ML20214C167

Amend 2 to License NPF-37,approving Changes to Tech Specs to Correct Typos & Grammatical Errors

ML20214C167
Person / Time
Site:	Byron
Issue date:	02/13/1986
From:	Noonan V Office of Nuclear Reactor Regulation
To:
Shared Package
ML20214C170	List: ML20214C167
References
NUDOCS 8602210037
Download: ML20214C167 (16)

Text

..

o UNITED STATES

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7,i -

NUCLEAR REGULATORY COMMISSION WASHINGTON D. C. 20555 -

COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-454 BYRON STATION, UNIT 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE" Amendment No. 2 License No. NPF-37 1.

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

A.

The application for amendment by Comonwealth Edison Company (the licensee) dated September 27, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Connission'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 Comission.

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) th,at such activities will be conducted in compliance with the Comission'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 Comission's regulation and all applicable requirements have been satisfied.

8602210037 860213 PDR ADOCK 05000454 P

PDR

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4e 2

' -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.

NPF-37 is hereby amended to read as follows:

(2) Technical Specifications

._The Technical Specifications contained in Appendix A, as revised through Amendm'ent No. 2, and the Environmental Protection Plan contained in Appendix B, bo.th of which are attached hereto, are hereby incorporated into this license. The licensee shall operate.

the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/5 Vincent S. Noonan, Director PWR Project Directorate #5 Division of PWR Licensing-A

Attachment:

Changes'to the Technical

. Specifications Date of Issuance: FEB 13 ES g

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ATTACHMENT TO LICENSE AMENDMENT N0. 2 FACILITY OPERATING LICENSE N0. NPF-37 DOCKET NO. STN 50-454 Revise 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 page (*) has been provided to maintain document completeness.

REMOVE INSERT 3/4 3-15 3/4 3-15 3/4 3-16*

3/4 6-7 3/4 6-7 3/4 6-8*

3/4 7-44 3/4 7-44 3/4 7-43*

. '3/4 8-5 3/4 8-5 t

3/4 8-6*

?

3/4 11-8 3/4 11-8 3/4 11-7*

3/4 6-13 3/4 6-13 3/4 6-14*

y -

y.

-e,

,4 e

ADMINISTRATIVE CONTROLS ONSITE (Continued) 3)

Review of all proposed changes to the Technical Specifications; 4)

Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety; 5)

Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the

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Division Vice President and General Manager Nuclear Stations and to the Supervisor of.the Offsite Review and Investigative Function; 6)

Review of-all REPORTA8LE EVENTS; 7)

Perfornance of special reviews and investigations and reports thereon as requested by the Supervisor of the Offsite Review and Investigative Function; 8)- Review of the Station Security Plan and implementing procedures and submittal of recommended changes to the Division Vice President and General Manager - Nuclear Stations; 9)

Review of the Emergency Plan and station implementing procedures and submittal of recommended changes to the Division Vice

. President and General Manager - Nuclear Stations; l

10) Review of Unit operations to detect potential hazards to nuclear safety;

-11) Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to i

i prevent recurrence and the forwarding of these reports to the Division Vice President and General Manager - Nuclear Stations and the Supervisor of the Offsite Review and Investigative j.

Function; and

12) Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE DOSE CALCULATION MANUAL, and the Radwaste Treatment Systems.

c.

Authority The Technical Staff Supervisor is responsible to the Station Superintendent and shall make recommendations in a timely manner in j

all areas of review, investigation, and quality control phases of plant maintenance, operation, and administrative procedures relating

?

to facility operations and shall have the authority.to request the action necessary to ensure compliance with rules, regulat. ions, and j

procedures when in his opinion such action is necessary.

The Station l

Superintendent shall follow such recommendations or select a course BYRON - UNITS 1 & 2 6-13 Amendment No. 2 i

..... _. -,..... - ~. - _... _ _ -. -... -.. _ _ -. _

ADMINISTRATIVE CONTROLS ONSITE (Continued) of action that; is more conservative regarding safe op'eration of the facility.

All such-disagreements shall be reported immediately to j

the Division Vice President and General Manager - Nuclear Stations and the Supervisor of the Offsite Review and Investigative Function.

i d.

Records

.i 1)

Reports, reviews, investigations, and recommendations shall be documented with copies to the Division Vice President and General Manager - Nuclear Stations, the Supervisor of the Offsite Review and Investigative Function, the Station Superintendent, and the Manager of Quality Assurance.

2)

Copies of all records and documentation shall be kept on file -

at the station.

f i

e.

Procedures I

Written administrative procedures shall be prepared and maintained for conduct of the Onsite Review and Investigative Function.

These procedures shall include the following:

1)

Content and method of submission and presentation to the Station Superintendent, Division Vice President and General Manager - Nuclear Stations, and' the Supervisor of the Offsite Review and Investigative Functf5n, 2)

Use of committees, 3)

Review and approval, i

4)

Detailed listing of items to be reviewed, t

5)

Procedures for administration of the quality control l

activities, and 6)

Assignment of responsibilities.

l f.

Personnel i'

1)

The personnel performing the Onsite Review and Investigative Function, in addition to the Station Superintendent, shall consist of persons having expertise in:

a)

Nuclear power plant technology, b)

Reactor operations, c)

Reactor engineering, d)

Chemistry e)

Radiological controls, f)

Instrumentation and control, and g)

Mechanical and electrical systems.

BYRON - UNITS 1 & 2 6-14

l TABLE 3.3-3 E

g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION s

MINIMUM i

g TOTAL NO.

CHANNELS CHANNELS APPLICA8LE q

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES-ACTION m

H 1.

Safety Injection (Reactor n=

Trip, Feedwater Isolation, Start Diesel Generators, m

i Containment Cooling Fans, i

Control Room Isolation,

'l Phase "A" Isolation Turbine Trip, Auxiliary Feedwater,

]

Containment Vent Isolation, and Essential Service Water).

i

\\

a.

Manual Initiation 2

1 2

1,2,3,4 18

[

b.

Automatic Actuation 2

1 2

1,2,3,4 14 Logic and Actuation i

i IN Relays c.

Containment 3

2 2

1,2,3 15*

Pressure-High-1 d.

Pressurizer Pressure-4 2

3 1, 2, 3#

19*

Low (Above P-11) l e.

Steam Line Pressure-3/ steam line 2/ steam line 2/ steam line 1, 2, 3#

15*

l Low (Above P-11) any steam line 2.

Containment Spray a.

Manual Initiation 2 pair 1 pair 2 pair 1, 2, 3, 4 18 ig b.

Automatic Actuation 2

1 2

1,2,3,4 14 Logic and Actuation m

Relays c.

Containment Pressure-4 2

3 1,2,3 16 High-3 i

1

TABLE 3.3-3 (Centinued)

{

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION I

e

-MINIMUM g

TOTAL No.

CHANNELS CHANNELS APPLICABLE q

FUNCTIONAL UNIT OF CHANNELS.

TO TRIP

, OPERABLE M00ES ACTION vi w

3.

Containment Isolation j

o.

~~

)

=

a.

Phase "A" Is'olation 1)

Manual Initiation 2

'1 2

1,2,3,4 18 i

2)

Automatic Actuation 2

1 2

1,2,3,4 14 Logic and Actuation i

Relays 3)

Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.

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b.

Phase "B" Isolation I

Y 1)

Manual Initiation 2 pair 1 pair 2 pair 1, 2, 3, 4 18 m

2)

Automatic Actuation 2

1 2

1,2,3,4 14 Logic and. Actuation Relays i

3)

Containment 4

2 3

1,2,3 16 Pressure-High-3

)

c.

Containment Vent j

Isolation 1)

Automatic Actuation 2

1 2

1, 2, 3, 4 17 Logic and Actuation Relays 2)

' Manual Phase "A" See Item 3.a.1 for all manual Phase "A" Isolation initiating Isolation functions and requirements.

3)

Manual Phase "B" See Item 3.b.1 for all manual Phase "B" Isolation initiating Isolation functions and requirements.

4)

Safety Injection See Item 1. above for all Safety Injection initiating i

functions and requirements.

e

l i

CONTAINMENT SYSTEMS j

AIR TEMPERATURE LIMITING CONDITION FOR OPERATION i

3.6.1.5 Primary containment average air temperature shall not exceed 120*F.

i APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

l With the containment average air temperature greater than 120*F, reduce the i

average air temperature to within the limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or be in at least i

HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following i

30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

1 i

f i

i j

SURVEILLANCE REQUIREMENTS i

4.6.1.5 The primary containment average air temperature shall be the arithmetical average of the temperatures of the running fans at the i

following locations and shall be determined at least once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

i l

Location l

l A RCFC Dry Bulb Inlet Temperature B RCFC Dry Bulb Inlet Temperature C RCFC Dry Bulb Inlet Temperature D RCFC Dry Bulb Inlet Temperature i

1 i

L i

i l

BYRON - UNITS 1 & 2 3/4 6-7 Amendment No. 2 4

, -.. - -.. ~.,.-.. -, -. -. -,.., -. - - -. -, -., -, _ _.

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, - -,..,,.. - -.,,.,.,,. +., - - -, - -, -,. -., -,...,, _... - -

1 1

CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY

)

LIMITING CONDITION FOR OPERATION

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3.6.1.6 The structural integrity of the containment vessel shall be maintained l

i at a level consistent with the acceptance criteria in Specifications 4.6.1.6.1, j

4.6.1.6.2, and 4.6.1.6.3.

APPLICABILITY:

MODES 1, 2, 3, and 4.

I i

ACTION:

With more than one tendon with an observed lift-off force between a.

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the predicted lower limit and 90% of the predicted lower limit or I

with one tendon below 90% of the predicted lower limit, restore the t

tendon (s) to the required level of integrity within 15 days and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 30 days in accordance with i

Specification 6.9.2 or be in at least H0T STAN08Y within the next j

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

i b.

With any abnormal degradation of the structural integrity other than

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ACTION a. at a level below the acceptance criteria of Specifications 4.6.1.6.1, 4.6.1.6.2, and 4.6.1.6.3, restore the containment vessel to i

the required level of integrity within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and perform an engi-neering evaluation of the containment and provide a Special Report to i

the Commission within 15 days in accordance with Specification 6.9.2 or be in at least HOT STAN08Y within the r. ext 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l SURVEILLANCE REQUIREMENTS

~

4.6.1.6.1 Containment Vessel Tendons.

The containment vessel tendons' struc-tural integrity shall be demonstrated at the end of 1, 3, and 5 years following the initial containment vessel structural integrity test and at 5 year intervals thereafter.

The tendons' structural integrity shall be demonstrated by:

i a.

Determining that a random but representative sample of at least 19 l

tendons (5 done, 6 vertical, and 8 hoop) each have an observed lift.-

l off force within predicted limits for each.

For each subsequent inspection one tendon from each group may be kept unchanged to develop a history and to correlate the observed data.

If the observed lift-off force of any one tendon in the original sample population lies between the predicted lower limit and 90% of the predicted lower limit, two tendons, one on each side of this tendon should be checked for their lift-off forces.

If both of these adjacent tendons are found to be within their predicted limits, all three tendons should be restored to the required level of integrity.

This afngle deficiency may be considered unique and acceptable.

Unless there is abnormal degradation of the containment vessel during the first three inspec-tions, the sample population for subsequent inspections shall include at least 10 tendons (3 done, 3 vertical, and 4 hoop);

BYRON - UNITS 1 & 2 3/4 6-8

t TA8LE 3.7-5b (Continued)

(Unit 2) i 1 a, FIRE HOSE STATIONS

. w:

l LOCATION ELEVATION H0SE RACK REEL ANGLE VALVE i*

Aux. 81do (Continued)

M-26:

Radweste Control Panel 387 109 0FP377 L!

M-18:

By Aux. Feedwater motor driven pump 1A 387 108 0FP383 I

jw N-23:

By remote shutdown panel U-1 387 111 0FP376 i s.

Q-15:

By 480V MCC 132X3 387 113 0FP382

m V-18

By letdoun heat exchanger 387 114 0FP379 P-29: West wall 6.9kV SWGR Rm U-2 455 29 0FP339 i

i L-11:

In UC HVAC Am OA of LCSR C-1 455 22 0FP332 L-25:

By 08VC HVAC Room 455 27 0FP335 M-8:

South wall of battery room 451 279 0FP638 j

M-26:

South wall of battery room 451 280 0FP639

}

M-18:

North wall U-1 AB by door 444 238*

0FP463 L-7:

East wall LCSR A-1 443 207*

OFP330 i

M-10:

In the southeast corner of LCSR 8-1 443 208*

OFP327 2

P-10:

In the southeast corner of LCSR 8-1 '

443 209*

OFP325

,Y M-13:

South wall of LCSR C-1 443 210*

0FP326

'C P-13:

West wall of LCSR D-1 443 211*

OFP328 M-18:

South wall of LCSR U-2 A8 By Door s

}

L-29:

East wall LCSR A-2 j

443 239*

OFP464 443 212*

OFP336 M-26:

In Northeast corner of LCSR 8-2 443 213*

OFP337 j

P-26:

In Northwest corner of LCSR 8-2 443 214*

OFP340 M-23:

North wall of LCSR C-2 443 215*

OFP341 i

P-21:

East wall of LCSR D-2 443 216*

OFP333 i

S-21:

By cabinet 2RY01EC (elec. pen. area) 431 229 0FP454 i

S-24:

By U-2 cont., shield wall (elec pen. area)-

431 230 0FP455 S-15:

By Pzr htr. transformer (elec pen area) 431 236 0FP461 S-12:

8,y U-1 cent shield wall (elec pen area) 431 237 0FP462

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M-18:

Rad Chesioffices 430 57 0FP323 P-11:

Outside Laundry Room 430 52 0FP313 i

Q-19:

By U-2 VCT valve aisle 430 54 0FP342 i

P-24:

By radwaste evaporator 430 55,

OFP343 V-17:

By east door to decon/ change area 430 58 0FP319 V-17:

By west door to decon/ change area 430 61 0FP320 3

CFire hoses that do not supply the prinary means of fire suppression.

=-

e

. ~ _ _ _ _. _ _ _ _ _. _ _ _ _ _ _ _. _ _ _. _..

i j

TABLE 3.7-5b (Continued)

(Unit 2) j e

FIRE HOSE STATIONS 5

LOCATION ELEVATION HOSE RACK REEL ANGLE VALVE Aux. 81ds. (Continued)

V-19: By West Door to Decon Pad & Storage 430 59 0FP344 d

Q-26: Back of Div. 21 SWGR Room 430 284 0FP641 i

i S-21: By U-2 Pzr. HTR. Transformer (elec. pen. area) 419 175 0FP347 l

]

[

Q-26: By U-2 Elect. Penetration Area 419 206 0FP346 l

1 L-11: By weste oil tank room 405 90 0FP315 i

P-18:

By elevator 405 91 0FP318 i

P-23:

By spent resin pumps 405 92 0FP349 i

Q-11:

By laundry tanks 405 93 0FP314 I

S-21:

East of U-2 hydrogen recombiner 405 94 0FP348 V-21: West of U-2 hydrogen recombiner 405 SS OFP345 V-15: West of U-1 hydrogen recombiner control j

panel 405 96 0FP316 l

}

S-15:

East of U-1 t i e; n recombiner 405 97 0FP317 s

?

N-11:

By the recycle holdup tanks 368 130 0FP373 l

l j

2 M-13: By the U-1 stirs 368 131 0FP374 P-13: By panel IPL84J8 368 132 0FP369 j~

L-20:

By the U-2 stairs 368 133 0FP355 P-21: By the blowdown condenser 368 134 0FP356 i

L-25: By_the PW M/U pumps 368 135 0FP361 N-25:

By chemical drain tank 368 136 0FP357 S-18: By panel IPL86J 368 138 0FP362 I

Q-11: By Aux. 81dg. floor drain tanks 368 139 0FP368 f

Q-25: By spent resin flushing pump 368 137 0FP360 I

U-21:

By U-2 spray add tank 368 142 0FP358 i

y V-18:

By U-2 cent. chg. pump room 368 141 0FP366 i

3 P-11: c'By recycle evaporator feed pumps 350 151 0FP381 l

3 M-13: By U-1 stairs 350 152 0FP370 i

E N-23:

By gas decay tanks 350 154 0FP352

?.

Q-19:

By "B" Aux. B1dg. Equip. drain tank 350 155 0FP365 Q-17: By "A" Aux. 81dg. Equip. drain tank 350 156 0FP371.

'^

M l

~

1 ELECTRICAL POWER SYSTEMS i

j SURVEILLANCE REQUIREMENTS (Continued) i

}

3)

Verifying the diesel generator capability to re[ect a load of i

5500 kW without tripping.

The generator voltage shall not 3

exceed 4784 volts during and following the load rejection, a

i 4)

Simulating a loss of ESF bus voltage by itself, and:

J a)

Verifying de-energization of the ESF busses and load l

shedding from the ESF busses, and l

b)

Verifying the diesel starts on the auto-start signal,

~

energizes the ESF busses with permanently connected loads within 10 seconds,. energizes the auto-connected safe shutdown loads through the load sequencing timer and

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operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads.

After I

energization, the steady-state voltage and frequency of the ESF busses shall be maintained at 4160 + 420 volts and 60 + 1.2 Hz during this test.

~

I 5)

Verifying that on ~an ESF Actuation test signal without loss 2

i of ESF bus voltages, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes.

Tl)e generator voltage and frequency shall be 4160 + 420 volts and 60 + 1.2 Hz within 10 seconds after the 4

auto-start signal; the generator stea @ state generator voltage and frequency shall be maintained within these limits during this test; I

6)

Simulating a loss of ESF bus voltage in conjunction with an ESF Actuation test signal, and a)

Verifying deenergization of the ESF busses and load j

shedding from the ESF busses; i,

b) i Verifhingthedieselstartsontheauto-startsignal, t

energizes the ESF busses with permanently connected loads within 10 seconds, energizes the auto-connected i

emergency (accident) loads through the LOCA sequencer and j

operates for greater than or equal to 5 minutes while l

its generator is loaded with emergency loads.

After energization, the stea @-state voltage and frequency of the ESF busses shall be maintained'at 4160 + 420 volts i

and 60 + 1.2 Hz during this test; and

~

i c) i Verifying that all automatic diesel generator. trips, i

except engine overspeed and generator differential, are i

automatically bypassed upon loss-of-voltage on the emergency bus concurrent with a Safety Injection Actuation signal.

l' i

l l

l BYR0N - UNITS 1 & 2 3/4 8-5 Amendment No. 2

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ELECTRICAL POWER SYSTEMS l

SURVEILLANCE REQUIREMENTS (Continued) i.

1 l

7), Verifying the diesel generator operates for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

i During the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of this test, the diesel generator load-ing shall be equivalent to the 2-hour rating of 6050 kW* and during j

the remaining 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> of this test, the diesel generator shall i

be loaded to greater than or equal to 5500 kW.

The generator 1

voltage and frequency shall be 4160 1 420 volts and 60 1 1.2 Hz I

within 10 seconds after the start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test.

Within 5 minutes.after completing

{

this 24-hour test, perform Specification 4.8.1.1.2f.6)b);**

8)

Verifying that the auto-connected loads to each diesel generator j

do not exceed the 2000-hour rating of 5935 kW; j

9)

Verifying the diesel generator's capability to:

j a)

Synchronize with the offsite power source while the j

generator is loaded with its emergency loads upon a i

simulated restoration of offsite power, 6

j b)

Transfer its loads to the offsite power source, and j

c)

Se restored to its standby status, c l 10)

Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection' signal I

overrides the test mode by: (1) returning the diesel generator to standby operation and (2) automatically energizing the

]

emergency loads with offsite power; i

11)

Verifying that the fuel transfer pump transfers fuel from each j

fuel storage tank to the day tank of each diesel via the installed cross-connection lines; 1

12)

Verifying that the automatic LOCA and Shutdown sequence timer 1

is OPERA 8LE with the interval between each load block within

• 10% of its design interval; and G

I i

• Instantaneous loads of 6050 kW (+0, -150) are acceptable as equivalent to the i

2-hour rating provided voltage and frequency requirements and cooling system functioning requirements are verified to be within design Ifmits at 6050 kW.

i

"*If Specification 4.8.1.1.2f.6)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test.

Instead, the diesel generator j

may be operated at 5500 kW for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or ' ntil operating temperature has u

j stabilized.

8-i g

i E

BYRON - UNITS 1 & 2 3/4 8-6 l

l u.

j l

. RADI0 ACTIVE EFFLUENTS i

1 l

LIQUID RADWASTE TREATMENT SYSTEM i

l LIMITING CONDITION FOR OPERATION i

j 3.11.1.3 The Liquid Radwaste Treatment System shall be OPERA 8LE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED i

i AREAS (see Figure 5.1-1) would exceed 0.06 area to the whole body or 0.2 aren j

to any organ in a 31-day period.

1 j

APPLICABILITY:

At all times.

i ACTION:

il With radioactive liqJid waste being discharged without treatment and a.

in excess of the above limits and any portion of the Liquid Radwaste l

Treatment System not in operation, prepare and submit to the f

Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the following information:

1 i

1.

Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability, j

2.

Action (s) taken to restore the Tnoperable equipment to OPERA 8LE i

j status, and A

l 3.

Summary description of action (s) taken to prevent a recurrence.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

i 0

)

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the 00CM when the Liquid Radwaste Treatment System is not being fully utilized.

4.11.1.3.2 The installed Liquid Radwaste Treatment System shall b.e considered j

OPERA 8LE by meeting Specifications 3.11.1.1 and 3.11.1.2.

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i BYRON - UNITS 1 & 2 3/4 11-7

RADI0 ACTIVE EFFLUENTS LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material, excluding tritium and dissolved or entrained noble gases, contained in any outside tanks shall be Ifmited to the following:

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a.

Primary Water Storage Tank

$ 2000 Curies, and b.

Outside Temporary Tank 1 10 Curies, i

APPLICA81LITY:

At all times, j

ACTION:

With the quantity of radioactive material in any of the above If sted a.

tanks exceeding the above Ifmit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the Ifmit, and describe the events leading to i

this condition in the next Semiannual Radioactive Effluent Release j

Report, pursuant to Specification 6.9.1.7.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

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SURVEILLANCE REQUIREMENTS 4

4.11.1.4 The quantity of radioactive material contaided in each of the above j

tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.

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BYRON - UNITS 1 & 2 3/4 11-8 Amendment No. 2

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ISSUANCE OF A'iE!!D'iEf!T I!0. 2 TO BYR0:1 UNIT 1 TECHNICAL SPECIFICATIONS - NPF-37 DISTRIBUTION:

Docket File 50-454

• NRC PDR*

LPDR*

PRC System

• NSIC*

PNR#5 Reading

4. Rushbrook (5)*

L. Olshan*

T.flovak*

S. Lewis, OELD*

OELD*

CMiles, PA HDenton R. Diggs, LFriB JPartlow*

B. Grimes

• EJordan*

L. Hamon*

V. Benaroya*

R. Perch

• TBarnhart(4)*

I. Bailey

• tlith Tech. Specs.

e