Text

Application for Amend to Licenses NPF-37,NPF-66 & NPF-72, Changing Tech Specs for one-time Extension of Interval for Performing Certain 18-month Instrument Surveillances Until First Refueling Outage.Fee Paid
	ML20238C890
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	09/03/1987
From:	Ainger K; COMMONWEALTH EDISON CO.
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20238C892	List: ML20238C890;
References
	3401K, NUDOCS 8709100434
	Download: ML20238C890 (3)
	v • d • e

I Comm:nw alth Edison 1

One First Nitiond Plaza, Chicago, Illinois V

Address Reply to: Post Office Box 767 Chicago, Illinois 60690 - 0767 September 3, 1987 l

U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Subject:

Byron Station Unit 2 Braidwood Station Units 1 and 2 Application for Amendment to Facility Operating Licenses NPF-37, NPF-66, l

and NPF-72 l

Appendix A, Technical Specifications NRC Docket Nos. 50-454, 50-455, 50-456 and 50-457 Gentlemen:

Pursuant to 10CFR 50.90, Commonwealth Edison proposes to amend Appendix A, Technical Specifications, of Facility Operating Licenses NPF-37, NPF-66, and NPF-72.

The proposed amendment requests changes to Technical Specifications 4.2.3.4, 4.4.4.1, 4.4.6.1, 4.4.9.3.1 and 4.5.1.2 and Technical Specification Tables 4.3-1, 4.3-2, 4.3-3, 4.3-6, 4.3-7, 4.3-8, 4.3-9 for a one time extension of the interval for performing certain 18 month instrument surveillance until the first refueling outages for Byron Unit 2 and Braidwood Unit 1, respectively. These changes apply to Braidwood Unit 2 as well.

l The revised Technical Specification pages are contained in Attachment A.

The basis for our request for Technical Specification schedular relief is discussed in Attachment B.

Attachment C documents our safety evaluation of the proposed amendment and Attachment D addresses instrument drift.

The proposed changes have been reviewed and approved by both On-site and Off-site review in accordance with Commonwealth Edison Company procedures. We have reviewed this proposed amendment in accordance with 10 CFR 50.92(c) and determined that no significant hazards consideration exists. This evaluation is documented in Attachment E.

Commonwealth Edison is notifying the State of Illinois of our application for this amendment by transmitting a copy of this letter and its attachments to the designated State Official.

8709100434 870903 PDR ADOCK 05000454-P PDR 0

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U.S. NRC September 3, 1987

]

In accordance with 10 CFR 170, a fee remittance in the amount of I

$150.00 is enclosed.

I The Technical Specification schedular relief requested in this amendment will reduce the duration of surveillance outages scheduled to begin in late October, 1987 for Braidwood Unit 1 and late November, 1987 for Byron Unit 2.

Accordingly, prompt NRC cction on this application is requested.

q l

Please direct any questions you may have concerning this matter to J

this office.

l Very truly yours, l

l K. A. Ainger Nuclear Licensing Administrator crs

Enclosure:

Fee Remittance I

Attachments (A): Proposed Technical Specification Changes l

(B): Basis for Technical Specification Schedular Relief i

(C): Safety Evaluation (D):

Instrument Drift (E):

Evaluation of Significant Hazards Consideration l

l l

cc: Resident Inspector - Byron Resident Inspector - Braidwood L. N. Olshan - NRR S. Sands - NRR Region III Office M. C. Parker - State of Ill.

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1 ATTACHMENT A PROPOSED CHANGES TO APPERDIX A, TECHNICAL SPECIFICATIONS OF FACILITY l

OPERATING LICENSES NPF-66 and g!PF-72 Byron Station Braidwood Station Revised Pages: 3/4 2-9 Revised Pages:

3/4 2-9 3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-10 l

3/4 3-11 3/4 3-11

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3/4 3-12 3/4 3-12 3/4 3-34 3/4 3-34 3/4 3-35 3/4 3-35 3/4 3-36 3/4 3-36 3/4 3-37 3/4 3-37 3/4 3-38 3/4 3-38 I

3/4 3-42 3/4 3-42 3/4 3-52 3/4 3-52 3/4 3-55 3/4 3-55 3/4 3-68 3/4 3 3/4 3-69 3/4 3-63 3/4 3-75 3/4 3-69 3/4 3-76 3/4 3-70 3/4 3-77 3/4 3-71 3/4 3-78 3/4 3-72 3/4 4-12 3/4 4-12 3/4 4-20 3/4 4-20 3/4 4-41 3/4 4-41 3/4 5-2 3/4 5-2 3401K l

i

0WER DISTR!BUTION LIMITS LIMITING CCNDITICN FOR OPERATION ACTION (Continued) b.

Witnin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limits, verify l

through incore flux mapping and RCS total flow rate comparison that N

the combination of F and RCS total flow rate are restored to aH within the above linits, or reduce THERMAL POWER to less than 5% of

(

RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; and

]

Icentity and correct the cause of the out-of-limit condition prior c.

to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b. above; subsequent POWER OPERATION may proceed provided that the combination of F and indicated RCS total flow rate are demonstrated, througn incere flux mapping and I

RCS total flow rate comparison, to be within the region of acceptable operation defined by Specification 3.2.3 prior to exceeding the fol-lowing THERMAL POWER levels:

j j

1.

A nominal 50% of RATED THERMAL POWER, 2.

A nominal 75% of RATED THERMAL POWER, and 3.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attailing greater than or equal to 95% of f

RATED THERMAL POWER.

l SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 The combination of indicated RCS total flow rate and F" shall be deter-mined to be within the region of acceptable operation of Specification 3.2.3:

Prior to operation above 75% of RATED THERMAL POWER after each fuel a.

loading, and I

b.

At least once per 31 Effective Full Power Days.

4.2.3.3 The indicated RCS total flow rate shall be verified to be within the region of acceptable operation of Specification 3.2.3,at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the most recently obtained value of F;N~g, obtained per Speci fication 4. 2. 3. 2, is assumed to exist.

4.2.3.4 The RCS total flow rate indicators shall ce subjected to a CHANNEL CALIBRATION at least once per 18 months?

4.2.3.5 The RCS total flow rate shall be cetermined by precision heat balance measurement at least once per 13 months.

The measurement instrumentation snall be calibrated within seven days prior to the performance of the calorimetric flow measurement.

Prior to the precision heat balance measurement, at least' two of the four feewater flow meter venturis, shall be-visually inspected anc,.

if fouling is found, all venturis shall be cleaned.

BYRON - UNITS 1 & 2 3/4 2-9

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TABLE 4.3-1 (Continued)

TABLE NOTATICNS "With the Reactor Trip System creakers closed and ne Control Rod Drive System capable of rod withdrawal.

- These channels also provide inputs to ESFAS.

The Operational Test Frequency l

for these channels in Table 4.3-2 is more cor.servative and, therefore,

i*n?ghkit w wh inktval may be oknded fu 32 mrrdts & Cyde / only, l

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

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

l l

(1)

If not performed in previous 7 days.

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

THERMAL POWER.

Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%.

The provisions of Speci-l fication 4.0.4 are not applicable for entry into MODE 2 or 1.

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

Recalibrates if the absolute difference is greater than or equal to 3%.

The provisions of Specification 4.0,4 are l

not applicable for entry into MOCE 2 or 1.

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

(5)

Initial plateau curves shall be measured for each detector.

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

(6)

Incore - Excore Calibration, above 75% of '.TED THERMAL POWER.

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

1 (7)

Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

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

(9) Surveillance in MODES 3", 4*, and S" shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

Surveil-4 lance shall include verification of the Baron Oilution Alarm Setpoint of less than or equal to an increase of twice the count rate within a 10-minute period.

(1]) 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 *.he Undervoltage and Shunt trips.

(12) At least once per 18 months during shutdown verify that on a simulated Boron Dilution Doucling test signal CVCS valves 1120 and E open and l

112B and C close within 30 seconds.

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

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TABLE NOTATICNS

.symified IB n enf4 in-krval rnay be, erknckd to 52rnom%s for e/ m /y.

(1) T e DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:

a.

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

i (2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control 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 j

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

Detector check source test failure, or l

d.

Detector channel out-of-service, or I

e.

Monitor loss of sample flow.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standarcs certified by the National Bureau of Standards (NBS) or using standards that have been outained 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 subsecuent 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.

BYRON - UNITS 1 & 2 3/4 3-69

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TABLE NO"'ATONS "At all times.

""During WASTE GAS HOLOUP SYSTEM operation.

All instrume

rt'The CPeciRehts required for Unit 1 or Unit 2 heration.fs 32 men %.T for Cycle

/8 mc>rm inkrya/ rnay be eHey,dc (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic 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.

l 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.

(2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control 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 Detector check source test failure, or c.

d.

Detector channe' 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 Sureau 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 anc measurement range.

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

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

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

BYRON - UNITS 1 & 2 3/4 3-78

REACTOR CCOLANT SYSTEM l

l 3/4.4.a RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION-With one or more PORV(s) inoperable because of excessive seat leakage, a.

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERA 8LE status or close the associated block valve (s);'otherwise be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the folleving 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With both PORV(s) inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore each of the PORV(s) to OPERABLE status or close their associated block valve (s) and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

]

d.

With one or more block valve (s) inoperable, within 1 hour:

1

1) restore the block valve (s) to OPERABLE status or close the block i

valve (s) and remove power from the block valve (s), or close the PORV and remove power from its associated solenoid valve; and 2) apply the ACTION of b. or c. above, as appropriate for the isolated PORV(s).

The provisions of Specification 3.0.4 are not applicable.

e.

SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, cach PORV shall be demonstrated OPERABLE at least once per la months by:

a.

Performance of a CHANNEL CALIBRATION, anc l

e b.

Operating the valve througn one complete cycle of full travel.

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve througn one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. of Specification 3.4.4

The.JPeriBed IB men % inktval may be ekknc/cd fs 32. menMS fby Cyc BYRON - UNITS 1 & 2 3/4 4-12

l 1

(

REACTOR COOLANT SYSTEM i

3/4.4.6 REACTOR COOLANT SYSTEM LEAXAGE

{

LEM/,GE DETECTICN SYSTEMS

)

i LIMITING CONDITION FOR OPERATION I

3. 4. 6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a.

The Containment Atmosphere Particulate Radioactivity Monitoring System, b.

The Containment Floor Drain and Reactor Cavity Flow Monitoring System, and c.

The Contai'nment Gaseous Radioactivity Monitoring System.

APPLICABILITY:

MODES 1, 2, 3, and 4.

1 ACTION:

a.

With a. or c. of the above required Leakage Detection Systems inoperable, I

operati2n may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed for gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required Gaseous or l

Particulate Radioactivity Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00VN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With b. of the above required Leakage Detection Systems inoperable be in et least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO' SHUT 00WN within tne i

following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i c.

With a. and c. of the above required Leakage Detection Systems inoperable:

1)

Restore either Monitoring System (a. or c.) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and 1

l 2)

Obtain and analyze a grab sample of the containment atmosphere for l

gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and l

l 3)

Perform a Reactor Coolant System water inventory balance at least once per B hours, i

Otherwise, be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

a.

Containment Atmosphere Gaseous and Particulate Monitoring System-performance of CHANNEL CHECX, CHANNEL CAL *BRATION, and OIGITAL CWA W '.

OPERATIONAL TEST at the frequencies specified in Taole 4.3-3, b.

Containment Floor Drain and Reactor Cavity Flow Monitoring System-performance of CHANNEL CALIBRATION at least once per 18 monthsi anc c.

Verify the oil separator portion of the containment floor drain collection sump has been filled to the level of the overflow to the containment floor drain unidentified leakage collection weir box once per 18 months, following refueling, and prior to initial startup.

y l

BYRON - UNITS 1 & 2 3/4 4-20

The sYeciRed te rw>,A inkaval omy be c& dad h, 3 2. meas fw Qelei cwk,, l

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated 0FERABLE by:

a.

Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE; b.

PerformanceofaCHANNELCSIBRATIONonthePORVactuationchannel g

at least once per 18 months; and 14 c.

Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used for overpressure protection.

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:

a.

For RHR suction relief valve 8708B:

1)

By verifying at least once per 31 days that RHR RCS Suction Isolation Valve RH8702A is open with power to the valve operator removed, and 2)

By verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that RH87028 is open.

b.

For RHR suction relief valve 8708A:

1)

By verifying at least once per 31 days that RH87018 is open f

with power to the valve operator removed, and 2).

By verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that RH8701A'is open.

c.

Testing pursuant to Specification 4.0.5.

4.4.9.3.3 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months " when the vent (s) is being used for overpressure protection.

i i

"Except when tne vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

< y Spcipieg te me>&L inkeval nay be e4 ended k 32 rnan#s L~

\\

Cyde / Mly.

BYRON - UNITS 1 & 2 3/4 4-41 i

I l

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) l 1

b.

At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution I

volume increase of greater than or equal to 70 gallons by

'l verifying the boron concentration of the accumulator. solution, and At least once per 31 days when the RCS pressure is above 1000 psig c.

by verifying that the MCC compartment is open and tagged out of se'rvi ce.

4.5.1.2 Each accumu'lator water level and pressure channel shall be demonstrated OPERABLE at'least once per 18 monthfey the performance of a CHANNEL CALIBRATION.

l l

l 1

l

.I

The SpeciRed 18 tha>&h Jn,Wr#l may be ek/esde) /s 32. rno>dk.r for Cycle /

Ow ly.

BYRON - UNITS 1 & 2 3/4 5-2

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION i

ACTION (Continued) b.

WithL1 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initial'y being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of F and RCS total flow rate are restored to I

H within the above limits, or reduce THERMAL POWER to less than 5% of i

RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; and I

i c.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b. above; subsequent POWER OPERATION may proceed provided that the combination of F and indicated RCS H

total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable i

operation defined by Specification 3.2.3 prior to exceeding the fol-l lowing THERMAL POWER levels:

1.

A nominal 50% of RATED THERMAL POWER, 2.

A nominal 75% of RATED THERMAL POWER, and 3.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.

SURVEILLANCE RE0VIREMENTS i

l l

4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 The combination of indicated RCS total flow rate and F shall be deter-H mined to be within the region of acceptable operation of Specification 3.2.3:

1 a.

Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and l

b.

At least once per 31 Effective Full Power Days.

f 4.2.3.3 The indicated RCS total flow rate shall be verified to be within the I

region of acceptable operation of Specification 3.2.3 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the most recently obtained value of F g, obtained per Specification 4.2.3.2, is assumed to exist.

4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATIONatleastonceper18 months.M 4.2.3.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months.

The measurement instrumentation shall oe calibrated within seven days prior to the performance of the calorimetric flow measurement.

Prior to the precision heat balance measurement, at least two of the four feedwater flow meter venturis shall t e visualiy inspected and, if fpyling is found, all venturis shall be cleaned.

M The. 5pc a +4

d i 2 mo n+ h i n+e r VA 1 ru y be - e y f e nd e_ d.

+o '3 7_ me M h s & Cyde i em) ;

BRAIDWOOD - UNITS 1 & 2 3/4 2-9 l

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TABLE NOTATIONS

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

- These channels also provide inputs to ESFAS.

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

th TA?

chkl It than f h l^ hrval may b r et fen de d te s z. m e dh.s 4v Cycle f

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

o^I Y j!

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

(1) If not performed in previous 7 days.

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER.

Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%.

The provisions of Spec 1-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 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.

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

Subsequent plateau curves shall be obtained, evaluated and compared to the initial For the Intermediate Range and Power Range Neutron Flux channels curves.

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

(5b) With the high voltage setting varied as recommended by the manufacturer, an initial discriminator bias curve shall be measured for each detector.

Sub-sequent 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.

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

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

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

(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.

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i TABLE 4.3-8 (Continued) nt Th c Sp et fic d

/8 mo n% jMer ug fr> q b c c>e kn N d Yo 32 MOO l TABLE NOTATIONS r

fe C y c. / t / en/y (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:

a.

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

1 (2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control l

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

i a.

Instrument indicates measured levels above the Alarm Setpoint, or l

b.

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

l i

c.

Detector check source test failure, or d.

Detector channel out-of-service, or 1

e.

Monitor loss of sample flow.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of l

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.

I i

(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, j

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1 TABLEE4.3-9 (Continuedl

[

TABLE NOTATIONS i

At all times.

- During WASTE GAS HOLDUP SYSTEM operation.

All instruments required for Unit 1 or Unit 2 operation.

d

$t 4k Tb e Spe-d-fle J /R m o n+h i n +e waI ma y be-eylen d ~h>3L monIhJ A #

(1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic Cycle. I; isolation of th's pathway and control room alarm annunciation occur if any of the following conditions exists:

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

or b.

Circuit faibra (manitor loss of communications - alarm only, detector loss of county, or monitor loss of power), or c.

Detector check source test failure, or d.

Detector chancel out-of-service, or i

e.

Monitor loss of sample flow.

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

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

I i

b.

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

Detector check source test failure, or l

d.

Detector channel out-of-service, or 1

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Monitor loss of sample flow.

(3) The initial CHANNEL CALIBRATION shall be performed using nne 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 l

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 samples containing oxygen and nitrogen.

i s

BRAIDWOOD - UNITS 1 & 2 3/4 3-72

[

l' REACTOR COOLANT SYSTEM l

3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block j

valves shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

1 ACTION:

With one or more PORV(s) inoperable because of excessive seat leakage, a.

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve (s); otherwise be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV to OPERABLE status or close the associated block valva and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With both PORV(s) inoperable due to causes other than excessive seat leakage, within 1 nour either restore each of the PORV(s) to OPERABLE status or close their associated block valve (s) and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d.

With one or more block valve (s) inoperable, within I hour:

1) rectore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s), or close the PORV and remove power from its associated solenoid valve; and 2) apply the i

ACTION of b. or c. above, as appropriate for the isolated PORV(s).

The provisions of Specification 3.0.4 are not applicable.

e.

SURVEILLANCE REQUIREMENTS I

1 l

4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by:

PerformanceofaCHANNELCALIBRATION!b a.

and b.

Operating the valve through one complete cycle of full travel.

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least onge per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements

{

of ACTION b. or c. of Specification 3.4.4.

4Y The SptCd-fied 13 M 0rtO i nttr Val mg 1 b( e,y fe nde d +0 n monas +ec e.m n o /4 4-12 n w.

i BRAIDWOOD - UNITS 1 & 2 3

l

,-Q a

'y 4

7' 3

3 REACTOR C00LANT' SYSTEM.

3/4.4.6. REACTOR COOLANT SYSTEM'LEl@ SE LEAKAGE DETECTION SYSTEMS 7

y LIMITING CONDITION FOR OPERATIGN

?

3.4.6.1 The' following Reactqr Caolany System i.cakage Detection Systsms shall f

De OPERABLE:

The Containment-Atmosphe're Particulat@ndicadtivdy McPtoring System, f

a.

a b.

The Containment Floor Dr.ain and Reactor, Cavity Flow.tordtoring System, and q

j, The Containment Gaseods Radioactivity Honitoring System.-

c.

APPLICABILITY:

MODES 1, 2, 3, and 4.

t i

ACTION:

7 L]

3 With a. or c. of the above, required Leakage Detection Systems inoperable, a.

1 operation may continue.for# up to 30 days provided grab samples of the containment atmosphere are obtained end analyzed fcr gaseorg and particulate radioactivity at least orice per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the requireg Gaseous or j

i Particulate Radioactivity Monitoring System is inoperable.otherwise, be in l

at least H0T STANDBY within* the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within '

the following 30' hours.

'I s

b.

With b. of the above required (cakage Detection. Systems inoperable be in at least HOT STANDBY within the next:6 hours and.in COLD SHUTDCWN within the e

following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(,

a With a. and c. of the above, required Leakage-Detection Systems inoperable:

l c.

,1)

Restore either Monitoring System (a. or c.) to OPERABLE status within t

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and

.a 2)

Obtain and analyze a grab sample of tne conttineent atmosphere for gaseous and particulate radioactivity a.t least orse per 24' hours, and 3)

Perform a Reactor Conlant System water invd tory balance at least once l

per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD /

-i SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by-Containment Atmosphere Gaseous and Particulate Mdnitoring System-a.

performance of CHANNEL CHECK, CHANNEL CALIBRATIQH, aa.S DIGITAL CHANNEL OPERATIONAL TEST at the frequencies specified in Tal/le 4.3-3, b.

Containment Floor Drain arid Reactor Cavity now Monitoring Sysgm-performance of CHANNEL CALIBRATION at least once per 18 months, and i

Verify the oil separator portion of the containment floor drain c.

collection sump.has been filled to the level of the overflow to the containment floor drain unidentiflec leakage collection weir box once

-per 18 months, following: refueling,' and prior to initial startup.

a,

1 s

BRAI,0 WOOD - UNITS 1 & 2 V4 4-20 L

TL speeMi.ed i E @th fin +w va i rnak 6. erfe n cled 40 l

3 7., n w n g

-f o r C y c l e.>

I only;

REACTORICOOLANTSYSNM SURVEILLANCE _, REQUIREMENTS

/(4.9.3.1 Eacn PORV shall be demonstrated OPERABLE by:

s h.

' Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and et least once per 31 days thereafter when the PORV is required OPERABLE; b.

Performance of a CHANNEL CA atleastonceper18 month 8gIBRATIONonthePORVactuationchannel

, and i

c.

Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used for overpressure protection.

4.4.9.3.2 Each RH9 suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used f or cold overpressure protection as follows:

j a.

For RHR suction relief valve 8708B:

l 1)

By verifying at least once per 31 days that RHR RCS Suction l

Isolation Valve PH8702A is open with power to the valve operator removed; and 2)

By verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that RH8702B is open

{

)

b.

For RHR suction relief valve 8708A:

1 1

l 1)

By verifying at least once per 31 dayr that RH8701B is open with power to the valve operator removed, and I

l 2)

By verifying at least once per 12 nours that RH8701A is open.

I c.

Testing pursuant to Specification 4.0.5.

l

)

4.4.9.3.3 The RCS vent (s) shall be verified to be open at least once per

]

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

when the vent (s) is being used for overpressure protection.

j Txcept when the vent pathway is provided with a valve'which is locked, sealed, or othofwise secured in the open position, then verify these valves open at 1 cant coce par 31 days, d The opcc., f icd i2 mon +h in i cr' Va l may bt. e v + e n c/ e d f o I

3t m on+ h.s b

C y c le I

only.

l BRAIDWOOD - UNITS 1 & 2 3/4 4-41 1

i I

J

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i b.

At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 70 gallons by verifying the boron concentration of the accumulator solution, and i

At least once per 31 days when the RCS pressure is above 1000 psig c.

by verifying that the MCC compartment is open and tagged out of service.

4.5.1.2 Each accumulator water level and pressurg channel shall be demonstrated OPERABLE at least once per 18 monthrby the performance of a CHANNEL CALIBRATION.

@ Th e-sr;e u ri e c!

12 m on +h i n h v( I rn a y b e eMen cle cl

+o st enc an s cych i

o,, / y i

BRAIDWOOD - UNITS 1 & 2 3/4 5-2

__o

ATTACHMENT B BASIS FOR TECHNICAL SPECIFICATION SCHEDULAR RELIEF The Byron and Braidwood Technical Specifications require many surveillance tests to be performed every 18 months (plus a maximum extension defined by Specification 4.0.2).

Several of these surveillance require the units to be shutdown to be performed. The proposed change requests approval to defer a selected number of 18 month frequency Technical Specification instrument surveillance for cycle 1.

The surveillance will be performed during the first refueling outage.

This proposed amendment will reduce the duration of surveillance outages scheduled this fall for Byron Unit 2 and Braidwood Unit 1.

It will alsa minimize the duration of a surveillance outage to be scheduled during cycle 1 of Braidwood Unit 2.

The 18 month surveillance interval was placed in the Standard Technical Specifications for PWR's to be consistent with the maximum anticipated interval between refueling outages for Westinghouse pWR's.

Specification 4.0.2 allows an extension to this 18 month frequency to accommodate scheduling considerations. However, this extension (25%) is limited to a maximum of 3.25 x 18 months for 3 consecutive surveillance intervals. A "due date" reflects the date 18 months Trom when the surveillance was previously performed. A calculated " critical date" includes the allowed extension per Specification 4.0.2.

At Byron Station, the earliest "due date" for the instrument surveillance is February 21, 1988 and its associated " critical date" is July 7, 1988.

The first refueling outage for Byron Unit 2 is scheduled to commence by January 28, 1989. The period of plant operation during the maximum requested extension would be approximately eleven months from the allowed "due date" or approximately six months from the " critical date".

Tha currently scheduled "due dates" for the instrument surveillance are distributed over a period from February 21, 1988 until November 11, 1988.

As such, over half of the instrument surveillance due dates are within six j

months of the start of the Byron Unit 2 refueling outage.

The preceding discussion provides a perspective on the instrument surveillance interval extensions being requested for Byron Unit 2.

The surveillance interval extensions are somewhat less for Braidwood Unit le We intend to not use the 4.0.2 surveillance extension for these first refueling cycle surveillance since this would result in scheduling concerns for j

subsequent cycles. The discussion on critical dates has been presented to indicate the required extension beyond the maximum allowed extension per the Technical Specifications. Upon acceptance of the extension, Technical Specification 4.0.2 would apply to the three subsequent refueling cycles; 2, 3, and 4.

Byron Unit 2 has been through a startup program which has involved several shutdowns and operation at less then 100% power. Braidwood Unit 1 is currently going through the same program.

The shutdowns and reduced power operations were part of a normal startup test program which resulted in extending the length of the first fuel cycle. This is typical for a nuclear unit's first cycle.

(For example, on Byron Unit 1, the first fuel cycle was 24 months and a six week surveillance outage was required in the fall of 1986).

I

1 ]

I The instrument surveillance require up to a 21 working day outage I

because of the limitations on the number of people and testing equipment

)

that can physically be used to work on an instrument rack at one time and j

since surveillance must be 'arformed sequentially rather than I

p concurrently. Surveillance are scheduled to optimize personnel and testing equipment available.

If it was a matter of doubling personnel to decrease i

the job duration, personnel and testing equipment would be obtained from I

other Commonwealth Edison sites. However, for the instrument surveillance, obtaining additional qualified people or equipment will not expedite the performance of the required surveillance. As scheduled, surveillance testing would be performed normally by three shifts covering 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day, 6 days a week and as a result, the outage period cannot be-further reduced.

1 I

1

l ATTACHMENT C l

l SAFETY EVALUATION i

I commonwealth Edison believes there is no safety significance in l

delaying these instrument surveillance until the first refueling outages l

for the Byron and Braidwood Stations. There are other functional tests that are performed on a shiftly, monthly or quarterly basis which would detect l

degradation or inoperability of a channel or component so appropriate actions can be initiated.

These other types of testing include an analog l

(or digital) channel operational test, a channel check, a source check or a l

trip actuating device operational test. The analog or digital channel operational test and the trip actuating device operational test verify operability of the entire channel except for the sensor. Therefore, it is verified that if a channel senses an input, it can process it such that the appropriate safety function is completed.

In addition, the channel checks and source checks perform a qualitative assessment of channel behavior.

If one channel's indication deviates from its other associated channels, this can be readily detected and corrective actions would be taken.

For those items where there is more than one channel, our previous channel calibrations will have been performed at different times so certain channels will remain in frequency for a longer period because of the staggered testing.

4 At Byron Station, certain channel calibrations had been routinely scheduled to be performed at power where it has been shown that the risk of inducing an associated transient on the unit is minimal. For some inst uments, the channel calibration, which includes verification of sensor oper bility, requires actuation of systems in a mode which is not possible during normal power plant operation. For other instrument channels, testing can be performed at power but it might involve placing a channel in test for longer than the Technical Specification limits allow. performing a channel calibration at power involves placing the unit in a halt trip condition.

The unit is much more susceptible to a noise spike or minor transient that can complete the logic coincidence for a reactor trip. This can result in an undesired and unnecessary transient on the plant, j

In addition, ALARA concerns dictate that containment entries be minimized. Some of the censors are located within the missle shield or are located in high radiation areas. personnel exposure would be unacceptable to try to perform these channel calibrations at power. This also reduces the number of surveillance that can be performed at power.

I Commonwealth Edison believes that the routine shiftly, monthly and quarterly surveillance that will continue to be performed during the 18 month surveillance extension will detect instrument failure mechanisms.

Therefore, any failures can be corrected so the Units will not operate in a degraded condition. Also, the probability of an accident occurring during this surveillance extension is minimal. Reactor trips can be initiated from a variety of trip signals.

Typically there are redundant trip signals to respond to an accident scenario.

It is highly unlikely these trip signals will not function as a result of this surveillance extension. Thus, Commonwealth Edison believes this one-time extension request should not significantly impact the safety of the Byron and Braidwood Stations.

ATTACHMENT D EVALUATION OF INSTRUMENT DRIFT l

The instrument channels affected by this amendment request were left in tolerance whan the last 18 month surveillance were performed. When Byron Station Unit I received its full power license Technical Specifications, the NHC Staff had approved a change from a monthly to quarterly surveillance interval for the analog channel operational test for the reactor protection instrumentation. When this change was approved, Byron Station agreed to trend instrument drift for a period of approximately one year. Because insufficient data was collected during the first year of Byron Unit 1 operation, due to increased surveillance frequencies required by the startup program, the trending of instrument drift was extended an-additional year.

The Station trended the information gathered from l

maintenance calibrations and quarterly analog channel operational l

surveillance and compared the values obtained with rack drift allowances outlined in the Westinghouse Statistical Setpoint Study. The conclusion of j

this study was that over 99 percent of all data taken remained within the drift tolerances of the statistical setpoint study.

The above study trended only instrument drift for certain surveillance that had been changed from a monthly to a quarterly frequency. The trending data indicated very little drift tendency in the loop modules. However, recognizing the quarterly surveillance do not include the process sensors, a short list of sensors has been selected for calibration during the surveillance interval extension.

All pressure and differential pressure sensors and pressure switches which provide logical input to the reactor protection or safety injection bystems have been included. The remaining sensors selected either provide interlocks for equipment operated during design basis events or provide key indication to the operator during design basis accidents. The selected sensors monitor the following parameters: steam generator pressure, narrow and wide range steam generator level, pressurizer level and pressure, auxiliary feedwater suction pressure, RWST level, containment pressure, and reactor coolant system loop flow.

Instrument calibrations were performed during the Byron Unit 1 first refueling outage and a review has been performed of the instrument drift data obtained from these calibrations. Of the channel calibrations performed as a Technical Specification requirement, approximately 9% were determined to have drifted beyond the Technical Specification allowable limit. Of the channels outside the limit, it was noted that all turbine trip emergency trip header pressure channels were outside the limits.

Therefore Byron and Braidwood Stations are not requesting an extension for these channels. These pressure channel calibrations will be performed in accordance with the existing 18 month Technical Specification requirements.

When these channels are removed from the data base, the drift observed only involved 5% of the sensors calibrated. The remaining sensors that were found outside the limit were considered isolated, independent examples. The other channels of that instrument function remair.ed operable.

l

ATTACHMENT E EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that it '-"olves no significant hazards consideration. According to 10 CFR 00.92(

- proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the 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 i significant reduction in a margin of safety.

This proposed amendment involves a one-time extension of the 18 month Technical Specification surveillance interval for certain instrument channels in the reactor protection system aad engineered safety features actuation system. This amendment would permit the surveillance (channel calibrations) to be performed at the first refueling outage rather than within 18 months from the last time the surveillance were performed.

Deferral of the 18 month channel calibrations may increase the potential for undetected instrument drift, but it will not increase the potential for undetected instrument failure.

Instrument failures will be detected by shiftly, monthly and quarterly Technical Specification surveillance that will continue to be performed. These surveillance (functional tests) include channel checks and analog / digital channel operational tests. The magnitude of undetected instrument drift will be small and have no effect on the probability of previously evaluated l

accidents. Therefore, deferral of the 18 month channel calibration l

surveillance of these instruments will have no effect on the probability of previously evaluated accidents.

l l

Within the group of instruments whose 18 nonth channel calibration is being deferred, the instruments which mitigate the consequences of previously evaluated accidents have redundant channels to perform their protective function. Additionally, the reactor protection system ani l

engineered safety features actuation system contain diverse features to mitigate the consequences of accidents. Potential drift of an instrument sensor within a certain protective function would be compensated for by l

redundant channels of that function or by another, diverse protective function within the reactor protection system or engineered safety features actuation system. Therefore, deferral of the 18 month channel calibrat$on surveillance of these instruments will not involve a significant increase in the consequences of previously evaluated accidents.

This proposed amendment does not add or modify any existing equipment, nor introduce a new mode of plant operation.

The operability of the instruments will continue to be verified by performing the other Technical Specification surveillance discussed above. Accordingly, this proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated, i

_2_

)

The setpoints and protective functions of the instrument channels whose 18 month surveillance is being deferred are not being changed as a result of this amendment. The setpoints and protective functions include j

margin to assure applicable safety limits will not be exceeded following

)

design basis accidents. The setpoints and protective functions will

]

continue to be checked for proper operation when the monthly and quarterly j

surveillance are performed. As a result, this proposed amendment will not involve a significant reduction in a margin of safety.

i For the reasons stated above, Commonwealth Edison believes this i

proposed amendment involves no significant hazards consideration.

1 I

l 3401K l

i l

1 I

_- - -

ML20238C890

Text

Subject:

Enclosure:

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