margin of s'afety since the proposed change does not affect system or personnel response to an accident.
Page 6 of 12 Revision 1
NO SIGNIFICANT HAZARDS CONSIDERATIONS BFN ISTS 5.5 - PROGRAMS AND MANUALS TECHNICAL CHANGES - LESS RESTRICTIVE L3 TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences .of an accident revious1 evaluated.~~
This change proposes to increase the standby gas treatment system test interval for pressure drop testing and inlet heater testing. The change to these testing frequencies is not an initiator of any analyzed accident. Therefore, the proposed change does not affect the probability of an accident previously evaluated. The proposed change follows the recommendations of ASME N510-1989, for establishing test frequencies for this system., The proposed change to the test intervals from once per year to once every 18 months will maintain system efficiency, based on test history. Therefore, this change does not involve a significant increase in the consequences of an accident previously evaluated.
2~ The ro osed amendment does not create the ossibilit of a new or different kind of accident from an reviousl evaluated.
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
~~3. The ro osed amendment does not involve a si nificant reduction in a mar in of safet .~~
This change does not involve a significant reduction in a margin of safety since the proposed change does not affect system or personnel response to an accident.
Page 7 of 12 Revision 1
NO SIGNIFICANT HAZARDS CONSIDERATIONS BFN ISTS 5.5 - PROGRAMS AND MANUALS TECHNICAL CHANGES - LESS RESTRICTIVE L4 TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92.
~~l. The ro osed amendment does not invo1ve a si nificant increase i the robabilit or conse uences of an accident reviousl evaluated.~~
This change proposes to increase'the allowable pressure drop for, the standby gas treatment system from 6 inches of water to 7 inches of water to account for the inclusion of the prefilter into the test boundaries.
The change to this testing requirement is not an initiator of any analyzed accident. Therefore, the proposed change does not affect the probability of an accident previously evaluated. Therefore, this change
. does not involve a significant increase in the consequences of an accident previously evaluated.
20 The ro osed amendment does not create the ossibilit of a new or di e ent kind of accident rom an reviousl evaluated.
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3~ The ro osed amendment does not involve a si nificant reduction in a mar in of safet .
This change does not involve a significant reduction in a margin of s'afety since the proposed change does not affect system or personnel response to an accident.
Page 8 of 12 Revision 1
NO SIGNIFICANT HAZARDS CONSIDERATIONS .
BFN ISTS 5.5 - PROGRANS AND HANUALS TECHNICAL CHANGES - LESS RESTRICTIVE L5 TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92.
I. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.
This 'proposed change will change'he test fr'equencies for the standby gas treatment system and the control room emergency ventilation system.
The requirement to test after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of operation will no longer be applied to the in-place testing of the HEPA filter or the charcoal adsorber; The CTS requirement to perform laboratory testing of the charcoal after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation is included in the proposed ITS 5.5.7.c. The change to these testing frequencies is not an initiator of any analyzed accident. Therefore, the proposed change does not affect the probability of an accident previously evaluated. The proposed change to the test intervals will not impair system efficiency.
Therefore, this change does not involve a significant increase in the consequences of an accident previously evaluated.
~~2. The ro osed amendment does not create the ossibilit of a new or different kind of accident from an reviousl evaluated.~~
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3~ The ro osed amendment does not involve a si nificant reduction in a mar in of safet .
This change does not involve a significant reduction in a margin of safety'ince the proposed change does not affect system or personnel response to an accident.
Page 9 of 12 Revision 1
NO SIGNIFICANT HAZARDS CONSIDERATIONS BFN ISTS 5.6 - REPORTING REgVIREHENTS TECHNICAL CHANGES - LESS RESTRICTIVE
~L1 TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(l), of the three standards set forth in 10 CFR 50.92.
The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.-
4 The proposed change relaxes the time allowed to submit a Special Report after the inoperability of the primary containment radiation monitor "
from within 7 days to within 14 days. This change will not result in operation that will increase the probability of initiating an analyzed event since the time frame for submitting an Special Report is not assumed in the initiation of any analyzed event. This change only affects the time frame for submitting the report after an equipment inoperability. This change will not alter assumptions relative to mitigation of an accident or transient event. This change will not alter the operation of process variables, structures, systems, or components as described In the safety analyses. Therefor e this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
20 The ro osed amendment does not create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change will not impose or eliminate any new or different requirements. Thus, this change does not create the possibility of a new or different kind of accident from any previously evaluated.
~~3. The ro osed amendment does not involve a si nificant reduction in a mar in of safet .~~
This change proposes to relax the time required for submittal of the Special Report following a period of inoperability of the primary containment radiation monitor from 7 to 14 days. Increasing the time for submitting a report does not affect the margin of safety since this change will not impact any safety analysis assumptions. As such, no question of safety is involved. Therefore, this change does not involve a significant reduction in a margin of safety.
Page 10 of 12 Revision 1
Il NO SIGNIFICANT HAZARDS CONSIDERATIONS BFN ISTS 5.6 - REPORTING REgUIRENENTS TECHNICAL CHANGES - LESS RESTRICTIVE L2 TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not nvolve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.~~
This change proposes to delete the requirement to report main steam relief valve operation annually. The submittal of any report does not affect the probability of an accident. Therefore, the proposed change will not increase the probability or consequences of any accident previously evaluated.
~~2. The ro osed amendment does not create the ossibilit of a new or different kind of accident from an reviousl evaluated.~~
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from 0 3~
any previously evaluated.
The mar ro osed amendment in of safet .
does not involve a si nificant reduction in a This change does not involve a significant reduction in a margin of safety since the proposed change does not affect system or personnel response to an accident.
Page ll of 12 Revision 1
NO SIGNIFICANT HAZARDS CONSIDERATIONS BFN ISTS 5.6 - REPORTING RE(UIRENENTS TECHNICAL CHANGES - LESS RESTRIC IVE L3 TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to technical specifications does not involve a significant hazards consideration. TVA's conclusion is based on its .
evaluation, in accordance with 10 CFR 50.91 (a)(l), of the three standards set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not i volve a si nificant increase the robabilit or conse uences of an accident reviousl evaluated.~~
This change proposes to delete the requirement to report fatigue usage annually. The submittal of any report does not affect the probability of an accident. Therefore, the proposed change will not increase the probability or consequences of any accident previously evaluated.
~~2. The ro osed amendment does not create the ossibilit of a new or different kind of accident from an rev ousl evaluated.~~
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from
'ny previously evaluated.
The ro osed amendment does not involve a si nificant reduction in a mar in of safet .
This change does not involve a significant reduction in a margin of safety since the proposed change does not affect system or personnel response to an accident.
Page 12 of 12 Revision 1
0 GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS ADMINiSTRATIVE CHANGES "Ax" Labeled Comments Discussions TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance wi.th the proposed change to the Technical Specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(l), of the three standards set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.~~
The proposed change involves reformatting, renumbering, and rewording of the existing Technical Specifications. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. As such, this 'change is administrative in nature and does not impact initiators of analyzed events. Therefore,'his change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
20 The ro osed amendment does not create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change will not impose or eliminate any new or different requirements. Thus, this change does not create the possibility of a new or different kind of accident from any previously evaluated.
~~3. The ro osed amendment does not involve a si nificant reduction in a mar in of safet .~~
The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumptions. This change is administrative in nature. As such, no question of safety is involved, and the change does not involve a significant reduction in a margin of safety.
BFN-UNITS 1, 2, 5, 3 Page I OF ll Revision 0
GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS TECHNICAL CHANGES - NORE RESTRICTIVE "Hx" Labe1ed Comments Discussions TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to the Technical Specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(l), of the three standards set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl eva1uated.~~
The proposed change provides more stringent requirements for operation of the facility. These more stringent requirements do not result in operation that will increase the probability of initiating an analyzed event and do not alter assumptions relati've to mitigation of an accident or transient event. The more restrictive requirements continue to ensure process variables, structures, systems and components are maintained consistent with the safety analyses and licensing basis.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
~~2. The ro osed amendment does not create the ossibilit of a new or different kind of accident from an accident revious1 evaluated.~~
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change does impose different requirements. However, these changes are consistent with the assumptions made in the safety analysis and licensing basis.
~
Thus, this change does not create the possibility of a new or different kind of accident from any previously evaluated.
~~3. The ro osed amendment does not involve a si nificant reduction in a mar in of safet .~~
The imposition of more restrictive requirements either has no impact on or increases the margin of safety. As provided in the justification for the change, each change in this category is by definition providing additional restrictions to enhance plant safety. The change maintains requirements within safety analyses and licensing basis. Therefore, this change does not involve a significant reduction in a margin of safety.
BFN-UNITS l, 2, 5 3 Page 2 OF ll Revision 0
GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS "GENERIC" LESS RESTRICTIVE CHANGES "LAx" Labeled Comments Discussions TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to the Technical Specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not involve a si nificant increase in t e robabilit or conse uences of an accident reviousl evaluated.~~
The proposed change relocates certain details from the Technical Specifications to the Bases, FSAR or procedures. The Bases, FSAR and procedures containing the relocated information will be maintained in accordance with 10 CFR 50.59. In additioh to 10 CFR 50.59 provisions, the Technical Specification Bases are subject to the change control provisions in the Administrative Controls section of the Technical Specifications. The FSAR is subject to the change control provisions of 10 CFR 50.71(e), arid plant procedures are subject to controls imposed by plant administrative procedures, which endorse applicable regulations and standards. Since any changes to the Bases, FSAR or procedures will be evaluated per the requirements of 10 CFR 50.59, no increase (significant or insignificant) in the probability or consequences of an accident previously evaluated will be allowed. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
20 The ro osed amendment does not create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.
The proposed change does.not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change will not impose any new or different requirements and adequate control of the information will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any previously evaluated.
3~ The ro osed amendment does not involve a si nificant reduction in a mar in of safet .
The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumptions. In addition, the requirements to be relocated from the Technical Specifications to the Bases, FSAR or. procedures are the same as the existing Technical Specifications. Since any future changes to these requirements in the Bases, FSAR or procedures will be evaluated per the requirements of 10 BFN-UNITS 1, 2, 5 3 Page 3 OF ll Revision 0
GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS "GENERIC" LESS RESTRICTIVE CHANGES (Continued)
"LAx" Labeled Comments Discussions CFR 50.59, no reduction (significant or insignificant) in a margin of safety will be allowed. The existing requirement for NRC review and approval of revisions, in accordance with 10 CFR 50.92, to these details proposed for relocation, does not have a specific margin of safety upon which to evaluate. However, since the proposed change is consistent with the BWR4 Standard Technical Specifications, approved by the NRC Staff (NUREG-1433), revising the Technical Specifications to reflect the approved level of detail ensures no significant reduction in the margin of safety.
BFN-UNITS 1, 2, 5 3 Page 4 OF 11 Revision 0
GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS "GENERIC" LESS RESTRICTIVE CHANGES "LBx" Labeled Comments Discussions TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to the Technical Specifications does not involve a significant hazards consideration. TYA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards. set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl ev'aluated.~~
The proposed changes increase the Surveillance Test Intervals (STIs) and Allowed Out-of-Service Times (AOTs) for instrumentation supporting a number of TS functions. There are no modifications to any of the affected systems. However, the changes are expected to reduce the test-related plant scrams and test-induced wear on the equipment. Therefore, there is no significant increase in the probability or occurrence of an accident previously evaluated.
General Electric Topical Reports NED0-30851-P-A, NED0-30936-P-A, NEDO-30S51-P-A, NED0-31677-P-A, GENE-770-06-1, and GENE-770-06-2 show that the effects of these extensions of STIs and AOTs are negligible and are bounded by previous analyses. Furthermore, the NRC has reviewed these reports and approved conclusions on a generic basis. Therefore, the change does not involve a significant increase in the consequences of an accident previously evaluated.
~~2. The ro osed amendment does not create the ossibil it of a new or different kind of accident from an accident reviousl evaluated.~~
The design and operational, function of the affected equipment are not changed by the proposed revisions. The proposed changes affect only the STIs and AOTs and will not impact the function of monitoring system variables over the anticipated ranges for normal operation, anticipated operational occurrences, or accident conditions. Furthermore, the proposed changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed), make changes in methods governing normal plant operation, make any physical modifications, or alter any operational setpoints. Therefore, the possibility of a new or different kind of accident from any previously evaluated is not created.
BFN-UNITS 1, 2, 5 3 Page 5 OF 11 Revision 0
GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS "GENERIC" LESS RESTRICTIVE CHANGES (Continued)
"LBx" Labeled Comments Discussions
~~3. The ro osed amendment does not involve a si nificant reduction in a mar in of safet .~~
The proposed changes do not alter the manner in which Safety Limits, Limiting Safety System Settings, or Limiting Conditions for Operation are determined. Reduced testing, other than as addressed above, allows a longer time interval over which instrument uncertainties (e.g., drift) may act. The current affected instrumentation setpoints account for the effects of drift and include a sufficient allowance to tolerate extensions of the STIs. Implementation of the proposed changes is expected to result in an overall improvement in safety for two reasons.
First, reduced testing will result in fewer inadvertent reactor trips, less frequent actuation of ESF components, and greater equipment availability. Second, reduced testing will result in less distractions of the operating staff from monitoring and controlling plant operations, thereby increasing the effectiveness of the operating staff. Therefore, the proposed changes do not significantly reduce the margin of safety.
BFN-UNITS 1, 2, 5 3 Page 6 OF 11 Revision 0
GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS "GENERIC" LESS RESTRICTIVE CHANGES "LCx" Labeled Comments Discussions TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to the Technical Specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(l), of the three standards set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not involve a si nificant incr ease in the robabilit or conse uences of an accident reviousl evaluated.~~
The proposed change relocates instrumentation requirements, which provide no post-accident function, from the Technical Specifications to the Bases, FSAR, or procedures. These requirements are part of routine operational monitoring and are not considered in the safety analysis.
The Bases, FSAR and procedures containing the relocated information will be maintained in accordance with 10 CFR 50.59. In addition to 10 CFR 50.59 provisions, the Technical Specification Bases are subject to the change control provisions in the Administrative Controls section of the Technical Specifications. The FSAR is subject to the change control provisions of 10 CFR 50.71(e), and plant procedures are subject to controls imposed by plant administrative procedures, which endorse applicable regulations and standards. Since any changes to the Bases, FSAR or procedures will be evaluated per the requirements of 10 CFR 50.59, no increase (significant or insignificant) in the probability or consequences of an accident previously evaluated will be allowed.
Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
~~2. 'he ro osed amendment does not create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.~~
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change will not impose any new or different requirements and adequate control of the information will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any previously evaluated.
BFN-UNITS 1, 2, 5 3 Page 7 OF 11 Revision 0
GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS "GENERIC" LESS RESTRICTIVE CHANGES (Continued)
"LCx" Labeled Comments Discussions
~~3. The ro osed amendment does not involve a si nificant reduction in a mar in of safet .~~
The proposed change will not reduce a margin of safety because no safety analysis assumptions are affected. In addition, the requirements to be relocated from the Technical Specifications to the Bases, FSAR or procedures are the same as the existing Technical Specifications. Since any future changes to these requirements in the Bases, FSAR or procedures will be evaluated per the requirements of 10 CFR 50.59, no r eduction (significant or insignificant) in a margin of safety will be allowed.
The existing requirement for NRC review and approval of revisions, in accordance with 10 CFR 50.92, to these details proposed for relocation, does not have a specific margin of safety upon which to evaluate.
However, since the proposed change is consistent with the BWR4 Standard Technical Specifications, approved by the NRC Staff (NUREG-1433),
revising the Technical Specifications to reflect the approved level of detail ensures no significant reduction in the margin of safety.
BFN-UNITS 1, 2, 5 3 Page 8 OF 11 Revision 0
GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS TECHNICAL CHANGES - RELOCATIONS "Rx" Labelled Comments Discussions TVA has concluded that operation of Browns Ferry Nuclear Plant in accordance with the proposed change to the Technical Specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the thr ee standards set forth in 10 CFR 50.92.
~~1. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.~~
This proposed change relocates requirements from the Technical Specifications to a licensee controlled document. The licensee controlled documents containing the relocated requirements will be maintained using the provisions of 10 CFR 50.59. Since any changes to a licensee controlled document will be evaluated per 10 CFR 50.59, no increase (significant or insignificant) in the probability or consequences of an accident previously evaluated will be allowed.
Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
~~2. The ro osed amendment does not create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.~~
This change relocates requirements to a licensee controlled document.
This change will not alter the plant configuration (no new or different type of equipment will be installed) or change any methods governing normal plant operation. This change will not impose different requirements and adequate control of information will be maintained.
'his change will not alter assumptions made in the safety analysis and
~
licensing basis. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
BFN-UNITS I, 2, & 3 Page 9 OF ll Revision 0
GENERIC NO SIGNIFICANT NZARDS CONSIDERATIONS TECHNICAL CHANGES - RELOCATIONS "Rx" Labelled Comments iscussions (continued)
~~3. The ro osed amendment does not involve a si nificant reduction in a mar in of safet .~~
This change relocates requirements from the Technical Specifications to a licensee controlled document. This change will not reduce a margin of safety since it has no impact on any safety analysis assumptions. In addition, the requirements to be relocated from the Technical Specifications to the licensee controlled document are the same as the existing Technical Specifications. Since any future changes to this licensee controlled document will be evaluated per the requirements of 10 CFR 50.59, no reduction (significant or insignificant) in a margin of safety will be allowed. Therefore, this change will not involve a significant reduction in a margin of safety.
The existing requirement in 10 CFR 50.90 for NRC review and approval of revisions to these details and requirements proposed for relocation does not have a specific margin of safety upon which to evaluate. However, since the proposed change is consistent with the BWR4 Standard Technical Specifications approved by the NRC Staff (NUREG-1433) and the change controls for proposed relocated details and requirements provide an equivalent level of regulatory authority, revising the Technical Specifications to reflect the approved level of detail and requirements ensures no reduction in the margin of safety.
BFN-UNITS I, 2, & 3 Page 10 OF ll Revision 0
GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS ENVIRONMENTAL IMPACT CONSIDERATION The proposed change does not involve a significant hazards consideration, a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Accordingly, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.
BFN-UNITS 1, 2, 5 3 Page 11 OF 11 Revision 0
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BFN UNIT 1 2 AND 3 CROSS-REFERENCE MATRIX RELOCATED RELOCATED RELOCATED CTS NUMBER [*] BFN ITS NUMBER NUREG NUMBER DELETED TO BASES TO TRM " TO PROC RELOCATED CONTROL 1.0.BB 5.5.1.a 5.5.1.a 1.0.BB 5.5.1.b 5.5.1.b 1.0.MM 5.5.6 5.5.7 1.0.MM.1 NONE NONE YES 1.0.MM.2 5.5.6.a 5.5.7.a 1.0.MM.3 5.5.6.b 5.5.7.b 1.0.MM.4 NONE NONE YES 1.0.MM.5 5.5.6.d 5.5.7.d 1.0.MM.6 NONE NONE YES 3.2.K NONE NONE YES 10 CFR 50.59 3.2.K Table NONE NONE YES 10 CFR 50.59 3.2.K Table Notes NONE NONE YES 10 CFR 50.59 3.7.B.2.a 5.5.7.a 5.5.8.a 3.7.B.2.a 5.5.7.b 5.5.8.b 3.7.B.2.b 5.5.7.c 5.5.8.c 3.7.B.2.c 5.5.7.a 5.5.8.a 3.7.B.2.c 5.5.7.b 5.5.8.b 3.7.B.2.c 5.5.7.d 5.5.8.d 3.7.E.2.a 5.5.7.a 5.5.8.a 3.7.E.2.a 5.5.7.b 5.5.8.b 3.7.E.2.b 5.5.7.c 5.5.8.c 3.8.A.5 5.5.8 5.5.9 3.8.A.6 NONE NONE YES 10 CFR 50.59 3.8.B.9 5.5.8.a 5.5.9.a YES 10 CFR 50.59 3.8.B.10 NONE NONE YES 10 CFR 50.59 4.2.K NONE NONE YES 10 CFR 50.59 4.2.K Table NONE NONE YES 10 CFR 50.59 4.2.K Table Notes NONE NONE YES 10 CFR 50.59 4.7.B.1 NONE NONE YES 4.7.B.1.a 5.5.7.d 5.5.8.d 4.7.B.1.b 5.5.7.e 5.5.8.e 4.7.B.1.c NONE NONE YES 4.7.B.2.a 5.5.7 5.5.8 4.7.B.2.b 5.5.7.a 5.5.8.a 1 of4 Revision 0
BFN UNIT 1 2 AND 3 CROSS-REFERENCE MATRIX RELOCATED RELOCATED RELOCATED CTS NUMBER [*] BFN ITS NUMBER NUREG NUMBER DELETED TO BASES TO TRM TO PROC RELOCATED CONTROL 4.7.B.2.c 5.5.7.b 5.5.8.b 4.7.E.1 5.5.7.d 5.5.8.d 4.7.E.2.a 5.5.7 5.5.8 4.7.E.2.b 5.5.7.a 5.5.8.a 4.7.E.2.c 5.5.7.b 5.5.8.b 4.8.A.6 NONE NONE YES 10 CFR 50.59 4.8.B.5 5.5.8.a 5.5.9.a YES 10 CFR 50.59 4.9.A.1.e 5.5.9.a 5.5.10.a 6.1.1 5.1.1 5.1.1 6.1.2 5.1.2 5.1.2 6.2.1 5.2.1 5.2.1 6.2.1.a 5.2.1.a 5.2.1.a 6.2.1.b 5.2.1.c 5.2.1.c 6.2.1.c 5.2.1.b 5.2.1.b 6.2.1.d 5.2.1.d 5.2.1.d 6.2.2.a NONE NONE YES 6.2.2.b NONE NONE YES 6.2.2.c 5.2.2.b 5.2.2.b 6.2.2.d 5.2.2.b 5.2.2.b YES 6.2.2.e 5.2.2.d 5.2.2.d 6.2.A Table 5.2.2.a 5.2.2.a YES 6.2.A Table 5.2.2.g 5.2.2.g 6.2.A Table Note a NONE YES 6.2.A Table Note b 5.2.2.c 5.2.2.c YES 6.2.A Table Note c NONE NONE YES 6.2.A Table Note d NONE NONE YES 6.3 5.3.1 5.3.1 6.8.1.1 5.4.1 5.4.1 6.8.1.1.a 5.4.1.a 5.4.1.a 6.8.1.1.b 5.2.2.e 5.2.2.e 6.8.1.1.c NONE NONE YES 6.8.1.1.f 5.4.1.d 5.4.1.d 6.8.1.1.i NONE NONE YES 6.8.1.1.j NONE NONE YES 6.8.2 NONE NONE YES 2of4 Revision 0
0 BFN UNIT 1 2 AND 3 CROSS-REFERENCE MATRIX RELOCATED RELOCATED RELOCATED CTS NUMBER ["] BFN ITS NUMBER NUREG NUMBER DELETED TO BASES TOTRM TO PROC RELOCATED CONTROL 6.8.3 NONE NONE YES 6.8.3.1 5.7.1 5.7.1 6.8.3.2 5.7.2 5.7.2 6.8.3.3 5.7.3 5.7.3 6.8.4.1 5.5.4 5.5.4 6.8.4.1.a 5.5.4.a 5.5.4.a 6.8.4.1.b 5.5.4.b 5.5.4.b 6.8.4.1.c 5.5.4.c 5.5.4.c 6.8.4.1.d 5.5.4.d 5.5.4.d 6.8.4.1.e 5.5.4.e 5.5.4.e 6.8.4.1.f 5.5.4.f 5.5.4.f 6.8.4.1.g 5.5.4.g 5.5.4.g 6.8.4.1.g.1 5.5.4.g.1 NONE 6.8.4.1.g.2 5.5.4.g.2 NONE 6.8.4.1.h 5.5.4.h 5.5.4.h 6.8.4.1.i 5.5.4.i 5.5.4.i 6.8.4.1.j 5.5.4 5.5.4.j 6.8.4.3 5.5.12 NONE 6.8.5 5.5.3 5.5.3 6.9.1 5.6 5.6 6.9.1.2.a 5.6.1 5.6.1 6.9.1.2.b NONE NONE YES 6.9.1.3 5.6.4 5.6.4 6.9.1.4 NONE YES 6.9.1.5 5.6.2 5.6.2 6.9.1.6 NONE NONE YES 10 CFR 50.59 6.9.1.7.a 5.6.5.a 5.6.5.a YES 10 CFR 50.59 6.9.1.7.b 5.6.5.b 5.6.5.b 6.9.1.7.c 5.6.5.c 5.6.5.c 6.9.1.7.d 5.6.5.d 5.6.5.d 6.9.1.8 5.6.3 5.6.3 6.9.2.1 NONE NONE YES 6.9.2.2 NONE YES 10 CFR 50.59 6.9.2.3 NONE YES 10 CFR 50.59 6.9.2.4 NONE NONE YES 10 CFR 50.59 3 of 4 Revision 0
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A
BFN UNIT 1 2 AND 3 CROSS-REFERENCE MATRIX RELOCATED RELOCATED RELOCATED CTS NUMBER t'] BFN ITS NUMBER NUREG NUMBER DELETED TO BASES TOTRM TO PROC RELOCATED CONTROL 6.9.2.6 NONE NONE YES 10 CFR 50.59 6.9.2.7 NONE NONE YES 10 CFR 50.59 6.9.2.9 5.6.6 5.6.8 6.9.2.10 NONE NONE YES 10 CFR 50.59 6.12.1 5.5.1.a 5.5.1.a 6.12.1.a 5.5.1.a.1 5.5.1.a.1 6.12.1.b 5.5.1.a.2 5.5.1.a.2 6.12.2 5.5.1.b 5.5.1.b 6.12.3 5.5.1.c 5.5.1.c NONE 5.2.2.a 5.2.2.a NONE 5.2.2.e 5.2.2.e 5.2.2.f 5.2.2.f NONE 5.2.2.g 5.2.2.g NONE 5.4.1.b 5.4.1.b NONE 5.4.1.c 5.4.1.c NONE 5.4.1.e 5.4.1.e NONE 5.5.2 5.5.2 NONE 5.5.5 5.5.5 NONE 5.5.6.c 5.5.7.c NONE 5.5.9 5.5.10 NONE 5.5.9.b 5.5.10.c NONE 5.5.10 5.5.11 NONE 5.5.11 5.5.12 NONE NONE 5.5.4.k NONE NONE 5.5.6 NONE NONE 5.5.9.b NONE NONE 5.6.6 NONE NONE 5.6.7 NONE NONE 5.5.10.b 4of4 Revision 0

ML18039A198

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