ML20036A611
| ML20036A611 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/05/1993 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20036A612 | List: |
| References | |
| NUDOCS 9305120248 | |
| Download: ML20036A611 (7) | |
Text
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[
t UNITED STATES
~
NUCLEAR REGULATORY COMMISSION o
g WASMNGTON, D. C. 20555
%...../
i TENNESSEE VALLEY AUTHORITY j
DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 213 License No. DPR-52 The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated December 23, 1992 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.
930St20248 9305d5 ADOCK05 GOGO PDR P
. 2.
Accordingly, the license is amended by changes to the Technical 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 and B, as revised through Amendment No. 2i3, 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 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Q
b
^
FrederickJ.Hebd/a, Director Project Dirt.ct orate 11-4 Division of keactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 5, 1993
V n
P-ATTACHMENT TO LICENSE AMENDMENT'NO.p n FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 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
- and spillover** pages are provided to maintain document completeness.
. REMOVE JNSERT 3.2/4.2-11a 3.2/4.2-11a 3.2/4.2-42 3.2/4.2-42**
3.2/4.2-43 3.2/4.2-43 3.7/4.7-35 3.7/4.7-35 1
TABLE 3.2.A (Continued)
PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION C to Minimum No.
E$
Instrument Channals Operable w
Per Trio Svsfi)(11)
Function Trio Level Settino Action (1)
Remarks 2
Instrument Channel 1 201.0'F C
Above Trip Setting Reactor Water Cleanup initiates Isolation of System Main Steam Reactor Water Cleanup valve Vault Lines to and from the (TIS-069-834A-D)
Reactor 2
Instrument Channel i 135.0'F C
Above Trip Setting Reactor Water Cleanup initiates Isolation of System Pipe Trench Reactor Water Cleanup
( TI S-069-8354-0)
Lines to and from the Reactor 2
Instrument Channel i 152.0'F C
Above Trip Setting Reactor Water Cleanup initiates Isolation of System Pump Room 2A Reactor Water Cleanup F
(TIS-069-836A-0)
Lines to and from the y
Reactor I
2 Instrument Channel i 152.0'F C
Above Trip Setting y
Reactor Water Cleanup initiates Isolation of System Pump Room 28 Reactor Water Cleanup 7
(TIS-069-837A-D)
Lines to and from the Reactor 2
Instrument Channel i 143.0*F C
Above Trip Setting Reactor Water Cleanup initiates Isolation of' g
System Heat Exchanger R6 actor Water Cleanup e
Room (TIS-069-838A-D)
Lines to and from the Reactor
- J 2
Instrument Channel i 170.0'F C
Above Trip Setting e
Reactor Vater Cleanup initiates Isolation of System Heat Exchanger Reactor Water Cleanup g
Room (TIS-069-839A-D)
Lines to and from the
'd Reactor
c: m TABLE 4.2.A (Cont'd)
Q]
SURVEILL M r REQUIREMENTS FOR PRIt%RY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION function functional Test Calibration Frecuency Ln.strument Check Group 1 (Initiating) Logic Checked during channel N/A N/A functional test. No further test required.(ll)
Group 1 (Actuation) Logic Once/ operating cycle (21)
N/A N/A Group 2 (Initiating) Logic Checked during channel N/A N/A functional test. No further test required.
Group 2 (RHR Isolation-Actuation)
Once/ operating cycle (21)
N/A N/A Logic Group 8 (Tip-Actuation) Logic Once/ operating cycle (21)
N/A N/A Group 2 (Drywell Sump Drains-Once/ operating cycle (21)
N/A N/A y
Actuation) Logic Z
Group 2 (Reactor Building and Once/ operating cycle (21)
N/A N/A Refueling floor, and Drywell
,y Vent and Purge-Actuation) Logic b
Group 3 (Initiating) Logic Checked during channel N/A N/A functional test. No further test required.
Group 3 (Actuation) Logic Once/ operating cycle (21)
N/A N/A Group 6 Logic Once/ operating cycle N/A N/A (18) g Group 8 (Initiating) Legic Checked during channel N/A N/A S
functional test. No g
further test required.
r+
Reactor Building Isolation Once/6 months (18)
(6)
N/A g
(refueling floor) Logic Reactor Building Isolation Once/6 months (18)
(6)
N/A (reactor zone) Logic
P TABLE 4.2.A (Cont'd) ew SURVEILLANCE REQUIREMENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION p$
Function fynttional Test Calibration Frecuency Instrument Check N
SGTS Train A Logic Once/6 months (19)
N/A N/A SGTS Train B Logic Once/6 months (19)
N/A N/A SGTS Train C Logic Once/6 months (19)
N/A N/A Instrument Channel -
(1)(27) 4 months N/A Reactor Water Cleanup System Main Steam Valve Vault (T15-069-834A-0)
Instrument Char.nel -
(1)(27) 4 months N/A Reactor Water Cleanup System Pipe Trench (TIS-069-835A-D) w Instr'.nnent Channel -
(1)(27) 4 months N/A
- w Reactor Water Cleanup System Pump Room 2A N*
(TIS-069-836A-0)
Instrument Channel (1)(27) 4 months N/A w
Reactor Water Cleanup System Pump Room 28 (TIS-069-837A-0)
Instrument Channel (1)(27) 4 months N/A 3=
Reactor Water Cleanup System Heat Exchanger 3
Room S
(TIS-069-838A-0) o S
Instrument Channel -
(1)(27) 4 months N/A Reactor Water Cleanup 3
System Heat Exchanger W
Room (TIS-069-839A-0) w w
, * - + -
3.7/4.7 EASES (Cont'd) in the system, isolation is provided by high temperature in the cleanup system area. Also, since the vessel could potentially be drained through the cleanup ]
system, a low-level isolation is provided.
Grouac 4 and 5 - Process lines are designed to remain OPERABLE and mitigate the consequcoces of an accident which results in the isolation of other process lines. The signals which initiate isolation of Groups 4 and 5 process lines are therefore indicative of a condition which would render them inoperable.
Group 6 - Lines are connected to the primary containment but not directly to the reactor vessel. These valves are isolated on resctor low water level (538"), high drywell pressure, or reactor building ventilation high radiation which would indicate a possible accident and necessitate primary containment isolation.
Group 7 - (Deleted)
Group 8 - Line (traveling in-core probe) is isolated on high drywell pressure or reactor low water level (538").
This is to assure that this line does not provide a leakage path when containment pressure or reactor water level indicates a possible accident condition.
The maximum closure time for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.
In satisfying this design intent, an additional margin has been included in specifying maximum closure times. This margin permits identification of degraded valve performance prior to exceeding the design closure times.
In order to assure that the doses that may result from a steam line break do not exceed the 10 CFR 100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate that fuel rod cladding perforations would be avoided for main steam valve closure times, including instru=ent delay, as long as 10.5 seconds.
e BFN 3.7/4.7-35 Amendment 213 Unit 2