Text

Amend 10 to Licenses NPF-37 & NPF-66,respectively,revising Condition 2.C.(6) & 2.E,removing Fire Protection Tech Specs & Revising Table 4.3-1
	ML20239A598
Person / Time
Site:	Byron
Issue date:	09/09/1987
From:	Muller D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20239A599	List: ML20239A598; ML20239A604;
References
	NPF-37-A-010, NPF-66-A-010 NUDOCS 8709170512
	Download: ML20239A598 (52)
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NUCLEAR REGULATORY COMMISSION 5

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COMMONWEALTH EDISON COMPANY DOCKET N0. STN 50-454 BYRON STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 10 License No. NPF-37 1.

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

A.

The applications for amendment by Commonwealth Edison Company (thelicensee)datedAugust 29, 1986 and September 15, 1986, supple-mented March 24, 1987 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the l

Commission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; i

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and l

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements I

have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraphs i

2.C(2) and 2.C(6) of Facility Operating Licem.,e No. NPF-37 are hereby amended to read as follows:

1 8709170512 870909 PDR ADOCK 05000454 P

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(2) Technical Specifications 1

The Technical Specifications contained in Appendix A as revised through Amendment No.10 and the Environment Protection Plant co6tained in Appendix B, both of. which are attached hereto, are' hereby incorporated into this lic'ense.. The licensee shall operate the facility in accordance with the Technical Specifications and

.the Environmental Protection Plan.

i (6).The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in'the licensee's Fire Protection Report through Amendment 8 and as approved in the 3

SER dated February 1987 through Supplement No 8 subject to the j

following provision:

1 The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

3.

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

l l

FOR THE NUCLEAR REGULATORY C0 I SION ps e A & d Daniel R. Muller, Director Project Directorate III-2 I

l Division of Reactor Projects - III,

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IV, V and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: September 9, 1987 i

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'o UNITED STATES

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NOCLEAR REGULATORY COMMISSION o

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WASHINGTON, D C. 20555

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COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-455 BYRON STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 10

~

License No. NFP-66 1.

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

A.

The applications for amendment by Commonwealth Edison Company (the licensee) dated August 29, 1986 and September 15, 1986, supple-mented March 24, 1987 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the l

Commission's rules and regulations set forth in 10 CFR Chapter 1; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of. the Comissien; C.

There is reasonable assurance (1) that the activities authorized I

by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-

'tions as indicated in the attachment to this license amendment, and paragraphs 2.C(2) and 2.E of Facility Operating License No. NPF-66 are hereby amended to read as follows:

{

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(2) Technical Specifications

)

The Technical Specifications contained in Appendix (NUREG-1113),

as revised through Amendment No. 10 and revised by Attachment 2 to WPF-60, and the Environmental Protection Plan contained in i

Appendix B, both of which were attached to License No. NPF-37, l

dated February 14, 1985, are hereby incorporated into this license. contains a revision to Appendix A which is hereby in-

corporated into this license. The licensee shall operate the facility in accordance with Technical Specifications and the Environment 1

Protection Plan.

E.

The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described.in the licensee's Fire Protection Report through Amendment 8, and as approved in the SER dated February 1987 through Supplement No. 8, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those i

changes would not adversely affect the ability to achi. eve and maintain safe shutdown in the event of a fire.

3.

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

FOR THE NUCLEAR REGULATORY COMMI S N l

4

//

W Daniel R. Muller, Director Project Directorate III-2 Division of Reactor Projects - III, i

IV, V and Special Projects

Attachment:

Changes to the Technical Specifications i

Date of Issuance: September 9, 1987 l

l l

7 -

1 ATTACHMENT TO LICENSE AMENDMENT NOS. 10 AND 10 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN-50-454 AND STN 50-455 1

i Revis*e Appendix A as follows:

Remove Pages Insert Pages V*

V*

VI VI VII VII VIII*

VIII*

XI XI l

l XII*

XII*

XVII XVII XVIII*

XVIII*

3/4 3-9 3/4 3-9 3/4 3-10*

3/4 3-10*

3/4 3-11*

3/4 3-12 3/4 3-12 3/4 3-12a l

3/4 3-56 through 3/4 3-81**

3/4 3-56 through 3/4 3-73**

3/4 7-30 through 3/4 7-49***

3/4 7-30 and 3/4 7-31***

B 3/4 3-5 B 3/4 3-5 B 3/4 3-6 B 3/4 3-6 B 3/4 7-7 8 3/4 7-7 i

6-1 6-1

)

6-2 6-2 6-13 6-13 6-14*

'6-14*

6-15*

6-16 6-16 i

Overleaf pages added for convenience Pages 3/4 3-56 through 3/4 3-63 were removed.

Pages 3/4 3-64 through j

3/4 3-81 have been renumbered to replace missing pages Pages 3/4 7-30 through 3/4 7-47 were removed.

Pages 3/4 7-4B and 3/4 7-49 have been renumbered to replace missing pages.

/

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3 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE j

3/4.2 POWER DISTRIBUTION LIMITS

)

3/4.2.1 AXIAL FLUX DIFFERENCE...........'.........................

3/4 2-1 FIGURE 3,2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER................................

3/4 2-3 3/4.2.2 HEAT FLUX H0T CHANNEL FACT 0R.............................

3/4 2-4 FIGURE 3.2-2 K(Z)-NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT...

3/4 2-5 9

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT0R................................................

3/4 2-8 l

-3/4.2.4

~}UADRANT POWER TILT RATI0................................

3/4 2-10 4

l' 3/4.2.5 DNB PARAMETERS...........................................

3/4 2-13 TABLE 3.2-1 DNB. PARAMETERS........................................

3/4 2-14 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................

3/4 3-1 i

l TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION...................

3/4 3-2 k

1 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES....

3/4 3-7 j

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................................

3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM i

INSTRUMENTATION........................................

3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.....................................

3/4 3-15 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS......................

3/4 3-23 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES.............

3/4 3-30 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM I

INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........

3/4 3-34

.l BYRDN - 45 TITS 1 & 2 V

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6 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3.3 MONITORING INSTRUMENTATION j

Radiation Mcnitoring for Plant Operations................

3/4 3-39 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION

)

l FOR PLANT 0PERATIONS................................

3/4 3-40 TABLE 4.3 d RADIATION MONITORING INSTRUMENTATION I

FOR PLANT OPERATIONS SURVEILLANCE f

REQUIREMENTS........................................

3/4 3-42 l

Movable Incore Detectors.................................

3/4 3-43

{

Seismic Instrumentation..................................

3/4 3-44 I

TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION....................

3/4 3-45 Meteorological Instrumentation...........................

3/4 3-47 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION.............

3/4 3-48 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................

3/4 3-49 Remote Shutdown Instrumentation..........................

3/4 3-50 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION............

3/4 3-51 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................

3/4 3-52 Accident Monitoring Instrumentation.......................

3/4 3-53 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION..................

3/4 3-54 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................

3/4 3-55 Loose-Part Detection System..............................

3/4 3-56 Radioactive Liquid Effluent Monitoring Instrumentation...

3/4 3-57 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION....................................

3/4 3-58 BYlWN - LMIT 1 VI ATENDMENT 10.10 L

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h LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 4

I PAGE SECTION 1

i TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING

]

INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........

3/4 3-60

{

Radioactive Gaseous Effluent Monitoring Instrumentation.

3/4 3-62

]

TABLE 3.6-13 RADIOACTIVE GASE0US EFFLUENT MONITORING j

INSTRUMENTATION....................................

3/4 3-63 l

.i TABLE 4.3-9. RADI0 ACTIVE GASE0US EFFLUENT MONITORING

.j INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........

3/4 3-67 1

TABLE 3.3-14 HIGH ENERGY LINE BREAK INSTRUMENTATION................

3/4 3-72 l

3/4.3.4 TURBINE OVERSPEED PROTECTION.............................

3/4 3-73 i

]

3/4.4 REACTOR COOLANT SYSTEM 1

1 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power 0peration..............................

3/4 4-1 Hot Standby..............................................

3/4 4-2 Hot Shutdown.............................................

3/4 4-3 Cold Shutdown - Loops Fi11ed.............................

3/4 4-5 1

Cold Shutdown - Loops Not Fi11ed.........................

3/4 4-6 i

Loop Isolation Valves-0peration..........................

3/4 4-7

]

l Loop Isolation Valves-Shutdown...........................

3/4 4-8 L

3/4.4.2 SAFETY VALVES Shutdown...............................................

3/4 4-9 l

0perating..............................................

3/4 4-10 l

3/4.4.3 PRESSURIZER..............................................

3/4 4-11 3/4.4.4 RELIEF VALVES............................................

3/4 4-12 3/4.4.5 STEAM GENERATORS..................-.......................

3/4 4-13 TABLE 4.4-1 MINIMUM NUMBER OF S1EAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION.........................

3/4 4-18 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION........................

3/4 4-19 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................

3/4 4-20 Ope rati ona l Le a ka g e......................................

3/4 4-21 BVMll - WITS 1 & 2 VII NEMNENT W. 3

i.,-

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLA'NCE REQUIREMENTS

)

I SECTION PAGE i

TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES......

3/4 4+23 l

I 3/4.4.7 CHEMISTRY................................................

3/4 4-24 j

TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS...............

3/4 4-25

)

l TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE 1

i REQUIREMENTS........................................

3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY........................................

3/4 4-27 1

FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT. SPECIFIC l

I ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR'C00LANT SPECIFIC ACTIVITY

>l pCi/ GRAM DOSE EQUIVALENT I-131..................

3/4 4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM....................................

3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System...................................

3/4 4-32 FIGURE 3.4-2a REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 32.EFPY (UNIT 1)......

3/4 4-33 FIGURE 3.4-2b REACTOR COOLANT SYSTEM HEATUP

\\

LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)......

3/4 4-34 FIGURE 3.4-3a REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)......

3/4 4-35 FIGURE 3.4-30 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)......

3/4 4-36 i

TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE.......................

3/4 4-37 i

Pressurizer..............................................

3/4 4-38 l

Overpressure Protection Systems..........................

3/4 4-39 FIGURE 3.4-4 NOMINAL PORV PRESSURE RELIEF SETPOINT VERSUS i

1 RCS TEMPERATURE FOR THE COLD OVERPRESSURE' i

l PROTECTION SYSTEM APPLICABLE UP TO 10 EFPY.........

3/4 4-40 3/4.4.10 STRUCTURAL-INTEGRITY.....................................

3/4 4-42 3/4.4.11-REACTOR COOLANT SYSTEM VENTS.............................

3/4 4-43 BVRBN-45117514 2 VIII

c

+.

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7.12 AREA TEMPERATURE M0NITORING..............................

3/4 7 TABLE 3.7-6 AREA TEMPERATURE M0NITORING............................

3/4 7-31 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0perating................................................

3/4 6 l ~

TABLE 4.8-1 DIESEL' GENERATOR TEST SCHEDULE........................

3/4 8-8 Shutdown..................................................

3/4 8-9 3/4.8.2 D.C. SOURCES

-Operating................................................

3/4 8 [

. TABLE 4.8 BATTERY SURVEILLANCE REQUIREMENTS.....................

3/4.8-12 Shutdown.................................................

3/4 8-13 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating................................................

3/4 8-14 Shutdown...............

3/4 8-16 3/4.8.4-ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent-Protective Devices.....................................

3/4 8-17 TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES......................

3/4 8-19

_l BYRON - Ani1T 1 II NelDFENT NO.10 4

o LIMITING CONDITIONS FOR OPEhATION AND SURVEILLANCE REQUIREMENTS SECTION PAG _E Motor-0perated Valves Thermal Overload Protection Devices....................................

3/4 8-39 TABLE 3.8-2a MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES (UNIT 1)....................................

3/4 8-40 TABLE 3.d-2b MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES (UNIT 2)....

3/4 8-44 1

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION..

3/4 9-1 l

3/4.9.2 INSTRUMENTATION....

3/4 9-2 3/4.9.3 DECAY TIME..........................

3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................

3/4 9-4 j

3/4.9.5 COMMUNICATIONS..........................................

3/4 9-6 3/4.9.6 REFUELING MACHINE...................................

3/4 9-7 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY...............

3/4 9-8 l

3/4.9.8 RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level.......

3/4 9-9 Low Water Leve1..................................

3/4 9-10 3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM......

3/4 9-11 j

3/4.9.10 WAT ER LEVEL - REACTOR VESSE L............................

3/4 9-12 3/4.9.11 WATER LEVEL - STORAGE P00L...............................

3/4 9-13 3/4.9.12 FUEL HANDLING BUILDING EXHAUST FILTER PLENUMS............

3/4 9-14 i

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l BYRON -- UNITS 1 & 2 X11

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1 BASES SECTION PAGE l

3/4.7 PLANT SYSTEMS 3/4.7.1 TURB1NE CYCLE......................'.......................

B 3/4 7 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION...........

B 3/4 7-3 3/4.7.3 G5MPONENTCOOLINGWATERSYSTEM............................

B 3/4 7-3 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM............................

B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK........................................

B 3/4 7-3 3/4.7.6 CONTROL ROOM VENTILATION SYSTEM...........................

B 3/4 7-4.

3/4.7.7 NON-ACCESSIBLE AREA EXHAUST FILTER PLENUM VENTILATION SYSTEM.................................................

B 3/4 7-5 3/4.7.8' SNUBBERS..................................................

B 3/4 7-5 3/4.7.9 SEALED SOURCE CONTAMINATION...............................

'B 3/4 7-6 3/4.7.12 AREA TEMPERATURE M0NITORING...............................

B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION...............................

B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES...................

B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.......................................

B 3/4 9-1 3/4.9.2 INSTRUMENTATION...........................................

B 3/4 9-1 3/4.9.3 DECAY TIME................................................

B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.........................

B 3/4 9-1 3/4.9.5 COMMUNICATIONS.............-...............................

B 3/4 9-1 i

BYRON-iAET 1 XVII NENDHT SE. 3D

BASES SECTION PAGE t

3/4.9.6 REFUELING MACHINE.........................................

B 3/4 9-2 3/4.9.7 C.RANE TRAVEL - SPENT FUEL STORAGE FACILITY...........,.....

B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION.............

B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM........................

B 3/4 9-2 3/4.9.10* and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE P00L.................................

B 3/4 9-3 3/4.9.12 FUEL HANDLING BUILDING EXHAUST FILTER PLENUM SYSTEM...... B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN...........................................

B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS....

B 3/4 10-1 3/4.10.3 PHYSICS TESTS.............................................

B 3/4 10-1 3/4.10.4 REACTOR COOLANT L00PS.....................................

B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM -

SHUTD0WN.....................

B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIOUID EFFLUENTS.........................................

B 3/4 11-1 I

I 3/4.11.2 GASLOUS EFFLUENTS........................................ 'B 3/4 11-3 3/4.11.3 SOLID RADI0 ACTIVE WASTES.................................

B 3/4 11-7 3/4.11.4 TOTAL D0SE...............................................

B 3/4 11-7 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM.......................................

.B 3/4 12-1 3/4.12.2 LAND USE CENSUS..........................................

B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM.......................

B 3/4 12-2

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TABLE NOTATIONS

With the Reactor Trip System breakers closed and the Control Rod Drive System capable of rod withdrawal.

i

- These channels also provide inputs to ESFAS.

The Operational Test Frequency for these~ channels in Table 4.3-2 is more conservative and, therefore, controlling.

1. Below P-6 (Interms.diate Range Neutron Flux Interlock) Setpoint.

1. 1. Below F-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1) If not performed in previous 7 days.

l i.

1 i

(2) Comparison of calorimetric to excore power indication above 15% of RATED I

THERMAL POWER.

Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Speci-fication 4.0.4 are not applicable for entry into MODE 2 or 1.

(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER.

Recalibrates if the absolute difference is j

greater than or equal to 3%.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

j (Sa) Initial plateau curves shall be measured for each detector.

Subsequent l

l plateau curves shall be obtained, evaluated and compared to the initial I

I curves.

For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(50) With the high voltage setting varied as recommended by the manufacturer, an initial discriminator bias curve shall be measured for each detector.

Subsequent discriminator bias curves shall be obtained, evaluated and compared to the initial curves.

(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

The provi-sions of Specification 4.0.4 are not' applicable for. entry into MODE 2 or 1.

l (7) Each train shall be tested at least every 62 days on a STAGGEREO TEST BASIS.

1 (8) With power greater than or equal to the interlock Setpoint the required ANALOS CHANNEL OPERATIONAL TEST shall consist of verifying that the inter-lock is in the required state by observing the permissive annunciator window.

I (9) Surveillance in MODES 3*, 4*, and 5* shall also include verification that l

l permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

Surveil-lance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to an increase of twice the count rate within a 10-minute period.

BYRON - UNITS 1 L 2 3/4 3-12 N O NENT W. 10

TABLE 4.3-1 (Continued)

TABLE NOTATIONS (10) Setpoint verification is not applicable.

(11) At least once per 18 months and following maintenance or adjustment of the Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include ~ independent verification of the Undervoltage and Shunt trips.

(12) At least once per 18 months during shutdown verify that on a simulated Boron, Dilution Doubling test signal CVCS valves 1120 and E open and 1128,and C close within 30 seconds.

(13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.

i BYRON - ISUT51 & 2 3/4 3-12a NEIDGIT 10.19 l

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INSTRUMENTATION LOOSE-PART DETECTION SYSTEM ej LIMITING COMDITION FOR OPERATION q

./

3.3.3.8 The Loose-Part Detection System sbeli be OFERABLE.

j

~

APPLICABILITY:

MODES 1 and 2.

ACTION: *.

With one or more Loose-Part Detection System channels inoperable for ~

1 a.

more than 30 days, prepare and submit a Special Report to the l

Commission pursuant to Specification 6.9.P'within'the next 10 days outlining the cause of the malfunction and the plans for restoring

~

the channel (s) to OPERABLE status, TheprovisionsofSpecifications'3.0.3and3.0.4arenotahplicable.

b.

f r

t i

l SURVEILLANCE REQUIREMENTS i

4.3.3.8 Each channel of the Loose-Part Detection' Systems shall be demonstrated j

OPERABLE by performance of:

a.

A CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , i

b.

An ANALOG CHANNEL OPERATIONAL TEST except for verification of setpoint at least once per 31 days, and c.

A CHANNEL CALIBRATION at least once per 18 months.

+

i I

i

/

1 l

mag - asuric 2 AJ 3/4 3 Amendment No. 10 l

__.-.___m'_.

m

_____U______.______

l INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION Th' radioactive liquid effluent ponitoring instrumentation channels 3.3.3.9-e shown in Table 3.3-12 'shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits.of Specification 3.11.1.1 are not exceeded. The Alarm /

Trip Setpoints 'of these channels shall be determined and adjusted in accordance with the, methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY:

At all times.

ACTION:

a.

With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above i

specification, immediately suspend the release of radioactive liquid effluents ennitored by the affected channel, or declare the channel inoperable, b.

With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12.

Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specifica-tion 6.9.1.7 why this inoperability was not corrected within the time specified.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.9-Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and DIGITAL and ANALOG CHANNEL OPERATIONAL TEST at the fre quencies shown in Table 4.3-8.

SYa0N - edNIIS.1 Al 3/4 3-57 Amendment No. 10 l

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TABLE 3.3-12 (Continued)

ACTION STATEMENTS ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels GPERABLE requirement, effluent releases via i

'this pathway may continue for up to 14 days provided that prior t.o initiating a release:

At least two independent s'amples are analyzed in accordance l

a.

with Specification 4.11.1.1.1, ard l

l ii.

At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.

Otherwise, suspend release of radioactive effluents via this pethway.

I ACTION 32 - With the number of channels OPERABLE less than required by the

/'

Miaimum Channels OPERABLE requirement, offluent releases via this j

pathway may continue for up to 30 days provided that, at least l

once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for j

radioactivity at a lower limit of detection of no more than 10 7

~

i l

microcurie /ml.

ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathedy may continue for up to 30 days provided the flow rate is l

estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Pump i

performar,ce curves generated in place may be used to estimate l

flow.

1

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i 5

1 l

, BYRON - MMITS.1 & 2 3/4 3-59 M=rt. No. 20.

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3 TABLE 4.3-8 (Continued)

TABLE NOTATIONS (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this r .way and control room alarm annunciation occur if any of,the follow 1r., conditions exists:

Instrument indicates measured le'vels above the Alarm / Trip Setyint, a.

or

b., Circuit failure (monitor loss of communications'- alarm only, detector, loss of counts, or monitor loss of power), or c.

Detector check source test failure, or d.

Detector channel out-of-service, or e.

Monitor loss of sample flow.

(2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

a.

Instrument indicates measured levels above the Alarm Setpoint, or b.

Circuit failure (monitor loss of communications - alarm only, detector loss of counts, or monitor loss of power), or Detector check source test failure, or c.

d.

Detector channel out-of-service, or e.

Monitor loss of sample flow.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS.

These standards shall permit calibrating the system over its intended range of energy and measurement range.

For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release.

CHANNEL CHECK shall be made at'least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on-days on which continuous, periodic, or batch releases'are made.

1 SVED01 - mITEl & 2 3/4.3-51 Amendment No. 10 l

D a

INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION l

3.3.3.10 The radioactive gaseous effluent' monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specifications 3.11.2.1 and 3.11.2.5 are not exceeded.

The Alarp/ Trip Setpoints of these channels meeting Specification 3.11.2.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY:

As shown in Table 3.3-13 ACTION:

a.

With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.

b.

With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE take the ACTION shown in Table 3.3-13.

Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.7 why this inoperability was not corrected within the time specified, c.

The provisions of Specifications 3.0.3, and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, 50l!RCE CHECK, CHANNEL CALIBRATION and DIGITAL CHANNEL OPERATIONAL TEST at tne fre-quencies shown in Table 4.3-9.

BYaDN - AiETS 1 & 2 3/4 3-62 Amendment No. 10 l

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4 TABLE 3.3-13 (Continued)

TABLE NOTATIONS

At all times.

- During WASTE GAS HOLDUP SYSTEM operation.

All instruments required for Unit 1 or Unit 2 operation.

ACTION STATEMENTS ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment for up to 14 days provided that prior to initiating the release:

At least two independent samples of the tank's contents are a.

analyzed, and b.

At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 36 - With the number of channels OPERABLE less than required'by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimat?d at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.

ACTION 38 - With the number of channels OPERABLE one less -than required by the Minimum Channels OPERABLE requirement, operation of this system may continue provided grab samples are taken and a'nalyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With both channels inoperable, operation may continue provided grab samples are taken and analyzed at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing operations and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during other operations.

ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

.syaou - UNITS.1 & 2 3/4 3-65 Amendnent No.10 l

j

' '2 d

TABLE 3.3-13 (Continued)

ACTION STATEMENTS (Continued)

{

ACTION 40 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the

,affected pathway may continue for up to 30 days provided samples j

are continuously collected with auxiliary sampling equipment as-required in' Table 4.11-2.

ACTION 41 - With the number 'of channels OPERABLE less than required by the i

Minimum Channels OPERABLE requirement, effluent releases via this

}

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TABLE 4.3-9 (Continued) 1 TABLE NOTATIONS 1

At all times.

I

- During WASTE GAS HOLDUP SYSTEM operation.

All instruments required for Unit 1 or Unit 2 operation.

'(1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic 4

isolation of this pathway and control room alarm annunciation occur if any "of. the following conditions exists:

Instrument indicates measured levels above the Alarm / Trip Setpoint, a.

or b.

Circuit failure (monitor loss of communications - alarm only, detector 4

l loss of counts, or monitor loss of power), or c.

Detector check source test failure, or j

j d.

Detector channel out-of-service, or l

l e.

Monitor loss of sample flow.

(2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate tha't control l

room alarm annunciation occurs if any of the following conditions exists:

Instrument indicates measured levels above the Alarm Setpoint, or a.

b.

Circuit failure (monitor loss of communications - alarm only, detector loss of counts, or monitor loss of power), or i

c.

Detector check source test failure, or-d.

Detector channel out-of-service, or e.

Monitor loss of sample flow.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate'in measurement assurance activities with NBS.

These standards shall permit calibrating the system over its intended range of energy and measurement range.

For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing hydrogen and nitrogen.

(5) The CHANNEL CALIBRATION shall include the use of standard gas semples containing oxygen and nitrogen.

. gygen - sairTEl A 2 3/4 3-70 Amendment No. 10 l

i INSTRUMENTATION HIGH ENERGY LINE BREAK ISOLATION SENSORS i

MMITING CONDITION FOR OPERATION 3.3.3.11 The high energy line break instrumentation shown in Table 3.3-14 shall be OPERABLE.

APPLICABILITY:

As shown in Table 3.3-14.

l ACTION:

With the number of OPERABLE auxiliary steam isolation instruments less a.

than the Minimum Channels OPERABLE as required by Table 3.3-14, restore the inoperable instrument (s) to OPERABLE status within 7 days, or suspend the supply of auxiliary steam to the Auxiliary Building, or establish a continuous watch in the affected area (s) until the inoperable sensors are restored to OPERABLE status.

b.

With the number of OPERABLE steam generator blowdown line isolation instru-ments less than the Minimum Channels OPERABLE as required by Table 3.3-14, restore the inoperable instrument (s) to OPERABLE status within.7 days, or limit the total steam generator blowdown flow rate to less than or equal to 60 gpm or e:tablish a continuous watch in the affected area (s) until the inoperable sensors are restored to OPERABLE status.

The provisions of Specifications 3.0.3'and 3.0.4 are not applicable.

c.

SURVEILLANCE REQUIREMENTS 4.3.3.12 Each of the above high energy line break isolation instruments shall be demonstrated OPERABLE by the performance of an ANALOG CHANNEL OPERATIONAL TEST AND CHANNEL CALIBRATION at least once per 18 months.

I L

l 1

Embl - INTTE.1.12 3/4 3-71 NUEllENT W. W l

N,

\\

Table 3.3-14 HIGH ENERGY LINE BREAK INSTRUMENTATION Isolation Instrument Minimum Channels Applicable Function Channel OPERABLE MODES 1.

Auxiliary Steam 0TS-AS031A.

1 Isolation OTS-AS032A OTS-AS031B 1

OTS-AS032B f

OTS-AS031C 1

l OTS-A5032C OTS-AS031D 1

OTS-A50320 OTS-AS031E 1

OTS-AS032E

^

0TS-AS031F 1

OTS-AS032F 2.

Steam Generator TS-SD045A 1

1,2,3,4 Blowdown Line TS-SD045B Isolation TS-SD046A 1

1,2,3,4 TS-SD046B TS-SD045C 1

1,2,3,4 TS-SD045D TS-SD046C 1

1,2,3,4 TS-SD046D

Required when auxiliary steam is being supplied, from any source, to the i

Auxiliary Building.

l BY20N - talITS 1 &.2 3/4 3-72 N BENEWT 18. 3 l

E-_

I',

=

1 INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OPERATION 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE.

i APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

a.

With one throttle valve or one governor valve per high pressure turbine steam line inoperable and/or with one reheat stop valve or one I

reheat intercept ialve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , l

or close at least one valve in the affected steam line(s) or isolate l

l the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l l

b.

With the above required' Turbine Overspeed Protection System otherwise I

inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months isolate the turbine from the steam supply.

l I

SURVEILLANCE REQUIREMENTS i

l 4.3.4.1 The provisions of Specification 4.0.4 are not applicable.

4.3.4.2 The above required Turbine Overspeed Protection System shall be demonstrated OPERABLE:

q During turbine operation at least once per 31 days by direct obser-a.

vation of the movement of the valves below through one complete cycle from the running position:

1)

Four high pressure turbine throttle valves, 2)

Four high pressure turbine governor valves,;

3)

Six turbine reheat stop valves, and j

4)

Six turbine reheat intercept valves.

l b.

Within 7 days prior to entering MODE 3 from MODE 4, by cycling each I

of the 12 extraction steam nonreturn check valves from the closed

position, During turbine operation at least once per 31 days by direct observa-I c.

tion, of freedom of movement of each of the 12 extraction steam non-return check valve weight arms, d.

At least once per 18 months by performance of CHANNEL CALIBRATION on the Turbine Overspeed Protection Systems, and At least once per 40 months by disass.embling at least one of each of e.

the valves given in Specifications 4.3.4.2a. and b. above, and per-forming a visual and surface inspection of valve seats, disks and stems and verifying no unacceptable flaws or corrosion.

l l

svuou - UNITS 1 & 2 3/4 3-73 Amenement inn. 33 l

___.___.__.____________.__s

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a. -

I PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

I b.

Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless

)

tested within the previous 6 months.

Sealed sources'and fission

.l

~

detectors transferred without a' certificate indicating the last test l

date shall be tested prior to being-placed into use; and

c.

Startup sources and fission detectors - Each. sealed startup source i

. and fission detector shall be tested within 31 days prior to being l

subjected to core flux or installed in the core and following repair l

or maintenance to the source.

]

.4.7.9.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable-contamination.

i l

AugDN 21115.1 & 2 3/4 7-29

j

?q PLANT SYSTEMS 3/4.7.12 AREATEMPERATUREMONITORQG i

LIMITING CONDITION FOR OPERATION 3.7.12 The temperature limit of each area shown in Table 3.7-6 shall not be

. i exceeded for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or by more than 30 F.

i APPLICABIl[ITY: Whenever the equipment in an affected area is required to be OPERABLE.

I ACTION:

3 a.

With one or more areas exceeding the temperature limit (s) shown in Table 3.7-6'for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area (s) exceeded the limit (s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment.

b.

With one or more areas exceeding the temperature limit (s) shown in

~

Table 3.7-6 by more than 30'F, prepare and submit a Special Report as required by ACTION a. above, and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either return the area (s) to within the temperature limit (s) or declare the equipment in the affected area (s) inoperable.

SURVEILLANCE REQUIREMENTS 4.7.12 The temperature in each of the areas shown in Table 3.7-6 shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

._._____..-___-_________.J

o'.

li TABLE 3.7-6 AREA TEMPERATURE MONITORING TEMPERATURE AREA LIMIT (*F) 1.

Misc. Electric Equipment and Battery Rooms 108 2.

ESF Syitchgear Rooms 108 3.

Division 12 for Unit 1 (Division 22 for Unit 2) Cable Spreading Room 108 4.

Upper Cabie Spreading Rooms 90 5.

~ Diesel-Generator Rooms 132 6.

Diesel Oil Storage Rooms 132 7.

Aux. Building Vent Exhaust Filter Cubicle 122 8.

Centrifugal Charging Pump Rooms 122 9.

Containment Spray Pump Rooms 130 10.

RHR Pump Rooms 130 11.

Safety Injection Pump Room 130 12.

Control Room 90 13.

Lower Cable Spreading Rooms 108

'S E - E '1 4 7 3M 7-M ht h.10

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V S *e

't 4

d i

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f s

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INSTRUMENTATION BASES l

SEISMIC INSTRUMENTATION (Continued)

)

The response spectrum analyzer computes the response spectrum of the event for two sensor locations, compares it to the design response spectra of the I

plant, and indicates whether the event exceeded the operating basis earthquake criteria or the safe shutdown earthquake criteria.

l This instrumentation.is consistent with the recommendations of Regulatory i

Guide 1.12, " Instrumentation for Earthquake, April 1974.

]

3/4.3.3.4 METEOROLOGICAL INSTRUMENTAL'ON t

The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the~ health and safety of the gublicandisconsistentwiththerecommendationsofRegulatoryGuide1.23, Onsite Meteorological Programs," February 1972.

3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION l

The OPERABILITY of the remote shutdown instrumentation ensures that l

l sufficient capability is available to permit shutdown and maintenance of HOT l

STANDBY of the facility from locations outside of the control room.

This i

capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50.

l l

3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.

This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," May 1983 and NUREG 0737, " Clarification of TMI Action Plan Requirements," November 1980.

BYRON - UNIT 1

'B ~3/4 3-5

ENDMENT NO.10

_ _ _ _ _ _ = _ _ _ - - _ _ _ _ - _

INSTRUMENTATION BASES l

3/4.3.3.8 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part detection system ensures that sufficient capability is available to detect loose metallic parts in the Reactor Coolant System and avoid or mitigate damage to Reactor Coolant System components.

The allowable eut-of-service times and Surveillance Requirements are consistent with the recomraendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981.

3/4.3.3.9 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrua.entation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.

The Alarm /

Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part'50.

3/4.3.3.10 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceedia; the limits of 10 CFR Part 20.

3/4.3.3.11 HIGH ENERGY LINE BREAK ISOLATION SENSORS The OPERABILITY of the high energy line break isolation sensors ensures that the capability is available to promptly detect and initiate protective action in the event of a line break.

This capability is required to prevent the potential for damage to safety-related systems and structures in the auxiliary building.

BYRON - UNIT 1 B 3/4 3-6 AMENDMENT NO. ID

i PLANT SYSTEMS I

BASES l

3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium.

This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group.

Those sources which are frequently handled are required to be tested more often than those which are not.

Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

I 3/4.7.12 AREA TEMPERATURE MONITORING l

The area temperature limitations ensure that safety-related equipment I

will not be subjected to temperatures in excess of their environmental qualification temperatures.

Exposure to excessive temperatures may de' grade equipment and can cause a loss of its OPERABILITY.

r, 1

BYRON - ASIITS 1 & 2 B 3/4 7-7 AMENDMENT ND. 10-t

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i ADMINISTRATIVE CONTROLS l

I J

6.1 RESPONSIBILITY 6.1.1 The Station Manager, Byron

Station, shall be responsible for overall unit operation and shall delegate in writing the succession to this respon-I sibility during his absence.

l l

6.1.2 The Shift Engineer (or during his absence from the control room, a designated individual) shall be responsible for the control room command function.

A management directive to this effect, signed by the Assistant Vice j

President and General Manager-Nuclear Stations shall be reissued to all 1

station personnel on an annual basis.

]

6.2 ORGANIZATION OFFSITE l

l 1

6.2.1 The offsite organization for unit management and technical support shall l

be as shown in Figure 6.2-1.

l I

UNIT STAFF 6.2.2 The unit organization shall be as shown in Figure 6.2-2 and:

Each on duty shift shall be composed of at least the minimum shift

)

a.

crew composition shown in Table 6.2-1; and l

b.

At least one licensed Operator shall be in the control room when i

fuel is in the reactor.

In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the centrol room; c.

A Radiation Chemistry Technician,* qualified in radiation protection procedures, shall be on site when fuel is in the reactor; d.

All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;

The Radiation Chemistry Technician may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.

BYEDN - 18ETS.1 A 2 6.1 AMENDMENT NO. 10

'a i

I ADMINISTRATIVE CONTROLS UNIT STAFF (Continued)

]

e.

. Administrative procedures shall be developed and implemented to l

limit the working hours of unit staff who perform safety-related functions; e.g., licensed Senior Operators, licensed Operators, health physics personnel, equipment operators, and key maintenance personnel.

The amount of overtime worked by Unit staff members performing l

tafety-related functions shall be limited in accordance with the

.NRC Policy Statement on working hours (Generic Letter No. 82-12).

6.2.3 ONSITENUCLEARSAFETYGROUP(0NSGJ FUNCTION 6.2.3.1 The 0NSG serves as an independent safety engineering group and shall function to examine plant operating characteristics, NRC issuances, industry advisories, REPORTABLE EVENTS and other sources of plant design and operating experience information, including plants of similar design, which may indicate areas for improving plant safety.

The 0NSG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, opera-tions activities or other means of improving plant safety to the Manager of Nuclear Safety, and the Station Manager, Byron Station.

l COMPOSITION 6.2.3.2 The 0NSG shall be composed of at least four, dedicated, full-time engineers located on site.

RESPONSIBILITIES 6.2.3.3 The 0NSG.shall be responsible for maintaining surveillance of plant activities to provide independent verification

that these activities are performed correctly and that human errors are reduced as much as practical.

RECORDS 6.2.3.4 Records of activities performed by the 0NSG shall be prepared, main-tained, and forwarded each calendar month to the Mahager of Nuclear Safety, and the Station Manager, Byron Station.

6.2.4 SHIFT TECHNICAL ADVISOR l

l The Station Control Room Engineer (SCRE) may serve as the Shift Technical Advisor (STA) during abnormal operating or accident conditions.

During these conditions the SCRL or other on duty STA shall provide technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering and plant analysis with regard to the safe operation of the unit.

Not responsible for sign-off function.

l BYRON - UNITS 1 & 2 6-2 AMENDMENT NO. 10

j S.~4 _,

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.L ADMINISTRATIVE CONTROLS ONSITE (Continued) 3)

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

Review of all proposed changes or modifications'to plant j

, systems or equipment that affect nuclear safety; 5)

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

1 Assistant Vice President and General Manager - Nuclear Stations j

and to the Supervisor of ths Offsite Review and Investigative e

Function; 6)

Review of all REPORTABLE EVENTS; 7)

Performance of special reviews ar,d 4vestigations and reports thereon as requested by the Supervisor of the Offsite Review and Investigative Function; 8)

Review of the Station Security Plan and implementing procedures and submittal of recommended changes to the Assistant Vice President and General Manager - Nuclear Stations; 9)

Review of the Emergency Plan and station implementing procedures and submittal of recommended changes to,the Assistant Vice President and General Manager - Nuclear Stations;

10) Review of Unit operations to detect potential hazards to nuclear j

safety; r

1

11) Review of any accidental, unplanned, or uncontrolled radioactive j

release including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Assistant Vice President and General Manager - Nuclear Stations and the Supervisor of the Offsite Review and Investigative Function; and

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

13) Review of the Fire, Protection Program and implementing instruc-tions and submittal of. recommended changes to the Offsite Review and Investigative Function, c.

Authority The Technical Staff Supervisor is responsible to the Station Manager and shall make recommendations in a timely manner in all areas of review, investigation, and quality control phases of plant maintenance, operation, and administrative procedures relating to facility operations and shall have the authority to request the' H

action necessary to ensure compliance with rules, regulations, and procedures when in his opinion such action is necessary. The Station Manager shall follow such recommendations or selrt:t a course-BYRON - 131175 2 & 2 6-13

ca m ENTtt). 30

ADMINISTRATIVE CONTROLS

~

ONSITE (Continued) of action that is more conservative regarding safe operation of the facility..All such disagreements shall be reported immediately to' the Assistant Vice President and General Manager - Nuclear Stations and the Supervisor of the Offsite Review and Investigative Function.

d.

Records

1)

Reports, reviews, investigations, and recommendations shall be documented with copies to the Assistant Vice President and General Manager < Nuclear Stations, the Supervisor of the Offsite Review and Invest,igative Function, the Station Manager, and the Manager of Quality Assurance.

2)

Copies of all records and documentation shall be kept on file at the station.

e.

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

These procedures shall include the following:

1)

Content and method of submission and presentation to the Station Manager, Assistant Vice President and General Manager - Nuclear Stations,'and the Supervisor of the~0ffsite Review and Investigative Function, 2)

Use of committees, 3)

Review and approval, 4)

Detailed listing of items to be reviewed, 5)

Procedures for administration of the quality control activities, and 6)

Assignment of responsibilities.

f.

Personnel 1)

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

a)

Nuclear power plant technology, b)

Reactor operations, c)

Reactor engineering, d)

Chemistry e)

Radiological controls, f)

. Instrumentation and control, and g)

Mechanical and electrical systems.

SYaDN-UllITS 1 & 2 6-14 Amendment ilo. S

,r 6.,,

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- J ADMINISTP.ATIVE CONTROLS ONSITE (Continued)_

l I

2)

Pirsonriel performing the Onsite Review and Investigative Function shall' meet minimum acceptable levels as descrit-ed I

in At451 N18.1-1971, Sectiens 4.2 and 4.4.

6.6 REPORTABLE. _EV.ENT ACTION ThefolIowingactionsshaY1betakenfor,PEPORTABLEEVENTS:

f 6.6.1 a.'

The Commission shall be riotified and a report submitted pursuant i

to the requirements of Section 50.73 to 10 CFR Part'50, and b.

Each REPORTABLE EVENT shall be reviewed by the Onsite Review and Investigative Function and the results of tnis review shs11 be l

1 submitted to the Offsite Review and Investigative Function and the Ass'.stant Vice President end General Mareger - Nuclear Stations.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The followirg actions shall be taken in the event a Safety Liurit is violated:

a.

The NRC Operations Center shall be r.otified by telephone ar soon as possible ar.d in a'l cases within I hour.

The Assistant Vice President j

and General Manager - Nuclear Stations and the Offsite Review and Invt.stig tive Function shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; b.

A Safety Limit Violation Report shall oe prepared. The report shall

> i be revie:wed by the Onsite Review and Investige.tiVe function.

This report shall describe: (1) applicable circumstances preceding the violation, (2) ef fects of the violation upor. f acility components, systems'or struct4res, and (3) corrective action taken to prevent l

rccur rence,'

s c.

The Safety t.imit Violation Report shall be submitted to the Commission, I

the Offd te Review and Investigative Function and the Assistant Vice j

Preside.r.t" and General Manager - Nuclear Stations within 14 days of the violation; and d.

Critical operttion of the Unit shall not be resumed until authorized by the Commission.

E 6J PROCEDURES AND,3 0 GRAMS' 6.8.1 Written procedures shalf te established, implemented, and maintained covering the activities referenced below:

The applicable procedures recommended in Appendix A, of Reguiatory a.

Guide 1.33, Revision 2, Februny 1978, d

E BVnll ~ niEIS.t A 2 6-15 Ammhaent no. 9

'N 4

q,.',

ADMINISTRATIVE CONTROLS i

PROCEDURES AND PROGRAMS (Continued) j i

b.

The emergency operating procedures required to implement the require-l ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Sec-i tion 7.1 of Generic Letter No. 82-33;

)

l c.

Plant Security Plan implementation, d.

Radiological Emergency Response Plan implementation, e.

fR0 CESS CONTROL PROGRAM implementation, f.

OFFSITE DOSE CALCULATION' MANUAL implementation, g.

Quality Assurance Program implementation for effluent and environ-l mental monitoring, and h.

Fire Protection Program implementation.

l 6.8.2 Each procedure of Specification 6.8.1 above, and changes thereto, shall be reviewed prior to implementation as set forth in Specification 6.5 above.

6.8.3 Temporary changes to procedures of Specification 6.8.1 above, may be made provided:

)

a.

The intent of the original procedure is not altered; b.

The change is approved by two members of_the plant management staff, at least one of whom holds a Senior Operator license on the Unit affected; and c.

The change is documented, reviewed by the Onsite Review and Investi-gative Function, and approved by the Station Manager within 14 days of implementation.

6.8.4 The following programs shall be established, implemented, and maintained:

a.

Reactor Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

The systems include the_ recirculation portion of the Containment Spray System, Safety Injection System, Chemical and Volume Control System, and RHR System.

The progran shall include the following:

1)

Preventive maintenance and periodic visual inspection requirements, and 2)

Integrated leak test requirements for each system at refueling cycle intervals or less.

BYIIDIl

.1EfIS 2 1.2 frJ.6 AMENDMENT NO. 10 4

ML20239A598

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