ML20246D509
| ML20246D509 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/22/1989 |
| From: | Black S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246D512 | List: |
| References | |
| NUDOCS 8908280151 | |
| Download: ML20246D509 (13) | |
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION i
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p TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
171 License No. OPR-52 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated February 17 and June 20, 1989, 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 assurante (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 comon 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.
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- 2.
Accordingly. the ' license'isl amended by changes to the' Technical
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Specifications as indicated in the attachment-to this license amendment and paragraph 2.C.(2) of Facility Operating License No.. DPR-52 is hereby amended to read as follows:
_ 2); Technical Specifications
(
The Technical Specifications contained in' Appendices A andLB, as revised through Amendnent No.171, 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 ard shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION o
h WC Suzanne Black. Assistant' Director for Projects TVA Projects. Division Office of Nuclear Reactor Regulation-
Attachment:
Changes to the Technical Specifications Date of Issuance: August 22, 1989
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ATTACHMENT TO LICENSE AMENDMENT NO. 171 FACILITY OPERATING LICENSE NO. DPR-52 l
DOCKET NO. 50-260 Revise the Appendix. A Technical Specifications by removing the pages-
, identified ~below and; inserting the enclosed pages. The revised pages l
are' identified by the captioned amendment number and contain marginal l
Ifnes indicating the area of change. Overleaf
- pages are provided for document integrity.
REMOVE INSERT 3.2/4.2-31 3.2/4.2-31*
3.2/4.2-32 3.2/4.2-32 3.2/4.2-33 3.2/4.2-33 3.2/4.2-33a 3.2/4.2-54 3.2/4.2-54
- 3.2/4.2-55 3.2/4.2-55 6.0-27 6.0-27*
6.0-28 6.0-28*
6.0-29 6.0-29 6.0-30 6.0-30*
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i NdTFS POR Tintt 3.2.F (1) From and af ter the date that one of these parameters is reduced to cne indication, continued operation is permissible during the succeeding 30 days unless such instrumentation is sooner made OPERABLE.
(2) Trem and after the date that one of these parameters is not indicated in the control room, continued operation is permissible during the succeeding seven days unless such instrumentation is sooner made OPERABLE.
(3) If the requirements of notes (1) and (2) cannot be met, and if one of the indications cannot be restored in (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(4)
These surveillance instruments are considered to be redundant to eadh other.
(5) From and after the date that both the acoustic monitor and the temperature indication on any one valve fails to indicate in the control room, continued operation is permissible during the succeeding 30 days, unless one of the two monitoring channels is sooner made OPERABLE.
If both the primary and secondary indication on any SRV tailpipe is inoperable, the torus temperature will be monitored at least once per shift to observe any unarplained temperature increase which might be indicative of an open SRV.
(6) A channel consists of eight sensors, one from each alternating torus bay.
Seven sensors must be OPERABLE for the channel to be OPERABLE.
(7) When one of these instruments is inoperable for more than seven days, in lieu of any other report required by Specification 6.9.1.4, l
prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next seven days outlining the action l
taken, the cause of inoperability, and the plans and schedule for restoring the system to OPERABLE status.
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(8) With the plant in REACTOR POWER OPERATION, STARTUP CONDITION, HOT STANDBY CONDITION OR HOT SHUTDOWN CONDITION and with the number of OPERABLE channels less than the required OPERABLE channels, either restore the inoperable channel (s) to OPERABLE status within 72 1
hours, or Laitiate the preplanned alternate method of monitoring the appropriate parameter.
(9) Noble Gas only BFN Unit 2 3.2/4.2-33 Amendment No. 154,171
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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.
Istigue 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, t
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 BFN 6.0-27 Amendment No. 145' Unit 2
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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.
Diesel Generator Reliability. Improvement Program Report shall be submitted within 30 days of meeting failure criteria'in~ Table 4.9.A.; As a minimum, the l
reliability Improvement Prograt report for NRC audit shall include:
A summary of all tests (valid and invalid) that a.
occurred within the time period over which the last 20/100 valid tests were performed.
b.
Analysis of failures and determination of root causes of failures.
c.
Evaluation of each of the recommendations of NUREG/CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability in Operating Reactors," with respect to their application to the plant.
d.
Identification of all actions taken or to be taken to (1) Correct the root causes of failures defined in b above and-(2) Achieve a general improvement of diesel' generator reliability, A supplemental report shall be prepared for an e.
NRC audit within 30 days after each subsequent failure during a valid demand, for so long as the affected diesel generator unit continues to violate the criteria (3/20 or 6/100) for the reliability improvement program remedial action.. The supplemental report need only update the failure / demand history for the affected diesel generator unit since the last report for that diesel generator..The supplemental report shall also present an analysis of the failure (s) with a root cause determination, if possible, and shall delineate any further procedural, hardware or operational changes to be incorporated into the site diesel I
generator improvement program and the schedule for implementation of those' changes.
)
BFN 6.0-28 Unit 2 Amendment No. 149
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8.
Secondary Containment 4.7.C.
Leak Rate Testing
- Within 90 days of completion of each test.
9.
High-Range Primary 3.2.T Containment Radiation Within 7 days Monitors after 7 days of inoperability.
- 10. Wide-Range Gaseous Effluent 3.2.F Within 7 days Radiation Monitor and recorder after 7 days of 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
.the test, concurrent reactor building pressure, and amergency ventilation flow rate.
The report shall also include analyses and interpretations of those data which demonstrate compliance with the specified leak rate limits.
6.10 STATION OpenATING RECDRDS 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 activisfes c.
Reportable Events 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 4
performed pursuant to Specification 3.8.D BrN 6.0-29 Amendment No. 128, 134,145, 149, 171 Unit 2
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4 h.
Record of annual physical inventory verifying accountabil,ity of sources on record 1.
Gaseous and liquid radioactive waste released to the environs 4
- j. Offsite environmental monitoring surveys k.
Fuel inventories and transfers 1.
Plant radiation and contamination surveys i
Radiation exposures for all plant personnel m.
Updated, corrected, and as-built drawings of the plant n.
Reactor coolant system inservice inspection o.
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Minutes of meetings of the NSRB j
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 1
ASME Boiler and Pressure Vessel Code Section III was performed for the conditions defited 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 feedvater nozzles 2.
The shall at or near the waterline 3.
The flange studs BFN 6.0-30 Amendment No. 128, 134, 145 Unit 2
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _