Text

Letter Submitting Amendments No. 38, 36 and 12 to Facility Licenses No. DPR-033, DPR-52 and DPR-68 for Browns Ferry Nuclear Plant, Units No. 1, 2 and 3 Consisting of Changes to TSs in Response to 05/11/1978 Request
	ML18283B732
Person / Time
Site:	Browns Ferry
Issue date:	06/22/1978
From:	Ippolito T; Office of Nuclear Reactor Regulation
To:	Niav Hughes; Tennessee Valley Archaeological Research
References
	Download: ML18283B732 (53)
	v • d • e

Distribution

, Pocket and M-VORB y3 Local PDR NRC PDR GLear Tennessee Valley Author2ty ITppolito

'T%:

Hr. N. 8. Hughes SSheppard f<anager of Po>(er RClark 330 Power Bui'tding VStello Chattanooga, Tennessee 37401 BGrimes.

Attorney, GELD Gentlemn:

OraE (5 BJones 12)

The Commission has issued the enclosed Aaandamts Dos. 38, 36 and 12 to Facility Licenses hos.

DPR-33, DPR-62 and DPR-68 for the Brefns Ferry Nuclear Plant, Units Nos. l. 2 and 3.

These amendaeats consist of changes to the Technical Specifications in response to your request of toy 11, 1978 (TN.BFflP TS 108).

Docket tlos. 6 59 BScharf (10)

JNcGough DEisenhut ACRS (16)

OPA (CHiles)

DRoss TBAbernathy JRBuchanan RDiggs Copies of the Safety Evaluation and Notice of Issuance are also enc1osed.

Sincerely, The amendrmts change the Technical Specifications to pere6t you to delete the requirer:.ants for the oxygen sensors as used 4n the con-taintaent atmosphere ren$ toring system.

Nth your concurrence.

ve have mdified your submittal to add additional surve)11ance requirer nts in Section 4.7.A.G.

Enclosures:

Amendment halo.

8S to DPR-33 2.

Jimen@nent tlo. >> to DPR-52 3.

Anendment Ho. IO-to DPR-68 4.

Safety Evaluation 6.

Notice cc v~/enclosu s:

See page 2 Thonas A. Xppolito, Chief Operating Reactors Branch 83 Division of Operating Reactors CVRNAMC~

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Tennessee Valley Authority 2

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H. S. Sanger, Jr.,

Esquire General Counsel Tennessee Valley Authority 400 Commerce Avenue E

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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 Mr. Charles R. Christopher

Chairman, Limestone 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. Walker Tennessee Valley Authority 400 Commerce Avenue W 9D199 C

Knoxville, Tennessee 37902 Athens Public Library South and Forrest

Athens, Alabama 35611 Chief, Energy Systems.

Analyses Branch (AW-459)

Office of Radiation Programs U.S. Environmental Protection Agency Room 645, East Tower 401 M Street, SW Washington, D.C.

20460 U. S. Environmental Protection Agency Region IV Office ATTN:

EIS Coordinator 345 Courtland Street Atlanta, Georgia 30308

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

38 License No.

DPR-33 1.

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

A.

The application for amendments by Tennessee Valley Authority (the licensee) dated Nay ll, 1978, 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 Spec-ifications 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 B, as revised through Amendment No. 38, 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 the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

June 22, 1978 Thomas A.

polito, Chief Operating Reactors Branch 83 Division of Operating Reactors

0 C

I

ATTACHMENT TO LICENSE AMENDMENT NO. 38 FACILITY OPERATING LICENSE NO.

DPR-33 DOCKET NO. 50-259 Revise Appendix A as follows:

1.

Remove the following pages and replace with identically numbered pages:

79/80 234/235 248/249 268/269 270/271 2.

Marginal lines indicate revised area.

Overleaf pages are. provided for convenience.

IC 1

TABLE 3.2.F Surveillance Instrumentation Minimum 4 of Operable Instrument Channels Instrument 4 H2M 37 H H 39 Instrument Drywell H Concentration Type Indication and Ha e

Q.l - 20)

Notes H M 76 - 38 2

Suppression Chamber g Concentration O.l 20X Amendment No.

38

I NOTES FOR TABLF. 3.2.P (1)

From and after the date that one of these parameters is reduced to one indi'cation, continued operation is permissible during the

~ucceeding thirty diys unless such instrumentation ia sooner made operable.

(2)

.From and after the date that one of these parameters is not indi-cited in the control room, continued operation is permissible during the succeeding seven days unless such instrumentation ia sooner made opersblc.

(3) If the requirements of notes (1) snd (2) canno't be met, either the requirements of 3,5.H shal'1 be complied vith or an orderly shutdovn shall be initiated and the reactor shall be in a Cold Condition vithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(4)

These surveillance instruments are considered to be redundant to each other.

80

l.lMlT[HO ~ Cll.'in'1('C FOR OPKl'ITlON SURVLTLI h.'iCF. kFI~'IIi~~',ITS 4.7.A Primary Containment ae Except as spec'fied in

3. 7.A. 3. b below, tvo pressure suppression chamber-resc tor building vacuum breakers shall be operable at all tbues when primary containment inte-trity is required.

The set point of thc d'ffcren-tial pressure instrur euta-tion which actuates the pressur~

supprcssicn cham-ber-reactor building vacuum breakers shall be Oe5 paid.

b.

From and after the date that one of the pressure suppression chamber-reactor building vacuum brenkero is mode or found to be inopera-ble for cny reason, reactor operation is permissible only during the succeeding seven

days, provided that thc repair procedure does not viola t c pr imary contain-ment integrity, 4.

~Dr ell-praeuu~rr Bu rue uinn Chamber Vacuum Breakers 3.

Prcssure Su rcosion Chomber-Beeccar Burldln-uecuun Breakere the reactor shall be placed in cold shutdown and the above inspection shal'e performed before the reactor is started up.

3.

Pressure Su ression Chamber Reactor Buildin Vacuum Breakers a.

The pressure suporession chamber-reactor building vacuum breakers shall be exercised and the associ-ated instrumentation including setpoint shall be functionaily tested for proper operation each three months.

r b.

A visual examination and detennina tion that the force required to open each vacuum breaker (check valve) does not'exceed 0.5 psid vill be made each refueling outage.

4.

Dr e11-Prcssure Suppression Chamber Vacuum Rreakers a.

Fach dryvell-suppression chamber vacuum breaker shall be excrcfscd through an opening-closing cycle eve ry mon th.

a ~

When primary containment is required, all dryvell-ouppresoion chamber vacuum breakero sho) 1 be operable

'nd pooitioned in the fully closed position (except during tooting) except os speci ficd in 3. 7.A. 4. b and c, below.

b.

When it is determined that tvo vacuum b re ak e rs or e inoperable for opening at a

time -when operability is rcquir all other vacuum breaker b.

One dryvell-suppression chamber vacuum breaker may be non-fully closed so long as it ic determined to be not more than 3" open as indicated by the position lights, 234 Amendment No.

38

LIHJTINC Ci.'.

TIOHS FOR OPERATION SURVEILLANCE RE UIREHEHTS

'3. 7. A Primnr Containment

4. 7.A primer Con cairn ent valves shall be exercised iuuaediately and every 15 days thereafter until the inoperable valve hab beer returned to noraal service.

c.

Two drywcl I-suppression chamber vacuum breakers may be determined to be inoperable for opening.

c ~

Once each operating cycle each vacuum breaker valve shall be inspected for proper operation of the valve and limit switches'.

d.

If specifications 3.7.A.4.a,

.b, or,c cannot be met, the unit shall be placed in a cold shutdown condition in an orderly manner within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months '.

A leak test of the drywell to suppression chamber structure shall be con-ducted during each operating cycle.

Accept-able leak rate is 0.14 ib/

sec of primary containment atmosphere Mith 1 psi differential.

5.

0 ea Concentration a.

After completion of the fire-related startup retestin

program, containment atmosphere shall be reduced to less than 4Z oxygen with nitro-gen gas during reactor power operation with reac-tor coolant pressure above 100 psig, except as speci-fied in 3.7.A.5.b.

b, Within the 24-hour period subsequent to placing the reactor in the Run mode following a shu down, thc containment atmosphere oxygen concentration shall be reduced to lees than 4Z by weight and maintained in this condition.

De-inert-ing may commence 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to a shutdown.

~.

The primary containment oxygen concentration shall be measured

'and recorded.daily, The oxygen measurement shall be ad)usted to account for the uncertainty of

'he method used by adding a

predetermined error function.

b.

The methods used to measure the primary containment oxygen con-centration shall be calibrated once every refueling cycle.

c. If spec ifice tion 3.7.A.5.a and 3.7.A.5. b cannot be met, an orderly 235 shutdown shall be initiated and the reactor shall be in a Colil Shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE UIRENENTS 3, 7 Cr)ilTAINLIK."lT S YSTF~lS 4.7 CONTAINMENT SYSTEMS G.

The Containment Atmosphere Dilution (CAD) System shall be cperable with:

a

~

Two independent systems capable of supp).yinq nitrogen to the drywell and torus.

A minimum supply of 2500 gallons of liquid nitrogen, per system.

Conta.'.nment A"mos here Dilution System CAD G.

a ~

At least once per month cycle each solenoid operated air/nitrogen valve through at least one complete cycle of full travel and verify that each manual valve in the flow path is open.

Containment Atmos-here Dxlutron Svs tern CAD 1.

S stem 0 erabilit 3.

The Containment Atmosphere Dilution (CAD) System shall be operable whenever the reactor mode switch is in the "RUN" posi tion.

If one system is inoperable, the reactor may remain in operation for a period of 30 days provid d all active components in the other system are operable.

b.

Verify that the CAD System contains a,

minimum suoply of 2500 gals of.

liquid nitrogen twice per week.

Xf Soeci ication 3.7.6.1 and 3.7 ~ G. 2, o

3.7.G. 3 cannot bo net, an orde"'

shutdc'n shall b.'nitio"ed and the reactor shall be ir. the Cold Shu down cond'ion wxthxn 2" hours.'rimary containment pressure shall be limited to a

maximum of 30 psig during repr essurization following a loss of coolant accident.

248 Amendment No.

38

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREiMNTS 3, 7 CONTAINMEiV1 S YSTFMS 4.7 CONTAINMENT SYSTEMS H.

Containment Atmosoh re Monitorin CAM S stem 2

H Anal zer 1.

Whenever the reactor is not in cold shutdown, two gas analyzer systems shall be operable for monitoring the drywell.

2.

Whenever the reactor is not in cold shutdown, one gas analyzer system shall be operable for monitoring thetorus.

H.

Containment Atmos here Monztorxn CAN S stem Anal zer 2

H Once per month perform a channel calxbrat2.on using standard gas samples containing a nominal:

Three volume percent

hydrogen, balance nitrogen 3.

If specification 3.7. H. 1 cannot be met, but one -system remains operable, the reactor may be operated for a period of 30 days.

If both systems are inoperable, the reactor should be placed in shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4. If specification 3.7.H.2 cannot be met, the reactor may be operated for a period of 30 days.

249

.BASES 3.7.A 5 4. 7. A Primar Containment The integrity of the primary cortainment and operation of'he core standby cooling system in combination, limit the off-site doses to values less than those suggested in 10 CFR 100 in the event of a break in the primary system piping:

Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists.

Concern about such a violation exists when-ever the reactor is critical and above atmospheric prossure.

An exception is made to this requ'.rement during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required.

There will be no pressure on the system at this time, thus oreatly reducing the chances of a pipe break.

The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occur ring.

Procedures and the Rod Worth Minimizer would limit control worth such that a

rod drop would not result in any fuel damage.

In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment

system, which shall be operational during this time, offer a sufficient barrier to keep offsite doses well below 10 CFR 100 limits.

The pressure suppression pool water proviges the heat sink for the reactor primary system energy release following a postulated rupture of the system.

The pressure suppression chamber water volume must absoro the associated decay and structural sensible heat released during primary system blowdown from 1,035 psig.

Since, all of the gases in ?he drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig,,the suppression chamber maximum pressure.

The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression

chamber, Using the minimum or maximum water volumes given in the specification, containment pressure duriqg the design basis accident is approximately 49 psig which is below the maximum of 62 psig.

Maximum water volume of 135,000 ft3 results in a downcomer submergence of 5'2-3/32" and the minimum volume of 123,000 ft3 results in submergence approx irately 12 inches less, The majority of the 8odega te~ts were run with a submerged length of 4 feet and with complete condensa

.ion.

Thus, with respect to downcomer submergence, this specification is adequate.

The maximum temperature at the end of blowdown tested during t'e Humoolt 8ay and Bodega 8ay tests was 170'F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condersa ion

'would occur for temperatures above 1'70'F.

Amendment No.

38 268

~

BASES Should it be necessary to drain the suppression chamber, this should only be done when there is no requirement for core standby cooling systems operatibility.

Under full power operation conditions, blowdown from an initial suppression chamber water temperature of 95'F results in a peak long term water temperature of 170'F which is sufficient for complete condensation; At this temperature and atmospheric

pressure, the available NPSH exceeds that required by both the RHR and core spray

pumps, thus there is not I

dependency on containment overpressure.

Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160'F during any period of relief valve operation with sonic conditions at the discharge exit.

Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressuirzed in a timely manner to avoid the regime of potentially high suppression chamber loadings.

Limiting suppression pool temperature to 105'F during RCIC, HPCI, or relief valve operation when decay heat and stored energy is removed from the primary system by discharging reactor steam directly to the suppression chamber assures adequate margin for controlled blowdown anytime during RCIC operation and assures margin for complete condensation of steam from the design basis loss-of-coolant accidents In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a

relief valve inadvertently opens or sticks open.

This action would include:

(1) use of all available means to close the valve, {2) initiate suppression pool water cooling heat exchangers (3) initiate reactor

shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open.relief valve,to assure mixing and uniformity of energy insertion to the pool.

If a loss-of-coolant accident were to occur when the reactor water

'emperature is below approximately 330'F, the containment pressure will not exceed the 62 psig code permissible pressures even if no condensation were to occur.

The maximum allowable pool temperature, whenever the reactor is above 212'F, shall be governed by this soecification.

Thus, specifying water volume-temperature requirements applicable for reactor-water temperature above 212'F provides additional margin above that

. available at 330'F.

~Incr tin The relatively small containment volume inherent in the GE-BMR pressure suppression containment and the large amount of zirconium in the co."e are such that the occurrence of a very limited (a percent or so) reaction o.

the zirconium and steam during a loss-of-coolant accident could lead to the liberation of hydrogen combined with an air atmosphere to result in a

flamnable concentration in the containment.

lf a sufficient amount cf hydrogen is generated and oxygen is available in stoichi'ometric quantities the subsequent ignition of the hydrogen in rapid recombination rate could lead to fai lure of the containment to maintain a low leakage

.'ntegrity.

The

~4% hydrogen concentration minimizes the possibility of hydrogen combustion following a loss-of-coolant accident.

BASES The occurrence of primary system leakage,ollowing a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the loss-of-coolant accident upon which the specified oxygen concentration limit is based.

Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety.

Thus, to preclude the possibility of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled during startup periods, when the primary system is at or near rated operating temperature and pressure.

The 24-hour period to provide inerting is judged to be sufficient to perform the leak inspection and establish the required oxygen concentration.

To ensure that thc hydrogen concentration is maintained less than 4% following an accident, liquid nitrogen is maintained on-site for containment atmosphere dilution.

About 2260 gallons would be sufficient as a 7-day supply, and replenishment facilities can deliver liquid nitrogen to the site within one day; therefore, a requirement oz 2500 gallons is conservative.

Following a loss of coolant accident the Containment Air Bonitoring (CAM) System continuously monitors the hydrogen concentration of the containment volume.

Two indepcndcnt systems

{ a system consists of one hydrogen sensing circuit) are installed in the drywell and one system is installed in the torus.

Each sensor and associated circuit is periodically checked by a calibration gas to verify operation.

Failure of a drywell system does not reduce the ability to monitor system atmosphere as a second independent and redundant system villstill be operable.

In terms of separability, redundancy for a failure of the torus system is based upon at least one operable drywell system.

The drywell hydrogen concentration can be used to limit the torus hydrogen concentration during post LOCA conditions.

Post LOCA calculations show that the CAD system initiated within two hours at a flow rate of 100 scfm will limit the peak drywell and wetwell hydrogen con-centration to 3.6% {at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) and 3.8%$ t 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months ), respectively.

T}is is based upon purge initiation after 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months at a flow rate of 100 scfm to maintain containment pressure belov 30 psig.'hus, peak torus hydrogen concentration can be controlled below 4.0 percent using either the direct torus hydrogen monitoring system or the drywell hydrogen monitoring system with appropriate conservatism (~ 3.8%),

as a guide for CAD/Purge operations.

Amendment No.

38 270

BABE'.i Vacuum Relief The purpose of the vacuum relief'a) ves is to equalize tlirt'icss>>rc between the drywell and suppression chamber. and reactor building so that the structural integrity of the containment is maintained.

The vacuum relief system from the pressure suppression chamber to reactor building consists of two 100$ vacuum relief breakers (2 parallel sets of 2 valves in series).

Operation of either system vill maintain the pressure differential less than 2 psig; the external design pressure.

One reactor building vacuum breaker may be out of service for repairs for a period of seven days. If repairs cannot be completed within seven days, the reactor coolant system is brought to a condition where vacuum relief is no longer required.

When a drywell-suppression chamber vacuum breaker valve is exercised "throuph an openinp-closinp cycle the position indicatinp liphts in the control room are desipned to function as specified belov:

Initial and Final Check On (Fully closed)

Condition Green On Red Off Opening Cycle Check Off Green - Off Red On (Cracked open)

(> 80 Open)

(> 3" Open)

Closing Cycle Check - On Green On Red

- Off (Fully Closed)

(< 300 Oper.)

< 3 Open)

The valve position indicating lights consist of one check light on the check light panel which confirms full closure, one green light next to the hand svitch vhich confirms 80 of full openinp, and one red lipht next o

to the hand switch which co>>firm" ">>car c'losure" (within 3" of full closure).

Hach lipht is on a separate svitch. If the check light circuit is oper~bio whc>> the valve is exercised l>v it" air ope) ator there exists a

confirmation that the valve will fplly close. If the red light circuit is

operable, there exists a confirmation that the valve will at least "nearly close" (vithin 3 of full closure).

The green light circuit con.irms the valve vill fully open.

If none of the lights chancre indication during the cycle, the air operator must be inoperable or the valve disc is stuck.

For this case, a check light on and red li~-,ht off confirms the di c is ir.

a near'y closed position even if one of the indications is in error.

Although the va've may be inoperable for full closure,.it does not consti-tute a safety threat.

If the red light circuit alor.e is inoperable, the valve shall sti' be considered fully operable.

Zf the f'reen and red or the preen light circuit along is inoperable the valve snail be sonsidered inoperable for 271

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+**y4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY. NUCLEAR PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

36 License No.

DPR-52 1.

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

A.

B.

The application for amendments by Tennessee Valley Authority (the licensee) dated May ll, 1978, 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; The facility will operate in conformity.with the application, the provisions of the Act, and the rules and regulations of the Commi.ssion; 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 Spec-ifications 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 Amendment'No. 36, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Spec.'fications.

Cl t

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

June 22, 1978 Thomas A.

po ito, Chief Operating Reactors Branch P3 Division of Operating Reactors

ATTACHMENT TO LICENSE AMENDMENT NO.

36 FACILITY OPERATING LICENSE NO.

DPR-52 DOCKET NO. 50-260 Revi se Appendi x A as fo 1 1 ows:

Remove the following pages and replace with identically numbered pages:

79/80 235/236 249/250 269/270 Marginal lines indicate revised arear'verleaf pages are provided for convenience.

TABEZ 3.2eF Surveillance Instrumentation Minimum g of Operable Instrument Channels Instrument g H2M 37 H M - 76 39 r

Instrument Drywell H Concentration Type Indication and Ra e

O.l -

2'M 38 Suppression Chamber H2 Concentration 0.1 -

2'mendment No.. 36 sos s'rsaNs>>>>ssGC'~sossass+~s~s'.s aea~vars>>ss:rheo>>abase>>o>>o "s.,s= sasar.

so,. >>ra

~.i >>..s.>>

~.~, ~

~

~=...

KOTES FOR TABLE

.2.F Prow and after the dnte that nnc of these parnm~t.er ~ in reduced tn one indication, continued operation is permissible during the succeeding thirty days unless such instrumentation is sooner made o erable.

p I

(2)

From and after the date that one of these parameters is not indi-cated in the control room, continued operation is permissible during the. succeeding seven days unless such instrumentation is sooner made opersblc.

(3)

If the requirements of notes (1) and (2) cannot be met, either the requirements of 3.5.H shall be complied Mith or an orderly shutdown shall be initiated and the reactor shall be in a Cold Condition vithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(4)

These surveillance instruments are considered to be redundant to each other.

80

3. 7.A Primar Containment SURVEILLANCE RE VIREMENTS

4. 7.A Pri>ear Containment

'alves shall be exercised immediately and every 15 days thereafter until the inoperable valve has been

~ returned to morsel service.

c.

TVo dryvcl 1-suppression chamber vacuum breakers may be d<<termined to be inoperable for opening.

c ~

Once each operating cycle each vacuum breaker valve shall be inspected for proper operation of the valve and limit svitch<<s.

d.

If specifications 3.7.A.4.a,

.b, or.c cannot be met, the unit ohall be placed in a cold shutdovn condition in an orderly manner vithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d.

A leak teat of the dryvell to suppression chamber structure shall be con-ducted during each operating cycle.

Accept-

," able leak rate is 0.14 lbl sec of primary containment atmosphere vith I psi differential.

5.

Ox cn Concentration 5.

0 ea Concentration a.

After completion of che fire-related startup retestin program>

containment atmosphere shall be reduced to less then 42 oxygen vith nitro-gen gas during reactor power operation vith reac-tor coolant pressure above 100 psig, except ae speci-fied in 3.7.A.5.b.

b.

Mithin the 24-hour period subsequent to placing the reactor in the Run mode folloving a shu dovn, thc containment atmosphere oxygen concentration shall be reduced co less than 4X by vei>ch>>

and >t>aincaincd in this condition.

De-incr t-ing may commence 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to a shutdovn.

a.

The primary containment oxygen concentration shall be measures and recorded daily.

The oxygen measurement shall be adjusted to account for the uncertainty of the method used by adding a

predetermined error function.

b.

The methods used to measure the primary containment oxygen con-centration shall be calibrated once every refueling cycle.

ce If specification 3,7.A.5.a and 3,7.A.5. b cannot be mec, an orderly 735 shutdown shall be initiated and the reactor r hall be in a Color Shucdovn condicion vithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Amendment No.

36

I.I'INC COiMDITIO'./S FOR OPEMETIO'3

.0 I

SU RYE ILLAiNCE R~Q"IR.".'"Z"ITS

l. 1

(:<ztdTA 1 <441".ilT S Y."TYHS i

.4. 7 COt'<TA I NNEHT SYSTEMS B.

St.lni!bv Gas Treatment Gv. t;.m Excel< as specified in Specification 3.7.B. 3 below, all three trains of the standby gas treatment system

~ and the diesel generator s required for operation of 'such trains shall be" operable at all times when secondary containment integrity is required.

B.

a ~

Pressure drop across the combined HG'A filters and charcoal adsorber Lanks is less than 6

inches of water at a flow of 9000 cfm

(+

10Ã).

I S-.ando Gas Treatment S vs t Bill At least once per yea r, the fol1owing conditions shall be demonstrat d.

b.

C ~

The inlet heaters cn each circuit a.

capab~

oc an output of at least 40 kw wnen tested in accordance with ANSI H510-1975.

Air dist"ibution is uniform within 20%

across HEPA filters and charcoal adscrbers.

236

LIMITING CONDITIONS POR OPERATION SURVEILLANCE REQUIREHENTS 7

CO57TAINMELLL'YSTEMS 4. 7 CONTAIN<lENT SYSTEMS 2

H Anal zer Whenever the reactor is not in cold shutdown, two gas analyzer systems shall be operable for monitoring the drywell.

Containment Atmosphere Monitorin CAM S stem-H.

Containment, Atmos here Monxtorxn CAN S stem-Anal zer Hz Once per month perform a channel calxbratxon using stand,ard gas samples contaxnzng a nominal Three volume percent hydrogen balance nitrogen 2.

Whenever the reactor is not in cold shutdown, one gas analyzer system (one hydrogen sensing circuit per system) shall be operable for monitoring the torus.

3.

If specification

3. 7. H. 1 cannot be met, but one system remains operable, the reactor may be operated for a period of 30 days.

If both systems are inoperable, the reactor should be placed in shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If specification 3.7.H.2 cannot be met, the reactor may be operated for a period of 30 days.

Amendment No.

36 249

TABLE 3.7.A P<fINARY CONThI~ ISOLATION VALVES

~Grou Valve Identification in steamline isolation valves (FCV-1-14,26,37,&5I jl-15, 27, 38,

& 52)

.'fain stea<<<line drain.isolation valves FCU-1-55

& 1-56 Reactor Mater sample line isola-tion valves Number of Pover crated Valves Inoo'rd Outboard 4

4

.'faximum

.Opera ting Tine (sec.)

3<T< 5 15 Aor<aal Pos<L<aa 0

hction on Initiating Ssi<nal SC SC RfMS sliutdovn cooling supply isolation valves FCV-74-48

& 47 40 SC 2

RfNS LPCI to reactor FCV-74-53, 67 30 SC Reactor vessel nead spray isola-tion valves FCV-74-77, 78 30 C

RffRS fluff< and drain vent to suppre<<sion chamber FCV-74-102,

103, 1.19 '

120 Suppression Ct<aaber Drain FCV-74-57, 58 DryM'-ll equip+ant drain discharge i<<olatio<l valves FCV-77-].5h

& ]5B 20 15 C

SC 2

Dry. ell floor drain discharge isolation valves FCV-77-2h

& 2B

BASES Should it be necessary to drain the suppression

chamber, this should only be done when there is no requirement for core standby cooling systems operatibility.

Under full power operation conditions, blowdown from an initial suppression chamber water temperature of 95'F results in a peak long term water temperature of 170'F which is sufficient for complete condensation.

At this temperature and atmospheric

pressure, the available NPSH exceeds that required by both the RHR and core spray

pumps, thus there is not dependency on containment overpressure.

Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160'F during any period of relief valve operation with sonic conditions at the discharge exit; Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressuirzed in a timely manner to avoid the regime of potentially high suppression chamber loadings.

Limiting suppression pool temperature to 105'F during RCIC, HPCI, or relief valve operation when decay heat and stored energy is removed from the primary system by discharging reactor steam directly to the suppression chamber assures adequate margin for controlled blowdown anytime during RCIC operation and assures margin for complete condensation of.steam from the design basis loss-of-coolant accident.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a

relief valve inadvertently opens or sticks open.

This action would include:

(1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers (3) initiate reactor

shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relic. valve to assure mixing and uniformity of energy insertion to the pool.

If a loss-of-coolant accident were to occur when the reactor water temperature is below approximately 330'F, the containment pressure will not exceed the 62 psig code permissible pressures even if no condensation were to occur.

The maximum allowable pool temperature, whenever the reactor is above 212'F, shall be governed by +his soecification.

Thus, specifying water volume-temperature requirements applicable for reactor-water temperature above 212'F provides additional margin above that available at 330'F.

~lnerti n The relatively small containment volume innerent in the GE-BMR pressure suppression containment and the large amount of zirconium in the co."e are such that the occurrence of a very limited (a oercent or so) reaction of the zirconium. and steam during a loss-of-coolant accident could lead to the liberation of hydrogen combined with an air atmosphere to result in a

flammable concentration in the containment.

If a sufficient amount of hydrogen is generated and oxygen is available in stoichi'ometric quanzi-ies the subsequent ignition of the, hydrogen in rapid recombination rate could lead to failure of the containment to maintain a low leakaoe

.'ntearitv.

The

~ 4% hydrogen concentration minimizes the possibility of hydrogen combustion following a loss-of-coolant accident.

269

8ASES The occurrence of primary system leakage ol iowing a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the loss-of=coolant accident upon which the specified oxygen concentration limit is based.

Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety.

Thus, to preclude the possibility of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled during startup periods, wren the primary system is at or near rated operating temperature and pressure.

The 24-hour period to provide inerting is judged to be sufficient to perform the leak inspection and establish the required oxygen concentration.

To ensuxe that the hydrogen concentration is maintained less than 4% following an accident, liquid nitrogen is maintained on-site goy containment atmosphere dilution.

About 2260 gallons would be sufficient as a 7-day supply, and replenishment facilities can deliver liquid nitrogen to the, site within one day; therefore, a requirement of 2500 gallons is conservative.

Following a loss of coolant accident the Containment Air Monitoring (CAN) System continuously monitors the hydrogen concentration of the containment volume.

Two independent systems

( a system consists of one hydrogen sensing circuit) are installed in the drywell and one system is installed in the torus.

Each sensor and associated cix'cuit is periodically checked by a calibration gas to verify operation.

Failure of a drywell system does not reduce the ability to monitor system atmosphere as a second independent and x'edundant system willstill be operable.

In terms of separability, redundancy for a failure of the torus system is based upon at least one operable drywell system.

The drywell hydrogen concentration can be used to limit the toxus hydrogen concentration during post LOCA conditions.

Post LOCA calculations show that the CAD system initiated within two hours at a flow rate of 100 scfm will limit the peak drywell and wetwell hydrogen con-centration to 3.6% (at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) and 3.8% at 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months ), respectively.

TKs is based upon purge initiation after 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months at a flow rate of 100 scfm to maintain containment pressure below 30 psig.

Thus, peak torus hydrogen concentration can be controlled below 4.0 percent using either the direct torus hydrogen monitoring system or the d~r e11 hydrogen monitoring system w'th appropriate conservatism

(

,3.8%),

as a guide for CAD/Purge operations.

270

<p.S ReCu<

(4 WpO UNITEDSTATES NUCLEAR REGULATORYCOMMISSION WASHlNGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BRONNS FERRY NUCLEAR PLANT UNIT NO.

3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 12 License No.

DPR-68 1.

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

A.

The application for amendments by Tennessee Valley Authority (the licensee) dated May ll, 1978, 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.. 'ccordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment. to this license amendment and paragraph 2.C(2) of Facility 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. 12, are hereby incorporated, in, the license.

The licensee shall, operate the facility in accordance with the Technical Specifications.

Cl I

3.

This 1icense amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY CONISSION 4

Thomas ppo1ito, Chief Operating Reactors Branch P3 Division of Operating Reactors

Attachment:

Changes to the Technica1 Specifications Date of Issuance:

June 22, 1978

"ATTACHMENT TO LICENSE AMENDMENT NO.

12 FACILITY OPERATING LICENSE NO.

DPR-68 DOCKET NO, 50-296 Revise Appendix A as follows:

Remove the following pages and replace with identically numbered pages:

82 245 261 286A 287 Marginal lines indicate revised area.

TABLE 3.2.P SURVEILLANCE INSTRUMENTATION Minfssss 6 of Operable Instrument Cluuuwls Instrument t HqM 76 - 37 HaÃ 39 Instrument Drywall H~

Concentration Type Indication and Range 0.1 - 20%

Notes HqM 76 - 3$

Suppression Chamber H~ Concentration 0.1 - 20%

Amendment No.

12

tINITZNG CONDITIONS FOR "'OPERATION SURVEILLANCE REQUIREMENTS

) ~ 7 NT NUGENT SYST MS

4. 7 NTAINMENT SYST S

d.

If speci fications 3 7A4 a~

~ b~

or.c, cannot be met, the unit shall be placed in a cold shutdown condition in an orderly manner within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

Within the 24-hour period subsequent to placing the reactor in the Run mode following a shutdown, the containment atmosphere oxygen concentration shall be reduced to less than 4%

by weight and aaintained in this condition.

De-inerting may commence 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to a shutdown.

5.

Oxygen Concent,rat,ion After completion of the 300-hour warranty run, conta inme nt atmosphere shall be reduced to less than 4%

oxygen with nitrogen gas during reactor power operation with reactor coolant pressure above 100 psig, except as specified in

3. 7 A.5.b.

245 d.

h leak test of the drywell to suppr essien chamber structure shall be conducted during each operating cycle.

Acceptable leak rate is 0.14 lb/sec of primary containment atmosphere with psi differential.

Oxygen Concentration a.

The primary containment oxygen concentration shall be measured and recorded daily.

The exygen measurement shall be adjusted to account for the uncertainty of the method used by adding a predetermined error function.

b.

The methods used to measure the primary con-tainment oxygen concen-tration shall 6e cali-brated once every refuel-ing cycle.

Amendment No.

12

'I]lITI N(p CONDITIOVS POR OPERATION 3.7

. CONTAINNEzza'YSTEMS SU RYEILLANC-RFOUIREAGENTS 4.7 CONTAINtlENT SYSTENS H

~Anal zer

'I ~

Whenever the reactor is not in cold

shutdown, two gas analyzer systems shall be operable for monitoring the drywell.

2 ~

Whenever the reactor is not in cold shutdown, one gas analyzer system shall be operable for monitoring the torus.

Containment Atmosohe e

Honitorin CA'I S stem-Containment Atmos here Monxtorxn CAN S stem H~

~Ana zer Once per month perform a channel calibration using standard gas samples containing a nominal three volume precent

hydrogen, balance nitrogen.

I 3 e If specification 3.7.H.

1 cannot be met, but one system remains operable, the reactor may be operated for a period of 30 days.

If both systems are inoperable, the reactor should he placed in shutdown condition within 2'ours.

If specification 3.7.8.2 cannot be met, the reactor may be operated for a period of 30 days.

Amendment No.

12 261

I

~nertin The relatively small containment volume inherent in the GE-BWR pressure suppression containment and the large amount of zirconium in the core are such that the occurrence of a very limited (a percent or so) reaction of the zirconium and steam during a loss-of-coolant accident could lead to the liberation of hydrogen combined with an air atmosphere to result in a flammable concentration in the contai.nment.

If a sufficient amount of hydrogen is generated and oxygen is available in stoichiometric quantities, the subsequent ignition of the hydrogen in rapia recombination rate could lead to failure of the containment to maintain low leakage integrity.

The (4% hydrogen concentration minimizes the possibility of hydrogen combustion following a loss-of-coolant accident.

The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the loss-of-coolant accident upon which the specified oxygen concentration limit is based.

.Permitting access to the drywell for leak inspections duri.ng a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety.

Thus, to preclude the possibility of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are acheduled during etartup

periods, when the primary system ie at or near rated operating temperature and pressure.

The 24-hour period to provide inerting is judged to be sufficient to perform the leak inspection and establish the required oxygen concentration.

To ensure that the hydrogen concentration is maintained less than 4%

following an accident, liquid nitrogen is maintained on-site for contain-ment atmosphere dilution.

About 2260 gallons would be sufficient as a

7-day supply, and replenishment facilities can deliver liquid nitrogen to the site within one day; therefore, a requirement of 2500 gallons is conservative.

Following a loss-of-coolant accident the Containment Air Monitoring (CAM)

System continuously monitors the hydrogen concentration of the.containment volume.

Two independent systems (a system consists of one hydrogen sensing circuit) are installed in the drywell and one system is installed in the torus.

Each sensor and associated circuit is periodically checked by a calibration gas to verify operation.

Failure of a drywell system does not reduce the ability to monitor system atmosphere as a second independent and redundant system villstill be operable.

In terms of separability, redundancy for a failure of the torus system is based upon at least one operable drywell system.

The drywell hydrogen concentration can be used to limit the torus hydrogen concentration during

,post LOCA conditions.

Post LOCA calculations show that the CAD system within two hours at a flow rate of 100 scfm will limit the peak drywell and wetwell hydrogen concentration to 3.9% (at 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ),and 3.9% (at 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months ) respectively.

This is based upon purge initiation after 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months at a flow rake of 100 scfm to maintain containment pressure below 30 psig.

Thus, peak torus hydrogen concentration can be controlled below 4.0 percent using either the direct torus hydrogen monitoring system or the drywell hydrogen monitoring system with appropriate conservatism

(~ 3.9%),

as a guide for CAD/Purge operations.

Amendment No.

12 286A

Vacuum Reli f The purpose of the vacuum relief valves is to equalize the pressure between the dryvell and suppression chamber and reactor building so that the structural integrity of the containment is

'aintained.

The vacuum relief system fran the pressure suppression chamber to reactor buiding consists of tvo 100%

vacuum relief breakers (2 parallel sets of 2 valves in series).

Operation of either system vill maintain the pressure differential less than 2 psig; the external design pressure.

One reactor building vacuum breaker may be out of service for repairs for a period of seven days.

If repairs cannot be completed vithin seven

days, the reactor coolant system is brought to a condition vhere vacuum relief is no longer required.

Shen a dryvell-suppression chamber vacuum breaker valve is exercised through an opening-closing cycle the position indicating lights in the control room are designed to function as opecified belov:

Amendment No.

12 287

gag RK0I

~o Cp I

~ y OO 0~-

v/g I~~

()w

~+*++

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO, 38 TO FACILITY OPERATING LICENSE NO.

DPR-33 AMENDMENT NO.

36 TO FACILITY OPERATING LICENSE NO.

DPR-52 AMENDMENT NO.

12 TO FACILITY OPERATING LICENSE NO.

DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT NOS.

1, 2 AND 3 DOCKET NOS. 50-259, 50-260 AND 50-296 1.0 Introduction By letter dated May ll, 1978, the Tennessee Valley Authority (the licensee or TVA) requested changes to the Technical Specifications (Appendix A) appended to Facility Operating Licenses Nos.

DPR-33, DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant, Units Nos.

1, 2 and 3.

The proposed amendments and revised Technical Specifica-tions would delete the surveillance requirements for the containment oxygen monitoring instrumentation on the basis that this instrumentation is not required for post-accident monitoring.

The Technical Specifications for the Browns Ferry Plant currently require, that specific instrumentation be used to monitor the oxygen concentration in the containment during normal plant operation, and identify the surveillance requirements for that instrumentation.

The primary containment is operated with an oxygen-deficient (i.e.,

inerted) atmosphere as part of those measures for combustible gas control following a postulated loss-of-coolant accident (LOCA).

Based on its reassessment of the design of the combustible gas control

system, the licensee has requested that the requirements for specific oxygen monitoring instrumentation be deleted from the plant's Technical Specifications.

Those Technical Specification require-ments which limit the maximum oxygen concentration in the containment during normal operation would be retained (i.e,, the containment will continue to be inerted during normal operation).

The licensee has proposed these changes to allow the use of alternate measurement techniques to establish the containment oxygen concentra-tion.

Past experience has shown that the existing oxygen monitoring subsystems have not demonstrated sufficiently reliable performance.

2.0 Discussion Following a postulated LOCA, hydrogen is generated from a reaction between the Zircaloy fuel cladding and the primary coolant (metal-water reaction),

and both hydrogen and oxygen are generated as a result of radiolysis of the primary coolant and the water in the suppression pool.

When a sufficient quantity of hydrogen has been generated in an oxygenated atmosphere, a combustible mixture of gasses is formed.

In order to protect the containment structure and engineered safety feature systems from the potential consequences of combustion, a Combustible Gas Control System (CGCS) is incorporated into the plant design.

{reference 1).

For Browns Ferry, the CGCS consists of inerting the containment atmosphere with nitrogen during normal operation and dilut-ing the containment atmosphere with nitrogen following a postulated LOCA to maintain the hydrogen concentration below the flammability 1 imit.

The operation of the CGCS is now to bebased on the measured hydrogen concentration in the containment following a postulated accident.*

The operation of the CGCS was previously based on the monitored oxygen concentration.

Control of either hydrogen or oxygen will assure that a combustible mixture of gases will not be formed.

The initially inerted containment atmosphere assures that the hydrogen released from any metal-water reaction will not exceed a combustible concentration, and it provides a longer period of time to the point at which the nitrogen dilution system is manually initiated.

Two inter-related

systems, Containment Atmosphere Nonitoring (CAN) and Containment Atmosphere Dilution (CAD) systems, are provided in the Browns Ferry Units 1-3 plant to monitor the concentration of oxygen and hydrogen and thereby prevent the creation of a combustible gas mixture.in the primary containment following a LOCA.

In order to en'sure that a com-bustible mixture is not created, either the hydrogen concentration must remain below 4X by volume or the oxygen concentration must remain below 5X by volume.

As discussed

above, the only significant sources of hydrogen and oxygen buildup in the containment following a LOCA are hydrogen evolution from metal-water reactions and the radiolysis of water.

Since the buildup of hydrogen following a LOCA will occur more rapidly than the oxygen buildup due to the early occurrence of a metal-water reaction, control based only on hydrogen monitoring is adequate if the H2 concentration is demonstrated by analysis to be less than 4X.

The hydrogen concentration is controlled and the containment pressure is maintained below 30 psig by operation of the CAD system and sub-sequent purge.

The offsite dose following the purgung must be less than or equal to the guidelines contained in 10 CFR 100.-

lo, 3.0 Evaluation As discussed

above, the licensee submitted an evaluation to demonstrate that the hydrogen sensors could be used exclusively to monitor contain-ment atmosphere for a combustible mixture following a postulated LOCA and that the present requirement for both hydrogen and oxygen sensors was unnecessary and redundant, provided other acceptable means of assur-ing that the oxygen concentration during normal operation can be demon-strated to be less than 4X.

As additional support, the licensee also submitted calculations to demonstrate that the potential hydrogen con-centration in containment would be below a combustible mixture follow-ing a postulated LOCA even if the containment was not inerted (i.e.,

if the atmosphere were not diluted with nitrogen as currently required by the Technical Specifications to reduce the oxygen concentration to less than 4X). The licensee did not request to delete the requirements for inerting and this subject is not covered by this safety evaluation.

In addition, the licensee has not proposed to change the requirement in Section 4.7.A.5 of the Technical Specifications that the primary con-tainment oxygen concentration be measured and recorded daily.

I We find that the current requirements for surveillance testing of the oxygen monitoring instrumentation are not necessary.

Therefore, the Technical Specifications have been modified to delete the specific requirements for oxygen instrumentation and to include the more general requirements for surveillance of and calibration by an. appropriate method which the licensee would choose to establish the oxygen concentra-'ion during normal operation.

The staff has

proposed, and the licensee has agreed, to modify the surveillance requirements in Section 4.7.A.5 of the Technical Specifications to specify that the uncertainty asso-ciated with the method be predetermined, and factored into the measure-ment of the oxygen concentration.

The measurement will be taken daily as previously discussed.

-These requirements assure that the CGCS for the Browns Ferry'lant will be able to maintain the hydrogen concentra-tion below a combustible level in the unlikely event of a LOCA.

Accord-ingly, we find the changes to the Technical Specifications are acceptable.

4.0 Environmental Considerations We have determined that these amendments do 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 these amendments involve 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 these amendments.

5.0 Conclusion We have concluded that:

(1) because the amendments do not involve a

significant increase in the probability or consequences of accidents previously considered and do not invol.ve a significant decrease in a safety margin, the amendments do not involve a significant hazards con-sideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

manner, and (3) such activities will 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.

Dated:

June 22, 1978

7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NOS. 50-259, 50-260 and 50-296 TENNESSEE VALLEY AUTHORITY

,I NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES The U.

S. Nuclear Regulatory Commission (the Coranission) has issued Amendment No.

38 to Facility Operating License No.

DPR-33, Amendment No.

36 to Facility Operating License No.

DPR-52 and Amendment No.

12 to Facility Operating License No.

DPR-68 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, located in Limestone County, Alabama, The amendments are effective as of the date of issuance.

The amendments change the Technical Specifications to delete the requirements for the-oxygen.sensors used in the containment atmosphere monitoring system and augment the surveillance requirements associated with the daily oxygen analyses of primary containment.

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 appro-priate 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 required 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 appraisal need not be prepared in connection with issuance of these amendments.

1 I

For further details with respect to this action, see (1) the appli-cation for amendments dated May ll, 1978, (2) Amendment No.

38 to License No.

DPR-33, Amendment No.

36 to License No.

DPR-52, and Amendment No.

12 to License No.

DPR-68, and (3) the Commission's related Safety Evaluation.

All of these items are available for public inspection at the Conmission'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 Commission, Washington, D.

C.

20555, Attention:

Director, Division of Operating Reactors.

Dated at Bethesda, Maryland, this 22nd day of June 1978.

FOR THE NUCLEAR REGULATORY COMMISSION Thomas A.

polito, Chief Operating Reactors Branch P3 Division of Operating Reactors

P

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