ML18283B749

Technical Specifications Changes to Decrease the Main Stream Line Isolation Pressure Setpoint. Jan. 26, 1977 as Supplemented Feb. 7, 1977

ML18283B749
Person / Time
Site:	Browns Ferry
Issue date:	02/24/1977
From:	Schwencer A Office of Nuclear Reactor Regulation
To:	Gerald Williams Tennessee Valley Authority
References
Download: ML18283B749 (31)

Text

DISTRIBUTION Dockets(3) ACRS{16)

NRC PDR{3) OPA(CHiles)

Local PDR DRoss Oockets Hos. 50>>25 260 296 February 24 l g77 ORB81 Reading BHarless VStel 1 o TBAbernathy Y>RGoller /TJCarter JRBuchanan SHSheppard Rzageml.T@ambach Tennessee Valley Authority OF LD ATTH; Pfr. God~in Nlliams. Jr. Oras{5)

Hpnager of Power BJones('12)

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Chattanooga, Tennessee 3720'1 JÃcGough ZKDXXNKN Qer itl QITIen DEi senhut The ComiSsion has issued the enclosed Amndmants fios. 29, 26 and,4 to Facility Licenses Hors. DPR<<33, DPR-52 and DPR-68 for the Gro.-es Ferry Nuclear Plant', Units 1. 2 and 3, respectively. These a~>>ndr>> nts consist of changes to the Technical SpecificaNons in response to your request of January 86, 1977 as supplenented February', l977.

These amendnants change the Technical Specifications to decrease the rmin steato line isolation pressure setpoint. You Mere previously notified of these changes by telephone and by tolecopy on February 7, 1977.

rr Copies of the Safety Evaluation and the Federal Register Hotice are also enclosed.

r Sincerely, r Or]ginal signed by A. Schwencer, Chief Operating Reactors Branch 01 Division of Operatinq Reactors Enclosures'a>>endrent No. 2g to DPR-33

2. Aavndmnt Ho. 26 to GPR-62

3. Ar~ndrrant Ho. 4 to GPR-68 4< Safety Evaluation 5, Federal Register Hotice cc M/enclosures:

See next page

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Tennessee Val 1 ey Authority February 24 1977 cc: H. S. Sanger, Jr., Esquire General Counsel Tennessee Valley Authority 400 Commerce Avenue E llB 33 C Knoxville, Tennessee 37902 Mr. D. McCloud Tennessee Valley Authority 303 Power Building Chattanooga, Tennessee 37401 Mr. William E. Garner Route 4, Box 354 Scottsboro, Alabama 35768 Athens Public Library South and Forrest Athens, Alabama - 35611 Mr. Charles R. Christopher Chairman, 'Lim'estone County Commission Post Office Box 188 Athens, Alabama 35611 Ira L. Myers, M.D.

State Health Officer State Department of Public Health State Office Building Montgomery, Alabama 36104 Mr. C. S. Ilalker Tennessee Valley Authority 400 Commerce Avenue M 9D199 C Knoxville, Tennessee 37902

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4**~4 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 29 License No. DPR-33

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Tennessee Valley Authority .

(the licensee) dated January 26, 197$ , as supplemented February 7, 1977, complies with tIIe standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission' 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.

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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 License No. DPR-33 is hereby amended to read as follows:

(2) Technical S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 29, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective February 7, 1977.

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Attachment:

Changes to the Technical Specifications Date of Issuance: February 24, 1977

ATTACHNENT TO LICENSE ANENDNENTS ANENDNENT NO, 29 TO FACILITY LICENSE NO, DPR-33 AMENDNENT NO 26 TO FACILITY LICENSE NO, DPR-52 ANENDNENT NO. "TO FACILITY LICENSE NO. DPR-68 DOCKETS NOS. 50-259, 50-260 5 50-296 Revise Appendix A of Units 1 & 2 as follows:

Remove pages 10, 24, 55 and 112 and replace with identically numbered pages.

Revise Appendix A of Unit 3 as follows.

Remove pages 13, 23, '57 and 110 and replace with identically numbered pages.

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t 2.1 BASES

2. Scram on loss of 'control oil pxessux'e The turbine hydraulic control system operates using high pressure oil. There are several points in this oil system where a loss of control oil pressure could result in a fast closure of the tu.bine valves. This fast closure of the turbine control valves is not px'otected by the generatox'oad re5ection scram, since failure of the oil system ~ould not result in the fast closure solenoid valves being actuated. Pox a turbine control valve fast closure, the core should be protected by the APRM and high reactor pressuze scx'ams. However, to provide the sane margins as provided

~ for the generator load re5ection sera~ on fast closure of the turbine control valves, a scxam has been added to the reactor protection system, which senses failuxe of control oil pressure to the tur-bine contx'ol system. This is an anticipatory scram and results in reactor shutdown before any significant increase in pressure oz neutx'on flux occurs. The transient response is very similar to that resulting from the generator load re5ection.

P. Main Condenser Low Vacuum Scram To protect the main condenser against overpxessuxe, a loss of con-denser vacuum initiates automatic closure o. the turbine stop valves and turbine bypass valves. To anticipate the transient and automatic scram resulting fran the closure of the turbine stop valves, low con" denser vacuum initiates a'cram. The low vacuum scx'am set point is selected to initiate a scram befc,.'e the closure of the turbine stop valves is in G. Ec H. Hain Steam Line Is~ ation on Low Pxessure and Main Steam Line Isolation Scram The low pressure isolation of the main steam lines at 825 psig vas provided to protect against rapid reactor depressurixation and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature that occuxs when the. main steam line isolation valves are closed, to provide for reactor shutdown so that high power opera-tion at low reactor pressure does not occur, thus providing protec'tion,.

for the fuel cladding integrity safety limit. Operation of the reac-tor "at pressures lover than 825 psig requires that the reactor mode switch be in the STARTUP posit'on, where protect'on of the fuel cladding integri y safety limit is provided by the IBM and APPA high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and.isolation valve closuza scram assures the availability of neutron flux scram protection ovex'he entire range of applicab'1'ty of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closuxe. Mith the sc'rams set at 10 percent of valve closure, neutron flux does not inczease.

Amendments Hos. 29, 26 8 0

TABLE 3.2.A PRIMARY CONTAINHEHT AND REACTOR BVILDIHC ISOLATION INSTRUHENTAYfON Mfnfaua No.

Operable Pcr Tri S s l Punction Trf Level Settfn Action I Remarks Instrument Channel- > 538" above vessel zero h or l. Below trfp setting does the Reactor Low Pater Level (6) (B and-E)

-following:

Initfates Reactor Building Isolation

b. Initiates Prfcaary Contahuaent Isola t ion c Initiates SGTS Instruaent Channel- l00 + 15 psig I. Above trfp setting faolatea the Bcactor High Pressure shutdovn cooling suction valves of the RHR system.

Instrument Channel > 49Q".above,vessel. zero. 1. :Bclcnr trip -.setting initiates Main Reactor I~~lIater Level Stcaa Linc .Isolation

'(LIS-3-'56h-'D, SM 'fl)

- Instrument Cfmnnel- < 2 psig h or 1, Above trfp setting does High DryMcll. Prcssure (6) (B and E) the'ollmrfng:

(PS-64-56A-D) a. Inftfatee Reactor Building Isolation

b. Initiatca Prfaary Containment Isolation O c Inftiaten SCTS CA Instrument Channel < g times normaL rated. 1 Above trip setting inftfates Main High Radiation Main Steam full power background Steam Line Isolation, Line Tunnel (6)

Inntrument Channcl- > 825 psfg (4) B 1. BeloM trip setting initiates Main Lov Prcsaurc Hain Steam Stcam Lfne Isolation Linc 2 (3) Instrument Channel- 140X of rated steam floor 1. Above trip setting initiates Hain llfgh Ylov Hain Stcam Line ~

Steam Linc Isolation

3..2 BASES LPCI loop selection logic and trips the recirculation pumps. The water level instrumentation that ia set to trip vhan reactor lov'eactor vatar level is 17.7" (378" above vassal ero) above the top of the active Eual (Table 3.2.8) initiates the LPCI, Cora Spray Pumps, contributes to AOS initiation and starts the diesel generators. These trip setting levels vere chosen to be high enough to prevent spurious actuation but lov enough to initiate CSCS operation so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 vill not be violated.

For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to neet the above criteria.

The high dryvell pressure instrumentation is a diverse signal to the

.uatcr level instrunentation and in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, this instrumentation vill initiate CSCS operation at about the same time as the lov water level instrumentation;. thus the results given-above are appl.icable herc also.

Vanturis are provided in the main steam lines as a means of measuring stcam Elov and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instru-mentation is to detect a break 'n the main stcam line. For the vorst case accident, main'team line brcak outside the dryvell, a trip setting of 140'f rated steam flow in con)unction with the Elov limi.ters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain belov 1000'F snd release of radioactivity to the enviions is vali below 10 CFR 100 guidelines. Rc Ear ence Section 14. 6. 5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line

, tunnel to detect leaks in these areas, Trips are provided on this instru-mentation and when exceeded, cause closure of isolation valves. The setting of 200'F for the main steam linc tunnel detector is lov enough to detect leaks of thc order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks. For large branks, the high stean flov instru; mentation is a backup to the temperature inatrunantation.

High radiation monitors in the main steam line'tunnel have been provided to detect gross fuel failure as in the control rod drop acc'dent. Mith *'ha established setting of 3 times normal background, and main steam line isolation valve closure, E'ssion product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident. Reference Secti~n 14.6.2 FSAR. n alam, '>th a.ncm mal sat point o" ".5 x nnaf fully. jo::er o" ck~."outed, 's prcrided aYso, Pressure instrumentat1on is provided to close the main steam isolation valves in Run Node vh n the maig steam line pressure drops belov 825 psig o Amendments Nos. 29, 26 8 4

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY D. Shutdown Condit1on Ci Scram and isola- 2 538 in tion reactor above Whenever the reactor is in low water vessel the shutdown condition level zero with irradiated fuel in the reactor vessel, the D Scram--turbine 5 10 per-water level shall not be stop valve cent. valve less than 17.7 in. above closure closure the top of the normal active fuel zone. E Scram--turbine control valve

1. Fast closure Upon trip ofsolenoid the fast acting valves

2. Loss of con- 2 550 psig trol oil pressure F.. Scram low con- 2 23 inches denser vacuum Hg vacuum G Scram--main 5 10 per-steam line cent. valve isolation closure H Main steam isola- 5 825 psig tion valve closure

--nuclear system low pressure Core spray and h 378 in.

LPCI actuation above

'reactor low 'water vessels level zero HPCI and RCIC 2'90 in.

actuation--reac- '-above tor low water vessel level zero K, Main steam .isola- I above 490 in.

tion valve closure--reactor vessel low water level zero Amendments Nos. 29, 26 & 4

setting and the fact that control valve closure time is approximately twice as long as that for the stop valves means that resulting transients, while similar, are less severe than for stop-valve closure. No fuel damage occurs, and reactor system pressure does not exceed the relief valve set point, wh'h is approximately 280 psi below the safety limit.

2. Scram on loss of control oil pressure The turbine hydraulic control system operates using high pressure oil. There are several points in this oil system where a loss of oil pressure could result in a fast closure of the turbine control valves. This fast closure of the turbine control valves is not protected by the generator load rejection scram, since failure of the oil system would not result in the fast closure solenoid valves being actuated. For a turbine control valve fast closure, the core would be protected by the APRM and high reactor pressure scrams. However, to provide the same margins as provided for the generator load rejection scram on fast closure of the turbine control valves, a scram has been added to the system, which senses failure of control oil reactor'rotection pressure to the turbine control system. This is an anticipatory scram and results in reactor shutdown before any significant increase in pressure or neutron flux occurs. The transient response is very similar to that resulting from the generator load rejection.

F. Main Condenser Low Vacuum Scram To protect the main concenser against ovexpressure, a loss of condenser vacuum initiates automatic closure of the turbine stop valves and turbine bypass valves. To anticipate the transient and automatic scram resulting from the closure of the turbine stop valves, low condenser vacuum initiates a scram. The low vaccum scram set point is selected to initiate a scram before the closure of the turbine stop valves is initiated.

G. 6 H. Main Steam Line Isolation on Low Pressure and Main Steam Line Isolation Scram The low pressure isolation of the main steam lines at 825 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature that occurs when the main steam line isolation valves are closed, to.

provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit.

Operation of the reactor at pressures lower than 825 psig requires that the reactor mode switch he in the STARTUP Amendments Nos, 29, 26 II 4

TABLE 3 2 A PRIMARY CONTAINMENT AND REACTOR BUILDING ISOZATION INSTRUMENTATION Minimum No.

Operable Per Tri S s 1 Function Tri Level Settin Action 1 Remarks Instrument Channel-  ? 538" above vessel zero A or 1. Below trip setting does the Reactor Low Water Level (6) (B and E) following:

a. Initiates Reactor Building Isolation

b. Initiates Primary Containment Isolation

c. Instigates SGTS Instrument Channel- 100 + 15 psig 1. Above trip setting isolates the Reactor High Pressure shutdown cooling suction valves of the RHR system.

Instrument Channel-  ? 490" above vessel zero 1. Below trip setting initiates Main Reactor Low Water Level Steam Line Isolation (LIS-3-56A-Di SW 41)

Instrument Channel- S 2 psig A or. l. Above trip setting does the High Drywell Pressure (6) (B and E) following:

(PS-64-56A-D) a. Initiates Reactor Building Isolation

b. Initiates Pre~i, y Containment Isolation

c. Initiates SGTS

0) Instrument Channel- 5 3 times normal rated l. Above trip setting initiates Main High Radiation Main Steam full power background Steam Line Isolation Line Tunnel (6) 2 Instrument Channel-  ? 825 psig (4) 1. Below trip setting initiates Main Low Pressure Main Steam Steam Line Isolation Line 0,

O 2(3) Instrument Channel-High Plow Main Steam Line 140% of rated steam flow B 1. Above Steam trip setting initiates Line Isolation Main 2 Instrument Channel- 5 ?00oF 1. Above trip setting initiates Main Steam Line Tunnel Main Steam Line Isolation.

High Temperature

Pressure instrumentation is provided to close the main steam isolation valves in Run Mode when the main steam line pressure drops below 825 psig.

The HpCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves.

Tripping logic for the high flow is a 1 out of 2 logic, and all

'sensors are required to be operable.

High temperature in the vicinity of the HPCI equipment is sensed by 4 sets of 4 bimetallic temperature switches. The 16 temperature switches are arranged in 2 trip systems with 8 temperature switches in each trip system.

The HPCX trip settings of 90 psi for high flow and 200oF for high temperature are such that core uncovery is prevented and fission product release is within limits.

The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 450" water for high flow and 200~F for temperature are based on the same criteria as the HPCI.

High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system. When high temperature occurs, the cleanup system is. isolated.

The instrumentation which initiates CSCS action is arranged zn a dual bus system. As for other vital instrumentation arranged in this fashion, the Specification preserves the effectiveness of the system even during periods when maintenance or testing performed. An exception to this is when logic functional is'eing testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to 1.05.

The trip logic for this function is 1 out of n: e.g., any trip-on one of six APRM's, eight IRM's,'r four SRM's will result in a rod block.

The minimum instrument channel requirements assure sufficient rm in strumentation to assure the single failure criteria is met.

The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration. This time period is only 3% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.

The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation at reduced flow. The APRM provides gross core protection; i.e.,

limits the gross core power increase from withdrawal of control 110 Amendments Nos. 29, 26 & 4

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TENNESSEE'VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 26 License No. DPR-52

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Tennessee Valley Authority (the licensee) dated January 26, 1977, as supplemented February 7, 1977, 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 2.C(2) of Facility License No. DPR-52 is hereby amended to read as follows:

(2) Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Ame'ndrpent No.26 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective February 7, 1977.

FOR THE NUCLEAR REGULATORY COMMISSION I I A. Sc wencer, Chief Operating Reactors Branch Pl Division, of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: February 24, 1977

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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 4 License No. DPR-68

l. The Nuclear Regulatory Commission (the Commission) has found that:I A. The application for amendment by Tennessee Valley Authority (the licensee) dated January 26, 1977, as supplemented February 7, 1977, 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, atld (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.

1

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 Faci)ity License No. DPR-68 is hereby amended to read as follows:

(2) Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 4 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective february 7, 1977.

FOR THE NUCLEAR REGULATORY COMMISSION A. Schwencer, Chief Operating reactors Branch Pl Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: February 24, 1977

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WASHINGTON, D. C. 20555 IP 4~*~4 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 29 TO FACILITY OPERATING LICENSE NO. DPR-33 AMENDMENT NO, 26 TO FACILITY OPERATING LICENSE NO. DPR-52 AMENDMENT NO. 4 TO FACILITY OPERATING LICENSE NO. DPR-'68 TENNESSEE VALLEY AUTHORITY k

BROWNS FERRY NUCLEAR GENERATING PLANT, UNITS 1 2 & 3 Introduction By letter dated January 26, 1977, as supplemented by letter dated February 7, 1977, the Tennessee Valley Authority (TVA) proposed changes to the Technical Specifications appended to Facility Operating Licenses Nos. DPR-33, DPR-52 and DPR-68. The proposed changes involve a reduction in the main steam line low pressure isolation setpoint and are the same as changes previously approved by the Nuclear Regulatory Commission (NRC) for other nuclear plants, (Hatch 1, Brunswick 2, Monticello and Oyster Creek).

Discussion and Evaluation Main Steam Line Pressure Isolation Set Point Reduction Installation of the main steam line low pressure sensors was required to provide reactor isolation in the event of an abnormal transient associated with the failure of the initial turbine pressure regulator in the open direction. This reactor isolation function was provided to limit the durati n and severity of system depressurization so that no significant the al stresses are imposed on the primary system.

No credit was taken for these low pressure sensors in any of the other postulated abnormal operating transients or accidents. The current isolation set point is 850 psig; the proposed setpoint is 825 psig.

TVA referenced Edwi I. Hatch Nuclear Plant Unit 1 (50-321) submittal ~

dated October 9, 19 I5 which provided a bounding analysis for a reduction in the main steam line low pressure setpoint from 880 psig to 825 psig.

The NRC staff has reviewed the Hatch I analysis and has determined that it is applicable to TVA's proposed changes. In both cases (Hatch and Browns Ferry) the additional temperature decrease and subsequent reactor vessel thermal stresses, resulting from the additional pressure reduction during the abnormal transient, are negligible. Because reduction of the low pressure isolation setpoint would not have significant effects on

previously analyzed 'transients, we have concluded that the proposed change is acceptable, Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

Conclusion We have concluded, based on the considerations discussed above, that>

( 1) because the amendments do not involve a singificant increase in the probabi lity or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by oper ation in the proposed manner, and (3) such activities wi'll be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public, Date: February 24, 1977

UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKETS NOS. 50-259 50-260 AND 50-296 TENNESSEE VALLEY AUTHORITY NOTICE OF ISSUANCE OF AHENDNENTS TO FACILITY 0 ERA ING LICENSES The U. S. Nuclear Regulatory Commission {the Commission) has issued Amendments Nos. 29, 26and 4 to Facility. Operating Licenses Nos. DPR-33, DPR-52 and DPR-68, respectively, issued to Tennessee Valley Authority (the licensee), which revised Technical Specifications for operation of the Browns Ferry Nuclear Plant, Units Nos, 1, 2 and 3 (the facility)

/

located in Limestone County, Alabama. The amendments are effective February 7, 1977.

These amendments change the Technical Specifications to decrease the main steam line isolation pressure setpoint.

The application for the amendments complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. ,The Commission has made appropriate.

findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments. Prior public notice of these amendments was not requi red since the amendments do not involve a significant hazards consideration.

The Commission has determined that the issuance of these amendments will not result in any significant environmental impact and that pursuant to 10 CFR 551.5(d)(4) an environmental impact statement or negative

. declaration and environmental impact appraisal need not be prepared in connection with issuance of these amendments.

For further details with respect to this action, see ( 1) the application for amendments dated January 26, 1977 as supplemented February 7, 1977, (2) Amendments Nos, 29, 26 and 4 to Licenses Nos.

DPR-33, DPR-52 and DPR-68, respectively and (3) the Commission's related Safety Evaluation. All of these items are available for public inspection at the Commission's Public Document Room, 1717 H Street, N.W., Washington, D.C. and at the Athens Public Library, South and Forrest, Athens, Alabama 35611. A copy of items (2) and (3) may be obtained upon request addressed .

to the U. S. Nuclear Regulatory'ommission, Washington, D. C. 20555, Attention: Director, Division of Operating Reactors.

Dated at Bethesda, Maryland, this 24th day of February 1977.

FOR THE NUCLEAR REGULATORY COMMISSION A, Schwencer, Chief Operating Reactors Branch fl Division of Operating Reactors

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