Text

Proposed Tech Specs Re Administrative Controls
	ML20149H179
Person / Time
Site:	Clinton
Issue date:	02/05/1988
From:	; ILLINOIS POWER CO.
To:	;
Shared Package
ML20149H162	List: ML20149H160; U-601118, Proposed Tech Specs Re Administrative Controls;
References
	U-601118, NUDOCS 8802190165
	Download: ML20149H179 (86)
	v • d • e

_ _ _ _ _ _ _ _ - _ _ _ _

Attachmsnt 3 to U-601118 Page 2 of 87 gMINISTRATIVECONTROLS UNIT STAFF (Continued)

f. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions e.g., licensed Senior Operators, licensed Operators, health physicists, aux.iliary operators, and key maintenance perstinnel.

The amount of overtime worked by unit staff members performing safety-related functions shall be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the unit is operating. However, in the event that unforeseen problems reauire substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major unit modifications, on a temporary basis the following guidelines shall be followed:

1. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months l

straight, excluding shift turnover time.

1

2. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any seven day period, all excluding shif t turnover time.

3. A break of at least eight hours should be allowed between work periods, including shift turnover time.

4. Except during extended shutdown periods, the use of overti.me should be considered on an individual basis and not for the entire staff on a shift.

Any deviation,from the above guidelines shall be authorized by the(Power [ i Plant Manager) or his deputy, or higher levels of management, in accordance g (with established procedures and with documentation of the basis for l' granting the deviation. Controls shall be included in the orocedures _such that individual overtime shall be reviewed monthly by the(Rower PlantF" Lanageffor his designee to assure that excessive hours have not been ,

assigned. Routine deviation from the above guidelines is not authorized,

~

repktcr taiTh " M.wty- Chtm Axe hhm "

880219016', 880205 PDR ADOCT 05000461 P DCD CLINTON - UNIT 1 6-2

Attachmsnt 3 to U-601118 Page 3 of 87 ADMINISTRATIVE CONTROLS AUTHORITY 6.5.2.9 The NRAG shall report to and advise the Vice President on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.

RECORDS 6.5.2.10 Records of NRAG activities shall be prepared, approved, and distrib-uted as indicated below:

a. Minutes of each NRAG meeting shall be prepared, approved, and forwarded to the Vice President within 14 days following each meeting.

b. Reports of reviews encompassed by Specification 6.5.2.7 shall be prepared, approved, and fo marded to the Vice President within 14 days following completion of the review.

c. Audit reports encompassed by Specification 6.5.2.8 shall be fo marded to the Vice President and to the management positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization.

62 5.3 TECHNICAL REVIEW AND CONTROL ACTIVITIES Procedures required by Technical Specification 6.8 and other orocedures which affect plant nuclear safety as determined by th Power Plant Manage % nd changes l thereto, other than editorial or typographical c anges, shall oe reviewed as follows:

6.5.3.1 TECHNICAL REVIEW

a. Each such procedure or procedure change shall be independently reviewed ,

by an individual knowledgeable in the area affected other Dn the in_d_ivid- 1 ual who prepared the procedure, or procedure change. The(Power Plant)"

A fanagej shall prior to implementation approve all plant procedures and changes thereto,

b. Individuals responsible for reviews performed in accordance with Item 6.5.3.la. above shall be members of the plant staff previously I designated by the4 Power Plant Manans f'Each such review shall include l ;

(a determination of whether or not additional, cross-disciplinary, review i is necessary. If deemed necessary, such review shall be performed by l the review personnel of the appropriate discipline. -

Individuals performing these revit.ws shall meet or exceed the qualifi-cations stated in A%I/ANS 3.1-1978 for the appropriate discipline.

a rt plxa. uA

M wy.r- Cli,b be.c bhon CLINTON - UNIT 1 6-13

Attachesnt 3 to U-601118 Pags 4 of 87 ADMINISTRATIVE CONTROLS TECHNICAL REVIEW (Continued)

c. When required by 10 CFR 50.59, a safety evaluation to determine whether or not an unreviewed safety question is involved shall be included in the procedure or the procedure change review. If it is determined that an unreviewed safety question is not involved, a written safety evaluation to support that decision will be prepared and submitted to the FRG for !

review. Pursuant to 10 CFR 50.59, NRC approval of items involving unre- , !

viewed safety questions shall be obtained prior to the(Eower Plant Managej ~ ll '

approval for implementation. ,, [ ,

d. Written records of reviews performed in accordance with Item 6.5.3.1.a.

above, including recomendations for approval or disapproval, shall be '

prepared and maintained.

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the FRG, and submitted to the NRAG and the Vice President.

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

a. In accordance with 10 CFR 50.72, the NRC Operations Center shall be noti-fied by telephone as soon as possible and in all cases within I hour after the violation has been determined. The Vice President and the NRAG shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. A License Event Report shall be prepared in accordance with 10 CFR 50.73.

c. A Safety Limit Violation Report shall be prepared. The report shall be re-viewed by the FRG. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon unit components, .,

systems, or structures, and (3) corrective action taken to prevent recurrence,

d. The Safety Limit Violation Report shall be submitted to the Comission within 30 days and to the NRAG, and the Vice President within 30 days of the violation. .

e. Critica; operation of the unit shall not be resumed until authorized F/

the Commission. l CLINTON - UNIT 1 6-14

Attachmsnt 3 to U-601118 Pago 5 of 87 1

PACKAGE NUMBER 2 Description and Justification of Proposed Change Illinois Power is requesting a change to Technical Specification 3.3.7.5. The proposed change would add a note to item 14 on Table 3.3.7.5-1 to allow inoperability of primary containment isolation valve position indication when the valve / valve operator is electrically deactivated in the isolated position, such as when a containment isolarAon valve is deactivated and secured in the isolated position (for maintaining containment integrity) according to the provisions of ACTION "a" under Specification 3.6.4.

When any motor-operated cantainment isolation valve is electrically deactivated, its breaker is usually switched to the "0FF" position.

Power is consequently lost to the position indication circuit and all position indication in the main control room is lost. ACTION 82 associated with iten 14 of Table 3.3.7.5-1 will require a plant shutdown after 7 days if all position indication to a primary containment isolation valve is lost. ,

The intent of Technical Specification 3.3.7.5 (Table 3.3.7.5-1 item 14) is to ensure that indication is available in the main control room to enable the control room operators to assess containment integrity. The intent of this specification can be satisfied if the affected valve (s) are placed in the isolated position prior to being deactivated, and if appropriate administrative controls are in place to ensure that the control room operators can determine the valve's position if needed.

Intermittent reopening of these valves under administrative controls should be allowed since the control room operators will be directly '

involved in the valve manipulation and since such a provision is currently specified in Technical Specification 3.6.4.

Basis For No Significant Haznrds Consideration According to 10CTR50.92, a proposed change to the license (Technical Specifications) involves no significant hazards consideration if operation of the facility it. accordance with the proposed change would ,

not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the l possibility of a new or different kind of accident from any accident l previously evaluated, or (3) involve a significant reduction in the !

margin of safety.

(1) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because, under the provisions of the proposed change, the affected valve (s) will be in the isolated / conservative position (except as limitedly provided) . With the proposed change a means still exists that allows plant operators to know the position of an affected containment isolation valve (s). The proposed change is consistent ;

with the provisions of Technical Specification 3.6.4. "Containment l Isolation Valves."

I

e Attech::nt 3 5 to U-601118 Page 6 of 87 (2) The proposed change does not create the possihA11ty of a new or different kind of accident from any accident previously evaluated because the proposed change introduces no new modes of operation with respect to existing requirements and provisions contained in the Technical Specifications. The proposed change does not involve any changes to the plant's as-built design.

(3) The proposed change does not involve a significa.qt rgduction in a ,

margin of safety because this change does not reduce the , capability of assessing containment isolation during accident conditions or i when otherwise required.

i P

0 i

4 1

i i !

l I

i l

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

5

.e .

Attachmint 3 .

to U-601118 Pega 7 of 87 INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION l LIMITING CONDITION FOR OPERATION :

3.3.7.5 The accident monitoring instrumentation channels shown in Table .

i

. 3.3.7.5-1 shall be OPERABLE. ;

4 APPLICABILITY: As shown in Table 3.3.7.5-1.

ACTION:

With one or more accident monitoring instrumentation channels inoperable, take the ACTION required by Table 3.3.7.5-1. l 4 SURVEILLANCE REQUIREMENTS 4.3.7.5 Each of the above required accident monitoring instrumentation ;

=

4 channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK 1

and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7 5-1.

1 4

i i

! I l !

+

(

i i

)

4 j rdo chu3c b E6 N~ 2 prov de.d odg be coa N 5 h < )

i 4

CLINTON - UNIT 1 3/4 3-86 3

TABLE 3.3.7.5-1 8 ACCIDENT MONITORING INSTRUMENTATION

'i -

o MINIMUM APPLICABLE 7 REQUIRED NUMBER CHANNELS OPERATIONAL c OF CHANNELS OPERABLE CONDITIONS ACTION 3 INSTRUMENT 2 1 1,2,3 80

- 1. Reactor Vessel Pressure 1,2,3 80

2. Reactor Vessel Water Level 2 1 4 2 1, 2, 3 80

3. Suppression Pool Water Level

4. Suppression Pool Water Temperature 2/quadrantt 1/quadrantt 1, 2, 3 80 2 1 1, 2, 3 80

5. Drywell Pres ure 1,2,3 80 2 1

6. Drywell Air Temperature

7. Drywell/ Containment Hydrogen and Oxygen Concentration Analyzer and Monitor 2 1 1,2,3 80

8. Containment Pressure ## 2/ division 1/ division 1, 2, 3 80 Containment Temperature 2 1 1,2,3 80

9. 80

10. Safety /Nelief Valve Acoustic Monitor 1/ valve *** 1/ valve *** 1, 2, 3 y

11. Containment /Drywell High Range Gross Gamma Radiation Monitors 4** 2* 1,2,3 81 y 1 1 1,2,3 81 3 12. HVAC Stack High Range Radioactivity Monitor #

13. SGTS Exhaust High Range Radioactivity Monitor # 1 1 1,2,3 81

14. Primary Containment Isolation Valve Position I Indication it 2/ valve 1/ valve # # 1,2,3 82 TABLE NOTATIONS

One each for containment and drywell.

- Two each for containment and drywell.

- - Thermocouples in the SRV discharge line can serve as backup to the acoustic tail pipe monitors indication should one channel of the position indication become inoperable.

High range noble gas monitors and iodine / particulate sampler. ,o>

1. For Divisions I and II only. "

a*

f f These instruments monitor suppression pool water temperature when pool water level is below instruments ?%

of Specification 3.5.3.1.

8; i tt One channel consists of the open Ifmit switch, and the other channel consists of the closed limit switch L for each automatic isolation valve in Table 3.6.4-1 Part 1, "Automatic Isolatinn Valves." &Cg

\ggg Insert attached h-n

w

Attachmint 3 to U-601118 Page 9 of 87 Nok 2pglic ablt i V2fVe fos ion indic2kion is unavailable b.cause +ke valve was d liberately deactivated, provid.J the valve is in the isolated position and adminisf rative controls are in p lace to ensure + Int +Le con 4rol room ep.rators can de+ ermine the valve's position , i{needed. Vaives close d in accordanc, with +hese conditions may be. reopened on an intermittent basis under adminis; rativ e.

controls .

l l

l

Attechm:nt 3 to U-601118 ,

Pag:10 of 87 1 TABLE 3.3.7.5-1 (Continued)

ACCIDENT MONITORING INSTRUMENTATION ACTION ACTION 80 - a. With the number of OPERABLE accident monitoring instrumenta-tion channels less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With the number of OPERABLE accident monitoring instrumen-tation channels less than the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1, restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 81 - With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable Channel (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or:

a. Initiate the preplanned alternate method of monitoring the appropriate parameter (s), and

b. Prepare and submit a Special Report to the Commission pur-suAnt to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inopera-bility and the plans and schedule for restoring the system to OPERABLE status.

ACTION 82 -

a. With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3.7.5-1, verify the valve (s) position by use of alter-nate indication methods; restore the inoperable channel (e) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the fol-lowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1, verify the valve (s) position by use of alternate indication methods; restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUT-DOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in 00LD SHUT 00WN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

%eby.w Ns m e. -

1 pmad enh fee con % ig CLINTON - UNIT 1 3/4 3-88

Attachm nt 3 to U-601118 CONTAINMENT SYSTEMS

11 f" 2/4.6.4 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.4 The containment isolation valves and the instrumentation line excess flow check alves shown in Table 3.6.4-1 shall be OPERABLE with isolation times less than or aval to those shown in Table 3.6.4-1.

APPLICABILITY: As shown in Table 3.6.4-1.

ACTION:

a. With one or more of the containment isolation valves shown in Table 3.6.4-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1. Restore the inoperable valve (s) to OPERABLE status, or

2. Isolate each affected penetration by use of at least one deactivated automatic valve secured in the isolated position,*t or

3. Isolate each affected penetration by use of at least one closed manual valve or blind flange.*t The provisions of Specification 3.0.4 are not applicable provided the affected penetration is isolated in accordance with ACTION a.2 or a.3 above, and provided the associated system, if applicable, is declared inoperable or appropriate ACTION statements for that system are performed.

Otherwise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Otherwise, in OPERATIONAL CONDITION **, suspend all operations involving CORE ALTERATIONS, handling irradiated fuel in the secondary containment, or with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

We e m p +o Ats p%t. -

pr. d d enty b cm%g ,

"Isolation valves closed to satisfy these requirements may be reopened on an intermittent basis under administrative controls.

- When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel, tContainment Isolation Valves can have dual functions in that they provide l both containment isolation and Emergency Core Cooling functions. Any l inoperable dual function vahe coulo degrade the valves' other function.

CLINTON - UNIT 1 3/4 6-29

)

1 1

. . i Attachment 3 to U-601118 Page 12 of 87 i

PACKAGE NUMBER 3 Description and Justification of Proposed Change This propored change request incorporates the Startup Test Program data into the CPS Technical Specifications. IP committed to provide this data within 90 days of the completion of the Startup Test Program. The specific changes are summarized below.

1) -page 3/4 3-21: item 2.d - delete the "**" from both the trip setpoint and allowable value columns.

2) page 3/4 3-22: item 4.a - change the trip setpoint from " $257.5" to " $110" and the allowable value from "$ 266" to " 1118.5."

Delete the "**" from both the trip setpoint and allowable value columns.

3) page ?/4 3-23: item 4.1 - delete the "**" from both the trip setpoint and allowable value columns.

4) page 3/4 3-24: item 5.e - delete the "**'" from both the trip setpoint and allowable value columns.

5) page 3/4 3-24: Note "**" - delete this entirely.

6) page 3/4 3-66: item 1.a - delete "(*)% of RATED THERMAL POWER" and insert "124.5 + 1.1, -0 pois*" in the trip setpoint column and delete "(*)% of RATED THERFAL POWER" and insert "124.5 + 50.5,

-23.0 pois*" in the allowable value column.

)

f 7) page 3/4 3-66: item 1.b - delete "(*)% of RATED THERMAL POWER" and I insere "364.3 + 0, -3.6 psig*" in the trip setpoint column and delete "(*)% of RATED THERMAL POWER" and insert "1400 psig*" in the allowable value column.

8) page 3/4 3-67: Note "*" - replace with "These values are for turbine first stage pressure."

9) page 3/4 4-2: item a.1.f - replace "33,000 spm*" with "31,341 sym*." ,

10) page 3/4 4-2: item a.1.g - change "50%" to "30%".

11) page3{44-2: items e and d - replace "(39)# 1" and "(39)% " with "35.5%

12) page 3/4 4-2: Note "*" - delete "The actual value to be applied will be determined during the Startup Test Program."

13) page 3/4 4-2t Note: "**" - delete "Initial values. Pinal values to be determined during Startup Testing based upon." Change "the" to "The" and leave the rest of the Note as is.

l

Attachr.ent 3

to U-601118 Page 13 of 87

14) page 3/4 4-2: Note "#" - delete "Value to be established during Startup Test Program." Delete the parenchesis from the remaining sentence.

t 1

15) page 3/4 4-3: item 4.4.1.1.3.c - replace "33,000 gpm**" with "31,341 gpm**." ,

16) page 3/4 4-3: item 4.4.1.1.3.d - replace "(39)#1" with "35.5%'."

17) page 3/4 4-3: Note "f" - delete "Value to be established during l Startup Test Program". Delete the parenthesis from the remaining 3

sentence. ,

10) page 3/4 4-4: item 4.4.1.1.4 - replace "50%" with "30%."

)

4 19) page 3/4 4-4t Note "*" - delete "Initial values. Final values to be determ&ned during Startup Testing based upon." Change "the" to '

"The" and leave the rest of the Note as is.

20) page 3/4 4-6: item 4.4.1.2.b.1,.2 3 - delete the "*" in each item.

21) page 3/4 4-6: Note "*" - delete this entirely.

22) page 3/4 4-19: ACTION c.1 - delete "*."

23) page 3/4 4-19: Note "*" - delete this entirely.

24) page B 3/4 4-1: second paragraph - delete the "( )" from around r

"(30%)*" and replace "(50%)*" with "30%*". Change footnote "*" by deleting "Initial Values. Final values to be determined during Startup Testing based on." Capitalize the word "the" and leave the rest of the sentence as is.

Basis for No Significant Hazards Consideration According to 10CTR50.92, a proposed change to the license involves no significant hazards consideratien if operation of the facility in ,

~

accordance with the proposed change would not (1) involve a significant '

increase in the probability or consequences of an accidant previously evaluated, or (2) create the possibility of a new or different kind of ,

accident from any accident previously evaluated, or (3) involve a significant' reduction in a margin of safety.

(1) The proposed changes do not involve a significant increase in the possibility or consequences of an accident previously evaluated :

because these changes are either (a) administrative (deleting extraneous or outdated information of ,

an administrative nature), or l r

i

Attacheont 3 i 1

to U-601118 Page 14 of 87 (b) are required to provide the correct setpoints and limits ,

determined to be appropriate for the as-built plant design and are consistent with the applicable safety analyses. They are based on the reviewed and approved test results from the Startup Test Program and have been incorporated into the appropriate plant design documents.

(2) The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed values are consistent with plant design and/or are based on the operational performance of the as-built plant :

during the Startup Test Program. The values support a plant-specific design basis for Clinton, so their incorporation into the plant design documents and the CPS Technical Specifications doas mot create the possibility of a new or different kind cf accident from ecy previously evaluated.

(3) The proposed changes do not involve a slynificant reduction in a margin of safety because the values proposed have been determined to be consistent with plant design and the corresponding accident analyses. In some cases, the proposed values are values established to support a margin of safety since such values had not been previously established. The proposed changes do not therefore constitute a reduction in a margin of safety.

t I

e i

a I

TABLE 3.3.2-2 (Continued)

E

~

CRVICS INSTRUMENTATION SETPOINTS

'i E ALLOWABLE VALUE

TRIP SETPOINT TRIP FUNCTION

$ 1. PRIMARY AND SECONDARY CONTAINMENT ISOLATION (Continued) i 3.00 psid

k. Containment Pressure - High 1 2.62 psid

[

Main Steam Line Radiation - High i 3.0 x full power background 1 3.6 x full power background 1.

m. Fuel Building Exhaust Radiation -

High 1 10 mR/hr i 17 mR/hr NA NA

n. Manual Initiation

2. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level - g -147.7 in.

Low Low Low, Level 1 1 -145.5 in.*

Main Steam Line Radiation - High 1 3.0 x full power background 1 3.6 x full power background b.

Main Steam Line Pressure - Low 1 849 psig 1 837 psig c.

Main Steam Line Flow - High i 170 psi d i 178 psf N l d.

Condenser vacuum - Low 1 8.5 in. Hg vacuum 1 7.6 in. Hg vacuum e.

f. Mair. Steam Line Tunnel leep. - High i 165'F i 176*F

g. Main Steam Line Tunnel a Temp. - High 1 54.S*F 1 60*F

h. Main Steam Line Turbine Bldg. ygg Temp. - High 1 131.2*F i 138'F Manual Initiation NA NA $

i.

o S, 8,

3. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. A Flow - High 1 59 gpa 1 66.1 gpa 5[

b. A Flow Timer . > 45 sec. 1 47 sec.

TABLE 3.3.2-2 (Continued)

CRVICS INSTRUMENTATIO84 SETPOINTS

i E

TRIP SETPOINT Alt 0WABLE VALUE TRIP FUNCTION

$ 3. REACTOR WATER CLEANUP SYSTEM ISOLATION (Continued)

[ c. Equipment Area Temp. - High

1. Pump Rooms - A, B, C -< 186.5"F -< 197.1*F

2. Heat Exchanger Rooms - East, West < 201*F 1 212*F

d. Equipment Area A Temp. - High

1. Pump Rooms - A, B, C 1 54.5*F < 60*F

2. Heat Exchanger Rooms - East, West 1 54.5'F i 60*F -

w e. Reactor Vessel Water Level -

1 Low Low, Level 2 1 -45.5 in." 1 -47.7 in.

w

f. Main Steam Line Tunnel Ambient Temp. - High 1 165'F < 176*F

g. Main Steam Line Tunnel a Temp. - High 1 54.5'i 1 60*F SLCS Initiation NA NA h.

i. Manual Initiation MA NA

4. REACTOR CORE ISOLATION COOLIFG SYSTEM ISOLATION [dllo $ ggg,3 l

a. RCIC Ste m Line Flow - High 6 257 in. H 2O (<266 fili.H0*2

b. RCIC Steam Line Flow - Hinh Timer 1 3 sec. s.13 sec. m, o n

c. RCIC Steam Supply Pressure - Low 10 6 psig 1 52 psig 5 c: "

- e :r

d. RCIC Turbine Exhaust Diaphragm *SS '

Pressure - High < 10 psig < 20 psig A"5g 2

4 x g : ?

p, ; i g d 5 i .

s ,

~

i- , .

TPa!E1.3.2-2 (Continued'j

~ '

9 t

!CRVICS INSTRUMENTATION SETPOINTS E

z

' TRIP SETPOIhT ALLOWABLE VALUE TRIP FUNCTICN E

Z 4. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION (Continued) e

e. RCIC Equipment Room Ambient Temp. - High 5 222.5'F 5 233.1*F

f. RCIC Equipment Room A Temp. - High 1 34.5*F $ 40 F

g. Main Steam Line Tunnel Ambient Temp. - High i 165*F $ 176*F 4

h. Main Steam Line Tunnel A Temp. - High 5 54.5 F 5 60 F w

1 Main Steam Line Tunnel w i.

Temp. Timer > 25 min. 5 28 min.

4 Drywell Pressure - High i 1.68 psig i 1.88 psig J.

NA NA

k. Manual Initiation l

1. RHR/HCIC Steam Lt.ne Flow - High i 179.5 in. H 2 5 188 in.2 H M

m. RHR Heat Exchanger ?, B Ambient Temperatu.-e High 5 138.5'F 5 149.6*F

n. RHR Heat Exchanger A, B A, Temp. - High 5 74.2*F i 79.6*F oa, on n :>.

5. RHR SYSTEM ISOLATION 5 "

c,., a

~ os 2

a. RHR Heat Exchanger Rooms A, B ~og Ambient Temperature - High 1 138.5*F 5 149.6*F o-a 4 ,- ~n

=

RHR Heat Exchanger Rooms A, B "

b. 0 A Temperature - High 5 74.2 F 5 79.6 F

_ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ _ _ _ _ _ - - _ _ _ _ - - - - - _ - _ _ _ _ _ m ___ ___-_ - _ _ . - - _ _ _

TABLE 3.3.2-2 (Continued)

?** CRVICS INSTRUNENTATION SETPOINTS

'i E ALLOWABLE VALUE TRIP FUNCTI9N TRIP SETPOINT E

21 5. RHR SYSTEM ISOLATION (Continued) w

c. Reactor Vessel Water Level -

Lov, Level 3 > 8.9 in.* 1 8.3 in.

d. Reactor Vessel Water Level - -

> -147.7 in.

Low Low Low, Level 1 1 -145.5 in.*

e. Reactor Vessel (RHR Cut-in 5150psifEE) l Permissive) Pressure - High 1135psidEE3

f. Drywell Pressure - High

< 1.68 psig < 1.88 psig is Ocr.t:fn=ent ('neav 51.88psig 51.68psig

~

}l 2) Fuel Pool Cooling

. NA Manual Initiation NA g.

See 8ases Figure 8 3/4 3-1.

Final setpoint to be determined during startup test program. Any required change

- I.iitial setpoint.

to this setpoint shall be submitted to the Commission within 90 days of test completion.

U O M

%c"

E

= w

TABLE 3.3.6-2 b CONTROL R00 BLOCK INSTRUMENTATION SETPOINTS

'i ALLOWABLE VALUE TRIP SETPOINT

@ TRIP FUNCTION

1. R00 PATTERN CONTROL SYSTEM geence wtt Amitco Low Power Setpoint *)% of RATED THERMAL POWER (*)% of RATED THERMAL POWER y a.

(*)% of RATED THERMAL POWER g b. RWL High Power Setpoint *)% of RATED THERMAL POWER

2. APRM

a. Flow Biased Neutron Flux

- Upscale < 0.66 p >;) + 42%** < 0.66 (W-AW) + 45%**

NA HA

b. Inoperative

c. Downscale 1 5% of RATED THERMAL POWER 1 3% of RATED THERMAL POWER

d. Neutron Flux - Upscale Startup 5 12% of RATED THERMAL POWER $ 14% of RATED THERMAL POWER

$ 3. SOURCE RANGE MONITORS NA Y a. Detector not full in NA

< 1 x 105 cps < 1.6 x 105 cps g b. Upscale HA

c. Inoperative NA 1 3 cps 2 1.8 cps

d. Downscale

4. INTERMEDIATE RANGE MONITORS .

NA

a. Detector not full in NA

< 110/125 division of full scale

b. Upscale < 108/125 division of full scale NA

c. Inoperative NA

d. Downscale 1 5/125 division of full scale 1 3/125 division of full scale

5. SCRAM DISCliARGE VOLUME Water Level-High, C11-N602A < 12" # < 19 7/8" #

a.

b. Water Level-High, C11-N602B 312"## 5197/8"##

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

[o{

a. Upscale 5 10S% of rated flow $ 111% of rated flow (({

@ O9

7. REACTOR MODE SWITCH o{$

NA m w

a. Shutdown Mode NA "

NA NA

b. Refuel Mode

'*' ' 'Attachmsnt 3 to U-601118 Pags 20 of 87 l

Trip SetooM MlowWa.Wla Q. b hoWel' bY f vhi . l7 4 5 +3.1 y -O p/3 IM.5 + 50.5; -23.0 pi3

b. bl. Nijh fonc $tpiht- %44 PDj -2.(e fai[ 4 1 400 fM 9

Attachmant 3 to U-601118 TABLE 3.3.6-2 (Continued) Page 21 of 87 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS l

TABLE NOTATIONS

ITo be determined during startup test program. The actual setpoints are

~

the corresponding values of the turbine first stage pressure for these Lpowerlevelsr~~~

3

- The Average Pover Range Monitor rod block function is varied as a function of recirculation loop flow (W). The trip setting of this function must be maintained in accordance with Specification 3.2.2, and note (a) of Table 2.2.1-1.

Instrument zero is 758' 5" msl.

1. Instrument zero is 758' 4 1/2" msl.

N% Yalues arc Opc he Line_, [icst stay 'Peteore .

l 4

i l

i I

CLINTON - UNIT 1 3/4 3-67 i

1

Attachmant 3 3/4.4 REACTOR COOLANT SYSTEM $g 2 87 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system . recirculation loops shall be in operation j with:

a. Total core flow greater than or equal to 45% of rated core flow, or

b. THERMAL POWER within the unrestricted zone of Figure 3.4.1.1-1, or

c. THERMAL POWER within the restricted zonet of Figure 3.4.1.1-1 and APRM or LPRMtt noise levels not larger than three times their established baseline noise levels.

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a. With one reactor coolant system recirculation loop not in operation:

1. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

a) Place the recirculation flow control system in the Local Manual (Position Control) mode, and b) Reduce THERMAL POWER TO < 70% of RATED THERMAL POWER, and c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.07 per Specification 2.1.2, and d) Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit to a value of 0.85 times the two-recirculation-loop operation limit per Specification 3.2.1, and e) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block Trip Setpoints and Allowable Values to those applicable for single-recirculation-loop operation per Specifications 2.2.1, 3.2.2, and 3.3.6, and -

See Special Test Exception 3.10.4.

tThe operating region for which monitoring is required. See Surveillance Requirement 4.4.1.1.2.

ttDetector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

do choge. h iLis [^3<-

providd ontg forconhou.

CLINTON - UNIT 1 3/4 4-1

Attachmsnt 3 to U-601118 Page 23 of 87 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION f) Reduce the volumetric f, low rate of the operating recirculation loop to < [33,000 gpfm"

, g l g) Perform Surveillance Requirement 4.4.1.1.2 if thermal power is

< 30%** of RATED THERMAL POWER r the recirculation loop flow In the operating loop is <-

of rated loop flow. l N 30%

2. The provisions of Specification 3.0.4 are not applicable.

3. Otherwise, place the unit in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With no reactor coolant system recirculation loops in operation, immediately initiate action to reduce THERMAL POWER so that it is in the unrestricted zone of Figure 3.4.1.1-1 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and initiate measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

c. With one or two reactor coolant system recirculatio loops in operation and total core flow less than 45% but greater than 3 % _of rated core flow l and THERMAL POWER within the restricted zone of igur'e 3.4.1.1-1, and with' the APRM or LPRMt neutron flux noise levels greater than three times their established baseline noise levels, immediately initiate corrective action l to restore the noise levels to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow or by reducing THERMAL POWER. ,

d. With one or two reactor coolant recircula n loops in operation, and total core flow less than or equal to ) % ,and THERMAL POWER within the l restricted zone of Figure 3.4.1.1-1, within 15 minutes initiate corrective action to reduce THERMAL POWER to within the unrestricted f Figure 3.4.1.1-1, or increase core flow to greater than ithin , l 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

This value represents the design volumetric recirculation loop flow whien ,

oduces 100% core flow at 100% THERMAL POWER. /The actual value to bey (-

applied will be determined during the_5tartup Test Program s j Initial values. Final values to be aetermined aurino startup iesting oas sweep l

go.D[fhe threshold THtKMAL POWER and recirculation loop flow whi~ch wii the cold water from the vessel bottom head preventing stratification.

_ a-.

(Value to be established during Startuo Test Proaram. 6 Core flow with both recirculation pumps at rated speed and minimum control valve position 7 ;

1 tDetector levels A and C of one LPRM string per core octant plus detectors !

A and C of one LPRM string in the center of the core should be monitored. l CLINTON - UNIT 1 3/4 4-2 l

Attachmint 3 8

REACTOR COOLANT SYSTEM [0 37 Ri! CIRCULATION LOOPS l l

SURVEILLANCE REQUIREMENTS 4.4.1.1.1 Each reactor coolant system recirculation loop flow control valve sha'Il be demonstrated OPERABLE at least once per 18 months by:

a. Verifying that the control vaive fails "as is" on loss of hydraulic ,

pressure at the hydraulic control unit, and l 1

b. Verifying that the average rate of control valve movement is:

1. Less than or equal to 11% of stroke per second opening, and 1

2. Less than or equal to 11% of stroke per second closing.

4.4.1.1.2 When THF.RMAL POWER it, within the restricted zone of Figure 3.4.1.1-1, and one or two pumps are in operation, establish a baseline APRM and LPRM* l neutron flux noise value within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of entering this operating region i unless baselining has previousiy been performed in the region since the last CORE ALTERATION, and

a. Determine the APRM and L?RM* noise levels at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

b. Determine the APRM and LPRM* noise levels within 30 minutes after the completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER. l 4.4.1.1.3 With one reactor system recirculation loop not in operation, at i least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that: ,

a. Reactor THERMAL POWER is 5 70% of RATED THERMAL POWER.

I

b. The recirculation flow control system is in the Local Manual ,

(Position Control) mode, si,wt p w

c. The volumetric flow rate of the operating loop is 1 C~33.000 gpm*[and l ss. s y,* 9

d. Coreflowif,greaterthanb9/%)whenTHERMALPOWERiswithintheun- l; restricted zone of Figure 3.4.1.1-1. l 4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, I within no more than 15 minutes prior to either THERMAL POWER increase or recir- l culation loop flow increase, verify that the following differential temperature i l

Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

- This value represents the design volumetric recirculation loop flow which )

produces 100% core flow at 100% THERMAL POWER.

6alue to be established durina Startuo Test Program. (tore flow with both recirculationpumpsatratedspeedandminimumcontrolvalveposition.p CLINTON - UNIT 1 3/4 4-3 l

Attachmsnt 3 to U-601118 REACTOR COOLANT SYSTEM Paga 25 of 87 RECIRCULATION LOOPS l

1 SURVEILLANCE REQUIREMENTS requirements are met if THERMAL POWER is < 30%* of RATED THERMAL POWER or the I

recirculation loop flow in the operating Toop is sgf ratgloop flow:

a. < 100*F between reactor vessel steam space coolant and bottom head Brain line coolant,

b. < 50*F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c. 5 50*F between the reactor coolant within the loop not in operation and the operating loop.

The differential temperature requirments of Specification 4.4.1.1.4.b and c do not apply when the loop not in operation is isolated from the reactor pressure vessel.

J2 81(1jtialvalues rical value: te be datermined durina Startup Testina basedl p e threshold THERMAL POWER and recirculation loop flow which will the cold water from the vessel bottom head preventing stratification.

T CLINTON - UNIT 1 3/4 4-4

Attachmsnt 3 to U-601118 REACTOR COOLANT SYSTEM Page 26 of 87 JET PUMPS LIMITING CONDITION FOR OPERATION _

3.4.1.2 All jet pumps shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.1.2 All jet pumps shall be demonstrated OPERABLE as follows:

a. EachoftheaboverequiredjetpumpsshallbedemonstratedOPERABLEkrior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by determining recirculation loop flow, total core flow and diffuser-to-lower plenum differential pressure for each jet pump and veri-fying that no two of the following conditions occur when both recircula-tion loop flows are operating at the same flow control valve position.

1. The indicated recirculation loop flow differs by more than 10% from the established flow control valve position-loop flow characteristics.

2. The indicated total core flow differs by more than 10% from the established total core flow value derived from recirculation loop flow measurements.

3. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established patterns by more than 10%.

b. During single recirculation loop operation, each of the above required jet pumps shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by veri-fying that no two of the following conditions occur:

1. The indicated recirculation loop R ow in the operating loop differs }

by more than 10% from established Tsingle recirculation flow control valve position-loop flow characteristics.

2. The indicate total core flow differs by more than 10% from the establishe

otal core flow value derived from single recirculation i loop flow measurements.

3. The indicated diffuser-to-lower plenum diff ential pressure of any individual jet pump differs from establishe

ingle recirculation l

loop patterns by more than 10%.

c. The provisions of Specification 4.0.4 are not applicable provided that this surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding 25% of RATED THERMAL POWER.

(*To be determined h ny thw stariup Lest, progra l CLINTON - UNIT 1 3/4 4-6

Attachment 3 REACTOR COOLANT SYSTEM to U-601118 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary coolant shall be limited to:

a. Less than or equal to 0.2 microcuries per gram DOSE EQUIVALENT I-131, and

b. Less than or equal to 100/E microcuries per gram.

APPLICABILITY: UPERATIONAL CONDITIONS 1, 2, 3, and 4.

ACTION:

a. In OPERATIONAL CONDITIONS 1, 2, or 3 with the specific activity of the primary coolant:

1. Greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTOOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2. Greater than 100R microcuries per gram, be in at least HOT SHUTDOWN with the main steamline isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. In OPERATIONAL CONDITIONS 1, 2, 3, or 4, with the specific activity of the primary coolant greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 or greater than 100 6 microcuries per gram, perform the sampling and analysis requirements of Item 4a of Table 4.4.5-1 until the specific activity of the primary co0lant is restored to within its limit.

c. In OPERATIONAL CONDITION 1 or 2, with:

1. THER A POWER changed by more than 15% of RATED THERMAL POWER in one hou * , or 1

2. The off gas level, at the off gas recombiner effluent, increased by more than 10,000 microcuries per second in one hour during steady state operation at release rates less than 75,000 microcuries per second, or ,

3. The off gas level, at the off gas recombiner effluent, increased by more than 15*4 in one hour during steady state operation at release rates greater than 75,000 microcuries per second, perform the sampling and analysis requirements of Item 4b of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.

(*Not applicable during the startup test program.] l CLINTON - UNIT 1 3/4 4-19

Attachment 3 )

to U4601118 l

Page 28 of 87 l

3/4.4 REACTOR COOLANT SYSTEM BASES _

3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively, MAPLHGR limits are decreased by the factor given in Specifica-tion 3.2.1, and MCPR operating limits are adjusted per Section 3/4.2.3.

l Additionally, surveillance on the volumetric flow rate of the operating recir- ;

culation loop is imposed to exclude the possibility of excessive o nternals !

vibration. surveillance on differential temperatures below ( 0 THERMAL 1 POWER or_ 0% rated recirculation loop flow is to sitigate the un thermal l 3og" stress on vessel nozzles, recirculation pump, and vessel bottom head during the :

extended operation of the single recirculation loop mode. l An inoperable jet pump is not, in itself, a sufficient reason to declare a re- l circulation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump failure can be detected by monitoring jet pump performance on a pre-scribed schedule for significant degradation. Recirculation loop flow mismatch limits are in compliance with ECCS LOCA analysis design criteria for two recir-culation loop operation. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other prior to startup of an idle loop. The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Sudden equilization of a tempera-ture difference > 100*F between the reactor vessel bottoa head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

The objective of GE BWR plant and fuel design is to provide stable operation with margin over the normal operating domain. However, at the high power /lo,< ,

flow corner of the operating domain, a small probability of neutron flux limit cycle oscillations exists depending on combinations of operating conditions 1 (e.g., rod pattern, power shape). To provide assurance that neutron flux limit !

cycle oscillations are detected and suppressed, APRM and LPRM neutron flux -

noise levels should be monitored while operating in this region. l (f AHadalues. Finalv41ues tn be determined during Startuo Testino based onb g e threshold THERMAL POWER and recirculation loop flow which will sweep the T cold water from the vessel bottom head preventing stratification.

CLINTON - UNIT 1 B 3/4 4-1

Attachmsnt 3 to U-601118 Pags 29 of 87 PACKAGE NUMBER 4 Description and Justification of Proposed Change This proposed change is an administrative type of change to add clarifying information to the existing CPS instrumentation Technical Specifications. Two changes are actually proposed, but each affects common instrumentation while both provide additional detail to clarify existing requirements.

The first proposed change is requested to clarify the divisional assignment for the reactor vessel water level and drywell pressure instruments listed on Technical Specification Table 3.3.2-1, items 1.c and 1.f, and Table 3.3.3-1, items C.1.a. C.1.b and C.1.c. This is consistent with other instrumentation for which the divisional assignment was specified (items 1.b and 1.e of Table 3.3.2-1)* and helps to emphasize that some of the instrumentation for High Pressure Core Spray System (RPCS) is associated with two divisions. The proposed change is indicated on the attached marked-up pages and is further explained below.

The channels or trip functions affected by the proposed change provide an isolation trip and they also provide an Emergency Core Cooling System (ECCS) actuation trip. These particular channels are associated with the actuation of HPCS and provide for automatic isolation of the associated test return line. Although HPCS is the Division III ECCS, the channels providing the automatic trip function are located within the Division III and Division IV Nuclear Systems Protection System (NSPS) cabinets or circuitry. Currently, the phrase, "(ECCS Division III)," is inserted adjacent to the functional description specified for the applicable channels in Table 3.3.2-1. It has been determined that specifying "(HPCS - NSPS Div. III and IV)" is more accurate as it indicates that some of these channels are in (NSPS) Division III and some are in (NSPS) Division IV even though HPCS is generally considered to be a Division III ECCS. A similar change is proposed for Table 3.3.1-1 by adding note (g) as shown.

It should be noted that Division IV instruments were taken into account when the Technical Specifications were developed and are a part of the Minimum Operable Channels Per Trip System requirement. They have always been taken into account in the Clinton surveillance procedures. The proposed change, therefore, does not reflect a new requirement and does not change the intent of the affected Technical Specifications or the manner in which they have always been implemented.

The affected instrumentation performs an isolation function upon a trip condition and so is appropriately listed in the Primary and Secondary Containment Isolation section of the noted Table (Table 3.3.2-1 CRVICS Instrumentation). Because several "Reactor Vessel Water Level-Low Low, Level 2" and "Drywell Pressure-High" channels or trip functions are listed in this section of the Table, clarifying information was provided to distinguish, for example, one "Drywell Pressure-High" trip function from another.

. . i Attachmsnt 3 I to U-601118 I Page 30 of 87 The second proposed change would provide clarification of the combinational logic scheme for the Reactor Vessel Water Level-Low, Low l Level 2 and Dyrwell Pressure-High channels. (Table 3.3.2-1, items 1.c j and 1.f, and Table 3.3.3-1, items C.1.a and C.I.b.) This enhancement is requested to ensure that the corresponding ACTION statements are l properly followed if one or more of the affected channels is declared l J

inope rable .

The logic scheme is unique in that although the protective action (trip) is effected upon either a high drywell pressure or' low reactor water level condition, the combinational logic is different than other one-out-of-two-twice logic schemes. The logic scheme is shown on attached Figure 7.3-8 from the CPS FSAR. Furthermore, for the isolation function, no outboard or inboard trip system configuration exists because. the protective action only closes the single test return line valve for HPCS within ECCS Division III. Clarification is therefore desired to specify how the term "trip system" applies to this configuration with respect to the ACTION statements presently specified in the Technical Specifications.

IP proposes that notes with the following wording be added to Tables 3.3.2-1 and 3.3.3-1 of the CPS Technical Specifications:

Four reactor vessel water level (drywell pressure) trip channels are logically combined in a one-out-of-two-twice configuration.

For the purposes of the associated ACTION, each one-out-of-two logic is defined as a separate trip system.

Separate notes, one for the drywell pressure channels and one for the reactor water level channels, are proposed to avoid confusing the logic with other one-out-of-two twice schemes in which drywell pressure channels and reactor water level channels are alternately combined within the one-out-of-two logics such as exists for the containment isolation logic and ECCS Division 1 and 2 actuation logic. The Minimum OPERABLE Channels Per Trip System requirement was based on the existing logic configuration. The proposed change therefore does not change the intent of the Technical Specification and is proposed only for the purpose of clarifying existing requirements.

Basis For No Significant Hazards Consideration for Change (1)

According to 10CFR50.92, a proposed change to the license (Technical Specifications) involves no significant hazards consideration if operation of the facility in accordance with the proposed change would not (1) involve a significant increase in the probability or ,

consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin 4

of safety.

l l

1

- - - .. , . . . - . - . - , . . -- - . - . - - - - - , ,--t ---.-,,--

. \

Attachm:nt 3 to U-601118 Page 31 of 87 (1) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because this change in merely administrative in nature and does not affect any accident analyses. The intent and implementation of the affected Technical Spacifications remains unchanged.

(2) The proposed change does not create the possibility of a new or different kind of accident from any previously evaluated because this change is administrative and is consistent with plant design.

(3) The nroposed change does not involve a significant reduction in a margin of safety because this is an administrative enhancement to operation that provides clarifying information consistent with plant design and the intent of the affected Technical Specification. The troposed change does not reflect any physical changes to the corresponding plant instrumentation. The proposed change therefore does not affect any margin of safety.

l 1

I l

l

TABLE 3.3.2-1 CRVICS INSTRUMENTATION

'i o MINIMUM OPERAHLE APPLICABLE z

ISOLATION CHANNELS PER TRIP OPERATIONAL CONDITION ACTION SIGNAL tt SYSTEM E TRIP FUNCTION I U 1. PRIMARY AND SECONDARY CONTAINMENT ISOLATION s*

a. Reactor Vessel Water level-Low low, 20 tevel 2 B(b)(f) 7(a) 1, 2, 3

25

b. Reactor Vessel Water level-Low Low, level 2 (ECCS Div. I and II) B ZI ") 1, 2, 3 29 Reactor Vessel Water Level-Low B 2f

N'") 1, 2, 3 29 c.

Low, Level 2-(CCCE "!-. I'!)Y

)( 2(*) 1, 2, 3 20

d. O NPr sifr - q L 1,2,3 29 Drywell Pressure - High L 2(a)

)" e.

m (ECCS Div. I and II)

Drywell Pressure - High ( HPCS -

h f.

L 2(,)(g 1,2,3 29

"!- II!f HSPS Div.m and N) 29

g. Containment Building fuel Transfer Z(b)(f) 2(*) 1, 2, 3

25 Pool Ventilation Plenum Radiation - High

h. Containment Building Exhaust Radiation - High 2(*} 1, 2, 3 29

1) Containment Bldg. HVAC (VR) and M(b)(f) 25 Drywell Purge (VQ) 1(k) 1, 2, 3 29

2) Containment Monitoring (CM) and M

25

. Process Sampling (PS) 29 oa>

S(b)(f) 7(a) 1, 2, 3

i. Containment Building Continuous Containment Purge (CCP) Lxhaust f 25 {[{ . n Radiau on - High 2

I) ] , 2, 3 29 ,

j. Reactor Vessel' Water level-Low U

25 mgn i

Low Low, level 1

1(k)(1) 1,2,3 29

k. Containment Pressure-High P

25

.

l Attachmsnt 3 l to U-601118 l Page 33 of 87 TABLE 3.3.2-1 (Continued) l CRVICS INSTRUMENTATION TABLE NOTATIONS

When handling irradiated fuel in the primary or secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

- When any turbine stop valve is greater than 95% open or the reactor mode switch is in the run position.

Main steam line isolation trip functions have 2-out-of-4 isolation logic except for the main steam line flow - high trip function which br: 29 Jt-t of-4 isolation logic for each main steam line.

1t See Specification 3.6.4 Table 3.6.4-1 for valves which are actuated by these isolation signals.

(a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped con-dition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

(b) Also actuates the standby gas treatment system.

(c) Deleted (d) Also trips and isolates the mechanical vacuum pumps.

(e) Isolates RWCU valves 1G33-F001 and 1G33-F004 only.

(f) Also actuates secondary containment ventilation isolation dampers per Table 3.6.6.2-1.

(g) Hanual Switch closes RWCU system inboard isolation valves F001, F028, F053, F040 and outboard isolation valves F004, F039, F034 and F054.

(h) Vacuum breaker isolation valves require RCIC system steam supply pressure low coincident with drywell pressure high for isolation of vacuum breaker isolation valves.

(i) A single raanual isolation switch isolates outboard steam supply line isolation valve (F064) and the RCIC pump suction from suppression pool valve (F031) only followirg a manual or automatic (Reactor Vessel Water Level 2) RCIC system initiation.

(j) Only actuates secondary containment ventilation isolation dampers per Table 3.6.6.2-1. Note ft is not applicable to this Trip Function.

(k) A channel may be place.d in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the trip con-dition provided that the redundant trip system is OPERABLE and monitoring

)

that parameter.

(1) Not required to be OPERABLE when valves IVR002A,B and IVR006A,B are sealed closed in accordance with Specification 3.6.4.

%ed AHacked 3/4 3-18 l CLINTON - UNIT 1 l

~ ,

Attachmant 3 to u-601118 TABLE 3.3.2-1 (Confinued) Page 34 of 87 Four reac4cr vessd water level + rip. channels are (m) logically combined in a one- out-o - 4wo - dwice CCH igu r akion. For 4he ptArgoses o +e associaled ACTION 3 e a ch on e - out- of - 4 W logic is de ined as a separ3de krip sysdem, (n) Four drywell pressure 4 rip channels are (ofically ecdined in a owe- ouf -f-4wo- +wice codigur ahon . For de purgoses d the associated ACTiot4 3 e ach one.- od- f- fwo logic is Jehned as 2 reparak hip system .

l i

i

TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION

'i E APPLICABLE e

MINIMUM OPERABLE CHANNELS PER TRIP OPERATIONAL c ACTION FUNCTION CONDITIONS

$ TRIP FUNCTION C. DIVISION III TRIP SYSTEM

1. HPCS SYSTEM

a. Reactor Vessel Water Level-Low, Low, Level 2 4 1, 2, 3, 4* , 5* 36 8 1,2,3 36

b. Drywell Pressure - High t 4(c)g ) 1, 2, 3, 4 * , 5* 32

c. Reactor Vessel Water Level-High, Level 8 2(d)( )

RCIC Storage Tank Level-tow 1, 2, 3, 4*, 5* 37

d. 2(d)(a) 1, 2, 3, 4 * , 5* 37 Suppression Pool Water Level-High,, 2

e. 1, 2, 3, 4*, 5* 40 1

f. HPCS Pump Discharge Pressyge-High 1, 2, 3, 4 * , 5* 40 HPCS System Flow Rate-Low 1 R*

g.

1 1, 2, 3, 4* , 5* 35

h. Manual Initiation t M MINIMUM APPLICABLE

[ TOTAL NO. CHANNELS OPERABLE OPERATIONAL OF CHANNELS TO TRIP CHANNELS CONDITIONS ACTION

0. LOSS OF POWER

1. 4.16 kV Emergency Bus Undervoltage (Loss of Voltage)

a. Divisions I & II 2/ Division 2/ Division 2/ Division 1, 2, 3, 4**, 5** 38 2 4 1, 2, 3, 4**, 5** 38

b. Division III 4

2. 4.16 kV Emergency Bus Undervoltage (Degraded Voltage) 2/ Division 2/ Division 2/0! vision 1, 2, 3, 4**, 5** 39

a. Divisions I & II 3 1, 2, 3, 4**, 5** 39

b. Division III tt 3 3 2' o R Y ?E

' 55

?OE*

Attachmsnt 3 to U-601118 Pags 36 of 87 TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during periods of required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

(b) Also actuates the associated division diesel generator.

(c) Provides signal to close HPCS pump injection valve only.

(d) Provides signal to HPCS pump suction valves only.

I i

I When the system is required to be OPERABLE per Specification 3.5.2 cr 3.5.3. )

i

- Required when ESF equipment is required to be OPERABLE.

l

Not required to be OPERABLE when reactor steam dome pressure is 1 100 psig.

1. These Trip Functions are not required for ECCS actuation. l 1

t The HPCS initiation functions of the Drywell Pressure - High and Manual Initiation are not required to be OPERABLE with indicated reactor vessel .

water level on the wide range instrument greater than the Level-8 setpoint l coincident with the reactor steam dome pressure less than 600 psig. l tt One relay with three inputs in 3 out of 3 logic. :

1 l

(e) Four reac4er vessel water level tri channels are. lo icall codined in a one-out-of-bo +wice confifaration. or 6e purposes - tk associatecl Action 3 each one.ouf-c{-bo logic is de{ined as a separaft Oip sysfem .

(f) Four de well ressure + rip channels are lo icall combirect 6 a one-out-c{-

, +wo-t e confifuration.For &c purposes the associated ACT10N sach one-3 out- ek-bo logic is dekined as a separafe trip system ,

(g) One hal k Atse Mp channels is associafed with Wuclear Systems Prratecken Sys b Ns7S) Divisten III. ; +he cher half is associated with NSPS Division TZ ,

CLINTON - UNIT 1 3/4 3-37 l

Attachssnt 3 to U-601118 Page 37 of 87 l

PACKAGE NUMBER 5 Description and Justification of Proposed Change Illinois Power (IP) is requesting a change to Clinton Power Station (CPS) Technical Specifications to remove the isolation requirements for isolating the Containment Monitoring (CM) and Process Sampling Systems (PS) upon receiving a Containment Building Exhaust High Radiation signal. This will require changes to Technical Specifications 3/4.3.2 (Table 3.3.2-1 item 1,h, lable 3.3.2-2 item 1.h, Table 3.3.2-3 item 1.h, and Table 4.3.2.1-1 icem 1.h) and 3/4.6.4 (Table 3.6.4-1) in accordance with the attached marked-up pages from the CPS Technical Specifications.

The current Technical Specifications require that the CM and PS systems automatically isolate from a Containment Building Exhaust High Radiation Signal. This trip function is required to be operable in OPERATIONAL CONDITIONS 1, 2, and 3, when handling irradiated fuel in the primary or secondary containment, during CORE ALTERATIONS, and during operations with a potential for draining the reactor vessel.

IP requests deletion of these requirements for the following reasons:

1) The CM and PS systems are designed to be operable at post-accident radiation levels. The CM and PS systems are closed loops and have small diameter (3/4 inch) containment penetrations, and they will isolate on a Loss of Coolant Accident (LOCA) signal. The LOCA isolation function will not be affected by this proposed change.

2) There are no regulatory bases or engineering justification for isolation of the CM and PS systems upon a Containment Building Ventilation Exhaust High Radiation signal. Discussionc with the Architect Engineer (Sargent and Lundy) indicate that addition of this isolation logic was an overconservative application of regulatory requirements regarding containment isolation capability.

Regarding the marked-up pages, the proposed change deletes the CM/PS isolation function as described so there would no longer be any need to distinguish the CM/PS isolation function (currently part (2) of item h]

from the VR/VQ isolation function [ currently part (1)). Therefore, "Containment Building Exhaust Radiation - High" adequately describes the isolation function since it would be understood that this isolation function only applies to the affected VR/VQ valves. That is, items (1) and (2) may be deleted because there would no longer be any need to subdivida item h. All of the remaining information lef t for item h on Tables 3.3.2-1, 3.3.2-2 and 3.3.2-3 applies to the "Containment Building Exhaust Radiation - High" description and should be brought into line with these words with respect to their vertical position on the tables.

i l

I

Attachmsnt 3' to U-601118 Page 38 of 87 Basis For No Significant Hazards Consideration t

According to 10CFR50.92, a proposed change to the license (Technical-Specifications) involves no significant hazards consideration if operation of the facility in accordance with the proposed change would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

(1) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the capability for automatic isolation of the CM and PS systems in response to a LOCA remaina unchanged. Furthermore, the CM and PS systems are designed to be operable and intended to be used under post-accident radiation conditions. Both of the systems are closed loops outside containment and have very small diameter (3/4 inch)containmentpenetrations.

(2) The proposed change does not create the possibility of a new or different kind of secident from any &ccident previously evaluated because the proposed change is applicable only to the design basis established for isolation of the ?S and CM systems with respect to design basis accidents previously evaluated. The proposed change involves no other design changes or introduces any new modes of ,

operation that would require evaluation of an accident not previously evaluated.

(3) The proposed change does not involve a significant reduction in a margin of safety since the original system design has been determined to be an overconservative application of the i requirements of NUREG 0737, section II.E.4.2. l

\

)

Attachm2nt 3 to U-601118 Pags 39 of 87 INSTRUMENTATION 3/4.3.2 CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.3.2 The containment and reactor vessel isolation control system (CRVICS) channels shown in Table 3.3.2-1 shall be OPERABLE

with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3.

APPLICABILITY: As shown in Table 3.3.2-1.

ACTION:

a. With a CRVICS channel trip setpoint less conservative than the value shown in the Allowable Value column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b. For CRVICS Main Sterm Line Isolation Trip functions:

1. With one of th.a four channels required for any Trip Function inopera-ble, operation may continue provided the inoperable channel is placed in the trippen condition within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The provisions of Speci-fication 3.0.4 are not applicable.

2. With two of the four channels required for any Trip Function inopera-ble, place one channel in the tripped condition within one hour provided no tripped channel for that Trip Function already exists.

The provisions of Specification 3.0.4 are not applicable.

3. With three of the four channels required for any Trip Function inoperable, take the ACTION required by Table 3.3.2-1.

c. For other CRVICS Isolation Trip Functions:

1. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel (s) and/or that trip system in the tripped condition ** within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The provisions of Specification 3.0.4 are not applicable. No ch%nge to +kis pagt -

proved only for centinmty-

For CRVICS Main Steam Line Isolation Trip Function, a channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance provided at least two OPERABLE channels are monitoring that parameter. .

For other CRVICS Isolation Trip Function, a channel may be placed in an in-operabin status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance provided the requirenents of Table 3.3.2-1 are fulfilled.

- An inoperable channel need not be placed in the tripped condition where this would i:ause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.

CLINTON - UNIT 1 3/4 3-11

Attachmznt 3 to U-601118 Page 40 of 87 INSTRUMENTATION CONTAIN3ENT AND REACTOR VESSEL ISCLATION CONTR01 SYSTEM LIMITING CONDITION FOR OPERATION (Continued) 3.3.2 ,'fTION(Continued):

2. With the number of OPERABLE channels less than required by the Minimu i OPERABLE Channels per Trip System requirement for both trip systems ,

place at least one trip system

in the tripped condition wi. thin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and take the ACTION required by Table 3.3.2-1.

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each CRVICS channel shall be demonstrated OPERABLE by the perfor! nance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS shall be performed at least once per 18 months. CRVICS main steam line isolation divisional logic and portions of the channel coincicent logic shall be manually tested independent of the SELF TEST SYSTEM curing each refueling outage. Each of the two trip systees er divisions of the CRVICS trip system logic shell be alternately and manually tested independent of the SELF TEST SYSTEli d ring every other refueling outa;e.

All manual testing shall be completed such that all trip functions are tested at least once every four fuel cycles.**

4.3.2.3 The CRVICS RESPONSE TIME of each CRVICS trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one logic train tested at least on:e per 35 conths, and one channel per trip fanction such that all channels are tested at least once every N times 18 m ths, where N is the total nur.ter of redundant channels in a specific CRVICS trip function.

No ebge. 4e Als pay-premde.d cdy Soc cenundh)

The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped conditio'n; if both systems have the same number of inoperable channels, place either trip system in the tripped condition.

"' Manual testing for the purpose of satisfying Specification 4.3.2.2 is not required until after shutdown during the first regularly scheduled refueling outage.

CLINTON - UNIT 1 3/4 3-12

TABLE 3.3.2-1 CRVICS INSTRUMENTATION b

'i O MINIMUM OPERABLE APPLICABLE 2

ISOLATION CHANNELS PER TRIP OPERATIONAL CONDIIION ACTION SIGNAL ti SYSTEM g TRIP FUNCTION

1. PRIMARY AND SECONDARY CONTAINMENT ISOLATION e

a. Reactor Vessel Water Level-Low low. 2(a) 1,2,3 20 Level 2 B(b)U) 25

b. Reactor Vessel Water Level-Low 2(*) 1, 2, 3 29 Low, level 2 (ECCS Div. I and II) 8 2(a) 1, 2, 3 29 B

c. Reactor Vessel Water Level-Low Low, level 2 (LCCS Div. Ill) 2(*} 1,2,3 20

d. Drywell Pressure - High L( }( }

2(a) 1, 2, 3 29 Drywell Pressure - High L

$ e.

w (ECCS Div. I and II) h f. Drywell Pressure - High (ECCS 2(,) 1,2,3 29 L

Div. III)

Certainment Building Fuel Transfer Z(D)(I) 2 I) 1, 2, 3 29

g. # 25 Pool Ventilation Plenum Radiation - llich

h. Containment Building Exhaust Radiation - High 2

__ h - --

1(k) 1,2,3 29

2) Containment Monitoring (CM) and M

- # /

. Process Sampling (PS 2(*) '1, 2, 3 29 m ,, >

i. Containment Building Continuous 5(b)(f) # 25 A o a, '

Containment Purge (CCP) Exhaust *

?%.

Radiation - High ***

2(k) 1,2,3 29 ~ S !!

j. Reactor Vessel Water level-tow U

25 ,, C ?,

Low Low, level 1 *

)II) 1, 2, 3 29 7

k. Containment Prese.ure-fliqh P 1 M 7 ',

Attachmsnt 3 to U-601118 l Page 42 of 87 TABLE 3.3.2-1 (Continued)

CRVICS INSTRUMENTATION 1

1 TABLE NOTATIONS I

f When handling irradiated fuel in the primary or secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel. l

With any ccatrol rod withdrawn. Not applicable to control rods removed per )

Specification 3.9.10.1 or 3.9.10.2.

- When any turbine stop valve is greater than 95% open or the reactor mode switch is in the run position.

t Main steam line isolation trip functions have 2-out-of-4 isolation logic except for the main steam line flow - high trip function which has 2-out-of-4 isolatioa logic for each main steam line.

tt See Specification 3.6.4 Table 3.6.4-1 for valves which are actuated by these isolation signals.

l (a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped con-dition provided at least one other OPERABLE channel in the same trip system is monitoring that pwineter.

(b) Also actuates the standby cas treatment system.

(c) Deleted i (d) Also trips and isolates the mechanical vacuum pumps. ,

(e) Isolates RWCU valves 1G33-F001 and 1G33-F004 only.

(f) Also actuates secondary containment ventilation isolation dampers per Table 3.6.6.2-1.

(g) Hanual Switch closes b'00 system inboard isolation valves F001, F028, F053, F040 and outdoard isclation valves F004, F039, F034 and F054.

(h) Vacuum breaker isolation vahes require RCIC system steam supply pressure low coincident with crywe i pressure high for isolation of vacuum breaker !

isolation valves.

(i) A single manual isolation switch isolates outboard steam supply line ,

J isolation valve (F054) and the RCIC pump suction f rom suppression pool valve (F031) only following a manual or automatic (Reactor Vessel Water Level 2) RCIC system initiation.

l (j) Only actuates secondary containment ventilation isolation dampers per Table 3.6.6.2 4. Note tt is not applicable to this Trip Function.

i (k) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the trip con-ditien provided that the redundant trip system is OPERABLE and monitoring that parameter.

(1) Not required to be OPERABLE when valves IVR002A,B and IVR006A,B are sealed closed in accordance with Specification 3.6.4.

No change.&c +his P % c.-

pc Med eMg b cc%g CLINTON - UNIT 1 3/4 3-18

,, Attachm nh.3 l to U-601118 l TABLE 3.3.2-1 (Continued)

CRVICS INSTRUMENTATION ,

ACTION ACTION 20 -

Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . <

ACTION 21 -

Deleted.

ACTION 22 - With one channel in either trip system inoperable rA tore the ,'

manual initiation function to OPERASLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months er '

be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

s ACTION 23 -

Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN wittiin '

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 24 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 25 - CORE ALTERATIONS, operations with a potential for draining the' ,

reactor vessel, and handling irradiated fuel in the primary or '

secondary containment may continue provided that SECONDARY CONTAINMENT INTEGRITY is established with the standby gas t treatment system operating within I hour.

ACTION 26 -

Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 27 - Close the affected system isolation valves within I hour and \

declare the affected system inoperable.

ACTION 28 - Lock the affected system isolation valves closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected system inoperable.

ACTION 29 - Operations may continue provided that the affected CPNICS isola- .

tion valve (s) are closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and, rs appropriate, declare the affected system or component inoperable anc follo=

any ACTIONS appropriate to Specifications of the affected system. Othe Nise, be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24; hours.

No e.hu.nSe 4e h PY-providt) e d y for cediwit3 r.

CLINTON - UNIT 1 3/4 3-19 ,

4 i

V

~

' ~

. TABIE 3.3.2-2

?

-o CRVICS INSTRUMENTATION SETPOINTS

'i z

TRIP SETPOINT A!.LOWABLE VALUE

TRIP FUNCTION ~

h 1. PRIMARY AND SECONDARY CQgAINMENT ISOLATION

[$ a. Reactor Vessel Water Level -

Low Low, level 2 1 -45.5 in.* 1 -47.7 in.

b. Reactor Vessel Water Level - Low Low, level 2 (ECCS Div. I and II) 1 -45.5 in.* 1 -47.7 in.

c. Reactor Vessel Water Level - Low Low, Level 2 (ECCS Div. III) 1 -45.5 in.* 1 -47.7 in.

Drywell Pressure - High i 1.68 psig i 1.88 psig d.

(( e. Drywell Pressure - High i 1.88 psig (ECCS Div. I and II) i 1.68 psig E5 f. Drywell Pressure - High i 1.88 psig (ECCS Div.III) i 1.68 psig

g. Containment Building Fuel Transfer Pool Ventilation Plenu=

Radiation - High i 100 mR/hr i 500 mR/hr

h. Contr.inment Building Exhaust 1 inne Drywell Purge (VQ)

[-

1 100 mR/hr 1 400 mR/hr

2) Containment Monitoring (CM) and _,

< 400 mR/hr

' Process Sampling (PS) ^- < 100 mR/hr

%- + ,L- lgg v

a, c n,

i. Containment Building Continuous y a g.

Containment Purge (CCP) Exhaust *Sg Radiation - High i 100 mR/hr i 400 mR/hr EC5 o, oo ,

j. React or Vessel Water i evel low 1-147.7 in.

Los low-tevel I 1-145.5 in.*

t ,

Attachesnt 3 to U-601118 Paga 45 of 87 TABLE 3.3.2-3

- , CRVICS INSTRUMENTATION RESPONSE TIME TPIP FUNCTION RESPONSE TIME (Seconds)

1. PRIMARY AND SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level - Low Low, level 2 NA

- , b. Reactor Vessel Water Level - Low Low, Level 2 NA (ECCS Div. I and II)

c. Reactor Vessel Water Level - Low Low, Level 2 NA

/ ' ./ . (ECCS Div. III) NA v . d. Drywell Pressure - High NA Drywell Pressure - High (ECCS Div. I and II)

- < e.

NA

f. Drywell Pressure - High (ECCS Div. III)

g. Containment Building Fuel Transfer Pool w Ventilation Plenum Radiation - High NA

- h. Containment Building Exhaust Radiation - High

- ---m

1) Containment Building HVAC (VR) and Drywell Purge (VQ) NA p

2) Containment Monitoring (CM) and Process Samplin PS - NA

i. Containmen uilding Continuous Containment

- Purge (CCP) Exhaust Radiatien - High NA

j. Reactor Vessel Water Level-Low Low Low, Level 1 NA

k. Containment Pressure - High NA Main Steam Line Radiation - High NA 1.

Fuel Building Exhaust Radiation - High NA w.

NA N nual Initiation n.

' - 2. M'AiN STEAM LINE ISOLATION

a. Reactor Vessel Water Level - Low Low Low, Level 1 < l.C*

b. Main Steam Line Radiation - High NA

c. Main Steam Line Pressure - Low < 1.0*

7 7 0.5'

. 'd. Main Steam Line Flow - High

' EA

e. Condenser Vacuum - Low

- f. Main Steam Line Tunnel Temp. - High NA

g. Main Steam Line Tunnel A Temp. - High NA Main Steam Line Turbine Bldg. Temp. - High NA

~~. b.

NA

1. Manual Initiation .

3. REACTOR WATER CLEANUP SYSTEM ISOLAT. ION '

NA i

a. A Flow - High l NA

b. A Flow Timer l NA

c. Equipment Area Temp. - High l Equipment Area a Temp. - High NA d.

e. Reactor Vessel Water Level - Low Low, Level 2 NA

f. Main Steam Line Tunnel Ambient NA l Temp. - High CLINTON - UNIT 1 3/4 3-25 l

)

TABLE 4.3.2.1-1 CRVICS INSTRUMENTATION SURVEILLANCE REQUIREMENTS

'i E CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH c- CAllBRATION SURVEILLANCE REQUIRED CHECK TEST

, TRIP FUNCTION

]-.

" PRIMARY AND SECONDARY CONTAINMENT ISOLATION 1.

a. Reactor Vessel Water levei - R ID) 1, 2, 3, #

Low Low, level 2 5 M

b. Reactor Vessel Water Level -

Low Low, Level 2 (ECCS (b)

M R- 1,2,3 Div. I and II) S

c. Reactor Vessel Water Level -

Low Low, level 2 (ECCS M R(b) 1, 2, 3 w Div. III) S Drywell Preserre - High S M R(b) 1, 2, 3 d.

[

4 e. Drywell Pressure - High (ECCS M R(b) 1, 2, 3 Div. I and II) S

f. Drywell Pressure - High R Ig 1,2,3 (ECCS Div, III) S M

g. Containment Building Fuel Transfer Pool Ventilation R 1,2,3,#

Plenum Radiation - liigh 5 M

h. Containment Building Exhaust Radiat. ion - liigh 1 Containment Building HVAC M R 1, 2, 3, #

(VR) and Drywell Purgr- (VQ) S m ,, >

2) Containme nt. Monitoring (CM) 1, 2, 3, # #a "

and Prncess Sampling (PS) S M R e o.

& & C.

i. Containment Building Contin-

S P, uous Containment Purge

, C ?,

(CCP) Exhaust Radiation -

5 M R 1,2,3,# . u High

t

Attachm2nt 3 to U-601118 Page 47 of 87 TABLE 4.3.2.1-1 (Continued)

CRVICS INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS

When handling irradiated fuel in either the secondary or the primary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

- When any turbine stop valve is greater than 95% open or the reactor mede switch is in the run position.

(a) Each train or logic channel shall be tested at least every other 31 days.

(b) Calibrate the analog trip modules at least once per 31 days.

Nochge.W AM Mt- '

providul ont 9 for Gntim h T

l I CLINTON - UNIT 1 3/4 3-32 i

t

Attachm2nt 3 to U-601118 Page 48 of 87 CONTAINMENT SYSTEMS 3/4.6.4 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.4 The containment isolation valves and the instrueentation line excess flow check valves shown in Table 3.6.4-1 shall be OPERABLE with isolation times less than or equal to those shown in Table 3.6.4-1.

APPLICABILITY: As shown in Table 3.6.4-1.

ACTION:

a. With one or more of the containment isolation valves shown in Table 3.6.4-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1. Restore the inoperable valve (s) to OPERABLE status, or

2. Isolate each affected pt.1etration by use of at least one deactivated automatic valve secured in the isolated position,*t or

3. Isolate each affected penetration by use of at least one closed manual valve or blind flange.*t The provisions of Specification 3.0.4 are not applicable provided the af fected penetration is isolated in accordance with ACTION a.2 or a,3 above, and provided the associated system, if applicable, is declared inoperable or appropriate ACTION statements for that system are performe).

Othemise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Othemise, in OPERATIONAL CONDITION **, suspend all oper;tions involving CORE ALTERATIONS, handling irradiated fuel in the seconcary containeent, or with a potential for draining the reactor vessel. 'he provisions of Specification 3.0.3 are not applicable.

t4o chuge. 4o ths Me-provided enh kr cnitnej "Isolation valves closed to satisfy these requirerents may be reopened on an intermittent basis under administrative controlt. j

- When handling irradiated fuel in the secondary containment and d'uring CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

tContainment Isolation Valves can have dual functions in that they previde both containment isolation and Emergency Core Cooling functions. Any inoperable dual function valve could degrr.de the valves' other function.

CLINTON - UNIT 1 3/4 6-29 4

Attachesnt 3 to U-601118 Pago 49 of 87 CONTAINMENT SYSTEMS CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION (Continued) 3.6.4 ACTION (Continued):

b. With one or more of the instrumentation line excess flow check valves shown in Table 3.6.4-1 inoperable, operation may continue provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1. The inoperable valve is returned to OPERABLE status, or

2. The instrument line is isolated and the associated instrument is declared inoperable.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.4.1 Each isolation valve shown in Table 3.6.4-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair, or replacement work is performed on the valve or its associated actuator, control, or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.

4.6.4.2 Each automatic isolation valve shown in Table 3.6.4-1 shall be demon-strated OPERABLE during COLD SHUTDOWN or REFUELING at least once per 13 months by verifying that on an isolation test signal each automatic isolation valve actuates to its isolation position.

4.6.4.3 The isolation time of each power operated or automatic valve shown in Table 3.6.4-1 shall be determined to be within its limit when tested purs; ant to Specification 4.0.5.

4.6.4.4 Each instrumentation line excess flow check valve shown in Table 3.6.4-1 shall be demonstrated OPERABLE at least once per 18 months y verifying that the valve actuates within the differential pressure range provided.

% ge +c *n pab' ~

provWto only be co@%M)

I CLINTON - UNIT 1 3/4 6-30

~

TABLE 3.6.4-1 (Continued) n CONTAINMENT ISOLATION VALVES C

'i MAXIMUM SECONDARY E APPLICABLE ISOLATION CONTAINMENT TEST s

ISOLATION OPERATIONAL TIME BYPASS PATH PRESSURE c VALVE PENETRATION (psig)*

HUMBER SIGNAlt CONDITIONS (Seconds) (YES/NO)

'_i NUMBER Automatic Isolation Valves (Continued) 1, 2, 3,# Yes 9.0 65

35) RWCU Transfer To Radwaste B, L, R 2 IWX019 2 8, L, R IWX020 1, 2, 3,# NA Yes 9.0

36) Process Sampling 68 B, L, M, R 1PS016 B, L, M, R 1P5017 M, D, L, R 1P5022 M, B, L, R t' IP5023 M, R

B, L, IP5034 B, L, M, R T IPS035 B, L, M, R 8' IP5055 M, R B, L, 1P5056 B, L, M. ,R 1PS069 1PS070 B L,(My R 1,2,3 No 9.0

37) DW/ Cont. Equip. Drain 69 B, L, R 16 1RE021 16 8 L, R 1RE022 1, 2, 3 No 9.0 70

38) DW/ Cont. Floor Drain B, L, R 16 1RF021 16 B, L, R 1RF022 1, 2, 3,# Yes 9.0

39) Hydrogen Recombiner Supply 71 B. L R 117 1HG001 1, 2, 3,# Yes 9.0

40) Hydrogen Recombiner Return 72 B,L,R 117 2"$

IHG004 5 ?!

$$l O ** 8 m r

?; e

~

TABLE 3.6.4-1 (Continued) n CONTAINMENT ISOLATION VALVES

'L 2 MAXIMUM SECONDARY APPLICABLE ISOLATION CONTAINMENT TEST

@ OPERATIONAL TIME BYPASS PATH PRESSU)

. VALVE PENETRATION ISOLATION NUMBER SIGNAlt CONDITIONS (Seconds) (YES/NO) (psig)'

c- NUMBER Automatic Isolation Valves (Continued)

48) Containment HVAC Supply 101 1, 2, 3,# Yes 9.0 B,L,M,Z,5,R 4 IVR001A B,L,M,Z,5,R 4 IVR0018 IVR002A,B P 1(9) 2(9)

, , 16 3(9) 4(9) ,#

102 1, 2, 3,# Yes 9.0

49) Containment HVAC Exhaust IVQ004A B,L,M,Z,5,R 10 m iv0004B B,L,M,Z,5,R 10 1 IVQ006A,8 P II9)'2(9) , 16 T 3(9}'4I9) ,#

"co 1, 2, 3,# Yes 9.0

50) Plant Chilled Water Supp!y 103 IW0001A L, U 44 IW0001B L, U 44 104 1, 2, 3,# Yes 9.0

51) Plant Chilled Water Return 44 IW0002A E. U IW0002B L, U 44

~3 F 52) Containment Bldg. HVAC 106 1, 2, 3,# Yes 9.0 R_ 9- IVR0078 B,L,M,Z,5,R 6 y .

IVR007A B,L,M,Z,5,R 6 107 1,2,3 No 9.0

't g- 53) DW Chilled Water Supply IVP004B L, U 74 D-# IVP005B L, U 74

" r. 8 n

(T 54) DWIVP014B Chilled Water Return 108 1,2,3 isc 9[{

= = o

> a L, U 74 IVP015B L, U 78 5 %l i: ?

O $

- n

< _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ __ _ _ _ _ mm_ , _ _ _ _ _, _ _ _ _ _ _ _ _ _ . _ _ _ _

TABLE 3.6.4-1 (Continued) n CONTAINMENT ISOLATION VALVES C MAXIMUM SECONDARY "i APPLICABLE ISOLATION CONTAINMENT TEST E ISOLATION OPERATIONAL TIME BYPASS PATH PRESSURE:

VALVE PENETRATION (psig)*

e NUMBER SIGNALt CONDITIONS (Seconds) (YES/NO) e NUMBER

] Automatic Isalation Valves (Continued) 1,2,3 No 9.0

55) OW Chilled Water Supply 109 L, U 74 IVP004A 74 IVP005A L, U 1, 2, 3 No 9.0 110

56) OW Chilled Water Return L, U 74 IVP014A 74 IVP015A L, U 1, 2, 3,# Yes 9.0

57) Cont. Bldg. HVAC 113 R IVR006A B, L, M, 2, 5, R 6 IVR0068 8, L, M, 2, 5, R 6

? 1,2,3 NA No 9.0 3 58) Cont. Monit. 153 8, L, M, R 1CM022

8. L, M, R ICM023 ICM025 8 L, M, R B, L, M, R ICM026 166 1, 2, 3,# Yes 9.0 ,

59) Hydrogen Recombiner Supply 117 8, L. R 1HG005 169 1, 2, 3 NA No 9.0

60) Containment HVAC 8,L,M,2,5,R IVR035 IVR036 B,L,M,Z,5,R IVR040 ,

8 L,M,Z,5,R IVR041 8 L,M,Z,5,R 4

173 1,2,3 NA No 9.0

61) Cont. Monit. B, L, M, R 1CM048 m ,, >

1CM047 8 L, M, R "

M #

ICM011 8,

8, LL,

, h, R R *?%

ICM012 8 if hb e -

___.-________--_____-__-i

TABLE 3.6.4-1 (Continued) n CONTAINMENT ISOLATION VALVES

'i SECONDARY MAXIMUM E APPLICABLE ISOLATION CONTAINMENT TEST a

ISOLATION OPERATIONAL TIME BYPASS PATH PRESSURE c VALVE PENETRATION (psig)*

SIGNAlt CONDITIONS (Seconds) (YES/NO)

{

NUMBER NUMBER Automatic Isolation Valves (Continued) 1, 2, 3,# Yes 9.0

62) Instrument Air Bottles 206 L, 8, R 19 11A013B 1, 2, 3,# NA Yes 9.0

63) Process Sampling 210 1P5038 B L, M R 8, L, M, R 1P5037 8, L, M, R IPSO 48 B, L, M, R IP5047 M, R B, L, R*

1PS004 8, L, M, R 195005 8, L, M, R

? 1P5010 B, L, M, R

$ IPS009 8, L, M, R IPS031 8 L,g R 1P5032 OO

%c"

' $bh oC$

t $"*

l

?

Attcchz:nt 3 to U-601118 Page 54 of 87 PACKAGE NUMBER 6 Description and Justification of Proposed Change This proposed change affects Technical Specification 3.3.2 (Table 3.3.2-1, item 1.g and Table 4.3.2.1-1 item 1.g) which requires the Containment Building Fuel Transfer Pool Ventilation Plenum Radiation Monitors (1RIX-PR008A-D) to be operable when handling irradiated fuel in primary or secondary containment, during CORE ALTERATIONS and operations with a potential for draining the reaesor vessel (note "f"), and during OPERATIONAL CONDITIONS 1, 2, and 3 (Power Operation, Startup and ilot Standby, respectively). Illinois Power requests that this Technical Specification be changed to be consistent with the intended purpose of these monitors by requiring operability when handling irradiated fuel in the primary containment (building) during CORE ALTERATIONS and operations with the potential for draining the reactor vessel. ACTION 29 associated with OPERATIONAL CONDITIONS 1, 2, and 3 should also be deleted.

Illinois Power proposes adding an additional note (if) to Tables 3.3.2-1 and 4.3.2.1-1 to replace the current "f" note in the respective "APPLICABLE OPERATIONAL CONDITIONS" and "0PERATIONAL CONDITIONS IN WHICH SURVEILLANCE REQUIRED" columns for the Containment Building Fuel Transfer Pool Ventilation Plenum Radiation Monitors. The note reads as follows:

f f When handling irradiated fuel in the primary containment (building) and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

As required by NUkEG-0800 (The Standard Review Plan) Subsection 15.4.7.III.4, process radiation monitors 1RIX-PR008A-D are provided only to mitigate the consequencas of a fuel handling accident inside the primary containment by providing prompt radiation detection and automatic containment isolation capability. These monitors are located as far upstream in the HVAC exhaust system as practical in order to ensure adequate response for prevention of a puff release of radioactivity due to a opent fuel accident from leaving the containment.

Two other process radiation monitors (Containment Building Exhaust and Containment Building Continuous Containment Purge Exhaust) are downstream of 1RIX-PR008A-D and monitor the entire containment exhaust air flow. Thus, the function of monitoring containment exhaust air under all normal operating conditions is adequately performed by two fully redundant safety-grade radiation monitoring systems. Therefore, IRIX-PR008A-b are not required for the purpose of monitoring containment exhaust air during conditions other than when fuel handling is in progress. The possibility of having a fuel handing accident inside the containment is directly related to fuel handling operations but has no direct relationship to plant operating conditions. Regardless, FSAR section 9.1.2.1.2 states that the upper fuel storage pool will contain no fuel during power operations; therefore, no fuel handling can take place inside the primary containment during OPERATING CONDITIONS 1, 2, or 3. CPS Procedure 3001.01, APPROACH TO CRITICALITY, step 6.5, ensures that this co=mitment is fulfilled.

l Attechn:nt 3 ;

to U-601118 l Pcge 55 of 07 l i

The proposed "##" note is a revision of the original "f" note in which the words "or secondary containment" were deleted and "(building)" was inserted. This was done because the affected monitors, i.e., the Containment Building Fuel Transfer Pool Ventilation Plenum radiation monitors, which are located inside the containment building, would not respond or provide the protective action in response to a fuel handling accident inside secondary containment (i.e. , the fuel building) . They, therefore, should not be required to be OPERA 3LE'when handling irradiated fuel in the secondary containment (as opposed to the primary containment).

Maintaining these instruments operable is relatively labor intensive.

Detector calibration, and therefore containment entry, is required every six months to meet Equipment Qualification requirements. Requiring these monitors to be operable when no fuel handling is in progress in the areas monitored by these instruments takes resources away from other -

tasks.

Basis For No Significant Hazards Consideration According to 10CTR50.92, a proposed change to the license (Technical Specifications) involves no significant hazards consideration if operation of the facility in accordance with the proposed change would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reductirn in a margin of safety.

(1) The proposed change does not involve a significant incresse in the probability or consequences of an accident previou91y evaluated because the affected radiation monitor will supply the necessary containment isolatien signals following a fuel hs.ndling cecident ,

inside the containment building. Other designated radiation monitors are the primary instruments for initiscing a containment isolation during GPERATING CONDITIONS 1, 2, and 3. Operability of these instruments is not affected by this change.

(2) The proposed cht.nge does not create the posribility of a new or different kind af accident from any previously evaluated because this change involves changing the operability requirements only for the noted monicor. The monitor will be r2 quired to be operable only when it is needed to perform its deoign function consistent with the accident (fuel handling accider.t inside the primary containment building) for which its rerponse is required. No other changes in piant operation are required. This change will make the Technical Specifications reflect the actual plant design and will ,

not change the plant's physical configuration.

(3) The propos2d change does not involve a significant reduction in the margin of safety because while in OPERATING CONDITIONS 1, 2 and 3 !

sufficiene redundant process radiation instrumentation is available to initis,te a containment isolation when a high radiation condition exists. The proposed change does not impact the operability or trip setpoints for these instruments.

Attachm:nt 3 l to U-601118 Pego 56 of 87 INSTRUMENTATION CONTAINMENT AND etEACTOR VESSEL ISOLATION CONTROL SYSTEM )

I LIMITING CONDITION FOR OPERATION (Continued) 1 l

3.3.2 ACTION (Continued):

2. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system

in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and take the ACTION required by Table 3.3.2-1.

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each CRVICS channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS shall be performed at least once per 18 months. CRVICS main steam line isolation divisional logic and portions of the channel coincident logic shall be manually tested independent of the SELF TEST SYSTEM during each refueling outage. Each of the two trip systems or divisions of the CRVICS trip system logic shall be alternately and manually tested independent of the SELF TEST SYSTEM during every other refueling outage.

All manual testing shall be completed such that all trip functions are tested at least once every four fuel cycles.**

4.3.2.3 The CRVICS RESPONSE TIME of each CRVICS trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one logic train tested at least once per 36 months, and one channel per trip function such that all channels are tested at least once every N times 18 months, where N is the total number of redundant channels in a spec 1fic CRVICS trip function.

So chattgc. 4o NiS Mt -

predtd4d cdy he cMb AOJ-

The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped conditio~n; if both systems have the same number of inoperable channels, place either trip system in the tripped condition.

- Manual testing for the purpose of satisfying Specification 4.3.2.2 is not required until after shutdown during the first regularly scheduled refueling -

outage.

CLINTON - UNIT 1 3/4 3-12

_ l

TABLE 3.3.2-1 b CRVICS INSTRUMENTATION

'i o

MINIMUM DPERABLE APPLICABLE

' ISOLATION CHANNELS PER TRIP OPERATIONAL SIGNAL tt SYSTEM CONDITION ACTION E TRIP FUNCTION

1. PRIMARY AND SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level-Low Low, Level 2 B(b)(f) 2(a) 1, 2, 3 20

25

b. Reactor Vessel Water Level-Low Low, Level 2 (ECCS Div. I and II) B 2I *) 1, 2, 3 29

c. Reactor Vessel Water Level-Low B 2I *I 1,2,3 29 Low, Level 2 (ECCS Div. III)

d. Drywell Pressure - High L(b)(f) p(a) 1, 2, 3 20

$ e. Drywell Pressure - High L 2(*) 1, 2, 3 29 w (ECCS Div. I and II) h f. Drywell Pressure - High (ECCS 2(,) 1,2,3 29 Div. III) L

g. Containment Building Fuel Transfer Z ID)CI) 2(a) 3o g Pool Ventilation Plenum ## 25 Radiation - High

h. Containment Building Exhaust Radiation - High

1) Containment Bldg. HVAC (VR) and M(b)(f) 2I ") 1, 2, 3 29 Drywell Purge (VQ) # 25

2) Containment Monitoring (CM) and M 1(k) 1, 2, 3 29

. Process Sampling (PS) # 25

i. Containment Building Continuous 5(b)(f) 2(*) l', 2, 3 29 ygR

25 a a Containment Purge (CCP) Exhaust c Radiation - High , a g.

j. Reactor Vessel Water Level-Low Low Low, Level 1 U 2(k) 1, 2, 3 29 25

[$!

ga Containment Pressure-High 1(k)(1) 1, 2, 3 29 0

k. P

25 l __- - __.-- _ -_ _ __. _ _ _ _ _

I Attachm10t 3 to U-601118 Page 58 of 87 TABLE 3.3.2-1 (Continued)

CRVICS INSTRUMENTATION TABLE NOTATIONS f When handling irradiated fuel in the primary or secondary containment and during CORE ALTERATIONS and operations with a potential for draining the n# [n'e's*k *5 5 W d ' l

With any control rod withdrawn. Not applicable to control rods remcved per Specification 3.9.10.1 or 3.9.10.2.

- When any turbine stop valve is greater than 95% open or the reactor mode switch is in the run position.

1 Main steam line isolation trip functions have 2-out-of-4 isolation logic except for the main steam line flow - high trip function which has 2-out-of-4 isolation logic for each main steam line, tt See Specification 3.6.4 Table 3.6.4-1 for valves which are actuated by tnese isolation signals.

(a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped con-dition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

(b) Also actuates the standby gas treatment system.

(c) Deleted (d) Also trips and isolates the mechanical vacuum pumps.

(e) Isolates RWCU valves 1G33-F001 and IG33-F004 only.

(f) Also actuates secondary containment ventilation isolation dampers per Table 3.6.6.2-1.

(g) Hanual Switch closes RWCU system inboard isolation valves F001, F028, F053, F040 and outboard isolation valves F004, F039, F034 and F054.

(h) Vacuum breaker isolation valves require RCIC system steam supply pressure low coincident with drywell pressure high for isolation of vacuum breaker isolation valves.

(i) A singla manual isolation switch isolates outboard steam supply line isolation valve (F064) and the RCIC pump suction from suppression pool valve (F031) only following a manual or automatic (Reactor Vessel Vcter Level 2) RCIC system initiation.

(j) Only actuates secondary containment ventilation isolation danpers per Table 3.6.6.2-1. Note it is not applicable to this Trip Function.

(k) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the trip con-dition provided that the redundant trip system is OPERABLE and monitoring that parameter.

(1) Not required to be OPERABLE when valves IVR002A,B and IVR006A,B are sealed closed in accordance with Specification 3.6.4.

CLINTON - UNIT 1 3/4 3-18

Attachmsnt 3 to U-601118 Page 59 of 87

]

1 5

i O When handling irradiated fuel in fke P imary containmnf (building) and during CORE ALTERATIONS and operations w&L a pohntial pr draining +he nactor venel.

4 i

f 1

6 1

),

1 i

-- - --,, --- -, ,---,-...-.,.,-w-,_.cr,. . . .,, , . _.._ _ . , , ,- - ,,,_,ww ..w.,,w,-- , - , .-.e , - - - , - -

Attachm2nt 3 1 to U-601118 !

Pags 60 of 87 j TABLE 3.3.2-1 (Continued)

CRVICS INSTRUMENTATION ACTION ACTION 20 -

Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUT 00WN l within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l ACTION 21 -

Deleted. ,

ACTION 22 - With one channel in either trip system inoperable restore the manual initiation function to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 23 -

Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 24 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 25 -

CORE ALTERATIONS, operations with a potential for draining the reactor vessel, and handling irradiated fuel in the primary or secondary containment may continue provided that SECONDARY !

CONTAINMENT INTEGRITY is established with the standby gas I treatment system operating within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . !

ACTION 26 -

Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 27 -

Close the affected system isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and f declare the affected system inoperable.

ACTION 28 -

Lock the affected system isolation valves closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months l and declare the affected system inoperable.

ACTION 29 -

Operations may continue provided that the affected CRVICS isola-tion valve (s) are closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and, as appropriate, declare the affected system or component inoperable and follow any ACTIONS appropriate to Specifications of the affected system. Othe mise, be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

No ch% k M W-t.

prouMtA tydy he conHQ l

J CLINTON - UNIT 1 3/4 3-19

l TABLE 4.3.2.1-1 CRVICS INSTRUMENTATION SURVEILLANCE REQUIREMENTS x

CHANNEL OPERATIONAL e

CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH

< c

CHECK TEST CALIBRATION SURVEILLANCE REQUIRED TRIP FUNCTION

1. PRIMARY AND SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water level - 1,2,3,#

Low Low, Level 2 S M R(b)

b. Reactor Vessel Water Level -

Low Low, Level 2 (ECCS Div. I and II) S M R(b) 1, 2, 3

c. Reactor Vessel Water Level -

Low Low, level 2 (ECCS Div. III) S M R(b) 1, 2, 3 w

R(b) 1,2,3

{4 d.

e.

Drywell Pressure - High Drywell Pressure - High (ECCS 5 M R Ib) 1, 2, 3 Div. I and II) S M

f. Drywell Pressure - High )

S M R 1,2,3 (ECCS Div. III) 9 Containment Building fuel Transfer Pool Ventilation Plenum Radiation - liigh 5 M R -1, 2, 3,F## l

h. Containment Building Exhaust Radiation - liigh

1) Containment Building HVAC (VR) and Drywell Purge (VQ) S M P 1,2,3,#

.a ,, >

2) Containment Monitoring (CM) "

and Process Sampling (PS) S M R 1, 2, 3, # j e e :r Containment Building Contia-i.

nous Containment Purge

~SS Z E, (CCP) Exhaust Radiation - 1,2,3,# =

w lii9 h 5 M R

Attachmsnt 3 to U-601118 i Psgs 62 of 87 .

TABLE 4.3.2.1-1 (Continued)

CRVICS INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS

When handling irradiated fuel in either the secondary or the primary containment and during CORE ALTERATIONS and operations with a potential -

for draining the reactor vessel. ,

With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

1

- When any turbine stop valve is greater than 95% open or the reactor mode switch is in the run position.

i (a) Each train or logic channel shall be tested at least every other 31 days.

1 (b) Calibrate the analog trip modules at least once per 31 days.

i I

inyrt N When Lang ircadiated feel . m pmarg cent.wne.rt ,

(ksthsn$) anb duciss Core Acruxr, ohs ana gucchons wiv a potubal {er drain,q %e.

reactor vc5Se). ,

i

) i 1

l i

A 1

i l

l CLIh CN - UNIT 1 3/4 3-32 l

. o Attcchsent 3 to U-601118 Page 63 of 87 PACKAGE NUMBER 7 Description and Justification of Proposed Change This proposed change is to clarify Technical Specification 3/4.7.2 "Control Room Ventilation System" by adding an Equipment Identification Number (EIN) to Specification 4.7.2.h. This will ensure that the visual inspection (to verify integrity of the noted flexible connection) is performed on the correct fan and thereby eliminates any possibility of misinterpretation of the Specification.

It should be noued that although both the OVC04CA(B) and OVC03CA(B) fans are associated with the recirculation filter system, the requirement for visually verifying the integrity of the Recirculation Filter Housing flexible connection is applicable only to the OVC03CA(B) fan because that fan is located at the exit of the filter housing. A loss of integrity of the floxible hose connection to the OVC03CA(B) fan could allow unfiltered air to enter the habitability boundary since such air would bypass the filter system. The OVC04CA(B) fan also has a flexible connection, but because it is located upstream of the filter housing, air leaking into the system through this connection would not bypass the filter system. In-leakage through the latter connection is a component of the leakage required to be verified as required by the first part of this surveillance requirament. That is, the latter connection is alresdy included within the test scope of the test described in the first part of Specification 4.7.2.h and therefore does not need to be visually inspected like the OVCO3CA(B) f an.

Basis For No Significant Hazards Consideration According to 10CFR50.92, a proposed change to the license (Technical Specifications) involves no significant hazards consideration if operation of the facility in accordance with the proposed change would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a cargin of safety. l (1) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because this proposed change is merely a clarification of an existing Specification and does affect any accident analyses.

(2) The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated I

because no design changes or new modes of operation are affected by the proposed change. The scope of the proposed change is strictly ;

limited to clarifying an existing Specification. No new accident l scenario is created.

(3) The proposed change does not involve a significant reduction in a margin of safety because this is an administrative change clarifying the intent and scope of an existing Specification and does not affect a margin of safety. l

PLANT SYSTEMS Attachm,nt 3 to U-601118 3/,4.7.2CCNTROLROOMVENTILATIONSYSTEM Page 64 of 87 LIMITING CONDITION FOR OPERATION _-

3.7.2 Two independent Control Room Ventilation Systems shall be OPERABLE.t APPLICABILITY: All SPERATIONAL CONDITIONS and *.

ACTION:

a. In OPERATIONAL C0tIDITION 1, 2 or 3 with one Control Room Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. In OPERATIONAL CONDITION 4, 5, or *:

1. With one Control Room Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the OPERABLE system in the high radiation mode of operation.

2. With both Control Rcom Ventilation Systems inoperable, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.

c. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION *.

SURVEILLANCE REQUIREMENTS 4.7.2 Each Control Room Ventilation System shall be demonstrated OPERABLE:t

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control room air tempera-ture is less than or equal to 86*F.

b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the makeup filter system operates continuously for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters operating; and with flow through the recircula-tion charcoal adsorber for at least 15 minutes.

When irradiated fuel is being handled in the secondary containment.

tAutomatic transfer to the chlorine mode is not required when chlorine containers having a capacity of 150 pounds or less are stored 100 meters or more from the control room or its fresh air inlets, b CLMys to fins t-CLINTON - UNIT 1 3/4 7-3 h*V'd'd odl1 b Coo %]

1

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

T Attcchcont 3 PLANT SYSTEMS to U-601118 Page 65 of 87 CONTROL ROOM VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.7.2 (Continued)

c. At least once per 18 months or (1) after any structural maintenance on the makeup or recirculation HEPA filters or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the makeup or recirculation filter system by:

1. Verifying that the makeup filter system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.S.a C.S.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978*, and the system flow rate is 3000 cfm i 10%.

2. Verifying that the recirculation filter system satisfies bypass leakage testing acceptance criteria of less than 2% total bypass and uses test procedure guidance in Regulatory Positions C.S.a and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978*, and the system flow rate is 64,000 cfm i 10%.

3. Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regula-tory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978*,

meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978*, for a methyl iodide penetre. tion of less than 0.175% for makeup filter system carbon ad-sorber and 6% for recirculation filter system carbon adsorber when tested; in accordance with ASTM D3803-79 methods, with the following parameters:

Make Up Filter System l a) Bed Depth -

4 inches 1 b) Velocity -

40 fpm c) Temperature -

30*C d) Relative Humidity -

70%

l Recirculation Filter System a) Bed Depth -

2 inches b) Velocity -

80 fps (

c) Temperature -

30'C l d) Relative Humidity -

70%

4. Verifying flow rate of 3000 cfm i 10% for the makeup filter system and 64,000 cfm i 10% for the recirculation filter system during opera-tion when tested in accordance with ANSI N510-1980.

ANSI N510-1980 shall be used in place of ANSI N510-1975 as referenced in Regulatory Guide 1.52, Revision 2, March 1978.

CLINTON - UNIT 1 3/4 7-4 %dia.npb h pp-p d e d eatg erf con h ,0. g

Attechmsnt 3 PLANT SYSTEMS to U-601118 Page 66 of 87 CONTROL ROOM VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.7.2 (Continued)

d. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978*, meets the laboratory test-ing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978*, for a methyl iodide penetration of less than 0.175% for the makeup filter system carbon adsorber and 6% for the recir-culation filter system carbon adsorber when tested; in accordance with ASTM D3803-79 methods, with the following parameters:

Make Up Filter System a) Bed Depth -

4 inches b) Velocity -

40 fpm c) Temperature -

30*C d) Relative Humidity -

70%

Recirculation Filter System a) Bed Depth -

2 inches b) Velocity -

B0 fpn c) Temperature -

30'C d) Relative Humidity -

70%

e. At least once per 18 monthr, by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the makeup filter system at a flow rate of 3000 cfm i 10%.

2. Verifying that on a high chlorine actuation ** and a manual initia-tion test signal, the system automatically ** switches to the chlorine mode of operation and the dampers close within 2 seconds.

3. Verifying that the control room leak rate is limited to < 4000 cfm I i 10% at > 1/8-inch Water Gauge (W.G.) with respect to a3jacent areas. l

4. Verifying that on a smoke mode actuation test signal, the system automatically switches to the smoke mode of operation at a flow rate less than or equal to 64,000 cfm i 10%.

5. Verifying that on a high radiation actuation test signal, the system ,

automatically switches to the high radiation mode of operation and j

- i

ANSI N510-1980 shall be used in place of ANSI N510-1975 as referenced in Regulatory Guide 1.52, Revision 2, March 1978. l

- Automatic transfer to the chlorine mode is not required when chlorine containers having a capacity of 150 pounds or less are stored 100 meters from the control room or its fresh air inlets.

CLINTON - UNIT 1 3/4 7-5 S " h 45 fn -

g,a ce r secceheeq l

. s Attachmtnt'3 to U-601118 PLANT SYSTEMS Page 67 of 87 CONTROL ROOM VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued) ,

4.7.2 (Continued) the control room is maintained at a positive pressure of at least 1/8-inch W.G. relative to the outside atmosphere during system operation at a flow rate less than or equal to 3000 cfm.

6. Verifying that the makeup filters heaters dissipate 16 i 1.6 kW when tested in accordance with ANSI N510-1980.

f. After each complete or partial replacement of a HEPA filter bank in the makeup filter system, by verifying that the HEPA filter bank satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 while operating the system at a flow rate of 3000 cfm i 10%.

g. After each complete or partial replacement of a charcoal adsorber bank in the makeup or recirculation filter systems, by verifying that the char-coal adsorber bank satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% for the makeup filter system and 2% total bypass leakage for the recirculati.on filter system in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the makeup system at a flow rate of 3000 cfm i 10%

and the recirculation filter system at a flow rate of 64,000 1 10%.

h. At least once per 18 months by verifying that the air inleakage rate into the negative pressure portions of the Main Control Room Ventilation System ductwork located outside the Main Control Room habitability boundary between fan OVC04CA(B) and isolation dampers OVC03YA(8) inclusive, and fire dampers OVC042YA(E), 0VC042YB(F), 0VC042YC(G) and OVC042YO(H) to be

< 650 cfm when tested in accordance with an NRC-approved test method. In addition, visually verif the integrity of the Recirculation Filter Housing flexible connection to a, ang-ovco3C A(3). l P

l CLINTON - UNIT 1 3/4 7 6 l

l l

l i

. 1 Attcchesnt 3 to U-601118 Page 68 of 87 PACKAGE NUM3ER 8 Description and Justification of Proposed Change The purpose of this proposed change to the CPS Technical Specifications is to revise the Trip Setpoint. Allowable Value and instrument-zero point specified for the suppression pool water level instrument associated with Table 3.3.3-2, (item C.1.e) and Table 3.3.5-2 (item d).

ISee attached marked-up pages 3/4 3-41, 3-42, and 3-61 from the CPS Technical Specifications). Initial performance of the plant calibration procedure revealed that, due to an error in the installation design, the desired calibration range for the noted instrument loop did not properly correspond to the calibration range of the installed transmitters. The effect of this was that the values indicated in inches (water column) by the analog trip modula (ATM) for the Trip Setpoint and A11cvable Value did not match the values specified in inches (water column) in the Technical Specifications. This was temporarily resolved by inserting a note into the calibration procedure acknowledging the apparent discrepancies. This reaolution was allowed because the instrument would still trip at the water level (in terms of elevation) for which the trip was required to occur.

A modification was originally proposed to relocate the transmitter, but it was decided that the problem could be remedied crithout relocating the sensor by redesignating the instrument-sero point and correspondingly respecifying the associated Trip Satpoint and Allowable Value. The instrvrent would still continue to operate satisfactorily for the minimum range of water level (elevation) required. The effect of redesignating the instrument-zero point is to shift the ATM-indicated water level values by 15 inches. The Technical Specification thus needs to be changed to reflect the ATM-indicated values. That is, under.the revised Technical Specification, the ATM-indicated values for the Trip Setpoint and Allowable Value would match the values indica:ed by the Technical Specification Table.

The change only electronically affects the instrument loop indication and does not change the design basis for the trip function. That is, the actual suppression pool level (in terms of elevation) at which the trip function should occur will remain unchanged.

Basis for No Significant Hazards Consideration According to 10CFR50.92, a proposed change to the license (Technical Specifications) involves no significant hazards consideration if operation of the facility in accordance with tne proposed change would not (1) involve a significant increase in the probability or consequences of an accident previously ev.luated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

l I

I f

. s Attachsont 3 ;

to U;601118- l Page 69 of 87 l (1) The proposed change does not involve s'significant increase in the probability or consequences of an accident previously evaluated ,

because the proposed changes do not change the actual level (in terms of elevation) at which the affected instrument is designed to trip. That is, the analytical basis from which the Trip Setpoint and Allowable Value were determined remains unchanged.

(2) It has been dstermined that the proposed change does not create the .

L possibility of a new or different kind of accident from any previously evaluated because no changes to plant operatioa or ~

changes to the design function of the instrument are proposed. No consideration of new or additional accident scenarios is required. 9 (3) The proposed change does not involve a reduction in a margin of safety because the proposed change does not change the relationship .

2 between the Trip Setpoint. Allowable Value, instrument-sero . point l and analytical limit. The design basis for the Trip Setpoint is t

unchanged. The proposed change to the Technical Specification reflects a change'to the instrument loop that will more closely match the sensor range to the range indicated by the trip unit thus ,

enhancing satisfactory operation of the instrument loop and satisfactory performance of the instrument loop calibration.

' i l k

! i 1 :

r I

L 7

k i 1 1

1 1

l i

TABLE 3.3.3-2 (Continued) b EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS

'i E ALLOWABLE TRIP SETPOINT */ALUF TRIP FUNCTION C. DIVISION III TRIP SYSTEM

1. HPCS SYSTEM

a. Reactor Vessel Water Level - Low Low, level 2 > -45.5 in.* > -47.7 in.

b. Drywell Pressure - High 31.68psig 31.88psig

c. Reactor Vessel Water Level - High, Level 8 < 52.0 in.* < 54.2 in.

d. RCIC Storage Tank Level - Low i 3b in.** 5 0 in.**

i

e. Suppression Pool Water Level - High 165$n.t _- s i 3 @ r.in.t 1 120 psig

_-3

f. HPCS Pump Discharge Pressurgj- High,, 2 145 psig HPCS System Flow Rate - Low 3 625 gpo 2 500 gpa

g. NA Manual Initiation NA w h.

~:

s w D. LOS3 0F POWER E 4.16-kV Emergency Bus Undervoltage (Loss of Voltage)#

1.

a. Divisions I and II 1. 4.16-kV Basis -

28701143.5 volts 28701525 volts

2. 120-volt Basis -

8214.1 volts 82115 volts

3. < 10 sec. time delay < 10-sec. time delay

b. Division III 1. 4.16-kV Basis -

25201175 volts 2520+210, -175 volts

2. 120-volt Basis -

7215 volts 72+6, -5 volts

> 3. < 2.5 1 0.075-sec. < 3.0-sec. time delay time delay ,o >

4.16-kV Emergency Bus Undervoltage a. 4.16-kV Basis - A*

l

2. 3797135 volts ?%

(Degraded Voltage) 3797135 volts i

b. 120 volt Basis - 3$[

108.511 volt 108.511 volt o?3

c. 15-sec. 10.5 sec. 15-sec. 11.0 sec. *G""

time delay time delay $

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

/

[ Attechmsnt 3 to U-601118 Page 71 of 87 TABLE 3.3.3-2 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS TABLE NOTATIONS

See Bases Figure B 3/4 3-1.

These are inverse time delay voltage relays or instantaneous voltage relays with a time delay. The voltages shown are the maximum that will not result in a trip. For the inverse time relays, lower voltage conditions i will result in decreased trip times. i

1. These Trip Functions are not required for ECCS actuation. ' ', ?

- Instrument zero is elevation 739' 10-3/4" as1. ,

0.

t Instrument zero is elevation (j31' 5"Jes1. l r73P_' 8 "

l l

)

l 1

J i

't CLINTON - UNIT 1 3/4 3-42 i i s;

TABLE 3.3.5-2 a

l-x REACTOR CORE ISOLATION COOLING SYSTEM ACTUAT}CH INSTRuhENTATION SETPOINTS z

ALLOWABLE c FUNCTIONAi_ UNITS -

TRIP 3rTP0lNT ,

VALUE z - -

a. Reactor Vessel Water Level - Low Low, Level 2 1 -45.5 in.* 1 -47.7 ini b Reactor Vessel Water Level - High, Level 3 $ S2.0 in." $ S2.6 in.

3 0 in.**

~

c. RCIC Storage Tank Level - Low 1 3\ in.**

d. Suppression Pool Water Level - High $ 6\ in.t $ 12 in.t l

e. Manual Initiation MA NA ,

-8 '/t ~3 w ~

See Bas 3s figure B 3/4'3-1.

w _

~

J. ** Instriment reco is 739' 10-3/4" asl.

l

+ Instrument zero is]31' 5"}iiisi.

\uc r 4

" c. ;o DJ O

[$

e-

, , , _ . - - , . -. ,- , n , . . - , , - - . ,. n . , - , a _-. _ . , ,

Attachmsnt 3

-l to U-601118 L Pega 73 of 87 y .PAgAGE NUMBER 9

' Description and Justification of Proposed Change

<; ?

This proposed change request affects two parts of Facility Operating s ' License NPF-62. The first change is to delete an Operation License

- condition that is no longer applicable. License Condition 2.C.6

- 25% of RATED THERMAL POWER. Adjust t APRM channel if the absolute difference is greater than 2% of RATED THLKMAL POWER. Any APRM channel gain adjustment made in compliance with Specifi-cation 3.2.2 shall not be included in determining the absolute difference.

~~(e) This calibration shall consist of a setpoint verification of the Neutron Flux-High and the Flow Biased Simulated Thermal Power-High trip functions.~~

~~The Flow Biased Simu'ated Thermal-High trip function is verified using a calibrated flow signal.~~

~~(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system.~~

~~(g) Calibrate the analog trip module at least once per 31 days. p l~~

~~N(h)[Verifymeasuredcore(totalcoreflow)flowtobegreaterthanorequalto (established core flow at the existing loop flow control (APRM % flow). j~~

~~!"MsfRT~~

~~)elete[ (i) This calibration shall consist of verifying the 610.6 second simulated thermal power time constant.~~

~~(j) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.~~

~~(k) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.~~

~~(1) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required to be OPERABLE per Special Test Exception 3.10.1. ,~~

~~(m) The CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION shall include the turbine first stage pressure instruments.~~

~~CLINTON - UNIT 1 3/4 3-10~~

o s Attachmtnt-3 to U-601118 Page 83 of 87 PACKAGE NUMBER 11 Description and Justification of Propose'd Change Illinois Power (IP) is requesting a change to Technical Specification Table 3.3.2-2, item 2.h to increase the Trip Setpoint and Allowable Value for Turbine Building Temperature Monitors 1E31-N563A, B, C, and D.

~~The proposed change would increase the Trip Setpoint from $ 131.2*F to~~

~~$143.2*F and the Allowable Value from 6138'F to S150*F.~~

~~Instruments 1E31.-N563A, B, C, and D monitor ambient temperatures in the Turbine Building Steam Tunnel between columns K-N, rows 109-116, and ,~~

elevations 762'-0" and 796'-0". These temperature monitors initiate a mair steam line isolation when the ambient temperature exceeds the trip setpoint. (The main steam line isolation will subsequently initiate a -

~~reactor scram when the reactor mode switch is in the RUN position.)~~

According to the design basis provided by the architectural engineer (AE) for Clinton, a maximum normal operating temperature of 122*F was assumed for the setpoint calculations. The temperature increase resulting'from a steam leak equivalent to 25 spa after two minutes was determined to be 16*F. These values, together with a drift allowance of 16.8'F, were included in the instrument setpoint calculations performed in accordance with Regulatory Guide 1.105, "Instrument Setpoints." ,

~~These calculations established the current'setpoint of 131.2*F for the ,~~

~~1E31-N563A, B, C and D instruments:~~

Maximum normal operating temperature = 122*F i Temperature increase for a 25 gpm leak @ 2 min. = 16'F Allowable value = 138'F Drift Allowance = - 6.8'F i

Resultant setpoint = 131.2'F The AE noted that radiological considerations were not included in the determination of this setpoint since it was not the bounding concern in this situation and since radiation monitors are installed in the exhaust ductwork from the tunnel and in the common station vent to which the exhaust is directed. (The exhaust process radiation monitor is designed i to detect an airborne concentration of 0.1 times the Maximum Permissible {

~~Concentration (MPC) and to alarm at 1.0 MPC to ensure that 10CFR20 limits are not exceeded at the restricted area boundary).~~

Operating experience to date has shown that the ambient temperature that typically exists in the area of the subject instruments is 120-125'F when the reactor is operating at 90-100% rated thermal power. Since Clinton has not yet operated at a sustained full reactor power level during summer conditions, the peak normal oper.cing temperature in the applicable area may be higher. If a drift of - 6.8'F from the currently specified setpoint of 131.2'F is assumed, resulting in an effective l l setpoint of 124.4'F, an unwanted trip or at least a very small operating j l margin is practically ensured. !

~~l I~~

~~- - , . - -- .-, -- -,_ ,.. _ . , _ .- , _ _ _ . - . . _ = _ _ _ _ _ - , . ,. ,..,~ .m- ,_ ,~~~

~~l * * \~~

~~Attachm:nc 3 !~~

~~to U-601118 Page 84 of 87 Illinois Power has determined that increasing the current trip setpoint~~

~~, and Allowable Value by 12'F will provide an adequate operating margin. .~~

I A 12"F increase is based on a re-evaluation of the setpoint using the same methodology as before except that the new setpoint corresponds to a l steam leak equivalent to 25 gpm after 5 minutes. As noted earlier, the !

~~original setpoint was based on a 25 gpm equivalent leak for 2 minutes, f~~

The time limit for the assumed leakage rate is established through sound engineering judgment by striking a balance between the minimum time needed before a leak can be detected without causing an inadvertent or unwanted isolation and the maximum time allowed for the leakage ;

~~condition to be recognized. A leakage time of 5 minutes is still considered acceptable and consistent with this intent. ,~~

~~The resultant Allowable Value of 150*F is justified for the following l reasons:~~

(1) Potential unnecessary channel trips due to high localized ambient temperature (especially caused by a reduction in or a loss of non-safety area cooling capacity) should be reduced. This will reduce the number of cycles and challenges to the Main Steam Isolation Valves, Safety / Relief Valves, Primary Containment, and other safety systems caused by the resultant inadvertent main steam line isolations and subsequent reactor scrams.

(2) Safety related components (Leak Detection sensors, pressure and limit switches) located in the Turbine Building Steam Tunnel between columns K-N and rows 109-116, from elevation 762'-0" to 796'-0" were evaluated for the higher temperature of 150'F. The results of this evaluation were that there were no adverse effects upon the affected components or structures. Therefore, localized temperatures of 150*F cause no adverse environmental impact on safety-related systems, structures, or components.

~~(3) Radiological protection is still provided by the Process Radiation Monitoring System and associated alarm functions.~~

(4) The assumed leakage rate of 25 gal / min. (which remains unchanged) is within the capacity of the Feedwater delivery system, which is designed to provide 115.5% of the full power feedwater flow of approximately 29,300 gal / min.

(5) The main steam line turbine building temperature monitors associated with the main steam line isolation trip function are located in each of five different areas of the turbine building (four sensors per location) near the main steam lines. The proposed change is applicable to only one of these areas. No changes are proposed at this time for the setpoints associated with the instruments in the other four areas.

4

Attachmsnt 3 to U-601118 I Page 85 of 87 Basis For No Significant Hazards Consideration According to 10CFR50.92, a proposed change to the license (Technical Specifications) involves no significant hazards consideration if operation of the facility in accordance with the proposed change would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated.' (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

(1) It has been determined that the proposed change poses no significant increase in the probability or consequences of an accident previously evaluated because a) localized temperatures of 150*F will cause no adverse environmental impact on systems, structures or components necessary for safe shutdown, (b) the proposed change does not impact the probability of occurrence of a leak, (c) only one of five turbine building areas in the vicinity of the main steam lines is affected by the proposed change, and (d) the trip setpoints for the monitors in the affected area are still based on the leakage rate assumed for the original setpoint and a leakage time determined to be acceptable such that the proposed change does not constitute a significant change in leak detection capability. In addition, the proposed change should decrease the probability of inadvertent main steam line isclations and reactor scrams. Radiation detection will continue to be provided as designed.

~~(2) The proposed changa does not create the possibility of a new or different kind of accident from any accident previously evaluated~~

~

because the proposed change introduces no new modes of operation, failure modes or changes to any equipment other than the affected monitors (for which the effects on the applicable accident scenarios have been evaluated).

(3) The proposed change does not involve a significant reduction in a margin of safety because the leak detection capability provided by the affected monitors will not be significantly reduced by the proposed change to the monitor setpoints. Other temperature j monitors, as well as the associated radiation monitors, will not be affected at all, l

l

~~6 TABLE 3.3.2-2 (Continued)~~

P

~

~~CRVICS INSTRUMENTATION SETPOINTS~~

'i E ALLOWABLE VALUE TRIP SETPOINT TRIP FUNCTION h 1. PRIMARY AND SECONDARY CNTAIRENT ISOLATION (Continued) 1 2.62 psid 5 3.00 psid

~~[ k. Containment Pressure - in3h Main Steam Line Radiation - High 1 3.0 x full power background i 3.6 x full power background 1.~~

~~m. Fuel Building Exhaust Radiation -~~
~~$ 10 mR/hr 5 17 mR/hr High NA NA~~
~~n. Manual Initiation~~
~~2. MAIN STEAM LINE ISOLATION~~
~~a. Reactor Vessel Water Level - 1 -147.7 in.~~
g Low Low Low, Level 1 1 -145.5 in.*
i 3.0 x full power background 1 3.6 x full power background
~~b. Main Steam Line Radiation - High 1 849 psig 1 837 psig~~
c. Main Steam Line Pressure - Low 1 170 psid** $ 178 psid**
~~d. Main Steam Line Flow - High 1 8.5 in. Hg vacuum 1 7.6 in. Hg vacuum~~
~~e. Condenser Vacuum - Low Main Steam Line Tunnel Temp. - High 1 165"F i 176*F f.~~
~~g. Main Steam Line Tunnel 5 60*F a Temp. - High i 54.5*F REPLACE Maid Steam Line Turbine Bldg.~~
w gg , h. -<'138 F Temp. - High -< 131.2'F .o on o n n
REACHED NA NA ?"
m e :r
~~i. Manual Initiatinn *Sg O e :3~~
~~3. REACTOR WATER CLEANUP SYSTEM ISOLATION *g" i 59 gpm 1 66.1 gpm = u~~
~~a. A Flow - High "~~
~~> 45 sec. 1 47 sec.~~
~~b. A flow Timer ,~~
1
h- Mai: . % ***s Line. TearM ne.1443 .
T inge. - Mh3h g136*F A, g (, o ( 131.2*F l) lE31- N559 t ESL - M56o A, e, C, o i E 31 - N 5tel A, 6, C, D 1 Est - N%2 A . 6, C., D C 150*F
~~2) ic3i - a ns A, e, c,o sie.t*F l~~
l 4
1 1
i l
t
?oN
.o i
1 . m
_ _ - - - _ _ _ _ _ _ _ _ - _ _ _ - _ - - - _ _ - - - _ _ _ - _ - _ _ - _ . - _ _ - - - - _ - _ _ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ _ . -

U-601118, Proposed Tech Specs Re Administrative Controls

Text

Navigation menu

Search