ML20235H114
| ML20235H114 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/11/1987 |
| From: | Zwolinski J NRC OFFICE OF SPECIAL PROJECTS |
| To: | |
| Shared Package | |
| ML20235H120 | List: |
| References | |
| NUDOCS 8709300381 | |
| Download: ML20235H114 (99) | |
Text
o
.. )
[e nc Io UNITED STATES s
g
[
3 g
NUCLEAR REGULATORY COMMISSION g
'j WASHINGTON, D. C. 20555
- %...../
TENNESSEE VALLEY AUTHORITY I
DOCKET NO. 50-?59
)
BPOWNS FEpRY NUCLEAR PLANT, UNIT 1 J
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.138 License No. DPP-33 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated September 27, 1984 as supplemented January 17, June 2 end December 10, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFP Part 51 of the Commission's regulations and all applicable reovirements have been satisfied.
2.
Accordingly, the license is amended by chances to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(?) of Facility Operating License No. OPR-33 is herehy amended to read as follows:
kk b
PR P
c
.2-(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. '38, are hereby incorporated in the license. The licensee shal', operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.
F0R THE NUCLEAR REGULATORY COMMISSION OmEw 8
oh A. Zwolinsk As t Director or Projects TVA Projects Division Office of Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: September 11, 1987
e.
J j
i ATTACHMENT TO LICENSE ANENDMENT NO. 138
-1 FACILITY OPERATING NO. DPR-33 DOCKET N0. 50-259 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by the i
captioned amendment number and contain marginal lines indicating the area of.
i change. Overleaf pages* are provided to maintain document completeness.
Remove Insert iii iii*
iv iv v
v vi*
vi vii vii viii viii 3.2/4.2-3 3.2/4.2-3 3.2/4.2-4 3.2/4.2-4*
3.2/4.2-5 3.2/4.2-5 3.2/4.2-6 3.2/4.2-6 i
3.8/4.8-1 3.8/4.8-1*
l 3.8/4.8-2 3.8/4.8-2 3.8/4.8-3 3.8/4.8-3 3.8/4.8-4 3.8/4.8-4 3.8/4.8-5 3.8/4.8-5 3.8/4.8-6 3.8/4.8-6 3.11/4.11-3 3.11/4.11-3 3.11/4.11-4 3.11/4.11-4*
I 3.11/4.11-5 3.11/4.11-5*
3.11/4.11-6 3.11/4.11-6 3.11/4.11-7 3.11/4.11-7 3.11/4.11-8 3.11/4.11-8 3.11/4.11-12 3.11/4.11-12 j
3.11/4.11-13 3.11/4.11-13*
l 3.11/4.11-14 3.11/4.11-14*
3.11/4.11-15 3.11/4.11-15 6.0-1 thru 6.0-23 6.0-1 thru 6.0-34
)
i
J Section Paac No.
C.
Coolant Leakage....
3.6/4.6-9 D.
Relief Valves......
3.6/4.6-10 3.6/4.6-11 E.
Jet Pumps.
F.
Recirculation Pump Operation.........
3.6/4.6-12 G.
Structural Integrity......
3.6/4.6-13 H.
Seismic Restraints, Supports-and Snubbers...
3.6/4.6-15 3.7/4.7 Containment Systems....
3.7/4.7-1 A.
3.7/4.7-1 B,
Standby Gas Treatment System..
3.7/4.7-13 C.
3.7/4.7-17 D.
Primary Containment Isolation Valves....
3.7/4.7-18 E.
Control Room Emergency Ventilation......
3.7/4.7-19 F.
Primary Containment Purge System.......
3.7/4.7-21 G.
Containment Atmosphere Dilution System (CAD) 3.7/4.7-22 H.
Containment Atmosphere Monitoring (CAM)
System Hg Analyzer.......
3.7/4.7-24 3.8/4.8 Radioactive Materials.....
3.8/4.8-1 A.
Liquid Effluents...............
3.8/4.8-1 B.
Airborne Effluents....
3.8/4.8-3 C.
Radioactive Effluents - Dose........
3.8/4.8 6 g
l D.
Mechanical Vacuum Pump.
3.8/4.8-6 E.
Miscellaneous Radioactive Materials Sources..
3.8/4.8-7 F.
Solid Radwaste.....
3.8/4.8-9 3.9/4.9 Auriliary Electrical System.
3.9/4.9-1 j
A.
Auxiliary Electrical Equipment.
- 3. 9 / 4. 9 -1 B.
Operation with Inoperable Equipment.
3.9/4.9-10 C.
Operation in Cold Shutdown......
3.9/4.9-17 3.10/4.10 Core Alterations..
3.10/4.10-1 A.
Refueling Interlocks.
3.10/4.10-1 111 gg3 Amendment No. 132 j
i, o
Paae No.
Section 3.10/4.10-4 B.
Core Monitoring..
3.10/4.10-7 C.
Spent Fuel Pool Water.
3.10/4.10-8 D.
Reactor Building Crane.
3.10/4.10-9 E.
Spent Fuel Cask.
3.10/4.10-10 F.
Spent Fuel Cask Handling-Refueling Floor...
3.11/4.11-1 3.11/4.11 Fire Protection Systems....
3.11/4.11-1 A.
High Pressure Fire Protection System.
3.11/4.11-4 B.
CO2 Fire Protection System..
3.11/4.11-6 C.
Fire Detectors..
3.11/4.11-7 D.
Deleted E.
Fire Protection Systems Inspection......
3.11/4.11-8 3.11/4.11-8 F.
Deleted G.
Air Masks and Cylinders.......
3.11/4.11-8 3.11/4.11-9 H.
I.
Open Flames, Welding and Burning in the 3.11/4.11-9 Cable Spreading Room.
5.0-1 5.0 Major Design Features.
5.0-1 5.1 Site Features..
5.0-1 5.2 Reactor..
5.0-1 5.3 Reactor Vessel.
5.0-1 5.4 Containment.
5.0-1 5.5 Fuel Storage...
5.0-2 5.6 Seismic Design.
is BFN Unit 1 Amendment No.138
)
l ADMINISTRATIVE CONTROLS SECTION
[ AGE 6.1 RESPONSIBILITY..........................................
6.0-1 6.2 ORGANIZATION............................................
6.0-1 6.2.1 Corporate...............................................
6.0-1 6.2.2 Plant Staff.............................................
6.0-1 6.3 PLANT STAFF QUALIFICATIONS..............................
6.0-4 6.4 TRAINING................................................
6.0-4 6.5 PLANT REVIEW AND AUDIT..................................
6.0-4 6.5.1 Plant Operation Review Committee (P0RC).................
6.0-4 6.5.2 Nuclear Safety Review Board (NSRB)......................
6.0-10 6.5.3 Technical Review and Approval of Procedures.............
6.0-26 t
6.6 REPORTABLE EVENT ACTIONS................................
- 6. 0-17 6.7 SAFETY LIMIT VIOLATION..............................,...
6.0-18 6.8 PROCEDURES / INSTRUCTION AND PROGRAMS.....................
6.0-19 6.8.1 Procedures..............................................
6.0-19 6.8.2 Dri11s..................................................
6.0-20 6.8.3 Radiation Control Procedures............................
6.0-20 6.8.4 Quality Assurance Procedures - Effluent and..............
Environmental Monitoring..............................
6,0-22 6.9 REPORTING REQUIREMENTS..................................
6.0-23 6.9.1 Routine Reports.........................................
6.0-23 Startup Reports.........................................
6.0-23 Annual Operating Report.................................
6.0-24 Monthly Operating Report................................
6.0-25 Reportable Events.......................................
6.0-25 Radioactive Effluent Release Report.....................
6.0-25 Source Tests............................................
6.0-25 6.9.2 Special Reports.........................................
6.0-26 6.10 STATION OPERATING RECORDS AND RETENTION.................
6.0-28 6.11 PROCESS CONTROL PR0 GRAM.................................
6.0-31 6.12 0FFSITE DOSE CALCULATION MANUAL.........................
6.0-31 6.15 RADIOLOGICAL EFFLUENT MANUAL............................
6.0-32 e
i BFN
(
Unit 1 Amendment No.138 i
i i
l J
t LIST OF TABLES Table Title Paae No.
1.1 Surveillance Frequency Notation..........
1.0-12 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 Circuits........
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.
3.2/4.2-7 3.2.B Instrumentation that Initittes or Controls the Core and Containment Cooling Systems..
3.2/4.2-14 3.2.C Instrumentation that Initiates Rod Blocks.
3.2/4.2-25 3.2.D 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 Control Room Isolation Instrumentation.......
3.2/4.2-34 3.2.H Flood Protection Instrumentation.
3.2/4.2-35 3.2.I Meteorological Monitoring Instrumentation.....
3.2/4.2-36 3.2.J Seismic Monitoring Instrumentation.........
3.2/4.2-37 3.2.K Radioactive Gaseous Effluent Monitoring Instrumentation....
3.2/4.2-38 4.2.A Surveillance Requirements for Primary Containment l
and Reactor Building Isolation Instrumentation..
3.2/4.2-40 1
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 J
l 4.2.D Radioactive Liquid Effluent Monitoring I
Instrumentation Surveillance Requirements.
3.2/4.2-51 i
1 vi Amendment No. 132 BFN Unit 1
}
LIST OF TABLES (Cont'd)
(
Table Title Page No.
l l
4.2.E Minimum Test and Calibration Frequency for Drywell Leak Detection Instrumentation..........
3.2/4.2-53 4.2.F Minimum Test and Calibration Frequency for Surveillance Instrumentation......
3.2/4.2-54 4.2.G Surveillance Requirements for Control Room Isolation Instrumentation............
3.2/4.2-56 4.2.H Minimum Test and Calibration Frequency for 3.2/4.2-57 Flood Protection Instrumentation...
4.2.J Seismic Monitoring Instrument Surveillance 3.2/4.2-58 Requirements..................
4.2.K Radioactive Gaseous Effluent Instrumentation 3.2/4.2-62 Surveillance 3.5-1 Minimum RHRSW and EECW Pump Assignment.
3.5/4.5-11 3.5.I MAPLHGR Versus Average Planar Exposure.
3.5/4.5-21 3.7.A Primary Containment Isolation Valves.....
3.7/4.7-25 3.7.B Testable Penetrations with Double 0-Ring Seals.
3.7/4.7-32 3.7.C Testable Penetrations with Testable Bellows..
3.7/4.7-33 3.7.D Air Tested Isolation Valves....
3.7/4.7-34 3.7.E Primary Containment Isolation Valves which Terminate below the Suppression Pool Water 3.7/4.7-37 Level......................
3.7 F Primary Containment Isolation Vales Located in Water Sealed Seismic Class 1 Lines.......
3.7/4.7-38 3.7.H Testable Electrical Penetrations........
3.7/4.7-39 4.9.A.4.C Voltage Relay Setpoints/ Diesel Generator Start.
3.9/4.9-16 3.11.A Fire Protection System Hydraulic Requirements...
3.11/4.11-10 6.2.A Minimum Shift Crew Requirements.
6.0-3 vil BFN Amendment No. 132,138 Unit 1
1 1
LIST OF ILLUSTRATIONS Figure Title Page No.
2.1.1 APRM Flow Reference Scram and APRM Rod Block Settings....................
1.1/2 1-6 2.1-2 APRM Flow Bias Scram vs. Reactor Core Flow....
1.1/2.5.-7 4.1-1 Graphic Aid in the Selection of an Adequate Interval Between Tests..........
3.1/4.1-13 i
4.2-1 System Unavailability...............
3.2/4.2-64 j
3.4-1 Sodium Pentaborate Solution Volume Concentrated 3.4/4.4-4 Requirements.....
l 1
3.4-2 Sodium Pentaborate Solution Temperature 3.4/4.4-5 Requirements.
1 3.5.K-1 MCPR Limits.
3.5/4.5-24 3.5/4.5-25 3.5.2 Kr Factor.
3.6-1 Minimum Temperature OF Above Change in 3.6/4.6-24 Transient Temperature.
3.6-2 Change in Charpy V Transition Temperature Vs.
3.6/4.6-25 Neutron Exposure.
4.8.1.a Gaseous Release Points and Elevations.
3.8/4.8-10 4.8.1.b Land Site Boundary 3.8/4.8-11
]
f 6.2-1 Offsite Organization for Facility Management j
and Technical Support 6.0-33 l
6.2-2 Facility Organization 6.0-34 1
}
1 i
l l
I vili i
)
BFN Amendment No. Z32,138 Unit 1 1
L__
L r
J l
l 3.2/4.2 PROTECTIVE INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.D Radioactive Liould Effluent 4.2.D Radioactive Liould Effluent Monitoring Instrumentation Monitoring Instrumentation i
1.
The radioactive liquid 1.
Each of the radioactive liquid effluent monitoring effluent monitoring instruments instrumentation listed in shall be demonstrated operable by j
l Table 3.2.D shall be operable performance of test in accordance I
with the applicability as with Table 4.2.D.
shown in Tables 3.2.D/4.2.D.
Alarm / trip setpoints will be set in accordance with guidance given in the ODCM to ensure that the limits of specification 3.8.A.1 are not exceeded.
2.
The action required when the number of operable channels is less than the minimum channels operable requirement is specified in the notes for Table 3.2.D.
Exert best efforts to raturn the instrument (s) to OPERABLE status within 30 days and if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.
3.
With a radioactive liquid effluent monitorin6 channel alarm / trip setpoint less conservative than required by these specifications, suspend the release without delay, declare the channel inoperable, or adjust the alarm / trip setpoint to establish the conservatism required by these specifications.
4.
The provisions of Specifications 1.0.C and 6.9.1.4 are not applicable.
3.2/4.2-3 BFN U"It 1 Amendment No. 132, 138
(
t 3.2/4.2 PROTECTIVE INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.E.
Drywell Leak Detection 4.2.E.
Drywell Leak Detection The limiting conditions of Instrumentation shall be operation for the calibrated and checked as l
instrumentation that monitors indicated in Table 4.2.E.
drywell leak detection are given in Table 3.2.E.
\\
F.
Surveillance Instrumentation F.
Surveillance Instrumentation l
The limiting conditions for the Instrumentation shall be instrumentation that provides calibrated and checked as surveillance information indicated in Table 4.2.F.
l readouts are given in I
Table 3.2.F.
G.
Control Room Isolation G.
Control Room Isolation The limiting conditions for Instrumentation shall be instrumentation that isolates calibrated and checked as the control room and initiates indicated in Table 4.2.G.
the control room emergency pressurization systems are given in Table 3.2.G.
H.
Flood Protection H.
Flood Protection The unit shall be shutdown and Surveillance shall be placed in the cold condition performed on the when Wheeler Reservoir lake instrumentation that monitors stage rises to a level such the reservoir level as that water from the reservoir indicated in Table 4.2.H.
begins to run across the pumping station deck at elevation 565.
Requirements for instrumentation that monitors the reservoir level are given in Table 3.2.H.
I.
Meteorological Monitoring I.
Meteorological Monitoring Instrumentation Instrumentation The meteorological monitoring Each meteorological monitoring instrumentation listed in Table instrument channel shall be 3.2.I shall be operable demonstrated operable by the at all tines.
performance of the CHANNEL CHECK at least once per l
DFN 3.2/4.2-4 Unit 1
e i
3.2/4.2 PROTECTIVE INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.I.
Meteorological Monitoring 4.2.I Meteorological Monitoring Instrumentation (Cont'd)
Instrumentation (Cont'd) 1.
With the number of operable 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the CHANNEL meteorological monitoring CALIBRATION at least once channels less than required each 6 months.
by Table 3.2.I resto: e the inoperable channel (s) to j
operable status within j
7 days.
2.
With one or more of the meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring I
the system to operable status.
J.
Seismic Monitoring 4.2.J.
Seismic Monitoring Instrumentation Instrumentation 1.
The seismic monitoring 1.
Each of the seismic j
i instruments listed in monitoring instruments Table 3.2.J shall be shall be demonstrated f
operable at all times, operable by performance of tests at the frequencies 2.
With the number of listed in Table 4.2.J.
j seismic monitoring j
instruments less than the 2.
Data shall be retrieved a
number listed in Table from all seismic instruments 3.2.J. restore the actuated during a seismic inoperable instrument (s) event and analyzed to to operable status within determine the magnitude 30 days.
of the vibratory ground motion. A Special Report
{
3.
With one or more of the shall be submitted to the l
instruments listed in Table Commission pursuant to j
3.2.J inoperable for more Specification 6.9.2 within i
than 30 days, submit a 10 days describing the
{
Special Report to the magnitude, frequency Commission pursuant to spectrum, and resultant Specification 6.9.2 within effect upon plant features j
the next 10 days describing important to safety.
the cause of the malfunction and plans for restoring the instruments to operable status.
3.2/4.2-5 BFN Amendment No. I32,138 Unit 1 i
n s
1 l
3.2/4.2 PROTECTIVE INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.K Radioactive Gaseous Effluent 4.2.K Radioactive Gaseous Effluent Monitoring Instrumentation Monitoring Instrumentation 1.
The radioactive gaseous 1.
Each of the radioactive effluent monitoring gaseous effluent monitoring instruments listed in instruments shall be Table 3.2.K shall be operable demonstrated operable by with the applicability as performance of tests in l
shown in Tables 3.2.K/4.2.K accordance with Table 4.2.K.
Alarm / trip setpoints will be set in accordance with guidance given in the ODCM to I
ensure that the limits of l
Specification 3.8.B.1 are not exceeded.
2.
The action required when the l
number of operable channels is less than the Minimum Channels Operable requirement is specified in the notes for Table 3.2.K.
Exert best I
efforts to return the l
instruments to operable status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Release Report why the inoperability was not corrected in a timely manner.
3.
With a radioactive gaseous effluent monitoring channel alarm / trip setpoint less conservative than required by these specifications, suspend the release without delay, declare the channel inoperable or adjust the alarm / trip setpoint to establish the conservatism required by these specifications.
4.
Both off-gas treatment monitors may be taken out of service for less than one hour for purging of monitors during SI performance.
5.
The provisions of Specifications 1.0.C and 6.9.1.4 are not applicable.
3.2/4.2-6 Amendment No. 232,138 BFN
o n
3.8/4.8 RADIOACTIVE MATERIALS j
LIMITING CONDITIONS FOR OPERATION' SURVEILLANCE REQUIREMENTS 3.8 Radioactive Materials 4.8 Radioactive Materials Applicability Applicability Applies to the release of Applies to the periodic test and 1
radioactive liquids and gases record requirements and sampling
)
from the faellity, and monitoring methods used for facility effluents.
Objective Objective i
To define the limits and To ensure that radioactive conditions for the release of liquid and gaseous releases from radioactive effluents to the the facility are maintained environs to assure that any within the limits specified by radioactive releases are as Specifications 3.8.A and 3.8.B.
low as reasonably achievable and within the limits of Specification 10 CFR Part 20.
The specifications j
except for 3.8.A.1 and 3.8.B.1 are A. Liquid Effluents I
exempt from the requirements of definition 1.0.C (Limiting
- 1. Facility records shall Condition for Operation).
be maintained of radioactive concentrations and Specification volume before dilution of each batch of liquid j
A. Liauld Effluents effluent released, and J
of the average dilution
- 1. The concentration of flow and length of time radioactive material over which each released at any time from discharge occurred.
]
the site to unrestricted areas (see Figure 4.8-Ib)
- 2. Radioactive liquid waste f
shall be limited to the sampling and activity
]
concentrations specified analysis of each liquid in 10 CFR Part 20, waste batch to be Appendix B. Table II, discharged shall be j
Column 2 for radionuclides performed prior to
{
other than dissolved or release in accordance i
entrained noble gases, with the sampling and For dissolved or entrained analysis program noble gases, the specified in the REM.
concentration shall be limited to 2E-4 pCi/ml
- 3. The operation of the total activity.
automatic isolation valves and discharge
- 2. If the limits of 3.8.A.1 tank selection valves are exceeded, appropriate shall be checked action shall be initiated annually.
without delay to bring
)
the release within BFN 3.8/4.8-1 Amendment No.132 Unit 1 l
3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.A.
Liauld Effluents 4.8 A. Liquid Effluents limits. Provide prompt
- 4. The results of the notification to the NRC analysis of samples pursuant to Section 6.9.1.4.
collected from release points shall be used
- 3. The doses or dose with the calculational commitment to a member of methodology in the ODCM the public from radioactive to assure that the materials in liquid concentrations at the effluents released from point of release are each unit to unrestricted maintained within the tress (See Figure 4.8-lb) limits of specification shall be limited:
3.8.A.I.
a.
During any calendar
- 5. Cummulative quarterly guarter to <1.5 mrem to and yearly dose the total body and <5 contributions from mrem to any organ and, liquid effluents shall be determined as b.
During any calendar specified in the ODCM at year to <3 mrem to the least once every 31 days.
total body and <10 mrem to any organ
- 6. The quantity of radioactive material
- 4. If the limits specified in contained in any 3.8 A.3 a & b above are outside liquid radwaste exceeded, prepare and storage tanks shall be submit Special Report determined to be within the pursuant to Section 6.9.1.4.
above limit by analyzing a representative sample of
- 5. The maximum activity to be the tank's contents at contained in one liquid least once per 7 days when radwaste tank or temporary radioactive materials are storage tank that can be being added to the tant, discharged directly to the environs shall not exceed 10 curies excluding tritium and dissolved / entrained noble gas.
- 6. With radioactive liquid waste exceeding 3.8.A.5 limits, without delay l
suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, I
reduce the tank contents to within the limit.
Events leading to this condition must be reported in the next Semiannual Radioactive Effluent Release Report (Section F.2 of the REM)
Amendment No. 132, 138 3.8/4.8-2 l
BFN Unit 1 l
t J
J
)
1 3.8/4.8 R/DI0 ACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.B.
Airborne Effluents 4.8.B.
Airborne Effluents
- 1. The dose rate at any time 1.
The gross S/y and to areas at and beyond the particulate activity of site boundary (see Figure gaseous wastes released 4.8-Ib) due to to the environment shall radioactivity released in be monitored and gaseous effluents from the recorded.
i I
site shall be limited to I
the following values:
- a. For effluent streams having continuous a.
The dose rate limit monitoring for noble gases shall capability, the be <500 mrem /yr to the activity shall be total body and <3000 monitored and flow mrem /yr to the skin, rate evaluated and j
and recorded to enable
)
release rates of j
b.
The dose rate limit gross radioactivity l
for I-131, I-133. H-3, to be determined at and particulate with least once per shift greater than eight day using instruments half-lives shall be specified in Table
<1500 mrem /yr to any 3.2.K.
i
- organ,
- b. For effluent streams
- 2. If the limits of 3.8.B.1 without continuous are exceeded, appropriate monitoring corrective action shall be capability, the immediately initiated to activity shall be bring the release within monitored and limits. Provide prompt recorded and the notification to the NRC release through pursuant to Section these streams 6.9.1.4.
controlled to within the limits specified in 3.8.B.
2.
Radioactive gaseous 1
waste sampling and j
activity analysis shall j
be performed in accordance with the j
sampling and analysis program specified in the REM. Dose rates shall
)
be determined to be I
within limits of 3.8.B using methods contained f
in the ODCM.
3.8/4.8-3
{
Amendment No. I32,138 BFN Unit 1
)
T
e 6
l l
3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l
3.8.B Airborne Effluents 4.8.B Airborne Effluents (Cont'd)
(Cont'd)
- 3. The air dose to areas at 3.
Cumulative guarterly and and beyond the site yearly dose contributions from boundary (see Figure 4.8-1b) due to noble gases gaseous releases shall be determined using released in gaseous effluents per unit shall methods contained in the be limited to the ODCM at least once every following:
31 days.
a.
During any calendar quarter, to 55 mrad for gamma radiation and $10 mrad for beta radiation; b.
During any calendar year, to $10 mead for gamma radiation and
$20 mrad for beta radiation.
1 I
- 4. If the calculated air dose exceeds the limits specified in 3.8 B.3 above, prepare and submit a special report pursuant to section 6.9.1.4.
- 5. The dose to a member of l
the public from radioiodines, radioactive l
materials in particulate i
form, and radionuclides l
other than noble gases with half lives greater than 8 days in gaseous effluent released per unit to areas at and beyond the site boundary (see Figure l
4.8-Ib) shall be limited to the following:
a.
To any organ during any calendar quarter to $7.5 mrem; b.
To any organ during any calendar year to
$15 mrem; 3.8./4.8-4 Amendment No. 232, 138 BFN Unit 1
.i 3.8/4.8 RADI0 ACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 1
3.8.B Airborne-Effluents 4.8.B Airborne Effluents (Cont'd)
(Cont'd)
- 6. If the calculated doses 4.
During operatien above exceed the limits of 25% power, the pnsition 3.8.B.5 above, prepare of the charcoal bed and submit a special bypass valve will be report pursuant to verified daily.
Section 6.9.1.4.
- 7. During operation above 5.
The concentration of 25% power the discharge hydrogen downstream of of the SJAE must be the recombiners shall be routed through the determined to be within charcoal adsorbers, the. limits of 3.8.B.9 by continuously monitoring
- 8. With gaseous waste being the off-gas whenever discharged for more than the SJAE is in servico 7 days without treatment using instruments described in through the charcoal Table 3.2.K.
Instrument adsorbers, prepare and surveillance requirements are submit a special report specified in Table 4.2.K.
pursuant to Section 6.9.1.4.
- 9. Whenever the SJAE is in service, the concentration of hydrogen in the offgas downstream of the recombiners shall be limited to $4% by volume.
- 10. With the concentration of hydrogen exceeding the limit of 3.8.B.9 above, restore the concentration to within the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
I i
i l
1 Amendment No. 132, 138 BFN i
Unit 1
\\
o s
3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.C Radioactive Effluents - Doae 4.8.C Radioactive Effluents - Dose
- 1. The dose or dose 1.
Cumulative dose commitment to a real contributions from individual from all liquid and gaseous uranium fuel cycle sources effluents shall be is limited to $25 neem to determined in the total body or any accordance with organ (Jxcept the thyroid, Specifications 3.8.A 3, which is limited to $75 3.8.B.3, and 3.8.B.5 arem) over a period of one and the methods in the calendar year.
ODCM.
- 2. With the calculated dose from the release of radioactive materials in liguld or gaseous effluents exceeding twice the limits of Specification 3.8.A.3, 3.8.B.3, or 3.8.B.5, prepare and submit a Special Report to the l
Commission pursuant to specification 6.9.1.4.
and limit the subsequent releases such that the limits of 3.8.C.1 are not exceeded.
3.8.D Mechanical Vacuum Pump 4.8.D Mechanical Vacuum Pump
- 1. Each mechanical vacuum pump At least once durint each shall be capable of being operating cycle verify automatic automatically isolated and securing and isolation of the j
secured on a signal or mechanical vacuum pump.
high radioactivity in the steam lines whenever the main steam isolation valves are open.
- 2. If the limits of 3.8.D.1 are not met, the vacuum pump shall be isolated.
4 3.8/4.8-6 Amendment No. 232,138 BFN Unit 1 l
c s
3.11/4.11 FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.A.
High Pressure Fire Protection 4.11.A.
High Pressure Fire System Protection System area where protection j
is lost is checked hourly.
3.
If only one high
- 3. Raw Service Water pressure fire pump is System Testing OPERABLE, the reactors may remain Item Frecuency in operation for a period not to exceed Simulated Once/ year 7 days, provided the automatic requirements of and manual Specification actuation 3.11.A.I.b above of raw service are met, water pumps s
and operation of tank level
- switches, j
4.
If Specification
- 4. The high pressure i
3.11.A.3 cannot be fire protection met, the reactors system pressure shall be placed in shall be logged the Cold Shutdown daily.
condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
5.
Removal of any
- 5. Principal header
)
component in the High and component i
Pressure Fire System isolation valves from service for any shall be checked reason other than open at intervals
]
testing or emergency no greater than operations shall 3 months.
require Plant Manager approval.
6.
The Raw Service Water
]
storage tank level j
shall be maintained above level 723'7" by the raw service water pumps.
3.11/4.11-3 BFN Amendment No. 138 Unit 1
e l
3.11/4.11 FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
/
3.11.A.
Hirh Pressure Fire Protection 4.11.A.
Hir.h Pressure Fire Protection System System 7.
If Specification 3.11.A.6 cannot be met a fire pump shall be started and run continuously until the raw service water pumps can maintain a raw service water storage tank level above 723'7".
8.
The fire protection water distribution system shall have e minimum capacity of 2664 spm at 250' head.
9.
The fire protection I
i system shall be capable of supplyint l
the individual loads 1
listed in Table 3.11.A.
B.
CO2 Fire Protection System B.
CO2 Figg_ Protection 1
S,,y tem 1.
The CO2 Fire 1.' CO2 Fire Protection Protection System Testing:
shall be OPERABLE:
Item Frecuency a.
With a minimum a.
Simulated Once/ year of B-1/2 tons automatic r.nd manual (0.5 Tank) CO2 in storage units actuation 1 and 2.
b.
With a minimum b.
Storage Checked of 3 tons (0.5 tank daily Tank) CO2 pressure storage unit 3.
and level 3.11/4.11-4 BFN Unit 1
4 4
1 3.11/4.11 FIRE PROTECTION SYSTEMS l
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l
l 3.11.B.
CO2 Fire Protection 4.11.B.
CO2 Fire Protection l
System System l
l 3.11.B.1 (Cont'd) 4.11.B.1 (Cont'd) c.
Automatic c.
CO2 Spray Once/3 initiation logic header and years OPERABLE.
nozzle inspection for blockage 2.
If Specifications 2.
When the cable 3.11.B.I.a or spreading room 3.11.B.1.b or CO2 Fire 3.11.B.1.c cannot be Protection met, a patrolling is INOPERABLE, fire watch with one 125-pound portable fire (or larger) equipment shall be portable fire established to ensure extinguisher that each area where shall be placed protection is lost at each entrance.
is checked hourly.
3.
If Specifications 3.11.B.1.a.
3.11.B.I.b, or 3.11.B.1.c are not met within 7 days, the affected unit (b) shall be in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.
If CO2 fire protection is lost to a cable spreading room or to any diesel generator building area a continuous fire watch shall be established immediately and shall be continued until i
CO2 fire protection is restored.
l l
l i
l BFN 3.11/4.11-5 Unit 1
)
l
a
.3.11/4.11 FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.B.
CO2 Fire Protection System 5.
Removal of any component in the CO2 Fire Protection System from service for any reason other than testing or emergency operations shall require Plant Manager approval.
C.
Fire Detectors 4.11.C.
Fire Detectors 1.
The fire detection 1
All heat and smoke system's heat and detectors shall be smoke detectors for tested in accordance all protected zones with industrial shall be OPERABLE standards or other except that one approved methods detector for a given semiannually.
protected zone may be INOPERABLE for a period no greater than 30 days.
2.
If Specification 2.
The non-Class A 3.11.C.1 cannot be supervised detector met, a patrolling circuitry for those fire watch will be detectors which established to ensure provide alarm only that *ach protected will be tested once zone or area with each month by INOPERABLE detectors actuating the l
is checked at detector at the und intervals no greater of the line or end than one each hour, of the branch such that the largest l
number of circuit conductors will be checked.
3.11/4.11-6 BFN Amendment No. 138 Unit 1 i
3.11/4.11' FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.11.C.
Fire Detectors 3.
The class A supervised detector alarm circuits will be tested once each two months at the local panels.
4.
The circuits between the local panels in 4.11.C.3 and the main control room will be tested monthly.
5.
Smoke detector sensitivity will be checked in accordance with manufacturer's instruction annually.
3.11.D.
ROVING FIRE WATCH D.
ROVING FIRE WATCH Deleted Deleted l
l 1
3.11/4.11-7 Amendment No. 138 BFN Unit 1
I 3.11/4.11 FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.E. Fire Frotection Systems 4.11.E.
Fire Protection Systems Inspection Inspections All fire barrier Each required fire penetrations, including barrier penetration shall cable penetration barriers, be verified to be J
fire doors and fire functional at least once dampers, in fire zone per 18 months by a visual boundaries protecting inspection, and prior to safety-related areas shall restoring a fire barrier be functional at all times.
to functional status With one or more of the following repairs or required fire barrier maintenance by i
penetrations nonfunctional performance of a visual within one hour establish inspection of the a continuous fire watch affected fire barrier on at least one side of penetration.
the affected penetration or verify the OPERABILITY of fire detectors on at least one side of the I
nonfunctional fire barrier and establish an hourly fire watch patrol until the j
4 work is completed and the barrier is restored to functional status.
)
F. Fire Protection Organization F.
Fire Protection Organization J
Deleted Deleted G. Air Masks and Cylinders G.
Air Masks and Cylinders A minimum of l'i air No additional masks and 30 500-cubic-inch surveillance required.
air cylinders shall be available at all times i
except that a time period of 48 hou s following emergency use is allowed to permit recharging or replacing.
3.11/4.11-8 BFN Amendment No. 138 Unit 1
r f
f 4
b 3.11 BASES q
e 1
The High Pressure Fire and CO2 Fire Protection specifications are provided
'in order to meet the preestablished levels of operability during a fire'in l
either or all of the three units. Requiring a patrolling fire watch with'
' portable fire equipment'if the automatic initiation is lost will provide (as
'does the automatic system) for early reporting and immediate fire fighting
]
' capability in the event of a fire occurrence.
i L
The High Pressure Fire Protection System is supplied by four pumps (three l
electric driven and one diesel driven) aligned to the high pressure fire header. The reactors may remain in operation for a period not to.ezceed 1
seven days if three pumps are out-of-service. If at least two pumps are not
)
made operable in seven days or if all pumps are lost during this seven-day l
period, the reactors will be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
For the areas of applicability, the fire protection water distribution system I
minimum capacity of 2,664 gpm at 250' head at the fire pump discharge consists of the following design loads:
2.
Sprinkler System (0.30 spm/ft /4440 ft2 area) 1332 spm 2
2.
I 1/2" Hand Hose Lines 200 gpm 3.
Raw Service Water Load 1132 apm TOTAL 2664 spm The CO2 Fire Protection System is considered operable with a minimum of 8 1/2 tons (0.5 tank) C02 in storage for units 1 and 2; and a minimum of 3 tons (0.5 tank) CO2 in storage for unit 3.
An immediate and continuous fire watch in the cable spreading room or any diesel generator building area will be established if CO2 fire protection is lost in this room and will continue until CO2 fire protection is restored.
To assure close supervision of fire protection system activities, the removal from service of any component in either the High Pressure Fire System or the CO2 Fire Protection System for any reason other than' testing or emergency operations will require Plant Manager approval.
Early reporting and immediate fire fighting capability in the event of a fire occurrence will be provided (as with automatic system) by requiring a patrolling fire watch if more than one detector for a given protected zone is inoperable.
l 1
3.11/4.11-12 BFN Amendment No. 138 Unit 1
3.11 BASES (Cont'd)
The fire protection system is designed to supply the required flow and pressure to an individual load listed on Table 3.11.A while maintaining a design raw service water load of 1132 spm.
4.11 BASES Periodic testing of both the High Pressure Fire System and the CO2 Fire Protection System will provide positive indication of their operability.
If only one of the pumps supplying the High Pressure Fire System is operable, the pump that is operable will be checked immediately and daily thereafter to demonstrate operability. If the CO2 Fire Protection System becomes inoperable in the cable spreading room, one 125-pound (or larger) fire extinguishers will be placed at each entrance to the cable spreading room.
Annual testing of automatic valves and control devices is in accordance with NFPA Code Vol. II, 1975, section 15, paragraph 6015. More frequent testing would require excessive automatic system inoperability, since there are a large number of automatic valves installed and various portions of the system must be isolated during an extended period of time during this test.
Wet fire header flushing, spray header inspection for blockage, and nozzle inspection for blockage will prevent, detect, and remove buildup of sludge or other material to ensure continued operability. System flushes in conjunction with the semiannual addition of blocide to the Raw Cooling Water System will help prevent the growth of crustaceans which I
could reduce nozzle discharge.
Semiannual tests of heat and smoke detectors are in accordance with NFPA Code.
With the exception of continuous strip heat detectors panels, all i
non-class A supervised detector circuits which provide alarm only are j
hardwired through conduits and/or cable trays from the detector to the main control room alarm panels with no active components between.
Nonclass A circuits also actuate the HPCI water-fog system, the CO2 system in the diesel generator buildings, and isolate ventilation in shutdown board rooms. The test frequency and methods specified are justified for the following reasons:
1.
An analysis was made of worst-case fire detection circuits at Browns Ferry to determine the probability of no undetected failure 1
of the circuits occurring between system test times as specified in the surveillance requirements. A circuit is defined as the wire connections and components that affect transmission of an alarm signal between the fire detectors and the control room annunciator.
Three circuits were analyzed which were representative of an alarm-only circuit, a water-fog circuit, and a CO2 circuit. The spreading room B smoke detector was selected as the worst-case alanm-only circuit because it had the largest number of wires and connections in a single circuit. The HPCI water-fog circuit was selected for analysis because it is the only water-fog circuit in BFN 3.11/4.11-13 Unit 1
9 4-4.11 BASES (Cont'd)
{
l l
i the area of applicability'for Technical Specifications. The Standby Diesel Generator Room A CO2 circuit was selected because it contained 2 out of 3 detector logic, the most complicated CO2 circuit logic. Calculations were based on failure rates for wires, connections, and circuit components as shown in Appendix III of WASM-1400.
Failure rates were considered for the following circuit components:
1.
Open circuit 2.
Short to ground 3.
Short to power 4.
Timing motor failure to start 5.
Relay failure to energize 6.
Normally open contact failure to close 7.
Normally open or normally closed contact short 8.
Normally closed contact opening 9.
Timing switch failure to transfer The calculated probabilities (Pf) for no undetected failure of the i
circuits occurring were as follows, based on the specified test I
frequency.
l AREA TEST FREQUENCY Pf Spreading Room B One Month 0.975287 HPCI Water Fog Six Months 0.977175 Six Months 0.957595 Standby Diesel Gen Room A CO2 The worst case of the three areas considered is Spreading Room B.
The probability of undetected failure is approximately 1/40, which means that one undetected failure will occur on the average every i
40 months over an extended period of time and that the failure could exist up to one month. The frequency of testing is thus much greater than the frequency of failure and produces circuits with adequate reliability.
2.
Circuits checks by initiation of end of the line or end of the branch detectors will more thoroughly test the parallel circuits than testing on a rotating detector basis. This test is not a detector test, but is a test to simulate the effect of electrical supervision as defined in the NFPA Code.*
3.
Testing of circuits which actuate CO, water, or ventilation 2
systems requires disabling the automatic feature of the fire protection system for the area. A surveillance program which disabled these circuits monthly would significantly reduce the ability of these circuits to provide fire suppression.
- Ref: NFPA Code 72D-9, paragraph Illi, Code 72D-15, paragraph 1312 for definition of Class A systems, and Code 72A-18, Article 240.
BFN 3.11/4.11-14 Unit 1
4.11 BASES (Cont'd) 4.
Daily tests of annunciation lights and audible devices are performed as a routine operation function.
5.
The CO2 system manufacturer reconenends semiannual testing of CO2 system fire detection circuits.
In addition, operating personnel periodically inspect the plant during their normal operating activities for fire hazards and other abnormal conditions.
Smoke detectors will be tested "in-place" using inert freon gas applied by a pyrotronics type applicator which is accepted throughout the industrial fire protection industry for testing products of combustion detectors or by use of the NSA chemical smoke generators. At the present time, the manufacturers have only approved the use of " punk" for creating smoke. TVA will not use
" punk" for testing smoke detectors.
3.11/4.11-15 Amendment No. 138 BFN Unit 1
BFN TECHNICAL SPECIFICATIONS 6.0 ADMINISTRATIVE CONTROLS 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY The Plant Manager has onsite responsibilities for the safe operation of the In the absence of facility and shall report to the Browns Ferry Site Director.
the Plant Manager, a Plant Superintendent will assume his responsibilities.
6.2 ORGANIZATION CORPORATE 6.2.1 The portion of TVA management which relates to the operation of the plant is shown in Figure 6.2-1.
PLANT STAFF The functional organization for the operation of the plant shall be shown 6.2.2 in Figure 6.2-2, Shift manning requirements, shall as a minimum, be as described in Table a.
6.2.A and below.
A licensed senior reactor operator shall be present at the site at all b.
times when there is fuel in the reactor.
6.0-1 Amendment No.138 BFN Unit 1
y 6.2.2 (Cont.)
A licensed reactor operator shall be in the control room whenever c.
there is fuel in the reactor.
d.
Two licensed reactor operators shall be in the control room during any cold startups, while shutting down the reactor, and during recovery from unit trip. In addition, a person holding a senior operator license shall be in the control room for that unit whenever it is in an operational mode other than cold shutdown or refueling.
A Health physics Technician
- shall be present at the facility at e.
all times when there is fuel in the reactor.
f.
A person holding a senior operator license or a senior operator license limited to fuel handling, shall be present during alteration of the core to directly supervise the activity and I
during this time shall not be assigned other duties.
g.
A site fire brigade of at least five members shall be maintained onsite at all times.* The fire brigade shall not include the Shift Engineer and the other members of the minimum shift crew 1
i necessary for safe shutdown of the unit, nor any personnel required for other essential functions during a fire emergency.
- The Health physics Technician and fire brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to acconunodate unexpected absence, provided insnediate action is taken to fill the required positions.
6.0-2 Amendment No. 138 BFN Unit 1 l
.l
c.
=
1.
Table 6.2.A Minimum Shift Crew Recuirementsb position Units in Operation Type of License E
1 1
1 i
Senior Operatora 1
1 1
1 SRO Senior Operator 0-1 2
2 SRO Licensed Operators 3
3 3
0 1
2 2
Additional Licensed Operators Assistant-Unit Operators (AUO) 4 4
5 5
None Shift Technical Advisor (STA) 0 1
1 1
None Health physics Technician 1
1 1
1 None 1
Note for Table 6.2.A a.
A' senior operator will be assigned responsibility for overall plant operation at all times there is fuel in any unit.
b.
Except for the senior operator discussed in note "a", the shift crew composition may be one less than the minimum requirements of Table 6.2.A for a period of time not to exceed two hours 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 of Table 6.2.A.
This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent, c.
One of the Additional Licensed Operators must be assigned to each control room with an operating unit.
d.
The number of required licensed personnel, when the operating units are controlled from a common control room, are two senior operators and four operators.
6.0-3 BFN Amendment No. 138 Unit I
_ - _ _ _ _ _ _ _ _ _ - _ = _ _ _ - _ _ _ _
6.3 PLANT STAFF QUALIFICATIONS Qualifications of the Browns Ferry Nuclear Plant management and operating staff shall meet the minimum acceptable levels as described in ANSI -
N18.1, Selection and Training of Nuclear Power Plant Personnel, dated l
March 8, 1971. The qualifications of the Health Physics Supervisor will j
meet or exceed the minimum acceptable levels as described in Regulatory f
Guide 1.8, Revision 1, dated September 1975. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and transient and accident response and analysis.
f l
6.4 TRAINING l
A retraining and replacement training program for station personnel sha33 be in accordance with ANSI - N18.1, Selection and Training of Nuclear Power Plant Personnel, dated March 8, 1971. The minimum frequency of the retraining program shall be every two years.
1 6.5 PLANT REVIEW AND AUDIT I
i I
6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) l FUNCTION The PORC shall function to advise the Plant Manager in all 6.5.1.1 a.
matters related to nuclear safety.
b.
This advisory function shall be performed by the PORC acting in a formal meeting or by members acting individually without j
a formal meeting.
6.0-4 Amendment No. 138 BFN Unit 1
COMPOSITION 6.5.1.2 The PORC shall be composed of the:
a.
Chairman:
Plant Manager Alternate Chairman:
Assistant to Plant Manager Alternate Chairman or Member: Technical Services Superintendent Member:
Unit Superintendents (3)
Member:
Maintenance Superintendent Member:
Quality Assurance Staff Supervisor Member:
Health Physics Supervisor i
b.
All alternate chairmen and alternate members shall be appointed in writing by the PORC chairman.
MEETING FREQUENCY 6.5.1.3 The PORC shall convene in a formal meeting at least once a month and as directed by the chairman. Other PORC meetings may be requested by the chairmen or members as required.
6.0-5 Amendment No. 138 BFN Unit 1
\\
6.5.1.4 For expedited meetings, when it is not practical to convene as a group, the chairman or alternate chairman may conduct committee business by polling the members individually (by telephone or in person) or via a serialized review.
QUORUM 6.5.1.5 The' quorum necessary for the PORC to act in a formal meeting shall consist of the chairman or alternate chairman and at least five members or their alternates. Members shall be considered present if they are in telephone communication with the committee.
RESPONSIBILITIES 6.5.1.6 The PORC shall be responsible for the activities listed below.
The PORC may delegate the performance of reviews, but will maintain cognizance over and responsibility for them, e.g.,
subcommittees.
a.
Review of administrative procedures for the control of the technical and cross-disciplinary review of (1) all procedures required by Specification 6.8.1.1, and changes thereto, (2) any other procedures and changes thereto determined by the Plant Manager to affect nuclear safety, b.
Review of the administrative procedures required by Appendix A of Regulatory Guide 1.33, Revision 2, j
i February 1978 and changes thereto.
l c.
Review of emergency operating procedures and changes thereto.
3 1
i d.
Review implementing procedures of the Radiological Emergency Plan and the Industrial Security Program.
Amendment No. 138 j
BFN Unit 1 6.0-6
i
- e-
-I'-
e.
Reviev of all proposed changes to the Technical
-l Specifications.
1 f.
Review of. safety evaluation for proposed tests or experiments to be completed under.the provisions cf 10 CFR
- c 50.59
- g.. Review proposed changes-to the Radiological Effluent Manual.
- h. -Review adequacy of the Process Control Program'and Offsite Dose Calculation Manual at least once every 24 months.
~ i.- Review changes to the radwaste treatment systems.
I j.- Review of every unplanned onsite release'of radioactive 1
material to the environs including the preparation and forwarding of reports covering evaluation, recommendation, and disposition of the corrective action to prevent i
recurrence to the Director, Nuclear Power and to the Nuclear Safety Review Board.
~ k.
Review of all safety evaluations for modifications to structures, systems or components that affect nuclear. safety
-to verify that such actions did not constitute an unreviewed safety question as defined in 10 CFR 50.59, or requires a change to these Technical Specifications, Amendment No. 138 f-BFN l
. Unit 1 6.0-7
1.
Review of reportable events, unusual events, operating anomalies, and abnormal performance of plant equipment.
Investigate reported or suspected incidents involving safety m.
questions or violat'ons of the Technical Specifications.
Review of unit operations to detect potential hazards to n.
nuclear safety. Items that may be included in this review are NRC inspection reports, QA audit, NSRB audit results.
American Nuclear Insurer (ANI) inspection results, and significant corrective action reports (CARS).
performance of special reviews, investigations, or analysis, o.
and report thereon as requested by the Plant Manager or the Nuclear Safety Review Board.
l l
l 6.0-8 BF Amendment No. 138
4 AUTHORI'IY 6.5.1'7 The PORC shall:
a.
Recommend to the Plant Manager in writing, approval, or disapproval of items considered under 6.5.1.6.a through i above.
1.
The recommendation shall be based on a majority vote of the PORC at a formal meeting.
2.
The recommendation shall be based on a unanimous vote of the PORC when the PORC members are acting individually.
3.
Each member or alternate member shall have one vote, b.
Furnish for consideration a determination in writing with
. regard to whether or not each item considered under 6.5.1.6.f above constitutes an unreviewed safety question.
c.
Make recommendations to the Plant Manager in writing concerning whether action reviewed under 6.5.1.6.k above did not constitute an unreviewed safety question.
d.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Site Director and the Nuclear Safety Review Board of disagreements between the PORC and the Plant Manager.
However, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.
BFN 6.0-9 Amendment No. IJ8 Unit 1 l
l-
1
' RECORDS 6.5.1.8 The PORC shall maintain written minutes of each PORC meeting including expedited meetings that, as a minimum, document the result of all PORC activities performed under the responsibility and authority provisions of these technical specifications.
Copies shall be provided to che Site Director and the Nuclear Safety Review Soard, 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB)
FUNCTION 6.5.2.1 The NSRB shall function to provide independent review and audit cognizance of designated activities in the areas of:
a.
Nuclear power plant operations b.
Nuclear engineering c.
Chemistry and radiochemistry d.
Metallurgy e.
Instrumentation and control f.
Radiological safety g.
Mechanical and electrical engineering, and
)
h.
Quality assurance practices i
COMPOSITION i
6.5.2.2 The NSRB shall be composed of at least five members, including I
the Chairman. Members of the NSRB may be from the Office of Nuclear Power or other TVA organizations, or external to TVA.
i-l-
6.0-10 f
BFN Amendment No. 138 Unit 1
(-
o
)
QUALIFICATIONS-1 6.5.2.3 The Chairman, members, alternate members of the NSRB shall be appointed in writing by the Manager of Nuclear Power and shall have an academic degree in engineering or a physical science
)
field, or the equivalent; and in addition, shall have a minimum of 5 years technical experience in one or more areas given in 6.5.2.1.
No more than two alternates shall participate as voting members in NSRB activities at any one time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized to provide expert advice as determined by the NSRB.
MEETING FREQUENCY 6.5.2.5 1he NSRB shall meet at least once per six months.
l QUORUM 6.5.2.6 The minimum quorum of the NSRB necessary for the performance of the NSRB review and audit functions of these technical specifications shall consist of more than half of the NSRB membership or at least five members, whichever is greater. The quorum shall include the Chairman or his appointed alternate and the NSRB members including appointed alternate members meeting the requirements of 6.5.2.3.
No more than a minority of the quorum shall have line responsibility for operation of the unit.
6.0-11 Amendment No. 138 BFN Unit 1
_________.____________________________________j
i i
REVIEW j
6.5.2.7 The NSRB shall review:
a.
The safety evaluations for:
(1) changes to procedures, equipment or systems, and (2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
J b.
Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
I c.
Proposed tests or experiments which involve an unreviewed safety guestion as defined in Section 50.59, 10 CFR.
1 d.
Proposed changes to Technical Specifications or this Operating License.
1 e.
Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or' instructions having nuclear safety significance.
f.
Significant operating abnormalities or deviations from l
normal and expected. performance of plant equipment that affect nuclear safety.
g.
All Reportable Events h.
All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and 1
1.
Reports and meeting minutes of the PORC.
6.0-12 I
BFA Amendment No. 138 Unit 1 L______________________
AUDITS i
6.5.2.8 Audits of unit activities shall be performed under the j
cognizance of the NSRB. These audits shall encompass.
1 The conformance of plant operation to provisions contained a.
within the Technical Specifications and applicable license conditions at least once per 12 months.
b.
The performance, training and qualifications of the entire plant staff at least once per 12 months.
c.
The results of actions taken to correct deficiencies q
occurring in site equipment, structures, systems or method
)
of operation that affect nuclear st.fety at least once per 6
- months, d.
The performance of activities required by the Operational Quality Assurance program to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months.
e.
The Site Radiological Emergency Plan and implementing procedures at least once every 12 months.
f.
The Plant Physical Security Plan and implementing procedures at least ence every 12 months.
I g.
Any other area of site operation considered appropriate by the NSRB or the Manager of Nuclear Power.
h.
The fire protection programmatic controls including the implementing procedures at least ones per 24 months.
6.0-13 BFN Amendment No. 138 Unit 1 1
)
- i. 'An independent fire protection and loss prevention program inspection and audit shall be performed annually utilizing either qualified offsite license personnel or an outside fire protection firm.
J.
An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervals no greater than 3 years.
k.
The Radiological Environmental Monitoring program and the results thereof at least once per 12 months, i
1.
The performance of activities required by the Quality
)
Assurance Program to meet the criteria of Regulatory Guide
+
4.15, December 1977,'or Regulatory Guide 1.21 Rev. 1, 1974, and Regulatory Guide 4.1, 1975, at least once every 12 months.
m.
The performance of activities required by the Safeguards j
Contingency Plan to meet the criteria of 10 CFR 73.40(d) at j
least once every 12 months.
n.
The Offsite Dose Calculation Manual and implementing procedures at least once per 24 months.
I o.
The Process Control Program and implementing procedures for solidification of wet radioactive wastes at least once per 24 months.
p.
The Radiological Effluent Manual and implementing procedures at least once per 12 months.
l 6.0-14 BFN Unit 1 Amendment No. 132, 138 I
l s
1
AUTHORITY 6.5.2.9 The NSRB shall report to and advise the Manager of Nuclear Power on those areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8.
RECORDS 6.5.2.10 Reports of activities shall be prepared, approved, and distributed as indicated below:
a.
Minutes of each NSRB meeting shall be prepared, approved and forwarded to the Manager of Nuclear Power within 14 days following each meeting.
b.
Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepered, approved and forwarded to the Manager of Nuclear Power within 14 days following completion of the
- review, c.
Audit reports encompassed by Specification 6.5.2.8 above, shall be forwarded to the Manager of Nuclear Power and to the management positions responsible for the areas audited within 30 days after completion of the audit.
6.0-15 BFN Amendment No. 3 38 Unit 1
6.5.3 TECHNICAL REVIEW AND APPROVAL OF PROCEDURES ACTIVITIES 6.5.3.1 Procedures required by Technical Specification 6.8.1.1 and other procedures which affect plant nuclear safety, and changes (other than editorial or typographical changes) thereto, shall be prepared, reviewed and approved. Each procedure or procedure change shall be reviewed by an individual other than the preparer. The reviewer may be from the same organization or from a different organization. Procedures other than Site Director Standard-Practices will be approved by the responsible Section Supervisor, or applicable Plant Superintendent.
6.5.3.2 Proposed changes or modifications to plant structures, systems and components that affect nuclear safety shall be reviewed as designated by the Plant Manager.
Each such modification shall be reviewed by an individual / group other than the individual / group which designed the modification, but who may be from the same organization as the individual / group which designed the modification. Proposed modifications to plant structures, systems and components that affect nuclear safety shall be approved by the Plant Manager, prior to implementation.
6.5.3.3 Individuals responsible for reviews performed in accordance l
with 6.5.3.1 shall be members of the site supervisory staff previously designated by the Plant Manager.
Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary.
If
)
deemed necessary, such review shall be performed by review j
'I personnel of the appropriate discipline.
6.5.3.4 The Plant Manager shall approve all administrative procedures requiring PORC review prior to implementation.
l l
BFN Unit 1 6.0-16 Amendment No. 138
c 1
6.6 REPORTABLE EVENT ACTION i
6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
i i
i a.
The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and i
b.
Each REPORTABLE EVENT shall be reviewed by the PORC and the results of this review shall be submitted to the NSRB and j
the Site Director.
i t
I 6.0-17 BFN Amendment No. 13E Unit 1
6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
The NRC Operations Center shall be notified by telephone as a.
soon as possible and in all cases within I hour. The Manager of Nuclear Power and the NSRB shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I 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 the c.
Consnission, the NSRB and the Manager of Nuclear Power within 14 days of the violation.
d.
Critical operation of the unit shall not be resumed until authorized by the Commission.
1
=
l 6.0-18 Amendnent No.138 1
BFN Unit 1 l
e 6.8 PROCEDURES / INSTRUCTIONS AND PROGRAMS 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, Febniary 1978.
b.
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.
Security plan implementation.
e.
Emergency plan implementation.
f.
Fire Protection Program implementation, g.
Radiological Effluent Manual implementing procedures.
h.
Process Control Program (PCP).
{
I
- j. Administrative procedures which control technical and cross-disciplinary review.
t 6.8.1.2 Each administrative procedure required by Section 6.8.1.1.a.
l shall be reviewed by PORC and all other procedures required i
I
(
by Section 6.8.1.1.a. shall be reviewed in accordance with Section 6.5.3.
Amendment No.138 BFN l
l Unit 1 6.0-19
}
l
6.8.1.3 Temporary changes to procedures of Specification 6.8.1.1 may be made provided:
- a. -The intent of the original procedure is not altered; b.
The change is approved by two members of the plant management' staff, at least one of whom holds a Senior Operator License on the unit affected; c.
The change is documented, reviewed by the PORC and approved by the Plant Manager within 14 days of implementation, for changes in administrative procedures requiring PORC review.
d.
The change is documented, reviewed per Specification 6.5.3, and approved by the responsible group section supervisor within 14 days of implementation, for changes to procedures other than administrative procedures.
DRILLS 6.8.2 Drills on actions to be taken under emergency conditions i
i involving release of radioactivity are specified in the i
Radiological Emergency Plan and shall be conducted
)
I annually. Annual drills shall also be conducted on the actions to be taken following failures of safety-related systems or components.
3 RADIATION CONTROL PROCEDURES i
6.8.3 Radiation Control Procedures shall be maintained and made available to all station personnel. These procedures shall show permissible radiation exposure and shall be consistent 6.0-20 BFN Amendment No.138 4
Unit 1
with the requirements of 10 CFR 20.
This radiation protection program shall be organized to mest the requirements of 10 CFR 20 except in-lieu of the " control device" or " alarm signal" required by paragraph 20.203 (c) of 10 CFR 20.
J 6.8.3.1 Each high radiation area in which the intensity of radiation is greater than 100 meem/hr but less than 1000 mrem /hr shall be i
barricaded and conspicuously posted as a high radiation area and j
entrance thereto shall be controlled by requiring issuance of a Radiological Work Permit.* 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 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 arem/br shall be subject to the provisions of (1) above; and, in addition, access to the source and/or ares i
l 6.0-21 BFN Anendment No. 130 Unit 1
shall be secured by lock (s), The key (s)'shall be under the administrative control of the shift engineer.
InLthe case of a high radiation area established for a period of 30 days or less, direct surveillance to prevent unauthorized entry may'be 3
. substituted for permanent access control.
- Health Physics personnel, or personnel escorted by Health
' Physics personnel, in accordance.with approved emergency l
procedures, shall be exempt from the RWP issuance requirement i
I during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection f
procedures for entry into high radiation areas, QUALITY ASSURANCE PROCEDURES - EFFLUENT AND ENVIRONMENTAL MONITORING b.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 Reguir.*ory Guide 4.1, Rev. 1,' April 1975 or Regulatory Guide 4.15. Dec. 1977.
4 l
l l
6.0-22 BFN Unit 1 Amsndment No. J32,138
l' 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title l
10, code of Faderal Regulations, the followinr, identified reports shall be submitted to the Director of the Regional Office of NRC, unless otherwise noced.
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 l
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 obtained during the test program and a comparison of these values with design I
predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.
1 4
s 6.0-23 Amendment No. 138
!$$t1
~w__-----_--
i t
i b.
Startup reports shall be st.br. 't.ted within (1) 90 days following completion of the 7,tartup test program, (2) 90 days following resumption or commencement of conenercial 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, l
completion of startup test program, and resumption or j
commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have beer completed.
6.9.1.2 ANNUAL 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 mrem /yr 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 80% of the total whole body dose 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 operater action to control reactor pressure shall be reported.
- A single submittal may be made for a multiple unit station.
- This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.
6.0-24 BFN Amendment No. 138 Unit I
6.9.1.3 MONTHLY.0PERATING REPORT Routine reports of operating statistice 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.
RADI0 ACTIVE EFFLUENT RELEASE REPORT 6.9.1.5 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.
l l
6.0-25 BFN Unit 1 Amendment No. 132,138
1 6.9.2 SPECIAL REPORTS l
1 Reports on the following areas shall be submitted in writing to the Director of Regional Office of Inspection and Enforcement:
1.
Fatigue Usage 6.10.1.g Annual ~
)
Operating Report
]
2.
Relief Valve Tailpipe 3.2.F Within 30 days
)
after inoper-i ability of thermocouple and acoustic monitor on one valve.
l
)
3.
Seismic Instrumentation 3.2.J.3 Within 10 days Inoperability after 30 days of inoperability.
1 l
4 Meteorological Monitoring 3.2.I.2 Within 10 days Instrumentation after 7 days of Inoperability inoperability.
1 i
5.
primary Containment 4.7.A.2 Within 90 days Integrated Leak. Rate of completion of j
Testing each test.
l 6.
Data shall be retrieved from all seismic instruments actuated during a seismic event and analyzed to determine i
the magnitude of the vibratory ground motion. A Special 6.0-26 BFN Amendment No. 138 Unit 1
h.
~ '
,,g
.I l
~
Report shall be submitted within 10 days after the event
{
describing the magnitude, frequency spectrum, and resultant i
effect upon plant features important to safety.
7.
Secondary Containment 4.7.C.
Within 90 days i'
Leak Rate Testing
- of completion of each test.
- Each integrated leak rate test of the secondary containment shall be the subject of a summary technical report. This report should include data on the wind speed, wind direction, outside and inside temperatures during the test, concurrent reactor building pressure, and emergency ventilation flow rate. The report shall also include analyses and interpretations of those data which demonstrate compliance with the specified leak rate limits.
6.0-27 BFN Amendment No. 138 Unit 1
6.10 STATION OPERATING RECORDS AND RETENTION 6.10.1 Records and/or logs shall be kept in a manner convenient for review as indicated below:
All normal plant operation including such items as power a.
I level, fuel exposure, ar.d shutdowns b.
Principal maintenance activities 1
c.
Reportable Events 1
J d.
Checks, inspections, tests, and calibrations of components l
and systems, including such diverse items as source leakage i
Reviews of changes made to the procedures or equipment or j
e.
reviews of tests and experiments to comply with 10 CFR 50.59 f.
Radioactive shipments g.
Test results in units of microcuries for leak tests performed pursuant to Specification 3.8.D i
l i
j l
l 6.0-28 Amendment No. 13E BFN Unit 1
c..
h.
Record of-annual physical inventory verifying accountability.
.of sources on record 1.
Gaseous and liquid radioactive waste released to the environs-
- j. 0ffsite environmental monitoring surveys k.
Fuel inventories and transfers 1.
Plant radiation and contamination surveys m.
Radiation exposures for all plant personnel Updated, corrected, and as-built drawings of the plant n.
Reactor' coolant system inservice inspection o.
p.
Minutes of meetings of the NSRB g.
Design fatigue usage evaluation Monitoring and recording requirements below will be met for various portions of the reactor coolant pressure boundary (RCPB) for which detailed fatigue usage evaluation per the ASME Boiler and' Pressure Vessel Code Section III was performed for the conditions defined in the derign specification.
In this plant, the applicable codes require fatigue usage evaluation for the reactor pressure vessel only. The locations to be monitored shall be:
1.
The feedwater nozzles l
2.
The shell at or near the waterline 3.
The flange studs 6.0-29 Amendment No. 138 BFN -
' Unit 1
_ _ _ _ _ _ _ _ - _. __ _ _j
s Transients that occur during plant operations.will be reviewed and.a cumulative fatigue usage factor determined.
For transients which are more severe than the transients i
evaluated-in the stress report, code fatigue usage calculations will be made and tabulated separately.
In the annual operating report, the fatigue usage factor determined for the transients defined above shall be added and a cumulative fatigue usage factor to date shall be reported. When the cumulative usage factor reaches a value of 1.0, an inservice inspection shall be included for the specific location at the next scheduled inspection (3-1/3-year interval) period and 3-1/3-year intervals thereafter, and a subsequent evaluation performed in accordance with the rules of ASME Section l
XI Code if any flaw indications are detected. The results of the evaluation shall be submitted in a i
Special Report for review by the Commission.
6.10.2 Except where covered by applicable regulations, items a through h above shall be retained for a period of at least 5 years and item i through q shall be retained for the life of the plant. A complete inventory of radioactive materials in possession shall be Maintained current at all times.
l l
l 1.
See paragraph N-415.2, ASME Section III, 1965 Edition.
l 6.0-30 l
l I
l BFN Amendment No.138 Unit 1 Y______-_-______
o i
l 6.11 PROCESS CONTROL PROGRAM (PCP)
- 1. The'PCP shall be approved by the Commission prior to Implementation.
- 2. Changes to ;sie PCP shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
a.
Sufficiently detailed information to totally support the change.
b.
A determination that the change did not change the overall conformance of the solidified product to existing criteria.
6.12 0FFSITE DOSE CALCULATIONAL MANUAL (ODCM)
- 1. The ODCM shall be approved by the Commission prior to implementation.
- 2. Changes to the ODCM shall be submitted to the Commission in the semi-annual Radioactive Effluent Release' Report for the period in which the change (s) was made. This submittal shall contain:
a.
Sufficiently detailed information to totally support the change.
6.0-31 BFN Unit 1 Amendment No. 132. 138
-. _. - - - - - - _. _ _ _. _ _ - _ ~
3 6.13 RADIOLOGICAL EFFLUENT MANUAL (REM) l i
- 1. The REN shall be approved by the Conunission prior to implementation.
- 3. Changes to the REM shall be approved by the Commission prior to l
1 l
implementation, I
l I
i 6.0-32 BFN Unit 1 Amendment No. 132.138
i)!!lj l1 l1 l
.e AC r
EU mL o
Cr Ua p
N p
uS la R RG c
O AN i
T EI n
CF LN h
EOCIA RI UR c
D NT eT t
n R RS e
O AE n
T EC r
CF LtV e
EOCR R UE g
ID NS a
n a
M y
n RM o
EE G
t t
i N
WG R
RI l
EAa U PA T EE c
U T
A O N O AR i
o OV M
CF nr LE a
ANOR n
oN EOCN o
AY A ET R
UI F
s M LU I
N G C
D N
P UE E
r ND o
f n
o YTD it ER R R F A a
A O O A S B T E z
CF LA i
N9 W
EOCO n
E R U M
D N a
I E
g t
r O
e t
N i
O s
4 R RT f
I O AC f
TCFEU O
L R EOCT R US 1
ID NN O
C 2
6 er u
sm m "^
g o2 i
w" F
r e
o""
r*"
r rs D72 m
- O{w o3OB$ +r h Cco 1:
C-4 C3*t r e I'
il-
,i l ll l
I Es o
D m
NGR A NE I
C 0" N SG NA ME c
EN CA u
UM
=
M G
A N
R I
G R
O E
RE N
O I
E PS N
- n D
R N
N Ms NE I
NE c
G T
G 0C u
N N
T. U R
A E
1 RN E
I A A*
gD r
C i
N E RE D
TOS N
D O M IRY I
]E R A E T
OT N N
c TT I
P R C R
SE E
E I
UF P
Y E P U U
DA S
R RO R R
N S T
FI O
T R
S T
N I
I I
R O A E
SR A
T T TG M
O C R EE SNA l
E F A E CG I AN H
E P N
A SLA L
C Y R O O
N SPM A
T I
C uR R T
E A A
IG O I I T L
BO O 2 A l
SM S
Z I
O R SNC 2, AN I
R L T NEA E
I O
I N OSE O
8 PR G
T DO S
ER S E P
AC I
RO RO-E R
U RO GY R
I NG C
AA A
LN F
PAM T
)
N O
E E CD R
NN S
A L
(
E-E NT RG I
EN R RR TI T
RAN E
TR NR FI O O O EI
)
I N E
E I
L R CE A
3 T T I
H
)N A A P
C C E EE MU T
(
S OIG)R R E U E JN S
RNOEE N
~
E,e
~
(ER P P S
RNI R N (O O G
PE RFRT T I
T DFN OE EI O NI I
EH N
T S
NSAUU R
GTR T I
Y O
NNENR S
T EAPAA F A M
I TS SU
)
F TOT 3
RO Q I
1 I
HR FITI I R
(
(
- SISX O
E SC HSNSU TN R E G P
SAUAA CO A A
O U EI E RlSL 5 N g
C A
I v DiU M
D N v?wb gggk 2oL$
E#~
Sp Ricy
'o UNITED STATES l/
g NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20$55
%..,* /
TENNESSEE VALLEY AUTH0 PITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.134 License No. DPR-5?
1.
-The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the l
licensee) dated September 27, 1984 as supplemented January 17, June ?
and December 10, 1986, complies with the standards and requirements 1
of the Atomic Energy Act of 1054, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I:
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The is'suance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
i 2.
Accordingly, the license is amended by changes to the Technical i
Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(?) of Facility Operating License No. DPR-52 is hereby amended to read gs follows:
l
2 i
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.134, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 90 days froa1 the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION CWE
+
irector hn A. Zwolinski, As f
Projects TVA Projects Division Office of Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: September 11, 1987 l
1 i
r I
p-i l-!
I
. ATTACHMENT TO LICENSE AMENDMENT NO. 134 FACILITY OPERATING NO. DPR-52
)
1 DOCKET NO. 50-260 Revise the Appendix A' Technical Specifications by removing the pages identified below and inserting the enclosed pagt4, -The revised pages are identified by the i
captioned amendment number and contain marginal lines indicating the area of j
change. Overleaf pages* are provided to maintain document completeness.
Remove Insert iii iii*
iv iv v
v vi-vi*
.vii vii viii viii 3.2/4.2-3 3.2/4.2-3 3.2/4.2-4 3.2/4.2-4*
3.2/4.2-5 3.2/4.2-5
.3.2/4.2-6 3.2/4.2-6 i
3.8/4.8-1 3.8/4.8-1*
3.8/4.8-2 3.8/4.8-2 3.8/4.8-3 3.8/4.8-3 3.8/4.8-4 3.8/4.8-4 3.8/4.8-5 3.8/4.8-5 3.8/4.8-6 3.8/4.8-6 3.11/4.11-3 3.11/4.11-3 3.11/4.11-4 3.11/4.11-4*
3.11/4.11-5 3.11/4.11-5*
3.11/4.11-6 3.11/4.11-6 3.11/4.11-7 3.11/4.11-7 3.11/4.11-8 3.11/4.11-8 3.11/4.11-12 3.11/4.11-12 3.11/4.11-13 3.11/4.11-13*
3.11/4.11-14 3.11/4.11-14*
3.11/4.11-15 3.11/4.11-15 q
6.0-1 thru 6.0-23 6.0-1 thru 6.0-34 l
' S^ction Pern No.
C.
Coolant Leakage.
3.6/4.6-9 D.
Re?.ief Valves.
3.6/4.6-10 E.
Jet Pumps...................
3.6/4.6-11 F..
Recirculation Pump Operation....
3.6/4.6-12 G.
Structural Integrity..
3.6/4.6-13 H.
Seismic Restraints, Supports and Snubbers..
3.6/4.6-15 3.7/4.7 Containment Systems............
3.7/4.7-1 A.
3.7/4.7-1
.B.
Standby Gas Treatment System....
3.7/4.7-13 C.
3.7/4.7-16 D.
Primary Containment Isolation Valves.....
3.7/4.7-17 E.
Control Room Emergency Ventilation.
3.7/4.7-19 F.
Primary Containment Purge System.
3.7/4.7-21 G.
Containment Atmosphere Dilution System (CAD) 3.7/4.7-22 H.
Containment Atmosphere Monitoring (CAM)
System H2 Analyzer........
3.7/4.7-24
~
3.8/4.8 Radioactive Materials.
- 3. 8/ 4. 8 -1 A.
Liquid Effluents...............
3.8/4.8-1 B.
Airborne Effluents...
3.8/4.8-2 C.
Radioactive Effluents - Dose.
3.8/4.8-6 D.
Mechanical Vacuum Pump.
- 3. 8 / 4. 8-6 E.
Miscellaneous Radioactive Materials Sources.
3.8/4.S-7 F.
Solid Radwaste.
3.8/4.8-9 3.9/4.9 Aurillary Electrical System...
- 3. 9 / 4.9-1 A.
Auxiliary Electrical Equipment.
3.9/4.9-1 B.
Operation with Inoperable Equipment.
3.9/4.9-8 l'
C.
Operation in Cold Shutdown.
3.9/4.9-15 3.10/4.10 Core Alterations....
3.10/4.10-1 A.
Refueling Interlocks.............
3.10/4.10-1 111 BFN Unit 2 Amendment ho.128
..w____.-__._
l l
l Section Page No.
B.
Core Monitoring...
3.10/4.10-4 C.
Spent Fuel Pool Water.........
3.10/4.10-7 1
D.
Reactor Building Crane..
3.10/4.10-8 1
l l
E.
Spent Fuel Cask.
3.10/4.10-9 F.
Spent Fuel Cask Handling-Refueling Floor...
3.10/4.10-10 3.11/4.11 Fire Protection Systems..............
3.11/4.11-1 A.
High Pressure Fire Protection System.....
3.11/4.11-1 B.
CO2 Fire Protection System.
3.11/4.11-4 i
C.
Fire Detectors...
3.11/4.11-6 D.
Deleted 3.11/4.11-7 i
E.
Fire Protection Systems Inspection.
3.11/4.11-8 F.
Deleted 3.11/4.11-8 G.
Air Masks and Cylinders............
3.11/4.11-8 H.
Continuous Fire Watch.............
3.11/4.11-9 I.
Open Flames, Welding and Burning in the Cable Spreading Room.
3.11/4.11-9 5.0 Maj or Design Features...............
5.0-1 5.1 Site Features........
5.0-1 5.2 Reactor..........
5.0-1 5.3 Reactor Vessel................
5.0-1 5.4 Containment.
5.0-1 5.5 Fuel Storage.................
5.0-1 5.6 Seismic Design................
5.0-2 i
i J
l it l
i BFN Amendment No.134 1
i Unit 2
{
l I
J
-o.
ADMINISTRATIVE CONTROLS PAGE
.SECTION 6.1 RESPONSIBILITY..........................................
6.0-1 6.2 ORGANIZATION...............................
6.0-1 6.2.1 Corporate...............................................
6.0-1 6.2.2
' Plant Staff.....'........................................
6.0-1 6:3 PLANT STAFF QUALIFI_ CATIONS..............................
6.0-4 6.4 T RA I N I N G................................................ 6. 0 - 4 6.5 PLANT REVIEW AND AUDIT..................................
6.0-4 6.5.1 Plant Operation Review Committee (P0RC).................
6.0-4 6.5.2 Nuclear Safety Review Board (NSRB)......................
6.0-10 6.5.3 Technica) Review and Approval of Procedures.............
6.0-16 6.6 REPORTABLE EVENT ACTIONS........m......
6.0-17 6.7 SAFETY LIMIT VIOLATION..................................
6.0-18 6.8 PROCEDURES / INSTRUCTION AND PR0 GRAMS.....................
6.0-19 6.8.1 Pr o c e d u r e s.............................................. 6. 0 -19 6.8.2 Dri11s..................................................
6.0-20 6.8.3 Radiation Control Procedures............................
6.0-20 6.8.4 Quality Assurance Procedures - Effluent and..............
l Environmental Monitoring..............................
6,0-22 6.9 REPORTING REQUIREMENTS..................................
6.0-23 6.9.1 Routine Reports.........................................
6.0-23 Startup Reports.........................................
6.0-23 Annual Operating Report.................................
6.0-24 Monthly Operating Report................................
6.0-25 Reportable Events.......................................
6.0-25 Radioactive Effluent Rolease Report.....................
6.0-25 Source Tests............................................
6.0-25 6.9.2 Special Reports.........................................
6.0-26 6.10 STATION OPERATING RECORDS AND RETENTION.................
6.0-28 6.11 PROCESS CONTROL PROGRAM.................................
6.0-31 6.12 0FFSITE DOSE CALCULATION MANUAL.........................
- 6. 0-31 6.13 RADIOLOGICAL EFFLUENT MANUAL............................
6.0-32 l
V Anend9ent No.134
'BFN Unit 2
LIST OF TABLES Table-Title Page No.
'1.1
~ Surveillance Frequency Notation.....
1.0-11 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 Circuits.................
3.1/4.1-8
'4.1. B -
Reactor Protection' System (SCRAM) Instrumentation Calibration' Minimum Calibration Frequencies for Reactor Prote; tion Instrument Channels....
3.1/4.1-11 3.2.A Primary Containment and Reactor Building Isolation Instrumentation..
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.....
3.2/4.2-25 3.2.D Radioactits Liquid Effluent Monitoring.
'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 3.2.0 Control Room Isolation Instrumentation......
3.2/4.2-34 3.2.H Flood Protection Instrumentation..........
'3.2/4.2-35 3.2.I Meteorological Monitoring Instrumentation.....
3.2/4.2-36 3.2.J Seismic Monitoring Instrumentation.
3.2/4.2-37 3.2.K Radioactive Gaseous Effluent Monitoring Instrumentation.................
3.2/4.2-38 4.2.A
' Surveillance Requirements for Primary Containment and Reactor Building Isolation Instrumentation..
3.2/4.2-40 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 i
- 4. 2.D -
Radioactive Liquid Effluent Monitoring I
Instrumentation Surveillance Requirements....
3.2/4.2-51 vi 8[$t2 Amendment No. 128 iL
e LIST OF TABLES (Cont'd)
Table Title Page No.
4.2.E Minimum Test and Calibration Frequency for Drywell Leak Detectica Instrumentation..........
3.2/4.2-53 4.2,F Minimum Test and Calibration Freguancy for Surveillance Instrumentation...
3.2/4.2-54 4.2.G.
Surveillance Requirements for Control Room Isolation Instrumentation............
3.2/4.2-56 4.2.H Minimum Test and Calibration Frequency for Flood Protection Instrumentation........
3.2/4.2-57 4.2.J Seismic Monitoring Instrument Surveillance Requirements..................
3.2/4.2-58 4.2 K Radioactive Gaseous Effluent Instrumentation Surveillance.
3.2/4.2-62 3.5-1 Minimum RHRSW and EECW Pump Assignment.
3.5/4.5-11
- 3. 5. I-MAPLHGR Versus Aserage Planar Exposure.
3.5/4.5-21 3.7.A Primary Containment Isolation Valves.......
3.7/4.7-25 3.7.B Testable Penetrations with Double 0-Ring Seals..
3.7/4.7-32 3.7.C Testable Penetrations with Testable Bellows.
3.7/4.7-33 3.7.D Air Tested Isolation Valves..
3.7/4.7-34 3.7.E Primary Containment Isolation Valves which Terminate below the Suppression Pool Water Level......................
3.7/4.7-37 3.7.F Primary Containment Isolation Vales Located in Water Sealed Seismic Class 1 Lines.......
3.7/4.7-38 3.7.H Testable Electrical Penetrations.........
3.7/4.7-39 4.9.A.4.C Voltage Relay Setpoints/ Diesel Generator Start.
3.9/4.9-16 3.11.A Fire Protection System Hydraulic Requirements.
3.11/4.11-10 6.2.A Minimum Shift Crew Requirements..........
6.0-3 l
l vii Amendment No. 128, 134 BFN Unit 2
i LIST OF ILLUSTRATIONS figure
. Title Page No.
.2.1.1 APRM Flow Reference Scram and APRM' Rod Block Settings.....................
1.1/2.1-6
'2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow....
1.1/2.1-7 l
i 4.1-1
' Graphic Aid in the. Selection of an Adequate
]
Interval Between Tests.......'.......
3.1/4.1-13 3.2/4.2-64 4.2-1 System Unavailability.
-3.4-1 Sodium Pentaborate Solution Volume Concentrated Requirements..................
3.4/4.4-4 3.4-2 Sodium Pentaborate Solution Temperature Requirements...
3.4/4.4-5 3.5.K-1 MCPR Limits...
3.5/4.5-22 l
3.5.2 Kr Factor....................
3.5/4.5-23.
)
J 3.6-1 Minimum Temperature 'F Above Change in l
Transient Temperature.
3.6/4.6-24
]
-)
3.6-2 Change in Charpy V Transition Temperature Vs.
)
Woutron Exposure.
3.6/4.6-25 l
1 4.8.1.a Gaseous Release Points and Elevations 3.8/4.8......
4.8.1.b Land Site Boundary.
3.8/4.8-11
)
6.2-1 Offsite Organization for Facility Management and-Technical Support 6.0-33 6.2-2 Facility Organization 6.0-34 I
I
)
'III Amendment No. 128.134 BFN Unit 2
3.2/4.2 PROTECTIVE INSTRUMENTATION LfMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.D Radioactive Liauld Effluent 4.2.D Radioactive Liquid Effluent Monitoring Instrumentation Monitoring Instrumentation
- 1. The radioactive liquid
- 1. Each of the radioactive effluent monitoring liquid effluent monitoring instrumentation listed in instruments shall be Table 3.2.D shall be OPERABLE demonstrated OPERABLE by with the applicability as performance of test in shown in Tables 3.2.D/4.2.D.
accordance with Table 4.2.D.
1 Alarm / trip setpoints will be set in accordance with guidance given in the ODCM to ensure that the limits of specification 3.8.A.1 are not exceeded.
- 2. The action required when the number of operable channels is less than the minimum channels operable requirement is specified in the notes for Table 3.2.D.
Exert best efforts to return the instrument (s) to OPERABLE status within 30 days and if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.
- 3. With a radioactive liquid effluent monitoring channel alarm / trip setpoint less conservative than required by these specifications, suspend the release without delay, declare the channel inoperable, or adjust the alarm / trip setpoint to establish the conservatism required by these j
specifications.
- 4. The provisions of specification 1.0.C and 6.9.1.4 are not applicable.
3.2/4.2-3 Amendment No. 228,134 BFN Unit 2
l l
3.2/4.2 PROTECTIVE INSTRUMENTATION 1
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.E.
Drrwell. Leak Detection 4.2.E. Drywell Leak Detection The limiting conditions of Instrumentation shall be operation for the calibrated and checked as instrumentation that monitors indicated in Table 4.2.E.
drywell leak detection are given in Table 3.2.E.
F.
Surveillance Instrumentation F. Surveillance Instrumentation The limiting conditions for the Instrumentation shall be instrumentation that provides calibrated and checked as surveillance information indicated in Table 4.2.F.
readouts are given in Table 3.2.F.
G.
Control Room Isolation G. Control Room Isolation The limiting conditions for Instrumentation shall be instrumentation that isolates calibrated and checked as the control room and initiates indicated in Table 4.2.G.
the control room emergency pressurization systems are given in Table 3.2.G.
H.
Flood Protection H. Flood Protection The unit shall be shutdown and Surveillance shall be placed in the cold condition performed on the when Wheeler Reservoir lake instrumentation that monitors stage rises to a level such the reservoir level as that water from the reservoir indicated in Table 4.2.H.
begins to run across the pumping station deck at elevation 565.
Requirements for instrumentation that monitors the reservoir level are given in Table 3.2.H.
I.
Meteorological Monitorinr.
I.
Meteorological Monitorinr.
Instrumentation Instrumentation The meteorological monitoring Each meteorological monitoring instrumentation listed in Table instrument channel shall be 3.2 I shall be OPERABLE demonstrated OPERABLE by the at all times, performance of the CHANNEL CHECK at least once per BFN 3.2/4.2-4 Unit 2
.o.2/4.2 PROTECTIVE INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.I.
Meteorological Monitoring 4.2.I Meteorological Monitoring Instrumentation (Cont'd)
Instrumentation (Cont'd)
- 1. With the number of operable 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the CHANNEL meteorological monitoring CALIBRATION at least once channels less than required each 6 months, by Table 3.2.I. restore the inoperable channel (s) to operable status within 7 days.
2.
With one or more of the meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the system to operable status.
J.
Seismic Monitoring 4.2.J.
Seismic Monitoring Instrumentation Instrumentation 1.
The seismic monitoring 1.
Each of the seismic instruments listed in monitoring instruments Table 3.2.J shall be shall be demonstrated operable at all times.
operable by performance of tests at the frequencies 2.
With the number of listed in Table 4.2.J.
seismic monitoring instruments less than the 2.
Data shall be retrieved number listed in Table from all seismic instruments 3.2.J. restore the actuated during a seismic inoperable instrument (s) event and analyzed to to operable status within determine the magnitude 30 days, of the vibratory ground motion. A Special Report 3.
With one or more of the shall be submitted to the instruments listed in Table Commission pursuant to 3.2.J inoperable for more Specification 6.9.2 than 30 days, submit a within 10 days describing Special Report to the the magnitude, frequency Commission pursuant to spectrum, and resultant Specification 6.9.2 within effect upon plant features the next 10 days describing important to safety.
the cause of the malfunction and the plans for. restoring the instruments to operable status.
3.2/4.2-5 BFN Amendment No. 128, 134 Unit 2
l I
l L
L l
l 3.2/4.2 PROTECTIVE INSTRUMENTATION j
i LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.K Radioactive Gaseous Effluent 4.2.K Radioactive Gaseous Effluent j
Monitoring Instrumentation Monitoring Instrumentation l
1.
The radioactive gaseous 1.
Each of the radioactive i
effluent monitoring gaseous effluent monitoring instruments listed in instruments shall be Table 3.2.K shall be OPERABLE demonstrated OPERABLE by I
with the applicability as performance of tests in shown in Tables 3.2.K/4.2.K accordance with Table 4.2.K.
1 Alarm / trip setpoints will be j
I set in accordance with guidance given in the ODCM to ensure that the limits of Specification 3.8.B.1 are not exceeded, l
2.
The action required when the number of operable channels i
is less than the Minimum l
Channels Operable requirement l
is specified in the notes for Table 3.2.K.
Exert best efforte to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Release Report why the inoperability was not i
corrected in a timely manner.
l 3.
With a radioactive gaseous effluent monitoring channel alarm / trip setpoint less conservative than required by these specifications, suspend the release without delay, declare the channel inoperable or adjust the alarm / trip setpoint to establish the conservatism l
required by these specifications.
I' 4.
Both off-gas treatment monitors may be taken out of service for less than one hour for purging of monitors during SI performance.
5.
The provisions of Specifications 1.0.C and 6.9.1.4 are not applicable.
3.2/4.2-6 t
BFN Amendment No. J28,134 i
Unit 2
a 3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8 RADIOACTIVE MATERIALS 4.8 RADIOACTIVE MATERIALS j
I i
Applicability Applicability l
Applies to the release of Applies to the periodic testing and radioactive liquids and gases record requirements and sampling from the facility, and monitoring methods used for facility effluents.
Objective objective To define the limits and conditions To ensure that radioactive for the release of radioactive liquid and gaseous releases from effluents to the environs to assure the facility are maintained that any radioactive releases are as within the limits specified by low as reasonably achievable Specifications 3.8.A and 3.8.B.
and within the limits of 10 CFR Part 20.
The specifications Specification except for 3.8.A.1 and 3.8.B.1 are exempt from the requirements of A. Liould Effluents definition 1.0.C (Limiting condition for Operation).
- 1. Facility records shell be maintained of Specification radioactive concentrations and A. Liould Effluents volume before dilution of each batch of liquid
- 1. The concentration of effluent released, and radioactive material of the average dilution released at any time from flow and length of time the site to unrestricted over which each areas (see Figure 4.8-lb) discharge occurred.
shall be limited to the concentrations specified
- 2. Radioactive liquid waste in 10 CFR Part 20, sampling and activity Appendir B. Table II, analysis of each liquid Column 2 for radionuclides waste batch to be other than dissolved or discharged shc11 be entrained noble gases.
performed prior to For dissolved or entrained release in accordance noble gases, the with the sampling and concentration shall be analysis program limited to 2E-4 mci /ml specified in the REM.
total activity.
- 3. The operation of the
- 2. If the limits of 3.8.A.1 automatic isolation are exceeded, appropriate valves and discharge action shall be initiated tant selection valves without delay to bring shall be checked the release within annually.
BFN 3.8/4.8-1 Amendment No. 128 Unit 2
1 3.8/4.8 RADI0 ACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.A.
Liould Effluents 4.8.A.
Liould Effluents limits. Provide prompt
- 4. The results of the notification to the NRC analysis of samples pursuant to section 6.9.1.4.
collected from release points shall be used
- 3. The doses or dose with the calculational commitment to a member of methodology in the ODCM the public from radioactive to assure that the materials in liquid concentrations at the effluents released from point of release aro each unit to unrestricted maintained within the areas (See Figure 4.P-lb) limits of specification shall be limited:
3.8.A.I.
a.
During any calendar
- 5. Cummulative guarterly guarter to <1.5 mrem to and yearly dose the total body and <5 contributions from mrem to any organ and, liquid effluents shall l
be determined as
{
b.
During any calendar specified in the ODCM at yeer to <3 mrem to the least once every 31 days.
total body and <10 prem to any organ
- 6. The quantity of j
radioactive material
- 4. If the limits specified in contI!ned in any 3.8.A.3 a & b above are outside liquid radwaste j
exceeded, prepare and storage tanks shall be submit Special Report determined to be within the pursuant to Section 6.9.1.4.
above limit by analyzing a representative sample of
- 5. The maximum activity to be the tank's contents at contained in one liquid least once per 7 days when j
radwaste tank or temporary radioactive materials are j
storage tank that can be being added to the tank.
I discharged directly to the environs shall not exceed 10 curies excluding tritium and dissolved / entrained I
noble gas.
- 6. With radioactive liquid waste exceeding 3.8.A.5 i
limits, without delay suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, reduce the tank contents to within the limit.
Events leading to this condition must be reported in the next Semiannual Radioactive Effluent Release Report (section F.2 of the REM) 3.6/4.8-2 Amendment No. I28, 134 BFN 0)ait 2
3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.B.
Airborne Effluents 4.8.B.
Airborne Effluents
- 1. The dose rate at any time
- 1. The gross S/y and to areas at and beyond the particulate activity of site boundary (see Figure gaseous wastes released 4.8-Ib) due to to the environment shall radioactivity released in be monitored and gaseous effluents from the recorded.
site shall be limited to the following values:
- a. For effluent streams having continuous a.
The dose rate limit monitoring for noble gases shall capability, the be <500 mrem /yr to the activity shall be total body and <3000 monitored and flow mrem /yr to the skin, rate evaluated and and recorded to enable release ratas of b.
The dose rate limit gross radioactivity for I-131 I-133, H-3, to be determined at and particulate with least once per shift greater than eight day using instruments half-lives shall be specified in Table
<1500 mrem /yr to any 3.2.K.
organ.
- b. For effluent streams
- 2. If the limits of 3.8.B.1 without continuous are exceeded, appropriate monitoring corrective action shall be capability, the immediately initiated to activity shall be bring the release within monitored and limits.
Provide prompt recorded and the notification to the NRC release through pursuant to Section these streams 6.9.1.4.
controlled to within the limits specified
{
I in 3.8.B.
- 2. Radioactive gaseous waste sampling and activity analysis shall be performed in accordance with the sampling and analysis program specified in the REM. Dose rates shall be determined to be within limits of 3.8.B using methods contained in the ODCM.
3.8/4.8-3 BFN Amendment No. I28,134 Unit 2
l 1
3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.B Airborne Effluents 4.8.B Airborne Effluents
,j (Cont'd)
(Cont'd)
- 3. The air dose to areas at 3.
Cumulative quarterly and i
and beyond the site yearly dose boundary (see Figure contributions from
{
4.8-1b) due to noble gases gaseous releases shall i
released in gaseous be determined using f
effluents per unit shall methods contained in the be limited to the ODCM at least once every following:
31 days.
I 1
l a.
During any calendar j
quarter, to $5 mrad j
for gamma radiation
)
and $10 mrad for beta l
radiation; b.
During any calendar year, to $10 mrad for gamma radiation and
$20 mrad for beta radiation.
- 4. If the calculated air dose I
exceeds the limits specified in 3.8.B.3 above, prepare and submit a special report pursuant to section 6.9.1.4.
- 5. The dose to a member of the public from radiciodines, radioactive materials in particulate form, and radionuclides
'other than noble gases with half lives greater than 8 days in gaseous effluent released per unit to areas at and beyond the site boundary (see Figure 4.8-1b) shall be limited 1
to the following:
i a.
To any organ during any calendar quarter to $7.5 arem; b.
To any ortan during l
any calendar year to
$15 mrem; i
3.8/4.8-4 BFN Amendment No. 128, 134 Unit 2
3.8/4.8 RADIOACTIVE KATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.B Airborne Effluents 4.8.B Airborne Effluents (Cont'd)
(Cont'd)
- 6. If the calculated doses 4.
During operation above exceed the limits of 25% power, the position 3.8.B.5 above, prepare of the charcoal bed and submit a special bypass valve will be report pursuant to verified daily.
Section 6.9.1.4
- 7. During operation above 5.
The concentration of 25% power the discharge hydrogen downstream of of the SJAE must be the recombiners shall be routed through the determined to be within charcoal adsorbers.
the limits of 3.8.B.9 by continuously monitoring
- 8. With gaseous waste being the off-gas whenever discharged for more than the SJAE is in service 7 days without treatment using instruments described in through the charcoal Table 3.2.K.
Instrument adsorbers, prepare and surveillance requirements are submit a special report specified in Table 4.2.K.
pursuant to Section 6.9.1.4.
- 9. Whenever the SJAE is in service, the concentration of hydrogen in the offgas downstream of the recombiners shall be limited to $4% by volume.
- 10. With the concentration of hydrogen exceeding the limit of 3.8.B.9 above, restore the concentration to within the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Amendment No. 228,134 BFN Unit 2
4 3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.C Radioactive Effluents - Dose 4.8.C Radioactive Effluents - Dose
- 1. The dose or dose 1.
Cumulative dose commitment to a real contributions'from individual from all liquid and gaseous uranium fuel cycle sources effluents shall be is limited to $25 mrem to determined in the total body or any accordance with organ (except the thyroid, specifications 3.8.A.3, which is limited to 175 3.8.B.3. and 3.8.B.5 arem) over a period of one and the sethods in the calendar year.
ODCM.
- 2. With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 3.8.A.3, 3.8.B.3, or 3.8.B.5, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.1.4 and limit the subsequent releases such that the limits of 3.8.C.1 are not exceeded.
3.8.D Mechanical Vacuum Pump 4.8.D Mechanical Vacuum Pump
- 1. Each mechanical vacuum pump At least once during each shall be capable of being operating cycle verify automatic automatically isolated and securing and isolation of the secured on a signal or mechanical vacuum pump.
high radioactivity in the steam lines whenever the main steam isolation valves are open.
- 2. If the limits of 3.8.D.1 are not met, the vacuum pump shall be isolated.
3.8/4.8-6 Amendment No. J78, 134 BFN Unit 2 1
l i
i l
l
\\
3.11/4.11 - FIRE PROTECTION SYSTEMS l
I LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
]
3.11.A.
High Pressure Fire Protection 4.11.A.
High pressure Fire System Protection System area where protection is lost is checked hourly.
3.
If only one high
- 3. Raw Service Water pressure fire pump is System Testing OPERABLE, the reactors may remain Item Freauency in operation for a period not to exceed Simulated Once/ year 7 days, provided the automatic requirements of and manual Specification actuation 3.11.A.I.b above of raw service are met, water pumps and operation of tank level switches.
4.
If Specification
- 4. The high pressure 3.11.A.3 cannot be fire protection set, the reactors system pressure shall be placed in shall be logged the Cold Shutdown
- daily, condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
5.
Removal of any
- 5. Principal header component in the High and component Pressure Fire System isolation valves from service for any shall be checked reason other than open at intervals testing or emergency no greater than operations shall 3 months.
require Plant Manager approval.
6.
The Raw Service Water storage tank level shall be maintained above level 723'7" by the raw service water pumpe.
l l
3.11/4.11-3 Amendment No. 134 BFN Unit 2 l
L_____._____.__
i i
3.11/4.11 FIRE PROTECTION SYSTEMS l
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.A.
High Pressure Fire Protection 4.11.A.
High Pressure Fire System Protection System 7.
If Specification 3.11.A.6 cannot be met a fire pump shall be started and run continuously until the raw service water pumps can maintain a raw service water storage tank level above 723'7".
q l
8.
The fire protection water distribution system shall have a minimum capacity of 2664 spm at 250' head.
9.
The fire protection system shall be capable of supplying the individual loads listed in Table 3.11.A.
B.
C0p Fire Protection System B.
CO2 Fire Protection System 1.
The CO2 Fire 1.
CO2 Fire Protection Protection System Testing:
shall be OPERABLE:
Item Frecuency a.
With a minimum a.
Simulated once/ year of 8-1/2 tons automatic and manual (0.5 Tank) CO2 in storage units actuation 1 and 2.
b.
With a minimum b.
Storage Checked of 3 tons (0.5 tant daily Tant) CO2 pressure storage unit 3.
and level BFN 3.11/4.11-4 Unit 2
1 3.11/4.11 FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.B.
CO2 Fire Protection 4.11.B.
CO2 Fire Protection i
System
. System l
t 3.11.B.1 (Cont'd) 4.11.B.1 (Cont'd) c.
Automatic c.
CO2 Spray Once/3
)
l initiation logic header and years OPERABLE.
nozzle inspection for blockage 2.
If Specifications 2.
When the cable l
3.11.B.I.a or spreading room 3.11.B.1.b or CO2 Fire 3.11.B.I.c cannot be Protection met, a patrolling is INOPERABLE, fire watch with one 125-pound portable fire (or larger) equipment shall be portable fire established to ensure extinguisher that each area where shall be placed protection is lost at each entrance.
is checked hourly.
3.
If Specifications 3.11.B.1.a.
3.11.B.1.b. or 3.11.B.1.c are not met within 7 days, the affected unit (s) shall be in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.
If CO2 fire protection is lost to a cable spreading room or to any diesel generator buildirg area a continuous fire watch shall be established immediately and shall be continued until CO2 fire protection is restored.
l BFN 3.11/4.11-5 Unit 2
l L
3.11/4.11 FIRE PROTECTION SYSTEMS I
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l
3.11.B.
C.02 fire Protection l-J System 5.
Removal of any component in the CO2 l
Fire Protection System l
from service for any l
reason other than i
tetting or emergency operations shall require Plant Manager
)
- approval, l
C.
Fire Detectors 4.11.C.
Fire Detectors
)
2.
The fire detection 1.
All heat and smoke system'c heat and detectors shall be smoke detectors for tested in accordance all protected zones with indastrial shall be OPERABLE standards or other except that one approved methods detector for a given semiannually.
protected zone may be INOPERABLE for a period no greater than 30 days.
2.
If Specification 2.
The non-Class A 3.11.C.1 cannot be supervised detector met, a patrolling circuitry for those fire watch will be detectors which established to ensure provide alarm only that each protected will be tested once zone or area with each month by INOPERABLE detectors actuating the is checked at detector at the end intervals no greater of the line or end than one each hour.
of the branch such that the largest number of circuit conductors will be checked.
3.11/4.11-6 BFN Amendment No.134 Unit 2
i 3 11/4.11 FIRE PROTECTION SYSTEMS i
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.11.C.- Fire Detectors
]
3.
The class A supervised detector alarm circuits will l
be tested once each two months at the local panels.
4.
The circuits between the local panels'in 4.11.C.3 and the main control room will be tested monthly.
5.
Smoke detector sensitivity will be checked in accordance with manufacturer's instruction annually.
3.11.D.
ROVING FIRE WATCH D.
ROVING FIRE WATCH Deleted Deleted 3.11/4.11-7 BFN Amendment No.134 Unit 2
l l
1 3.11/4.11 FIRE PROTECTION SYSTEMS
. LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.E. Fire Protection Systems 4.11.E.
Fire Protection Systems Inspection Inspections All fire barrier Each required fire penetrations, including barrier penetration shall cable penetration barriers, be verified to be l
fire doors and fire functional at least once dampers, in fire zone per 18 months by a visual boundaries protecting inspection, and prior to i
l safety-related areas shall restoring a fire barrier l
be functional at all times, to functional status l
With one or more of the following repairs or j
required fire barrier maintenance by penetrations nonfunctional performance of a visual within one hour establish inspection of the a continuous fire watch affected fire barrier on at least one side of penetration.
the affected penetration or verify the OPERABILITY of fire detectors on at least one side of the nonfunctional fire barrier and establish an hourly fire watch patrol until the work is completed and the barrier is restored to functional status.
F. Fire Protection Organization F.
Fire Protection Orr.anization Deleted Deleted G. Air Masks and Cylinders G.
Air Masks and Cylinders A minimum of 15 air No additional masks and 30 500-cubic-inch surveillance required.
air cylinders shall be available at all times except that a time period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following emergency use is allowed to permit recharging'or replacit.g.
1 l
3.11/4.11-8 BFN Amendment No. 134 i
Unit 2
}
1
3.11, BASES The High Pressure Fire and'CO2 Fire Protection specifications are provided.
'in order to meet the preestablished levels of operability during a fire in either or all of the three units.
Requiring a patrolling fire watch with portable fire equipment if the automatic initiation is lost will provide (as
'does the automatic system) for early reporting and isunediate fire fighting capability in the event of a fire occurrence.
The High Pressure Fire Protection System is supplied by four pumps (three electric driven and one diesel driven) aligned to the high pressure. fire header. The reactors may remain in operation for a period not to exceed seven days if three pumps are out-of-service.
If at least two pumps are not made operable in seven days or if all pumps are lost durint, this seven-day period, the reactors will be placed in the cold shutdown condition within I
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
For the areas of applicability, the fire protection water distribution system minimum capacity of 2,664 spm at 250' head at the fire pump discharge consists of the following design loads:
2 2 area) 1332 gpm 1.
Sprinkler System (0.30 gpm/f t /4440 f t
'2.
I 1/2" Hand Hose Lines 200 gpm 3.
Raw Service Water Load 1132 apm 1
TOTAL 2664 spm The CO2 Fire Protection System is considered operable with a minimum of 8 1/2 tons (0.5 tank) CO2 in storage for units 1 and 2; and a minimum of 3 tons (0.5 tank) CO2 in storage for unit 3.
An immediate and continuous fire watch in the cable spreading room or any diesel generator building area will be established if CO2 fire Protection is lost in this room and will continue until CO2 fire protection is restored.
To assure close. supervision of fire protection system activities, the removal from service of any component in either the High Pressure Fire System or the CO2 Fire Protection System for any reason other than testing or emergency operations will require Plant Manager approval.
Early reporting and immediate fire fighting capability in the event of a fire occurrence will be provided (as with automatic system) by requiring a patrolling fire watch if more than one detector for a given protected zone is inoperable.
1 3.11/4.11-12 1
l BFN Amendnent No.134 l
I Unit 2 L_-_______-_-___-______-_____..
n 3.11 BASES (Cent'd)
The fire protection system is designed to supply the required flow and pressure to an individual load listed on Table 3.11.A while maintaining a design raw service water load of 1132 gpm.
4.11 BASES Periodic testing of both the High Pressure Fire System and the CO2 Fire Protection System will provide positive indication of their operability.
If only one of the pumps supplying the High Pressure Fire System is operable, the pump that is operable will be checked immediately and daily thereafter to demonstrate operability. If the CO2 Fire Protection 4
System becomes inoperable in the cable spreading room, one 125-pound (or larger) fire extinguishers will be placed at each entrance to the cable spreading room.
Annual testing of automatic valves and control devices is in accordance with NFPA Code Vol. II, 1975, section 15, paragraph 6015. More frequent testing would require excessive automatic system inoperability, since there are a large number of automatic valves installed and various portions of the system must be isolated during an extended period of time during this test.
1 l
Wet fire header flushing, spray header inspection for blockage, and nozzle inspection for blockage will prevent, detect, and remove buildup of sludge or other material to ensure continued operability. System flushes in conjunction with the semiannual addition of biocide to the Raw
{
Cooling Water System will help prevent the growth of crustaceans which i
could rede e nozzle discharge.
l l
Semiannual :.ests of heat and smoke detectors are in accordance with the NFPA Code.
With the exception of continuous strip heat detectors panels, all non-class A supervised detector circuits which provide alarm only are hardwired through conduits and/or cable trays from the detector to the main control room alarm panels with no active components between.
Nonclass A circuits also actuate the HPCI water-fog system, the CO2 system in the diesel generator buildings, and isolate ventilation in shutdown board rooms. The test frequency and methods specif.ted are justified for the following reasons:
1.
An analysis was made of worst-case fire detection circuits at Browns Ferry to determine the probability of no undetected failure of the circuits occurring between system test times as specified in the surveillance requirements. A circuit is defined as the wire connections and components that affect transmission of an alarm signal between the fire detectors and the control room annunciator. Three circuits were analyzed which were representative of an alarm-only circuit, a water-fog circuit, and a CO2 circuit. The spreading room B smoke detector was selected as the worst-case alarm-only circuit because it had the largest number l
l of wires and connections in a single circuit. The HPCI water-fog l
circuit was selected for analysis because it is the only water-fog circuit in the area of applicability for Technical Specifications.
BFN 3.11/4.11-13 Unit 2
i 4.11 EASES (Cont'd)
The Standby Diesel Generator Room A CO2 circuit was selected because it contained 2 out of 3 detector logic, the most complicated CO2 circuit logic. Calculations were based on
-]
failure rates for wires, connections, and circuit components as j
shown in Appendix III of WASH-1400. Failure rates were considered 1
for the following circuit components:
J 1.
Open circuit 2.
Short to ground
)
3.
Short to power l
4.
Timing motor failure to start
)
5.
Relay failure to energize
)
6.
Normally open contact failure to close 1
7.
Normally open or normally closed contact short 8.
Normally closed contact opening 9.
Timing switch failure to transfer The calculated probabilities (Pf) for no undetected failure of the I
circuits occurring were as follows, based on the specified test f
frequency.
)
AREA TEST FREQUENCY Pf Spreading Room B One Month 0.975287 HPCI Water Fog Six Months 0.977175 Standby Diesel Gen Room A CO2 Sir Months 0.957595 The worst case of the three areas considered is Spreading Room B.
The probability of undetected failure is approximately 1/40, which means that one undetected failure will occur on the average every 40 months over an extended period of time end that the failure could exist up to one month. The frequency of testing is thus much greater than the frequency of failure and produces circuits with adequate reliability.
i 2.
Circuits checks by initiation of end of the line or end of the i
branch detectors will more thoroughly test the parallel circuits j
than testing on a rotating detector basis. This test is not a detector test, but is a test to simulate the effect of electrical supervision as defined in the NFPA Code.*
3.
Testing of circuits which actuate My, water, or ventilation systems requires disabling the autyutic feature of the fire protection system for the area. A surveillance program which disabled these circuits monthly would significantly reduce the ability of these circuits to provide fire suppression.
l
- Ref: NFPA Code 72D-9, paragraph Illi, Code 72D-15, paragraph 1312 for definition of Class A systems, and Code 72A-18, Article 240.
l 1
BFN 3.11/4.11-14 Unit 2
\\
a I
4.13 BASES' (Cont'd) 4.
Daily tests of' annunciation lights and audible devices are performed as a routine operation function.
5.
The CO2 system manufacturer recommends semiannual testing of CO2 system fire detection circuits.
In addition, operating personnel periodically inspect the plant during their normal operating activities for fire hazards and other abnormal conditions.
Smoke dotectors will be tested "in-place" using inert freon gas applied by a pyrotronics type applicator which is accepted throughout the industrial fire protection industry for testing products of combustion detectors or by use of' the NSA chemical smoke generators. At the present time, the manufacturers i
.have only approved the use of " punk" for creating smoke. TVA will not use
" punk" for testing smoke detectors.
]
3
.l 1
i 1
l 1
3.11/4.11-15 Amendment No. 134 BFN Unit 2
j 1
1 1
BFN TECHNICAL SPECIFICATIONS 6.0 ADMINISTRATIVE CONTROLS 1
i 6.0 ADMINISTRATIVE CONTROLS j
6.1 RESPONSIBILITY The Plant Manager has onsite responsibilities for the safe operation of the facility and Zhall report to the Browns Ferry Site Director.
In the j
absence of the Plant Manager, a Plant Superintendent will assume his j
responsibilities.
l 6.2 ORGANIZATION CORPORATE l
6.2.1 The portion of TVA management which relates to the operation of the plant is shown in Figure 6.2-1.
]
t PLANT STAFF 6.2.2 The funct.$onal organization for the operation of the plant shall i
be shown in Figure 6.2-2.
- a. Shift manning requirements, shall ss a minimum, be as described in Table 6.2.A and below, i
1 b.
A licensed senior reactor operato: shall be present at the site at all times when there is fue) in t:se reactor.
6.0-1 Amendment No. 134 arN Unit 2
i 6.2.2 (Cont.)
l k
c.
A licensed reactor operator shall be'in the control room whenever.
there is fuel in the reactor.
d.
Two licensed reactor operators shall be in the control room during any cold startups, while shutting down the reactor, and during j
recovery from unit trip.
In addition, a person holding a senior operator license shall be in the control room for that unit whenever it is in an operational mode other than cold shutdown or refueling.
j i
e.
A Health Physics Technician
- shall be present at the facility at all times when there is fuel in the reactor.
f.
A person holding a senior operator license or a senior operator license limited to fuel handling, shall be present during l
alteration of the core to directly supervise the activity and during this time shall not be assigned other duties.
g.
A site fire brigade of at least five members shall be maintained onsite at all times.* The fire brigade shall not include the Shift Engineer and the other members of the minimum shift crew necessary for safe shutdown of the unit, nor any personnel required for other essential functions during a fire emergency.
l
- The Health Physics Technician and fire brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to acconrnodate unexpected absence, provided inunediate action is taken to fill the required positions.
l 1
6.0-2 BFN Amendment No. 134 Unit 2
Table 6.2.A Minimum Shift Crew Requirements position
-Units in Operation Type of License E
1
.gd g
a 1
1 1
1 SRO Senior Operator Senior Operator.
0 1
2 2
SRO Licensed Operators 3
3 3
0 1
2 2
RO or SRO Additional Licensed Operators Assistant Unit Operators (AUO) 4 4
5 5
None Shift Technical Advisor (STA) 0 1
1 1
None Health physics Technician 1
1 1
1 None Note for Table 6.2.A A senior operator will.be assigned responsibility for overall plant a.
operation at all times there is fuel in any unit.
b.
Except for the senior operator' discussed in note "a", the shift crew composition may be one less than the minimum requirements of Table 6.2.A for a period of time not to exceed two hours 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 of Table 6.2.A.
This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.
One of the Additional Licensed Operators must be assigned to each control c.
room with an operating unit.
d.
The number of required licensed personnel, when the operating units are controlled from a common control room, are two senior operators and four operators.
6.0-3 Amendment No. 134 BFN Unit 2
N 1
j 6.3 PLANT STAFF QUALIFICATIONS Qualifications of the Browns Ferry Nuclear Plc.nt management and operating staff shall meet the minimum acceptable levels as described in ANSI -
N18.1, Selection and Training of Nuclear Power Plant Personnel, dated March 8, 1971. The qualifications of the Health Physics Supervisor will meet or exceed the minimum acceptable levels as described in Regulatory Guide 1.8, Revision 1, dated September 1975. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or i
engineering discipline with specific training in plant design and transient and accident response and analysis.
6.4 TRAINING A retraining and replacement training program for station personnel shall be in accordance with ANSI - N18.1, Selection and Training of Nuclear Power Plant Personnel, dated March 8, 1971. The minimum frequency of the retraining program shall be every two years.
6.5 PLANT REVIEW AND AUDIT f
6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC)
FUNCTION The PORC shall function to advise the Plant Manager in all 6.5.1.1 a.
matters related to nuclear safety.
b.
This advisory function shall be performed by the PORC acting in a formal meeting or by members acting individually without a formal meeting.
I 6.0-4 BFN Amendment No. 134 Unit 2
l o
COMPOSITION 6.5.1.2 The PORC shall be composed of the:
i
- a. -Chairman:
Plant Manager 1
Alternate Chairman:
Assistant to Plant Manager 1
Alternate Chairman or Member: Technical Services-Superintendent Member:
Unit Superintendents (3) l Member:
Maintenance Superintendent Member:
Quality Assurance Staff Supervisor Member:
Health Physics Supervisor b.
All alternate chairmen and alternate members shall be appointed in writing by the PORC chairman.
MEETING FREQUENCY 6.5.1.3 The PORC shall convene in a formal meeting at least once a month and as directed by the chairman. Other PORC meetings may be requested by the chairmen or members as required.
l t
6.0-5 Amendment No. 134 BFN
- Unit 2
6.5.1.4 For expedited meetings', when it is not practical to convene as a group, the chairman or alternate chairman may conduct committee business by polling the members individually (by telephone or in person) or via a serialized review.
QUORUM 6.5.1.5 The quorum necessary for the PORC to act in a formal meeting shall consist of the chairman or alternate chairman and at least five saembers or their alternates. Members shall be considered present if they are in telephone communication with the committee.
RESPONSIBILITIES 6.5.1.6.The PORC shall be responsible for the activities listed below.
The PORC may delegate the performance of reviews, but will maintain cognizance over and responsibility for them, e.g.,
subcommittees.
a.
Review of administrative procedures for the control of the technical and cross-disciplinary review of (1) all procedures required by Specification 6.8.1.1, and changes thereto, (2) any other procedures and changes thereto determined by the Plant Manager to affect nuclear safety.
b.
Review of the administrative procedures required by Appendix A of Regulatory Guide 1.33, Revison 2, February 1978 and changes thereto.
c.
Review of emergency operating procedures and changes thereto.
d.
Review implementing procedures of the Radiological Emergency Plan and the Industrial Security Program.
BFN Amendment No. 134 Unit 2 6.0-6
i e.
Review of.all proposed changes to the Technical Specifications.
f.
Review of safety evaluation for proposed tests or experiments to be completed under the provisions of 10 CFR 50.59 g.
Review proposed changes to the Radiological Effluent Manual.
h.
Review adequacy of' the Process Control Program and Offsite Dose Calculation Manual at least once every 24 months,
- i. Review changes to the radwaste treatment systems.
- j. Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendation, and disposition of the corrective action to prevent recurrence to the Director, Nuclear Power and to the Nuclear Safety Review Board, k.
Review of all safety evaluations for modifications to structures, systems or components that affect nuclear safety to verify that such actions did not constitute an unreviewed safety question as defined in 10 CFR 50.59, or requires a change to these Technical Specifications.
l l
BFN Amendment No.134 Unit 2 6.0-7 i
e 1.
Review of reportable events, unusual events, operating anomalies, and cbnormal performance of plant equipment.
Investigate reported or suspected incidents involving safety m.
questions or violations of the Technical Specifications, n.
Review of unit operations to detect potential hazards to nuclear safety. Items that may be included in this review are NRC inspection reports, QA audit NSRB audit results, American Nuclear Insurer (ANI) inspection results, and significant corrective action reports (CARS).
performance of special reviews, investigations, or analysis, o.
and report thereon as requested by the plant Manager or the Nuclear Safety Review Board.
6.0-8 BFN Amendment No. 134 Unit 2
- AlflHORITY f
L 6.5.1.7 The PORC shall:
1.
a.
Recommend to the Plant Manager in writing, approval, or disapproval of items considered under 6.5.1.6.a through l
1 above.
l 1.
The recommendation shall be' based on a majority vote of the PORC at a formal meeting.
l l
2.
The recommendation shall be based on a unanimous vote of I
the PORC when the PORC members are acting individually.
3.
Each member or alternate member shall have one vote.
b.
Furnish for consideration a determination in writing with I
regard to whether or not each item considered under 6.5.1.6.f above constitutes an unreviewed safety question.
l c.
Make recommendations to the Plant Manager in writing concerning whether action reviewed under 6.5.1.6.k above did not constitute an unreviewed safety question, d.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Site Director and the Nuclear Safety Review Board of disagreements between the PORC and the Plant Manager.
However, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.
^**" **"
BFN 6.0-9 Unit 2
i a
l RECORDS 6.5.1.8 The PORC shell maintain written minutes of each PORC nesting including expedited meetings that, as a minimum, document the result of all PORC activities performed under the responsibility and authority provisions of these technical specifications.
Copies shall be provided to the Site Director and the Nuclear Safety Review Board.
6.5.2 NUCLEAR SAFETY REVIEW BOARD FUNCTION 6.5.2.1 The NSRB shall function to provide independent review and audit cognizance of designated' activities in the areas of:
a.
Nuclear power plant operations b.
Nuclear engineering c.
Chemistry and radiochemistry d.
Metallurgy e.
Instrumentation and control f.
Radiological safety g.
Mechanical and electrical engineering, and h,
Quality assurance practices COMPOSITION 6.5.2.2 The NSRB shall be composed of at least five members, including the Chairman. Members of the NSRS may be from the Office of Nuclear Power or other TVA organizations, or external to TVA.
6.0-IC Amendment No.134 BFN Unit 2
QUALIFICATIONS 6.5.2.3 The Chairman, members, alternate members of the NSRB shall be appointed in writing by the Manager of Nuclear Power and shall have an academic degree in engineering or a physical science field, or the equivalent; and in addition, shall have a minimum of 5 years technical experience in one or more areas given in 6.5.2.1.
No more than two alternates shall participate as votint, members in NSRB activities at any one time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized to provide expert advice as determined by the NSRB.
MEETING FREQUENCY 6.5.2.5 The NSRB shall meet at least once per six months.
QUORUM 6.5.2.6 The minimum quorum of the NSRB necessary for the performance of the NSRB review and audit functions of these technical specifications shall consist of more than half of the NSRB membership or at least five members, whichever is greater. The quorum shall include the Chairman or his appointed alternate and the NSRB members includinr, appointed alternate members meeting the requirements of 6.5.2.3.
No more than a minority of the quorum shall have line responsibility for operation of the unit.
/-
6.(d11.'
y Amendment No. 134 BFN j>
, <[
i Unit 2 y
O '
[-
REVIEW 6.5.2.7 The NSRB shall review:
a.
The safety evaluations for:
(1) changes to procedures, equipment or systems, and (2) tests or experiments. completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
b.
Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
c.
Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
i d.
proposed changes to Technical Specifications or this Operating License.
e.
Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
f.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
g.
All Reportable Events h.
All recognized indications an unanticipated deficiency in some aspect of design or opwcation of structures, systems, or components that could affect nuclear safety; and 1.
Reports and meeting minutes of the PORC.
6.0-12 BFN Amendment No. 134 Unit 2
AUDITS
)
6.5.2.8 Audits of unit activities shall be performed under the I
cognizance of the NSRB. These audits shall encompass:
The conformance of plant operation.to provisions contained a.
within the Technical Specifications and applicable license conditions at least once per 12 months.
b.
The performance, training and qualifications of the entire plant staff at least once per 12 months.
c.
The results of actions taken to correct deficiencies occurring in site equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.
d.
The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months.
e.
The Site Radiological Emergency Plan and implementing procedures at least once every 12 months.
f.
The Plant Physical Security Plan and implementing procedures at least once every 12 months, g.
Any other area of site operation considered appropriate by the NSRB or the Manager of Nuclear Peder.
h.
The fire protection programmatic controls including the implementing procedures at least once per 24 months.
6.0-13 Amendment No.134 BFW Unit 2
i
- i. An independent fire protection and loss prevention program inspection and audit shall be performed annually utilizing either qualified offsite license personnel or an outside fire protection firm.
J.
An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervals no greater than 3 years.
k.
The Radiological Environmental Monitoring program and the results thereof at least once per 12 months.
1.
The performance of activities required by the Quality Assurance program to meet the criteria of Regulatory Guide 4.15, December 1977, or Regulatory Guide 1.21, Rev. 1, 1974, and Regulatory Guide 4.1, 1975, at least once every 12 months.
m.
The performance of activities required by the Safeguards contingency Plan to meet the criteria of 10 CFR 73.40(d) at least once every 12 months.
n.
The Offsite Dose Calculation Manual and implementing procedures at least once per 24 months.
o.
The Process Control Program and implementing procedures for solidification of wet radioactive wastes at least once per 24 months.
p.
The Radiological Effluent Manual and implementing procedures at least once per 12 months.
f 6.0-14 BFN Unit 2 Amendment No. 228,134
a AUTHORITY 6.5.2.9 The NSRB shall report to and advise the Manager of Nuclear Power on those areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8.
RECORDS 6.5.2.10 Reports of activities shall be prepared, approved, and distributed as indicated below:
a.
Minutes of each NSRB meeting shall be prepared, approved and forwarded to the Manager of Nuclear Power within 14 days following each meeting.
b.
Reports of reviews encompassed by'Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Manager of Nuc; ear Power within 14 days following completion of the review.
c.
Audit reports encompassed by Specification 6.5.2.8 above, shall be forwarded to the Manager of Nuclear Power and to
.the management positions responsible for the areas audited within 30 days after completion of the audit.
6.0-15 BFN Amendment No.134 Unit 2
i I
6.5.3 TEONICAL REVIEW AND APPROVAL OF PROCEDURES ACTIVITIES 6.5.3.1 Procedures required by Technical Specification 6.8.1.1 and other procedures which affect plant nuclear safety, and changes (other than editorial or typographical changes) thereto, shall be prepared, reviewed and approved. Each procedure or procedure change shall be reviewed by an individual other than the preparer. The reviewer may be from the same organization or from a different organization. Procedures other than Site Director Standard Practices will be approved by the responsible Section Supervisor, or applicable Plant Superintendent.
6.5.3.2 Proposed changes or modifications to plant structures, systems and components that affect nuclear safety shall be reviewed as designated by the Plant Manager. Each such modification shall be reviewed by an individual / group other than the individual / group which designed the modification, but who may be from the same organization as the individual / group which designoi the modification.
Proposed modifications to plant structures, systems and components that affect nuclear safety shall be approved by the Plant Manager, prior to implementation.
l 1
6.5.3.3 Individuals responsible for reviews performed in accordance i
with 6.5.3.1 shall be members of the site supervisory staff previously designated by the Plant Manager. Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary.
If deemed necessary, such review shall be performed by review personnel of the appropriate discipline.
l 1
6.5.3.4 The Plant Manager shall approve all administrative l
procedures requiring PORC review prior to implementation, i
BFN Amendment No. 134 Unit 2 6.0-16
6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
t l
- a. 'The Commission shall be notified and a report submitted
]
pursuant to the.roguirements of Section 50.73 to 10 CFR Part 50, and i
b.
Each REPORTABLE EVENT shall be reviewed by the PORC and the j
i i
results of this review shall be submitted to the NSRB and the Site Director.
i 4
k f
6.0-17 Amendment No. 134 BFN Unit 2
6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
The NRC Operations Center shall be notified by telephone as a.
soon ar possible and in all cases within I hour. The Manager of Nuclear Power and the WSRB shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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 (?) effects of the violation upon facility components, systems, or structuree, and (3) corrective action taken to prevent recurrence.
The Safety Limit Violation Report shall be submitted to the c.
Commission, the NSRB and the Manager of Nuclear Power within i
4 14 days of the violation.
d.
Critical operation of the unit shall not be resumed until authorized by the Commission.
1 l
1 l
6.0-18 Amendment No. 134 BFN Unit 2
.s
- 6. 8 PROCEDURES / INSTRUCTIONS AND-PROGRAWa 6.8.1 PROCEDURES
.6.8.1.1 Written procedures shall be established, implemented and
. maintained covering the activities referenced below:
.j a.
The applicable procedures recommended ~in' Appendix A of-Regulatory Guide 1.33, Revision 2, February 1978.
- b.. Limitations on the amount of overtime worked by
.j individuals performing safety-related functions in accordance with NRC Policy statement on working hours l
1 (Generic Letter No. 82-12).
Surveillance and test activities.of safety-related c.
equipment.
j i
I d.. Security plan implementation.
e.
Emergency plan impicmentation.
f.
Fire Protection Program implementation.
g.
Radiological Effluent Manual implementing procedures.
h.
Process Control Program (PCP).
-j. Administrative procedures which control technical and cross-disciplinary review.
6.8.1.2 Each administrative procedure required by Section 6.8.1.1.a.
shall be reviewed by PORC and all other procedures required by Section 6.8.1.1.a. shall be reviewed in accordance with l
Section 6.5.3.
Amenduent No.134 BFN Unit 2 6.0-19
6.8.1.3 Temporary changes to procedures of Specification 6.8.1.1 may be made provided:
a.
The intent of the original procedure is not altered; b.
The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator License on the unit affected; c.
The change is documented, reviewed by the PORC and approved by the Plant Manager within 14 days of implementation, for changes in administrative procedures requiring PORC review.
d.
The change is documented, reviewed per Specification 6.5.3, and approved by the responsible group section supervisor within 14 days of implementation, for changes to procedures other than administrative procedures.
DRILLS 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 Radiation Control Procedures shall be maintained and made available to all station personnel. These procedures shall show permissible radiation exposure and shall be consistent l
6.0-20 BFN Amendment No. 134 Unit 2
ei t-with the requirements of 10 CFR 20.
This radiation protection program shall be organized to meet the requirements of 10 CFR 20 except in lieu of the " control device" or " alarm signal" required by paragraph 20.203 (c) of 10 CFR 20.
6.8.3.1 Each high radiation area in which the intensity of radiation is greater than 100 mrom/hr but less than 1000 ares /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radfulogical Work Permit.* Ary 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 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 mrom/hr shall be subject to the provisions of (1) above; and, in addition, access to the source and/or area 6.0-21 Amendment No. 134 BFw Unit 2
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.
f
(
i
- 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 l
l duties, provided they comply with approved radiation protection procedures for entry into high radintion areas.
QUALITY ASSURANCE PROCEDURES - EF/LUENT 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.
I 1
1 l
l l
1 6.0-22 l
BFN U"
Amendment No. Z28,134
I j
i
(
)
I 6.9 REPORTING REQUIREMENTS l
I ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements cf Title 10, Code of Federal Regulations, the following identified l
4 reports shall be submitted to the Director of the Regional Office of NRC, unless otherwise noted.
6.9.1.1 STARTUP REPORT r
A summary report of plant startup and power escalation a.
testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a 1
planned increase in power level, (3) installation of fuel i
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 obtained 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 described. Any additional specific details required in license conditions based on other conunitments shall be included in this report.
6.0-23 BFN Amendment No. 134 Unit 2
f o
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 f
all three events have been completed.
1 l
l 6.9.1.2 ANNUAL OPERATING REPORT
- i A tabulation on an annual basis of the number of station, a.
utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rom exposure according
- .o 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 80% of the total whole body dose received from external cources shall be acsigned 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.
- A single submittal may be made for a multiple unit station.
- This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.
6.0-24 Amendment No. 134 BFN Unit 2
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 RADI0 ACTIVE EFFLUENT RELEASE REPORT Deleted (See REM section F-2) 6.9.1.6 SOURCE TESTS Resultc A' required leak tests perf'ormed on sources if the tests rever1 the presence of 0.005 microcurie or more of removable contamination, l
l 6.0-25 BFN Amendment No. 128,134 Unit 2 i
i 6.9.2 SPECIAL REPORTS Reports on the following areas shall be submitted in writing to the Director of Regional Office of Inspection and Enforcement:
1.
Fatigue Usage 6.10.1.q Annual Operating Report 2.
Relief Valve Tailpipe 3.2.F Within 30 days after inoper-ability of thermocouple and acoustic monitor on one valve.
3.
Seismic Instrumentation 3.2.J.3 Within 10 days Inoperability after 30 days of inoperability.
4.
Meteorological Monitoring 3.2.I.2 Within 10 days Instrumentation after 7 days of Inoperability inoperability.
5.
Primary Containment 4.7.A.2 Within 90 days Integrated Leak Rate of completion of Testing each test.
6.
Data shall be retrieved from all seismic instruments actuated during a seismic event and ar.alyzed to determine the magnitude of the vibratory ground motion. A Special 6.0-26 BFN Amendment No. 134 Unit 2
s o
Report shall be submitted within 10 days after the event describing the magnitude, frequency spectrum, and resultant effect upon plant features important to safety.
7.
Secondary Containment 4.7.C.
Within 90 days Leak Rate Testing
- of completion of each test.
8.
High-Range primary 3.2.F Within 7 days Containment Radiation after 7 days of Monitors inoperability.
9.
High-Range Gasesous 3.2.F Within 7 days Effluent Radiation after 7 days of Monitors inoperability.
- Each integrated leak rate test of the secondary containment shall be the subject of a summary technical report. This report should include data on the wind speed, wind direction,-outside and inside temperatures during l
the test, concurrent reactor building pressure, and emergency ventilation flow rate. The report shall also include analyses and interpretations of I
those data which demonstrate compliance with the specified leak rate j
limits.
1 J
6.0-27 Amendment No. 134 BFN L_ __ ___ _ frhrnA
s 6.10 STATION OPERATING RECORDS AND RETENTION 6.10.1 Recorde and/or logs shall be kept in a manner convenient for review as indicated below:
a.
All normal plant operation including such items as power level, fuel exposure, and shutdowns b.
principal maintenance activities c.
Reportable Events d.
Checks, inspections, tests, and calibrations of components and systems, including such diverse items as source leakage e.
Reviews of changes made to the procedures or equipment or reviews of tests and experiments to comply with 10 CFR 50.59 f.
Radioactive shipments g.
Test results in units of microcuries for leak tests performed pursuant to Specification 3.8.D l
6.0-28 BFN Amendment No. 134 Unit 2
.)
h.
Record of annual physical inventory verifying accountability of sources on record
- i. Gaseous and liquid radioactive waste released to the environs
- j. Offsite environmental monitoring surveys I
k.
Fuel inventories and transfers 1.
Plant radiation and contamination surveys m.
Radiation exposures for all plant personnel J
n.
Updated, corrected, and as-built drawings of the plant o.
Reactor coolant system inservice inspection p.
Minutes of meetings of the NSRB q.
Design fatigue usage evaluation Monitoring and recording requirements below will be met for various portions of the reactor coolant pressure boundary (RCPB) for which detailed fatigue usage evaluation per the ASME Boiler and Pressure Vessel Code Section III was performed for the conditions defined in the design specification.
In this plant, the applicable codes require fatigue usage evaluation for the reactor pressure vessel only. The locations to be monitored shall be:
1.
The feedwater noztles 2.
The shell at or near the waterline 3.
The flange studs 6.0-29 BFN Amendment No. 134 Unit 2
e e
Transients that occur during plant operations will be reviewed and a cumulative fatigue usage factor determined.
For transients which are more severe than the transients evaluated in the stress report, code fatigue usage calculations will be made and tabulated separately.
In the annual operating report, the fatigue usage factor determined for the transients defined above shall be added and a cumulative fatigue usage factor to date shall be reported. When the cumulative usage factor reaches a value of 1.0, an inservice inspection shall be included for the specific location at the next scheduled inspection (3-1/3-year interval) period and 3-1/3-year intervals thereafter, and a subsequent evaluation performed in accordance with the rules of ASME Section XI Code if any flaw indications are detected. The results of the evaluation shall be submitted in a Special Report for review by the Commission.
6.10.2 Except where covered by applicable regulations, items a through h above shall be retained for a period of at least 5 years and item i through q shall be retained for the life of the plant. A complete inventory of radioactive materials in possession shall be maintained current at all times, i
l 1.
See paragraph N-415.2, ASME Section III, 1965 Edition.
J 0.0-30 I
Amendment No.134 l
BFN Unit 2
i
.c t
6.11 PROCESS CONTROL PROGRAM (PCP)
- 1. The.PCP shall be approved by the Commission prior to implementation.
- 2. Changes-to the PCP shall be submitted to the commission in the semi-annual Ra.dioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
a.
Sufficiently detailed information to totally support the change.
b.
A determination that the change did not change the overall conformance of the solidified product to existing criteria.
- 3. Changes to the PCP shall become effective upon review and
)
acceptance by PORC.
6.12 0FFSITE DOSE CALCULATIONAL MANUAL (ODCM)
- 1. The ODCM shall be approved by the Commission prior to implementation.
- 2. Changes to the ODCM shall be submittod to the Conunission in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made, This submittal shall contain:
a.
Sufficiently detailed information to totally support the l
change.
- 3. Changes to the ODCN shall become effective upon review end acceptance by PORC.
6.0-31 BFN Unit 2 Amendment No. 128, 134
t t
-6.13 RADIOLOGICAL EFFLUENT MANUAL (REM)
(
- 1. The REM shall be approved by the Commission prior to implementation.
l l
implementation.
- 3. Changes to the REM shall be approved by the Commission prior to implementation.
1 1
l
)
)
l l
l 4
1 i
l l
6.0-32 BFN Unit 2 Amendment No. 228, 134
n FN Ar o
N er I
gore e
AC EU p
L C&
U N
R RG O A N T
EI CF LN I EO CIA R UR ID NT R RS O AE T
EC CF LIV I
EO C R IR U E D NS E
t G
R R
N G
o R
RI A
A mA O AR N
n
_OAI.
M oF w CFEE A
T V
LE EO C N Ot I
N G y
R UI r
R u
D N
E e
N E
E G
o I EO C O R R O A T E CF L A R U I
N D
N O
R RT I
O AC T EU I
CF L R I
EO C T R US D NNO C
sn n ma l
o r
e.
ca nu oo" s
r*"
us U=
o.OeMm r a3m3" z? wZ 3,,
Cm
+c N I
g Ibo D
N N G R "ms AME i
f N
E 5G SNU EA n NA C
EN u
CA M
u=
N M
G A
N R
I G
R O
E R E N
I E
P S N
O
'"En R
N N
o I
NE NE T
G O C c
G N
N T
UL I U A
E mJ RN L
L D
E C
s I
A IA*
C N
E RE A
No N
D I
T.
OT N E
TOS M
RY l
H N
N O
A E c
I t
M T T C
R liR C P
l SE E
E UF P
Y E P U U
R R O DA S
R I
R N S T
T FR O S
T N
I I
M OC R AE I
R O A T
SR T TG E
F A EE SNAl E
ECG I AN H
E P I
TI A SLA
_ L C
Y R O I V A
T SRN SPM C
U RR E A A
I L BO O I
R SM s
G O I
I T ET l
R I
O R SNC O
L T NEA o
I N OSE O
P R TCl DO S-S E P
AC E
I R
RO R
RO I
- SR D
R TE NG I
MA N
A M
T
)
l EEN O
CD R
NN S
A E Es L
(
NT n
EN o
TCO ER TI RAN s
MR i
FINS E
mR O EI f E v)
IH H V )O E
A I
L R P
n (3 I
N C C E E E MU E
T SCD G
N S
P E U En.
IG u
EA R N N
S RNI s
T (E
I G
N E
T DFN RF EI OE EH l
S NS S
R GT I
N O NN O
I S TS EA
)
TIV)
F W )3 T
F A 3
R l
(
.l A R (3 I H R( FI IE TS RO Q R
I S O
E E E S C HS TN R EI EJ -
O U I Q RISL S N S
C A
I U
DIV M
D N m 4s
[=@k o$
E ;; ro
'l
n.
1
- pm ascoq'o, '
. UNITED STATES
'i M
NUCLEAR REGULATORY COMMISSION
[
g 5
t WASHINGTON, D. C. 20555
)
. TENNESSEE VALLEY AUTHORITY l
a DOCKET NO. 50-?96 i
BROWNS FERRY NUCLEAR PLANT, UNIT 3 i
i AMENDMENT.TO FACILITY OPERATING LICENSF Amendment No.109 i
License No. OPR-68
~1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for' amendment by Tennessee Valley Authority (the licensee) dated September 27, 1984 as supplemented January 17. June ?
and December 10, 1986, complies with the standards and requirements of.
the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set:forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the
- provisions.of the' Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by.
this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and-E.
The. issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical
. Specifications as indicated in the attachment to this license amendment and paragraph ?.C.(?) of Facility Operating License No. DPP-68 is hereby amended to read as follows:
i A
.2-(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as i
revised through Amendment No. 109, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of. issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 4
Jo
. Zwolinski, ssist t r
for Projects VA ojec,ts Division 0
ce of Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 11, 1987 k
ATTACHMENT TO LICENSE AMENDMENT NO. 109 FACILITY OPERATING NO. DPR-68 DOCKET NO. 50-296 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness.
Remove Insert iii iii*
iv iv v
v vi vi*
vii vii viii viii 3.2/4.2-3 3.2/4.2-3 3.2/4.2-4 3.2/4.2-4*
3.2/4.2-5 3.2/4.2-5 3.2/4.2-6 3.2/4.2-6 3.8/4.8-1 3.8/4.8-1*
3.8/4.8-2 3.6/4.8-2 3.8/4.8-3 3.8/4.8-3 3.8/4,8-4 3.8/4.8-4 3.8/4.8-5 3.8/4.8-5 3.8/4.8-6 3.8/4.8-6 3.11/4.11-3 3.11/4.11-3 3.11/4.11-4 3.11/4.11-4*
3.11/4.11-5 3.11/4.11-5*
3.11/4.11-6 3.11/4.11-6 3.11/4.11-7 3.11/4.11-7 3.11/4.11-8 3.11/4.11-8 3.11/4.11-12 3.11/4.11-12 3.11/4.11-13 3.11/4.11-13*
3.11/4.11-14 3.11/4.11-14*
3.11/4.11-15 3.11/4.11-15 6.0-1 thru 6.0-23 6.0-1 thru 6.0-34
- - - - - - - - - ~ - - - - _ - _ _ _ _ _ - - _ _. _.. _ _ _
Sectirn Pato No.
C.
Coolant Leakage....
3.6/4.6-9 D.
Relief Valves.................
3.6/4.6-10 l
l E.
Jet Pumps...................
3.6/4.6-11 1
F.
Recirculation Pump Operation.........
3.6/4.6-12 G.
Structural Integrity.
3.6/4.6-13 l
H.
Seismic Restraints, Supports and Snubbers...
3.6/4.6-15 3.7/4.7 Containment Systems................
3.7/4.7-1 A.
3.7/4.7-1 B.
Standby Gas Treatment System.......
3.7/4.7-13 C.
3.7/4.7-16 l
D.
Primary Containment Isolation Valves.....
3.7/4.7-17 E.
Control Room Emergency Ventilation......
3.7/4.7-19 F.
Primary Containment Purge System.
3.7/4.7-21 G.
Containment Atmosphere Dilution System (CAD) 3.7/4.7-22 H.
Containment Atmosphere Monitoring (CAN)
System H2 Analyzer............
3.7/4.7-23 3.8/4.8 Radioactive Materials..........
3.8/4.8-1 l
A.
Liquid Effluents.............
3.8/4.8-1 B.
Airborne Effluents..............
3.8/4.8-2 C.
Radioactive Effluents - Dose.
3.8/4.8-6 D.
Mechanical Vacuum Pump............
3.8/4.8-6 E.
Miscellaneous Radioactive Materials Sources 3.8/4.8-7 F.
Solid Radwaste................
3.8/4.8-9 3.9/4.9 Auxiliary Electrical System...........
3.9/4.9-1 A.
Aurillary Electrical Equipa.ent.
3.9/4.9-1 B.
Operation with Inoperable Equipment.
3.9/4.9-8 C.
Operation in Cold Shutdown........
3.9/4.9-14 3.10/4.10 Core Alterations....
3.10/4.10-1 A.
Refueling Interlocks.............
3.10/4.10-1
]
BFN Amendment No. 103 Unit 3
r-
-}
.y, O
g_,.f
' uh p/
' Y l1 i U '.3 (*
l r
n I
s
- . J j,
Section
.f.
f
).
Page No.
j
,/'
,(
g B.
Core Mori.toring.
'.i.,.J.
3.10/4.10-4
+
+"
' ' l'
- fM Sperit, Fuel Poal Water.............
.3.10/4.10-7
,e i
/
6 'i 3.10/4.10-8 h.
.!(4Geor. Building Crans,
t
.).
~t y
~}
e E.
5eenEfuel'Clik.................
3.10/4.10-9 j
3 i'f t.
F fj engf'luel Cask Handling-Refueling Floor...
k 3.10/4.10-9 j f 41 i-1 3.31/d.11 Firo Pnfo tion Systems..............
3.11/4.11-1 "s,
y 3 j
. t
.e A.
High Pressure Fire Protection System. >...
3.11/4.11-1
+
't B '.
CO2 Fire Protection System...
3.11/4.11-4 e.
J n C.
Fire Detectors'................
3.11/4.11-6 c
7;
,y D.
Deleted..........
/.
3.11/4.11-7
,L f
4 i
E.
Fire Protection Systems Inspec'.to.n,..
3.11/4.11-8
< +
y l
3.11/4.11-8 c
F.
Deleted.
3.................
. s i
t G.
Air Mas.6s and Cylinders............
3.11/4.11-8 H.
Continuou s Fire Watch.
3.11/4.11-9 e
Y I.
Open Flawes, Welding and Burning in the A
Sabl.e Spreading Room.
3.11/4.11-9 4
1 a
k 5.(b Major Design Features........ !......
5.0-1
.r o
5.1 Site Features......
5.0-1 5.2 Reactor...'..............'...
5.0-1
+
5.3 Ret: tor Vessel.........
5.0-1 5.0-1
/
5.4 Cont ainment.
q 5.5 Fuel Storage...........^......
5.0-1 i
5.6 Seismic Design................
5.0-2 t
%L kY Amendment No. 109 Bht 3 4
k 1
ADMINISTRATIVE CONTROLS PAGE SECTION 6.3 R ES PONS I B I LITY..........................................
6. 0 -1 6.2 ORGANIZATION............................................
6.0-1 6.2.1 Corporate...............................................
6.0-1 6.2.2 P3 ant Staff.............................................
6.0-1 6.3 PLANT STAFF QUALIFICATIONS..............................
6.0-4 6.4 T RA I NI NG.,..............................................
6. 0 - 4 6.5 P LANT REVI EW AND AUDIT..................................
6. 0 -4 6.5.1 Plant Operation Review Committee (P0RC).................
6.0-4 6.5.2 Nuclear Safety Review Board (NSRB)......................
6.0-10 6.5.3 Technical Review and Approval of Procedures.............
6.0-16 6.6 REPORTABLE EVENT ACTIONS................................
6.0-17 6.7 S AF ETY LIMIT VIO LATION..................................
6. 0-18 6.8 PROCEDURES / INSTRUCTION AND PROGRAMS.....................
6.0-19 6.8.1 Procedures..............................................
6.0-19 6.8.2 Dri11s..................................................
6.0-20 6.8.3 Radiation Control Procedures............................
6.0-20 6.8.4 Quality Assurance Procedures - Effluent and..............
Environmental Monitorin6..............................
6,0-22 6.9 REPORTING REQUIREMENTS..................................
6.0-23 6.9.1 Routine Reports.........................................
6.0-23 Startup Reports.........................................
6.0-23 Annual Operatin6 Report.................................
6.0-24 Monthly Operating Report................................
6.0-25 Reportable Events.......................................
6.0-25 Radioactive Effluent Release Report.....................
6.0-25 Source Tests............................................
6.0-25 6.9.2 Special Reports.........................................
6.0-26 6.10 STATION OPERATING RECORDS AND RETENTION.................
6.0-28 6.11 PROCESS CONTROL PR0 GRAM.................................
6.0-31 6.12 OFFSITE DOSE CALCULATION MANUAL.........................
6.0-31 6.13 RADIOLOGICAL EFFLUENT MANUAL............................
6.0-32 v
BFN Amendment No.109 Unit 3
LIST OF TABLES Table' Title Paae No_.
- 1.1 Surveillance Frequency Notation..........
1.0-12 3.1. A '
. Reactor Pro';ection System (SCRAM)
Instrumentation Requirements............
3.1/4.1-2 4.1.A Reactor Protection System (SCRAN) Instrumentation Functional Tests Minimum Functional Test Frequencies for Safety Instr. and Control Circuits.................
3.1/4.1-7 4.1.B Eeactor Protection System (SCRAM) Instrumentation Calibration Minimum Calibration Frequencies for Reactor. Protection Instrument Channels....
3.1/4.1-10 3.2.A Primary Containment and Reactor Building Isolation l
Instrumentation.................
3.2/4.2-7 I
3.2.8 Instrumentation that Initiates or Controls the Core and Containment Cooling Systems..
3.2/4.2-14 3.2.C Iristrumentation that Initiates Rod Blocks.....
3.2/4.2-24 3.2.D Radioactive Liquid Effluent Nonitoring Instrumentation..................
3.2/4.2-27 1
3.2.E Instrumentation that Monitors Leakage Into Drywell.
3.2/4.2-29 j
3.2.F Surveillance Instrumentation............
3.2/4.2-30 3.2.G Control Room Isolation Instrumentation.......
3.2/4.2-33 3.2.H Flood Protection Instrumentation..
3.2/4.2-34 I
l 3.2.I Noteorological Monitoring Instrumentation.....
3.2/4.2-35 3.2.J Seismic Monitoring Instrumentation..
3.2/4.2-36 3.2 K Radioactive Gaseous Rffluent Nonitoring Instrumentation.................
3.2/4.2-37 4.2.A Surveillance Requirements for Primary Containment and Reactor Building Isolation Instrumentation..
3.2/4.2-39
- 4.2.8 Surveillance Requirements for Instrumentation that L
-Initiate or Control the CSCS...........
3.2/4.2-43 1
4.2.C Surveillance Requirements for Instrumentation that Initiate Rod Blocks.......
3.2/4.2-49 4.2.D Radioactive Liquid Effluent Monitoring
)
Instrumentation Surveillance Requirements....
3.2/4.2-50 j
l 71 BFN Amendment No. 103 i
Unit 3 1
1
4 LIST OF TABLES (Cont'd)
Table Title Page No.
4.2.E Minimum Test and Calibration Frequency for Drywell Leak Detection Instrumentation..........
3.2/4.2-52 4.2.F Minimum Test and Calibration Frequency for Surveillance Instrumentation..........
3.2/4.2-53 4.2..G Surveillance Requirements for Control Room I
Isolation Instrumentation............
3.2/4.2-55 4.2.H Minimum Test and Calibration Frequency for j
Flood Protection Instrumentation........
3.2/4.2-56 1
l 4.2.J Seismic Monitoring Instrument Surveillance i
3.2/4.2-S7 l
Requirements...
4'.2.K Radioactive Gaseous Effluent Instrumentation Surveillance 3.2/4.2-61 3.5-1 Minimum RHRSW and EECW Pump Assignment.
3.5/4.5-11 i
)
3.5.I MAPLHGR Versus Average Planar Exposure......
3.5/4.5-21 l
J 3.7.A Primary Containment Isolation Valveb 3.7/4.7-25 3.7.B Testable Penetrations with Double 0-Ring Seals..
3.7/4.7-31 3.7.C Testable Penetrations with Testable Bellows..
3.7/4.7-32 3.7.D Air Tested Isolation Valves............
3.7/4.7-33 3.7.E Primary Containment Isolation Valves which Terminate below the Suppression Pool Water Level......................
3.7/4.7-36 3.7.F Primary Containment Isolation Vales Located in Water Sealed Seismic Class 1 Lines.
3.7/4.7-37 3.7.H Testable Electrical Penetrations.........
3.7/4.7-38 4.9.A.4.C VoJ9 age Relay Setpoints/ Diesel Generator Start.
3.9/4.9-15 3.11.A Fire Protection System Hydraulic Requirements...
3.11/4.11-10 6.2.A Minimum Shift Crew Requirements..........
6.0-3 Vil Amendment No. 103, 109 l
l sFw Unit 3
LIST OF ILLUSTRATIONS Figure Title Page No.
1 2.1.1 APRM Flow Reference Scram and APRM Rod Block i
Settings....................
1.1/2.1-6 i
2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow....
1.1/2.1-7 4.3-1 Graphic Aid in the Selection of an Adequate Interval Detween Tests...........
3.1/4.1-12 4.2-1 System Unavailability......
3.2/4.2-63 3.4-1 Sodium Pentaborate Solution Volume Concentrated i
Requirements..................
3.4/4.4-4 3.4-2 Sodium Pentaborate Solution Temperature l
Requirements...........
3.4/4.4-5 3.5.K-1 NCPR Limits.......
3.5/4.5-25 3.5/4.5-26 3.5.2 Kr Factor...
3.6-1 Minimum Temperature OF Above Change in Transient Temperature..
3.6/4.6-24 3.6-2 Change in Charpy V Transition Temperature Vs.
Neutron Exposure.....
3.6/4.6-25 4.8.1.a Gaseous Release Points and Elevation.....
3.8/4.8-10 4.8.1.b Land Site Boundary.
3.8/4.8-11 6.2-1 Offsite Organization for Facility Management and Technical Support 6.0-33 6.2-2 Facility Organization 6.0-34 I
i Amendment No. 103, 109 BFN Unit 3
e o
3.2/4.2 PROTECTIVE INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.D Radioactive Licuid Effluent 4.2.D Radioactive Licuid Effluent Monitoring Instrumentation Monitoring Instrumentation
- 1. The radioactive liquid
- 1. Each of the radioactive effluent monitoring liquid effluent monitoring instrumentation listed in instruments shall be Table 3.2.D shall be OPERABLE demonstrated OPERABLE by with the applicability as performance of test in shown in Tables 3.2.D/4.2.D.
accordance with Table 4.2.D.
Alarm / trip setpoints will be set in accordance with guidance given in the ODCM to ensure that the limits of specification 3.8.A.1 are not exceeded.
- 2. The action required when the number of operable channels is less than the minimum channels operable requirement is specified in the notes for Table 3.2.D.
Exert best efforts to rsturn the instrument (s) to OPERABLE status within 30 days and if l
unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.
- 3. With a radioactive liquid effluent monitoring channel alarm / trip setpoint less conservative than required by these specifications, suspend the release without delay, declare the channel inoperable, or adjust the alarm / trip setpoint to establish the conservatism required by these specifications.
- 4. The provisions of Specification 1.0.C and 6.9.1.4 are not applicable.
3.2/4.2-3 Amendment No. 203, 109 BFW Unit 3
c.
o 3.2/4.2 PROTECTIVE INSTRUMENTATION 1
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.E.
Drywell Leak Detection 4.2.E.
Drywell Leak Detection The limiting conditions of Instrumentation shall be operation for the calibrated and checked as instrumentation that monitors indicated in Tabla 4.2.E.
drywell leak detection are given in Table 3.2.E.
i F.
Surveillance Instrumentation F.
Surveillance Instrumentation
]
The limiting conditions for the Instrumentation shall be instrumentation that provides calibrated and checked as surveillance information indicated in Table 4.2.F.
readouts are given in Table 3.2.F.
G.
Control-Room Isolation G.
Control Room Isolation The limiting conditions for Instrumentation shall be instrumentation that isolates calibrated and checked as the control room and-initiates indicated in Table 4.2.G.
the control room emergency pressurization systems are given in Table 3.2.G.
H.
Flood Protection H.
Flood Protection The unit shall be shutdown and Surveillance shall be placed in the cold condition performed on the when Wheeler Reservoir lake instrumentation that monitors stage rises to a level such the reservoir level as 1
that water from the reservoir indicated in Table 4.2.H.
begins to run across the pumping station deck at elevation 565.
I Requirements for instrumentation that monitors the reservoir level are given in Table 3.2.H.
I.
Noteorological Monitoring I.
Noteorological Nonitoring Instrumentation Instrumentation The meteorological monitoring Each meteorological monitoring i
instrumentation listed in Table instrument channel shall be
(
I 3.2.1 shall be operable demonstrated operable by the at all times, performance of the CHANNEL CHECK at least once per SFN Unit 3 3.2/4.2-4 j
c 3.2/4.2 PROTECTIVE INSTRUMENTATION t
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS I
3.2.I.
Meteorological Monitoring-4.2.I.
Meteorological Monitoring
. Instrumentation (Cont'd)
Instrumentation (Cont'd) 1.
With the number of operable 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the CHANNEL meteorological monitoring CALIBRATION at least once channels less than required each 6 months.
by Table 3.2.I restore the inoperable channel (s) to operable status within 7 days.
2.
With one or more of the meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the 1
malfunction and the plans for restoring the system to operable. status.
J.
Seismic Monitoring 4.2.J.
Seismic Monitoring Instrumentation Instrumentation The seismic monitoring 1.
Each of the seismic instruments listed in monitoring instruments Table 3.2.J shall be shall be demonstrated operable at all times, operable by performance of tests at the frequencies listed in Table 4.2.J.
2.
With the number of 2.
Data shall be retrieved seismic monitoring from all seismic instruments J
instruments less than the actuated during a seismic number listed in Table event and analyzed to 3.2.J restore the determine the magnitude 3
inoperable instrument (s) of the vibratory ground l
to operable status within motion. A Special Report 30 days.
shall be submitted to the Conunission pursuant to 3.
With one or more of the Specification 6.9.2 instruments listed in Table within 10 days describing j
3.2.J inoperable for more the magnitude, frequency l
than 30 days, submit a spectrum, and resultant
]
Special Report to the effect upon plant features l
consnission pursuant to important to safety.
Specification 6.9.2 within the next 10 days describing the cause of the malfunction I
and plans for restoring the instruments to operable status.
3.2/4.2-5 Amendment No. 23. 109 Nt3 L
__ a
s 3.2/4.2 PROTECTIVE INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.K.
Radioactive Gaseous Effluent 4.2.K.
Radioactive Gaseous Effluent Monitoring Instrumentation Monitoring Instrumentation 1.
The radioactive' gaseous 1.
Each of the radioactive l
effluent monitoring gaseous effluent monitoring instruments listed in instruments shall be i
Table 3.2.K shall be operable demonstrated operable by with the applicability as performance of tests in shown in Tables 3.2.K/4.2.K accordance with Table 4.2.K.
l Alarm / trip setpoints will be set in accordance with guidance given in the ODCM to 1
4 ensure that the limits of specification 3.8.B.1 are not exceeded.
2.
The action required when the number of operable channels is less than the Minimum Channels Operable requirement is specified in the notes for Table 3.2.K.
Exert best efforts to return the instruments to operable 3
l status within 30 days and, if unsuccessful, explain in the next, Semiannual Radioactive J
Release Report why the inoperability was not corrected in a timely manner.
3.
With a radioactive gaseous effluent monitoring channel alarm / trip setpoint less conservative than required by these specifications, suspend the release without delay, declare the channel inoperable or adjust the alarm / trip setpoint to establish the conservatism required by these i
specifications.
1 4.
Both off-gas treatment monitors may be taken out of service for less than one hour for purging of monitors during SI performance.
5.
The provisions of Specifications 1.0.C and 6.9.1.4 are not applicable. 3.2/4.2-6 Amendment No. 203, 109 BFN Unit 3
1 g
3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8 Radioactive Materials 4.8 Radioactive Materials Applicability Applicability Applies to the release of Applies to the periodic test and radioactive liquids and gases record requirements and sampling L
from the facility.
and monitoring methods used for I
facility effluents.
Objective Objective To define the limits and To ensure that radioactive
~~
conditions for the release of liquid and gaseous releases from radioactive effluents to the the facility are maintained environs to assure that any within the limits specified by radioactive releases are as Specifications 3.8.A and 3.8.B low as reasonably achievable and within the limits of Speelfication IO CFR Part 20.
The specifications except for 3.8.A.1 and 3.8.B.1 are A.
Liould Effluents exempt from the requirements of definition 1.0.C (Limiting 1.
Facility records shall Condition for Operation).
be maintained of radioactive Specification concentrations and volume before dilution A.
Liould Effluents of each batch of liquid effluent released, and 1.
The concentration of of the average dilution radioactive material flow and length of time releared at any time from over which each the site to unrestricted discharge occurred.
areas (see Figure 4.8-1b) shall be limited to the 2.
Radioactive liquid waste concentrations specified sampling and activity in 10 CFR Part 20, analysis of each liquid Appendir B. Table II.
waste batch to be Column 2 for radionuclides discharged shall be other than dissolved or performed prior-to entrained noble gases.
release in accordance For dissolved or entrained with the sampling and noble gases, the analysis program concentration shall be specified in the REN.
limited to 2E-4 mci /ml total activity.
3.
The operation of the automatic isolation valves and discharge 2.
If the limits of 3.8.A.1 tank selection valves are exceeded, appropriate shall be checked action shall be initisted
- annually, without delay to brinc the release within j
BFN Unit 3 3.8/4.8-1 Amendment No. 103
o o
3.8/4.8 RADIOACTIVE KATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
\\
3.8.A.
Liauld Effluents 4.8.A.
Liauld Effluents l
limits. provide prompt 4.
The results of the notification to the NRC analysis of samples pursuant to Section 6.9.1.4.
collected from release points shall be used
- 3. The doses or dose with the calculational commitment to a member of methodology in the ODCM the public from radioactive to assure that the materials in liquid concentrations at the effluents released from point of release are each unit to unrestricted maintained within the areas (See Figure 4.8-lb) limits of Specification shall be limited:
3.8.A.I.
a.
During any calendar 5.
Cummulative quarterly quarter to <1.5 mrem to and yearly dose the total body and <5 contributions from mrem to any organ and, liquid effluents shall be determined as b.
During any calendar specified in the ODCM at year to <3 mrem to the least once every 31 days.
total body and <10 mrem to any organ 6.
The quantity of radioactive materiel
- 4. If the limits specified in contained in any 3.8.A.3 a & b above are outside liquid radwaste exceeded, prepare and storage tanks shall be submit Special Report determined to be within the pursue.nt to Section 6.9.1.4.
above limit by analyzing a representative sample of
- 5. The maximum activity to be the tar.k's contents at contained in one liquid least once per 7 days when radwaste tank or temporary radioactive materials are storage tank that can be being added to the tank.
discharged directly to the environs shall not exceed 10 curies excluding tritium and dissolved / entrained noble gas.
- 6. With radioactive liquid waste exceeding 3.8.A.5 limits, without delay suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, I
reduce the tank contents to l
within the limit. Events leading to this condition must be reported in the next Semiannual Radioactive Effluent Release Report (Section F.2 of the REM) 3.8/4.5-2 Amendment No. 203, 109 BFN Unit 3
4 0
3;8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS.
3.8.B.
Airborne Effluents 4.8.B.
Airborne Effluents
- 1. The dose rate at any time
- 1. The gross S/y and to areas at and beyond the particulate activity of site boundary (see Figure gaseous wastes released 4.8-lb) due to to the environment shall radioactivity released in be monitored and gaseous effluents from the recorded.
site shall be limited to the following values:
- a. For effluent streams having continuous a.
The dose rate limit monitoring for noble gases shall capability, the l
be <500 mrem /yr to the activity shall be total body and <3000 monitored and flow mrem /yr to the skin, rate evaluated and and recorded to enable release rates of b.
The dose rate limit gross radioactivity for I-131, I-133. H-3, to be determined at and particulate with least once per shift greater than eight day using instruments half-lives shall be specified in Table
<1500 mrem /yr to any 3.2.K.
organ.
- b. For effluent streams
- 2. If the limits of 3.8.B.1 without continuous are exceeded, appropriate monitoring corrective action shall be capability, the immediately initiated to activity shall be bring the release within monitored and limits. Provide prompt recorded and the notification to the NRC release through pursuant to Section these streams controlled to within 6.9.1.4.
the limits specified in 3.8.B.
- 2. Radioactive gaseous waste sampling and activity analysis shall be performed in accordance with the sampling and analysis program.specified in the j
)
REM. Dose rates shall be determined to be within limits of 3.8.B using methods contained in the ODCM.
3.8/4.8-3 Amendment No. 203,109 i
BFN Unit 3
3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.B.
Airborne Effluents 4.8.B. Airborne Effluents (Cont'd)
(Cont'd)
- 3. The air dose to areas at 3.
Cumulative quarterly and and beyond the site yearly dose boundary (see Figure contributions from 4.8-lb) due to noble gases gaseous releases shall released in gaseous be determined using effluents per unit shall methods contained in the be limited to the ODCM at least once every following:
31 days, s.
During any calendar quarter, to $5 mrad for gamma radiation and $10 mrad for beta radiation; b.
During any calendar year, to $10 mrad for gamma radiation and
$20 mrad for beta radiation.
- 4. If the calculated air dose exceeds the limits specified in 3.8.B.3 above, prepare and submit a special report pursuant to Section 6.9.1.4.
- 5. The dose to a member of the public from radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases with half lives greater than 8 days in gaseous effluent released per unit to areas at and beyond the site boundary (see Figure 4.8-lb) shall be limited i
to the following-a.
To any organ during j
any calendar quarter j
I to $7.5 mrem; b.
To any organ during any calendar year to
$15 mrem; j
3.8/4.8-4 Amendment No. J03,109 BFN Unit 3 l
q o
3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.B. Airborne Effluents 4.8.B. Airborne Effluents (Cont'd)
(Cont'd)
- 6. If the calculated doses 4.
During operation above exceed the limits of 25% power, the position 3.8.B.5 above, prepare of the charcoal bed and submit a special bypass valve will be report pursuant to verified daily.
Section 6.9.1.4.
- 7. During operation above 5.
The concentration of 25% power the discharge hydrogen downstream of of the SJAE must be the recombiners shall be routed through the determined to be within charcoal adsorbers, the limits of 3.8.B.9 by continuously monitoring
- 8. With gaseous waste being the off-gas whenever discharged for more than the SJAE is in service 7 days without treatment using instruments described in through the charcoal Table 3.2.K.
Instrument adsorbers, prepare and surveillance requirements are submit a special report specified in Table 4.2.K.
pursuant to Section 6.9.1.4.
- 9. Whenever the SJAE is in service, the concentration of hydrogen in the offgas downstream of the recombiners shall be limited to 54% by voinme.
- 10. With the concentration of hydrogen exceeding the limit of 3.8.B.9 above, restore the concentration to within the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
3.8/4.8-5 Amendment No. JD3,109 BFN Unit 3
+
r i
3.8/4.8 RADIOACTIVE MATERIALS LIMITING CONDITIONS FOR OPERATION._.
SURVEILLANCE REQUIREMENTS 3.8.C.
Radioactive Effluents - Dose 4.8.C.
Radioactive Effluents - Dose
- 1. The dose or dose 1.
Cumulative dose commitment to a real contributions from l
individual from all liquid and gaseous uranium fuel cycle sources effluents shall be is limited to $25 mrem to determined in I
the total body or any accordance with organ (except the thyroid, Specifications 3.8.A.3, which is limited to $75 3.8.B.3, and 3.8.B.5 mrem) over a period of one and the methods in the calendar year.
ODCM.
- 2. With the calculated dose from the releare of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 3.8.A.3, 3.8.B.3, or 3.8.B.5, prepare and submit a i
Special Report to the Commission pursuant to Specification 6.9.1.4 and limit the subsequent releases such that the limits of 3.8.C.1 are not exceeded.
3.8.D Mechanical Vacuum Pump 4.8.D Mechanical Vacuum Pump
- 1. Each mechanical vacuum pump At least once during each shall be capable of being operating cycle verify automatic automatically isolated and securing and isolation of the secured on a signal or mechanical vacuum pump.
high radioactivity in the steam lines whenever the main steam isolation valves are open.
- 2. If the limits of 3.8.D.1 are not met, the vacuum pump shall be isolated, i
3.8/4.8-6 Amendment No. 203,109 p
Unit 3 I
i
3.11/4.11 FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.A.
High Pressure Fire Protection 4.11.A.
High Pressure Fire System Protection System area where protection is lost is checked hourly.
3.
If only one high
- 3. Raw Service Water pressure fire pump is System Testing OPERABLE, the reactors may remain Item Freauency in operation for a period not to exceed Simulated Once/ year 7 days, provided the automatic requirements of and manual Specification actuation 3.11.A.1.b above of raw service are met.
water pumps and operation of tank level switches.
4.
If Specification
- 4. The high pressure l
3.11.A.3 cannot be fire protection met, the reactors system pressure shall be placed in shall be logged
)
the Cold Shutdown daily.
I condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
5.
Removal of any
- 5. Principal header component in the High and component Pressure Fire System isolation valves from service for any shall be checked reason other than open at intervals testing or emergency no greater than operations shall 3 months.
require Plant Manager approval.
6.
The Raw Service Water storage tank level shall be maintained above level 723'7" by the raw service water pumps.
3.11/4.11-3 Amendment No.109 BFN Unit 3 j
e a
3.11/4.11 FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.A.
High Pressure Fire Protection 4.11.A.
High Pressure Fire System Protection System 7.
If Specification 3.11.A.6 cannot be met, a fire pump shall be started and run continuously until the raw service water pumps can maintain a raw service water storage tank level above 723'7".
8.
The fire protection water distribution system shall have a minimum capacity of 2,664 gpm at 250' head.
9.
The fire protection system shall be capable of supplying the individual loads i
listed in Table I
3.11.A.
B.
CO2 Fire Protection System B.
CO2 Fire Protection I
System 1.
The CO2 Fire 1.
CO2 Fire Protection Protection System Testing:
shall be OPERABLE:
Item Freauency i
a.
With a minimum a.
Simulated Once/ year
)
of 8-1/2 tons automatic l
(0.5 Tank) CO2 and manual in storage units actuation 1 and 2.
1 i
b.
With a minimum b.
Storage Checked of 3 tons (0.5 tank daily Tank) CO2 pressure storage unit 3.
and level BFN Unit 3 3.11/4.11-4 l
____-----------------------_---_a
i 4
3.11/4.11 FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.B.
C07 Fire Protection 4.11.B.
CO2 Fire Protection System System 3.11.B.1 (Cont'd) 4.11.B.1 (Cont'd)
I c.
Automatic c.
CO2 Spray Once/3 l
initiation logic header and yeers OPERABLE.
nozzle inspection for blockage 2.
If Specifications 2.
When the cable 3.11.B.1.a or spreading room 3.11.B.I.b or CO2 fire 3.11.B.I.c cannot be protection met, a patrolling is INOPERABLE, fire watch with one 125-pound portable fire (or larger) equipment shall be portable fire established to ensure extinguisher that each area where shall be placed protection is lost at each entrance.
is checked hourly.
3.
If Specifications 3.11.B.1.a.
3.11.B.1.b, or 3.11.B.1.c are no' met within 7 days, the affected unit (s) shall be in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.
If CO2 fire protection is lost to a cable spreading room or to any diesel l
generator building area a continuous i
fire watch shall be I
established immediately and shall be continued until CO2 fire protection is restored.
BFN Unit 3 3.11/4.11-5
3.11/4.11 FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.B.
CO2 Fire Protection System 5.
Removal of any component in the CO2 Fire Protection System from service for any recson other than testinb or emergency operatione shall require Plant Manager approval.
C.
Fire Detectors 4.11.C.
Fire Detectors 1.
The fire detection 1.
All heat and smoke system's heat and detectors shall be smoke detectors for tested in accordance all protected zones with industrial shall be OPERABLE standards or other except that one approved methods detector for a given semiannually.
protected zone may be INOPERABLE for a period no greater than 30 days.
2.
If Specification 2.
The non-Class A 3.11.C.1 cannot be supervised detector met, a patrolling circuitry for those fire watch will be detectors which established to ensure provide alarm only that each protected will be tested once zone or area with each month by INOPERABLE detectors actuating the is checked at detector at the end intervals no greater of the line or end than one each hour.
of the branch such that the largest number of circuit conductors will be checked.
I 1
3.11/4.11-6 BFN Amendment No. 109 Unit 3 l
-c E.11/4.11 FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.11.C.
Fire' Detectors 3.
The class A supervised detector alarm circuitr will l
be tested once each two months at the-local panels.
4.
The circuits between the local panels in 4.11.C.3 and the main control room will be tested monthly.
5.
Smoke detector sensitivity will be checked in accordance with manufacturer's instruction annually.
3.11.D.
ROVING FIRE WATCH D.
ROVING FIRE WATCH Deleted Deleted k
3.11/4.11-7 BFN Amendment No.109 Unit 3 i
1.
l 3.11/4.11 FIRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPEFATION SURVEILLANCE REQUIREMENTS 3.11.E. Fire Protection Systems 4.11.E.
Fire Protection Systems inspection Inspections All fire barrier Each required fire penetrations, including barrier penetration shall cable penetration barriers, be verified to be fire doors and fire functional at least once dampers, in fire zone per 18 months by a visual boundaries protecting inspection, and prior to Fafety-related areas shall restoring a fire barrier be functional at all times.
to functional status With one or more of the following repairs or required fire barrier maintenance by penetrations nonfunctional performance of a visual within one hour establish inspection of the i
a continuous fire watch affected fire barrier on at least one side of penetration, the affected penetration or verify the OPERABILITY of fire detectors on at least one side of the nonfunctional fire barrier and establish an hourly fire watch patrol until the work is completed and the barrier is restored to functional status.
F. Fire Protection Organization F.
Fire Protection Organization I
Deleted Deleted G. Air Masks and Cylinders G.
Air Masks and Cylinders A minimum of 15 air No additional masks and 30 SCO-cubic-inch surveillance required.
l air cylinders shall be available at all times except that a time period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following emergency use is allowed i
to permit recharging or replacing.
l 3.11/4.11-8 BFN Amendment No. 109 Unit 3 l
l
a 3.11 BASES The High Pressure Fire and CO2 Fire Protection specifications are provided in order to meet the preestablished levels of operability during a fire in either or all of the three units. Requiring a patrolling fire watch with portable fire equipment if the automatic initiation is lost will provide (as does the automatic system) for early reporting and immediate fire fighting capability in the event of a fire occurrence.
The High Pressure Fire Protection System is supplied by four pumps (three electric driven and one diesel driven) aligned to the high pressure fire header. The reactors may remain in operation for a period not to exceed seven days if three pumps are out-of-service.
If at least two pumps are not made operable in seven days or if all pumps are lost during this seven-day period, the reactors will be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
For the areas of applicability, the fire protection water distribution system minimum capacity of 2,664 spm at 250' head at the fire pump discharge consists of the following design loads:
1.
Sprinkler System (0.30 gpm/f t /4440 f t2 area) 1332 spm 2
2.
I 1/2" Hand Hose Lines 200 spm 3.
Raw Service Water Load 1132 tom TOTI.L 2664 spm Fire Protection System is considered operable with a minimum of The CO2 8 1/2 tons (0.5 tank) CO2 in storage for units 1 and 2; and a minimum of 3 tons (0.5 tank) CO2 in storage for unit 3.
An immediate and continuous fire watch in the cable spreading room or any diesel generator building area will be established if CO2 fire protection is lost in this room and will continue until CO2 fire protection is restored.
To assure close supervision of fire protection system activities, the removal from service of any component in either the High Pressure Fire System or the j
Fire Protection System for any reason other than testing or emergency j
CO2 operations will require Plant Manager approval.
Early reporting and immediate fire fighting capability in the event of a fire occurrence will be provided (as with automatic system) by requiring a patrolling fire watch if more than one detector for a given protected zone is inoperable.
'l 3.11/4.11-12 Amendment No. 109 BFN Unit 3
3.11 BASES (Cent'd)
The fire protection system is designed to supply the required flow and pressure to an individual load listed on Table 3.11.1 while maintaining a design raw service water load of 1132 spm.
4.11 BASES Periodic testing of both the High Pressure Fire System and the CO2 Fire Protection System will provide positive indication of their operability.
-If only one of the pumps supplying the High Pressure Fire System is OPERABLE, the pump that is OPERABLE will be checked immediately and daily thereafter to demonstrate operability.
If the CO2 Fire Protection System becomes INOPERADLE in the cable spreading room, one 125-pound (or larger) fire extinguisher will be pieced at each entrance to the cable spreading room.
Annual testing of automatic valves and control devices is in accordance.
l with NFPA Code Vol. II, 1975, section 15, paragraph 6015. More frequent i
testing would require excessive automatic system inoperability, since l
there are a large number of automatic valves installed and various I
portions of the system must be isolated during an extended period of time l
during this test.
1 Wet fire header flushing, spray header inspection for blockage, and J
nozzle inspection for blockage will prevent, detect, and remove buildup of sludge or other material to ensure continued operability.
System flushes in conjunction with the semiannual addition of blocide to the Raw Cooling Water System will help prevent the growth of crustaceans which could reduce nozzle discharge.
Semiannual tests of heat and smoke detectors are in accordance with NFPA Code.
With the exception of continuous strip heat detector panels, all j
non-clas,s A supervised detector circuits which provide alarm only are hardwired through conduits and/or cable trays from the detector to the q
main control room alarm panels with no active components between, j
Nonclass A circuits also actuate the HPCI watcr-fos system, the CO 2 system in the diesel generator buildings, and isolate ventilation in i
shutdown board rooms. The test frequency and methods specified are i
justified for the following reasons:
1.
An analysis was made of worst-case fire detection circuits at Browns Ferry to determine the probability of no undetected failure of the circuits occurring between system test times as specified in j
the surveillance requirements. A circuit is defined as the wire 1
connections and components that affect transmission of an alarm
}
i signal between the fire detectors and the control room annunciator. Three circuits were analyzed which were representative of an alarm-only circuit, a water-fog circuit, and a 1
CO2 circuit. The spreading room B smoke detector was selected as i
i the worst-case alarm-only circuit because it had the largest number of wires and connections in a single circuit.
The HPCI water-fog l
circuit was selected for analysis because it is the only water-fog BFN Unit 3 3.11/4.11-13 I
j 4.11 BASES (C:nt'd) circuit in the area of applicability for Technical Specifications.
The Standby Diesel Generator Room A CO2 circuit was selected because it contained 2 out of 3 detector logic, the most j
complicated CO2 circuit logic. Calculations were based on l
failure rates for wires, connections, and circuit components as i
shown in Appendix III of WASH-1400. Failure rates were considered for the following circuit components:
J 1.
Open circuit 2.
Short to ground 1
3.
Short to power 4.
Timing motor failure to start 5.
Relay failure to energize 6.
Normally open contact failure to close 7.
Normally open or normally closed contact short 8.
Normally closed contact opening 9.
Timing switch failure to transfer The calculated probabilities (Pf) for no undetected failure of the circuits occurring were as follows, based on the specified test frequency.
j l
AREA TEST FREQUENCY Pt 1
Spreading Room B One Month 0.975287 HPCI Water Fog Six Months 0.977175 i
Standby Diesel Gen Room A CO2 Six Months 0.957595
{
l The worst case of the three areas considered is Spreading Room B.
The probability of undetected failure is approximately 1/40, which means that one undetected failure will occur on the average every 40 months over an extended period of time and that the failure could exist up to one month. The frequency of testing is thus much greater than the frequency of failure and produces circuits with
)
adequate reliability.
2.
Circuits checks by initiation of end of the line or end of the branch detectors will more thoroughly test the parallel circuits than testing on a rotating detector basis. This test is not a detector test, but is a test to simulate the effect of electrical supervision as defined in the NFPA Code.*
3.
Testing of circuits which actuate CO, water, or ventilation 2
systems requires disabling the automatic. feature of the fire protection system for the area. A surveillance program which disabled these circuits monthly would significantly reduce the I
ability of these circuits to provide fire suppression.
- Ref: NFPA Code 72D-9, paragraph lill, Code 72D-15, paragraph 1312 for definition of Class A systems, and Code 72A-18 Article 240.
BFN Unit 3 3.11/4.11-14 I
4.11 BASES (C@nt'd) 4.
Daily tests of annunciation lights and audible devices are performed as a routine operation function.
5.
The CO2 system manufacturer recommends semiannual testing of CO2 system fire detection circuits.
In addition, operating personnel periodically inspect the plant during their normal operating activities for fire har.ards and other abnormal conditions.
Smoke detectors will be tested "in-place" using inert freon gas applied by a pyrotronics type applicator which is accepted throughout the industrial fire protection industry for testing products of combustion detectors or by use of the MSA chemical smoke generators. At the present time, the manufacturers have only approved the use of " punk" for creating smoke. TVA will not use
" punk" for testing smoke detectors.
1 l
l 4
l l
3.11/4.11-15 Amendment No. 109 BFN Unit 3
I i
J f
l BFN TECHNICAL SPECIFICATIONS l
r 6.0 ADMINISTRATIVE CONTROLS l
l l
-6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY The Plant Manager has onsite responsibilities for the safe operation of the facility and shall report to the Browns Ferry Site Director. In the absence of the Plant Manager, a Plant Superintendent will assume his responsibilities.
6.2 ORGANIZATION CORPORATE 6.2.1 The portion of TVA management which relates to the operation of the plant is shown in Figure 6.2-1.
PLANT STAFF 6.2.2 The functional organization for the operation of the plant shall be shown in Figure 6.2-2.
- a. Shift manning requirements, shall as a minimum, be as described in Table 6.2.A and below.
b.
A licensed senior reactor operator shall be present at the site at all times when there is fuel in the reactor.
6.0-1 Amendment No. 109 BFN Unit 3 i
6.2.2 (Cont.)
c.
A licensed reactor operator shall be in the control room whenever there is fuel in the reactor, d.
Two licensed reactor operators shall be in the control room during any cold startups, while shutting down the reactor, and during recovery from unit trip.
In addition, a person holding a senior operator license shall be in the control room for that unit whenever it is in an operational mode other than cold shutdown or refueling.
e.
A Health physics Technician
- shall be present at the facility at all times when there is fuel in the reactor.
f.
A person holding a senior operator license or a senior operator license limited to fuel handling, shall be present during alteration of the core to directly supervise the activity and during this time shall not be assigned other duties.
g.
A site fire brigade of at least five members shall be maintained onsite at all times.* The fire brigade shall not include the Shift Engineer and the other members of the minimum shift crew necessary for safe shutdown of the unit, nor any personnel required for other essential functions during a fire emergency.
- The Health physics Technician and fire brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.
6.0-2 knendment No. 109 BFN Unit 3 l
)
Table 6.2.A Minimum Shift Crew Reautrementsb Position Units in Operation Type of License d
a 1
Z 3
Senior Operatora 1
1 1
1 SRO Senior Operator 0
1 2
2 SRO Licensed Operators 3
3 3
0 1
2 2
RO or SRO Additional Licensed Operators Assistant Unit Operators (AUD) 4 4
5 5
None Shift Technical Advisor (STA) 0 1
1 1
None Health Physics Technician 1
1 1
1 None Note for Table 6.2.A a.
A senior operator will be assigned responsibility for overall plant operation at all times there is fuel in any unit.
b.
Except for the senior operator discussed in note "a",
the shift crew composition may be one less than the minimum requirements of Table 6.2.A for a period of time not to exceed two hours in order to acconnodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2.A.
This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.
c.
One of the Additional Licensed Operators must be assigned to each control room with an operating unit.
d.
The number of required licensed personnel, when the operating units are controlled from a common control room, are two senior operators and four operators.
6.0-3 BFN Amendment No.109 Unit 3
-~
.,;~~
'a s
[~
'6.3 PLANT STAFF QUALIFICATIONS
.. L J
Qualifications of the Browns Ferry Nuclear Plant management and' operating staff shall meet the minimum acceptable levels as described in ANSI -
h18.1, Selectsion: and Training of Nuclear Power Plant Personnel, dated a
.t March 8",;1971) The qualifications of the Health Physics Supervisor will meet'or"e$did'theminimumacceptablelevelsasdescribedinRegulatory
.i Guide 1.8, Sevision 1, dated September'1975. The Shift Technical Advisor i
shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training.in plant design and transient and accident response and analysis.
.'t 6.4 TRAINING i
A retraining and replacement training program for station personnel shall V
be in accordance with ANSI - N18.1, Selection and Training of Nuclear 1
d' Power Plant Personnel, dated March 8, 1971. The minimum frequency of the retraining program shall be every two years.
6.5 PLANT REVIEW AND AUDIT 3
6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) r FUNCTION
]
t W
l 6.5.1.1 a.
The PORC shall function to advise the Plant Manager in all watters relaf.ed to nuclear safety.
r
./
b.
This advisory function shall be performed by the PORC acting j
in a formal meeting or by members acting individually without j
m 1
a formal meeting.
.,s.
]
l l
a 6.0-4 g
Amendment No. 109 BFN l
Unit 3 LY _ -_ - _--_ -_ __ _ _ _
{
COMPOSITION 6.5.1.2 The PORC shall be composed of the:
- a. Chairman:
Plant Manager Alternate Chairman:
Assistant to Plar.t Manager Alternate Chairman or Member:
Technical Services Superintendent Member:
Unit Superintendents (3)
Member:
Maintenance Superintendent Member:
Quality Assurance Staff Supervisor Member:
Health Physics Supervisor
- b. All alternate chairmen and alternate members shall be appointed in writing by the PORC chairman.
MEETING FREQUENCY 6.5.1.3 The PORC shall convene in a formal meeting at least once a month and as directed by the chairman. Other PORC meetings may be requested by the chairmen or members as required.
1 6.0-5 Amendment No. 109 BFH Unit 3
)
6.5.1.'4.L For expedited meetings, when it' is not practical to convene as a group, the chairman or alternate chairman may conduct committee business by polling the members individually-(by telephone or.in
.. person) or via.a serialized review.
QUORUM 6.5.1.5 'The quorum necessary for the PORC to act in a formal meeting shall consist of the chairman or alternate. chairman and at least five members or their alternates. Members shall be considered present if they are in' telephone communication with the committee.
RESPONSIBILITIES 6.5.1.6 The PORC shall be responsible for the activities listed below.
The PORC may delegate the performance of reviews, but will maintain cognizance over and responsibility for them, e.g.,
subcommittees.
Review of administrative procedures for the control of the a.
technical and cross-disciplinary review of (1) all procedures required by Specification 6.8.1.1, and changes thereto, (2) any other procedures and changes thereto determined by the Plant Manager to affect nuclear safety, b.
Review of the administrative procedures required by Appendix A of Regulatory Guide 1.33, Revision 2, February 1978 and changes thereto.
I Review of emergency operating procedures and changes thereto, c.
d.
Review implementing procedures of the Radiological Emergency Plan and the Industrial Security Program.
BFN Amendment No. 109 Unit 3 6.0-6
_._-. _ -. - - - _ - - - - - - - - = - - - =
'a e.
Review of all proposed changes to the Technical Specifications.
l f.
Review of safety evaluation for proposed tests or experiments to be completed under the provisions of 10 CFR 50.59 g.
Review proposed changes to the Radiological Effluent Manual.
h.
Review adequacy of the Process Control Program and Offsite Dose Calculation Manual at least once every 24 months.
- i. Review changes to the radwaste treatment systems.
- j. Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendation, and disposition of the corrective action to prevent recurrence to the Director, Nuclear Power and to the Nuclear Safety Review Board.
k.
Review of all safety evaluations for modifications to structures, syster.s or components that affect nuclear safety to verify that such actions did not constitute an unreviewed safety question as defined in 10 CFR 50.59, or requires a change to these Technical Specifications, i
BFN Amendment No. 109 Unit 3 6.0-7 1
1.
Review of reporteble events, unusual events, operating anomalies, and abnormal performance of plant equipment.
u Investigate reported or suspected incidents involving safety m.
questions or violations of the Technical Specifications.
Review of unit operations to detect potential hazards to n.
nuclese safety.
Items that may be included in this review are NRC inspection reports, QA audit. NSRB audit results, American Nuclear Insurer (ANI) inspection results, and significant corrective action reports (CARS).
Performance of special reviews, investigations, or analysis, o.
and report thereon as requested by the Plant Manager or the Nuclear Safety Review Board.
l l
l l
6.0-8 Amendment No.109 BFN Unit 3
AUTHORITY 6.5.1.7 The PORC shall:
a.
Recommend to the Plant Manager in writing, approval, or disapproval of items considered under 6.5.1.6.a through i above.
1.
The recommendation shall be based on a majority vote of the PORC at a fomal meeting.
2.
The recommendation shall be based on a unanimous vote of the PORC when the PORC members are acting individually.
3.
Each member or alternate member shall have one vote.
b.
Furnish for consideration a determination in writing with regard to whether or not each item considered under 6.5.1.6.f above constitutes an unreviewed safety question.
c.
Make recommendations to the Plant Manager in writing concerning whether action reviewed under 6.5.1.6.k above did not constitute an unreviewed safety question.
d.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Site Director and the Nuclear Safety Review Board of disagreements between the PORC and the Plant Manager.
However, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.
BFN 6.0-9 Amendment No.109 Unit 3
RECORDS 6.5.1.8 The.PORC shall maintain written minutet of each PORC meeting including expedited moetings that, as a minimum, document the result of all PORC activities performed under the responsibility and authority provisions of these technical specifications.
Copies shall be provided to the Site Director and the Nuclear Safety Review Board.
6.5.2 NUCLEAR SAFETY REVIEW BOARD FUNCTION l
6.5.2.1 The NSRB shall function to provide independent review and audit I
cognizance of designated activities in the areas of:
a.
Nuclear power plant operations b.
Nuclear engineering c.
Chemistry and radiochemistry d.
Metallurgy e.
Instrumentation and control f.
Radiological safety g.
Mechanical and electrical engineering, and h.
Quality assurance practices COMPOSITION 6.5.2.2 The NSRB shall be composed of at least five members, including the Chairman. Members of the NSRB may be from the Office of Nuclear Power or other TVA organizations, or external to TVA.
6.0-10 BFN Amendment No. 109 Unit 3
r
- 3 a.-
4 I
l QUALIFICATIONS 6.5.2.3 The Chairman, members, alternate members of the NSRB shall be 1
appointed in writing by the Menager of Nuclear power and shall have an academic degree in engineering or a physical science field, or the equivalent; and in addition, shall have a minimum
)
I of 5 years technical experience in one or more areas given in 6.5.2.1.
No more than two alternates shall participate as voting members in NSRB activities at any one time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized to provide expert advice as determined by the NSRB.
NEETING FREQUENCY 6.5.2.5 The NSRB shall meet at least once per six months.
QUORUM 6.5.2.6.The minimum quorum of the NSRB necessary for the performance of the NSRB review and audit functions of these technical specifications shall consist of more than half of the NSRB membership or at least five members, whichever is greater. The quorum shall include the Chairman or his appointed alternate and the NSRB members including appointed alternate members meeting the requirements of 6.5.2.3.
No more than a minority of the quorum shall have line responsibility for operation of the unit.
6.0-11 l
1 BFN Amendment No. 1091 l
i Unit 3 j
=
REUIEW 6.5.2.7 The NSRB shall review:
a.
The safety evaluations for:
(1) changes to procedures, equipment or systems, and (2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
b.
Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
)
Proposed tests or experiments which involve an unreviewed c.
safety question as defined in Section 50.59, 10 CFR.
d.
Proposed changes to Technical Specifications or this Operating License, Violations of Codes, regulations, orders, Technical l
e.
Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
f.
Significant operating abnormalities or deviations fro:n normal and expected performance of plant equipment that affect nuclear safety.
g.
All Reportable Events h.
All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and 1.
Reports and meeting minutes of the p0RC.
)
6.0-12 Amendment No.109 BFN Unit 3
AUDITS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the NSRB. These audits shall encompass:
The conformance of plant operation to provisions contained a.
within the Technical Specifications and applicable license conditions at least once per 12 months, b.
The performance, training and qualifications of the entire plant staff at least once per 12 months.
The results of actions taken to correct deficiencies c.
occurring in site equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.
i d.
The performance of activities required by the Operational Quality Assurance Program to moet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months.
The Site Radiological Emergency Plan and implementing e.
procedures at least once every 12 months.
f.
The Plant Physical Security Plan and implementing procedures at least once every 12 months.
g.
Any other area of site operation considered appropriate by the NSRB or the Nanager of Nuclear Pcwor.
h.
The fire protection programmatic controls including the implementing procedures at least once per 24 months.
6.0-13 BFN Amendment No. 109 Unit 3
s 1.
An independent fire protection and loss prevention program inspection and audit shall be performed annually utilizing either qualified offsite license personnel or an outside fire protection firm.
J.
An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervals no greater than 3 years.
k.
The Radiological Environmental Monitoring program and the results thereof at least once per 12 months.
1.
The performance of activities required by the Quality Assurance Progran to meet the criteria of Regulatory Guide 4.15, December 1977, or Regulatory Guide 1.21, Rev. 1, 1974, and Regulatory Guide 4.1, 1975, at least once every 12 months.
m.
The performance of activities required by the Safeguards Contingency Plan to meet the criteria of 10 CFR 73.40(d) at least once every 12 months.
n.
The Offsite Dose Calculation Manual and implementing i
procedures at least once per 24 months.
1 o.
The Process Control Program and implementing procedures for solidification of wet radioactive wastes at least once per 24 months.
p.
The Radiological Effluent Manual and implementing procedures at least once per 12 months.
I 6.0-14 BFN U"It 3 Amendment No. 103, 109
)
s AUTHORITY.
i 6.5.2.9 The NSRB shall report to and advise the Manager of Nuclear Power on those areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8.
RECORDS 6.5.2.10 Reports of activities shall be prepared, approved, and distributed as indicated below, a.
Minutes of each NSRB meeting shall be prepared, approved and forwarded to the Manager of Nuclear Power within 14 days following each meeting, b.
Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Manager of Nuclear Power within 14 days following completion of the
- review, c.
Audit reports encompassed by Specification 6.5.2.8 above, shall be forwarded to the Manager of Nuclear Power and to the management positions responsible for the areas audited within 30 days after completion of the audit.
1
)
6.0-15 BFN Amendment No. 109 Unit 3
l s
0 l
6.5.3 TEQiNICAL REVIEW AND APPROVAL OF PROCEDURES ACTIVITIES 6.5.3.1 Procedures required by Technical Specification 6.8.1.1 and other procedures which affect plant nuclear safety, and changes (other than editorial or typographical changes) thereto, shall be prepared, reviewed and approved. Each procedure or procedure change shall be reviewed by an individual other than the preparer. The reviewer may be f rom the same organization or from a different l
organization. Procedures other than Site Director Standard Practices will be approved by the responsible Section Supervisor, or applicable Plant Superintendent.
6.5.3.2 Proposed changes or modifications to plant structures, systems and components that affect nuclear safety shall be reviewed as designated by the Plant Manager. Each such modification shall be reviewed by an individual / group other than the individual / group which designed the modification, but who may be from the same organization as the individual / group which designed the modification. Proposed modifications to plant structures, systems and components that affect nuclear safety shall be approved by the Plant Manager, prior to implementation.
6.5.3.3 Individuals responsible for reviews perfomed in accordance with 6.5.3.1 shall be members of the site supervisory staff previously designated by the Plant Manager.
Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary.
If deemed necessary, such review shall be performed by review personnel of the appropriate discipline.
6.5.3.4 The Plant Manager shall approve all administrative procedures requiring PORC review prior to implementation.
hnt3 6.0-16
i j
1 6.6 ' REPORTABLE EVENT' ACTION
)
1 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
a.
The Comission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part
{
50, and b.
Each REPORTABLE EVENT shall be reviewed by the PORC and the i
results of this review shall be submitted to the NSRB and the Site Director.
1 6.0-17 BFN Amendment No. 109 Unit 3
s a
6.7 SAFETY LIMIT VIOLATION l
6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
I The NRC Operations Center shall be notified by telephone as a.
soon as possible and in all cases within I hour. The Manager of Nuclear Power and the NSRB shall be notified j
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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 the c.
Commission, the NSRB and the Manager of Nuclear Power within 14 days of the violation.
d.
Critical operation of the unit shall not be resumed until authorized by the Commission.
l l
i i
1 1
6.0-18 Amendment No. 109 BFN l
Unit 3
- 4 v
'6.8 -PROCEDURES / INSTRUCTIONS AND PROGRAW) 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.
b..
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.
Security plan implementation.
e.
. Emergency plan implementation.
f.
Fire Protection Program implementation.
g.
Radiological Effluent Manual implementing procedures.
h.
Process Control Program (PCP).
i.
Offsite Dose Calculation Manual.
- j. Administrative procedures which control technical and cross-disciplinary review.
6.8.1.2 Each administrative procedure required by Section 6.8.1.1.a.
shall be reviewed by PORC and all other procedures required by Section 6.8.1.1.a. shall be reviewed in accordance with Section 6.5.3.
j 1
BFN Amendment No. 109 I
Unit 3~
6.0-19 i
I 1
6.8.1.3 Temporary changes to procedures of Specification 6.8.1.1 may be made provided:
i The intent of the original procedure is not altered; j
a.
l l
b.
The change is approved by two members of the plant management staff, at least one of whom holds a Senior i
Operator License on the unit affected; The change is documented, reviewed by the PORC and approved 1
c.
by the Plant Manager within 14 days of implementation, for changes in administrative procedures requiring PORC review.
i I
I d.
The change is documented, reviewed per Specification 6.5.3, and approved by the responsible group section supervisor within 14 days of implementation, for changes to procedures other than administrative procedures.
DRILLS 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 Radiation Control Procedures shall be maintained and made available to all station personnel. These procedures shall show permissible radiation exposure and shall be consistent 6.0-20 Amendment No. 109 BFN Unit 3
h 1
g t.
I with the requirements of 10 CFR 20.
This radiation protection program shall be organized to meet the requirements of 7.0 CFR 20 except in lieu of the " control device" or " alarm signal" required by paragraph 20.203 (c) of 10 CFR 20.
i 6.8.3.1 Each high' radiation area in which the intensity of radiation is l
greater than 100 mrom/hr but less than 1000 mrem /hr 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.* Any individual or group of f
individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
l A radiation monitoring device which continuously indicates a.
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, An individual qualified in radiation protection procedures c.
i who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the arca 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 mrom/hr shall be subject to the provisions of (1) above; and, in addition, access to the source and/or area 6.0-21 Amendment No.109 BFN
-Unit 3
0 1
shall be secured by lock (s). The key (s) shall be under the administrative control of the shift engineer. In the case of a l
l 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.
I QUALITY ASSURANCE PROCEDURES - EFFLUENT AND ENVIRONMENTAL
{
l MONITORING i
Quality Assurance procedures shall be established, implemented, 6.8.4 and maintained for effluent and environmental monitoring, using l
the guidance in Regulatory Guide 1.21, Rev. 1, June 1974 and j
Regulatory Guide 4.1, Rev. 1. April 1975 or Regulatory Guide 4.15, Dec. 1977.
I I
i I
l 1
6.0-22 BFN Unit 3 Amendment No. 103,109
6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting require.nents of Title 10, code of Federhl Regulations, the following identified reporte shall be submitted to the Director of the Regional Office of NRC, unless otherwise noted.
6.9.1.1 STARTUP REPORT A summary report of plant startup and power escalation a.
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 l
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 f:ests identified in the FSAR and shall include a desc[iptionofthemeasuredvaluesoftheoperating conditions or characteristics obtained 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 descrited. Any additional specific details required in license conditions based on other commitments shall be included in this report.
6.0-23 Amendment No.109 BFN Unit 3
F b.
Startup reports shall be submitted within (1) 90 days following completion of the startup test pcogram, (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 ANNUAL OPERATING REPORT
- A tabulation on an annual basis of the number of station, a.
utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr 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 l
1 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, et least 80% of the total whole body dose 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 l
reactor pressure shall be reported.
- A single submittal may be made for a multiple unit station.
- This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.
6.0-24 Amendment No. 109 BFN Unit 3
r D'
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 Conunission.
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) l 6.9.1.6 SOUFCE TESTS Results of required leak tests performed on sources if the tests reveal the presence of 0.005 microcurie or more of removable contamination.
i l
I L
t L
j l
6.0-25 BFN Unit 3 Amendment No. 103,109 l
_______ ___ _ __ j
l o
l l
l l
l 6. 9. '2 SPECIAL REPORTS l
1 Reports on the following areas shall be submitted in writing to f
the Director of Regional Office of Inspection and Enforcement' i
1.
Fatigue Usage 6.10.1.q Annual Operating l
Report i
l l
2.
Relief Valve Tailpipe 3.2.F Within 30 days after inoper-ability of thermocouple and acoustic monitor on one valve.
3.
Seismic Instrumentation 3.2.J.3 Within 10 days Inoperability after 30 days of inoperability.
4.
Meteorological Monitoring 3.2.I.2 Within 10 days Instrumentation after 7 days of Inoperability inoperability.
5.
Primary Containment 4.7.A.2 Within 90 days Integrated Leak Rate of completion of Testing each test.
6.
Data shall be retrieved from all seismic instruments
[
actuated during a seismic event and analyzed to determine the magnitude of the vibratory ground motion. A Special i
6.0-26 i
BFN Amendment No. 309 L_
___. 3DNR3.
Report shall be submitted within 10 days after the event describing the magnitude, frequency spectrum, and resultant effect upon plant features important to safety.
7.
Secondary Containment 4.7.C.
Within 90 days Leak Rate Testing
- of completion of each test.
l
- Each integrated leak rate test of the secondary containment shall be the subject of a summary technical report. This report should include data on the wind speed, wind direction, outside and inside temperatures during the test, concurrent reactor building pressure, and emergency ventilation flow rate. The report shall also include analyses and interpretations of those data which demonstrate compliance with the specified leak rate limits.
6.0-27 BFN Amendment No. 109 Unit 3
a It 6.10 STATION OPERATING RECORDS AND RETENTION 6.10.1 Records and/or logs shall be kept in a manner convenient for review as indicated below:
All normal plant operation including such items as power a.
level, fuel exposure, and shutdowns b.
Principal maintenance activities c.
Reportable Events i
d.
Checks, inspections, tests, and calibrations of components and systems, including such diverse items as source leakage Reviews of changes made to the procedures or equipment or e.
reviews of tests and experiments to comply with 10 CFR 50.59 f.
Radioactive shipments g.
Test results in units of microcuries for leak tests performed pursuant to Specification 3.8.D 1
i 6.0-28 i
BFN Amendment No. 109 Unit 3
R s
h.'
Record of annual physical inventory verifying accountability of sources on record
- i. Gaseous and liquid radioactive waste released to the environs J.
Offsite environmental monitoring surveys k.
Fuel inventories and transfers 1.
Plant radiation and contamination surveys m.
Radiation exposures for all plant personnel n.
Updated, corrected, and as-built drawings of the plant o.
Reactor coolant system inservice inspection p.
Minutes of meetings of the NSRB g.
Design fatigue usage evaluation Monitoring and recording requirements below will be met for various portions of the reactor coolant pressure boundary (RCPB) for which detailed fatigue usage evaluation per the ASME Boiler and Pressure Vessel Code Section III was performed for the conditions defined in the design specification.
In this plant, the applicable codes require fatigue usage evaluation for the reactor pressure vessel only. The locations to be monitored shall be:
1.
The feedwater nozzles 2.
The shell at or near the waterline 3.
The flange studs i
i i
6.0-29 Amendment No. 109
'BFN Unit 3
~
Transients that occur during plant operations will be reviewed and a cumulative fatigue usage factor determined.
For transients which are more severe than the transients evaluated in the stress report, code fatigue usage calculations will be made and tabulated separately.
In the annual operating report, the fatigue usage factor determined for the transients defined above shall be added and a cumulative fatigue usage factor to date shall be reported. When the cumulative usage factor reaches a value of 1.0, an inservice inspection shall be included for the specific location at the next scheduled inspection (3-1/3-year interval) period and 3-1/3-year intervals thereafter, and a subsequent evaluation performed in accordance with the rules of ASME Section XI Code if any flaw indications are detected. The results of the evaluation shall be submitted in a Special Report for review by the Commission.
6.10.2 Except where covered by applicable regulations, items a through h above shall be retained for a period of at least 5 years and item i through g shall be retained for the life of the plant. A complete inventory of radioactive materials in possession shall be maintained current at all times.
1.
See paragraph N-415.2, ASME Section III, 1965 Edition.
6.0-30 Amendment No. 109 BFN Unit 3
6.11 PROCESS CONTROL PROGRAM (PCP)
- 1. The PCP shall be approved by the Commission prior to implementation.
- 2. Changes to the PCP shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
a.
Sufficiently detailed information to totally support the change.
b.
A determination that the change did not change the overall conformance of the solidified product to existing criteria.
6.12 0FFSITE DOSE CALCULATIONAL KANUAL (ODCM)
- 1. The ODCM shall be approved by the Commission prior to implementation.
- 2. Changes to the ODCM shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
a.
$2fficiently detailed information to totally support the change.
6.0-31 BFN Unit 3 Amendment No. 103, 109
Ly y
[
8 '
VIa t
n.
l@
.g.
6.23'.
RADIOLOGICAL EFFLUENT MANUAL (REN)
H J'
1,The REM shall be approved by the Commission prict to-C implementation.
- d
.n
k
- 3. Changes to the REM shall be approved by the Commission prior'to b
implementation, p
Z b,
.4 e
g't s
?
.. { Q f ' '
g.
?
,f {> "l Y e '
lf
)
i
,,P
>[)
'O.'
I l
l s
1 i
1 1
1 6.0-32 3
BFN Unit 3 Amendment No. 203, 109
[
m i1 ji'
,ll l4 i
,A m
n YT EC n
F*
A o S3 rer W t
tor I
n, EN r
g A
o L o
CU*
p N
p u
S l
a R RG c
O AN i
T EI n
CYLN I
EOCI h
A RI UR c
D NT eT ca t
n R RS e
O AE n
T EC CY r
LI I
V e
EOCR R UE g
ID NS a
n a
M y
ME G
t i
G R RIN l
_O-A r
N O AR i
A T EE c
CY F
oM EOCN LE a
Y U
I UI F
T R
G P
D N N D
E E
r o
f n
o 1
i r_ -
i t
R R r
O A a
AS T E z
R CYLA i
A~
EOCQ n
E R U U~
I N
a t.C D
g N
r O
e t
N i
O s
R RT f
I O AC f
TCFEU O
L EOCR l
T R US 1
ID HNO C
2 8
er u
R g
-W"-
.A
~
iF w"
tr*
e o.WW Fnak3" z? w8 mz& M g
i bo~~
o o
i G R N
r"us-seNE n
t N
3G e
EJ N A L
s c
EN s
u CA 6
UM a
s u
N M
G A
. N R
I G
R O
E RE N
E P S s
l O
N N
n" o
I R I
NE n
_ G OC iOi I
AG T
e N
N r
uL U
C A I
c T
O RY l 7,E nJ E
FN L
D E
C N
ERE n"
s I A A*
E TOS i
N D
I M
T.
OTN o
M R A E c
M I
TT R
P R C f
SE 7E E
P Y
EP U UF U
T 1
R N S T
FR O N
I l
IS T
O A E
R SR A
T T TG M
OCR E
F A P E
EE S NA CG I AN H
E V A SLA L
C Y R O I
T A
R N SPM.
I R R C
U A
~
IG O t5NC L
BO O E A R
SM F
I T s
O R NEA O
Lr L T e
T O N OSE o
I P R TC[
DO S-P AC E
E-R RO R~
~
RO I
T
)
N E
O E
cD R
HN S
E L
(
A CE fE l -
NT RG I
EN R
T
)
I N E
O EI 3
L R A
H
)N P
(
C C E EE MU T
S OIG Ek JN S
R N EU E
S N
OIG J
T (ER RNI R N I
G P E RF T
I S R GT RO Q R
I
(
.I MS O
NN I
r EAP V) w TO F A
)
TS NER (3 l N
3 FIT I S sc HS O
E TN R G
N Q O U A
EI EJ.
N S
V DI U M
D N
%=
& w*
f=f g g' h 5u I