ML18033B419
| ML18033B419 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/03/1990 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18033B418 | List: |
| References | |
| RTR-NUREG-0737 GL-83-36, NUDOCS 9007100091 | |
| Download: ML18033B419 (27) | |
Text
ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION REVISIONS BROWNS FERRY NUCLEAR PLANT (BFN)UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 284)O9i 9OO703 gp PG CK pDC 9007 05000pgp P UHIT 2 EFFECTIVE PAGE LIST REMOVE IHSERT v vi 3.7/4.7-27 3.7/4.7-28 3.7/4.7-29 3.7/4.7-30 3.7/4.7-33 3.7/4.7-34 6.0-23 6.0-24 6.0-25 6.0-26'3.7/4.7-27 3.7/4.7-28*
3.7/4.7-29*
3.7/4.7 30 3.7/4.7-33 3.7/4.7-34*
6.0-23 6'-24 6.0-25*6.0-26**Denotes overleaf or spillover page.
DMINISTRATIVE CONTROLS SECTIO PAGE ES 0 S~~~I~~~~~~~~~~~~~~~~~~~~~~~~~~~I~~~~~~~~~~~6.0-1 ORGA ZATIO....~~.........................,............
~6.0-1 6.2.1 6.2.2 lant Staff~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~P PLANT STAFF UALIFICATIOHS...............................
TRAINX Go~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~o~~~~~~~~~~P ANT REV EW AHD AUDIT~~~~~~~~~.~~~~~~~.~~~~~~~~~~~~~~~~~6.0-2 6.0-5 6.0-5 6.0-5 Offsite and Onsite Organizations........................'.
6.0 1 6.5.1 6.5.2 Plant Operations Review Committee (PORC).................
6,0-5 Nuclear Safety Review Board (HSRB)......................, 6,0-11 6.5.3Technical Review and Approval of Procedures..............
6.0-17 6.8.1 6.8.2 6.8.3 6.8.4 REPORTABLE ACTIO S~~~~~o~~~~o~~~~~~~~~~~~~~~~~~~~~SAFETY LI VIOLATIOHo
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~PROCED S HS RUCTIO S A PROGRAMS...........
Procedures....;...........................................
rills~~~~~~~~.~~~~~~~~~~~~o~a~~~-~~~~~~~~~~~~~~~~~~e~~~~~~D 6.0-18 6.0-19 6.0-20 6.0-20 6.0-21 Quality Assurance Procedures
-Effluent and Environmental Monitoring..............................
6,0-23 Radiation Control Procedures.............................
6.0-22 6.8.5 6.9.1 rograms~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~e~~~~~~~~~~~~P REPORTING RE UIREMENTS........................,........
outine Reports...............................
R Startup Reports..........................................
Annual Operating Report.....;......................,.....
Monthly Operating Report.................................
Reportable Events...,...,...........,....
'.0-23 6.0-24 6.0-24 6.0-24 6.0-25 6;5-26 6.0-26 Radioactive Effluent Release Report......................
6.0-26 6.,9.2 6.10~611 ource Tests.............................................
S Special Reports..............................
6.0-26 6.0-27 STATIO OPERATING RECORDS AND RETENTION........,,...,....
6.0 29 PROCESS CONTROL PROGRAM..................................
6.0-32 OFFSITE DOSE CALCULATION MANUAL....
~..~~.............
~~~~6~0 32 6.13 RADIOLOGICAL EFFLUENT MANUAL.......................,.....
6.0-33 BFH Unit 2 v LIST OF TABLES Table Title~Pa e He.Surveillance Frequency Notation..........1.0-13 3.1.A Reactor Protection System (SCRAM)Instrumentation Requirements.
3.1/4.1-3 4.1.A Reactor Protection System (SCRAM)Instrumentation Functional Tests Minimum Functional Test Frequencies for Safety Instr.and Control Circuitse~~~~~~~~~~~~~~~~3.1/4.1-8 4.1.B Reactor Protection System (SCRAM)Instrumentation Calibration Minimum Calibration Frequencies for-Reactor Protection Instrument Channels.3.1/4.1-11 3.2.A Primary Containment and Reactor Building Isolation Instrumentation
~~~~~~~e~~~~~~~~~3.2/4.2-7 3.2.B Instrumentation that Initiates or Controls the Core and Containment Cooling Systems.3.2/4.2-14 3.2.C Instrumentation that Initiates Rod Blocks....3 2/4'-25 3.2.D 1 Radioactive Liquid Effluent Monitoring Instrumentation
.3.2/4.2-28 3.2.E Instrumentation that Monitors Leakage Into Drywell.3.2/4.2-30 3.2.F Surveillance Instrumentation.
3.2/4.2-31 3.2.G 3.2eH 3.2.I Control Room Isolation Instrumentation.
......3.2/4.2-34 Flood Protection Instrumentation.
.........3.2/4.2-35 Meteorological Monitoring Instrumentation
.....3.2/4.2-36 3.2eJ Seismic Monitoring Instrumentation.
........3.2/4.2-37 3.2eK Radioactive Gaseous Effluent Monitoring Instrumentation
.3 2/4A-38 3.2.L ATWS-Recirculation Pump Trip (RPT)Surveillance Instrumentation
.3.2/4.2-39b 4.2.A Surveillance Requirements for Primary Containment and Reactor Building Isolation Instrumentation.
.'.2/4.2-40 4.2.B Surveillance Requirements for Instrumentation that Initiate or Control the CSCS.3.2/4.2-44 4.2.C Surveillance Requirements for Instrumentation that Initiate Rod Blocks 3.2/4.2-50 4.2.D Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements
.3.2/4.2-51 BFN Unit 2 vi AMENDMEQ'gp y 69 TABLE 3.7.A (Continued)
Vlv Idnifi a i n Number of Power Operated Valves Inboard ggtbo~rd Haximum Operating Time~Action on Normal Ini ti ating~Pn i i n~anal~N Torus hydrogen sample line valves-Analyzer A (FSV-76-55, 56)Drywell hydrogen sample line valves-Analyzer A (FSV-76-49, 50)Sample return valves-Analyzer A (FSV-76-57, 58)Torus hydrogen sample line valves-Analyzer B (FSV-76-65, 66)Drywell hydrogen sample line valves-Analyzer B (FSV-76-59, 60)Sample return valves-Analyzer B (FSV-76-67, 68)PASS RHR Liquid Sample (FSV-43-50, 56)PASS Liquid/Gas Return to Torus (FSV-43-40, 42)N/A N/A N/A N/A N/A N/A N/A N/A 0/C 0/C 0/C 0/C GC/SC GC/SC GC GC/SC GC/SC GC SC SC 6, 1 6, 1 6, 1 6, 1 Ve~lv Iden ificd~ion TABLE 3.7.A (Continued)
Number of Power Operated Valves~Inbo rd O~bo rd Haximum Operating~Tim Normal 2QXU~n Action on Ini ti ating~in 1 Suppression chamber purge inlet (FCV-64-19)
Drywell/suppression chamber ni trogen make-up inlet (FCV-76-17)
Drywell exhaust valve bypass to standby gas treatment system (FCV-64-31)
Suppression chamber exhaust valve bypass to standby gas treatment system (FCV-64-34)
System suction isolation valves to air compressors"A" and"B" (FCV-32-62, 63)1 2.5 C 15 SC SC GC GC GC TIP guide tubes (5)(FCV-94-501, 502, 503, 504, 505)1 per guide tube N/A GC TABLE 3.7.A (Continued)
N/A N/A N/A N/A N/A N/A V~vdui~QS agog Standby liquid control system check valves (CV 63-526&525)Feedwater check valves (CV-3-558, 572, 554&568)Control rod hydraulic return check valves (CV-85-576)
RHRS-LPCI to reactor check valves (CV-74-54&68)CAD system torus/drywell exhaust.to standby gas treatment (FCV-84-19)
Drywell/suppression chamber ni trogen Purge Inlet (FCV-76-24)
Core spray discharge to reactor check valves (FCV-75-26, 54)Drywell dP air compressor suction valve (FCV-64-139)
Dryweil dP air compressor discharge valve (FCV-64-140)
Drywell CAH suction valves (FCV-90-254A and 2548)Drywell CAH discharge valves (FCV-90-257A and 2578)Drywell CAH suction valve (FCV-90-255)
CAD system torus/drywell exhaust to standby gas treatment (FCV-84-20)
Number of Power Operated Valves~I b~rd~b'g Haximum Operating Limmm~N/A N/A N/A N/A 10 10 10 10 10 10 Normal hmiiigB Action on Ini ti ating gi~n~N Process Process Process Process SC SC Process SC SC GC GC GC SC TABLE 3.7.A (Continued)
N/A N/A N/A N/A N/A N/A N/A N/A N/A V lv Identification Core spray discharge to reactor isolation valves (75-25, 75-53)PSC return line check valves (12-738,12-741)Suppression chamber sample RHR pumps A&C isolation valves (43-28A,43-288)Suppression chamber sample RHR pumps B&D isolation valves (43-29A,43-29B)PSC head tank tie-in to RHR check valves (74-803,74-804, 74-792,74-802)PSC head tank tie-in to CS check valves (75-606,75-609, 75-607,75-610)TIP nitrogen purge check valve (76-653)CAD crosstie to W control air check valve (84-617)CAD Crosstie to DW control air check valve (84-680)Number of Power Operated Valves Inboard guuboard Haximum Operating Normal~Time~.~Posi i n N/A C N/A N/A N/A N/A N/A N/A N/A N/A Action on Ini ti ating~inal~N SC 3 N/A SC SC Process 3 Process 3 Process Process Process TABLE 3'.A (Continued)
Valv Iden ifi i n Demineralized water supply check valve (2-1192)Number of Power Operated Valves~Tnb ard g~niniard Haximum Operating Normal~Tim~~~Pnai i n N/A C Action on Ini ti ating~in 1~oe N/A 1 Demineralized water supply isolation valve (2-1383)Service air supply isolation valves (33-1070)Service air supply check valve (33-785)Drywell control air inlet header check valves (32-2163,32-336)Drywell control air inlet header check valves (32-2516, 32-2521)Suppression chamber vacuum relief (64-20, 64-21)Suppression chamber vacuum relief check valves (64-800,64-801)Recirculation pump A seal injection check valves (68-508,68-550)Recirculation pump B seal injection check valves (68-523,68-555)Reactor water cleanup system discharge check valve (69-579)Reactor building closed cooling water drywell return isolation valve (70-47)N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Process Process N/A Process Process Process Process GC 1, 4 1, 4 1, 4 NOTES FOR TABLE A Key: 0=C=SC=GC=Open Closed Stays Closed Goes Closed Note: Isolation groupings are as follows: Group 1: The valves in Group 1 are actuated by any one of the following conditions:
1.Reactor Vessel Low Low Water Level (378")2.Main Steamline High Radiation 3.Main Steamline High Flow 4.Main Steamline Space High Temperature 5.Main Steamline Low Pressure Group 2: The valves in Group 2 are actuated by any of the following conditions:
1.Reactor Vessel Low Water Level (538")2.High Drywell Pressure Group 3: The valves in Group 3 are actuated by any of the following conditions:
1.Reactor Low Water Level (538")2.Reactor Water Cleanup (RWCU)System High Temperature in the main steam valve vault, 3.RWCU System High Temperature in RWCU pump room 2A, 4.RWCU System High Temperature in the RWCU pump room 2B, 5.RWCU System High Temperature in RWCU heat exchanger room, 6.RWCU System High Temperature in the space near the pipe trench containing RWCU piping.Group 4: The valves in Group 4 are actuated by any of the following conditions:
Group 5: 1.HPCI Steamline Space High Temperature 2.HPCI Steamline High Flow 3.HPCI Steamline Low Pressure 4.HPCI Turbine Exhaust Diaphragm High Pressure The valves in Group 5 are actuated by any of the following conditions:
1.RCIC Steamline Space High Temperature 2.RCIC Steamline High Flow 3.RCIC Steamline Low Pressure 4.RCIC Turbine Exhaust Diaphragm High Pressure BFN Unit 2 3.7/4.7-34 positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiological Work Permit.6.8.3.2 Each high radiation area in which the intensity of radiation is greater than 1,000 mrem/hr shall be subject to the provisions of (1)above;and, in addition, access to the source and/or area shall be secured by lock(s).The key(s)shall be under the administrative control of the shift engineer.In the case of a high radiation area established for a period of 30 days or less, direct surveillance to prevent unauthorized entry may be substituted for permanent access control.*Health Physics personnel, or personnel escorted by Health Physics personnel, in accordance with approved emergency procedures, shall be exempt from the RWP issuance requirement, during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection.
procedures for entry into high radiation areas.QUALITY ASSURANCE PROCEDURES
-EFFLUENT AND ENVIRONMENTAL MONITORING 6.8.4 Quality Assurance procedures shall be established, implemented, and maintained for effluent and environmental monitoring, using the guidance in Regulatory Guide 1.21, Rev.1, June 1974 and Regulatory Guide 4.1, Rev.1, April 1975 or Regulatory Guide 4.15, Dec.1977.6.8.5 PROGRAMS Postaccident Sam lin Postaccident sampling activities will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and BFN Unit 2 6.0 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 I 6.9 REPORT G RE UXREMEHTS ROUTIHE REPORTS 6.9.1 In 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 STARTUP REPORT a.A summary report of plant startup and power escalation testing shall be submitted following (1)receipt of an operating license, (2)amendment to the license involving a planned increase in power level, (3)installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4)modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.The report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics ebtained during the test program and a comparison of these values with design predictions and specifications.
Any corrective actions that were required to obtain satisfactory operation shall also be BFN Unit 2 6.0-24 6.9.1.1 STARTUP REPORT (Continued) described.
Any additional specific details required in license conditions based on other commitments shall be included in this report.b.Startup reports shall be submitted within (1)90 days following completion of the startup test program, (2)90 days following resumption or commencement of commercial power operation, or (3)9 months following initial criticality, whichever is earliest.If the Startup Report does not cover all three events (i.e.,'initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.
6.9.1.2 AHHUAL OPERATING REPORT*a.A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mremlyr and their associated man rem exposure according to work and job functions,**e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.
The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.
Small exposures totaling less than 20/of the individual total dose need not be accounted for.In the aggregate, at least 80K of the total whole body dose*A single submittal may be made for a multiple unit station.**This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.BFN Unit 2 6.0-25 received from external sources shall be assigned to specific major work functions.
b.Any mainsteam relief valve that opens in response to reaching its setpoint or due to operator action to control reactor pressure shall be reported.6.9.1.3 MONTHLY OPERATING REPORT Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Office of Inspection and Enforcement, U.S.Nuclear Regulatory Commission, Washington, D.C.20555, with a copy to the Regional Office, to be submitted no later than the fifteenth of each month following the calendar month covered by the report.A narrative summary of operating experience shall be submitted in the above schedule.6.9.1.4 REPORTABLE EVENTS-Reportable events, including corrective actions and measures to prevent re-occurrence, shall be reported to the NRC in accordance with Section 50.73 to 10 CFR 50.6.9.1.5 RADIOACTIVE EFFLUENT RELEASE REPORT Deleted (See REM section F-2)6.9.1.6 SOURCE TESTS, Results of required leak tests performed on sources if the tests reveal the presence of 0.005 microcurie or more of removable contamination.
BFN Unit 2 6.0-26
ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT (BFN)UNIT 2
SUMMARY
OF CHANGES Summa of Chan es 1~Revision to Table 3.7.A,"Primary Containment Isolation Valves".Proposed change to Table 3.7.A would add: Post-Accident Sampling System (PASS)liquid and gas return to torus valves (FSV-43-40 and FSV-43-42) and the PASS Residual Heat Removal liquid sample valves (FSV-43-50 and FSV-43-56) are added to Table 3.7.A.These valves are located outside of the primary containment.
These valves receive a Group 6 isolation signal, do not have a specified maximum operating time, are normally closed, and stay closed upon receipt of the isolation signal.2~Revision to Table 3.7.A,"Primary Containment Isolation Valves".Proposed change to Table 3.7.A would add: Drywell Control Air (DCA)system inlet header check valves (32-2516 and 32>>2521)and Containment Atmospheric Dilution (CAD)crosstie to DCA check valve (84-680)are added to Table 3.7.A.Valve 32-2516 is located inside the primary containment.
Valves 32-2521 and 84-680 are located outside of the primary containment.
The two DCA check valves do not have a specified maximum operating time, are normally open, and either stay open or close after an accident depending on the operational status of the process system.The CAD check valve does not have a specified maximum operating time, is normally closed, and will either stay closed or open after an accident depending on the operational status of the process system.3~Revision to Section 6.0,"Administrative Controls".
Proposed change to Section 6.0 would add: 'I Specification, 6.8.5, which requires BFN establish, implement, and maintain 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 (1)training of personnel, (2)procedures for sampling and analysis, and (3)provisions for maintenance of sampling and analysis.
ENCLOSURE 3 BROWNS FERRY NUCLEAR PLANT (BFN)UNIT 2 REASON AND JUSTIFICATION FOR THE PROPOSED CHANGES Reason for Chan es The BFN Unit 2 Technical Specifications are being revised: (1)to reflect the addition of four new primary containment isolation valves associated with the installation of the BFN Post-Accident Sampling System (PASS), (2)to reflect the addition of three new primary containment isolation valves associated with a modification to increase the Automatic Depressurization System (ADS)accumulator reliability, and (3)to add an administrative requirement associated with the PASS.Justification for Chan es (1)Zn order for licensees to ensure that their plants have the capability to obtain and perform radiological and chemical analyses of reactor coolant and containment atmosphere samples after a design basis accident, NRC requested in NUREG-0737, TMI Action Plan, Item II.B.3,"Post-Accident Sampling Capability"g that licensees install such a capability.
In response, TVA committed to install the BFN PASS prior to restart.This commitment is documented in TVA's December 28, 1989 letter to NRC.Equipment to perform this sampling and analytical function is required to be capable of providing at least one sample per day for seven days following onset of the accident and at least one sample per week until the accident condition no longer exists.Zn order to perform this function, the sample panel must be located outside of primary containment to allow for panel operation and for the performance of required maintenance.
Therefore, primary coolant and containment atmosphere sample lines must be routed through primary containment to the sample panel.In order to assure containment integrity and maintain post-design bask.if'ccident fission product releases below 10 CFR 100 limits, containment isolation valves are provided on each sample line which is considered an extension of primary containment (i.e., the sample line from the Residual Heat Removal (RHR)System and the PASS sample return line to the torus).These containment isolation valves are sub)ect to the testing provisions of 10 CFR 50, Appendix J and are required to be listed in BFN Technical Specifications, Table 3.7.A, Primary Containment Isolation Valves.
ENCLOSURE 3 (Continued)
BROWNS FERRY NUCLEAR PLANT (BFN)UNIT 2 REASON AND JUSTIFICATION FOR THE PROPOSED CHANGES Justificatio or C a es Continued (2)In order to verify that the accumulators for the ADS valves are provided with sufficient capacity to cycle the valves open five times at design pressure and to still perform their function for 100 days following an accident, NRC requested in NUREG-0737, TMI Action Plan, Item ZI.K.3.28, Verify Qualification of Accumulators of Automatic Depressurization System Valves, that licensees verify this capability.
Zn response, TVA committed to install an upgraded capability prior to restart.This commitment is documented in TVA's December 28, 1989 letter to NRC.The detailed description of the proposed modification is contained in TVA letters to NRC, dated July 12, 1984 and July 11, 1985.The NRC accepted this proposed modification, as documented in the Safety Evaluation which was included in NRC's July 24, 1985 letter to TVA The ADS accumulator system is capable of five actuations at design pressure.If the nitrogen supply is lost during an accident, and assuming maximum allowable accumulator leakage, sufficient pressure for ADS actuations could be maintained for six hours.Zn order for these valves to remain functional for periods up to 100 days following an accident, TVA committed to separate the Drywell Control Air (DCA)system supply into two separate trains.Each train is capable of being supplied with pressurized nitrogen from the Containment Air Dilution (CAD)system or from the DCA system.Therefore, an additional supply line must be routed through primary containment to provide a redundant supply of nitrogen to the ADS accumulators.
In order to assure containment integrity and maintain post-design basis accident fission product releases below 10 CFR 100 limits, containment isolation valves are provided on these lines.A single check valve is provided on the supply line inside containment and a check valve is provided on each of the CAD and DCA lines outside containment.
This arrangement is identical to the existing CAD/DCA line into the drywell.These containment isolation valves are sub)ect to the testing provisions of 10 CFR 50, Appendix J and are recpxired to be listed in BFN Technical Specifications, Table 3.7.A, Primary Containment Isolation Valves.
ENCLOSURE 3 (Continued)
BROWNS FERRY NUCLEAR PLANT (BFN)UNIT,2 REASON AND JUSTIFICATION FOR THE PROPOSED CHANGES Justif cat, on for C an es Conti ued (3)NRC requested in Generic Letter 83-36, NUREG-0737 Technical Specifications, dated November 1, 1983, that licensees establish Technical Specifications to ensure that their plant has the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.
TVA committed to submit the required Technical Specification Amendment prior to restart.This commitment is documented in TVA's December 28, 1989 letter to NRC.This submittal is in response to that commitment.
The Generic Letter states that this capability should be established, implemented and maintained.
The activities associated with this capability shall include (1)training of personnel, (2)procedures for sampling and analysis, and (3)provisions for maintenance of sampling and analysis.In Generic Letter 83-36, NRC stated that it was acceptable for licensees to refer to this capability in the administrative controls section of the Technical Specifications, and include and maintain the detailed descriptions in plant procedures.
BFN will meet the requirement in this manner.
ENCLOSURE 4 BROWNS FERRY NUCLEAR PLANT (BFN)UNIT 2 PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Descri tion of Pro osed Technical S ecification Amendment The BFN Unit 2 Technical Specifications are being revised as follows'able 3.7.A,"Primary Containment Isolation Valves," is being revised to include the Post-Accident Sampling System (PASS)licpxid and gas return to torus valves (FSV-43-40 and FSV-43-42) and the PASS Residual Heat Removal licpxid sample valves (FSV-43-50 and FSV-43-56).
2~Table 3.7.A,"Primary Containment Isolation Valves," is being revised to include the Drywell Control Air (DCA)System inlet header check valves (32-2516.and 32-2521)and the Containment Atmospheric Dilution (CAD)crosstie to DCA check valve (84-680).3~Section 6.0,"Administrative Controls," is being revised to add a new specification, 6.8.5, which requires BFN establish, implement and maintain the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.
The activities associated with this capability shall include (1)training of personnel, (2)procedures for sampling and analysis, and (3)provisions for maintenance of sampling and analysis.Basis for Pro osed No Si nificant Hazards Consideration Determination NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c)~A proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1)involve a significant increase in the probability or consequences of an accident previously evaluated, (2)create the possibility of a new or different kind of accident from an accident previously evaluated, or (3)involve a significant reduction in a margin of safety.The proposed change does not significantly increase the probability or consecpxences of an accident previously evaluated.
The PASS containment isolation valves satisfy the design criteria for BFN containment isolation valves.This modification does not significantly affect=the ability of the-containment to perform its ultimate safety ob)ective, which is to assure containment integrity and maintain post-design basis accident fission product releases below 10 CFR 100 limits.These containment isolation valves.will be provided with automatic isolation signals from the Primary Containment Isolation System.These valves are powered from Class 1E power supplies.
1 ENCLOSURE 4 (Continued)
BROGANS FERRY NUCLEAR PLANT (BFN)UNIT 2 PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Basis for Pro osed No Si nificant Hazards Consideration Determination Continued The addition of the PASS containment isolation valves by this proposed amendment does not eliminate or modify any requirement or commitment to comply with the provisions of 10 CFR 50, Appendix J.The operation of these containment isolation valves or the failure of any single valve to operate does not affect the Final Safety Analysis Report (FSAR)analysis of design bases accidents.
Therefore, the addition of these valves does not significantly increase the probability or consequences of an accident previously evaluated.
The DCA and CAD containment isolation valves satisfy the design criteria for BFN containment isolation valves.This modification does not significantly affect the ability of the containment to perform its ultimate safety ob)ective, which is to assure containment integrity and maintain post-.design basis accident fission product releases below 10 CFR 100 limits.These containment isolation valves are check valves and will automatically perform their isolation function if backflow through these lines is sensed.This arrangement is identical to the existing CAD/DCA line into the drywell.The addition of the DCA and CAD containment isolation valves by this proposed amendment does not eliminate or modify any requirement or commitment to comply with the provisions of 10 CFR 50, Appendix J.The operation of these containment isolation valves or the failure of any single valve to operate does not affect the Final Safety Analysis Report (FSAR)analysis of design bases accidents.
Therefore, the addition of these valves does not significantly increase the probability or consequences of an accident.previously evaluated.
The purpose of requiring the ADS valves be operable for 100 days following an accident is to provide the operators with the ability to manually decrease the pressure of the primary system.The addition of a redundant supply line to the ADS accumulators will aid in mitigation of the consequences of an accident by enhancing the reliability of ADS operation.
~~%W ENCLOSURE 4 (Continued)
BROWNS FERRY NUCLEAR PLANT (BFN)UNIT 2 PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Basis for Pro osed No Si nificant Hazards Consideration Determination Continued The purpose of the PASS is to perform radiological and chemical analyses of reactor coolant and containment atmosphere samples after a design basis accident has occurred.Otherwise, the PASS is only operated for short periods of time to support training and calibration activities.
The operation of the PASS or the failure of the PASS to operate does not affect the Final Safety Analysis Report (FSAR)analysis of design bases accidents.
Therefore, the addition, of an administrative requirement to establish and maintain a post-accident sampling capability does not involve a significant increase in the probability or consequences of an accident previously evaluated.
This sampling and analytical capability will aid in mitigation of the consequences of an accident by providing timely information to operating personnel on certain plant parameters.
2~This'proposed change does not create the possibility of a new or different kind of accident from an accident previously evaluated.
The addition of four PASS primary containment isolation valves and the three primary containment isolation valves associated with the redundant nitrogen supply line to the ADS accumulators by this proposed amendment does not eliminate or modify any requirement or commitment to comply with the provisions of 10 CFR 50, Appendix 8, or BFN Technical Specification 3/4.7.A.These additional containment isolation valves satisfy the design criteria for the containment isolation valves for BFN systems which penetrate primary containment and are of comparable size and configuration.
The addition of this equipment will not require any existing equipment to operate in a different manner from which it was designed to operate.The operation of these containment isolation valves or the failure of any single valve to operate does not affect the Final Safety Analysis Report (FSAR)analysis of design bases accidents.
No operation outside the plant design basis is introduced, so there is no possibility for creation of a new or different kind of accident from any previously evaluated.
~(
ENCLOSURE 4 (Continued)
BROWNS FERRY NUCLEAR PLANT (BFN)UNIT 2 PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Basis for Pro osed No Si nificant Hazards Considerat o Dete i ation continued The addition of a requirement to establish and maintain a post>>accident sampling capability is administrative and does not add any new equipment to the plant or require any existing equipment to be operated in a different manner from which it was designed to operate.Therefore, there is no effect on the Final Safety Analysis Report (FSAR)analysis of design bases accidents.
No operation outside the plant design basis is introduced, so there is no possibility for creation of a new or different kind of accident from any previously evaluated.
3 0 This proposed change does not involve a significant reduction in a margin of safety.The addition of four PASS primary containment isolation valves and the three primary containment isolation valves associated with the redundant nitrogen supply line to the ADS accumulators by this.proposed amendment is consistent with the existing BFN Safety Analysis.No adverse safety impact or significant reduction in safety margins occur due to the addition of these valves.The changes do not significantly affect the ability of the containment to perform its ultimate safety objective to assure containment integrity and maintain post-design basis accident fission product releases below 10 CFR 100 limits.BFN commitments to comply with the provisions of 10 CFR 50, Appendix J, and BFN Technical Specification 3/4.7.A are not altered.Therefore, this proposed change does not involve a significant reduction in a margin of safety.The addition of a requirement to establish and maintain a post-accident sampling capability is administrative.
No adverse safety impact or reduction in safety margins occur due to this change.This change does not physically modify any equipment, setpoints, or initiation sequence of equipment.
Therefore, this proposed change does not.involve a significant reduction in a margin of safety.Determination of Basis for Pro osed No Si nifidhnt Hazards Since the application for amendment involves a proposed change that is encompassed by the criteria for which no significant hazards consideration exists, TVA has made a proposed determination that the application involves no-significant hazards consideration.
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