ML20155J205

Proposed TS Converting to Its,Rev R

ML20155J205
Person / Time
Site:	Byron, Braidwood
Issue date:	11/05/1998
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20155J154	List: ML20155J151 ML20155J205
References
NUDOCS 9811120058
Download: ML20155J205 (140)

Text

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MSIVs 3.7.2 f

SURVEILLANCE RE001REMENTS V

SURVEILLANCE FREQUENCY SR 3.7.2.-1 NOTE Only required to be performed in MODES 1 and 2.

Verify closure time of each MSIV is In accordance s 5 seconds.

with the Inservice Testing Program O

i N

SR-3.7.2.2 NOTE 3-

_g Only required to be performed in MODES 1 and 2.

H k

Verify each MSIV. actuates to the isolation 18 months position on an actual or simulated actuation signal.

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. BYRON - UNITS 1 & 2

3. 7.2 - 2 11/5/98 Revision R 9811120058 991106 P

PDR ADOCK 05000454.-

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MSIVs B 3.7.2 i

BASES e

(v SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR verifies that MSIV closure time is s 5 seconds. The MSIV closure time is assunied in the accident and containment e

analyses. This Surveillance is normally performed upon o

returning the unit to operation following a refueling

. outage.

Based on ASME Code Section XI (Ref. 5), the MSIVs 9

are not closure time tested at power.

b ni The Frequency is in accordance with the Inservice Testing Program. This test is conducted in MODE 3 with the unit at h

operating temperatore and pressure.

This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of-testing until MODE 3. to establish conditions consistent with those under which the acceptance criterion was generated.

SR 3.7.2.2 1his SR verifies that each MSIV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage. The frequency of MSIV testing is every 18 months. The 18 month Frequency for testing is based on 7 $,

the refueling cycle. Operating experience has shown that sV these components usually pass the Surveillance when O

performed at the 18 month Frequency. Therefore. this 8 I Frequency is acceptable from a reliability standpoint.

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This SR is modified by a Note that allows entry into and ai operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3. to establish conditions y

consistent with those under which the acceptance criterion

.-.D was generated.

REFERENCES 1.

UFSAR. Section 10.3.

2.

UFSAR. Section 15.1.5.

3.

UFSAR. Section 6.2.

4.

10 CFR 100.11.

5.

ASME. Boiler and Pressure Vessel Code.Section XI.

,.q V

BYRON - UNITS 1 & 2 B 3.7.2 - 6 11/5/98 Revision R

BRWD ITS O

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MSIVs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 NOTE Only required to be performed in MODES 1 and 2.

i Verify closure time of each MSIV is In accordance s 5 seconds.

with the Inservice Testing Program h10 I

N SR 3.7.2.2 NOTE N

Only required to be performed in MODES 1 M

and 2.

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Verify each MSIV ac' uates't'o the isolation 18 months t

position on an actual or simulated

' actuation signal.

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BRAIDWOOD - UNITS 1 & 2 3.7.2 - 2 11/5/98 Revision R

l MSIVs B 3.7.2 BASES

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lURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR verifies that MSIV closure time is s 5 seconds. The MSIV closure time is assumed in the accident and containment analyses. This Surveillance is normally performed upon 9

O returning the unit to operation following a refueling i

outage. Based on ASME Code Section XI (Ref. 5), the MSIVs N

are not Closure time tested at power.

K The Frequency is in accordance kith the Inservice Testing h.

Program. This test is conducted in MODE 3 with the unit at h

l operating temperature and pressure. This SR is modified by a Note that allows entry into and operation in MODE 3 prior T

k to performing the SR. This allows a delay of testing until MODE 3. to establish conditions consistent with those under which the acceptance criterion was generated.

SR 3.7.2.2 This SR verifies that each MSIV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a 7:c refueling outage. The frequency of MSIV testing is every f-N 18 months. The 18 month Frequency for testu g is based on k-)

the refueling cycle. Operating experience has shown that

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these components usually pass the Surveillance when D

3erformed at the 18 month Frequency. Therefore, this

'I requency is acceptable from a reliability standpoint.

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Ny This SR is modified by a Note that allows entry into and q

operation in MODE 3 prior to performing the SR. This allows c

a delay of testing until MODE 3, to establish conditions h

consistent with those under which the acceptance criterion was generated.

l REFERENCES 1.

UFSAR. Section 10.3.

2.

UFSAR. Section 15.1.5.

3.

UF5AR. Section 6.2.

4.

10 CFR 100.11.

i 5.

ASME Boiler and Pressure Vessel Code.Section XI.

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BRAIDWOOD - UNITS 1 & 2 B 3.7.2 - 6 11/5/98 Revision R i

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SURVEILLANCE FREQUENCY l

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Verify each MSIV actuates to the isolation 18 months position on an actual or simulated actuation' signal.

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LC0 3.7.2 INSERT 3.7 9C (A )

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DISCUSSION OF CHANGES TO CTS ITS SECTION 3.7 - PLANT SYSTEMS A,

CTS SR 4.7.1.3.1 and SR 4.7.1.3.2 are conditional dependin on whether the CST or Essential Service Water (SX) System is the supp y source for the AF pumps.

CTS 3.7.1.3 Action b directs the demonstration of the SX System as operable for a backup supply to AF. ITS SR 3.0.1 indicates that SRs do not have to be performed on inoperable equipment and ITS 3.7.6 Condition A includes the requirement to demonstrate by administrative means the operability of the backup system. which in this case would be the SX System. Therefore. deletion of the references to CST or SX System as the source for AF is an administrative change associated with ITS format. During this reformatting, no technical changes (either actual or interpretational) were made to the TS. unless identified and justified.

A CTS 3.7.1.5 Mode 1 Actions address an inoperable but open MSIV.

Byron a

and Braidwood are not licensed for three loo) operation. Therefore indefinite operation would not permitted wit 1 a closed MSIV. Therefore, the deletion of the words "but open" is administrative in nature. There is no technical change (either actual or interpretational) made.

The CTS Mode 1 Actions include a requirement to be in Hot Standby (Mode

3) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Hot Shutdown (Mode 4) within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if an inoperable MSIV is not restored to OPERABLE r

status. Once the unit is in Mode 2 (within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />), the CTS Action for

'j, Modes 2 and 3 would apply. CTS 3.7.1.5 includes the same Actions for Modes 2 and 3.

ITS. Conditions A and B are associated with Mode 1. while Conditions C and D are associated with Modes 2 and 3.

In ITS if Condition B was entered (Be in Mode 2). as soon as Mode 2 was reached Condition D would be applicable and would require shut down to Mode 4 (per the Required Actions of Condition E). Therefore the requirements of CTS Actions are covered by ITS 3.7.2 Actions and this change is administrative in nature. No technical changes were made to the TS, unless identified and justified.

M' o

A CTS 3.7.1.5 Modes 2 and 3 Actions state that the provisions of CTS n

Specification 3.0.4 are not applicable.

In addition CTS SR 4.7.1.5 9

states that the provisions of Specification 4.0.4 are not applicable.

b The CTS requirements allow entry into the Mode of Applicability to M

perform the surveillance to verify perability. The CTS has been i

revised to delete the reference to pecification 3.0.4. and 4.0.4.

This D

has been replaced with the Note for ITS SR 3.7.2.1 and SR 3.7.2.2 which g

requires the SRs only in Modes 1 and 2.

ITS relies upon the guidance of SR 3.0.4 and wording of the SRs to allow the performance of these SRs in Mode 3.

This is merely a reformatting of existing requirements.

During this reformatting, no technical changes (either actual or interpretational) were made to the TS unless identified and justified.

BYRON /BRAIDWOOD - UNITS 1 & 2 3.73 11/5/98 Revision R

l DISCUSSION OF CHANGES TO CTS r~'S ITS SECTION 3.7 PLANT SYSTEMS lV l

l Ma CTS LCO 3.9.11 APPLICABILITY states "Whenever irradiated fuel assemblies are in the storage 2001."

ITS LC0 3.7.15 APPLICABILITY revises the CTS by stating. "W1enever fuel assemblies are stored in the spent fuel pool." The CTS APPLICABILITY was only when irradiated fuel g

was stored in the pool. The ITS is more restrictive since the 3

elimination of the word " irradiated" now requires that the APPLICABILITY yg-is for anytime new or irradiated fuel is in the pool.

This change is consistent with NUREG-1431.

H, (Byron Only) CTS LCO 3.7.5 Action e.l.a requires restoration of the g

level switch to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or verification that the UHS basin level is greater than or equal to 90% within the next hour after the time one UHS cooling tower b min switch is determined to be inoperable.

ITS 3.7.9 Required Acti

? ' requires this action to be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (versus R %% in the CTS) after the Z

determination that one SX makeup pump 1s inoperable, whether due to an L

-inoperable basin level switch or other cause. The reduction in j

g Completion Times for the inoperable tower basin switch is a more restrictive change to the requirements.that currently apply to the facility.

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(Byron Only) CTS LCO 3.7.5 Action e permits contin'ued' operation for an a

7.s indefinite period with one cooling tower basin level switch ino)erable.

ITS 3.7.9 Required Action C.2 requires verification of an operaale i

associated makeu) source for the UHS cooling tower basin within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and C.3 limits t1e time the plant may continue to operate with one basin g

level switch inoperable. RA C.3 limits plant operations to 7 days if l

both units are in Mode 1. 2. 3. or 4 and 14 days if one unit is in Mode 5, 6. or defueled with an inoperable SX makeup aump. This limit on k

the amount of time that a unit may operate with one JHS cooling tower basin level switch inoperable is an additional restriction on plant operations.

M (Byron Only) CTS LC0 3.7.5 Action e.2.a indicates that the provisions u

of Specification 3.0.4 are not a]plicable when both UHS cooling tower l-basin level switches are inoperaale.

ITS 3.7.9 Action D, which would be 2

applicable if two UHS cooling tower basin level switches are inoperable.

does not provide a similar allowance. This is an additional restriction j

k on plant operations.

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H Verify each MSIV actuates to the isolation 18 months position on an actual or simulated actuation signal, t

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Isj SURVEILLANCE SR 3.7.2.1 (continued)

REQUIREMENTS The Frequency is in accordance with the Inservice Testing yI"r:rogram er [M] menth:K The [10] = nth Frequency for vake ti:: la "*

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B OP Ntin-3 0x;;rien-a h u 9 til-th:t th::: cer;enerte "eu2H y p::: the Cneveillanen when perfgmed at the f_181 mnnth reannoney, Th;raf;re, th; Trcqu n:3 i:==pt:Me 4re : ; reliability

t;ndpeint.

I Mhis test is conducted in MODE 3 with the unit at operating

$g temperature and pressure as din = n d in ",efe.;a n 5 m

s ;rci;ia; rev :....u ts. This SR is modified by a Note that 4 j allows entry into and operation in MODE 3 prior to 3 cc performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.

C E nsert 8 3.7-rid REFERENCES

• 1.

UFSAR,Section)10.3[

fR.UFSAR,Section\\6.2

)g f,'t:,

d. UFSAR, Section I15.1.5],,.

4.

10 CFR 100,11.

5.

ASME, Boiler and Pressure Vessel Code Section XI.

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Q WOG STS B 3.7-12 Rev 1, 04/07/95 6o R

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I fy BASES INSERT (S) l SECTION 3.7 Bases 3.7.2 l

INSERT B 3.7 12A (P )

7 SR 3.7.2.2 l

This SR verifies that each MSIV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to l

S operation following a refueling outage. The frequency of MSIV testing is I

O every 18 months.

The'18 month Frequency for testing is based on the refueling I

cycle. Operating experience has shown that these components usually pass the N

Surveillance when performed at the 18 month Frecuency. Therefore, this N

Frequency is acceptable from a reliability stancpoint.

6 This SR is modified by a Note that allows entry into and operation in MODE 3 Q

prior to performing the SR, This allows a delay of testing until MODE 3 to k

establish conditions consistent with those under which the acceptance criterion was generated.

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11/5/98 Revision R O

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Commonwealth Edis:m Company tiraidwom! Generatiri 4 Mation Route s1, llox 84 liraceville,11. N)4074 i19 Tcl 815-45&2801 November 6,1998 United States Nuclear Regulatory Commission Attn: Document Contral Desk Washington D. C. 20555 - 000)

Subject:

Revision Q to the Improved Technical Specifications (ITS) Submittal Byron Nuclear Power Station, Units 1 and 2 Facility Operating Licenses NPF-37 and NPF-66 NRC Docket Numbers: 50-454 and 50-455 5

Braidwood. Nuclear Power Station, Units 1 and 2 Facility Operating Licenses NPF-72 and NPF-77 NRC Doc ket Numbers: 50-456 and 50-457 C')

C

Reference:

G. Stanley and K. Graesser (Commonwealth Edison) letter to NRC Document Control Desk, " Conversion to the Improved Standard Technical Specifications," dated Decemb:r 13,1996 The purpose of this letter is to provide Revision Q to the referenced ITS submittal. ITS Revision Q (Enclosure 1) contains minor miscellaneous cleanup items for ITS Sections 3.3,3.6,3.7,3.8,3.9, and 5.0.

These Revisions are being provided in the same ten-section format as the initial ITS submittal:

i

1. Byron ITS

2. Braidwood(Brwd)ITS

3. Byron CTS Markups

4. Brwd CTS Markups

5. CTS Discussion of Chrnges (DOCS)

6. LCO Marl.ups

7. LCO Justification for Differences (JFDs)

8. Bases Marktps A

9. Bases JFDs V

10. No Significant Hazards Consideration (NSHC)

A. Unicom Company

i Document Control Desk

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November 6,1998 Page 2 Please address any comments or questions regarding this matter to our Nuclear Licensing

. Department.

Sincerely, E

T thy J. Tulon te Vice President Braidwood Nuclear Generating Station : ITS Revision Q cc:

NRC Regional Administrator - Region III Senior Resident Inspector - Braidwood

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Senior Resident Inspector-Byron Office of Nuclear Facility Safety -IDNS nrc/980710t. doc O=

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I LlO REVISED RAI RESPONSE J

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Response to NRC RAI For ITS Section 3.6 05-Nov-98

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NRC RAI Number NRC issued Date RAI Status b

3.6.3-08 11/5/97 Closed NRC Description of Issue JFD B3 Bases JFD B3 Bases JFD Pl3 STS 3.6.3 Action E -

STS SR 3.6.3.1 and associated Bases discussions ITS 3.6.3 Action D ITS SR 3.6.3.1 and associated Bases discussions STS 3.6.3 RA E.1, E.2 and E.3 have been revised in ITS 3.6.3 ACTION D to delete the option ofisolating a penetration flow path with a purge valve not within the leakage limits and associated ras. The only option is the current licensing basis of restoring the valve to OPERABLE status in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or begin an orderly plant shutdown. Also, the STS SR 3.6.3.1 exception for opening one purge valves while in Condition E is deleted. Justification B.3, Bases B.3 and Bases P.13 state this exception is already accommodated in Condition A Required Actions. This is not understood because if a purge valve is leaking, then Condition A cannot be entered. Thejustification must be further explained. Comment: Provide additional discussion and justification for this CTS change.

Comed Response to issue-Revised Response: Comed's response at the 4/1/98 meeting was to evaluate adopting the NUREG for ITS 3.6.3 Required Action E and SR 3.6.3.1. Comed is retaining the changes associated with ITS Revision A based on current licensing basis.

7 Original Response: " Condition A" has been changed to " Condition D" in Section 3.6 LCO JFD B3 and Bases JFD B3. The JFDs have been revised to state, "This exception is adequately addressed by Condition D Required Actions, without the necessity for the complexity introduced by this exception." The exception in SR 3.6.3.1 is unnecessary since SRs do not f

have to be performed on inoperable equipment or components, i.e., when Condition D is entered.

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ENCLOSURE 1 O

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ITS REVISION Q l

ITS SECTIONS 3.3/3.6/3.7/3.8/3.9/5.0 i

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RTS Instrumentation

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SURVE!LLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY NOTE--

.SR 3.3.1.9 Verification of setooint is not required.

Perform TADOT 92 days NOTE l' SR 3.3.1.10

-G This Surveillance shall include 2

F

. verification that the time constants are i -

adjusted to the prescribed values.

]

m Perform CHANNEL CALIBRATION.

18 montns W

--SR_ 3.3.1.11 NOTE Neutron detectors are excluded from CHANNEL 3g CALIBRATION.

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Perform CHANNEL CALIBRATION.

18 months le

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SR 3.3.1.12.

Perform COT.

18 months SR -3.3.1.13 NOTE Verification of setpoint is not required.

Perform TAD 0T.

18 months

-(continued) l p-V

BYRON UNITS 1 & 2-3.3.1 - 12 10/27/98 Revision 0 L

.Il RTS Instrumentation 3.3.1

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Taole 3.3 1 1 (page a of 6)

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Reactor Trto System Instrumentattor.

APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION

. CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE 17 Reactor Trip System Interlocks a.

Source Range Block 2(d) 2 0

SR 3.3.1.11 a 6E-11 ame Perm 1551ve. P-6 SR 3.3.1.12 0.

Low Power Reactor Trips Block. P-7 "o

(1) P-10 Inout 1

3 P

SR 3.3.1.11 NA SR 3.3.1.12 3

-[ E (2) P 13 Inout 1

2 P

SR 3.3.1.13 NA SR 3.3.1.1" J

c.

Power Range 1

3 P

SR 3.3.1.11 s 32.11 RTP Neutron Flux. P-B SR 3.3.1.12

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Power Rance 1.2 3

0 SR 3.3.1.11 a 7.9% RTP and i

Neutron Flux. P 10 SR 3.3.1,12 s 12.1% RT?

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Turbine Imoulse 1

2 P

SR 3.3.1.10 s 12.11 Pressure. P-13 SR 3.3.1.12 turoine power J

IB-Reactor Trio 1.2 2 trains N

SR 3.3.1.4 NA Breaxers (RTBs)(9)

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3(a) 4(a) 5(a) 2 trains C

SR 3.3.1.4 NA l

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SR 3.3.1.4

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undervoltage and Shunt l_

Trio Meenanisms 3(a) 4(a) 5(a) 1 each per RTB C

SR 3.3.1.a NA 4

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20. Automatic Trto Logle 1.2 2 trains M

SR 3.3.1.5 NA l

3(a) 4(a) 5(a) 2 trains C

SR 3.3.1.5 NA l ta) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(d) :Below the P 6 (Source Range Block Permissive) Interlock.

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(g) Including any reactor trip bypass breakers that are racked in and closeo for bypassing an RTB.

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' BYRON - UNITS 1 & 2

- 3. 3.1.- 17 10/27/98 Revision 0

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ESFAS Instrumentation 3.3.2 (3

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FRE0VENCY SR 3.3.2.4 Perform ACTUATION LOGIC TEST.

31 days on a STAGGERED TEST BASIS SR -3.3.2.5 Perform MASTER RELAY TEST.

31 days on a STAGGERED TEST BASIS SR 3.3.2.6 Perform COT.

92 days SR 3.3.2.7 Perform SLAVE RELAY TEST.

92 days j

SR -3.3.2.8 NOTE Verification of relay setpoints not required.

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Perform TAD 0T.

92 days i

NOTE SR 3.3.2.9 l

Verification of setpoint not required.

Perform TADOT.

18 months (continued) w,;

l BYRON - UNITS 1 & 2

3. 3.2 - 7 7/9/98 Revision E

ESFAS Instrumentation 3.3.2

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SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.10 NOTE This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

18 months I

SR 3.3.2.11-Ve'rify ESFAS RESPONSE TIMES are within 18 months-limit.

SR 3.3.2.12 Verify ESFAS RESPONSE; TIMES are within 18 months on a limit.

STAGGERED TEST BASIS t

9 BYRON - UNITS 1 & 2

3. 3. 2 - 8 10/29/98 Revision 0

,m--

o n,

y

~,

.~

v

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.9 SR 3.3.1.9 is the performance of a TADOT every 92 days. as justified in Reference 7.

4 The SR is modified by a Note that excludes verification of-setpoints from the TADOT.

Since this SR applies to RCP undervoltage and underfrequency relays, setpoint verification requires elaborate bench calibration and is accomplished during the CHANNEL CALIBRATION.

SR 3.3.1.10 A CHANNEL CALIBRATION is performed every 18 months. or approximately at every refueling.

CHANNEL CALIBRATION is a complete check of the instrument loop. including the sensor.

s The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

yvt CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the plant specific setpoint methodology. The difference between the current "as found" values and the he previous test "as left" values must be consistent with the calculated normal uncertainties consistent with the setpoint methodology.

y g

j The Frequency of 18 months is based on the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint methodology.

E SR 3.3.1.10 is modified by a Note stating that this test f

shall include verification that the time constants are 6

adjusted to the prescribed values where applicable.

BYRON - UNITS 1 & 2 8 3.3.1 - 56 10/26/98 Revision 0

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.2.9 SR 3.3.2.9 is the performance of a TADOT.

This test is a 5

check of the Manual Actuation Functions and P-4 Reactor Trip

@W Interlock.

It is performed every 18 months.

Each Manual Actuation Function is tested up to and including, the master relay coils.

In some instances, the test includes actuation of the end device (i.e.. pump starts, valve cycles, etc.). The Frequency is adequate, based on industry operating experience and is consistent with the typical 3

refueling cycle.

The SR is modified by a Note that' excludes

@Wj verification of setpoints during the TADOT.

The Functions have no associated setpoints.

SR 3.3.2.10 SR 3.3.2.10 is the performance of a CHANNEL CALIBRATION.

A CHANNEL CALIBRATION is performed every 18 months. or approximately at every refueling.

CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor.

The test verifies that the channel responds to measured 7 w) parameter within the necessary range and accuracy.

i%pu CHANNEL CALIBRATIONS must be performed consistent with the; * > %

assumptions of the plant specific setpoint methodology. The difference between the current "as found" values and the 1

previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.

>db

'The Frequency of 18 months is based on the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint methodology.

This SR is modified by a Note stating that this test should 3

m include verification that the time constants are adjusted to De, the prescribed values where applicable.

bd BYRON.-' UNITS 1 & 2 B 3.3.2 - 54 10/29/98 Revision 0

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RTS Iristrumentation i

l 3.3.1 L.

1

(

SURVEILLANCE REQUIREMENTS icontinued) l SURVEILLANCE FREQUENCY L

SR 3.3.1.9

---NOTE -

l Verification of setpoint is not required.

l t

Perform TADOT.

92 days SR 3.3.1.10 NOTE

~ ~,

This Surveillance shall include A

verification that the time constants are t-!

adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

18 months i

10 SR 3.3.1.11 NOTE -

l Neutron detectors are excluded from CHA'lNE;.

l.FQ CALIBRATION.

lV o

L Perform CHANNEL CALIBRATION.

18 months I

W l-l s

SR 3.3.1.12 Perform COT.

18 months NOTE SR. 3.3.1.13 l

Verification of setpoint is not required.

l-Perform TADOT.

18 months o

(continued) 1 i,.v BRAIDWOOD - UNITS 1 & 2 3.3.1 - 12 10/27/98 Revision 0 l

l

~

RTS. Instrumentation 3.3.1 A.

Teole 3.3.. 1 (cage 4 c' 6 "eactor Trip System Instrw entation APP ICASLE MODES OR L

OTHER SPECIFIED-REOUIRED SURVE!LLANCE ALLOWABli FUNCTION CONDITIONS CHANNELS

. CON 0!TICNS REQUIREMENTS VALUE 15 Eeactc Trie j

System Interlocks l

M a,

Source Range Bicct 3

2 0

SR 3.3 1.1'.

a 6E.11 amo Sermissive. F.6 SR 3.3.1.1:

3 I

c, Lc= Po.er Geactc-Trips Block. F~

(1) P.10 Input 1

3-P SR 3.3.1 11 NA SR 3.3.1 1:

1

~

(h --

2) P.13 Input 2

SR 3.3.1;10 NA a

i SR 3 3 1.1:

c, Power Range 3

P SR 3.3.1.11 s 32.1% RTP Neutron Flux P.8 SR 3.3.1.12

. Power Range 1.2 3

0 SR 3.3.1 11

= 7.91 RTP and

.j Neutron Flux, F 10 SR 3.3.1.12 s 12.1% RTP l-e.

. Turbine impulse 1

2 F

SR 3.3.1.10 s 12.1 Pressure, P.13 SR 3.3.1.12 turd 1ne powea

.l'

18.. Reactor Trio 1,2 2 trains N

SR 3.3.1.4 NA Breakers (RT85)(9)

.pN 3(a) 4(a) 5(a) 2 trains C

SR 3.3.1.4 NA

/

i

-t-Q.

l : 19. Reactor Trip Breaker 1.2 1 each per RTB 0

SR 3.3.1.4 NA Undervoltage and Shunt

.l Trip Meer.anisms.

3(a) 4(a) 5(a)

I eacn per RTB C

SR 3.3 1.4 NA i

l

20. Automatic Trio Logic.

1.2 2 trains M

SR 3.3.1.5 NA

.l 3(a) 4(a) $(a) 2 trains C-SR 3.3.1.5 NA i

[ '(a)

With Rod Control System capable of red witnerawal or one ce more rocs not fully inserted.

)

( d)' Below the P.6 (Source Range Block Permissive) interlock.

-(g).

Including any reactor trip bypass creakers that are racked in and closed for bypassing an RTB.

I L'

l.

l t

m Q

BRAIDWOOD - UNITS 1 & 2 3.3.1 - 17 10/27/98 Revision 0 i'

I

~

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.4 Perform ACTUATION LOGIC TEST.

31 days on a STAGGERED TEST BASIS I

l SR 3.3.2.5 Perform MASTER RELAY TEST.

31 days on a STAGGERED TEST BASIS i

i SR 3.3.2.6 Perform COT.

92 days m

SR 3.3.2.7

. Perform SLAVE RELAY TEST.

92 days i

h SR 3.3.2.8 NOTE'

-Verification of relay setpoints not required.

Perform TADOT.

92 days NOTE SR 3.3.2.9

!p Verification of setpoint not required.

Perform TAD 0T.

18 months (continued)

BRAIDWOOD - UNITS 1 & 2

3. 3.2 - 7 7/9/98 Revision E

1 l

l ESFAS Instrumentation 3.3.2 l

O SURVEILLANCE REQUIREMENTS (continued)

Tj SURVEILLANCE FREQUENCY L-SR 3.3.2.10 NOTE--

This Surveillance shall include verification.that the time constants are adjusted to the prescribed. values.

Perform CHANNEL CALIBRATION.

15 months SR 3.3.2.11 verify ESFAS RESPONSE TIMES are within 18 months limit.

'SR 3.3.2.12 Verify ESFAS RESPONSE TIMES are within 18 months on a

. limit.

STAGGERED TEST BASIS

)

I BRAIDWOOD - UNITS 1 &.2 3.3.2 - 8 10/29/98 Revision 0

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

RTS Instrumentation B 3.3.1 BASES

' SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1 9 SR 3.3.1.9 is the performance of a TADOT every 92 aays, as justified in Reference 7.

The SR is modified by a Note that excludes verification of setpoints from the TADOT.

Since this SR applies to RCP undervoltage and underfrequency relays, setpoint verification requires elaborate bench calibration and is accomplished during the CHANNEL CALIBRATION.

SR 3.3.1.!0 A CHANNEL CALIBRATION is performed every 18 months. or l

approximately at every refueling.

CHANNEL CALIBRATION 1s a complete check of the instrument loop. including the sensor.

2 The test verifies that the channel responds to a measured

{'

parameter within the necessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the plant specific setpoint methodology. The difference between the current "as found" values and the

(],,

previous test "as left" values must be consistent with the

'g calculated-normal uncertainties consistent with the setpoint methodology.

~

The Frequency of 18 months is based on the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint methodology.

3~

SR 3.3.1.10 is modified by a Note stating that this test shall include verification that the time constants are i

.E adjusted to the prescribed values where applicable.

O BRAIDWOOD - UNITS 1 & 2 B 3.3.1 - 56 10/26/98 Revision 0 n

.r

,e

ESFAS Instrumentation B 3 3.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3 3.2 9 SR 3.3.2.9 is the performance of a TADOT.

This test is a

.C check of the Manual Actuation Functions and P-4 Reactor Trip CW Interlock.

It is performed every 18 months.

Each Manual Actuation Function is tested up to, and including. the master relay coils.

In some instances, the test includes actuation of the end device (i.e.. pump starts. valve cycles. etc.).

The Frequency is adequate, based on industry operating experience and is consistent with the typical refueling cycle.

The SR is modified by a Note that excludes gJJ l verification of setpoints during the TADOT.

The Functions have no associated setpoints.

SR 3.3.2.10 SR 3.3.2.10 is the performance of a CHANNEL CALIBRATION.

A CHANNEL CALIBRATION is performed every 18 months. or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor.

The test verifies that the channel responds to measured parameter within the necessary range and accuracy.

M CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the plant specific setpoint methodology.

The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.

M The Frequency'of 18 months is based on the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint methodology.

This SR is modified by a Note stating that this test should k

include verification that the time constants are adjusted to

'c the prescribed values where applicable.

O

'BRAIDWOOD - UNITS 1 & 2 B 3.3.2 - 54 10/29/98 Revision 0

r 4

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t e

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O TABLE 4.3-1 s.

s.- -

s 6-Se termos,oaly he ' \\ REACTOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE REQUIRENENTS

-final num ber ir shs,ie

)

3 l - => 52 3. 3.1.1 o-4 I

ANALOG ATING MODES FOR e 7 - =)

s st 3 3.17, e+c -

CHANNEL DEVICE WHICH b

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE' t

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED __

p I

-1.

Manual Reactor Trip N.A.

N.A.

N.A.

13-R(14 tol N.A. 1, 2, 3', 4', 5' [ 2. Power Range, Neutron' Flux a: High Setpoint I-S 1-Q N.A. N.A. 1, 2 '\\! z-D(2,4)h, T 3 - M(3, 4 o! A, c,- Q(4, 6) I ?' 1 - R(4, 5a) 7ni b. Low Setpoint I-S ii-R(4) E-Q-N.A. N.A. 1", 2 is[ JJaPower Range, Neutron Flux, High N.A. 11-R(4) 7-Q N.A. N.A. 1, 2 Qi j Positive Rate 't i A'Jb Power Range, Neutron Flux, High N.A. is-R(4) 7-Q N.A. N.A. 1, 2 [ Negative Rate E.4 Intermediate Range, Neutron Flux l-S 88-R(4, Sa) 6Q N.A. N.A. 1", 2 .f g E.5 Source Range, Neutron Flux SandLS .3an N R,(4, Sb) g(9) N.A. N.A. 2",3,4,5 3 h I.G Overtemperature AT l-S io-R 7-Q N.A. N.A. 1, 2 E.7 0.verpower AT l-S 'o-R 7-Q N.A. N.A. 1, 2 [ SbaPressurizer Pressure-Low l-S lo-R 7-Q N. A'. N.A. I (Above P-7) 188b Pressurizer Pressure-High I-S to-R 7-Q N.A. N.A. 1, 2 { H 3 Pressurizer Water Level-High t-S lo-R 7-Q N.A. H.A. I { (Above P-7) {st 3 3.1.o ibc - hm Luc.He'e_. s b*n wwdc ' Ag } j Vuontanm Tum vie mme r ~ w 3 n g ;i .O consFa nfs .4 e c Ai vwea O g BYRON - UNITS I & 2 W T"' f umb*1 u n \\v s. /4 3-9 ANENONENT NO. 55 e os " 3 Ln - m 4 C i rx )

.$.3.1-1 TABLE d i t (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4 -1 RIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE inflCH t CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE I FilNCTIONAL UNIT CHECK .CAllBRATION TEST TEST LOGIC TEST IS REQUIREQ_. i !l J, tic Reactor. Coolant Flow-Low 1-S 10 - R 7-Q N.A. N.A. I M14 Steam Generator Water Level-Low-l- S to-R 1-Q N.A. N.A. 1, 2 ll Low-i Mt2. Undervoltage-Reactor Coolant N.A. lo-R N.A. 7-Q(10) N.A. 1 l Pumps (Above P-7) 1513 Underfrequency-Reactor Coolant N.A. 10-R N.A. 1-Q(10) .N.A. 1 Pumps (Above P-7) NL5 Turbine Trip.(Above P-8) [

a. Emergency Trip Header N.A.

lo-R N.A. It( - S/U(1, 10) N.A. 1 [ Pressure q

b. Turbine Throttle Valve N.A.

10- R N.A. 14 - S/U(1, 10) N.A. l' Closure %[ R IG Safety Injection Input form ESF N.A. M.A. N.A. 13 - t-78 N.A. 1, 2 Jfril Reactor Coolant Pump Breaker N.A. N.A. N.A. 13 - R O A N.A. 1 Position Trip (Above P-7) b -\\ -itr7 Reactor T,rigSystem Interlocks

a. Ints.udiate.Rarige Neui.run N.A.

Il-R(4) i t-R N.A. N.A. 2" F+tm, P-6 (A zq )

b. Low Power Reactor Trips N.A.

11-R(4) iz-R N.A. N.A. I q l' Block, P-7 lU

c. Power Range Neutron Flux, P-8 N.A.

Il-R(4) 12.- R N.A. N.A. 1 1.sc $3.1 s M T 5 M if M W u w.iCJi N y %Aarion wr. w & ) Az r BYRON - UNITS 1 & 2 \\ % """ ^ a id e- - I 3/4 3-10 AMENDMENT NO. 55 J j h me raam.i <a nim. o (%- /

_ _. DW TABLE 44 Continued) p0 (s./ ~ 1 ENGINEERED SAFETY FEATURES ACTUA SYSTEM INSTRUMENTATION TRIP SETPOINTS / ALLOWABLE TRIP FUNCTIONAL UNIT SETPOINT VALUE i

6. Auxiliary Feedwater(Continued)

Lu ov ocvsm: bx d.-f- ~ 11 for Unit 1 L 81, - ~ ~ p' ( (Divi f ) -Driven 287 volts 2730 volts try t O. g-Auxiliary Feedwater Pump Suction Pressure-Low (Transfer to A Essential Service u Water) ( 218. psia 217.4 psia ,g,, 1

7. Automatic Opening of Containment Sump Sistion isolation Valves w

a. AutomaticActuation p' -' g-5 o Logic and Actuation in3rument f @ N.A. sp N.A. Relays R [ b. RWST Level-Low-Low 1 Coincident with l Safetyinjection See item 1. above for Safety injection Trip Setpoints and Allowable Values. v. i s ll h AMENOMENT N0.10 104 ,f ON - UNITS 1 & 2 3 27 O I .m m.

e- ' h,. jR mar up N, o M 1-l ' TABLE 4-3-e i j bg ) ebu h 9 e *

• I I

'> 5g 3 3 2.) o 3-ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION -\\ SURVEILLANCE REQUIREMENTS 3 > Sg 3.3. 2S e4c~ - i TRIP u ANALOG AC1UATING HODES CNANNEL DEVICE ACTUATION stASTER SLAVE FOR INIICN CNAIIIIEL CNAINIEL WERAll0NAL OPERAfl0NAL 10Glc RELAY RELAY SURVElttANCE FlatCT AN6AL UNif CNECK CAtlBRAflou Test TEST TEST TEST TEST 15 RENUIRED_ 1. !*sefety Injection (Reactor irlp. Feeduster Isolation, Start Sleset I Generators, Contalrunent Cooting i Fene, Control Room teoletion, Phase "A" lootetton, furbine irlp, i Aunttlery feedeeter, Containment vent teoletion and Essentist Service Weter) e. Manuet initiation N.A. N.A. II. A. 9-R N.A. N.A. N.A. 1,2,3,4 b. Autoestic Actuation Logic II. A. N.A. N.A. II. A. 4-M(1) fr'- M(1) 7-e 1,2,3,4 and Actuotton Releye. ? c. Conteltment Pressure-NfWi-1 l-S // -R (7-6 N.A. N.A. N.A. N.A. 1, 2, 3 l d. Pressuriter Preneure-Lois 35 /O-R ' h-e ll. A. M.A. N.A. fl. A. 1,2,3 l f (Above P-11) e. Steen Line Pressure-Loss. I-S /O-R 47 - 6 ' II. A. N.A. N.A. N.A. 1, 2, 3 l { (Above P-11) r 2. Conteltment Spray I

• e.

Manuel initletion II. A. N.A. II. A. fR ll. A. II. A. W.A. 1,2,3,4 b. Automatic Actuetlen Logic N.A. II. A. II. A. II. A. Ej -M(1) f M(1) 7-0 1, 2, 3, 4 and Actuation Reteye i. c. Contaltment Pressure-liigh 3 1-S /O - R f,- e N.A. N.A. II. A. al. A. 1,2,3 l 3. Conteltment teolation s. Phase "A" Isolation rW Of 1) Meruset Initletion ll. A. II. A. M.A. $R N.A. W.A. II. A. 1, 2, 3, 4 (4t 2) Automatic Actuetten N.A. N.A. N.A. N.A. y M(1) f M(1) 7-0 1, 2, 3, 4 g[ togle and Actuation i Releye Ni ' { rvue.a r 5 ft 3 3 2.lo Ac V A 3 L7 e I ' BYRON - UNITS 1 & 2 3/4 3-34 mruriuror un cc

( 3.72-1 TABLE M -2 iContinued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE' REQUIREMENTS TRIP AllALOG ACTUAilNG lumES CilANNEL DEVICE ACTUAft0N MASTER SLAVE FOR WWICN CMAlHIEL CNANNEL OPERATIONAL OPERATIONAL LOGlc RELAY RELAY SURVEILLANCE FUNCi}0NALLNilf CNECK. CAllBRAfl0N TEST TEST Test TElf... J1H_ IS REGUIRED 3.e.** Phase "k" lootation (continued) 3) Safety Injection Sec Item 1. above for att safety Injection Surveiltence Rewirements. b. Phase "B" Isotation 1) Manuel initiation N.A. N.A. N.A. (-R II. A. N.A. N.A. 1, 2, 3, 4 2) Automette Actuetlen N.A. N.A. N.A. N.A. '/- M(1) [-M(1) 7-4 1, 2, 3, 4 Logic Actuetten Relaye l 3) Contelrusent Pressure.

1. S

/p-R f,- e N.A. N.A. N.A. N.A. 1, 2, 3 Nigh-3 LCO J, Contaltenent Vent teoletten $.39 Automatic Actuetlen N.A. N.A. N.A. N.A. 343 3 & 2-M(1) Sp.3 313 3. M(1) W 7 L5e 1,2,3,4 Logle and Actuetlen Retere lE Manuel Phase "A" See Item 3.e.1 ebove for att menuet Phase "A" lootetton Survelttence Reviremente. lootetton .:2M Manuel Phase "S" See item 3.b.1 above for att manuel Phase "s" leoletion Survelliance Requirements. lootetlen I -41' safety injection See item 1. above for ett safety injection Surveltlance Re m irements. 4 Steam Line footetton e. Manuel Initletion N.A. N.A. II. A. 4-R N.A. N.A. N.A. 1,2,3 b. Automatic Actuation Logle M.A. N.A. N.A. N.A. e/. M(1) 6'- M(1) '7-Q 1,2,3 I~ l and Actuotton Reteye gll(o c. Contelrunmt Preneure-Nigh-2 5-S fo-R f- 0 N.A. M.A. M.A. M.A. 1, 2, 3 d, g# Steam Line Pressure-Lou l-S g-R [, - 4 N.A. N.A. N.A. N.A. 1, 2, 3 4 (Above P-11) O IA y (7 Ducer S P 3.3.2.lo ador]e 1 3 BYRON - UNITS I & 2 3/4 3-35 AMENDNTNT NO. 55 ..-..t...

( i '3,7.;2-1 TABLE 4.1 2 (Continued 1 LNGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION N SURVEILLANCE REQUIREMENTS IRIP AllALOG ActuAllNG NODES CllANNEL DEVICE ACTUAf!DN MASTER SLAVE FOR UNICN CHANNEL CHANNEL OPERAllONAL OPERAi!ONAL LOGIC RELAT RELAf SURVEILLANCE FUNCi b AL tNili CHECIC CAtteRAfl04 TEST TEST TEST TEST 1[$1_ 15 REeulRED 4 Steam Line Isolation (continued) d hf Steam Line Pressure - l+ $ /#- R 6-0 N.A. N.A. N.A. N.A. 3 f Negettve Rate - Nigh (Belou i P-11) 5. Turbine irlp and Feedwater lootetton e. Automatic Actuation Logic N.A. II. A N.A. N.A. tf-N(1) f-N(1) ~7-e 1, 2 I and Actuatim Relay b. Steen Generetor Water levet-1. S 80- R [r- 0 II. A. 8l-N(1) f-M(1) 7"e 1, 2 Nigh-Nigh (P-14) c. Safety injection See item 1. above for att Safety injection Surveillance Reptrements. t 6. Auntilary Feedester e_ 1 M !n!tiet!=

1. _.^.

A N;A. E.A. E.i 17273- [kg (1 I Automette Actuetten Logle M.A. N.A. N.A. N.A. Al-N(1) g-M(1) 7-e 1,2,3 and Actuation Relay i b..ef" Steam Generator Water Level-4-S /O - R [f- 0 N.A. N.A. N.A. W.A. 1,2,3 l' Low-Low eg Undervoltage-RCP Bus N.A. lO-R ll.A. 7-0(3) N.A. N.A. N.A. 1, 2 I C p/' Safety injection See item 1. above for o!! Safety Injection Surveillence Re@ lrements. I+% & O%'.W ~skwe i d [. ivls 11 for t1 II. A. /F-R N.A. 3-N(2, 3) N.A. N.A. N.A. 1,2,3,4 (of slon 21 Unit 2 SF undervo see h' Auxillery Feedseter Ptap J-S /A" R 2M N.A. N.A. N.A. M.A. 1,2,3 Suction Pressure-Lou n! " 3 R 3 3.7. 10 %] [l (@ r, e p# BYRON - UNITS 1 & 2 3/4 3-36 AMENDMINT No. 55

3.3.2-t TABLE 4-3=f (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUAIING MODES CNAIIIIEL DEVICE ACTUAfl0N MASTER SLAVE FOR WHIEN CNANNEL CHANNEL OPERAffollAL OPERATlollAL LOGIC RELAY RELAY SURVEILLANCE FUNCil($1ALUNIT CNECE Call 5RATICII TEST TEST TEST TEST TE1L is AtauiREo 7. Automatic Opening of Contaltesent stop Suction leotation Vetwee e. Automatic Actuetten Logic M.A. M.A. II. A. M.A. b M(1) f-M(1) 1-e 1,2,3,4 and Actuation Releys l b. RUST Level-Lou-Lou l -. 5 jo-R /,- G N.A. N.A. N.A. M.A. 1,2,3,4 j Colncident With Safety injection see item 1. above for ett safety injection Survelltence RegJirements tro 33d X. Loss of Pouer 5 o n. ESF Sue Undervettege N.A. R - SR 13.5.2 N.A N(2, 3).jst331r 3 N.A. N.A. N.A. 1,2,3,4 b. Grid Degraded Voltage N.A. R - SR3.3. f.1. N.A. M(3) 5A33 5".1 N.A. II. A. M.A. 1,2,3,4 4 9. Engineered Safety feature Actuation System Interlocke e. Pressurizer Pressure, P *1 N.A. (e-R (p- 0 N.A. N.A. N.A. N.A. 1,2,3 l [ b. Reactor Trip, P-4 N.A. N.A. N.A. 7-R N.A. N.A. N.A. 1,2,3 c. Low-Lou T. P-12 N.A. lo _ R h0 N.A. N.A. W.A. M.A. 1,2,3 l I TABLE NOTATION RT Each train shall be tested at least every EB days on a STAGGERED TEST BASIS. Undervoltage relay operability is to be verified independently. An inoperable channel inay be bypassed for up to 2 hours for surveillance testing of the OPERABLE channel per Specification 4.3.2.1. F37 etpoint verification is not applicable. l SA 3.% 2.7 tbte ~ ~ ~.. _ - - - 3R.3.3.5.8 Nore SR 3.3. 2,C M3TE. 93 e.m s c R.3.2 'O g ,r ; Ico 3 35 RA A.I Al r O Ol W 5.52. A4 F.l /&te Lu M~ SR 13.ZA NOTE- ~TA15Es12_T.3.3 - J 74 {~ O RYRON IINITS 1 17 3/4 % 37 AMrnnMrur Hn rA

i BRWD CTS MARKUPS O O l O

7O. O V 33.1-I TABLE & F b ~ REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS L SPs varmos,wly -lhe Li.ia l ' numbe h h, it.., TRIP ANALOG ACTUATING MODES FOR l - ) se 3 3.l.1.s CilANNEL DEVICE WHICH 7 J> S R 3'3' 1' 7 e 4 "- CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE 1; -? ' FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TESI IS REQUIRED-1, 2, 3*, 4*, S' f l '. Manual Reactor Trip N.A. N.A. 'N.A. 13-R(14) (lo) - N.A. i 2. Power Range, Neutron Flux g, . i l a. High Setpoint t-S J-D[2,4) 7-Q N.A. N.A. 1, 2 i (3-ML3,4f 9 4 ~~7 t-- Q(4, 61),~) II n R(T; 5a b. Low Setpoint 1-S u-R(4) 3-Q N.A. N.A. 1", 2 \\t i ,3'.hPower Range, Neutron Flux, High N.A. 0- R(4) 7-Q N.A. N.A. 1, 2 Positive Rate A'.3bPower Range, Neutron Flux, High N.A. Il R(4) 7-Q N.A. N.A. 1, 2 l Negative Rate E.4 Intermediate Range, Neutron Flux I-S 11' R(4, Sa) E-Q N.A. .N.A. 1", 2 lu l N.A. N.A. 2", 3, 4, 5 }l lyR4,5b) -](9) J ,6'.5 Source Range, Neutron Flux ,5 t-S 10-7-Q N.A. N.A. 1, 2 J.'b overtemperature AT E.7 Overpower AT I-S to-R 7-Q N.A. N.A. 1, 2 l J.T.sPressurizer Pressure-Low l-S 10 R 7-Q N.A. N.A. I l (Above P-7) 10/1 Pressurizer Pressure-High 1-S to R 7-Q N.A. N.A. 1, 2 8C l' if.lPressurizer Water level-High l-S 10 - R 7-Q N.A. N.A. I rr (Above P-7)

p. u

[': Sit 3 3.l.s o l*3OE" - h Swwe.dNcZTm 'j U l

• • ~ ~ ~ ~ r,w #2ia,de n'

g go 4 BRAIDWOOD - UNITS 1 & 2 N$ NIQ1Q Oa AMENDMENT NO.44

() V(Q p V d

3. 3.1-l TABLE G fContinued)

REACTOR TRIP SYSTEM INSTRUNENTATION SURVEILLANCE REQUIRENENTS TRIP ANALOG ACTUATING NODES FOR CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED- .lf.fc Reactor Coolant Flow-Low l-S 8* R 7 Q N.A. N.A. 1 ,13'J4 Steam Generator Water level-1-S lo-R 7-Q N.A. N.A. 1, 2 f low-Low l J4'.12 Undervoltage-Reactor Coolant N.A. 10- R N.A. 't - Q(10) N.A. 1 Pumps (Above P-7) HJ.'Underfrequency-Reactor Coolant N.A. 10 R N. /.. 'i-Q(10) N.A. I 3' Pumps (Above P-7) ,16M Turbine Trip ('Above P-8) l [

a. Emergency Trip Header N.A.

lo-R N.A. I'{ - S/U(1, 10) N.A. 1 Pressure

b. Turbine Throttle Valve N.A.

80-R N.A. 14-S/U(1, 10) N.A. 1 Closure 17JfoSafety Injection Input from ESF N.A. N.A. N.A. 13 - Gol N.A. 1, 2 .18.11 Reactor Coolant Pump Breaker N.A. N.A. N.A. 13-R OO 'u N.A. 1 I 2 Position Trip (Above P-7) 49.r7 Reactor Trip System Inter 1 gig

a. inn M iale been N.A.

ll-R(4)

1. 2-R N.A.

N.A. 2" Fhx, P-6 g Kl

b. Low Power Reactor Trips N.A.

II-R(4)

1. z-R N.A.

N.A. 1 h; Block, P-7 O'

c. P_ower Range Neutron Flux, P-8 N.A.

11-R(4) iz-R N.A. N.A. 1 33l i ij} @ 31.{-s o Were-h Swe.e.ue<uc_ a uw "8 mcLu se_ pn 1 m.n <o m._ m m BRAIDWOO,0 - UNITS 1 & 2 7 [ j g g y'(* g } 3/4 3-10 AMENDMENT NO. 44 i

3.3,2-l - 'O TABLE 3.- (Continued) U ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS \\ TRIP ALLOWABLE SETPOINT VALUE FUNCTIONAL UNIT 6. Auxiliary Feedwater(Continued) (dis, Ortsms bea. dg ~ ion 11 for Unit M, ~ ff (Dhiis f . et 2) pg c ESF B age-Motor-Driven 287 volts 2 730 volts A .Q y Auxiliary Feedwater Pump Suction Pressure-ly m An Low (Transfer to Essential i c Service Water) 218.1 s' 217.4 psia 7. Automatic Opening of 'I .s Containment Sump Suction isolation Valves a. AutomaticActuation .A. N.A. Logic and Actuation Relays g g m3rit am en t Q \\ 44.7 % D g b. RWST t'evel-Low-Low 46.7 % Coincident with Safetyinjection See item 1. above for Safety injection Trip Setpoints and Allowable Values. 't ) Y Y t O E DWOOD-UNITS 1 & 2 3, 27 AMENDMENT. 96; 96 O m-..~_..m

,\\ O b) 23.L-l TABLE m 5,- sa ur..g y\\3 Q n,.b.<- a s w a i e, ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS \\ _ <> 3R 3 3 2.1 M %. d) qR 3 3. 2 3 e 4* - TRIP g ANALOG AC1UAIING N0DE5 CNANNEL DEVICE ACTUATION MASTER SLAVE FOR IMllCN CHANNEL CHANNEL OPERAtl0NAL CPERAfl0NAL LOGIC RELAY RELAY SURVERLLANCE FUNCfl0NAL UNIT CNECK CAlltaAfl0N 1897 1EST TEST TEST _ ILSI_ IS REeulRED 1. Safety injection (Reector trip, Feedwater teolation, Start Dieset Generatore, Conteirenent Cooling Fene, Control Room Isoletion, Phase *A" leolation, Turbine frlp, Auxillery Feeduster, Conteltsment Vent teoletion and Essentlet t Service Water) l e. Manuel Initletion N.A. N.A. N.A. f-R N.A. N.A. N.A. 1, 2, 3, 4 b. Autountic Actuetlen Logle N.A. N.A. N.A. N.A. t/ --M(1) y M(1) 7-e 1,2,3,4 l end Actuation Relays C. Contelryment Pressure-Nigh 1 l-S lp-R [,- G N.A. N.A. N.A. M.A. 1,2,3 l d. Preseuciter Pressure-Lou lS lp-R (,- N N.A. N.A. N.A. N.A. 1,2,3 (Above P-11) e. Steam Line Pressure-Low l-3 /p. R g4 N.A. N.A. M.A. N.A. 1,2,3 (Above P-11) f 2. Centalesment Sprey e. Manuel Inittetlen N.A. N.A. N.A. fR N.A. N.A. N.A. 1, 2, 3.-4 { b. Automette Actuation Logle N.A. N.A. N.A. N.A. Y -M(1) $;M(1) 7-0 1, 2, 3, 4 and Actuation Reteye c. Contelrunent Pressure-Illsh 3 1-5 fo R 6-4 N.A. N.A. N.A. N.A. 1,2,3 3. Centelrunent teoletion e. Phase "A= lootetton 1) Manuet Inittetter; N.A. N.A. N.A. 7-R N.A. M.A. N.A. 1,2,3,4 l t 2) Automatic Actuellen N.A. N.A. N.A. N.A. '[-M(1) /:' M(1) 7-e 1, 2, 3, 4 g f LoSle and Actuetten p i Relays O ! ^% Y" I hny se 3.t.L.su dTt [# 69 i BRAIDWOOD - UNITS 1 & 2 3/4 3-34 AMENDMENT NO. 44

p ( b 3 3. !: 1 TABLE -4r3-E (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATLON SURVEILLANCE REQUIREMENTS 1 RIP ANALOC AttUATING. MODES CNANNEL DEVICE ACTUATION MASTER SLAVE FOR initCN CMANNEL CMANNEL OPERAll0NAL OPERATIONAL LOGIC RELAY RELAY SURVEILLANCE FUNCfl0NAL UNIT CMECI( __ CAllBRAfl04 TEST TEST TEST TElf TEST IS REeUIRED. 3.s. Phase "A" leoletion (continued) 3) Safety Injection See itese 1. above for ett Safety injection Survelltence Requiremente. b. Phase '9" Isoletion 1) Manuel Inttletten N.A. N.A. N.A. k-" R N.A. N.A. N.A. 1,2,3,4 2) Automette Actuetten N.A. M.A. N.A. N.A. Y -M(1) M M(1) Ne 1,2,3,4 Loele Actuation Releye 3) Contalrunent Pressure-l-S /O-R 6-0 N.A. M.A. N.A. M.A. 1, 2, 3 L c o 3. 3..'.- .r:/ Contalrsment vent f eeletten 3 J)" Autemetle Actuetten R.A. N.A. B.A. N.A. 3* lp 1 - M(1) SR3 3 6.3 M(1)$93%fe 1,2,3,4 togle and Actuotten Releye l.Jt Manuel Phase "A" - See item 3.s.1 above for ett manuel Phase "A" leolatlen Surveltlence Re ptremente. lealetion '2. M Manuel Phase "B" See item 3.b.1 above for ett moroset Phase "0" leolation Surveltlance Requirements. lootetten b [Af~ Safety injectlen See Itee 1. above for ett Safety injection Survelltence Rew iremente. 4. Steen Line teolation s. Manuel Initletion N.A. N.A. N.A. f-R N.A. N.A. N.A. 1,2,3 b. Automette Actuetten Leste N.A. N.A. II. A. N.A. y-M(1) 5"-M(1) 7-0 1, 2, 3 and Actuation Releys 8Ol c. Contalrunent Pressure-Mi@-2 l-S /#' R [-G N.A. M.A. N.A. N.A. 1, 2, 3 d, l A' Steam Line Pressure-Lew l-S .U R [a - t N.A. N.A. N.A. N.A. 1,2,3 y,g (Above P 11) a w; 4 'J.VOe. I'LT S R. 3 3 2. t 0 A O Qri 1I : o BRAIDWOOD - UNITS 1 & 2 3/4 3-35 AMENDHENT NO. 44

3.32-1 TABLE-4-9-2 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOS. ACTUATING ItucES CNANNEL DEVICE ACTUATION MASTER SLAVE FOR WHICN CHANNEL CHANNEL OPERATIONAL OPERATIONAL LOGIC RELAY RELAY SURVEILLANCE FUNCit0NAL UNIT CNECI( CALIBRAfl0N TEST UST TEST TEST HSJ_ 15 REeUIRED 4. Steam Line f ootetlen (contirwed)

d. Lac Steam tine Preneur -

1-S lo-R G-e N.A. N.A. N.A. N.A. 3 Negettve Rete - Nigh (Selou P-11) 5. Turbine Trip and Feeduster footetton e. Autemelle Actuellen Legle N.A. N.A. N.A. N.A. Y -M(1) [- M(1) 7-0 1, 2 and Actuation Retey l b. , Steam Generator Water Level-l-S N -R d-G N.A. f - M(1) f-M(1)- 7-0 1, 2 Nigh-Nigh (P-14) c. Safety injection See itse 1. above for att safety injectlen Survelttence Requiremente. 6. Atmittery Feeduster _ N A.. m E.A. A. N.A 1, 2, 3 I e,-~ Manuel-Init tet!r N.A. 21 M Automatic Actuotton Lesle N.A. N.A. N.A. N.A. y -M(1) g M(1) 7-e 1,2,3 and Actuotton Retey bp Steam Generator Water Levet-l-S fp-R Me M.A. W.A. N.A. N.A. 1,2,3 Low-Lou e.4/ Undervottege-RCP Sue M.A. D-R N.A. b O(3) N.A. N.A. N.A. 1, 2 C s: Safety injection See item 1. above for ett Safety injection Surveltlance Requiremente. tou of oWeite fosMs,- dJf Il for UV' N.A. /O-R N.A. 8-M(?,3) N.A. N.A. N.A. 1,2,3,4 (Olvlel Unit 2) ESF rvettese h.gf Atstillery Feedseter Ptap l-S IO - R 2-Il N.A. N.A. N.A. N.A. 1,2,3 g Suction Pressure-Low 3 Qnanse.3.7,eo rhw { 'rg u y, (n (D y 9 BRAIDWOOD - UNITS 1 & 2 3/4 3-36 AMENDHENT NO. 44

7-x v () ~ .3.5. z l TABLE 4d (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG AttUATING seuDES CNANNEL DEVICE ACTUATION MASTER SLAVE FOR ifMICE CNANNEL CNANNEL OPERAll0NAL OPERATIONAL LOCic RELAY RELAY SURVEILLANCE FUNCTIONAL UNif CNECE _ CAllORAfl0Il TEST TEST TEST TEST IEE. IS REGUIRES y. Automette Opening of Contelrument Staup Suetion Isolation Vetwee e. Automatle Actuation Legle N.A. N.A. N.A. N.A. k-- M(1) 8-M(1) 7-N 1,2,3,4 and Actuation Relays b. RWST Level-Low-low l-5 /O R 4,- G N.A. N.A. N.A. N.A. 1,2,3,4 Colncident With Safety injection See Ites 1. above for ett safety injection Surveittence Requiremente J. Loes of Power a. Esf Nun Undervottoge N.A. R -SR 3 3.f.1 N.A M(2, 3P$RM F 8 N.A. N.A. N.A. 1,2,3,4 b. Grld Degraded Vettege N.A. R -SP 13 5.1. N.A. M(3)-SRn5 I N.A. N.A. N.A. 1,2,3,4 9. Engineered Safety feature Actuation System Intertecits e. Presourf ter Pressure, P-11 N.A. f 0- R (o - 0 N.A. N.A. N.A. N.A. 1,2,3 e b. Reactor Trip, P-4 N.A. M.A. N.A. 9-R N.A. N.A. N.A. 1,2,3 c. Low low T. P-12 N.A. 16 R 6-4 N.A. N.A. N.A. N.A. 1,2,3 gi TABLE NOTATION R) Each train shall be tested at least every days on a STAGGERED TEST BASIS. 2T Undervoltage relay operability is to be verified independently. An inoperable channel may be bypassed for up to 2 houre [/for surveillance testing of the OPERABLE channel per Specification 4.3.2.1. [p) etpoint verification is not applicable. O \\ R 3.3. 6.1 tJote,) \\ 5t 3.3.2.8A 7.s..t 3 t4 ate >> { ,w s q 3 3,7,,fo ,g4, A.)e OO < Lco 3.3 5 RA A,l NoTc ) Leo 3.Y.2 6A F. I tMa27 " ' ' * ~ y' SR 3. 3 2.1 #bTE e lldSElrr 3 3 - 37A e GRAIDWOOD - UNITS 1 & 2 3/4 3-37 AMENDHENT NO. 44

k' bh# de M M Me m,p m gg_ ~~BM- t. ( t r i CTS. DOCS [g i l l-l i. b: l i 1 O

. ~. _. DISCUSSION OF CHANGES TO CTS i ( ITS SECTION 3.3 INSTRUMENTATION 3 k g.! A By letter dated October 6. 199S. Comed received License Amendment n 104 (Byron) / 96 (Braidwood) which changed CTS Table 3.3-4 Functional Unit 6.g. Auxiliary Feedwater Pump Suction Pressure - Low. Trip Setpoint and Allowable Value (and LC0 3.7.1.3 to raise the CST level, which is addressed ir another ITS Section). The clouded portions of the CTS l markup reflect this change. This change was requested as a result of l Comed identifying an operability concern involving the postulated i l failure of Safety Category II CST piping in the turbine building during i a seismic event. This postulated failure of the non-seismic piping N l could eventually result in atmospheric pressure (14.7 psia) in the AF C suction line. This would minimize the potential for an automatic switch over of the AF water supply from the CST to SX water, since the previous 9 Trip Setpoint value was 14.1 psia (1.22" Hg vac). In response to the N operability concern, the minimum administrative CST level and the 1 M physical height of the CST were raised. Additionally, the above I mentioned CTS License Amendment increased the AF pump Suction Pressure - 1 ai Low Trip Setpoint and Allowable Value to greater than or equal to 1 3 18.1 psia and 17.4 psia, respectively. as well as increasing the minimum required CST level. These setpoints ensure that the automatic switch over of the AF supply would occur when required, j A CTS Table 3.3-1 Actions 10 and 13 are revised (see also DOC L ) to include a Note precluding a MODE change into MODE 5 with the kod Control n iO System capable of rod withdrawal or all rods not fully inserted for ITS Functions 18. 19. and 20. This Note is inserted in conjunction with a revision to LCO 3.0.4 (refer to Section 3.0 DOC ls). As a result of the-change to LCO 3.0.4 all ITS Actions were evaluated for individual acceptability of this change. Based on this evaluation where MODE change restrictions were determined to be required in MODES 5 and 6. or W in MODES 1. 2. 3. and 4 during unit shutdown Notes containing the appropriate MODE change restrictions are added to the individual Specifications. The Note that is added to ITS 3.3.1 Condition C is a result of this evaluation. Since the technical aspects of this change are addressed in Section 3.0. this change is considered administrative in this Section. N A CTS and STS use the term " Intermediate Range Neutron Flux" for the P-6 3 0 RTS interlock. In ITS, the P-6 interlock is referred to as " Source .l Range Block Permissive." This change was made so that the TS agree with 6 plant design and terminology. The Byron /Braidwood Main Control Room annunciator and Bypass Permissive Panel windows, as well as plant g procedures, reference " Source Range Block Permissive" for the P-6 4 interlock. This change is considered editorial in nature and does not g involve a technicai change (either actual or interpretational) to the TS. BYRON /BRAIDWOOD UNITS'1 & 2 3.3 Sa 10/27/98 Revision 0

DISCUSSION OF CHANGES TO CTS ITS SECTION 3.3 INSTRUMENTATION L lb l A CTS Table 2.2-1 for Functional Unit 19.b (P-7) lists the inputs into 3 ,c l P-7. namely Functional Unit 19.b.1 (P-10) and Functional Unit 19.b.2 (P-13). Functional Unit 19.b.1 and Functional Unit 19.b.2 have been 7.! deleted since they are redundant to CTS Table 2.2-1 Functional Units t ; 19.d (P-10) and 19.e (P-13). The Allowable Values in CTS Table 2.2-1 .d ! for Functional Unit 19.d and Functional Unit 19.e have been retained in j ITS Table 3.3.1-1 for Function 17.d (P-10) and Function 17.e (P-13). % i During this reformatting. no technical changes (either actual or 1 interpretational) were made to the TS unless they were identified and justi fied. The change is consistent with NUREG-1431. l A In CTS Table 3.3-3. Functional Unit 1.d for " Pressurizer Pressure-Low 3 l (Above P-11)" and Functional Units 1.e and 4.d for " Steam Line Pressure-Low (Above P-11)" have an Applicability of Modes 1. 2. and 3# where footnote (#) states that the trip function may be blocked in Mode 3 below the P-11 setpoint. Functional Unit 4.e for " Steam Line Pressure Negative Rate - High (below P-11)" has an Applicability of Mode 3## where footnote (##) states that the trip function is automatically blocked above P-11 and may be blocked below P-11 when the steam line pressure-low SI is not blocked. In ITS Table 3.3.2-1. Function'1.d for " Pressurizer Pressure - Low" and Function 1.e for " Steam Line Pressure - Low" have an Applicability of (c) Modes 1. 2. and 3(a) where footnote (a) states above the P-11 interlock, Function 4.d.1 for " Steam Line Pressure - Low" has an Applicability of v D Modes 1. 2(g), and 3(a)(f)(g) where footnote (a) states above the P-11 interlock. footnote (g) states except when all MSIVs and MSIV bypass y ti valves cre closed, and footnote (f) states below the P-11 interlock with Qg Function 4.d.2 not enabled. Function 4.d.2 for " Steam Line Pressure tj Negative Rate - High" has an Applicability of Mode 3(d)(g) where footnote (d) states below the P-11 interlock with Function 4.d.1 blocked. This change is necessary due to the reformatting of the requirements j contained in the ITS. This change is perceived as the intent of the CTS i wording, is considered editorial in nature and does not involve a technical change (either actual or interpretational) to the TS unless otherwise noted. A CTS Table 4.3-1 Functional Units 7 through 16. and CTS Table 4.3-2 n Functional Units 1 through 7 and 9. have been revised by a Note stating. 3 "This Surveillance shall include verification that the time constants 'h C-- are adjusted to the prescribed values." This Note is consistent with current operating practices and is therefore consistent with NUREG-1431. L l BYRON /BRAIDWOOD UNITS 1 & 2 3.3 5b 10/29/98 Revision Q l l f

O A4yA--6mv SR 3.3.2

---

NOTE--------------------

r Verification of setpoint not required diec IIR A E E 1 E mE E IIE I I EAEE E WE Perform TADOT. g18gmonths i -SR 3.3.2

---

NOTE-------------------

This Surveillance shall include 1 fS verification that the time constants are

• O

E Q

adjusted to the prescribed values. 4 Perform CHANNEL CALIBRATION. {18Amonths J r \\ Ml2 SR.3.3.2. W TE---- No requir d to b perfo for th t rbine iven pump til [24] ours 04 fter press e is it 000] psi. l td Verify ESFAS RESPONSE TIMES are within 418jLmonths on 3

limit, a S'AGGERED v

TEST 8 ASIS t Q'(A 3.3.2. ll Mth 65F45 RE5904M 77ME5 on. WiMiri ( (continued) i limih. IEmm6 i 99 O WOG STS 3.3-30 Rev 1, 04/07/95 R :- V O e

M-4 d4 4dhs4mL M -hs AMJ m M e +,4 ef d4 y-4 deAW-*h b h p 4.=8 6e.d#dM Jh4 W aeA4hua M A*p. M144m S-^4h.S h,ma d.46e6 Mcm7 ~we -'M.4 hee h,4e me s.hs h e gg,ma,. ,i,,,mge.%.e. ee LCO JFDS l t b a i E ~ ,,,,,n-,

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS O SECTION 3.3 INSTRUMENTATION v Ps ITS SR 3.3.3.2 Note for PAM Instrumentation exempts neutron detectors 3 from a Channel Calibration. This Note is revised to replace " neutron detectors" with " radiation detectors." None of the PAM Instrumentation listed either in CTS Table 3.3-10 or in ITS Table 3.3.3-1 contains any neutron detectors. Revision of this note is consistent with the plant design and consistent with ;he general intent of NUREG-1431. Ps. ITS 3.3.1 Conditions 0 and P. and ITS LCO 3.3.2 Condition L is revised to reflect Condition entry based on "one or more" inoperable interlock channels. This is consistent with the CTS allowance for these actions. h) Ps, ITS Table 3.3.1-1 Function 17.b for the P-7 Low Power Reactor Trip Block is split into two line items in order to reflect the (as in CTS) the 3 o difference in the number of channels for this interlock depending on E whether referring to the P-10 input into P-7 or the P-13 input into P-7. P, NUREG Table 3.3.3.1-1 Function 18.d (P-9 Interlock) is deleted The 3 Byron and Braidwood design does not include this interlock. 3 b[ P Not used. g P. This Note is inserted in conjunction with adopting NUREG LC0 3.0.4 (refer to NUREG LC0 3.0.4 Reviewer's Note). As a result of adopting p) LCO 3.0.4 all ITS Actions were evaluated for individual acceptability (' of any increased flexibility beyond CTS allowances. Based on this y evaluation where MODE change restrictions were determined to be required in MODES 5 and 6. or in MODES 1, 2. 3. and 4 during unit shutdown. Notes d containing the appropriate MODE change restrictions are added to the individual Specifications. The Note ITS 3.3.1 Condition C is added as a e result of this evaluation. o O V BYRON /BRAIDWOOD UNITS 1 & 2 3.3;12a 10/29/98 Revision Q

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RTS Instrumentation B 3.3.1 BASES O SURVEILLANCE SR 3.3.1.9 (continued) REQUIREMENTS The SR is modified by a Note that excludes verification of setpoints from the TADOT. Since this SR applies to RCP undervoltage and underfrequency relays, setpoint verification requires elaborate bench calibration and is accomplished during the CHANNEL CALIBRATION. I.!) SR 3.3.1.10 A CHANNEL CALIBRATION is performed every (184. months, or i 1 approximately at every refueling. CHANNEL CALIBRATION is a l complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. -plad CHANNEL CALIBRATIONS must be performed consistent with the assumptions of theES specific setpoint methodology. The difference between the current "as found" values and the rgy 3,w previous test "as left" values must be consistent with the p -- e m - -

= the setpoint methodology.

C'dI8N"Y Ot-The Frequency of 18 months is based on the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint methodology. SR 3.3.1.10 is modified ~b~y a Note stating that this test "O ~ shall include verification that the time constants are ') adjusted to the prescribed values where applicable. ,E SR 3.3.1.11 SR 3.3.1.11 is the performance of a C NEL CALIBRATION, as describedinSR3.3.1.10,every918 nths. This SR 1's modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the power range neutron detectors consists of a normalization of the detectors based on a p I;J calorimetric and flux map performed above 15% RT. e 3 CHANNEL CALIBRATION for the source range #end. intermediate g rangejieutrondetectorsconsistsofobtainingthedetector f IMERX B 3 3~S4A) P (continued) WOG STS B 3.3-56 Rev 1, 04/07/95 2_c: \\) Q

ESFAS Instrumentation B 3.3.2 BASES O SURVEILLANCE SR 3.3.2. P-4 RWN Mf IAMor k SR 3.3.2.S'@iT the performance of a TADOT. REQUIREMENTS check of the Manual Actuation Functions and fT" :- i{ This test is a (continued) +-4; c' eii uru 7.-- It is performed every 4 aa gl8[ months. Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device. (i.e., pump starts, valve cycles, etc.). The Frequency is adequate, based on industry operating experience and is consistent with the typical refueling cycle. The SR is modified by a Note that excludes verification of setpoints during the TAD 0Tgfe-r 21 initi.ti a T. xti n :. The r.nl '-iti:ti:n Functions have no associated setpoints. SR

3. 3. 2.'1L SR 3.3.2.'9 is the performance of a CHANNEL CALIBRATION.

A CHANNEL CALIBRATION is performed every @8,Leonths, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. d The test verifies that the channel responds to measured \\ parameter within the necessary range and accuracy. ,Jriontl } CHANNEL CALIBRATIONS aust De performed consistent with the g assumptions of the u@ specific setpoint methodology. The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology. The Frequency of 98kmonths is based on the assumption of an(184 sonth cal bration interval in the determination of the magnitude of equipment drift in the setpoint methodology. This SR is modified by a Note stating that this tesii should 7 ~ ~ include verification that the time constants are adjusted to 8 ~~ the prescribed values where applicable. SR 3.3.2.1/// dad 54 5.5,2./ 2. } T Shnsu es the individual channel ESF RESPONSE TIMES are less than or equal to the maximum values assumed in the g 3 (continued) E O WOG STS B 3.3-118 Rev 1, 04/07/95 ReV Q .e et.---- v m-. p e

^4

• • a4.'.4 A

.,i.AmAme# 4._,Wi._, ,A, 4,.,M.g, _p ,4,g, ,g j_ _g i fa-( I - 4 i BASES JFDS jg i l 8

JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 BASES !(] SECTION 3.3 - INSTRUMENTATION 1G P Condition B of ITS LCO 3.3.2 and associated Bases were revised to u replace the' reference to " trains" with " channels" and to replace the references to " channels" to " switches." These changes are consistent with the SI. Containment Spray Phase A Isolation, and Phase B Isolation manual initiation design descriptions. P Consistent with the addition of ITS SR 3.3.2.11. ITS SR 3.3.2.11 Bases g are revised to reflect ITS SR 3.3.2.11 and ITS SR 3.3.2.12. All Res)onse Time testing is performed on an 18 month staggered test basis. l wit 1 the exception of the motor-driven pump auxiliary feedwater start c.) bus 141(241) undervoltage. y P Not used. g P ITS SR 3.3.1.11 Bases is revised to add "..and obtaining the detector g plateau curves, evaluating those curves. and comparing the curves to the manufacturer's data." This change provides a more accurate description of how this SR is accomplished. P NUREG-1431 SR 3.3.1.16 Bases states. "The analyses model the overall or u total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor to the point at which the equipment i reaches the required functional state (i.e., control and shutdown rods i < -m fully inserted in the reactor core)." The statement. "(i.e., control '( ) and shutdown rods fully inserted in the reactor core)" is being deleted. R./ The ITS Definition for Reactor Trip System (RTS) Response Time specifically states. "The RTS Response Time shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the LU channe1 sensor until loss of stationary gripper coil voltage." This u conflicts with the example provided in the SR, which identifies the 3 final measurement state as " rods fully inserted." When the stationary o gripper coil voltage is lost the control rods are not yet fully i h-inserted. therefore, this change eliminates potential confusion. and establishes consistency with the Definition. Pu ITS SR 3.3.3.2 Note for PAM Instrumentation exempts neutron detectors from a Channel Calibration. This Note is revised to replace " neutron detectors" with " radiation detectors." None of the PAM Instrumentation listed either in CTS Table 3.3-10 or in ITS Table 3.3.3-1 contains any neutron detectors. Revision of this note is consistent with the plant design and consistent with the general intent of NUREG-1431. l l U BYRON /BRAIDWOOD - UNIT 5 1 & 2 3.3 13 10/29/98 Revision Q t l i

a -aa %a _a__J-._. sA_ m uA -4 _.a A.4 _..e4r.Ay O ITS SECTION 3,6 4 4 [

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. ~. _ _. _... Containment Air Locks B 3.6.2 l- . BASES ACTIONS (continued) L D.1 and D.2 If the inoperable containment air lock cannot be restored to OPERABLE status within the required Completion Time the unit must be brought to a MODE in which the LCC does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on oper'ating experience, to reach the recuired unit conditions from full power conditions in an orcerly manner and without challenging plant systems. L SURVEILLANCE' SR 3.6.2.1 REQUIREMENTS E Maintaining containment air locks OPERABLE requires l compliance with the leakage rate test requirements of the l Containment Leakage Rate Testing Program. This SR reflects the leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The acceptance criteria were established during initial air lock and containment OPERABILITY testing. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is required by the Containment Leakage Rate Testing Program. t The SR' has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous-successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is 4 capable of providing a fission product barrier in the event 4 of a DBA. Note 2 has been added to this SR requiring thb results to be evaluated against the acceptance criteria

• % *7 which is applicable to SR 3.6.1.1.

This ensures that air lock leakage is properly accounted for in determining the combined Type B and C containment leakage rate. l BYRON:- UNITS 1 &-2 B 3.6.2 - 9 10/26/98 Revision 0 K

1 Containment Spray and Cooling Systems B 3.6.6 BASES SURVEllLANCE REQUIREMENTS (continued) SR 3.6.6.2 Operating each containment cooling train fan unit (in slow speed) for 2 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. j It also ensures that blockage, fan or motor failure. or excessive vibration can be detected for corrective action. The 31 day Frequency was developed considering the known reliability of the fan units and controls, the two train redundancy available. and the low probability of significant degradation of the containment cooling train occurring between surve111ances. It has also been shown to be acceptable through operating experience. SR 3.6.6.3 Verifying that each containment cooling train SX cooling flow rate to each cooling unit is a 2660 gpm provides assurance that the design flow rate assumed in the safety analyses will be achieved. The Frequency was developed considering the known reliability of the SX System, the two (' train redundancy available, and the low probability of a significant degradation of flow occurring between \\ surveillances. SR 3.6.6.4 Verifying each containment spray pump's developed head at the flow test point is greater than or equal to the required l developed head (resulting in 265 psig discharge pressure) ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by SECTION XI of the ASME Code (Ref. 8). Since the containment spray pumps cannot be tested with flow through the spray headers they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY trend performance, and detect incipient failures by abnormal performance. The Frequency of the SR is in accordance with the Inservice Testing Program. A BYRON - UNITS 1 & 2 B 3.6.6 - 10 10/30/98 Revision Q

4 i BRWD ITS lO 4 a 4 4 4 I l t. O O

.y Containment Air Locks B 3.6.2 l BASES- . ACTIONS-(continued) 0.1 and 0.2 t If the inoperable containment air lock cannot be restored to OPERABLE status within the required Completion Time. the unit must be brought to a MODE in which the LCO does not l apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience. to reath the required unit j conditions from full power conditions in an orderly manner and without challenging plant systems. i SURVEILLANCE SR 3.6.2.1 j REQUIREMENTS Maintaining containment air locks OPERABLE requires l compliance with the leakage rate test requirements of the Containment Leakage Rate Testing Program. This SR reflects l the leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The acceptance i l criteria were established during initial air lock and s containment OPERABILITY testing.- The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is required by the Containment Leakage Rate Testing Program, e The SR has been modified by two Notes. Note 1 states that l an inoperable. air lock door does not invalidate the previous successful. performance of the overall air lock leakage test. i T This is considered reasonable since either air lock door is J. capable of providing a fission product barrier in the event J of a DBA. Note 2 has been added to this SR requiring th'e 40 results to be evaluated against the acceptance criteria $7 which is applicable to SR 3.6.1.1. This ensures that air w dE lock leakage is properly accounted for in determining the combined Type B and C containment leakage rate. l I f BRAIDWOOD - UNITS 1 & 2 B 3.6.2 - 9 10/26/98 Revision 0 l -.,y .ye. y- +. -.

Containment Spray and Cooling Systems B 3.6.6 em b) BASES SURVEILLANCE REQUIREMENTS (continued) l SR 3.6.6.2 Operating each containment cooling train fan unit (in slow speed) for a 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure. or excessive vibration can be detected for corrective action. The 31 day Frequency was developed considering the known reliability of the fan units and controls, the two train redundancy available, and the low probability of significant degradation of the containment cooling train occurring between surveillances. It has also been shown to be acceptable through operating experience. SR 3.6.6.3 Verifying that each containment cooling train 5 - A ng flow rate to each cooling unit is a 2660 gpm ' a.i @ assurance that the design flow rate assumed ir..... oafety analyses will be achieved. The Frequency was developed considering the known reliability of the SX System, the two Di train redundancy available. and the low probability of a O significant degradation of flow occurring between surveillances. SR 3.6.6.4 Verifying each containment spray pump's developed head at the flow test point is greater than or equal to the required l developed head (resulting in 265 psig discharge pressure) ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by SECTION XI of the ASME Code (Ref. 8). Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY trend performance, and detect incipient failures by abnormal performance. The Frequency of the SR is in accordance with the Inservice Testing Program. (w BRAIDWOOD - UNITS 1 & 2 B 3.6.6 - 10 10/30/98 Revision 0

ll BASES MARKUPS e O f e O

1 i g Containment Air Locks gat:c;nnuic. =t;;;,,as;nent, ::: cancen::- 2nc n m B 3.6.2 BASES (continued) SURVEILLANCE 5_R 3.6.2.1 REQUIREMENTS Maintaining containment air locks OPERABLE requires compliance with the leakage rate test requirements of b-) i 10 KR 50, Appendix /J (Ref.1), as/ modified by aggirovedl l exemptions.1 This SR reflects the leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The acceptance criteria were established during i t meu m-m) initial air lock and containment OPERABILITY testing. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is recuired by Appeydix J (Ref.- 1),/ as modified by/ approved exem Thuf, SR 3.0.2 (whith allows Frequency extensions)ptions. does not apply.l The SR has been modified by two Notes. Note I states that an inoperable air lock door does not invalidate the previous 1 successful performance of the overall air lock leakage test. o' This is considered reasonable since either air lock door i's 5 capable of providing a fission product barrier in the evert h of a DBA. Note 2 has been added to this SR requiring the i e results to be evaluated against the acceptance criteria @ F, Unse eh-ne] v5R 3.6.1.1. This ensures that air lock leakage is properly gI accounted for in determining the 6 mcontainment leakage 4, rate. (c,,b,n,d me eaudc) SR 3.6.2.2 j The air lock interlock is designed to prevent simultane:ous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containmept % pressure, closure of either door will support containmant y OPERABILITY. Thus, the door interlock feature supporta containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the intericck will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur. Due to the purely mechanical nature of this interlock, and given c no+ norman y that the interlock mecnanism is eH+) challenged when the containment air lock door is I;;;r. d, tt,1: t::t i: Only ,quwed t; b; perfered upen enterini; er ;;iti;s ; o n m e ar.- a.c ) cent:i mert zi-leck b"t ir met re;" ired me e #re;" eat h O) (continued) V WOG STS B.3.6-27 Rev 1, 04/07/95 Rea Q

BASES INSERT (S) l( ) SECTION 3.6 Bases 3.6.2 INSERT B 3.6 27A (C ) 1 ... Containment Leakage Rate Testing Program. fi J-Ap7INSERTB3.6278 (C ) 3 u 3 5d..whichisapplicableto INSERT B 3.6 27C (C ) 2 ... used for entry'and exit (procedures require strict adherence to single . s' door opening) this test is only required to be perforned every 24 months. i r-s 4 .The 24 month Frequency is based on the need to perform this Surveillance under (j 3 the conditions that apply during a plant outage, and the potential for loss of M containment OPERABILITY if the Surveillance were performed with the reactor at S power. The' 24 month Frequency for the interlock is justified based on generic 'Q operating experience. t G 10/26/98 Revision Q

l ContainmInt Spray and Cooling SystemsG merened rd DuM D n 7 8 3.6.69 /~ SASES >(L l n SURVEILLANCE SR 3.6 @ 2 D Ili""'"d REQUIREMENTS (continued) Operating each ' ' containment cooling train fan unit for 215 minutes ensures that all trains are OPERABLE and l that all associated controls are functioning properly. It l also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. The 31 day Frequency was developed considering the known reliability of the fan units and_ controls, the two train l redundancy available, and the low probability of significant degradation of the containment cooling train occurring between surveillances. It has also been shown to be 1 acceptable through operating experience. g SR 3.6.603 h (zt, o ) .SX Verifying that each Grecu;r;:c occntainment coolind train G cooling flow rate to each cooling unit is 2 [4WTgpm provides assurance that the design flow rate assumed in the safety analyses will be achieved 6. The Frequency ) was developed considering the known reliability of the ,e h u eui mu W i.e6 System, the two train redundancy available, and the low probability of a significant degradation of flow s occurring between surveillances. h ~. (recuNv3 m zt.s pq ch:. charge pusure) h SR 3.6.5%)4 7 Verifying each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head' ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by Section XI of the ASME Code (Ref. 8). Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by abnormal performance. The Frequency of the SR is in accordance with the Inservice Testing Program. l. l

e t

(continued) WOG STS B 3.6-72 Rev 1, 04/07/95 %Q

.,--c. -m,.., .-w-um .m 4 .. m a J s a A + .4.+44 m. w. _.m-a d i O ITS SECTION 3,7 i 4 4 1 1 T l 6 O

Y 2 1-o 4 ' 4 l BYRON CTS MARKUPS 4. i / e i, : 1 i 9 k d .d i i k d d+ 1 4 e , ~ -

, 3,m ang " ~ ~ ~ ~ ~d~$ . ~ _ - - 3 ;,12. "A7.7 NON-Af'f'#RSIRI N AREA EXHAUST Fil mM PLFNUM VENTILAT10N S LIMmNG CONDmON FOR OPERATION !s g Qienvete kwe t*,ms) g

7. i 2, -31:7 Thnsef. 7.

N =I HH' ""M d=_ ngiidin --~; - .ilshall be OPERABLEnvim two 7 - F and oni ' - m standFA APPLICABillTY uMODEWa, una 4._ st e ~N ej l .4CI!,Qti: Qecurmn Wnm 7 nam) Lo With one non-accessible area exhaust filter steriu'n'iiE-ante, restore tre in:-:-erable olenum n: c. _ OPERABLE s ruJs withir 7 dawn ariaa m at east ru( CANDSY veisei me next 5 hours and i be g L COLD E HUTDOWh withn the fol!r as ao asass.,4 7- _9,i c., c ;: =e errc ;ii a W l m mi. 6 '- e-d t: ?:: 0~;.." m ::c':nr2e :ni.v SURVEll I ANCE REOulREupNTS W Each non-accessible area exhaust filter plenumilhall be demonstrated OPER ~ Qe nru amn Suwm Twe ; At least once car 31 dayst::n : =r" a.

n 3 a. a.1 r3,;,,,;,w,g;; 7,;g, ;;,, ;;,,,;.=s; :== T Wr;"r, ;r.re ;,, L,, r L

g 6 se.e., ,3.,-m G M a J)n= 1= cc-7-m er at '==* 15 nunla= ;,, - ; ; 7 -_;.; ;= :.siiy

O.%IZ L b.

At lessi onos per 18 months, or (1) after any atmetural maintenance on t filter or charcoal adsorber housings, or (2) followm' g paintmg, fire, or in any ventilation zona communicating with the exhaust filter plenum by 1) b-;ij;r that the exhaust filter plenum satisfies the in-pisce penetrabon a testing acca stanos enterta of less than 1% when using the test stocedure O-in Ragu story Podhons C.5.a. C.S.:: and C.S.d of and 68,200 cim forthetrain;e 1.52, Riivision 2. March 197 55 60 2) , within 31 days afterremoval, that a analyss of a carbon from each bank of o inedin accordanos Regu rs of the train Regu. Revision 2, March 1978, meets Position C.6.b of Regulatory Guide 1.52 latory Position C 6.a of Re gulatory Guide 1. testing emeria of of 30*C and a relative humidity of 70%;for methyl iodido ~e AgeM%D m A % 0h fiO so m o. kn en c. o ' N f G BYRON-UNITS 1 & 2 3/4 7-19 AMENDMENT NO.105;105 Aev Q

BRWD CTS MARKUPS O O O

___.m_.._ %t-19-68 07:47aa Fron-CCWED T-100 W20 F-219 m % 37. 7.1 %m n r 7, q .7 PLANT SYSTEMS 3 7 J 7_ SM-t7 NON-ACCESSim F AREA EMfAUST FILTER PLENUM VENTILATION SYs u mA 1.lMITING CONDITION FOR OPERATION b=0a SvmD .d./ocessele/reaexhadJ er.ums ' /. : ' - 2-c-r -- I ---m;

rd :

snall be OPERABLE 1lWm trea r u.='igned fbr operiitson una o % CM'GD q",,', APPLICABILITY: MODE 81,2,3, and 4. ACTION: 4 maw gb vtm.n e With one non ar===Ma ares -4 ust_ filter,;,unfinoperable, restore the InopeM plenum I mo /- 1 to OPERABLE status witten 7 days wlue in at least HOT STANDBY within the next 8 houra and in COLD SHUTDOWN within the fcilomno 30 hours.^r. fn. :":: ;:::n _r::.- 1.c.. : a l 6 - :1_2: :: T : ;;.; dT:c 5-^* " d2. h ^-_ C;. SURVE1LLANCE REculppuMMTS tNewmn Sou erT) 4-7-7 Each non-accassible area exhaust filter plenum 4 hall be demonstrated OPEPAgl C-6-M sR. 3 7.12 1 a. At toestonosperm " " ~^ -- ln..".r.":: _r.0 " ".7

. c. h r

~ that sp.isui, ~ - ~ atleast 15 minutes; u% 53u2 2. qnseer s.w s i% n. Atloestenos paraa,,,,4,;i, or (1) anerany seustural malmenance on me - HEPA iller or charcoal adsorter housings, or (2) lbiomng painting. fire, or chemical release in any ventihmon zone communicoting with the exhaust filter plenum by-1) Venfying that the exhaust filter plenum satisfies the in-place penetranon and bypass leakage testing acceptance cresna of less then 1% when using the test procedure guidance in Reguistory Positions C.S.a. C.5.c and C.5.d of Regulatory Guide 1.52. Revision 2, March 1978, and the flow rate is 68,900 cfm t10% for the train and 22,300 cfm 110% per bank: 2) Venfying, within 31 days amor removal, that a laboratory analysas of a representutve cert:on sample from each bank of adsorbers of the train obtained in encordance with Regulatory Position C.S.b of Reguistory Gulds 1.52, Revision 2. March 1978, meets the laboratory testing criterte of ReGuistory Position C.6.a of Regulatory Guide 1.52. Rewsion 2, Man

• 1978, for methyl lodido penetration of less then 1% when tested at the temperature of 30'C and a reiettve humidity of 70%;

Aoorumn en Suss .o hbb %kmnn 'jkGi6Qsh-E: M D 9 * (A$) G BRAIDWOOD - UNITS 1 & 2 3/4 7-17 AMENDMENT NO. 97: 97 y hJ 9

CTS DOCS l 9

i DISCUSSION OF CHANGES TO CTS ITS SECTION 3.7 PLANT SYSTEMS A CTS LC0 3.7.1.1 Action a requires restoration of an inoperable MSSV.. u NUREG-1431 does not contain the restoration action, but rather relles upon the guidance of LCO 3.0.2 which allows the restoration of a parameter within the time limits of the specified Required Actions. The CTS has been revised to delete this restoration action since the option to restore an inoperable component currently exists in the ITS. During this reformatting, no technical changes (either actual or interpretational) were made to the TS. unless identified and justified. The change is consistent with NUREG-1431 as modifiea by WOG-83. A CTS SR 4.7.3.1 requires verification that each valve (manual. y power-operated, or automatic) servicing safety-related is in its correct position. The Byron and Braidwood CC System does not include any automatic valves. Therefore, the CTS Reference to " automatic" does not have any technical meaning and ITS SR 3.7.7.1 does not include the words automatic. Because the ITS SR accurately reflects the design and intent of the CTS requirement. no technical changes were made to the TS. unless identified and justified. Au By letter dated October 15, 1998, a revision to the Byron and Braidwood CTS for the Nonaccessible Area Exhaust Filter Plenum Ventilation System was received (Amendment # 105 for Byron / Amendment # 97 for Braidwood). Any revisions to the request as a result of the conversion to the ITS are annotated and justified separately. A. A note is added to CTS LCO 3.7.7 (ITS LCO 3.7.12) that allows the Nonaccessible Area Exhaust Filter Plenum Ventilation System alignment requirement to be suspended intermittently under administrative controls for purposes of train realignment. The Nonaccessible Area Exhaust Filter Plenum Ventilation System is required to have two trains aligned for operation and one train aligned in standby because the system design cannot support operation of three trains simultaneously. This note is added to clarify that during realignment of the system there may be a short time period when three inlet dampers may be open or two inlet dampers may be closed. This is an administrative change with no impact on safety because the clarification provided by the Note is consistent with the existing interpretation of the CTS. l lV BYRON /BRAIDWOOD UNITS 1 & 2 3,7 11 10/29/98 Revision Q \\

a l i t' ITS SECTION 3,8 4 J J l l. 1 J 1 3 i

i I

i 6 1 i 1 \\ i I { i I i

BYRON ITS Q O O

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.8.1.10 This Surveillance demonstrates the DG capability to reject a full load without overspeed tripping or exceeding the i predetermined voltage limits. The DG full load rejection L may occur because of a system fault or inadvertent breaker tripping This Surveillance ensures proper engine / generator response under the simulated test conditions. This test l simulates a full load rejection and verifies that the DG does not trip upon loss of the load. These acceptance criteria provide for DG damage protection. While the DG is not expected to experience this transient during an event l and continues to be available, this response ensures that l the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be i corrected or isolated. l 7 The 18 month Frequency is consistent with the recommendation 7 of Regulatory Guide 1,9 (Ref. 3) and is intended to be 7 consistent with expected fuel cycle lengths. l g $l } This SR has been modified by two Notes. Note 1 states that L t momentary transients above the stated voltage limit L' J 1mmediately following a. load rejection (i.e., the DG full l load rejection) do not invalidate this test. The momentary transient is that which occurs immediately after the circuit Eg ' breaker is opened, lasts a few milliseconds, and may or may not be observed on voltage recording or monitoring instrumentation. The reason for Note 2 is that during 7 operation with the reactor critical, performance of this .6 - SR could cause perturbation to the electrical distribution 6 systems that could challenge continued steady state u j operation and, as a result plant safety systems, i l l \\ l i I (! J. BYRON - UNITS 1 & 2~ B 3.8.1 - 21 10/26/98 Revision 0 l _~

BRWD ITS 4 O t b O

AC Sources-Operating B 3.8.1 ( BASES l SURVEILLANCE REQUIREMENTS (continued) i SR 3.8.1.10 i This Surveillance demonstrates the DG capability to reject a full load without overspeed tripping or exceeding the predetermined voltage limits. The DG full load rejection l may occur.because of a system fault or inadvertent breaker tripping. This Surveillance ensures prnper engine / generator response under the simulated test conditions. This test simulates a full load rejection and verifies that the DG l does not trip upon loss of the load. These acceptance L criteria provide for DG damage protection. While the DG is not expected to experience this transient during an event and continues to be available this response ensures that the DG is not degraded for future application. including reconnection to the bus if the trip initiator can be corrected or isolated. ~ L M The 18 month Frequency is consistent with the recommendation of Regulatory Guide 1.9 (Ref. 3) and is intended to be t) j consistent with expected fuel cycle lengths. l1 1 Q This SR has been modified by two Notes. Note 1 states that e momentary transients above the stated voltage limit V immediately following a load rejection (i.e.. the DC full load rejection) do not invalidate this test. The momentary Sg. transient is that which occurs immediately after the circuit i t breaker is opened. lasts a few milliseconds, and may or may _i. not'be observed on voltage recording or monitoring 'y instrumentation. The reason for Note 2 is that during 5 operation with the reactor critical, performance of this -9 SR could cause perturbation to the electrical distribution 3 systems that could challenge continued steady state operation and, as a result, plant safety systems. l LQ V 'BRAIDWOOD - UNITS 1 & 2 B 3.8.1 - 21 10/26/98 Revision 0

4 ema-.am-.,we'4RAmV wA, M ,4XI-e = i-44 M-d 4 4-6 4 & 6.>.b.a a4 A-5.T-w - G -e A-bm b 4Lkb.63 A AE6 4lILLkm4Ast4 AmmaM4 96AA,mn--H KM An n >R MA4e & A# beard-e B44 44 e&AM 'Ame de *6.4 w ase nan n & m-a 6 o m m A M< l 1 i 4 1. BASES MARKUPS J 4 I i 1.; 1 4 1 5 k k, i 1 i 4 4 2 b i l a,. I o 1 1-1 ii a i i 4-i .s f 2 i 4 t j 4 e a 1 l l 4

BASES INSERT (S) ) SECTION 3.8 (J- \\_ Bases 3.8.1 ~ DI INSERT B 3.8 23A (P ) g m W Note i states that momentary transients above the stated voltaae limit immediately following a load rejection (i.e., the DG full load' rejection) do "2 27 not invalidate this test. The momentary transient is that which occurs immeclately after the circuit breaker is opened, lasts a few milliseconds and may or may not be observed on voltage recording or monitoring instrumentation. s .,m %s. l l ,f 's - 10/26/98 Revision Q l

e_a.x ~m,.e e m M &-,-aJ -O mas 4

• ..m'.._m-a w is..&,+e,oa4.,,_..am.mmap_me.,--n<-.wm m.aa.mus.eMa me

-4< maa m 4 Ln-awe .m,A.mAmsew-i e l( ITS SECTION 3,9 t n 4 2 i 4 4 i i i 4 { 1 t I 1 l l )

--w,:,-w-w mm- - -= emmm,-ra, -w~ - - - O BYRON ITS O O 9 O

Boron Concentration 3.9.1 l 3.9 REFUELING OPERATIONS -3.9.1 Boron Concentration i LCO 3 9.1 Boron concentrations of the Reactor Coolant System, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the COLR. I nPPLICABILITY: MODE 6. lTACTIONS: CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration A.1 Suspend CORE Immediately not within limit. ALTERATIONS. AND A.2 Suspend positive Immediately ,p reactivity additions. MD A.3 Initiate action to Immediately restore boron concentration to within limit. l ' SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR-3.9.1'.1 Verify boron concentration is withii1 the 72 hours limit.specified in the COLR. L 1

O BYRON'- UNITS 1 & 2 3.9.1 - 1 10/26/98 Revision 0 l-

= RHR and Coolant Circulation-Low Water Level 3.9.6 7 3.9 REFUELING OPERATIONS (V 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level LCO 3.9.6 Two RHR locps shall be OPERABLE. and one RHR ioop snall be in operation. -NOTE One required RHR loop may be removed from operation and considered OPERABLE: a. To support filling and draining the reactor cavity when al.igned to or during transitioning to or from. the refueling water storage tank provided the required RHR loop is capable of being realigned to the Reactor Coolant System (RCS): or b. To support required testing provided the required RHR loop is capable of being realigned to the RCS. APPLICABILITY: MODE 6 with the water level < 23 ft above the top of reactor vessel flange. U l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RHR loops A.1 Initiate action to Immediately inoperable. restore RHR loop (s) to OPERABLE status. O_R. A.2 Initiate action to Immediately establish a 23 ft of water above the top of reactor vessel flange. L (continued) l m .] BYRON - UNITS 1 & 2 3.9.6 - 1 10/26/98 Revision 0 1 l p l

_.... _ _ ~. _ _ _ _ _. _. 3oron Concentration 1 E 3.9.1 BASES LCO The LCO requires that a minimum boron concentration De maintained in all filled portions of the RCS. the refueling canal. and the refueling cavity, that are hydraulically coupled to the reactor core, while in MODE 6. The boron concentration limit specified in the COLR ensures that a core k,,, of s 0.95 is maintained during fuel handling operations. Violation of the LC0 could lead to an inadvertent criticality during MODE 6. APPLICABILITY This LC0 is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical. The required boron concentration ensures a k.,, s 0.95. In MODES 1 and 2 a 1.0. LCO 3.1.4. " Rod' Group Alignment Limits." with k,',5. " Shutdown Bank Insertion. Limits." and LCO 3.1.6. LCO 3.l. " Control Bank Insertion Limits." ensure an adequate amount of negative reactivity is available to shutdown the reactor. In MODE 2 with k, < 1.0 and MODES 3. 4. and 5. LCO 3.1.1. " SHUTDOWN MARGIN,,(SDM)." ensures that an adequate amount of negative reactivity is available to shut down the reactor i l end maintain it subcritical. O, @GjACTIONS A.1. A.2. and A.3 l Continuation of CORE ALTERATIONS or positive reactivity d cdditions (including actions to reduce boron concentration) I: _f is contingent upon maintaining the unit in compliance with d the LCO. A Heu L l l 1. i. 4 ( BYRON - UNITS 1 & 2' B 3.9.1 - 3 10/26/98 Revision 0 l m.

RHR and Coolant Circulation-Low Water Level B 3.9.6 BASES

ACTIONS A.1 and A 2 1

With one or more RHR loops inoperable. the RHR System may not be capable of removing decay heat and mixing the borated coolant. Therefore. action shall be immediately initiated and continued until the required number of RHR locps are restored to OPERABLE status or until = 23 ft of water level is established above the reactor vessel flange. When the l water level is a 23 ft above the reactor vessel flange. the Applicability changes to that of LCO 3.9.5. and only one RHR loop is required to be OPERABLE and in operation. An immediate Completion Time is necessary for an operator to initiate corrective actions. B.1. B.2. and B.3 ^ If no RHR loop is in operation there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations can occur by the addition of water with a lower boron concentration than that contained in the RCS. Therefore, actions that would result in a reduction in the coolant boron concentration must be suspended immediately. r In addition, with no forced circulation, any decay heat removal occurs by ambient losses only. Therefore. action shall be initiated immediately to restore one RHR loop to operation. Once initiated, actions shall continue until one RHR loop has been restored to operation. 4 O BYRON - UNITS 1 & 2 B 3.9.6 - 3 10/26/98 Revision 0

BRWD ITS Q O 1 l O i

..-..-.-.--.-.._..- - -..- -.~.- - -.--.==- L l. Boron Concentration 3.9.1 I

3.9 REFUELING OPERATIONS 3.9 1 Baron Concentration.

LCO 3.9.1 Boron concentrations of the Reactor Coolant System. the refueling canal. and the refueling-cavity shall be maintained within the limit specified in the COLR. l p APPLICABILITY: MODE 6. l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME LA. Baron' concentration A.1 Suspend CORE Immediately not within-limit. ALTERATIONS. AND I' A.2 Suspend positive. Immediately reactivity additions. A.3 Initiate action to Immediately restore boron, concentration to l within limit. e i. SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY L

SR 3.9.1.-1 Verify boron concentration is within the 72 hours limit specified in the COLR.

j BRAIDWOOD'- UNITS 1 & 2-3.9.1 - 1 10/26/98 Revision 0 -4.

. - - -.. - _ _ =... - RHR and Coolant Circulation-Low Water Level 3.9.6 O 3.9 -REFUELING OPERATIONS V 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level LCO 3.9.6 Two RHR loops shall be OPERABLE, and one RHR loop snall be in operation.

---

NOTE =

==--- One reauired RHR loop may be removea from operation anc considered OPERABLE: a. To support filling and draining the reactor cavity when aligned to, or during transitioning to or from. the refueling water storage tank provided the requirea RHR loop 1s capable of being realigned to the Reactor Coolant System (RCS): or b. To support required testing provided the required RHR _ loop 1s capable of being realigned to the RCS. APPLICABILITY: MODE 6 with the water level < 23 ft above the top of reactor ,p vessel flange. ,V lA30NS l C0fDITION REQUIRED ACTION COMPLETION TIME A. One or nare RHR loops A.1 Initiate action to Immediately inoperable, restore RHR loop (s) i to OPERABLE status. DB A.2 Initiate action to Immediately establish a 23 ft of water above the top i of reactor vessel flange. (contirued) BRAIDWOOD - UNITS 1 & 2 3.9.6 - 1 10/26/98 Revision 0 i'

Boron Concentration B 3.9.1 BASES LCO The LCO requires that a minimum boron concentration be maintained in all filled portions of the RCS, the refueling ccnal. and the refueling cavity. that are hydraulically coupled to the reactor core, while in MODE 6. The boron concentration limit specified in the COLR ensures that a core k,,,' of s 0.95 is maintained during fuel handling operations. Violation of the LCO could lead to an inadvertent criticality during MODE 6. APPLICABILITY This LC0 is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subtritical. The required boron concentration ensures a k, s 0.95. In MODES 1 and 2 with k,, a 1.0. LCO 3.1.4. " Rod,, Group Alignment Limits." LCO 3.1.5. " Shutdown Bank Insertion Limits." and LCO 3.1.6. " Control Bank Insertion Limits." ensure an adequate amount of negative reactivity is available to shutdown the reactor. In MODE 2 with k,,, < 1.0 and MODES 3. 4. and 5. LC0 3.1.1. " SHUTDOWN MARGIN (SDM)." ensures that an adequate amount of negative reactivity is available to shut down the reactor and maintain it subcritical. O> 4Gl ACTIONS A.1. A.2. and A.3 Continuation of CORE ALTERATIONS or positive reactivity j additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with 6 the LCO. m ~~ O BRAIDWOOD - UNITS 1 & 2 B 3.9.1 - 3 10/26/98 Revision 0

RHR and Coolant Circulation-Low Water Level B 3.9.6 (^)T BASES c l ACTIONS A.1 and A.2 With one or more RHR loops inoperable. the RHR System may not be capable of removing decay heat and mixing the borated coolant. Therefore, action shall be immediately initiated and continued until the required number of RHR loops are restored to OPERABLE status or until a 23 ft of water level is established above the reactor vessel flance. When the water level is a 23 ft above the reactor ves5el flange. the Applicability changes to that of LC0 3.9.5 and only one RHR loop is required to be OPERABLE and in operation. An immediate Completion Time is necessary for an operator to initiate corrective actions. B.1. B.2. and B.3 If no RHR loop is in operation, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations can occur by the addition of water with a lower boron concentration than that contained in the RCS. Therefore, actions that would result in a reduction in the coolant boron concentration p. must be suspended immediately. In addition, with no forced circulation, any decay heat removal occurs by ambient losses only. Therefore, action shall be initiated immediately to restore one RHR loop to operation. Once initiated, actions shall continue until one RHR loop has been restored to operation. /~T i ,1 BRAIDWOOD - UNITS 1 & 2 B 3.9.6 - 3 10/26/98 Revision 0

x_,mra f,&Y.L4,4AO4&ew,4,,A,n as,,AeM-m te,he, -- iQ M4K,d sn. Lim-A 4ko a - 9,o 9 ,+44 AA A M sa. e nge s,p. 3 2ms,namm4 m +,4skmsm,k.e,.m-4m, g,,a,LM.n,an 4mmam am-mm*AAme mm'a--c.ina-s'n m asasamm--mms-maapesi e h ?' e BYRON CTS MARKUPS 1 I Y 1 1 I' 4 a i N I C 1 l

3 q GM:9 REFUELING OPERATIONS LCo 3 11 .3 9. l C M. O. ! BORON CONCENTRATION 2039.2 '*? " ' Y ^ " ' * " O LIMITING CONDITION FOR OPERATION >\\ '/ (._/ LCO 3.9.1 The boron concentration

- 7 ' = - ~

'"--M o the Reactor Coolant System and the refueling canab shall be maintained /ani.. .. i c i ent

T e that tha mer; r;;trictie: cf the fel: wir.; re::ti.'t: :;ndition; i;

) i kinc % l.b.4 me h lm & COLP,) {e. f ,,, : f 0. 95 e r l e s s ~ } hi. 2)" A b q d.ncentration of grc;ter th;n or cau:1 1: 2000 p;m._) 'b h (2 At;rd;;;,;;ntrationofgrc;terthen;r:;;;! to 2300 ppm. ] ,c, APPLICABILITY: MODE g n.: ACTION: oce e reh borm cwearm )

• '""I'"-

1 A With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate /end ;;,tinu; t;r:ti;r :t grc ter th:n er ::;c 1 t: 30 gpd {of : ::! 14on ::ntnnino-creat-er-thrn or eoual to 7000 com boron or iti ,::uivaleagruc r.

:::::: i

th:n er :::. to0. m---

2r= 5 icor,CcntrollCn ;, rbt0rt: !: gre !:r !n:C cr 00 l :: 000 ppi GZJU 'JDfh, hichevar i:=th; more restrictive./ As Ai gg O SURVEILLANCE RE0VIREMENTS O de erm Dr t h. Remov4, 39-er--enb:1 ting the-Fe&C4GT-ve56ebhead, and, (b. Wit rawa 'of any full- ~ ngth ontro rod ~n ex ss o 57 lvteps [ ( proxi tely feet from s fu y in rte posi on yfthi the peactor vessel / n gyq,,, c d we. rebet.as cavi+C A9 _ _ _ - The baron concentration of the Reactor Coolant System and the refueling canales all be Sur "rd 5- " r:1 :=i nid at least once per Q i h,m,.,mll,;,;,hecacJio&cotg 72 hours. SR.3 A 2..I i -s Valves WB4Wishall be verified closed and secured in position "" - - " ' " ^~ ^ &c r==

c f n i - r?

(electri::1 ::: r/at least once per 1 days. 3 (.DM RT 3.9-1 A Lco.3.9.2 Acnoos morf ( 1 CONS A j 2 be afintai fed infl0DE 6jkheney6r ftte'l is jn thg reaptor f

• The eactor shal f

3 ve sel w' h th vesse hea#'clostt/e bolt 4 lesVthan4ullyXensi/5ned Ar wMh j itp he remn ed. 1 q# i- / #Not applicable to Unit 1. Applicable to Unit 2 Lntil the completion of cycle 5. A3

• • Applicable to Unit 1.

Not applicable to Unit 2 until after cycle 5. BYRON - UNITS 1 & 2 3/4 9-1 AMENDMENT NO. 65 l hO

Leo 3.% 3 9 REFUELING OPERATIONS ~'

• • d O, 33 g/ LOW WATER LEVEL g

LIMITING CONDITION FOR OPERATION LCO.3 94 Two residual heat removal (RHR) loops shall be OPERABLE, ano at least 2 one RHR loop shall be in operation. L.C.n H w s C ruar 3.a-n o s APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet. IT i ACTION: I' Cosa A / With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or, establish greater than or equal to 23 feet of water above the reactorT vessel flange > -- ---- - - m-- limedi.binn%ak mb-k CM, m m e d/.+,4 v M ) With no RHR loop in operat1Tn,/susp(end all operations involving a Gad 6 y. reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. SURVEILLANCE REQUIREMENTS M SR 3.9 4. I 43F:N:39 At least once per 12 hours one RHR loop shall be verified in operation H and circulatino coolant at a flowrate of areater than or equal to 1000 gpm@ MCS t::::r ::r: 1::: th:r er ::::1 to 1402f $W L A.. SR 3 9.t. 2- / r.us ser.3.9-iO A [ BYRON - UNITS 1 & 2 3/4 9-10 AMENDMENT NO. 38 k&Y &

i i L CTS INSERT (S) ( ) SECTION 3.9 LC0 3.9.6 INSERT 3.9 10A (M ) 3 SURVEILLANCE FREQUENCY 1 SR 3.9.6.2 Verify correct breaker alignment and 7 days indicated power available to the required RHR pump that is not in operation. ] l INSERT 3.9 10B'-(Ln)- l ( )' NOTE One required RHR loop may be removed from operation and considered OPERABLE ..t glL3 a. To support filling and draining the reactor cavity when aligned to, or during transitioning to or from, the refueling water storage tank provided the required RHR loop is capable of being realigned to the Reactor l Coolant System (RCS): or b. _ To support required testing provided the required RHR loop 1s capable of being realigned to the RCS. i l INSERT 3.9-10C i i Deleted in Revision 0 1 l k-- 10/26/98 Revision Q b I,

i i i BRWD CTS MARKUPS 4 O O O

L.c.C,3,. r. i i

1. '. a' ^-

l J.4 0 /.;. ;) REFUELING OPERARQ11 . 4. l (3 /4. 0. D BORON CONCENTRATION ,/-

1. 'b AN LIMTTING CONDITION FOR OPERATION

pv

{D / fi, W 3.9.1 The boron concentration W""--^'--do the Reactor Coolant System and the refueling canal"shall be maintainedf;nif:

:e?+1: tert t:T Q1 f::: r: th:t th: ;;; : ::tri:tte: Of th: f:ll: in; r:::tivity :::diti::: i:

/ 4 6% -k 16. q.c:4,eA 'm +L e M1, . _. ~. .ma ,p 2, /.. .e __ s.._.. J' t a . 4 - --[ t. 1) S tecer. ::::streti:n ;f ;;r::ter then er - ee1 te 2000 aa".) v:./ N ,....__.&._........,-_.,..,,nn... m MODE b APPLICABILITY: c sb. r v s ( A'. > acbn e recre. cora c. enc.ent.,r:or )

d ACTION:

L% 4. tR, w Wi +s,. 8 le C.At49 A \\ With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity v changes and initiatef::: :::::::: ::r:tter at ;reater ther er : r:1 t: 20 ;-- ' ef : ::lutter ::st : 4r; ;retter this er e-" 1 te ?000 --- her:: Or it: : ui"- n a n g alget fin + i' " t rdx--d t !- r *'-- ^- --"-l +a I concentra!!a"'!: re:10re !O Or0:!=* '"5" ^" ^^"I' T'S C300 - ; , whick="-- i! th-50 0 estricti"- I L / N h b Lh3 EVETLLANCEREOUTREMENTS ,,m ... -, m

___m,

.m..... C= m -s i. 4 4 . is wivi s is,6siw6svu vi seis .v,. ..u .. w... .g sviw.. vua hw IW[ .4. (E. .;^.17. er r.t$ltin; th' 700 t:r V::::1 h::d, ;.d) ~ 3 vi.ny i.ii-sun men wo u s,,,,,...,.....,,,,,, .. m.. n u i.... n (.approa4mately-4-feet)-from-its--fe'1" inter-ted ; :it40tHetthin th: (Ly C. 7:^ ter vesset.___...__. sg 3,4,I, I , --( ae we reGAM k.v e g

1. g

(,' 0.1.D The boroniconcentration of the Reactor Coolant 51 stem and the refueling canaleshall be(de^ers.tr.ed tv chnical r.:1v:i2 at least once per Cw.m +L4 c.a+ epe.g.aA w w cer.a.1 .Le 72 hours. 4 $4. M. A, I,,,J Val ves m, m.,,,. (,..... -,.,.J shall be veri fied. ,m __m ,o m.... , closed and secured in position &y --rherical IthT e-by - rs! cf 21-s- f 3,electri::1 ::c rjat least once per 31 days. Lco 3.5 SL LA pc.%s9 wrc. 1 N M.K Y 3.4-lh f ONP A It r:::ter : E '~5: estata+ nee-4r..es:. t = :::ver ?ee: 1: tr in re:04er g vessel with the-vesse' head-closure belt: le:S th:n fell" t:::i:::d r eith th: he-d -- red 1 ._.o.. ......2.. ._.m.. .....,...., _ _......,_.._,_ e.

• • n_yy iu.m,_

.u n v.. + ___.+4..,n u_ 4. + k.., u r. t.._ r.

1. _1. 4. +. 1 _.. 4 t_i. 4. +. 9

,+ hj ,,.1 4.., k i.. Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 3/4 9-1 Unit 2 - Amendment No. 55 2cJ @

L C o 3 't 4 e -?.4 REFUELING OPERATIONS ~ "" J.'t,G]LOWWATERLEVEL E LIMITING CONDITION FOR OPERATION Lc4 3.4.G G'3FIElb Two residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop 'shall be in operation.Ms.cr M-tos 3 p l wy y APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet'. ? 2 ACTION: ,g gg, CONpA 4. With less than the required RHR loops OPERABLE, immediately initiate correctiv action to return the required RHR loops to OPERABLE i status, orkestablish greater than or equal to 23 feet of water above i the reactor vessel flanoet :: :::: :: ::::ib ~. Mf7 With no RHR loop in ope (rat. ion,/suspen$~aeset f.ot@ 6' fr. d all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR ) loop to operation. Close all containment penetrations providing l -direct access from the containment atmosphere to the outside r' atmosphere within 4 hours. . O] w 1 - I-f

• SURVEILLANCE REQUIREMENTS td e; e

cR 3,4.4,l (f2FE'B At least once per 12 hours one RHR loop shall be verified in operation H and circulatina coolant at a flowrate of areater than or equal to 1000 gpm @ g CC: t---- :r;t re k.: th;n ;r ::::1 t: 110*FJ oc LA sa, 3.9.4. 2 ~-cru su.r s.9-ic e f BRAIDWOOD - UNITS 1 & 2 3/4 9-10 AMENDMENT NO. 25 R 6\\/ Q

CTS INSERT (S) D SECTION 3.9 .!O LC0 3.9.6 INSERT 3.9-10A '(M ). 3 SURVEILLANCE FREQUENCY SR 3.9.6.2 Verify correct breaker alignment and 7 days indicated power available to the required RHR pump that is not in operation. ~ INSERT 3.9 10B (Lu) NOTE f One required RHR loop may be removed from operation and ok considered OPERABLE: a. To support filling and draining the reactor cavity when aligned to, or during transitioning to or _from, the refueling water storage tank provided the required RHR loop is capable of being realigned to the Reactor Coolant System.(RCS): or b. To support required testing provided the required RHR loop is capable of being realigned to the RCS. INSERT 3.9 10C Deleted in Revision 0 f~ k 10/26/98 Revision Q

o 4 i t. 1-CTS DOCS ig s 4 3 4 f a i, t i s .t ' G-E i. 1-l' 1 1-1i l i-i> I e d' !O l O

p 'r - i i. o . DISCUSSION OF CHANGES TO' CTS [ ITS SECTION 3.9 REFUELING OPERATIONS e i :-' ' 1 -As Not:used. t }L 4 i;;y 4 1 1: d.1. t-O i BYRON /BRAIDWOOD UNITS 1 &"2 3.9 3a 10/26/98 Revision Q "*--s--w ~ m y.

4 i I 4 4 i LCO MARKUPS Y id' t i + 4 t f it i 4 4 t 1 e i t i 4 5 I I J ' 4, f i i i n J + a 1' 4 d 1 d 1 i 4 A k. 4 1 s 4 1 i l

Baron Concentration -3.9.1 O 3.9 REFUELING OPERATIONS 3.9.1 Baron Concentration LCO 3.9.1 . Boron concentrations of the Reactor Coolant System, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the COLR. i, . APPLICABILITY: MODE 6. i '. \\+ ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration .A.1 Suspend CORE Immediately not within limit. ALTERATIONS. AN,Q A.2 Suspend positive Immediately reactivity additions. AtiQ A.3 Initiate action to Immediately restore boron concentration to within limit. SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY 72 hours SR 3.9.1.1 Verify baron concentration is within the limit specified in.COLR. Y+he) O 3.9-1 Rev 1, 04/07/95 'WOG STS 0

RHR and Coolant Circulation-Low Water Level 3.9.6 '/ 'iA 3.9 REFUELING OPERATIONS '3.9.6 Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level LCO 3.9.6 Two RHR loops shall be OPERABLE, and one RHR loop shall be in operation. Nofe. (rn + cu v - to A ' APPLICABILITY: MODE 6 with the water level < 23 ft above the top of reactor vessel flange. g <d ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i-A. fleaa then the required A.1 Initiate action to Imediately n...ber of n"R h:p: ) . restore (seev+r-ee RHR OPERA 9tf. / Pv' loops to OPERABLE status. One. e,r-M>re. Mf,' b l00f4r 'l Mf444. f-.j ( ( P A.2 Initiate action to Imediately establish 2 23 ft of water above the top l of reactor vessel fl ange. +* B.. No RHR loop in B.1 Suspend operations Imediately operation. involving a reduction in reactor coolant boron concentration. AND (continued) L l l [) <\\J WOG STS 3.9-10 Rev 1, 04/07/95 $JAA Q y-

,g,. m, O LCO JFDS 1 I d O O

gm JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS t,j SECTION 3.9 REFUELING OPERATIONS P Condition A of ITS LCO 3.9.6 was revised to clarify the description of s the Condition and the associated Required Action. This change i eliminates the potential for misinterpretation of the Action requirements and provides consistency with the style and format of other similar ITS Action requirements. This is an additional enhancement only and does not involve a technical issue. lP Not used. 6 { \\_/ i p-BYRON /BRAIDWOOD UNITS 1 & 2 3.9 3 10/26/98 Revision Q

n BASES MARKUPS l 4 4 4 d i i i i I 1 J 3 4; ii i 1. ll n W

{ -

4 J l 1 i. 4 i d i 4 I a B

1 Baron Concentration 5 3.9.1 -.S.a an LCO s 0.95 is maintained during #uel handling operaticn:. (continued) Violation of the LCO could lead to an inadvertent criticality during MODE 6. APPLICABILITY This LCO is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical. Tne recuired boron concentration ensures a k., s 0.95. fAsev: " 0E 5, [mW rLcc ;,;,2, '= ,,, e ; ;,g,; , ;;0 ," ;ne (69,4-M y L CO 2.1. 2. "!HUTDC'.l" "^ RC !" '50") T. $ 200*r " ^ cure j that an adequate amount of negative reactivity is available 1 F4 to shut down the reactor and maintain it subtritical. "6 k e ACTIONS A.1(and)A.2 V$ Continuation of CORE ALTERATIONS or positive reactivity

dditions (including actions to reduce boron concentration) n OP2 is contingent upon maintaining the unit in compliance with lN i

the LCO. If the baron concentration of any coolant volume 3 V g gg g,,,p T inethe RCS, the refueling canal, or the refueling cavity is J less than its limit, 11 operations involving CORE i q ALTERATIONS S positi reactivity additions must be h suspended iately. gegT j Suspension of CORE ALTERATIONS and positive reactivity R ^ additions shall not preclude moving a component to a safe [f positiong 7 y g i 4 or %== <.mughe c-ht v \\u m e +oc u gc l s (emhe popese of sys+em wpua+a cu+ rot / u g amme me.ouw. j g In addition to immediately suspending CORE ALTERATIONS positive reactivity additions, Leprn:cquo restore the+ concentration must be initiated immediately g eq bore d CD" c - ~ "'" "'" ~ '" -' ' ^ dr deter-4-ine the re ::r:d creir2tice of boration flow Md ) rate and concentrationL c ur:::: De:te-5 sir Eve- % ust be satisfied. The only recuirement is to restore the boron concentration to its required value as soon as possible. In i order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions. 4 (cont' l l WOG STS B 3.9-3 Rev 1, 0-

~ ~. m BASES INSERT (S) N SECTION 3.9. [O. ~ Bases 3.9.1 INSERT' B 3.9 3A. (Pu) In MODES 1 and 2 with k,, a 1.0. LCO 3.1.4 " Rod Group Alignment Limits." LCO 3.1.5. " Shutdown Ba,nk Insertion Limits." and LCO 3 1.6. " Control Bank -Irsertion Limits " ensure an adequate amount of negative reactivity is available.to shutdown.the reactor. In MODE 2 with k,,,'< 1.0 and MODES 3. 4 and 5. LCO 3.1.1. " SHUTDOWN MARGIN (SDM)." ensures INSERT B 3.9 3 (P ) 4 an inadvertent criticality may occur due to.an incorrect fuel loading. To . minimize the. potential of an inadvertent criticality resulting from a fuel loading error. ' INSERT B 3.9 3C Deleted in Revision 0 i l i ~ l L. r rb 10/26/98 Revision 0 g l -. j.

RHR and Coolant Circulation-Low Water Level B 3.9.6 l U BASES (u alt,%4 one loop (EB@must be in operation in order LCO """4+4--"'a (continued) to provide: l a. Removal of decay heat; s i Ho i b.- Mixing of borated coolant to minimize the possibility >a i of criticality; and yp m l c. Indication of reactor coolant temperature, yc* An OPERABLE RHR loop consists of an RHR pump, a heat exchanger, valves, piping, instruments-and controls to l ensure an OPERABLE flow path."" * ^ "-+- 'm: t.5 c 10. rdi t tc := =J The flow path starts in one of the RCS hot i ' legs and is returned to the RCS cold legs. [.In3,,. # e 2. q.n e l 1 b l APPLICABILITY Two RHR loops are required to be OPERABLE, and one RHR loop l must be in operation in MODE 6, with the water level < 23 ft r ...), above the top of the reactor vessel flange, to provide decay L[q] e heat removal +. Requirements for the RHR System in teteee l %> g.___:rc:=,r:d by LCO: Srtie-2.4, R rt f C '=t 4 [Lp(u g/ 3 i n,. r r__i2_ ., e c______,. r_. c__,:_ l j ibh$t$$;i!!O) [RU' lool ' requirements in30bE USiEh the iY', J water level 2 23 ft are located in LCO 3.9.5, " Residual Heat g Removal (RHR) and Coolant Circulation-High Water Level." ( F' [N RM Sy,w,, y e I. cap +t,te of re=H T 't 4mv W Y Aal d'O af & kar* *L *I*". %ube;, l ACTIONS A.1 and A.2 @:(If 1::s than the recuired nebm u a RHR loopf WM," Ok ouse L t we m o u y m oc o action shall be immediately initiated and continued until h .; ~" lecc i:2 restored to OPERABLE status (=c 1; mr;;;;n) or until 2 23 ft of water level is established above the reactor vessel flange. When the water level is 2 23 ft above the reactor vessel flange, the Applicability changes to that of LC0 3.9.5, and only one RHR loop is required to i be OPERABLE and in operation. An immediate Completion Time i is necessary for an operator to initiate corrective actions. s b*'8*"'d._Y_".df* Y#b A (continued) i i l WOG STS B 3.9-22 Rev 1, 04/07/95 WQ u

l' 1 BASES INSERT (S) ,~V(N- 'SECTION'3.9 L Bases 3.9.6 t = INSERT B 3.9 22A (P )- a o L l'. 2. 3. 4.'and 5 are covered by LCO 3.4.6. "RCS Loops - MODE 4." LCO 3.4.7 "RCS Loops - MODE 5. Loops Filled." LCO 3.4.8. "RCS Loops'- MODE 5. Loops Not Filled." LC0 3.5.2. "ECCS - Operating." and LCO 3.5.3. "ECCS - Shutdown. " l l. INSERT B 3.9 22B (C and P ) 3 3 ~ However..the LCO is modified by a Note that permits the required RHR loop to l -be removed from operation and considered OPERABLE when aligned to, or during transitioning to or from. the Refueling Water Storage Tank (RWST) to support filling'or draining the refueling cavity or to support required testing. if capable of being realigned to the RCS. i I INSERT B 3.9 22C . Deleted in Revision 0 ? l. L l l l l l t I '.O V 10/26/9B Revision Q I L. !-r-

k. f 4 s. f i BASES JFDS-t 4 k I. - ii Y. b 'h 4 s 3 f 1 A l' i 4 4 i I I I I 1 4 1 l l f t

JUSTIFICATION'FOR DIFFERENCES TO NUREG 1431 BASES SECTION 3.9 REFUELING OPERATIONS P The-Applicable Safety Analyses section of the Bases for NUREG 3.9.2 23 incorrectly refers to the RCS boron concentration (which is addressed in LCO 3.9.1.) The statement is revised to refer to the unboratea water source isolation valves which are the subject of LCO 3.9.2. P,, 'The Bases for LCO 3.9.6 is revised to reflect changes made to LCO 3.9.6. 'Specifically, the Note to SR 3.9.6.1. added by TSTF-21, Revision 1 is positioned following-the LC0 and is reformatted consistent with other LC0 notes. This is an editorial enhancement only and does not involve any technical changes. s g c_ l P Not used. 3 P The Bases for SR 3.9.4.2 is revised by deleting the sentence. "The 3 . system actuation response time is demonstrated every 18 months. during refueling on a. STAGGERED TEST BASIS." These valves will continued to d be tested on an 18 month frequency but our current licensing basis does a not allow the flexibility of a STAGGERED TEST BASIS. This change is L4 consistent with Byron and Braidwood current licensing basis. O i l l l A iV BYRON /BRAIDWOOD UNITS 1 & 2 3.9 6 10/26/98 Revision Q g i

4 4 a m+ 4 J,_-e a4-aa h-u.e+*A-SA44 .-. ~ mee a-ham-e.A ep-- Eu--.8**cs**M-Ae-4--m em. 4 mm4m 4 -w s**-e-wwm-a----Aea,. L-m au O ITS SECTION 5,0 l 1 l 1 l i 1 i l i i i f 4 6 1 4 h 4 i - ?

h 4 ) I. CTS DOCS f f d I 4 W 4 9 i ' 4 N P. 4 4 N e we.. --,ee., t--ken _,-

l DISCUSSION OF CHANGES TO CTS ITS SECTION 5.0 ADMINISTRATIVE CONTROLS ADMINISTRATIVE CHANGES (A) A All reformattiag, renumbering and editorial rewording is in accoroance l 1 with the Westinahouse Standard Technical Specifications. NUREG-1431. During the development certain wording preferences or English language conventions were adopted. As a result. the Technical Specifications (TS) should be more readily readable, and therefore understandable, by plant operators and other users. During the reformatting. renumbering, and rewording process, no technical changes (either actual or interpretational) to the TS were made unless they were identified and Justi fied. ! A, CTS Definitions 1.20.a. Footnotes

• and **. and LC0 6.8.4.9 provide (in l

part) values and limitations applicable to Unit 1 operation prior to completing Cycle 7 (Braidwood) and Cycle 8 (Byron). These values and 3 limitations have been deleted. The deleted cycle specific values are 4 for fuel cycles that will no longer be applicable to operation of the y units since the applicable cycle will have been completed. The change is editorial in nature and does not involve a technical change (either actual or interpretational) to the TS. This change is consistent with u NUREG-1431. l A CTS Specification 6.2.2.d and Table 6.2-1. Footnote (e) have been 3 g deleted. The requirement for a licensed operator observing Core Alterations is contained in 10 CFR 50.54. Since conformance to 10 CFR is a condition of the license. specific identification of this requirement in the TS would be duplicative and is not necessary. Since the plant requirements remain the.same. the change is considered to be a change in presentation only. During this reformatting, no technical 3 changes (either actual or interpretational) were made to the TS unless i :-! they were identified and justified (Ref. Section 5.0. DOC L,). This ~ change is consistent with NUREG-1431. A, CTS Specification 6.6.1.a has been deleted. The requirement related to i reportable event action notification and submittal is contained in 10 CFP,50.73. Since conformance to 10 CFR is a condition of the I license, specific identification of this requirement in the TS would be duplicative and is not necessary. Since the plant requirements remain the same, the change is considered to be a change in presentation pnly. During this reformatting. no technical changes (either actual or interpretational) were made to the TS unless they were identified and l h-l justified (Ref. Section 5.0. DOC L,). This change is consistent with NUREG-1431. BYRON /BRAIDWOOD - UNITS 1 & 2 5.0 1 10/26/98 Revisien 0 ... ~ - -

t-DISCUSSION OF CHANGES TO CTS ITS SECTION 5.0 ADMINISTRATIVE CONTROLS 3 Ai CTS Specifications 6 8.1.c and 6.8.1.d have been deleted. Procedures to .I implement the station security plan and emergency response plan are required by 10 CFR 50 Appendix E and 10 CFR 50.54(p). Since conformance to 10 CFR is a condition of the license, specific i identification of.these plans in the TS would be duplicative and is not-necessary. Since the plant requirements remain the same.~the change is I considered to be a: change.in presentation only. During this reformatting no technical changes (either actual or interpretational) to - the.TS were made unless they were identified and justified (Ref. -- J Section 5.0. 00C L,). _This change is consistent with NUREG-1431. A - Not used.' 6 I. i O o L l l l'

i. :

f i f i l l,. l. F BYRON /BPAIDWOOD - UNITS 1 & 2 5.0 1a 10/26/98 Revision 0 ~ ,4,- .+c ..n.

1 DISCUSSION OF CHANGES TO CTS p ITS SECTION 5.0 ADMINISTRATIVE CONTROLS a A, Consistent with NUREG-1431. CTS 3.6.1.3.b is modified to add a "a" when referring to the air lock test pressure. The intent of tne CTS 1s to use a pressure sufficient to evaluate leakage not test at exactly P,. This change is administrative and is consistent with the CTS ~ Surveillance. A. (Byron Only) CTS Specification 6.9.1.4 has been revised to delete the reporting requirement for the initial annual report. This requirement is no 1.onger applicable to the operation of the units since the initial reporting period has been completed. This change is considered editorial in nature and does not involve a technical change (either actual or interpretational) to the TS. This change is consisttent with NUREG-1431. A, Not used. A CTS Specification 6.9.1.5 has been revised to add a clarification u statement denoting that a single annual report may be made for the facility. This change is necessary to eliminate the potential for misinterpretation of the reporting requirement. This change is perceived as the intent of the CTS requirements. is considered editorial in nature and does not involve a technical change (either actual or interpretational) to the TS. This change is consistent with NUREG-1431. A CTS S3ecifications 6.9.1.7 has been revised to modify the submittal date i u for t1e annual radioactive effluent release report. This change provides a reference to 10 CFR 50.36a in addition to the specified date contained in 10 CFR. This change does not involve a technical change since the reporting frequency has not changed. This change is considered ec..orial in nature and does not involve a technical change (either actual or interpretational) to the TS. This change is consistent with NUREG-1431. A CTS.Speci fications 6.9.1.8. 6.9.1.9 and 6.9.2 have been revised to u delete the reference to the submittal location for the monthly report, operating limits report. and special reports. The requirements related to report submittal are contained in 10 CFR. Since conformance to 10 CFR is a condition of the license. specific identification of this requirement in the TS would be duplicative and is not necessary. Since the plant requirements remain the same, the change is considered to be a change in presentation only. During this reformatting, no technical changes (either actual or interpretational) were made to the TS unless l-they were identified and justified (Ref. Section 5.0 DOC L,). This change is consistent with NUREG-1431. BYRON /BRAIDWOOD UNITS 1 & 2 5.0 2 10/27/98 Revision Q

DISCUSSION OF CHANGES TO CTS l(N-ITS SECTION 5.0 ADMINISTRATIVE CONTROLS , w) A CTS Specification 6.9.1.7. footnote **- has been revised to delete u information that'is not applicable to Byron and Braidwood. The radwaste l systems are common to the units, therefore, reporting releases from eacn l unit is not applicable. This change is considered editorial in nature and does not involve a technical change (either actual or interpretational) to the TS. This change is consistent with NUREG-1431. j , 7! An Not used. A CTS Specifications 6.12.1' and 6.14.1 have been revised to incorporate 3 references consistent with 10 CFR Part 20. Since the plant requirements remain the same, the change is considered to be a change in presentation only. During this reformatting no technical changes-(either actual or interpretational) to the TS were made unless they were identified and justified. This change is consistent with NUREG-1431. s l hiA Not used. 3 A,7 CTS Specification 4.0.5.a has been deleted. The requirement to perform ^ ASME Section XI testing.is denoted in 10 CFR 50.55a(g). Since conformance to 10 CFR is a condition of the license, specific identification of this requirement in the TS would be duplicative and is ln not necessary. Since the plant requirements remain the same. the change i () is considered to be a change in presentation only. During this reformatting, no technical changes-(either actual or interpretational) l were made.to the TS unless they were identified and justified d c-(Ref. Section 5.0. DOC L,). This change is consistent with NUREG-1431. i A CTS Specification 4.0.5.b has been revised to add a definition of the 3 biennially frequency for the IST 3rogram. This change provides only a t l clarification of the meaning of tie term and does not add any new requirement. This change is considered editorial in nature and does not i involve a technical change (either actual or interpretational) to the TS. -This change is consistent with NUREG-1431. i l l i I i BYRON /BRAIDWOOD - UNITS 1 & 2 5.0 3 10/27/98 Revision 0 l.-

DISCUSSION OF CHANGES TO CTS ITS SECTION 5.0 ADMINISTRATIVE CONTROLS ~ A The Explosive Gas and Storage Tank Radioactivity Monitoring Program n r includes a clarification statement denoting that the provisions of e (R 3.0.2 and SR 3.0.3 are applicable to this Program. This statement of applicability clarifies the allowance for surveillance frequency extensions and allowance to perform missed surveillances. This change -j is necessary since the CTS requirements. CTS LCO 3.11.1.4 and CTS 3 LC0 3.11.2.5. are being relocated from the TS, and the program described d in ITS Specification 5.5.12 is being added where the statements of applicability are generally not applied. Since this change maintains current requirements, it is considered a change of presentation method only. During this reformatting, no technical changes (either actual or interpretational) were made to the TS unless they were identified and justified. This change is consistent with NUREG-1431. 'A CTS 6.2.2.b is being deleted. The requirements of 10 CFR 50.54(m)(iii) 3 and 50.54(k) adequately provide for shift manning. These regulations. 50.54(m)(iii). require. "when a nuclear power unit is in an operational MODE other than cold shutdown or refueling. as defined by the unit's Technical Specifications. each licensee shall have a person holding a i senior operator license for the nuclear power unit in the control room ) g at all times. In addition to this senior operator, for each fueled nuclear power unit. a licensed operator or senior operator shall be o present at the controls at all times. Further, 50.54(k) requires. "An O4 o]erator or senior operator licensed ursuant to part 55 of this chapter H s1all be present at the controls at a 1 times during the o>eration of j the facility." STS 5.2.2b requirements will be met througl compliance with these regulations and is therefore, deleted from the CTS and not incorporated into the ITS. This is consistent with a letter from NRC. s 'T Chris Grimes to J. Davis dated April 9.1997 as stated in TSTF-258. . f G. j Reference Section 5.0. DOC L,. A (Byron Only) Consistent with the rest of CTS Section 6. CTS 6.9.1.4 is 3 revised to reference the " facility" versus " unit." Specification 6.1.9 l covers annual reports for which one report is prepared. covering the i operation of both units or the " facility." This change is considered a format change and is administrative. During this reformatting, no technical changes (either actual or interpretational) were made to the l TS unless they were identified and justified. This change is consistent j l' with NUREG-1431 philosophy. A CTS LC0 6.9.1.9 details the general topic associated with the listed 3 topical reports that are applicable to the COLR. These general topic 7 details are deleted. This is information-only content and does not ? reflect and technical or interpretational guidance. This change is r3 consistent with NUREG-1431. BYRON /BRAIDWOOD UNITS 1 & 2 5.0 6 10/26/98 Revision 0

DISCUSSION OF CHANGES TO CTS Q'O ITS SECTION 5.0 ADMINISTRATIVE CONTROLS A Based on utilization of WCAP-10216 as a basis for the F-Surveillance n Frequency in Specification 3.2. and the requirement stated in the WCAP's SER. reference to the WCAP 1s added to CTS Specification 6.9.1.9. Tnis is an administrative change reflecting commitments contained in the NRC's evaluation of the WCAP. Any technical changes made are discussed in the CTS markups for ITS Specification 3.2. A The CTS Table 4.4-2 10 CFR 50.72 (b)(2) reporting requirement is 'y deleted. Deletion of the CTS requirement does not change the requirement to report results that satisfy the criteria of 10 CFR 50.72 (b)(2). Therefore this change is considered a change of presentation method only. During this reformatting no technical changes (either .s interpretational) were made to the TS unless they were identifled and QN Justified (Ref. Section 5.0. DOC L,). This change is consistent with NUREG-1431. A Reference to CTS 4.0.5 in CTS 4.4.5.0 is deleted. The requirement to ~y perform ASME Section XI testing is denoted in 10 CFR 50.55a(g). Since conformance to 10 CFR is a condition of the license. specific identification of this requirement in the TS would be duplicative and is not necessary. Since the plant requirements remain the same. the change g is considered to be a change in presentation only. During this 3 reformatting, no technical changes (either actual or interpretational) p]' v A were made to the TS unless they were identified and justified (Ref. Section 5.0. DOC L,). This change is consistent with NUREG-1431. rl A Not used. 3 fA Consistent with NUREG-1431. as modified by TSTF-52. CTS 1.7 item d is 3 deleted. Item d simply provided a reference to another TS. The technical requirements remain the same. Therefore this change is considered a presentation preference. During this reformatting, no technical changes (either actual or interpretational) were made to the 9 TS unless they were identified and justified. la Hj A Not used. y BYRON /BRAIDWOOD - UNITS 1 & 2 5.0-7 10/26/98 Revision 0 i

I l DISCUSSION OF CHANGES TO CTS ITS SECTION 5.0 ADMINISTRATIVE CONTROLS l l L, CTS LCOs 3.7.6. 3.7.7. and 3.9.12 have been revised adding a statement j that verification of the specified flow rates may be accomolished during ( the performance of other specified surveillances within the ITS. Although ' accepted practice as related to the CTS.1t is not specifically l stated as in the ITS. Specifically stating that credit may be taken for l successful performance for the same Surveillance in another section of I the ITS is considered to be a less restrictive change. In addition. CTS I SRs.4.7.6.c.2). 4.7.6.d. 4.7.6.h.2). 4.7.6.j. 4.7.7.b.2). 4.7.7.c. 4.9.12.b.2), and 4.9.12.c have been revised to change the methyi lodine penetration values for the VC filtration system (makeup) from-0.175% to 9' 0.5%. VC Filtration System (Recirculation) from 1% to 4%, and the Nonaccessible Area Exhaust Filter Plenum Ventilation System from 1% to ) [ 4.5%. During the conversion process from CTS to ITS and additional discussion with the NRC reviewer. Comed committed to comply with l ASTM D 3803-1989 standards. Based on this standard, the subject j penetration values were relaxed resulting in a less restrictive change. l This change is consistent with NUREG-1431. L. CTS Section 5.0 contains several statements referenc' ng specific NRC 6 Regulations and Standards (i.e. 10 CFR 20. Regulator, Guide 1.52, or 5 ASTM-975). Since these are redundant to the appropriate NRC Standard, Q the NRC has agreed to remove the redundant information from the CTS. l Although no changes in Comed's commitment to the appropriate NRC Standards is being made, removal of the information is considered to be l a Less Restrictive change. This change is consistent with NUREG-1431. v ( l l I l BYRON /BRAIDWOOD UNITS 1 & 2 5.0-23a 10/26/98 Revision 0

".4 3 h 4 4enWM -W4 A4 e #qink.m A *. M M d Wh we,W.4-Ah A4. + E ww we w-4--_.A_v- _w--+ww-,s_wm B NSHCS I i 4 I iI. 4 3 r s 4 i i: 4 d 2 i i ,k I ei-i l

r NO SIGNIFICANT HAZARDS EVALUATION /O ITS SECTION 5.0 ADMINISTRATIVE CONTROLS V l TECHNICAL CHANGE LESS RESTRICTIVE "Soecific" _ l ("L," Labeled Comments / Discussions) ' Commonwealth Edison Company (Comed) has evaluated each of the proposed Technical Specification changes identified as " Technical Change - Less Restrictive (Specific)" in accordance with the criteria set forth in ! 10 CFR 50.92 and has determined that the proposed changes do not involve a l significant hazards consideration. The bases for the determination that the proposed changes do not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the evaluation are presented below. 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? Some of the information in the CTS is descriptive in nature and incorporates redundant information as stated in the 10 CFR. Standard Review Plan. Regulatory Guides or other NRC regulations. This redundant information pertains to various equipment. system (s). actions. Programs or surveillances. The NRC has previously approved removing this q redundant information and detail from the CTS to a licensee controlled Cf document. reference the appropriate NRC regulation in a Program, or simply referencing the approariate NRC Standard (i.e. 10 CFR or Regulatory Guide). The NRC 1as agreed that the inclusion of this redundant information in the ITS is not necessary to adequately protect the health and safety of the public. Although this information is being moved or referenced. 10 technical requirements or commitments for compliance are being changed. The only change is that instead of providing this redundant information in the ITS. credit is being taken for the same information and requirements in the appropriate NRC . Standards. This is considered to be a Less Restrictive change since the information is being removed from the CTS and not being included in the ITS. This relaxation will not alter the operation of any plant equipment. reduce any Comed commitments to appropriate NRC Standards, nor increase any failure probability for evaluated accidents in the UFSAR. The probability that equipment failures resulting in an analyzed event will occur is unrelated to this change. As such, the probability of occurrence for a previously analyzed accident is not significantly increased. \\ 'l BYRON /BRAIDWOOD - UNITS 'l & 2 5.0 21c 10/26/98 Revision Q

i l i'n NO SIGNIFICANT HAZARDS EVALUATION Q ITS SECTION 5.0 ADMINISTRATIVE CONTROLS l l 2. Does the change create the possibility of a new or different kind of l accident from any accident previously evaluated? i The proposed change does not involve a physical alteration to the plant. No new equipment is being introduced and no installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated nor in the setpoints which initiate protective or mitigative actions. No technical change is being proposed to the procedures governing normal plant operation of those relied upon to mitigate a design basis event. l Removing the redundant information and requirements from the CTS and l l taking credit for the appropriate NRC Standard (i.e. 10 CFR or j Regulatory Guides} does not alter any Comed commitments for compliance with the appro]riate NRC Standards nor introduce any new failure modes. In addition, t1e changes do not alter assumptions made in the safety analysis and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the change involve a significant reduction in a margin of safety? The margin of safety is determined by the design and qualification of / the plant equipment, the operation of the plant within analyzed limits, \\ and the point at which protective or mitigative actions are initiated. Removing the duplicative information and requirements from the CTS and taking credit for the appropriate NRC Standard (i.e. 10 CFR or Regulatory Guides) does not alter any Comed commitments for compliance with the NRC Standards. The proposed change has no effect on the assumptions of the design basis accident. This change has no impact on the safe operation of the plant. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change Therefore. this change does not involve a significant reduction in the margin of safety. l l B'rRON/BRAIDWOOD UNITS 1 & 2 5.0 21d 10/30/98 Revision 0

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