Text

Proposed Tech Specs Sections 1,3/4.3,3/4.11,5,6 & Bases for Section 3/4.11
	ML20045E993
Person / Time
Site:	Byron, Braidwood
Issue date:	06/29/1993
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20045E988	List: ML20045E987; ML20045E993;
References
	NUDOCS 9307060282
	Download: ML20045E993 (66)
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.

ATTACHMENT B PROPOSED CHANGES TO APPENDIX A,'

TECHNICAL SPECIFICATIONS, OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, and NPF-77 Revision to:

I II XVIII XX 1-2 1-2a 1-3 1-3a 1-4 1-5 1-5a 1-6 3/4 3-40 3/4 3-41 3/4 11-1

]

3/4 11-3 J

B 3/411 B 3/411-2 5-1 3

5-2 5-4 6-18 6-19 6-20 6-21 6-22 6-24 6-25 i

6-26 6-27 9307060202 930629 a.

PDR ADOCX 05000454 fJ' P

PDR Q

A:nta:zionrtrtsc2.:15

i INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS 1.1 ACTI0N........................................................

1-1 1.2 ACTUATION LOGIC TEST..........................................

1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST...............................

1-1

1. 4 AXIAL FLUX DIFFERENCE.........................................

1-1 1.5 CHANNEL CALIBRATION...........................................

1-1

1. 6 CHANNEL CHECK.................................................

1-1

1. 7 CONTAINMENT IN1EGRI 2cs.v=% -,mm 1-2 s.7. a 1.8 CONTROLLED LEAMGE.. con T RO LLO A R E A.

,, 1-2

.....r..... w e.,......

1-2 1.9 CORE ALTERATION..............................................

1-2 1.9.a CRITICALITY ANLLYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS.,................ m.....u...

1-2

1. 3. s,

DIGITAL CHANNEL OPERATIONALIE$

......... i.2 1-23 T'

Am 1.10 DOSEEQUIVALENTI-131....h[1 l'

1-2a l

tJ "

1.11 1.12 E-AVERAGE DISINTEGRATION ENERGY..............................

1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME.....................

1-3 1.14 FREQUENCY NOTATION.

.... emm..

1-3

1. M

+

w. a,A m A w m sun.........

.to 1.15 IDENTIFIED LEAKAGE.r....... h reer.s m e.

1-3 1.16 MASTER RELAY TEST............................................

1-3 1.17 MEMBER [S-) 0F THE PUBLIC......................................

1-39 1.18 0FFSITE DOSE CALCULATION MANUAL..............................

1-4 1.19 OPERABLE - OPERABILITY.......................................

1-4 1.19.a OPERATING LIMITS REP 0RT.....................................

1-4 l

1.20 OPERATIONAL MODE - H0DE......................................

1-4 1.21 PHYSICS TESTS................................................

1-4 1.22 PRESSURE BOUNDARY LEAKAGE....................................

1-4 1.23 PROCESS CONTROL PR0 GRAM......................................

1-5 l

1.24 PURGE - PURGING..............................................

1-5 1.25 QUADRANT POWER TILT RATI0....................................

1-5 1.26 RATED THERMAL P0WER..........................................

1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME............................

1-5 1.28 REPORTABLE EVENT.............................................

1-5

~

BYRON - UNITS 1 & 2 I

AMENDMENT NO. 46

DEFINITIONS SECTION PAGE

-v Cl. 2.8.=R e snic A te A.

-s

.. - ~

~ -

1.29 SHUTDOWN MARGIN....................................... 7.....

1-59 1.30 SITE B00NDARY................................................

1-6 1.31 SLAVE RELAY TEST.............................................

1-6 1.32 DELETED......................................................

1-6 1.33 SOURCE CHECK.................................................

1-6 1.34 STAGGERED TEST BASIS.........................................

1-6 1.35 THERMAL P0WER................................................

1-6 1.36 TRIP ACTUATING DEVICE OPERATIONAL TEST.......................

1-6 1.37 UNIDENTIFIED LEAKAGE.........................................

1-6 1.38 UNRESTRICTED AREA............................................

1-6 1.39 VENTILATION EXHAUST TREATMENT SYSTEM.........................

1-7 1.40 VENTING......................................................

1-7 1.41 WASTE GAS HOLDUP SYSTEM......................................

1-7 TABLE 1.1 FREQUENCY N0TATION......................................

1-8 TABLE 1.2 OPERATIONAL M0 DES.......................................

1-9 BYRON - UNITS 1 & 2 II AMENDMENT NO. '46,

DESIGN FEATURES SECTION PAGE 5.1 SITE D E L. E'T E D 5.1.1 4 EXCLUSION - AREA..............................................

5-1 5.1.2 LOW POPULATION 20NE............. g.g.g...................

5-1 5.1.3 MAPDEFININGUNRESTRICTEDAREAhAANDSITEBOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS..............

5-1 FIGURE 5.1-1 fXCLUSIEo m,n, w >kAREA-AND-UNRESTRICTED-AREA-FOR aws c RADI0 ACTIVE-GASEOUS-AND-LIQUID-EFFLUENTS.............

5-2 FIGURE 5.1-2 LOW POPULATION 20NE..................................

5-3 5.2 CONTAINMENT 5.2.1 CONFIGURATION...............................................

5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE'.............................

5-1 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES.............................................

5-4 5.3.2 CONTRO L ROD ASSEMB LI ES......................................

5-4 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE.............................

5-4 5.4.2 V0LUME......................................................

5-4 bttcTed

5. 5 METEOROLOGICAL-TOWER-LOCATION.................................

5-4 5.6 FUEL STORAGE 5.6.1 CRITICALITY.................................................

5-5 5.6.2 DRAINAGE....................................................

5-5 5.6.3 CAPACITY....................................................

5-5 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT...........................

5-5 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS..................

5-6 i

j BYRON - UNITS 1 & 2 XVIII AMENDMENT NO. 46, 1

l l

1 ADMINISTRATIVE CONTROLS SECTION PAGE 6.7 S AF ETY L I M I T V I O LAT I O N........................................

6-15 6.8 PROCEDURES AND PR0 GRAMS.......................................

6-16 J

t 6.9 REPORTING REQUIREMENTS........................................

6-20 6.9.1 ROUTINE REP 0RTS.............................................

6-20 Startup Report..............................................

6-20 i

AnnualIku.ueseports..............................................

6-20 i

Annual Radiological Environmental Operating Report..........

6-22

)

-Semiannual Radioactive Effluent Release Report..............

6-22 Monthly Operating Report....................................

6-22 Operating Limits Report.....................................

6-22 Criticality Analysis of Byron and Braidwood I

Station Fuel Storage Racks................................

6-23 6.9.2 SPECIAL REPORTS..........

6-23 6.10 RECORD RETENTION.............................................

6-23 I

6.11 RADIATION PROTECTION PR0 GRAM.................................

6-24 6.12 HIGH RADIATION AREA..........................................

6-25 6.13 PROCESS CONTROL PROGRAM (PCP)................................

6-26 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM).......................

6-26 BYRON - UNITS 1 & 2 XX AMENDMENT NO. 50

DEFINITIONS t

CONTAINMENT INTEGRITY

1. 7 CONTAINMENT INTEGRITY shall exist when:

All penetrations required to be closed during accident conditions a.

are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.

b.

All equipment hatches are closed and' sealed, Each air lock is in compliance with the requirements of jM c.

Specification 3.6.1.3, CoNTRotted MEA dum b.

The containment leakage rates are within the limits of Specification d.

c 3.6.1.2, and

{ ru.- Imut The sealing mechanism associated with each penetration (e.g., welds,

'A" e.

bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE

1. 8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION

1. 9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS 1.9.a The CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS, is a document that provides the maximum allowable fuel enrichment for storage.

These limits shall be determined and submitted in accordance with Specification 6.9.1.10.

Plant operation within these limits is addressed in individual Specifications.

move DIGITAL CHANNEL OPERATIONAL TEST 1.10 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digital 1-2. a computer hardware using data base manipulation and injecting simulated process (data to verify OPERABILITY of alarm and/or trip functions.

BYRON - UNITS 112 1-2 AMENDMENT NO. 2R

DEFINITIONS T - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include di.esel generator starting and sequence loading delays where applicable.

FRE00ENCY NOTATION I

g 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

gARF A IDENTIFIED LEAKAGE f, dEWh y

l 1.15 IDENTIFIED LEAKAGE shall be:

\\

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAVAGE, or c.

Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each. relay.

The MASTER _ RELAY TEST shall include a continuity check of each associated slave relay.

MEMBER (5) 0F THE PUBLIC dckt 1.17 MEMBER OF THE PUBLI shall include il persons wh re not qccupationallykssociatedwithgtheplant.

Ttlis category do not include employees of the\\ licensee, its contractors or bendors and persqns who enter thesitetoservibeequipmentor\\((omakedelivei%es.

This cate cry does s

inctyde persons who\\use portions o the site for recreational, o upational, s

1er other purposes not, associated with the plant.

\\

BYRON - UNITS 1 & 2 1-3 A

0""

}yiced N

ge 1-h Rs Shown iw InSed 3'

l Insert "A" Insert on page 1-2 before CONTROLLED LEAKAGE CONTROLLED AREA 1.7.a The CONTROLLED AREA shall be an area, outside of a RESTRICTED AREA but inside the SITE BOUNDARY, access to t

which can be limited by the licensee for any reason.

Insert on top of page 1-2a DEEP DOSE EOUIVALENT 1.9.b DEEP DOSE EQUIVALENT, which applies to external whole-body exposure, shall be the DOSE EQUIVALENT at a 2

tissue depth of 1 cm (1000 mg/cm),

Insert on page 1-2a after DIGITAL CHANNEL OPERATIONAL TEST I

DOSE EOUIVALENT 1.10.b DOSE EQUIVALENT shall be the product of the absorbed j

dose in tissue, quality factor, and all other necessary modifying factors at the location of interest.

The unit of E

DOSE EQUIVALENT is the rem.

i Insert on p.

1-3 before IDENTIFIED LEAKAGE HIGH RADIATION AREA 1.14.a A HIGH RADIATION AREA shall be an area, accessible i

to individuals, in which radiation levels could result in an individual receiving a DOSE EQUIVALENT in excess of 100 mrem t

in one hour at 30 cm from the radiation source or from any surface that the radiation penetrates.

i Insert on page 1-5 before SHUTDOWN MARGIN RESTRICTED AREA 1.28.a A RESTRICTED AREA shall be an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials.

RESTRICTED AREAS do not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a RESTRICTED AREA.

i

DEFINITIONS 1

--+

DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

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B 1

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J BYRON - UNITS 1E2.

1-2a AMENDMENT NO. $

4 Insert "B"

l DEFINITIONS

?

MEMBER OF THE PUBLIC 1.17 A MEMBER OF THE PUBLIC shall be an. individual.in a l

CONTROLLED or UNRESTRICTED AREA.

An individual is not a l

MEMBER OF THE PUBLIC during any period in which the individual receives an occupational dose.

i l

.i 1

BYRON - UNITS 1 &2 1-3a AMENDMENT NO.

l v

-p i

DEFINITIONS 0FFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs l

required by Sections 6.8.4e and f, and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.7.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its cpecified function (s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perfone its function (s) are also capable of performing their related support function (s).

OPERATING LIMITS REPORT 1.19.a The OPERATING LIMITS REPORT is the unit-specific document that provides l

operating limits for the current operating reload cycle. These cycle-specific operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9.

Plant operation within these operating limits is addressed in individual specifications.

OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

BYRON - UNITS 1 & 2 1-4 AMENDMENT NO. M

DEFINITIONS PROCESS CONTROL PROGRAM j

1.23 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.24 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.25 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.26 RATED THERMAL POWER shall be a total core heat transfer rate to the reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from Cy 4 when the monitored parameter exceeds its Trip Setpoint at the channel sensor

/m"en e t> until loss of stationary gripper coil voltage.

F AG A 44 REPORTABLE EVENT sk-a wg w 1.28 A REPORTABLE EVENT shall be any of those conditions specified in

~

.Section 50.73 of 10 CFR Part 50.

t SHUTDOWN MAP, GIN mut b nu 1.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be suberitical from its present condition P7 M assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity (worth which is assumed to be fully withdrawn.

BYRON - UNITS 1 & 2 1-5 AMENDMENT NO. 46

DEFINITIONS SITE BOUNDARY 1.30 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

SLAVE RELAY TEST 1.31 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay.

The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION 1.32 Deleted SOURCE CHECK 1.33 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.34 A STAGGERED TEST BASIS shall consist of:

A test schedule for systems, subsystems, trains, or other designated a.

components obtained by dividing the specified test interval into n equal subintervals, and b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.35 THERMAL POWER shall be the total core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.36 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.

UNIDENTIFIED LEAKAGE 1.37 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

g%E UNRESTRICTED AREA L/~~)

1.3B An UNRESTRICTED REA shall be any area,at-ot'beyond-the-SITE-BOUNDARY-access to which is not controlled by the licenseeAo& purposes-of-protection

-of-individuals-from-exposure-to-radiation-and-radioactive-materials, or-any-

-a rea - wi thi n - the -SITE-BOUNDARY-us ed - fo r-res i denti al-quarters -or-for-i ndus tri aly

-commercialrinstitutionalrand/or-recreational-purposes.

BYRON - UNITS 1 & 2 1-6 AMENDMENT NO. '46.

TABLE 3.3-6 E

RAJIATIONMONITORINGINSTRUMENTATIONFORPLANTOPERATIONS C*U HINIMUM

[

FUNCTIONAL UNIT CilANNELS CHANNELS APPLICABLE ALARM / TRIP TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION

[

1.

Fuel Building Isolation-Radioactivity-High and Criticality (ORE-AR055/56) 1 2

<5 mR/h,,

2.

Containment Isolation-29 Containment Radioactivity-liigh a) Unit 1 (IRE-AR011/12) 1 2

All

-** 5 'oo *Elb 44 b) Unit 2 (2RE-AR011/12) 1 2

All

+* s in mW*26 26 R

3.

Gaseous Radioactivity-RCS Leakage Detection T

a) Unit 1 (IRE-PR0118)

N.A.

I 1, 2, 3, 4 N.A.

28 b) Unit 2 (2RE-PR0118)

N. A.

1 1,2,3,4 N.A.

28 4.

Particulate Radioactivity-RCS Leakage Detection a) Unit 1 (1RE-PR011A)

N.A.

I 1,2,3,4 N.A.

28 b) Unit 2 (2RE-PR011A)

N.A.

I 1, 2, 3, 4 N.A.

28 5.

Main Control Room Isolation-Outside Air Intake-Gaseous Radioactivity-High a) Train A (ORE-PR0318/328) 1 2

All

< 2 mR/h 27 b) Train B (ORE-PR033B/348) 1 2

All 32mR/h 27 m

5

TABLE NOTATIONS

With new fuel or irradiated fuel in the fuel storage areas or fuel building.

-** T ri p-S etpo i nt-is-to-be-e s tabl i s hed d uc h-th a t-t he-actu a l-s ubme rs i on-dos e-ra t e

-would-not-exceed-20-mR/hr-in-the-containment building for-containment-purge-

-omvent-the-Setpoint-value-may-be-increased-up-to-twice-the maximum-concentra-

-ti o n-a cti v i ty-i n -the-con ta i nme nt-de t e rmi ned-by-the-s amp l e - ana lysi s-pe r formed-

-p ri e r to-e a c h-rel e a s e-i n-a cc o rdanc e-wi th -Tabl e-4 r11 p rovi ded -t he-v al ue-doe

-not-ex c eed-10%-o f-the-equiva l e nt-l i mi ts-o f-Spec i f i cat i o n -3 r11r2rir a -i n-a c cord--

-ance-with-the-methodology-and-parameters-in-the-CDCM.

Tr;p sd pidl aI e rdWd bp*dy dtm/^3,. I b

d e ne - me u cack a

ACTION STATEMENTS in p, q; u ; J Ae r e mi yw ACTION 26 With less than the Minimum Channels OPERABLE requirement operation may continue provided the containment purge valves are ma,intained closed.

ACTION 27 With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months switch to the redundant train of Control Room Ventilation, provided the redundant train meets the Minimum Channels OPERABLE requirement or isolate the Control Room Ventilation System and initiate operation of the Control Room Make up System.

Restore the inoperable monitors to OPERABLE status within 30 days or submit a Special Report to the Commission pursuant to Specification 6.9.2 within the following 30 days that provides the cause of the inoperability and the plans for restoration.

ACTION 28 Must satisfy the ACTION requirement for Specification 3.4.6.1.

ACTION 29 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, ACTION a. of Specification 3.9.12 I

must be satisfied.

With both channels inoperable, provide an appropriate portable continuous monitor with the same Alarm Set-point in the fuel pool area with one Fuel Handling Building Exhaust filter plenum in operation.

Otherwise satisfy ACTION b.

of Specification 3.9.12.

l BYRON - UNITS 1 & 2 3/4 3-41 AMENDMENT NO.48

3/4.11 RADI0 ACTIVE EFFLUENTS

~

3/4.11.1 LIQUID EFFLUENTS LIQUID HOLDUP TANKS i

LIMITING CONDITION FOR OPERATION

)

3.11.1.1 Deleted 3.11.1.2 Deleted j

3.11.1.3 Deleted gdedei) 3.11.1.4 The quantity of radioactive material, excluding tritium and dissolved or entrained noble gases, contained in1ny*ou side tanks shall be limited to l

.the-foWwing - less nn.< ept 4e b Cunes.

Deimary w ter Storage _ Tank

< 2000 Curicyard-a.

a

-b.

Outside-Temporary-Tank 110-Curies:

APPLICABILITY:

At all times.

ACTION:

a.

With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to l

i this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS l

l 4.11.1.4 The quantity of radioactive material contained in each of the above 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.

BYRON - UNITS 1 & 2 3/4 11-1 AMENDMENT NO. 49.

1 RADI0 ACTIVE EFFLUENTS

~

GAS DECAY TANKS t

LIMITING CONDITION FOR OPERATION I

3.11.2.6 The quantity of radioactivity contained in each gas decay tank shall be limited to less than or equal to 5x10* Curies of noble gases (considered as Xe-133 equivalent).

APPLICABILITY:

At all times.

~

ACTION:

With the quantity of radioactive material in any gas decay tank a.

exceeding the above limit, immediately suspend all additions of radioactive material to the tank and, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7.

b.

The provisions of Specification 3.0.3 are not appli. cable.

SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas decay tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank.

w 4

BYRON - UNITS 1 & 2 3/4 11-3 AMENDMENT NO. 49

1 3/4.11 RADIDACTIVE EFFLUENTS-j BASES 3/4.11.1 LIQUID EFFLUENTS l

3/4.11.1.1 DELETED

'j 3/4.11.1.2 DELETED 3/4.11.1.3 DELETED 3/4.11.1.4 LIQUID HOLDUP TANKS l

The tanks 1 bt:d in this specification include all those outdoor radwaste tanks that are.not surrounded by liners, dikes, or. walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.

Restricting the quantity of radioactive material contained in the specified tanks i

tanks'provides assurance that in the event of an uncontrolled release of the contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table M, Column 2, at the nearest potable water.

supply and the nearest surface wate supply in an UNRESTRICTED AREA.

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BYRON - UNITS 1 & 2 B 3/4 11-1 AMENDMENT NO. 44 i

l RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2 GASEOUS EFFLUENTS 3 /4.11.2.1 DELETED 3[4.11.2.2 DELETED 3/4.11.2.3 DELETED l

3/4.11.2.4 DELETED 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3/4.11.2.6 GAS DECAY TANKS The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification.

Restricting the quantity of radioactivity contained in each gas storage tankprovidesassurancethatintheeventofanuncontrolledreleaseofthe tanks contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC

-at-the-nearest-SITE-BOUNDARY will not exceed 0.5 rem.

This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5,

" Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure,"

in NUREG-0800, July 1981.

BYRON - UNITS 1 & 2 B 3/4 11-2 AMENDMENT NO. 4 j

l

5.0 DESIGN FEATUREf, 5.1 SITE

-EVEtesiGN-AREA-5.1.1 -The-Exclusion-Aree-sha44-be-et-shown-in figure-E-1 DELETet)

LOW POPULATION ZONE 5.1.2 The Low Population Zone shall be as shown in Figure 5.1-2.

,(gNTR9LLO ANA)

MAP DEFINING UNRESTRICTED AREAS ^ AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIOUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluentsy-which

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wil4-e14ow-ident4f4 cation-of-structures-and--re4 ease-points es-well--as-.

defini t i ons f-UNR ESTR ICT ED-AR EAS-wi thi n-t he-SIT E-BOUNDA RY-tha t-ere-ec ces s i ble-to-MEMB ER S-O F-T H E-PUBblC-s h ell-be-e s-s hown-i n-Fi g u re-5tl-1.

The-definition of UN REST RI CTED-AREA-us ed-i n-i mpl eme nti ng-thes e-Tec hni c al-S pecific a ti ons-hes-been

-exp a nded-ov e r-tha t-i n-10-CF R -20.-3-(a )-( 17).

T he-UN R EST R I CT E D-AR E A-b ound a ry--may-coi nc i de-w i th - the-Excl usi on-{f enc ed)-Area-bound a ryr-e s-defi ned-i n-10-GFR-100. 3 ( 0, --

-but-the-UNRE STR IGTED-AREA-does-not-i nc lude-a rea s-over-wa te r--bod i es.

The-concept-o f-UNR ESTR ICT ED-AR EA S ;-es t a b l i s hed -e t-o r-bey ond -the-SI TE-BOUNDA R Y r-i s-ut4442ed-4n-the-bi mi ti ng-Condi t ions-for-Opera tion-to-keep-l eveis-o f-rad ioactive-materia 41r-f 4e-4iquid-end-gaseous-ef-f4eents-es-low-as-is-reasonably-eehievable; pursuant-to

-1G-GR-- 50. 3f e.

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5.2 CONTAINMENT i

CONFIGURATION

5. 2.1 The containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

L a.

Nominal inside diameter = 140 feet,

[

b.

Nominal inside height = 222 feet, c.

Nominal thickness of concrete walls = 3.5 feet, d.

Nominal thickness of concrete dome = 3 feet, e.

Nominal thickness of concrete base slab = 12' feet, f.

Nominal thickness of steel liner = 0.25 inch, and 6 cubic feet.

g.

Net free volume = 2.8 x 10 DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a rnaximum internal pressure of 50 psig and a temperature of 250 F.

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DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4, except that limited substitution of fuel rods by filler rods consisting of Zircaloy-4 or stainless steel or by vacancies may be made if justified by a cycle specific reload analysis.

Each fuel rod shall have a nominal active fuel length of 144 inches.

The initial core loading shall have a maximum enrichment of less than 3.20 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading.

The enrichment of any reload fuel design shall be determined to be acceptable for storage in either the spent fuel pool or the new fuel vault.

i Such acceptance criteria shall be based on the results of the CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS.

CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies.

The full-length control rod assemblies shall contain a nominal 142 inches of absorber material.

All control rods shall be hafnium, silver-indium-cadmium, or a mixture of both types.

All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a.

In accordance with the Code requirements specified in Section 5.2 of theUFSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of 2485 psig, and c.

For a temperature of 650 F, except for the pressurizer which is 680 F.

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,257 cubic feet at a nominal T f 588.4 F.

avg tELETED 5.5 MEM GROEOG R AL-TOWE M OGATM N

-5.5.1---The meteorchrgit & tower shah-beiccated as shown on-Hgure 5.1 1.

i AMENDMENT NO. 30 BYRON - UNITS 1 & 2 5-4 1

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 2)

Identification of the procedures used to measure the values of the critical variables, 3)

Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, 4)

Procedures for the recording and management of data, 5)

Procedures defining corrective action for all off-control point chemistry conditions, and 6)

A procedure identifying: (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.

d.

Post-accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

The program shall include the following:

1)

Training of personnel, 2)

Procedures for sampling and analysis, and 3)

Provisions for maintenance of sampling and analysis equipment.

e.

Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable.

The program (1) shall be contained in the ODCM, (2) shall be implemented by station procedures, and (3) shall include remedial actions to be taken whenever the program limits are-exceeded.

The program shall include the following elements:

1)

Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, 2)

Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B. Table M, Column 2, 2

3)

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with IO-CFR-20-106-and-with the methodology and parameters in the ODCM, BYRON - UNITS 1 & 2 6-18 AMENDMENT NO. '60 3

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 4)

Limitations on the annual and quarterly doses or do:e-

-commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit-to-UNRESTRIGTED-AREAS conforming to Appendix I to 10 CFR Part 50, 5)

Determinationofcumulativeandprojecteddosecontributions from radi0 active effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days, 6)

Limitations on the operability and use of the liquid and gaseous effluent treatment. systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose er-dose-commitment-conforming to Appendix I to 10 CFR Part 50, 7)

Limitations on the dose rate resulting from radioactive material released in gaseous effluents to -areas-beyond-the-SHE BOUNDARY-conf o rmi n g-to-the-dose s-a s s oci ated-wi th-10-CFR-P art-20r

-Appendix-B-Table-H rColumn-1; membra o uv eunic, 8)

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas oeyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, 9)

Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to-areas-4eyond-the-SHE-BOUNDARY conforming to Appendix I to 10 CFR Part 50, and

10) Limitations on the annual dose-or-dose-commitment-to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

f.

Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant.

The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of i

environmental exposure pathways.

The program shall (1) be contained 1

in the ODCM, (2) confom to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1)

Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM, BYRON - UNITS 1 & 2 6-19 AMENDMENT NO. p

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 2)

A Land Use Census to ensure that changes in the use of areas -et-

-and-beyond-the-SITE-BOUNDARY are identified and that modifica-tions to the monitoring program are made if required by the results of this census, and 3)

Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the NRC Regional Office unless otherwise noted.

STARTUP REPORT

6. 9.1.1 A sumary report of plant startup and power escalation testing shali be submitted following:

thelicenseinvolvingaplannedincreaseinpowerlevel,(3}(2)amendmentto (1) receipt of an Operating License installation of fuel that has a different design or has been manufactured by a different fuel-supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

6. 9.1. 2 The Startup Report shall address each of the tests identified in the Final Safety Analysis Report FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifica-tions.

Any corrective actions that were required to obtain satisfactory, opera-tion shall also be described.

Any additional specific details required in license conditions based on other comitments shall be included in this report.

6.9.1.3 Startup Reports shall be submitted within: (1) 90 days following com-pletion of the Startup Test Program, (2) 90 days following resumption or com-mencement of comercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

If the Startup Report does not cover all three events (i.e., initial criticality completion of Startup Test Program, and resumstion or comencement of comercial operation) supplementary reports shall be su)mitted at least every 3 months until all three events have been completed.

ormw ANNUAla REPORTS G. 9. i. 4-Deatd c

6.9.1.45 Annual Reports covering the activities of the unit es-described-below for the previous calendar year,.shall be submitted prior to March 1 of each year.

The-initial-reportshaH beaubmitted-prior-to-March-1-of-the-year-

-foHowing-initial-crit 4ca14ty.

Tnt nyb s60 iuW 2 *5

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d '

BYRON - UNITS 1 & 2 6-20 AMENDMENT NO. 'StL l

ADMINISTRATIVE CONTROLS i

REPORTING REOUIREMENTS (Continued) 9-hE-Reports-requi red-on-an-annual-ba si s-shall-i ncl ude r-?W a.

Tabulation +n :n :nnu:1 bas 4 of the number of tation, utility, and other personnel (including contractors) recei ing exposures greater than 100 mremt/yrfand tt>eir associated annarok exposure according to (Eam@

wort and job functions,9 e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimetere'TLD,-or-f-ilm bad ~e measurements.

Small exposures totalling less than 20% of the individual total dese need not be accounted for.

hEEY DME \\/ In the aggregate, at least 8df the total-whole body dese-receid

[EamWGM -from-external-sources, should be. assigned to specific major work functions.

b.

The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8.

The following information shall be included:

(1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit.

Each result should include date and time of sampling and the radiciodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radiciodine isotcpe concentration in microcuries per 4

gram as a function of time for the duration of the specific activity i

above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

l

]

2T hi s-tabul a t i o n-s upplemen ts-the-requi rements-o f-620:407-o f-10 -U R-Part BYRON - UNITS 1 & 2 6-21 AMENDMENT NO. '50

ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS (Continued)

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

j 6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

The report shell include summaries, interpreta-tions, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period.

The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

SEMIANNUAt. RADI0 ACTIVE EFFLUENT RELEASE REPORT **

cAA y TheSemiennua4RadioactiveEffluentReleasefReportcoveringtheopera-6.9.1.7 tion of the unit during the previous & months-of-ope 4etion shall be submitted b A( 1 within-60-dayrreftec4anuary-1-and, July 4 of each year.

The report shall in 7 +

clude a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be (1) con-sistent with the objectives. outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

MONTHLY OPERATING REPORT 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or RCS safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555, with a copy to the Regional Administrator of the NRC Regional Office, no later than the 15th of each raonth following the calendar month covered by the report.

OPERATING LIMITS REPORT

6. 9.1. 9 Operating limits shall be established and documented in the OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.

The analytical methods used to determine the operating limits shall be those previously reviewed and approved by the NRC in Topical Reports:

1) WCAP 9272-P-A Westinghouse Reload Safety Evaluations Methodology" dated July 1985,

2) WCAP-8385 " Power Distribution Control and Load Following Procedures" dated September 1974, 3) NFSR-0016 " Benchmark of PWR Nuclear Design Methods" dated July 1983, and/or 4) NFSR-0081 " Benchmark of PWR Nuclear Design Methods Using the PHOENIX-P and ANC Computer Codes" dated July 1990.

The operating limits shall be determined so that all app)icable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The OPERATING LIMITS REPORT, including,any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

"A single suomittal may be made for a multi-unit station.

- A single submittal may be made for a multi-unit station.

The submittal should combine those sections that are comon to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

BYRON - UNITS 1 & 2 6-22 AMENDMENT NO. to h

ADMINISTRATIVE CONTROLS RECORDRETENTION,(Continur,dj roAolopulhyeshd c.

Records of radiation exposure for all individuals entering radiation

-control =r:ks; d.

Records of gaseous and liquid radioactive material released to the environs; Records of transient or operational cycles for those unit components e.

identified in Table 5.7-1; f.

Records of reactor tests and experiments; g.

Records of training and qualification for current members of the unit staff; h.

Records of in-service inspections performed pursuant to these Technical Specifications; i.

Records of Quality Assurance activities required by the QA Program; j.

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59; k.

Records of meetings and results of reviews and audits performed by the Offsite Review and Investigative Function and the Onsite Review and Investigative Function; 1.

Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.8 including the date at which the service life commences and associated installation and maintenance records; Records of secondary water sampling and water quality; m.

Records of analysis required by the Radiological Environmental n.

Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed, and o.

Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

BYRON - UNITS 1 & 2 6-24 AMENDMENT NO. 10 j

ADMINISTRATIVE CONTROLS 6_.12 HIGH RADIATION AREA g

mot 6.12.1 Pursuant to Paragraph 20.203(c)(5).of 10 CFR Part 20, in lieu of the em J

" control device" or " alarm signal" required by paragraph 20.203(c), each high HM

% radiation-area -as-4efined-in-10-CFR-Part-20r in which the intensity-of-radia-am 4 e

r J

@~'6Drce 6r ff5m any surf ace wh~Tch th,mR/he-at-45-cm-(-18-in.-)- from the radiation v

-tion is equal to or less than 1000 s

e radiation penetrates shall be barricaded pr u and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Pemit (RWP).

Individuals qua l i fi ed-i n-ra d ia ti on-p ro tec ti on-procedu re s-or-pe rsonnel-cont inuous ly-es corted-

- by-s uc h-i nd i vi dua l s-may4e-exempt-f ros--the-RWp-i s s uance-requi rement4uri ng-the-

-pe r fo ma nce-o f-thei r-a s s ig ned-duti e s-i n41gh-radiation-a rea s-wi th-expo s ure-ra tes-

-equal-to-or-less-than-1000-mR/h,-provided-they-are-otherwise-feHowing ple.

+a di a ti on-p ro tec ti on-p rocedu re s-fo r-entry-into-suc h-high-radi ati on-a rea s.

Any individual or group of individuals permitted to enter such areas shall be pm-vided with or accompanied by one or more of the following:

a.

A radiation monitoring device which continuously indicates the radiation dose rate in the area; or i

b.

A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and 4

~;

shall perform periodic radiation surveillance at the frequency specified in the Radiation Work Permit.

ffi.12. 2 In addition to the requirements of Sp'ecification 6.1'2,1, areas accessible t'o, personnel with. radiation levels greater thai l000 mR/h at 45 cm (18 in.) f' rom k

the radiation source or from any' surface which the radiation pe'netrates shall be I

s provided with locke'd doors to prev'ent unauthorizekentry, and thEskeys shall be maintainedundertheadministrativecontroloftheShiftForemanon.dutyand/or\\

s i

health ' physics supervision.

Doors sh'all remain locked except duringgeriods of j

access byspersonnel under an approved RWP which shall specify the dosi rate s

levels in'the immediate work areas and the maximum allowable stay time'for direct or remote (such as cTosed circuit TV cameras) continuous surveilla(nceind I

s i

t

]'

Nmay be made by\\ personnel qualified in radiation protection procedures to pr' yide c

positive exposure control over'the activities being performed'within the are d

Ddring emergency 'sjtuations which involve personnel injury or actions taken to s

j' pre' vent major equipment damage, continuous surveillance and radihtion monitorings of the work area by 'a qualified ind'iyidual may be substituted for't.he routine I

RWP pr'oc_edure.

'\\

\\

't

\\

1 C f Rd VJA n l g S e d "{

l BYRON - UNITS 62 6-25 AMENDMENT NO. 50.

Insert "C"

6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels greater than 1000 mrem /h at 30 cm (12 in.) from the radiation source or from any surface which the radiation penetrates shall require the following:

a.

Doors shall be locked to prevent unauthorized entry.

The keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or health physics supervision.

b.

Personnel access and exposure control requirements of activities being performed within these areas shall be specified by an approved RWP.

t c.

Each person entering the area shall be provided with an alarming radiation monitoring device that continuously integrates the radiation dose rate (such as an electronic dosimeter).

Surveillance and radiation monitoring by a radiation protection technician may be substituted for an alarming dosimeter.

d.

During emergency situations which involve personnel injury or actions taken to prevent major equipment damage, surveillance and radiation monitoring of the work area by a qualified individual may be substituted for the routine RWP procedure.

For individual high radiation areas accessible c.

to personnel with radiation levels of greater than 1000 mrem /h at 30 cm (12 in.) that are located within large areas where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual areas, then such individual areas shall be barricaded (by an object more substantial than rope), conspicuously posted, and a flashing light shall be activated as a warning device.

I l

J ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Cohtinued)

\\

Forindividualhighradia\\$onareasaccessibletopersonnelwithra ation levels of grkater than 1000 mR/h that are located'withih large areas, such as I

g

\\PWR containment, where no enclosure exists for purposes 'af locking, and where no enclosure enq be reasonab constructed round the individual area, 'that ihdividual area ghall be barri ed(byamo'resubstantia1\\obstaclethan\\ rope),

conspicuously posted, and a fla \\ g light shhl1 be activathd as a warning \\

in

\\

-\\

devn;e.

'\\

6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 Changes to the PCP:

a.

Shall be documented and records ?of reviews perfomed shall be l

retained as required by Specification 6.10.2o.

This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s)

and, 2)

A determination that the change will maintain the overall r

conformance of the solidified waste product to. existing requirements of Federal, State, or other applicable regulations.

b.

Shall become effective after review and acceptance by the Onsite Review and Investigative Function (Onsite Review) and the approval of the Station Manager.

-i t

6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) i 6.14.1 Changes to the ODCM:

i a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.20.

This documentation shall contain:

1)

Sufficient.information to support the change together with the appropriate analyses or evaluations justifying the change (s) l

and, 2)

A detemination that the change will maintain the level of radioactive effluent control-required by 10 CFR 20:-1M, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, f

dose, or setpoint' calculations.

j i

b.

Shall become effective after. review and acceptance by the Onsite' Review and Investigative Function and the approval of the Station i

Manager on the date specified by the Onsite Review and Investigative l

Function.

i i

BYRON - UNITS 1 & 2 6-26 AMENDMENT NO. pd

{

ADMINISTRATIVE CONTROLS r

OFFSITE DOSE CALCULATION MANUAL (00CM) (Continued) c.

Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the

-Seelannuah Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made effective.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indical.3 the date (e.g., month / year) the change was implemented.

O e

d

)

BYRON - UNITS 1 & 2 6-27 AMENDMENT NO. '50 l

.I N__DEX DEFINITIONS SECTION PAGE

1. 0 DEFINITIONS 1.1 ACTI0N........................................................

1-1 1.2 ACTUATION LOGIC TEST..........................................

1-1

1. 3 ANALOG CHANNEL OPERATIONAL TEST...............................

1-1 1.4 AXIAL FLUX DIFFERENCE.........................................

1-1 1.5 CHANNE L CALIB RATION...........................................

1-1 Wh

1. 6 CHANNEL CHECK..............................................a..

1-1 CONTAINMENT INTEGRI

............... C 1-2 CONTROLLED LEAKAGE.....v,ic. cut.co A c c A......D....

1.7

^

con

..........t-2.

/ 7. <.

e1.8 1-2 ww _v~

-y.

.ms 1.9 CORE ALTERATION...............................................

1-2 1.9.a CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL

(

DIGITAL CHANNEL.~ beck o..?:n.d;r@nn..nr.C7...... I - Z r%

1-2 STORAGE RACKS.i?:nr

~

ost E ur t.e n t.

l (l.q' b e

0PERATIONAL-TEST.............m..m. m _d 1-h 1.10 4 bos EG 4 I v%I.6.O T*

... p 2 0-

/. g. a 00?A EQUIVALENT I-131... %....E 1-a.

1.11 c -. m e m._. e......

1.12 E-AVERAGE DISINTEGRATION ENERGY..............................

1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME.....................

1-3 FREQUENCY NOTATION.. C.~.T.T.T.T. C. r.b...........c.

1-3

')j.l m 1.14

~

e

( Hr M o xa r ror) A R.E

.............l.3 j.g M1. 5 IDENTIFIED LEAKAGE..k..c.H 1-3 4

....... u....................

1.16 MASTER RELAY TEST............................................

1-3,

1.17 MEMB E MOF TH E P UB LIC......................................1-h 1.18 0FFSITE DOSE CALCULATION MANUAL..............................

1-4 1.19 O P E RAB LE - O P E RAB I LI TY.......................................

1-4 4

1.19. a O P E RATI NG L IM I TS R E P0 RT.....................................

1-4 2

1. '!0 OPERATIONAL MODE - M00E......................................

1-4 1.21 PHYSICS TESTS................................................

1-4 1.22 PRESSURE BOUNDARY LEAKAGE....................................

1-4 i

1.23 P ROC E S S CONT RO L P R0G RAM......................................

1-5 E>

?

1.24 PURGE - PURGING..............................................

1-5 1.25 QU ADRANT POWER TI LT RATI0....................................

1-5 1.26 RATED THERMAL P0WER..........................................

1-5 1.27 REACTOR TRIF SYSTEM RESPONSE TIME............................

1-5 1.28 REPORTABLE EVENT.............................................

1-5 BRAIDWOOD - UNITS 1 & 2 I

AMENDMENT NO. g

DEFINITIONS

~#

' ' ~' ~ ~ ~ ~ h -PAGE ECTION ~

~~~

~

1-1ser* :

1.Zf.a Acsre cTc.D u tA f.S

%IM9 ~ SHUTDOWN-MARGIN. c ver;T..... =....,w,, m m e.

w n 1

o.

1.30 SITE B0VNDARY................................................

1-6 if 1.31 SLAVE RELAY TEST.............................................

1-6 i;

1.32 DELETED......................................................

1-6 1.33 SOURCE CHECK.................................................

1-6 1.34 STAGGERED TEST BASIS.........................................

1-6 1.35 THERMAL P0WER................................................

1-6 1.36 TRIP ACTUATING DEVICE OPERATIONAL TEST.....................'..

1-6 1.37 UNIDENTIFIED LEAKAGE.........................................

1-6 1.38 UNRESTRICTED AREA............................................

1-6 1.39 VENTILATION EXHAUST TREATMENT SYSTEM.........................

1-7 1.40 VENTING......................................................

1-7 1.41 WASTE GAS HOLDUP SYSTEM......................................

1-7 TABLE 1.1 FREQUENCY N0TATION......................................

1-8 TABLE 1.2 OPERATIONAL M0 DES.......................................

1-9 Seh t

1 l

BRAIDWOOD - UNITS 1 & 2 II AMENDMENT NO. N

1 DESIGN FEATURES l

i SECTION PAGE t

5.1 SITE

- ~ ~w 5.1. r[b*~etusicN =A. ".'.W).....................

5-1

-~na LOW POPULATION Z0NE....................ed na 5.1. 2 5-1 cour u u.

5.1. 3 MAP DEFINING UNRESTRICTED AREAS /AND SITE BOUN ARY FOR RADIDAGTIVE-GASEOUS-AND-LIQUID-EFFLQENTS.,............

5-1 un r1s ne. uor no FIGURE 5.1-1 EXCLUSION-AREA-AND-UNRESTRIGTED-AREA-FDR

' G 0 ACTIVE-GASEOUS-AND-tIQUID-EFFtUENT&............

5-2 itADI

~ ~ -

FIGURE 5.1-2 LOW POPULATION Z0NE..................................

5-3 5.2 CONTAINMENT 5.2.1 CONFIGURATION...............................................

5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE.............................

5-1 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES............................................

5-4 5.3.2 CONTROL ROD ASSEMBLIES......................................

5-4 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE.............................

5-4 5.4.2 V0LU _ME.......... _..... _......................................

5-4 I

5.5\\ 44ETEOROLOGIEAL-TOWER-t00ATIOH.utcrcoj 5-4 5.6 FUEL STORAGE 5.6.1 CRITICALITY.................................................

5-5 5.6.2 DRAINAGE....................................................

5-5 5.6.3 CAP.YTTY....................................................

5-5 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT...........................

5-5 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS..................

5-6 I

m BRAIDWOOD - UNITS 1 & 2 XVIII AMENDMENT NO. %

J a

ADMINISTRATIVE CONTROLS SECTION PAGE 6.7 S AF ETY LI MI T VI O LATION........................................

6-15 6.8 PROCEOURES ANO PR0 GRAMS.......................................

6-16 6.9 REPORTING REQUIREMENTS........................................

6-20

(

6.9.1 ROUTINE REP 0RTS.............................................

6-20

}-

,, Star epgr..........................

6-20 c

(g_nnua} eportsk.........................

)

A 6-20 An.nual_ Rad ological Environmental Operating Report..........

6-22

(

-j s

Gem 4 annual Radioactive Effluent Release Report...............

6-22

(

J

~-

?

Monthip' Ope, rating Report....................................

6-22 Operating Limits Report.....................................

6-22 Criticality Analysis of Byron and Braidwood I

e Station Fuel Storage Racks................................

6-23 6.9.2 SPECIAL REP 0RTS..............................................

6 -(

1

(

6.10 RECORD RETENTION.............................................

6-23 1

(

)

6.11 RADIATION PROTECTION PR0 GRAM.................................

6-24 t

4 6.12 HIGH RADIATION AREA..........................................

6-25

(

)

6.13 PROCESS CONTROL PROGRAM (PCP)................................

6-26

.6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM).......................

6-26 i

h t

BRAIDWOOD - UNITS 1 & 2 XX AMENDMENT NO.)$f l

~E.

~ -,.

.O

)

DEFINITIONS CONTAINMENT INTEGRITY i

1. 7 CONTAINMENT INTEGRITY shall exist when:

All penetrations required to be closed during accident conditions a.

are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.

b.

All equipment hatches are closed and sealed, Each air lock is in compliance with the requirements of c.

Specification 3.6.1.3, d.

The containment leakage rates are within the limits of Specification g

3.6.1.2, and

< c u n eu.c4 6 tA e.

The sealing mechanism associated with each penetration (e.g., welds, r

g,g;;n i.u,

bellows, or 0 rings) is OPERABLE.

ONTROLLED LEAKAGE

1. 8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS

1. 9a The CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE c

RACKS, is a document that provides the maximum allowable fuel enrichment for storage. These limits shall be determined and submitted in accordance with Specification 6.9.1.10.

Plant operation within these limits is addressed in

& ksWb"0bONN0' h0llhos Y'..~

Lu V

meu OIGITAL CHANNEL OPERATIONAL TEST i' *d 1.10 C*Ak A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digital computer hardware using data base manipulation and injecting simulated process

(*t6 m.3 data to verify OPERABILITY of alarm and/or trip functions.

}1mer best. coauwr adiMie s sss g.u DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134, and 1-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in.

Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

BRAIDWOOD UNITS 1 & 2 1-2 AMENDMENT NO M e

m

t DEFINITIONS

~

Insert DEEP DOSE EOUIVALENT 1.9.b DEEP DOSE EQUIVALENT, which applies to external whole-body ggg exposure, shall be the DOSE EQUIVALENT at a tissue depth of 1 cm 2

\\(1000 mg/cm),

pge l-t x

DIGITAL CHANNEL OPERATIONAL TEST 1.10 A DIGITAL CHANNEL OPERATIONAL TEST shall....

Insert N

DOSE EOUIVALENT 1.10.a DOSE EQUIVALENT shall be the product of the absorbed dose in tissue, quality factor, and all other necessary modifying factors at the location of interest.

The unit of DOSE EQUIVALENT is the rem.

DOSE EOUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall....

~

f f

Insert on page 1-2_before CONTROLLED LEAKAGE CONTROLLED AREA 1.7.a The CONTROLLED AREA shall be an area, outside of a RESTRICTED AREA but inside the SITE BOUNDARY, access to which'can be limited by the licensee for any reason.

J'

~

t i

tlL4 DJtth \\ow 9%6 BRAIDWOOD UNITS 1 &2 1-2a AMENDMENT NO.

DEFINITIONS f

MEMBER OF THE PUBLIC 1.17 A MEMBER OF THE PUBLIC shall be an individual in a p

CONTROLLED or UNRESTRICTED AREA, An individual is not a MEMBER Y p3 OF THE PUBLIC during any period in which the individual receives an occupational dose.

f i

Insert on page 1-3 before IDENTIFIED LEAKAGE HIGH RADIATION AREA 1.14.a A HIGH RADIATION AREA shall be an area, accessible to individuals, in which radiation levels could result in an individual receiving a DOSE EQUIVALENT.in excess of 100 mrem in

+

one hour at 30 cm from the radiation source or from any surface.

that the radiation penetrate" h

e NtAOMhl94 M3 e-BRAIDWOOD UNITS 1 &2 1-3a AMENDMENT NO.

U

DEFINITIONS I - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

m

[ g qua caua m o Ac A c

nn.cw uu IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a.

Leakage (except CONTROLLED LEAXAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or c.

Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

X

-~~___x-v f MEMBER (G OF THE PUBLIC ma rwised atM4;on += new wed kw py 1-3s \\)

j 1.17 MEMBER (S) 0F THE PURLIC shall i clude all pd sons who a not-4

\\' 'dccupationallkassociatedViththeplat.

This ca gory does t include

(

employeesofthelicensee,ytscontractrsorvendo and perso s who enter the site to ser'vice equipment or to make deliveries.

This categ ry does

)

s s

( or ott}er purposes \\not associainciydepersonswhouseportinsofthes'{ plant.te

)

f d with the 3

- n

~~x

~-

~f

~

BRAIDWOOD UNITS 1 & 2 1-3 gem.sr oo.

DEFINITIONS OFFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology C

and parameters used in the calculation of offsite doses resulting from radio-(

active gaseous and liquid effluents, in the calculation of gaseous and liquid

(

effluent monitoring alarm / trip setpoints, and in the conduct of the Environ-t mental Radiological Monitoring Program.

The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs h

required by Sections 6.8.4.e and f, and (2) descriptions of the information i

cthat-should be included in the Annual Radiological Environmental Operating and Semiannuab Radioactive Effluent Release Reports required by Specification

' 6. 9 rl. 6-and ' 6. 9.1. 7.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their relate) support function (s).

OPERATING LIMITS REPORT 1.19.a The OPERATING LIMITS REPORT is the unit-specific document that provides I

operating limits for the current operating reload cycle.

These cycle-specific operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9.

Plant Operation within these operating limits is addressed in individual specifications.

OPERATIONAL H0DE - H0DE l

1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Comission.

PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube l

leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

BRAIDWOOD UNITS 1 & 2 1-4 AMENDMENT N0. Jd i

DEFINITIONS PROCESS CONTROL PROGRAM 1.23 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State i

regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.24 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

/

QUADRANT POWER TILT RATIO 1.25 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.26 RATED THERMAL POWER shall be a total core heat transfer rate to the reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT C'

-1,28 - A-REPORTABLE-EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50. " "-

(Twt resTA2cre e APeA debiN LM4

)

[T'YUTDOWN MARGIN ~ - -

1.29

b SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition t 1-5 assuming all full-length rod cluster assemblies (shutdown and control) are p3 t

fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdriewn.

l l

BRAIDWOOD UNITS 1 & 2 1-5 AMENDMENT NO. )!I l

{

\\

l DEFINITIONS s

Plewd ffe SHUTDOWN MARGIN page I-6 1.29 SHUTDOWN MARGIN shall....

i l

~~

i r

Insert on page 1-5 before SHUTDOWN MARGIN RESTRICTED AREA i

1.28.a A RESTRICTED AREA shall be an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials.

RESTRICTED AREAS do not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a RESTRICTED AREA.

F k

a f

1 i

r I

r b

6

. 1 i

w edo-a j

P%c BRAIDWOOD UNITS 1 & 2 1-Sa AMENDMENT NO.

[

a I

DEFINITIONS SITE BOUNDARY 1.30 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

SLAVE RELAY TEST 1.31 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay.

The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION g

4 1.32 Deleted j

SOURCE CHECK 1.33 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.34 A STAGGERED TEST BASIS shall consist of:

A test schedule for systems, subsystems, trains, or other designated a.

components obtained by dividing the specified test interval into n equal subintervals, and b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.35 THERMAL POWER shall be the total core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.36 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions.

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.

UNIDENTIFIED LEAKAGE 1.37 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA

..;nuIMhA w 1.38 An UNRESTRICTE AREA'shalT be any area,at-or-beyond-the-SITE-BOUNDARV-access to which is ne controlled by the licensee,for-purposes of protection-

+f-i ndiv iduals-f rom-exposruce-to-eadia t-ion-a nd-red ioact4 ve-materials, -or-any-a rea-wi thin-the-SIT E-BOUNDAR Y-esed-fo r-res iden t4 al-qu artersr-o r-for-indus tf4ab comme rci a lvi n sti tuti onalv-a nd/o r - rec re a t i ona l-pu rpo s es:

BRAIDWOOD UNITS 1 & 2 1-6 AMENDMENTNO.)$

TABLE 3.3-6 9

RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS 7

MINIMUM CHANNELS CHANNELS APPLICABLE ALARM / TRIP 5

FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION

--e

[

1.

Fuel Building Isolation-Radioactivity-High and l

<5 mR/h,,

29 Criticality (ORE-AR055/56) 1 2

m 2.

Containment Isolation-Containment Radioactivity-High

-a* * ~@" 26 a) Unit 1 (IRE-AR011/12) 1 2

All b) Unit 2 (2RE-AR011/12) 1 2

All

- See -@

26 3.

Gaseous Radioactivity-RCS Leakage Detection a) Unit 1 (1RE-PR0118)

N.A.

1 1,2,3,4 N.A.

28 b) Unit 2 (2RE-PR0118)

N.A.

I 1,2,3,4 N.A.

28 o

4.

Particulate Radioactivity-RCS Leakage Detection i

a) Unit 1 (IRE-PRO 11A)

N.A.

1 1, 2, 3, 4 N.A.

28 b) Unit 2 (2RE-PR011A)

N.A.

I 1, 2, 3, 4 N.A.

28 5.

Main Control Room Isolation-Outside Air Intake-Gaseous Radioactivity-High a) Train A (ORE-PR0318/328) 1 2

All

< 2 mR/h 27 I -7 b) Train 8 (ORE-PR0338/348) 1 2

All 32mR/h 27

,l P

t

TABLE NOTATIONS

'With new fuel or irradiated fuel in the fuel storage areas or fuel building.

% rip SetpoiWIFtii bieTsfablished~s'Uph that the'aituitliubmeVion dose rate vould not exceed 10 mR/hr\\in the containment building. For co

'orivent the Setpoint value\\may be increased up to Kwice the ma)ptainme (imum concentra-

k. ion activity in the containment detersjned by the sample analysis performed

' rior to each release in accordance with Table 4.1142 provided the value dohs p

not ' exceed 10% oY the equivalent limits hf Specification 3.11.2.1.a in accorb-lance hith the methodology andparameters in the ODCM.\\

\\

r m a *~rm 3 & u 3 p,.m d we w +-w Aw d o,. % % w;w

.a r a. wr Yr.:

vk Mf C t'

+h'

*M *'#-

L, k. w.J ie ACTION STATEMENTS With less than the Minimum Channels OPERABLE requirement, operation ACTION 26 may continue provided the conta.inment purge valves are maintained closed.

With the number of OPERABLE channels less than the Minimum ACTION 27 Channels OPERABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months switch to the redundant train of Control Room Ventilation, provided the

)

redundant train meets the Minimum Channels OPERABLE requirement

( )

or isolate the Control Room Ventilation System and initiate operation of the Control Room Make-up System.

Restore the

)

inoperable monitors to OPERABLE status within 30 days or submit a Special Report to the Commission pursuant to Specification 6.9.2 within the following 30 days that provides the cause of

(

the inoperability and the plans for restoration.

Must satisfy the ACTION requirement for Specification 3.4.6.1.

ACTION 28 With the number of OPERABLE channels one less than the Minimum ACTION 29 Channels OPERABLE requirement, ACTION a. of Specification 3.9.12

~

r,ust be satisfied. With both channels inoperable, provide an appropriate portable continuous monitor with the same Alarm Set-point in the fuel pool area with one Fuel Handling Building t

Exhaust filter plenum in operation.

Otherwise satisfy ACTION b.

l of Specification 3.9.12.

l c

C

(

BRAIDWOOD - UNITS 1 & 2 3/4 3-41 AMENDMENT NO. )7' l

1

3/4.11 RADI0 ACTIVE EFFLUENTS l

3/4.11.1 LIQUID EFFLUENTS LIQUID HOLOUP TANXS LIMITING CONDITION FOR OPERATION 3.11.1.1 Deleted 3.11.1.2 Deleted 3.11.1.3 Deleted

<~~~%

(a b*b or entrained noble gases, contained in.any6;L%ra 3.11.1.4 The quantity of radioactive materia xcluding tritium and dissolved tside tanks shall be limited to

-the-foMewing+-

11. % Au ept

+e to c meh s.

er

-a.

Primary-Water-Storage-Tank c 2000-Cur 4es -and-r t.

Outside-Temporary-Tank 104ur4est APPLICABILITY:

At all times.

ACTION:

With the quantity of radioactive material in any of the above listed a.

tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit (and' describe the events leading to this condition in the next -Semiennual Radioactive Effluent Release Report, pursuant to Specification 6.9.f.7.

W W' b.

The provisions of Specification 3.0.3 are not' applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive matsrial contained in each of the above 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.

BRAIDWOOD - UNITS 1 & 2 3/4 11-1 AMENDMENT NO. g

RADI0 ACTIVE EFFLUENTS GAS DECAY TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas decay tank shall be limited to less than or equal to 5x104 Curies of noble gases (considered as Xe-133 equivalent).

APPLICABILITY: At all times.

ACTION:

With the quantity of radioactive material in any gas decay tank a.

exceeding the above limit, immediately suspend all additions of radioactive material to the tank and, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the tank contents to within the limit ~; ind' describe the events leading to this condition in the next Semiannual Ridioactive Effluent Release Report, pursuant to Specification 6.9.1.7.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas decay tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank.

b BRAIDWOOD - UNITS 1 & 2 3/4 11-3 AMENDMENT NO. 38'

3/4.11 RADI0 ACTIVE EFFLUENTS

- BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1, DELETED 2

3/4.11.1.2 DELETED b

3/4.11.1.3 DELETED 3/4.11.1.4 LIQUID HOLDUP TANKS The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting conctntrations would be less than the limits of 10CFRPart20,AppendixB,Tabf'aterl,s'plyinanUNRESTRICTEDAREA.

e -If Column 2, at the nearest potable water supply and the nearest surface w e

BRAIDWOOD - UNITS 1 & 2 B 3/4 11-1 AMENDMENTND.J!f -

3/4.11 RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2 GASEOUS EFFLUENTS

/

3/4.11.2.1 DELETED

(

3/4.11.2.2 DELETED C

L 3/4.11.2.3 DELETED g

3/4.11.2.4 DELETED Y

3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits.

These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3/4.11.2.6 GAS DECAY TANKS The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification.

P stricting the quantity of radioactivity contained in each gas storage tank _pa>vides-assurance ~tttAt in the event of an uncontrolled release of the

-tints' contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC f

[et-the-nearest-SITE-BOUNDAR% wjl1 not exceed 0.5 rem. This is consistent wi_th Standard _ Review _P.lan 11.3i Branch Technical Position ETSB 11-5, i

"Poituin el Radioactive Releases Due to a Waste Gas System Leak or Failure,"

in NUREG-0800, July 1981.

L I

(

i(

1 BRAIDWOOD - UNITS 1 & 2 B 3/4 11-2 AMENDMENTNO.jHI

5.0 DESIGN FEATURES 5.1 SITE

-EXCLUSION "'EA 5.1.1 Jhe-Excimon-Area 4ha14-be-as-shown-in Figure 5. I 1.

beoeTcD LOW POPULATION ZONE 5.1. 2 The Low Population Zon -shall'be~af ihown in Figure 5.1-2.

cc.n c ou.c o a c e 4, ')

MAP DEFINING UNRESTRICTED AR_ S,^AND SITE BOUNDAR'Y FOR RADI0 ACTIVE GASE0US AND LIQUID EFFLUENTS

^%~7 5.1. 3 Information regarding radioactive gaseous and liquid effluents, ahich uill-a14ow-4dentM4 cation-of-struc4ures-and-release-points as ucll as,

-defini t i on-o f-UNR E STR ICTED-AR EA S-wi t h i n -t he-S I TE-BOUNDAR Y-tha t-a re-accessible-4e-MEMBER SW-THE40BL I C,- s ha l l-be-a s-s hown--i n4 i gu re-5-14.--The-def-in i+ ion-of-

-tHtRESTRICTED--AREA-used-in-imp 1ementing-these-Technical-Specifications-has been

-expanded-over-that-in-10-CFR-20.-3-(a)(17).-The-UNRESTRICTED-AREA-boundary may

-coincide-with-the-Exclusion -(-fenced)- Area-boundary,-as-defined-in-10-CFR-100,-3(+);-

but-the-UNR ESTR I CTEO-AR E A - does - no t-i nc l ude-a rea s -ove r-wa te r-bod i es r--The<encept-

-of-UNRESTR IOTED-AR E AS;-es t etrii s hed -a t-or-beyond-the-SITE-BOUNDARY;-i s -u t ihred-4n-the-Limiting-Conditions-for-Operation -to-keep-levels-of-radioactive-fdatef4ah

+n44 quid-and-gaseous-ef fluents-as -low-as - i s -reasonably-achievable -pursuant-to--

r ifMF R-5Dr3 &a :-For-the-B rai dwood-Sta t-ion ;-the-Exchsion-A re a-li es-withifWe-UNRESTRICTED-AREA-boundary-shcJ1 be h a+c A ir.

+he IWsik bn cdcA+i e iba l.

i 5.2 CONTAINMENT CONFIGURATION

5. 2.1 The containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

a.

Nominal inside diameter = 140 feet, b.

Nominal inside height = 222 feet, Nominal thickness of concrete walls = 3.5 feet, c.

d.

Nominal thickness of concrete dome = 3 feet, Nominal thickness of concrete base slab = 12 feet, e.

f.

Nominal thickness of steel liner = 0.25 inch, and g.

Net free volume = 2.8 x 106 cubic feet.

l DESIGN PRESSURE AND TEMPERATURE

'f 5.2.2 The containment building is designed and shall be maintained for a maximum internal pressure of 50 psig and a temperature of 250 F.

l i

BRAIDWOOD - UNITS 1 & 2 5-1 AmceJhvM.NT h)D, l

.a N

f

/

w......... -. u

'N J::

x

/

x 7

N.

< -*

1

., -r..v. m. 2 N

d -=

\\

Y:

m

./

s

x UNRESTRICTED ARE A BOUNDARY

,,eA~o s,re.ouuo.ar g

'x

,y namnen

.f/ '

, _/

QM,3 O*SE

4' a

mee e-

<9

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Wyg, 1

'* n'

,,v

_.a,,lggf;gyfs~ _ _ &

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w, s

g i

y xY,gg f'*..'**...

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u.

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N s\\

I N-Mw gi

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s

=

r --

s l **

w w-sw sw w-*wp.,,t

-a

'x*oOfL

/

coDLtv 6

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%. 4._ _... '".x* _ -- _:-

,p

.g f

g

,i

~

m taClution,aata se,i,, r,,,'

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susum unas sf asit o sat a eovmoaav

/

GAtt0Vs GF7tutwf mrtf ast PO58tf-

- s M

i j utettant etntpowo vtwf 9T ACE

/

FIGURE 5.1-1

-EXEtOSiON-AR EA-AND-UNRESTRIGTEO-AREA-

-f0R-R A010AGT4 VE -14 QUI D-EFFtuENM-(vas cusa esd>

ru v BRAIDWOOD - UNITS 1 & 2 5-2 thew trec am

tao,

OESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4, except that limited substitu-tion of fuel rods by filler rods consisting of Zircaloy-4 or stainless steel or by vacancies may be made if justified by a cycle specific reload analysis.

Each fuel rod shall have a nominal active fuel length of 144 inches.

The initial core loading shall have a maximum enrichment of less than 3.20 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading.

The enrichment of any reload fuel design shall be determined to be acceptable for storage in either the spent fuel pool or the new fuel vault.

Such acceptance criteria shall be based on the results of the CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS.

CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies.

The full-length control rod assemblies shall contain a nominal 142 inches of absorber material.

All control rods shall be hafnium, silver-l indium-cadmium, or a mixture of both types.

All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

In accordance with the Code requirements specified in Section 5.2 of a.

the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of 2485 psig, and For a temperature of 650*F, except for the pressurizer which is c.

680*F.

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,257 cubic feet at a nominal T,yg of 588.4 F.

5.5 44ETEOR0t0GICAL-TOWER-tOCAT40N-bELETED E5rl-Themeteorologi-cal-tower-shah-be-locate <f-asnhcwn en-FitJure 5.1-1.

BRAIDWOOD - UNITS 1 & 2 5-4 Amendment No. }4

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Conti med) 2)

Identification of the procedures used to measure the values of the critical variables, 3)

Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of ;ondenser in-leakage,

,)

Grocedures for the recording and management of data, frocedures defining :orrective action for all off-control point C

Aamistry conditions, and 6) i stocedure ide..tifying: (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.

d.

Post-accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and conta e x it atmosphere samples under accident conditions. The program shai! 'aclude the following:

(

1)

Training of personnel, 2)

Procedures for sampling and analysis, and 3)

Provisions for maintenance of sampling and analysis equipment.

e.

Radioactive Effluent Controls Prooram A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by station procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the followireg elements:

1)

Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, 2)

Limitations on the concentration; of radioactive material released in liquid effluents te UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B Table -(t,1 Column 2,

(")

3)

Monitoring, sampling, and analysis of Yadioactive liquid and gaseous effluents in accordance with GFR-20.-106-and-wit.h-the methodology and parameters in the ODCM, BRAIDWOOD - UNITS 1 & 2 6-18 AMENDMENT NO. JI

[

l ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 4)

Limitations on the annual and quarterly doses oe-dose-

-commitment, to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to-LWRESTRIGTE&

-AREAS conforming to Appendix I to 10 CFR Part 50, 5)

Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days, 6)

Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day peried would exceed 2 percent of the guidelines for the annual dose -or-dose-tommitment conforming to Appendix I to 10 CFR Part 50, 7)

Limitations on the dose rate resulting from radioactive material released in gaseous effluents to -ereas-beyond-the-SHE-

-BOUNDARY-con forming-to-the-dos es-es s oci a ted-wi th-10-GFR-Part-20r Appendix-8,-Table'11,-Column 1, Men cr we nea6 a

8)

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, 9)

Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit -to-areas-

-beyond-the-SHE-BOUNDARY confarming to Appendix I to 10 CFR Part 50, and

10) Limitations on the annual dose -or--dose-commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

f.

Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant.

The. program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitorirg program and modeling of environmental exposure pathways.

The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1)

Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM, BRAIDWOOD - UNITS 1 & 2 6-19 AMENDMENT NO. 7I

ADMINISTRATIVE C0NTROLS 0

PROCEDURES AND PROGRAMS (Continued) 2)

A Land Use Census to ensure that changes in the use of areas-at-

-and-beyond-the-SFTE-BOUNDAR% are identified and that modifica-tions to the monitoring program are made if required by the results of this census, and 3)

Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the NRC Regional Office unless otherwise noted.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall r

be submitted following: (1) receipt of an Operating License, (2) amendment to

('

the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

6.9.1.2 The Startup Report shall address each of the tests identified in the Final Safety Analysis Report FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifica-tions.

Any corrective actions that were required to obtain satisfactory opera-tion shall also be described.

Any additional specific details required in license conditions based on other commitments shall be included in this report.

6. 9.1. 3 Startup Reports shall be submitted within: (1) 90 days following com-pletion of the Startup Test Program, (2) 90 days following resumption or com-

~

mencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed.

Df E(ATNG ANNUAL ^ REPORTS (c. 9. 1. 4 I;M rE D p

6.9.1M5 Annual ' Reports covering the activities of the unit-as-descr4 bed below for the previous calendar year shall be submitted prior to March 1 of each year. -The-initial-reportM b 4ollowing-initial-critica14ty psubmitted-prior-to-March +of-the-yeat e &Mc A A tb ucHos TM n yi Mi AML:

BRAIDWOOD - UNITS 3 & 2 6-20 AMENDMENT NO. dif t

ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS (Continued) 0.0.1.5 Reports-required-on-an-a nnual-basis-sha44-iec4ude*

ces%- re. m )

a.

Tabulation-en er canual-basis-of the numbe tath n, utility, and w

other personnel (including contractors) re iving exposures greater E" N "\\

bece rec..

than 100 mrems/yr,and their associated - exposure according to work and job functions,+ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance i

(describe maintenance), waste processing, and refueling.

The dose assignments to various duty functions may be estimated based on pocket dosimeterf TLDr-ee-f44m-badge-measurements.

Small exposures total 4ing w) less than 20% of the individua1A stal dO;c need not be accounted for.

occe tee 5

N In the aggregate, atleast 80% of the totcl whole bcdy de:c received from-external-sourc+s should be assigned to specific major work functions.

b.

The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8.

The following information shall be included:

(1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit.

Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

l Vgn

_~

^ ^

l Ehis-tabulation-supplements-the-requirements-of-620AN-of-10 CFR Poet- !

BRAIDWOOD - UNITS 1 & 2 6-21 AMENDMENT NO. J!I l

l

ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS (Continued)

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted j

prior to May 1 of each year. The report shall include summaries, interpreta-tions, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided-shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

SEMIANNUAt-RADI0 ACTIVE EFFLUENT RELEASE REPORT **

cJends yw

6. 9.1. 7 The -Semiarnual-Radioactive Effluent Release tion of the unit during the previous months-of-ope / Report coveririg the opera-raten-shall be submitted p oc to W

Apl 1 -within-60-days-af ter-January-1-and-July of each year.

The report shall in-clude a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be (1) con-sistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

MONTHLY OPERATING REPORT

6. 9.1. 8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or RCS safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555, with a i

copy to the Regional Administrator of the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report.

OPERATING LIMITS REPORT

6. 9.1. 9 Operating limits shall be established and documented in the OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.

The analytical methods used to determine the operating limits shall be those previously reviewed and approved by the NRC in Topical Reports:

1) WCAP 9272-P-A " Westinghouse Reload Safety Evaluations Methodology" dated July'1985,

2) WCAP-8385 " Power Distribution Control and Load Following Procedures" dated

~

September 1974, 3) NFSR-0016 " Benchmark of PWR Nuclear Design Methods" dated July 1983, and/or 4) NFSR-0081 " Benchmark of PWR Nuclear Design Methods Using the PHOENIX-P and ANC Computer Codes" dated July 1990.

The operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

^A single submittal may be made for a multi unit station.

- A single submittal may be made for a multi unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

BRAIDWOOD - UNITS 1 & 2 6-22 AMENDMENTNO.JI

j ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued)

I Records of radiation exposure for all individuals entering --""'giall rad' o lo f

c.

pshJ cc. trol areas;

\\

d.

Records of gaseous and liquid radioactive material released to the

[

environs; r

Records of transient or operational cycles for those unit components e.

identified in Table 5.7-1; i

f.

Records of reactor tests and experiments; 1

g.

Records of training and qualification for current members of the unit staff; b.

Records of in-service inspections performed pursuant to these Technical Specifications; i.

Records of Quality Assurance activities required by the QA Program; j.

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59; 1

k.

Records of meetings and results of reviews and audits performed by I

g the Offsite Review and Investigative Function and the Onsite Review r

and Investigative Function, 1.

Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.8 including the date at which the service life commences and associated installation and maintenance 4

records; Records of secondary water sampling and water quality; l

m.

Records of analysis required by the Radiological Environmental' l

n.

Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date.

This should include procedures effective at specified times and QA records showing that these procedures were followed, and o.

Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

i 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent

[

with the requirements of 10 CFR Part 20 and shall be approved, maintained and i

adhered to for all operations involving personnel radiation exposure.

}

l i

i BRAIDWD0D - UNITS 1 & 2 6-24 AMENDMENT NO. //

f c

r ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA Wol LO T^)

6.12.1 Pursuant to Paragraph 20.403(c}{&} of 10 CFR Part 20,k n lieu of the i

%,g " control device" or " alarm signal" required by paragraph 20.203(r), each hipHM l

e - -r-adiation-area -as-def4ned-in-10-CFR Part-2Or in which the 4ntensrity-of-radie-h rde r

N -t4en-is equal to or less than 1000 mR/hr-at-45-cm-(-18-4n-} from the radiation f

mm(e.)j source or from any surface which the radiation penetrates shall be barricadedJand ym/s J

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,/ controlled by requiring issuance of a Radiation Work Permit (RWP). 4ndiv44uals-

' qualified-in-radiation-protection-procedures-cr-personnel-cont 4nuously-escoeted-

-by-such-individuals-may-be~ exempt-from-the-RWP-issuance-requirement-during the

-pe r f o rma nce-o f-t he i r-a s s i gned-du t-i es-i n-h ig h-rad i a t i on-a reas-with-exposuee-cetes-

-equa l - to-o r-les s-tha n-1000-mR/h,-p rov ided-they-a re-otherwise-f o14 ewing-plant

-radiation-protection-procedures-for-entry-into-such-high-radiation-aeeas-Any individual or group of individuals permitted to enter such areas shall be pro-vided with or accompanied by one or more of the following:

A radiation monitoring device which continuously indicates the a.

radiation dose rate in the area; or b.

A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the Radiation Work Permit.

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Sa iwh rne. d4v Pay

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\\6.12.2 In ddition to he requirerdents of Spe ification 6.12.1, areas ccessible to personnel with radia 'on levels greater than 1000 mR/h a 45 cm (18 'n.) from the radiatio source or rom any sur{ ace which t e radiatio penetrates shall be provided with locked door,s to prevent unauthoriz d entry, an,d the keys spall be malintained un. r the admi'gistrative c'ontrol of t e Shift Foreman on duty \\and/or health physics \\supervisiod, Doors shall remain ihcked except acc'ess by personnel under an approved RWP which sh'all specify\\during peri'pds o the dose rate levdis in the immediate work areas and 'the maximum \\ allowable s'tay time for\\

individuals in that area.

In lieu of the stay time \\ specification of the RWP, direc'torremote'(suchascibsedcircuitiTVcameras)'ectionproce\\surveillanc'e continuous may bd made by pehsonnel qualified in radiation prot dures to pr6 vide s

positike exposure ' control over the activities being performed within the are'a.

During ' emergency situations which involve' personnel ihjury or actyons taken to prevent'majorequipmentdamagelcontinuous\\surveillanceandradiationmonitor"ng t

of the wkrk area by'a qualified individual (nay be subskituted for the routine RWP procetture.

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BRAIDWOOD - UNITS 1 & 2 6-25 AMENDMENT NO. J5f

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ADMINISTRATIVE CONTROLS

-HIGH-R ADI AT40N-AREA-(Eentintred)-

TMs sum

<t e. nee eA ntu d l-Set- - E { L i~9 iwd FM C-For individEl high radiVtion areas gccessibie th personnel ith radiati evels of gr' ater than 1000 mR/h thatgare locatedgwithin lar areas, su as e

R containmegt, where no%nclosure exists for purposes of lo ing, and w ere s

no enclosure can be reasonably constru ed around he individ 1 area,thah conspicuously p\\shall be ba ricaded (

injvidualarea a more subs antial obst cle than to e),

otted, and a lashing li t shall be ctivated a a warning devibg.

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6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 Changes to the PCP:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2o.

This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s)

and, 2)

A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

k b.

Shall become effective after review and acceptance by the Onsite Review and Investigative Function (Onsite Review) and the approval of the Station Manager.

6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 Changes to the ODCM:

a.

Shall be documented and records of reviews performed shall be-retained as required by Specification 6.10.20.

This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s)

and, 2)

A determination that the change will maintain the levelvof radioactive effluent control required by 10 CFR 20-196, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFRdart'50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b.

Shall become effective after rr<iew and acceptance by the Onsite Review and Investigative Function and the approval of the Station Manager on the date specified by the Onsite Review and Investigative Function.

BRAIDWOOD - UNITS 1 & 2 6-26 AMENDMENTNO.yI

I l

ADMINISTRATIVE CONTROLS Insert in place of 6.12.2 6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels

. greater than 1000 mrem /hr at 30 cm (12 in.) from the radiation source or from any surface which the radiation penetrates shall require the following:

a.

Doors shall be locked to prevent unauthorftzed entry.

The keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or health physics supervision.

b.

Personnel access and exposure control requirements of activities being performed within these areas shall be specified by an approved RWP.

c.

Each person entering the area shall be provided with an alarming radiation monitoring device that continuously integrates the radiation dose rate (such as an electronic dosimeter).

Surveillance and radiation monitoring by a radiation protection technician may be substituted for an alarming dosimeter.

d.

During' emergency situations which involve 4

personnel injury or actions taken to prevent major equipment damage, surveillance and radiation monitoring of the work area by a qualified individual may be substituted for the routine RWP procedure.

e.

For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mrem /hr at 30 cm (12 in.) that are located within large areas where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual areas, then such individual areas shall be barricaded (by an obstacle more substantial than rope), conspicuously posted, and a flashing light shall be activated as a warning device, t

b.

ADMINISTRATIVE CONTROLS

' k' I

0FFSITE DOSE CALCULATION MANUAL (ODCM) (Continued)

Shall be submitted to the Commission in the form ofLa complete, c.

[ legible copy of the entire ODCM as a part of or concurrent with the5::

i vepott in which any change to the ODCM was made effective.

Each change shall be identified by markings in the margin of the affected-pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was t

implemented.

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i BRAIDh00D - UNITS 1 & 2 6-27 AMENDMENTNO.)HI i

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ATTACHMENT C EVALUATION OF SIGNIFICANT IIAZARDS CONSIDERATIONS Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to 10CFR50.92(c), a proposed amendment to an operating license involves no significant hazards if operation of the facility in accordance with the proposed amendment would not:

1.

Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2.

Create the possibility of a new or different kind of accident from any accident previously evaluated; or 3.

Involve a significant reduction in a margin of safety.

The proposed amendment makes several changes to Byron Station's and Braidwood Station's Technical Specifications. These changes are: (1) adding definitions for controlled area, deep dose equivalent, dose equivalent, high radiation area, and restricted area; revising definitions for member of the public and unrestricted area; (2) reducing the curie content limit of the primary water storage tank and updating the radioactive effluent holdup timk limit reference; (3) relocating certain information in Section 5; (4) revising high radiation area controls; (5) extending the Radioactive Efiluent Release Report submittal frequency from semiannual to annual; and (6) revising radiation monitoring instrumentation requirements; and (7) editorial changes.

A.

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes to the definitions, radioactive effluent holdup tank limits and location of previous 10CFR20 requirements do not impact previously evaluated accidents because there is no change in the types and amounts of effluents that will be released. There will be no increase in individual or cumulative occupational radiation exposures. The proposed changes to the high radiation area controls provide more controls for enhanced exposure monitoring; they do not change the effluents or exposures.

Relocating information to the Offsite Dose Calculation Manual (ODCM) and the editorial changes are administrative in nature. The proposed changes do not reduce the requirements of any Technical Specification requirement. The changes provide consistency and improve readability. The information deleted from Section 5 is covered in more detail in the ODCM. Therefore, the level of control is maintained.

A:nts:ren:ctrtst2.:16

i Changing the frequency of submitting the Radiological Effluent Release Report i

from semiannual to annual is consistent with the revised requirements of 10CFR50.36a. The change does not adversely impact the ability to meet applicable regulatory requirements related to liquid and gaseous effluents. The report is a historical record of station effluents and has no impact on the actual release process. The NRC will continue to receive the same information, only on-a different, approved schedule.

The changes to the radiation monitor setpoints are consistent with the monitoring requirements of the applicable accident (fuel handling accident) described in Chapter 15 of the UFSAR. The setpoints and revised table note provide more specific requirements.

B.

The proposed changes do not create the possibility of a new or different kind of accident from any accident prc-iously evaluated.

The changes in radiation monitoring instrumentation requirements are more specific. Calculations were performed to determine the appropriate setpoints based on the current plant design and operation. The changes provide better control over the instrumeetation.

The remaining proposed changes have no effect on the probability of an accident.

The changes are administrative in nature and do not affect plant design or operation. There is no change to the types and amounts of effluent that will be released, nor is there an increase in individual or cumulative occupational radiation exposures.

C.

The proposed changes do not involve a significant reduction in a margin of safety.

Reducing the activity limit for the primary water storage tank from 2000 curies to 10 curies is conservative. The 10 curie limit, which would apply to all unprotected outside tanks is within the revised limits in 10CFR20. Compliance with the limits of 10CFR20.1301 (revised) will be demonstrated by operating within the limits of 10CFR50 Appendix I, and 40CFR190.

The proposed changes to the radiation monitoring instrumentation provide additional controls over the current requirements. The instrumentation continues to meet the requirements described in the bases. The bases themselves are unchanged.

The remaining changes are editorial and have no effect on the margin of safety for any Technical Specification.

Therefore, based on the above evaluation, Commonwealth Edison has concluded that these changes do not involve significant hazards considerations.

kmla;r en.ctrtsc2.;17

ATTACHMENT D ENVIRONMENTAL ASSESSMENT Commonwealth Edison has evaluated the proposed amendment against the criteria for and identification oflicensing and regulatory actions requiring environmental t

assessment in accordance with 10CFR51.2L It has been determined that the proposed change meets the criteria for a categorical exclusion as provided for under 10CFR51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10CFR50 and the amendment meets the following specific criteria:

(i) the amendment involves no significant hazards considerations As demonstrated in Attachment C, this proposed amendment does not involve any significant hazards considerations.

(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite As documented in Attachment A, there will be no change in the types or significant increase in the amounts of any efiluents released offsite.

(iii) there is no significant increase in individual or cumulative occupational radiation exposure The proposed change will not result in c' anges in the operation or configuration of the facility. There will be no change in the level of controls or methodology used for processing of radioactive efIluents or handling of solid radioactive waste, nor will the proposal result in any change in the normal radiation levels within the plant. Therefore there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.

IC !I 3'4

ML20045E993

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