Text

Proposed Tech Specs Re Crvics Instrumentation,Remote Monitoring Instrumentation,Containment Isolation Valves & Radiation Monitoring Instrumentation
	ML20236K708
Person / Time
Site:	Clinton
Issue date:	10/30/1987
From:	; ILLINOIS POWER CO.
To:	;
Shared Package
ML20236K690	List: ML20236K708; U-601048, Application for Amend to License NPF-62,consisting of 24 Separate Tech Spec Changes,Including Crvics Instrumentation, Remote Shutdown Monitoring Instrumentation & Containment Isolation Valves.Fee Paid;
References
	NUDOCS 8711090316
	Download: ML20236K708 (153)
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PACKAGE NUMBER 3 DESCRIPTION AND JUSTIFICATION OF CHANGE n

The attached Technica,1 Specification pages all contain l

typographical errors. The following changes are being requested:

,.i (1) On page 3/4 3-16 and 3/4 3-17, Table 3.3.2-1, remove the "(a)"

just af ter " system" in the column heading " Minimum Operable

{

' Channels per Trip System". The "(a)" that appears adjacent to the MINIMUM OPERABLE CHAMELS PER TRIP SYSTEM heading on CPS'

. Technical Specification pages 3/4.3-15 and 3/4 3-17 is a i

typographical error handed down'from the draft. Technical l

Specifications. The purpose of note."(a)" was to allow a channel (s) to 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 performing. required surveillance (s) without having i

to place.the' trip system in the. tripped cond1 tion "provided at j

least one other OPERABLE channel in the' same trip ~ system is uonitoring that parameter." Note "(a)" clearly applies for one-out-of-two twice logic, but is not appropriate for one-out-of-two logic. That is, note "(a)" would be applicable to trip channels having a "2" listed in the MINIMUM OPERABLE

. CHANNELS PER TRIP SYSTEM column, but it would not be appropriate for trip channels having an "1" listed in the column. For the latter, note "(k)" is appropriate in the one-channel-per-trip system configuration having redundant-we trip systems.

I Since note "(a)" is applicable in some ense, while note "(k)"

'l (or no ' note) is applicable. in others, the "(a)" was removed 1

(or should have been removed) from the column heading to be placed adjacent to t,he number "2" representing the number of operable channels per trip system for each applicable trip channel.

(Note "(k)" was also applied'as appropriate.)

l Leaving the "(a)" next co.the column heading was clearly'a typographical error.

(Leaving it in place is either redundant or could result 1n an incorrect application of the noted provisions depending on the trip channel logic.)

(2) On page 3/4 3-82, Table 3.3.7.4-l', Item 7 and 8 should read 1C61-R501 and IC61-Rf02 respectively. A recent site walkdown l

of the Remote ShutLcsm Panel drywell temperature instrumentation verified that IC61-R501 and IC61-R$02 have been identified as the uppet drywell temperature and lower This is drywell temperature instruments, respectively.

. opposite from what is presently specified in the CPS Techt.ical l

Specifications under Table 3.3.6.4-1.

The instrument t

equipment numbers should therefore be transposed to correct this typographical error.

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to U-601048 Page 7 of 157 (3) On p. 3/4 6-33, Table 3.6.4-1, delete the third "41" listed

[

o for Item 18 under the MAXIMUM ISOLATION TIME column.

i Currently, two "4'1"s appear in'the MAXIMUM ISOLATION TIvr column for valve 1E51-7064. Two rows or lines were needed to list all of the characters identifying the isolation signals l

applicable to valve 1E51-F064. Only one line.(per valve) is needed however to specify the maximum isolation time since only one value for the maximum isolation time is applicable to a valve. The maximum isolhtion time value (41 seconds), was j

inadvertently repeated on the line or row containing the "F,X" l!

isolation signal characters. This is a typographical error i

that was introduced when the Full Power Operating License for l

Clinton was prepared and issued.

(The typo did not exist in the low power license.) The extra "41" should therefore be deleted.

4.6.6.3.c, (4). On page 3/4 6-71,, under Survef11ance Requirement delete the heading "Make Up Filter System".

These words are

-l The not applicable to the Standby Gas Treatment System.

. phrase may have appeared because this portion of the Standby Gas Treatment System Ev pacification is similar in format to the f

corresponding section of the Control Room Ventilation System Specification (4.7.2.d).

The latter may have been used originally as an example during the preparation of the draft Technical Specifications or the phrase may have just been l

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carried-over (by mistake) from the latter to the forrer.

phrase is not applicable to Specification 4.~6.6.3.c since the Standby Gas Treatment System (unlike the Control Room Ventilation System) only operates in an exhaust mode _of operation and, by design, does not contain a make-up system.

The inclusion of the noted phrase in Specification 4.6.6.3.c l

is a typographical error; the phrase should thus be deleted.

(5) Under ACTION a.1.g (Specification 3.4.1.1) on p. 3/4 4-2.

I l

change "4.4.1.1.2 " to "4.4.1.1.4."

ACTION a.1.g under Specification 3.4.1.1 currently and It erroneously refers to Surveillance Requirement 4.4.1.1.2.

should instead refer to Surveillance Requirement 4.4.1.1.4

}

I This typographical error was made when the specifications permitting single-loop operation were incorporated into the j

F 11 Power CPS Technical Specifications. The order (and

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consequert.ly, the numbering) of the Surveillance Requirements j

was different in the original submittal than it appears now.

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The order now is appropriate, but the number of the l

Surveillance Requirement required to be performed under ACTION l

a.1.g should have been correspondingly changed.

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Page 8 of 157 It is clear the Surveillance Requirement 4.4.1.1.4 is

-associated with the noted ACTION based at least on the 4

congruity in applicability. That is, they are both applicable l

with one recirculation loop not in operation and if thermal power is less than or equal to 30% of rated thermal power or

'the recirculation loop flow in the' operating loop is less than-or equal.to 50% of rated loop flew.

Basis for No Significant Hazards Consideration According to 10CFR50.92, a proposed change to the license

]

(Technical Specifications) involves no significant hazards l

a 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 i

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 all corrections of typographical errors and do not affect any previous analyses nor do they change or affect the intent or implementation of the applicable Technical Specifications.

(2) The proposed change does not create the possibility of ai new J

or different kind of accident from any previously evaluated

.)

because the scope of these proposed changes is limited to the correction of typographical errors. None of the changes affects the plant design or operation.

(3) The proposed change'does not involve a significant reduction in a marg 1n of safety because no changes to any margin of q

safety will result from the correction of these typographical 4

errors.

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Attacheent 3 to U-601048 Page 13 of 157 CONTAINMENT SYSTEM STANDBY GAS TREATMENT SYSTEM

)

SURVEILLANCE REQUIREMENTS (Continued) 4.6.6.3 (Continued)

Verifying that the subsystem satisfies the in place penetration and 1.

bypass leakage testing acceptance criteria of less than 0.05% and uses l

the test procedure guidance in Regulatory Positions C.5.a, C.S.c and 5

C.S.d of Regulatory Guide 1.52, Revision 2, March 1978*, and the system flow rate is 4000 cfm i 10%.

Verifying, within 31 days af ter remeval, that a laboratory analysis of 2.

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 testing criteria of Regulatory Position C.6.a of Regula-tory Guide 1.52, Revision 2, March 1978*, for a methyl iodide penetra-tion of less than 0.175%; when tested in accordance with ASTM D3803-79 methods, with the following parameters:

4 inches a) Bed Depth 40 fpm b) Velocity 80*C c) Temperature 70%

d) Relative Humidity and Verifying a subsystem flow rate of 4000 cfm 10% during system 3.

operation when tested in accordance with ANSI N510-1980, Af ter 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 c.

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%;

in accordance with ASTM D3803-79 methods, with the following parameters:

u (MakeUpFilterSys 4 inches a) Bed Depth 40 fpm b) Velocity 30 C c) Temperature 70%

d) Relative Humidity

ANSI H510-1980 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 6-71

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- Ac'tachmeriti 3

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.Page.14 of 157

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3/4.4 ' REACTOR COOLANT SYSTEM 3/4.4.1' RECIRCULATION SYSTEM'

^

RECIRCULATION LOOPS

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LIMITING CONDITION FOR OPERATION' f)

Reduce the' volumetric flow rate of the operating. recirculation loop to 1 33,000 gpm*,

fg). Perform' Surveillance Requirement 4.4.1.1 thermal power is b

i

< 30%** of= RATED THERMAL POWER or the recirculation loop flow

(

In the operating loop is 1.50%** of rated loop flow.

2.

The provisions of. Specification 3.0.4 are.not applicable.

3.

Otherwise, place the unit in HOT SHUTOOWN within -12 hours.

b.

With no reactor coolant system recirculation loops in operation, immediately:

initiate action to reduce THERMAL POWEP so that it'is in the unrestricted 1'

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. hours and in HOT SHUT 00WN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .'

i With one or two reactor coolant system recirculation # oops in operation and.

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c.

total core. flow less than 45% but. greater than (39)% of rated core flow.

'l and THERMAL POWER within the restricted zone of Figure 3.4.1.1-1, and with l

the APRM or LPRMt neutron flux noise. levels' greater than three times their established baseline noise levels, immediately initiate corrective action 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 i

by increasing core. flow or'by reducing THERMAL POWER.

.l on loops in operation, and With one or two reactor coolant recirculate %, and THERMAL POWER within' the d.

total' core flow less than or' equal to (39) restricted zone of Figure 3.4.1.1-1, within 15 minutes initiate corrective i

1 action to reduce THERMAL POWER to within the unrestricted zoDe of Figure 3.4.1.1-1, or increase core flow to greater than (39)"% within 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 which i

produces 100% core flow at 100% THERMAL POWER. The actual value to be applied will be determined during the Startup Test Program.

j I

Final values to be determined during Startup Testing based

- Initial values.

' ~

upon the threshold THERMAL POWER and recirculation loop flow whi'ch will sweep the cold water from the vessel bottom head preventing stratification.

Value to be established during Startup Test Program.

(Core flow with both recirculation pumps at rated speed and minimum control valve position.)

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.

CLINTON - UNIT 1 3/4 4-2 l

l

4 to U-601048 Page 15 of 157 PACKAGE NMGER 4

Background

3 1

According to standard instrument setpoint methodology, the Determination of an instrument trip setpoint begins with the analytical limit j

established as part of a safety analysis or design basis. Instrument.,

l loop accuracy and calibration capability is then taken into account to establish an allowable value (or Technical Specification limit as referred to in the G.E. Design Specification Data Sheets). A drift allowance is accounted for.co determine the trip setpoint.

For an increasing trip function in which the trip should occur before ij the process variable or parameter increases to a certain value, a "s "

sign appears'in front of the value specified in the TRIP SETPOINT and E

ALLOWABLE VALUE columns in the Technical Specification table.

(Similarly, for a decreasing trip function, a "e" sign appears in front of. TRIP SETPOINT and ALLOWABLE VALUE.) Operation with a trip set less

. conservative than its TRIP SET?0IMT but vithin the ALLOWABLE VALUE is peruitted on the basis that the difference between the TRIP SETPOINT rnd ALLOWABLE VALUE is less than or equal to the drif t-allowance assumed for

~

the instrument.

It is-the ALLOWABLE VALUE that is specified as the Technical Specification acceptance criteria identified in the surveillance procedures for the channel calibration and channel-functional test / trip-unit calibration. The resuleing operating band is single-ended in that the upper.(or lower) limit is bounded by the ALLOWABLE VALUE (as determined by the methed described above).

I The values specified in the design documents are usually specified according to this format. That is, a Trip Setpoint and an Allowable t

Value-(or Technical Specification limit) are usually specified with inequality signs appropriately applied.

Description and Reason for Change Many of the time limits specified for the timers in the applicabla design documents are not presented according to the above-described format. Often only a minimum and maximum time-out value or an acceptabla band is specified. Scme values are typically not specified with respect to an analytical limit and engineering judgement sometimes provides a basis for a time limit.

timers are independent of the process loop in that they are Furthermore, independent of the assumptions or allowances assumed for set determination of the trip setpoint with respect to the process variable.

Timers 3

(Timers are not actuated until trip requirements are saciafied.)

to the same conditions or considerations as the are not subject The drift transmitter or sensor which are in the process loop.

allowance assumed for the trip setpoint of a process variable instrument loop is not applicable to the associated timers.

3 d

-- ----_-_----__.-__-______j

to U-601048 Page 16 of 157

{

(

Because timers must be given special consideration in placing them in a format geared'to process instrumentation, the design bases'for the j

timers listed in Technical Specification 3.3.2, Table 3.3.2-2, Items 3.b, 4.b and 4.1 were reviewed to determine if the " a " or "4 " were appropriately applied to the specified va'ues.

This review determined that the following changes'should be made to the Technica.

1 Specifications for the noted timers, j

(1) For'the Reactor Water Cleanup System Isolation Differential Flow Timer (Item 3.b of Table 3.3.2-2), delete the " e" sign for the q

TRIP SETPOINT so that only a value (45 sec.) is specified. The J.1 proposed change is consistent with the design specification which

-[ l states, "the bypass timer shall be set at 45 seconds with &n m.

allowable valys of 47 seconds." The specification and the FGAR

,i j

note that the timer is provided to override the isolation during

.sW system pump and valve surge conditions. Operation with the timer less'than 45 seconds could lead P.o inadvertent Reactor Water set

' Cleanup Isolations and would therefore be discouraged for this The safety concern, however, is that an isolation occurs reason.

to preva.nt unacceptable radioactivity releases to the environment,

[

y 47 seconds is therefore justified as a' Technical Specification limit (i.e., ALLOWABLE VALUE).

f (2) For the Reactor Core Isolution Cooling System Isolation (RCIC)

Steam Line Flow - High Timer (Itam 4.b of Table 3.3 2-2), delete

- I the ".2 " sign for the TRIP SETPOINT so that only e value (3 sec.)

is specified. Change the ALLOWABLE VALUE column such that "3+10,-0 sec." is specified. From a safety standpcint, the RCIC Uteam Line l

Flow-High trip should occur'soon enougi. so that for postulated i

system pipe breaks, the system will isolate in time to prevent i

unacceptable radioactivity releaces to the environment, fee as discussed in Appendix D (Item II.K.3.15) of the FSAR, a'J-second delay is needed to preven't spurious isolations of the RCIC systec

.l due to pressure spikes which may occur during system startup.

Because the FSAR places equal emphasis on thesa conceins and in j

view of the fact that the design specification for the instrumentation specifies a " minimum setpoint" of 3 seconds and a

" maximum setpoint"'of'13 seconds for the timer, uge of the current sign in the TRIP CETPOINT column for the lower 11m2.t together "2"

uith a " 5" sign in the ALLOJABLE VALUE column for the upper limit is inappropriate. That is, both endpoints should be considered Technical Specification 11 nits so that an acceptable band is specified in the ALLOWABLE VALUE column.

1 For the RCIC Main Steam Line Tunnel TemperatureiTimer (Item 4.1 of i

f (3)

Table 3.3.2-2), delete the "d " sign for the TRIS SETPOINT so that only a value (25 min.) remains. Change the ALLOWABLE VALUE column such that "25+3,-0 min." is specified. A;n upper limit of 28 minutec is specified per the calculation performed to determine this value, but the FSAR (Section 7.6.1.4.3.3.1) states that the main steam line tunnel anbient temperature trjip is time delayed to prevent 7,CIC system isolation that could otherwise be caused by t

J ll

y.

to U-60.1048 Page 17 of 157 leakage'from other systen piping; i.e., the delay < keeps the RCIC.

-system available for water makeup.for a period of time longer in

. case the high temperature condition is caused by leakage from another system. Thie places equal emphasis on the need to specify

~

a minimum value. Therefore, for reasons similar to item (2) above, an acceptable band should be specified in the ALLOWABLE VALUE j

.p 4,,

i

' column.

7y, h

. Basis for No Significant Hazar s Consideration d

y According to 10CFR50.92, a proposed change to the license (Technical Specifications) involves no significant hazards consideration if

. (I operation of the f acility in accordance with the proposed change would

. w (1) involve a Jignificarit increase in the probability or

,4Q./a not consequences,.af an accident previously evaluated; or (2) create the 1, '

[ d. -

possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The pieposed changes do not involve a significant increase in the.

(1) probability'or consequences of an accident previously evaluated bec3use the tfm -linit values specified for the timers remain 1

unchanged. The'cimers will continue to be tested and calibrated according to plant surveillance pucedures in the same manner as that the proposed change will place additional before, except emphasis on some of the values with respect to the Technical Specification requirements.

(2) The proposed changes do not crr.te the possibility of a new or kind of ar.icf. dent fxem any previously evaluated, because i

different nochangestoplant'//r.ign,areproposedandbecausethescopeof the proposed changas are specifically directed towards achieving Technical Specification danformance to the plant design already evaluated wirA respect tn specific accident. conditions.

t The proposed changes do not involve a significant reduction in a (3) margin of safety because they are jply changes to reconcile the 7j f, < >

format of the time limits as they appear in the Technical Specif1cacion table with the design documents and bases which k'1 specify those limits. No changes to the values of the specified 3

limits are proposed.

fherefore, the proposed changes do<not involve a significant hazards i

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4 to U-601048 Page'21 of 157 i

PACKAGE NUMBER 5 The changes proposed for Technical Specifications 3/4.3.7.1 (along with j

4.11.2.7.1 and 4.11.2.7.2), 3/4.3.7.11 and 3/4.3.7.12 as indicated on l

the attached marked-up pages may be addressed as four changes. These are summarized below. A complete description, justification and basis.

for no significant hazards consideration is provided for each of the four changes (or categories of changes) on the following pages.

I (1) The first proposed change is a group of changes (summarized below) to account and allow credit to be taken for redundancy of the j

.cotmon Central Control Termina.ls (CCTs) where process radiation monitor status and indications are provided.* This is discussed in furthcr detail on the following pages.

7 1

(a) A change is proposed to include the CCTs in the OPERABILITY

{

requirements for certain radiation monitor channels required l

to be OPERABLE by the CPS Technical Specifications.

l OPERABILITY of these channels would include communication with l

either CCT.

(b) A change to the ACTIONS is proposed, as applicable, to account for inoperability of the CCTs versus inoperability of the monitor itself that provides input to the CCTs. The ACTIONS would be revised to address inoperability at two levels:

(1) ACTION to be taken if both CCTs are inoperable with respect to the applicable radiation monitor channel, (ii) ACTION to be taken if the applicable channel (s) are inoperable for reasons other than inoperability of the CCTs.

(This. level of ACTION proposed is the same as the current ACTICNs specified in the Technical j

Specifications.)

j I

(c) The CHANNEL CHECK for the applicable radiation monitors would f

be enhanced to ensure that channel communication is l

J established to the Main Control Room-CCT or Radiation

-Protection-CCT.

(The CCTs will normally be used to perform CHANNEL CHECKS since the CCTs provide channel indication and status).

The expanded CHANNEL FUNCTIONAL TEST requirements for the (d) radiation monitors would be revised to_make the wording of the l

requirement based on the Standard Technical Specifications l

more specific and applicable to the Clinton design without changing the intent of the requirement. The changes are based on the as-built capabilities and features unique to the monitors and the CCTs used at Clinton.

t.he other One CCT (the MCR-CCT) is located in the Main Control Room; (the RP-CCT) is located in the continuously canned Radiation Protection office.

to U-601048 Page 22 of 157 l

(

(2) Under the second proposed change, the CRANNEL FUNCTIONAL TEST requirement would be revised for the Liquid Radwaste Discharge Monitor. The current requirement requires a demonstration of automatic isolation of the release pathway with the monitor controls not set in the OPERATE mode. The proposed change'would delete this specific requirement since the monitor is not designed to;effect an isolation for this specific condition.

(3) 'The-third proposed change consists of specific changes to make the.

channel / instrument descriptions for the Standby Gas Treatment System (SGTS) Exhaust Process Radiation Monitor (PRM) agree with the HVAC Exhaust PRM descriptions since they are designed and operated in a similar manner. This is.only a word change and would not change the intent of the Specification or the manner in which the. surveillance are conducted.

l (4) The fourth change consists of several changes to ACTION 72 of Table 3.3.7.1-1 (in addition to the changes proposed for the ACTION (s) as specified in Change (1)]. The changes would make the ACTION consistent with other applicable Specifications including other ACTIONS. To support those changes related to OPERABILITY (or inoperability) of the Pre-Treatment Off-Gas process radiation monitor, changes to Specifications 4.11.2.7.1 and 4.11.2.7.2 are also proposed.

Change (1)

Background

the The radiation monitoring system at Clinton is designed such that various process (and area) radiation monitors provide their information Monitor via data links to two common Central Control Terminals (CCTs).

One CCT indication and status is provided through either of the CCTs.

(MCR-CCT) is located in the main control room and the other (RP-CCT) is located in the Radiation Protection Office.

(See FSAR sections f

7.7.1.19.7 and 7.7.1.9.5.2).

The CCTs are functionally equivalent.

Because the Radiation Protection Office is continuously manned and its the RP-CCT personnel communicate frequently with the main control room, may be considered redundant and equivalent to the MCR-CCT with respect to verifying monitor status, checking monitor indications, and i

performing surveillance on the radiation monitors.

itself is a departure from earlier plant designs in that The CCT concept each CCT is a common point for providing annunciation, indication, and status of all of the various monitors providing input to the CCT(s);

i.e., the radiation monitors are not individually annunciated by dedicated Main Control Room annunciator windows.

l a

\\

to U-60104S Page 23 of 157 Description and Justification for Changes Made under Change (I)

For certain radiation monitor channels, the Technical Specifications (4. 3. 7.1, 4. 3. 7.11,- 4. 3. 7.12) require that control room alarm annunciation be demonstrated as part of the CHANNEL FUNCTIONAL TEST.*

This expanded CRANNEL FUNCTIONAL TEST requirement (as specified by

tnnotation to the CHANNEL FUI CTIONAL TEST frequencies specified in Tables 4.3.7.1-1, 4.3.7.11-1 and 4.3.7.12-1) placen additional emphasis on the capability to remotely annunciate an alarm condition of the appropriate monitor.

l If th. RP-CCT is considered equivalent to the MCR-CCT, then this requirement should not refer specifically to " control room alarm ann'unciation," and the requirement should be changed to refer to the IICR-CCT or the RP-CCT.

That'is, inoperability of one CCT does not constitute inoperability of a monitor since the redundant CCT can provide the necessary status and indications.

Since the CRANNEL FUNCTIONAL TEST requirement for the applicable

-radiation monitors places special emphasis on the capability of the monitors to previde remote alarm annunciation, OPERABILITY of a radiation monitor channel will be extended to include communication from t he monitor to either of the CCTs. This would be effected by the incorporation of-a note affixed to the MINIMUM CHANNELS OPERABLE requirements for these channels.

(See Note (b) for Table 3.3.7.1-1, rote (a) for Table 3.3.7.11-1, and Note (a) for Table 3.3.7.12-1).

A radiation monitor channel must be considered inoperable (in view of the above discussion) if both CCTs are iroperable even though the nonitor itself may still be operable and still capable of providing local alarm annunciation. Because the CCTs (either one) provide annunciation, indication and status for all of the applicable radiation uenitors, some level of ACTION must be provided for the unlikely event

-that both CCTs are completely incapabic of providing annunciation, indication and status. The ACTION should be prudent and appropriate for the number of monitors that would be affected. The current ACTIONS (ACTIONS 72 and 73 for Table 3.3.7.1-1, ACTION 111 for Table 3.3.7.11-1, and ACTION 121 for Table 3.3.7.12-1) should therefore be revised. Under the propcsed change (s), each ACTION should be split to provide ACTION with both CCTs inoperable and to provide ACTION with an applicable-nonitor channel that is otherwise inoperable (i.e., due to a problum with the detector / monitor itself). The former ACTION would ensure that the monitor itself is operable (and not in an alarm state) by checking j

local status / indication. This check would be performed at a frec.uency 4.ppropriate to the monitor function yet commensurate with the ssmpling-imd analysis that would otherwise have to be performed. Ve rifying the nonitor is still working and checking its alarm status is preferred to I

collecting and analyzing a sample at least for the period when both CCTs l

Applicable to items 3 and 4.a of Table 4.3.7.1-1; items 1.a, 2.a, 2.b and 2.c of Table 4.3.7.11-1; items 1.a. 1.b, 2.a and 2.c of Table 4.3.7.12-1.

I I

L L__ ____

l

4

. Attachment'3 I

to10-601048 L

Page 24 of 157 H

are inoperable.** If the monitor / detector itself is determined to be inoperable, then the latter ACTION.(which would be equivalent to the ACTION that currently appears in the Technical Specifications) would be I

placed into effect by performing the. sampling and analysis as' required.

Since the CCT(s) provide indication and status of a monitor, and since monitor OPERABILITY includes communication between the monitor and the CCT(s), the routine CHANNEL CHECK surveillance requirements for these

~ Specifications will be annotated to note that the CHANNEL CHECK will include determination that chantel communication is established to the i

MCR-CCT or'the RP-CCT.

The last specific change lunder Change (1) revises the CHANNEL FUNCTIONAL' l

TEST requirement (as specified and expanded by Note (1) for Table 4.3.7.1-1, Note (1) and (2) for Table 4.3.7.11-1 and Note (1) for Table 4.3.7.12-1] to cake the wording agree more closely with the capabilities and actual indication provided by the Eberline system at Clinton. The intent of.the Specification is not affected by this proposed change.

The terms specified in parenthesis on the marked-up pages are described in FSAR Section 7.7.1.9.5.2.2.

Basis for No Significant Hazards Consideration for Chance (1) i

'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 tha proposed chang: would (1) involve a significant increase in the probability or not consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The proposed change does not involve a significant increase in the (1) probability or consequences of an accident previously evaluated for the following reasons:

The proposed change does not change the intent of the affected l

(a): Technical Specifications; it brings the Standard Technical Specification format more in line'with the radiation monitoring system employed at Clinton. The revised Specifications will ensure that effluent releases from Clinton are adequately monitored. In this respect, the proposed change does not involve a significant increase in the probability or consequences of an " accident" previously evaluated.

(b) The process radiation monitors to which the proposed change applies are, as noted in FSAR Section 7.7.1.19.1.1, not required for safety. This is supported by the fact that the l

l applicable Technical Specifications all contain exceptions to Specification 3.0.3.

However, some of the monitors do provide

)

j This check would not be a substitute for performing the sampling and analysis required by 5pe~cifications 3/4.11.1 and 3/4.11.2.

l r

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co U-601048 Page 25 of 157

,j support for ensuring that the limits of 10 CFR Part 20 are not exceeded. The proposed change does not significantly reduce the capability of ensuring that these limits will not be exceeded l

' because adequate monitoring will be provided in accordance with the l

revised Specifications. The intent of the applicable. Technical Specifications including the OPERA 3ILITY requirements and associated ACTIONS remains unchanged.

(2) The proposed change does not create the possibility of a new or different kind of e; accident from any accident previously evaluated because the proposed change does not involve any design changes and does not introduce any new mode of operation such that a different basis " accident" should be considered.

(3) ~The proposed change does,not involve a significant reduction in a margin of safety because the intent of the applicable Technical Specifications and the monitoring capability of the applicable

. instrumentation is not diminished by this. change. No changes to i

the instruments' setpoints are effected by the proposed change.

The OPERABILITY requirements specified for the instrumentation in

'the original and revised specification are equivalent.

Change (2)

Description and Justification for Change (2) 1 According to plant design (as discussed in FSAR Section 11.5.2.2.5 and 11.5.2.2.6) the liquid radwaste discharge radiation monitor shuts off radwaste discharge flow when the HIGH alarm setpoint is exceeded or when monitor failure occurs. The current Technica1' Specification (Note (1) to Table 4.3.7.11-1] implies that radwaste discharge f]ow is also automatically terminated when.the monitor is taken out of the OPERATE mode. A review of the monitoring system design confirms that this is not the case.

(This condition was reported in LER 87-046.) That is,

,the condition identified by item 4 in Note (1) on p. 3/4 3-100 does not cause an automatic isolation of the release pathway to occur, although the statue of this condition is conveyed by the CCT(s). The proposed change would reflect the actual capabilities of the monitoring system with respect to the noted condition (instrument not set in OPERATE mode).

The specific discrepancy between plant design and the current Technical Specification is due to an oversight at the time the Technical Specification was developed. The basis provided for no significant hazards consideration explains why current design is adequate and no design changes should be required.

Basis for No Significant Hazards Consideration for Change (2)_

1 According to 10CFR50.92, a proposed change to the license ('

thnical Specifications) involves no significant hazards consideratic if operation of the facility in accordance wi-h the proposed change would not (1) involve a significant increase in tae probability or consequences of an accident previously evaluated; or (2) create the r

p,ssibility of a new or different kind of accident from any accident previously evaluated; or (?) involve a significant reduction in a margin of safety.

i l

to U-601048 Page 26 of 157

.(1) It has been determined that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change is consistent with plant design which has been determined to be adequate.

The CPS Technical Specifications specify the OPERATIONAL CONDITIONS for which the'radwaste discharge radiation monitor is required to be.0PERABLE. The monitor is not censidered to be OPERABLE when the j

monitor controls are not set to the OPERATE position. Plant procedures (to ensure compliance with the Technical Specifications)

~

require the monitor controls to be set to the OPERATE position during the applicable OPERATIONAL CONDITIONS (i.e., during releascs via the radwaste discharge pathway). Thus, the monitor's

.CFERABILITY is administrative 1y controlled by compliance with the Technical Specifications (Limiting Conditions for Operation.and corresponding ACTIONS) according to plant procedures.

In addition, indications of monitor status are provided at the Central Control Terminals located in the control room and in the Radiation

' Protection office.

(2) The proposed _ change does not create the possibility of a new or different kind of accident from any accident previously evaluated since no changes to plant design or operation are required for the proposed change.

(3) It has been determined that the proposed change does not involve a l

significant reductien in a margin of safety. No changes to the alarm and trip setpoints for the radwaste discharge radiation monitor are proposed, and no changes to monitor operation or monitoring capability will be effected by the proposed change.

Change (3)

Description and Justification for Change (3) l The changes proposed under this change apply only to Tables 3.3.7.12-1 and 4.3.7.12-1.

The changes will make the channel descriptions for the Standby Gas Treatment System Exhaust Process Radiation Monitor and the Station HVAC Exhaust Process Radiation Monitor agree since they are equivaler.t monitors. Under the revised Specification, both monitors

'l will have a "High-Range" and a " Low-Range" Noble' Gas Activity Monitor channel listed, and both will have an." Iodine Sampler" listed.

I Basis for No Significant Hazards Consideration for Change (3) j

}

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 l

consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident l

previously evaluated; or (3) involve a significant reduction in a margin j

J of safety.

1 i

l

t i

-l 1

to U-601048 4

Page 27 of 157 (1) The proposed change does not involve a significant increase in the' l

probability or consequences of an accident previously evaluated

.j I

because the change is only a clarification of existing requirements 4

and is proposed for the sake of achieving consistency in the text of the Technical Specification document itself. The intent of the Specification (s) remains unchanged. Surveillance testing.will continue to be' performed in the same manner for the same intended i

e instrument channels.

I (2) The proposed. change does not create the possibility of a new or

)

different kind of accident from any accident previously evaluated since no enanges to plant design or operation are effected by the proposed change.-

(3) The proposed change does nct. involve a significant reduction in a margin of safety since no changes to the design, operation or testing of the applicable monitors are involved. The scope of the proposed change is strictly linited to changes in the wording currently used to describe the applicable monitor channels as they f

l currently appear in the text of the applicable Technical Specification (s).

The NRC has provided guidance concerning the application of the standards in 10CFR50.92 by providing certain examples of actions notThis likely to involve a significant hazards consideration (51FR7751).

guidance states that "A proposed amendment io an operating license for a facility...will likely be found to involve no significant hazards considerations, if operation of the facility in accordance with the proposed amendment 1nvolves... (1) a purely administrative change to Technical Specifications. For example, a change to achieve consistency throughout ' the Technical Specifications, correction of an error, or a change in nomenclature." The proposed change falls within the envelope of this example since the change would'make the wording as noted This therefore supports the conclusion that the proposed consistent.

change does not involve a significant hazards consideration.

]

Change (4)

.i f

Description and Justification for Change (4)

I Several. changes are proposed for ACTION 72.

To support those changes associated with the OPERABILITY of the Pre-Treatment Off-Gas process radiation monitor, changes are also proposed for Technical 4.11.2.7.1 and 4.11.2.7.2 due to the association between Specifications these Specifications and Specification 3.3.7.1 (specifically ACTION 72).

l These changes are discussed in more detail below.

- ACTION 72 currently states that (with the Pre-Treatment Off-gas Noble Gas Activity Monitor inoperable) " gases from the main condenser off-gas treatment system may be released to the environment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 4

l provided:

1 l

a.

The off-gas treatment system is not bypassed, and I

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3 j

i f

4 to U-601048 c

Page 28 of 157 1

b.

The post-treatment air ejector of f-gas PRM high range noble gas activity monitor is OPERABLE."

q q

J No shutdown requirement is specified if the 72-hour limit is exceeded.

This is consistent with the ACTIONS specified in the Limiting Condition j

for Operation section which state that the provisions of Specifications 3.0.3 and 3.0.4 are not applicable. "The 3.0.3 exemption suggests that it is not the intent of the Specification (3.3.7.1) to require a plant-shutdown if the radiation nonitoring instrumentation listed in che specification ir inoperable. With no plant shutdown required, it may be concluded that the plant should be able to continue to operate under the provisions of the ACTION statement (to ensure that adequate monitoring is still provided) and that the 72-hour time limit is not applicable.

(This conclusion has been previously diacussed with NRC.)

A discrepancy is introduced, however, because extended operation with the' Pre-Treatment Off-Gas process radiation monitor inoperable would not j

t support implementation of Technical Specification Surveillance Requirements-4.11.2.7.1 and 4.11.2.7.2.

These surveillance (which l

the require monitoring the radioactivity rate of noble gases measured at off-gas recombiner effluent) assume the retreatment monitor is performing its monitoring function. To support possible periods of operation with the retreatment monitor inoperable (as provided by 1

Specification 3.3.7.1), a provision for sampling and analyzing the offgas (from the sam.e point in the system as the retreatment monitor) will slso be included in ACTION 72.

The proposed change to ACTION 72 would therefore delete the words "for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months " and insert a third provision ("c.") stating that " grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed for gross noble gas activity within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ."

(This sampling frequency and analysis time limit is consistent with other ACTION requirements appearing in the Technical Specifications.)

To ensure consistency between the OPERABILITY requirements for the retreatment monitor specified in Table 3.3.7.1-1, including the provisions of ACTION 72, and the requirements of Specification 3.11.2.7.

an additional change is proposed for the Surveillance Requirements associated with Specification 3.11.2.7.

The words "by the Pre-Treatment Off-Gas process radiation monitor required to be OPERABLE or as otherwise provided by Table 3.3.7.1-1" would be inserted into Specifications 4.11.2.7.1 and 4.11.2.7.2 as shown. This would allev the sampling and analysis provision to be an alternate means of supporting the implementation of these Specifications. The expenditure of man-hours and utilization of equipment in order to support the sampling and analysis provides incentive fer returning the retreatment monitor I

4 to OPERABLE status as soon as possible.

Finally, a clarification would be inserted into the provision under ACTION 72 which requires the post-treatment air ejector off-gas PRM high Since ACTION 73 range noble gas activity monitor to be OPECABLE.

contains provisions for collecting and analyzing grab samples in lieu of j

an OPERABLE post-treatment air ejector of f-gas PRM, the provision should be recognized in ACTION 72.

This provision is consistent with other radiological effluent Technical Specifications in which sampling and analyses are included as provi.tions within the applicable ACTIONJ l

required in lieu of OPERABLE radiation ronitoring instrumentation.

t

q ';

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to U-601048 Page 29 of 157 It should also be noted that the off-gas. system effluentHis directed to the HVAC common stack which is also bonitored-(by the instrumentation identified in. Technical Specification 3.3.7.12) and for which requirements are imposed to enoure that the limits of 10CFR'Part 20 are not exceeded.

The intent of the provisions contained in ACTION 72 is to ensure that adequate monitoring is still provided (and that the off-gas treatment i

system is not bypassed) with the Pre-treatment off-gas PRM inoperable.

l 7

The changes proposed for ACTION 72 support this intent.

li

' Basis for No Significant Hazards Consideration for Change (4)'

According to 10CFR50.92, a proposed change to the license (Technical Specifications) involves no significant hazards consideration-if j

operation of the facility in accerdance with the proposed chance vaald (1) involve a significant increase in the probability or not 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.

(l') The proposed-change does not involve a significant increase in the probability or consequences of an accident'previously evaluated because the af fected ACTION statement contains provisions to ensure that adequate monitoring is maintained. The proposed change to the ACTION statement would support these provisions. Inoperability of the monitor does not create a condition when off-gas radioactivity would be unmonitored because the ACTION, as revised, provides for sampling and, analysis to verify that the limit specified in Specification 3.11.2.7 (289 mC1/sec) :is not exceeded.*

Regarding deletion of the 72-hour time limit, the proposed change is consistent with the 3.0.3 exemption currently specified.

Deleting the 72-hour limit has no impact on plant operation since exceeding the 72-hour time limit does not require the plant to operate in any condition different than before the time limit would be' exceeded.

(2) The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because no design changes are involved and no new mode of operation is introduced that requires a new or different kind of accident to be considered.

\\

This limit ensures that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of tne is limits of 10CFR100 in the unlikely event off-gas elfluent inadvertently discharged directly to the environment without treatment.

g O

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Attachment.3-

-l to U-601048 Page'30 of 157

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(3) JThe proposed-change does not involve a significant reduction in a-

. margin of. cafety since, no changes. co1the limit specif f ed -in Specification 3.11.2.7.are proposed and because the capability to

,1

-verify that'the off+ gas radioactivity is within this~ limit is not significantly reduced.

(Sampling and analysis.is an accepted

. compensatory measure specified as a provision.in applicable ACTICN

-i statements.).- Deleting the 72-hour. time limit does not. involve a I

reduction in a margin of safety since it has been determined that 1

the.72-hour time limit does not place a meaningful time limit o'n d

-inoperability of the applicable nonitor, t

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Attacheent 3

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to U-601048

(

Page 31 of 137 i

j' INSTRUMENT ATI0t! -

i 3/4.3.7~ MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION l

i LIMITING CONDITION FOR OPERATION i

\\

3.3.7.1 'The radiation monitoring instrumentation cnannels shown in Table L

3.3.7.1-1 shall Le OPERABLE, with their alarm / trip setpoints within the J

specified limits.

t APPLICABILITY:

As shown in Table 3.3.7.1-1.

j ACTION-I With a radiation monitoring instrumentation channel alarm / trip setpoint a.

exceeding the value shown in Table 3.3.7.1-1, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

l With one or more radiation conitoring channels inoperable, take the ACTION j

b.

J required by Table 3.3.7.1-1.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

c.

)

1 3RVEILLANCEREQUIREMENTS 4.3.7.1 Each of the above rec. < ed radiation monitoring instrumentation l

channels shall be demonstrate-OPERABLE by the performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operatio for the conditions and at the frequencies shown in Table 4.3.7.1-1.

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to U-601048 TABLE 3.3.7.1-1 (Continued)

P"de 33 o f 15 7 RADIATION HONITORING INSTRUMENTATION TABLE NOTATIONS When irradiated fuel is being handled in the secondary containment.

Alarm only.

- - During oparation of the main condenser air ejector.

.With fuel in the new fuel storage vault.

f

1. With irradiated fuel in the spent fuel storage pool.

t Reactivity concentration expected at the monitor location is a noble gas j

mix with a 2.9 minute de::ay.

tt Radioactivity concentration expected at the monitor location is a noble gas mix released from the off gas treatment system.

(a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for re-quired surveillance without placing the trip system in the tripped condi-tion provided at least one other OPERABLE charnel in the same trip system is monitoring that parameter.

(b)

T.nsert crHoche.A.

ACTION ACTION 70 -

a.

With one of the required monitors inoperable, place the inoperable channel in the (downscale) tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; restore the inoperable channel to OPERABLE status within 7 days, or, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , initiate and maintain operation of the control room emergency filtration system in the recirculation mode of Operation.

b.

With both of the required monitors inoperable, initiate and maintain operation of the control room emergency filtration

' system in the recirculation mode of operation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 71 -

With the required monitor inoperable, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 72 -

[WiththenumberofchannelsOPERABLElessthanrequiredbythe Minimum Channels OPERABLE requirement, gases from the main con-denser off gas treatment system may be released to environment peg A for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided:

g The off gas treatment systcm is not bypassed, and a.

l b.

The post-treatment air ejector off gas PRM high range noble 1

(

gas activi_ty monitor is OPERABLE.

ACTION 73 -

' With tne number of channels OPHnRE less than required by t e Minimum Channels OPERABLE requirement, effluent releases via DC8 O this pathway may continue provided grab samples are.taken at

& ched least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed for gross noble gas activity l

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months y ACTION 74 -

With the number of channels OPERABLE less than required by the i

Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

CLINTON - UNIT 1 3/4 3-72

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(h Ch2nnel OPERABILlhY Shall. include fbe capability f $$$

h3 eYther Yhe - M ain Conicol Room Cenfral Confrol Termin 1

a (McR-CC r) or. the Radiatien Protection O(( ice Central Confrol'

(

Terminal- (RP-CCT) ic provide fhe alarm statusofthe app icable radiation rnonifor channel (s).

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~ ACTION 72 -

a.

w;th both ike Mc R-CCT' sad RP - CcT ino pe r a ble, using local Monidor-

)

Per{orm a CHANNEL CHECK 1.

indication' within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 4hereahfer )

and 2.

Re s4cre. fhe McR - cCT er RP-CCT fo OPERA 3LE i

aff cable channel (s) udbin de status hr da li nest 30 days, and ih unsuccessful, prepare.

a and. su bm;4 a

special Reperf pursuant to

{

Spe.cikicahon G.9.2 wdhia fbe next IO' days outlining the cause f 4he. CcT {ailure er mal { unction and the action taken to reslor inoperable equipment to OPERABLE statu s.

'o,

With the Pre.-freatment of -fas PRM - Noble 62s Achity Mon il-o r oiber wise inoperable, gases krom fhe main condenser eff-fas treatment s7.sfts be released to 4ke environment provided :

L rn ay 1.

The ch-(as treabeni sysfem is not bypassed, and i

The post-freatment air ejector ch-fas PRM hi[

2.

ranfc nob {e gas activity menifor is OPERABLE, or q

the provisions e-ACTION 73-b are in efect and 3

3.

G ra b samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analysed he fross noble fas l

acNty vi+L:n 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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CHANNEL CHECK' bsing locak on dor' p-1;

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inclination within 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s: and at 'leasi ence 8' heurs 4here afte r,

and 1

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Re 44cre. the MC R - CCT er RP - CCT - to ' OPERABLE j

licable-channek(s) wItbin Lfbe.

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next so days, and ik unsuccess{al., pregari j

and su b-rf a special Report pursuant to spec $cabn G.9.2 wRhin - +Le.next to days outlining 4he cause ch4he - CcT {aildre or.

mal {unct!on ancl the ac4 ion taken Y f

inoperabfe equipment to O P E R ABLE..sl afu s..

H b.

With the. Posf-treabent Ch -jas.PRM ' High Rnfe j

Ac41vMy Monifer c4herwise inoperable)

Noble Gas e,hlued releases via fbb gaby may cc&ue provided g b saugles are taken d least once per 8 q

and anal red he gross noble gas ac+i a.

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"*TU

1 to U-601048 Page 37 of 157 TABLE 4.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATION

When irradiated fuel is being handled in the secondary containment.

- 0uring operation of main condenser air ejector.

With fuel in the new fuel storage vault.

1. With irradiated fuel i.n the spent fuel storage pool.

tAutomatic isolation.of valve IN66-F060 shall be demonstrated during the CHANNEL CALIBRATION

_f J

[1) The CHANNEL. FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation < occurs if any of the following conditions exist; REPLACE '

1.

Instrument indicates measured levels above the alarm setpoint.

WITH 4 ATTACHED' 2.

Circuit failure.

3.

Instrument indicates a downscale failure.

t 4.

Instrument controls not set in operate mode.

The initial CHANNEL CALIBRATION shall be performed using one or more of the j

(2) reference standards certified by the National Bureau of Standards or using 1

standards that have been obtained from suppliers that participate in measure-ment assurance activities with NBS. These standards shall permit calibrat-ing the system over its intended energy range and measurement range.

Subsequent CHANNEL CALIBRATION shall be performed using the initial radio-active standards or other standards of equivalent quality or radioactive sources that have beeri related to the initial calibration.

(3) The CHANNEL. CHECX shall tiro debrmine bd channel communical established to 4he MCR-CCT or RP-CC"T.

l i

4 CLINTON - UNIT 1 3/4 3-74

m cc si y

j

-f al' Attachment.3-D to U-601048

. Page 38-of:157-

")

i

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F.i J f' FUNCTIONAL TEST.shall also !clemensirate 'fkd

' (1)

The

'.C H ANN E L MC R'- CCT or R P i CCT - responds wik snnune'iaNee and

Vent k, 3 printout to each.4 the.jollowig cdndihons 1

~ 1,.

Thsfrument.indicales measured jevels above hke abrmfMp l

(HIGHI sef o'inf, p

Deitet$r:. failure (l.ow FAILj HI FAIL).

I'

2.,

3.

Sampl~c. flow jailure.(EXTERNAL FAIL).

I4.. InstrumenI noi se.Y in norm al operahf modt ( UNI N ITI ALI E E P

.C AL. iB RAT E, MAINTENANC E ) or STA'N D5Y ).

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Attachment. 3 to U-601048 Page 39 of 157 i

INSTRUMENTATION-

-RADIOACTIVE LIOUID EFFLUENT HONITORING INSTRUMENTATION

. LIMITING CONDITION FOR' OPERATION' L

3:3.7.11 The radioactive liquid effluent monitoring-instrumentation channels shown in Table 3.3.7.11-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined and adjustea.in.

'accordance with the methodology-and parameters in the OFFSITE 00SE CALCULATION MANUAL (ODCM).

APPLICABILITY:

Ouring releases via this pathway.

ACTION:

l With a radioactive liquid efflueat monitoring instrumentation channel a,

alarm / trip setpoint less conservative than required by the above specifi-cation, without delay suspend the release of_ radioactive liquid effluents y

j monitored _ by the affected channel, or ' declare the channel inoperable, or l

change the setpoint so it is acceptably conservative.

b.

With less.than-the minimum number of radioa::tive liquid effluent monitoring instrumentation channels OPERABLE,-take the ACTION shown in Table 3.3.7.11-1.

Restore the inoperable instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

c.

SURVEILLANCE REQUIREMENTS 4.3.7.11= Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.7.11-1.

)

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At tachtaent 3 to U-601048 Pige 41' o f 15 7 TABLE 3.3.7.11-1 (Continued) 1 :%

' RADIOACTIVE-L10010 EFFLUENT MONITORING' INSTRUMENTATION 1.

L3ert-w bc.hel (a.)

ACTION

' ACTION 110 -

With.the-number of' channels OPERABLE less'than required by the-Minimum Channels OPERABLE requirement, effluent. releases may l

' continue via this pathway provided that. prior to initiating.

L a release:

At least two independent samples are analyzed in accordance I

a.

with Specification 4.11.1.1.1, and

'b.

At least -two techn.ically qual'ified members _of the Facility Staff independently verify..the release rate calculations and discharge line valving:

Otherwise,' suspend release of radioactive effluents via this.

. pathway, bh'the number of channels OPERABLE less than. required by the

' ACTION 111 I Minimum Channels OPERABLE requirement, effluent releases via..this degau w4. pathway may continue provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and-analyzed for radioactivity'at;a j

g.ffac/,gg Qimit of detection of at least 10 7 pCi/ml.

f With the number of channels OPERABLE less than required by the

' ACTION 112 -

Minimum Channels _ OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases, Pump perfor-mance ' curves generated in place may be used to estimate flow, With the number of channels OPERABLE less than required by the ACTION 113 -

Minimum Channels OPERABLE requirement, liquid additions to-this tank may continue provided the tank liquid level is estimated during all -liquid additions to the tank, J

e i

!~

CLINTON - UNIT 1 3/4 3-97 I

to U-601048 Page 42 of 157 l

i TABLE N OT AT IO N l

1 (a)

C hannel OPE R ABlLITY Shall inclucle ike capability o 1

eiher -ihe-Main ' Conirol Room Cenhal.Confrol Terminal (MCR-CCT) or the Radiahon Prohction O(( ice Central Confrol Terminal (RP-CCT) io provide Yhe alarm sfafusof!Yhe

]

appica.ble.. radiation moniYor cbannej (s).

l 1

l L

j d

ACTION lil -

a.

With both ihe MCR - CCT and RP - CCT inoperah perf orm a CHANNEL CHECK using local.monifor indic afion taif hin 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ancl at least.ence B hours f here a [-f e r.

J per b.

With the monitor otherwise inoperable ehluen#l 3

re ke2 5c5 via dis pahway

>>7 cowbnut p ro ve' des l2.

ho o u, r S )

IbY af

}e.afY ence fef~

grab sa ;les are collected and analy:ed for radioa:tivity at a j

limit of detection of at least 10 7 pCf/ml.

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'Page 45 of 157 j.

i 'i

,2 Q

l' c TABLE 4.3.'741-1 (Continued)

I m

/

h,) / ')4('-

(,pSURMILLANCEWf0TRERENf5e.[v l

i rah 10 ACTIVE LIOUID EFFLUENTrMONITORING INSTP.UMENTATION o

y wp s

q./

[k N,w-P TABLE NOTATIONS

/ /3 N&"r f

f

'f '

(1h The CHANNEL FUNC710NAL TEST shall.also demonstrate that automatic isolation

.r-of this pathway and control room alarm annunciatie occufif any'of the U

following conditions exist.

pfq[k ?t' instrument indicates measured levels above.the tiarm/ trip setp6 int.

P 2.

Circuit failure.

,t]dM

/

/.y 3.

Instrument indicates a downscale failure.

'J 4.

Instrument controls not set in cperate mode.

p (2)

The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm r

annunciation occurs if any of the following conditions exist:

1.

Instrument indicates measured levels above the alarm setpoint.

A*keplcd gif 2.

Circuit failure, j

W 3.

Instrument indicates! a downscale failure.

p A4 Q.

Instrument controls not set in oper. ate mode.

%/

The initial CHANNEL CALIBRATION shall be performed using one or more of kX (3) the reference standards certi4ied by the National Bureau of Standards or

.,6

,i using standards that have been,obtained from suppliers that participate e

in measurement'assurancrtzctivities with NB5. These standards shall r

( r; permit calibrating the gystem byer its intended range of energy and l

. g.,..

n 3-measurement range.

Frp,idheauent CiANNEL CALIBRATION, sources that have l

been related to the injuak calibration shall be used.

I if, w

f (4) CHANNEL CHECK shall coa;l,t o' verifying indication of flow durgog periods

/ <l' of release.

CHANNEL r/iECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days Q' -

,. hich continuous, pepio#c,'or batch releases are mode.

w j

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'($)zY%e CH A NN EL FUNCTIONA YEST sNIl also demonstrate thaY

^l 3

5 lau $ & ic iso ladon b biis path ant occurs $nd fat +ke McR~ ccT o

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.provides annunciafhn and event peptout in re sponse

'ar 8.P - CCT YE.' ea dk the followie[ ccnditions

o 1

1:

'. Ecyr'ument indba}te,peasured hvels deve tnt ektrm/ frip

1..

(HIGH ) ret ernt:,

p

>$ 2, L.D ef ector fai{ure [LCb FAIL.) MIFAIL),

l 3

s 3', 1S uple (Iow failure (EXTERNAL Ff!L),.

4, InstrumenI noY Set in norm al o{erahh m9dt (UNIN I Ti ALI E E D,,

1 s.

T C AllB R A T E '3 MMNTENANCE, or STA N DBY ).

l s

y 1

i-

[2)

The C H anele L ! 7 UNCTION AL YEST shall a}so demensf raf e dat Inc M d R - CCT or RP-CCT respencIs wik anrunciafion and e Vent i

prinioui Yo each of tke foNoMngcchdMs 1,

i.nstru rnent indicalce i rnetsured levels above +he alarrn

't

(

(HiGH ) set oinf.

p 2.

D ef e :f o r failure (tow PAtt, Hi FAIL),

f 3.

S u ple { low f alldre (EXTERNAL FAIL),

4.

Instrument noY scf m nera al operai-t mode (UNINITlALIEE P, C ALIBRATE, M MJTENANCE or STA N D B T ).

3 f

I demo # strat'iori ok sukemafic isolSon of +he r elease pafmay A

is not appmMe b +L:s corJ'4;cn.

ny v(

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i, to U-601048 i

Page 47 of 157 i

INSTRUMENTATION!

V

.RADI0 ACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION u

p; LIMITING CONDITION FOR OPERATION l-3.3.7.12 The radioactive gaseous effluent monitoring instrumentation channels I

shown in Table 3.3.7.12-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 and 3.11.2.6 are not

.s' exceeded. The alarm / trip setpoints of these channels shall be determined and 1

adjusted in accordance with the methodology and parameters in the 00CM.

APPLICABILITY:

As shown in Table 3.3.7.12-1, i

'l ACTION:

)

~

With a radioactive gaseous effluent monitoring instrumentation channel a.

alarm / trip setpoint less conservative than required by the above Specifi-cation, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.

With less than the minimum number of radioactive gaseous effluent monitoring b.

instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.12-1.

i Restore the'inoperabic instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.7 wny this inoperability was not corrected in a timely manner.

q

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

l c.

1

)

fuRVEILLANCEREQUIREMENTS Each radioactive gaseous effluent monitoring instrumentation channel i

4.3.7.12 shall be demonstrated OPERABLE by performance of the CHANNEL CHECX, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.7.12-1.

2 V

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Attachment.3' to-U-601048 Page 30..of 157 TABLE 3.3.7.12-1 (Continued)-

RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION' I

TABLE NOTATIONS l

At all times.

During standby gas treatment system' operation.

During operation of. the main condenser air ejector.

A channel may.be placed in an inoperable status for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for.the-purpose of performing surveillance.

b) _Lset A hkd l

ACTION ACTION 121 -

With the number of channels OPERABLE less than required by the Minimum. Channels OPERABLE requirement, effluent releases via kep l, Ace-this pathway may continue provided grab samples are taken at-least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed for gross noble gas activity, d a'k'g within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 122 -'

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue.provided that, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after

'the channel has been declared inoperable, samples are contin-uously collected with auxiliary sampling equipment as required in Table 4.11.2-1.

ACTION 123 -

With the. number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at 3 east once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 124 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of the main condenser. off gas treatment system may continue provided grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , t

ACTION 125 -

Deleted i

With the number of channels OPERABLE less than required by the I

ACTION 126 -

Minimum Channels OPERABLE requirement, suspend release of radio-activity effluents via this pathway.

f

/

I l

CLINTON - UNIT 1 3/4 3-104 i

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Page 51 of 157

'( a)

Ch2nne1 OPERABILITY Shall includc. Ybe capability of either +he Main. Confrol Room Central Confrol Terminal (McR-CCT) or the. Radiation Protection O(( ice Central Confr Terminal (RP-CCT} io provide Ybe alarm SYatus ? o-f the app icable radiation rnonitor channel (s).

l

{

With both the. MCR-CCT anc} RP-CCT inoperabl l ACTIO N 121 -

a.

per{orm a CHANNEL CHECK using loca} monito indication wi+hin 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; and at leasi oncepej 8 - hours there a{4er.

h.

With the ' noble fas activity monitor channel (s) e.

luent releastS-via.

otherwise inoperable 3

this pathway may continue provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed for gross noble gas activ.'

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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Attachment.3 to U-601048;

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Page 54;of'157 TABLE 4.3.7.12-1 (Continued) 1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE' NOTATIONS

At all times.

During operation of the standby gas treatment' system.

'Ouring operation of the main condenser air: ejector.

.y (1) he CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm-annunciation occurs.if any of the following conditions exist; lac e 1.

Instrument indicates measured levels above the alarm setpoint.-

[ Rep %

{I w

2.

Circuit failure.

gg 3.

Instrument indicates.a downscale failure.

j 4.

Instrument controls. not set in operate mode.

t The initial CHANNEL CALIBRATION shall be performed using one or more of' thel (2): reference standards certified by the National' Bureau of Standards or using standards that have been obtained from suppliers that participate in measure -

ment assurance activities with NBS. These standards shall. permit calibrat--

l ing the system'over its intended range uf energy and measurement range.

Subsequent CHANNEL CALIBRATION shall:be performed using the initial radio-active standards or~other st.andards of equivalent' quality or radioactive.

j sources that'have been related'to the initial calibration.

l The CHANNEL CALIBRATION shall include the use of standard ' samples contain-(3) ing a nominal:

l 1.

1.0 vol. % hydrogen, balance nitrogen, and 2.

4.0 vol. %. hydrogen, balance nitrogen.

6 hall Al5c> dcYc m'ine Maf c.bne }

e (4) The CH h M N EL CRECK.

C o m mu icc8co i5 edOh5kd 40 the. M Cf(- CCT o r RP-ce t

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to.U-601048 77 Page.55lof;157' i

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Th e.; CHANNEL FUNCTl0N ALf TEST 1 shall: also demonstrate dat ifke

~

L.M'c R" CCT or; R P-CCT responds wi}k 2 annunciation andjeNent printouti Yo_. eac O of.tk~e foNowing condMsf

'i li-lhsYrumenk indicades mea.sured ltvels above tbe :aka s

(HIGH).sek oinf.

p D ef ector ' failure'-(1.OW FAIL].l Hl FAIL [..

2..

S ample' ikou; f ailu,re (EXTERNAL PAIL)'.

3.

4. JnsIrument noY sef-in n$rM al operatc. mode (UNINITIALIEEP, CA't.lB RATE, ; MAINTENANCE or - STA N DBY ).

' I 7

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Attachment.-3

'j to U-601048 Page 56 of 157 RADIOACTIVE EFFLUENTS

)

MAIN' CONDENSER LIMITING CONDITION FOR OPERATION 3.11.2.7 The radioactivity' rate of noble gases measured at the offgas recom-biner effluent shall be limited to less than or equal to 289 millicuries /sec after 30 minutes' decay.

APPLICABILITY:

During operation of the main condenser air ejector.

ACTION With the radioactivity rate of noble gases at the offgas recombiner effluent exceeding 289 millicuries per second after 30 minutes decay, restore the gross I

radioactivity rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 .

SURVEILLANCE RE0VIREMENTS 4.11.2.7.1 The radioactivity rate of noble gases at the offgas recombiner ef fluent shall be continuously monitored b h h+reht of-6as peacess race X/

m.aiw reqvM +e be OP M e>t,g er as erkuu m. FeMed % %,le. u.7,i-l.

I The radioactivity rate of noble gases from the offgas recombiner.

4.11.2,7.2 effluent shall be determined to be within the limits of Specification 3.11.2.7 at the following frequencies by performing an isotopic analysis of'a represen-tative sample of gases taken at the discharge (prior to dilution and/or dis-charge) of the offgas recombiner; At least once per 31 days.

a.

b.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase, as indicated by the Pr3 treatment Off-Gas process radiation monitor Oisted in Table 3.3.7.1

, of greater j

than 50%, af ter factoring out increases due to changes}in HERMAL POWER level, in the nominal steady state fission gas release from the primary coolant.

(

replo.cc wih "rquined b be. CFEMSLdcras crwerwise gended g Table. 3.n !-l "

CLINTON - UNIT 1 3/4 11-17 l'L--

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to U-601048 Page 57 of 1571

. PACKAGE NUMBER 6 1

Summary of Proposed Changes -

The changes proposed for the Accident Monitoring Instrumentation

Technical Specifications are addressed as foe.r changes and are

. summarized below. - A description and justification as well as a basis for no significant hazards consideration is provided for each of the four change 9 on the following pages.

(1) Under the'first proposed change, exceptions to Specification 3.0.4

-would be inserted into the' ACTIONS associated with Table 3.3.7.5-1.

The exceptions would permit entry into OPERATIONAL CONDITIONS 1, 2, 3 with an accident monitbring instrumentation channel (s) inoperable as provided in the individual ACTION statements.

l (2)- The second proposed change corrects a typographical error appearing

{

on p. 3/4 3-87.of the Technical Specifications.

(3) The third proposed change resolves an inconsistency existing between ACTION 81 associated with Table 3.3.7.5-1 and the general j

4 ACTION and Limiting Condition for Operation specified under l

Specification 3.3.7.5.

This appears to have been an oversight in to i

the Standard Technical Specification or a problem in adapting it f

the Clinton plant design.

i (4) The fourth proposed change would delete the safety / relief valve.

1 acoustic monitors frem the Accident Monitoring Instrumentation of j

the CPS Technical Specification on the basis that the 3.3.7.5 and 4.3.7.5 Specifications are redundant to the requirements In Specifications 3.-4.2.1 and 4.4.2.1.1.

Without this change, some l

minor administrative / technical differences between the two Technical Specification sections may be needed to be corrected in a later license amendment.

l l

Description and Justification for Change (1)

Specification 3.0.4 states that entry into an OPERATIONAL CONDITION or other specified condition shall not be made unless the conditions for

he Limiting Condition for Operation are met without reliance on pcovisions contained in the ACTION requirements This specification entry into OPERATIONAL CONDITIONS 1, 2, 3 when an ACTION would prohibit statement under Specification 3.3.7.5 is in effect due to an inoperable j

accident monitoring instrument.

Although the instrumentation is required to be OPERABLE in OPERATIONAL l

CONDITIONS 1, 2, 3, the instrumentation is designed and intended to be used to " assess plant and environs conditions during and following an l

accident" (Ref. Reg. Gu1de 1.97).

Thus, entry into Conditions.1, 2, 3 l

does not necessarily correspond to entry into the conditions in which l

}:

the accident monitoring instrumentation will be needed. Conditions 1,

-1 1

j-2, and 3 do correspond, however, to the OPERATIONAL CONDITIONS in which

]

With chese l

l a design basis accident is most likely to be initiated.

considerations in mind, entry into Conditions 1, 2, and 3 should be l

I l

l l

1 l

l-I

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i to U-601048 I

Page 58 of"157:

']

~ permitted with the'. number of OPERABLE channels less than_the REQUIRED

=

' tiUMBER '0F CHANNELS requirement but not with the number 'of OPEFABLE channels less than the MINIMUM CHANNELS' OPERABLE requirement (except when the ACTION provides compensatory measures as' discussed below).

Exceptious to Specification 3.0.4 would therefore be inserted into

-ACTIONS 80, 81 and 82 as shown.

Y For ACTION 80, a 3.0.4 exception vould be inserted only into that

>f portion of the ACTION addressing inoperability with the. number of OPERABLE channels less than the REQUIRED NUMBER OF CHANNELS in '

'j accordance with the discussion above. Entry into OPERATIONAL CONDITION 1, 2, or 3 with th; number of 0PERABLE channels less chan the.MINDfUM CHANNELS OPERABLE requirement would still be prohibited. Entr'y with j

less than the REQUIRED NUMBER OF CHANNELS would be allowed with part "c"

{

of ACTION 80 in effect.

(This would require the-inoperable channel (s)

~

to be restored to OPERABLE status within 7 days'or the plant would have to be placed in ROT SMUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .)

For each of the less severe ACTIONS, 81 and 82,.the 3.0.4 exception would be applicable to the entire ACTION because the ACTION time is longer and because' compensatory measures (preplanned alternate method of monitoring under' ACTION 81 and use of alternate indication methods under ACTION 82) are provided according to the ACTION. That'is, it should be permissible to allow the plant to enter OPERATIONAL CONDITION 1, 2, or 3' with the channel (s) associated with these ACTIONS inoperable since the ACTIONS require compensatory measures to be taken as applicable'.

In addition, like most of the instrumentation identified in Technical Specification Section 3/4.3.7, MONITORING INSTRUMENTATION, the accident monitoring instrumentation does not provide any eutomatic actuation or isolation trip functions. All of'the other subsections under section 3.3.7 (Specifications 3.3.7.1,.3.3.7.2, 3.3.7.3, 3.3.7.4, 3.3.7.6, j

3 3.3.7.8, 3,3.7.10, 3.3.7.11, 3.3.7.12) contain exceptions to Specification 3.0.4.

IP feels that the 3.0.4 exception has been

]

that conservatively applied to th1s Specification in view of the fact entry into OPERATIONAL CONDITION 1, 2, or 3 is permitted only if the y

-l minimum number of channels are operable or if compensatory actions are taken (within the required time limits as applicable) according to the less severe ACTION statements.

1i It should also be noted that a 3.0.4 exception appears in the Accident Monitoring Instrumentation Technical Specification approved for the l

Perry Nuclear Power Station.

3 asis for No Significant Hazards Consi eration for Change (1) l d

According to 10CFR50.92, a proposed change to the license (Technical l

I Specifications) involves no significant hazards consideration if I

operation of the facility in accordance with the proposed change would (1) involve a significant increase in the probability or l

not l

consequences of an accident previously evaluated; or (2) create the l

possibility of a new or different kind of accident from any accident reduction in a margin previously evaluated; or (3) involve a significant l

of safety.

l 1

l L

____________a

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j

<'i s

c i,.-

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to U-601048 Page,59 of 157 (1) Although the accident' monitoring instrumentation is required to provide operator information that may be needed to mitigate the.

-consequences of an accident (such as providing the indications needed by an operator to enable the' operation of manually initiated safety systems), the proposed change does not involve a significant

]

increase in the probability or consequences of an accident j

previously evaluated for the following reasons.

l 3

i (a)- The 3.0.4 exception would be conservatively ' applied such that

-i

'd the minimum number of accident monitoring channels would still-be required to be OPERAELE (for those accident monitoring f

instrumenc channels associated with ACTION 80) thus ensuring q

essential monitoring capability upon entry into OPERATIONAL l

l CONDITION 1, 2, or 3.

(

f (b) ACTIONS 81 and 82 currently allow certain instruments.to be inoperable provided that compensatory meaaures are taken (either immediately as in ACTION 82 or within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months as in

]

ACTION 81)..

1 (c) The difference in unavailability of an accident monitoring instrument (s) in OPERATIONAL CONDITION 1, 2, or 3 without the i

exception to Specification 3.0.4 versus including the

{

exception is associated with the out-of-service time that would otherwise be allowed as provided in the ACTION statement upon entry into OPERATIONAL CONDITION 1, 2, or 3.

(Inclusion of a 3.0.4 exception has no impact on the probability of i

failure of an instrument beyond this initial allowed out-of-service period each time an entry is made into OPERATIONAL CONDITION 1, 2, or 3.)

The allowed out-of-service time for a single inoperable redundant channel under ACTION 80 is seven days. In view of the fact that the surveillance test

-1 j

interval for-the CHANNEL CHECK is 31 days ("M"), the ACTION time is relatively severe (short). When considered together d

with the probability of occurrence of an accident condition

]

that would require use of the instrument (s) relative to the l

cotal amount of time that the plant is operated in OPERATIONAL i

CONDITION 1, 2, or 3, the application of a 3.0.4 exception should not significantly increase the unavailability of an accident monitoring instrument (s) and therefore does not l

involve a significant increase in the consequences of an accident previously evaluated.

(The proposed change for the noted monitoring instrumentation Technical Specification has I

no impact on tha probability of occurrence or the initiation l

of an accident.)

l (2)

It has been determined that the proposed change does not create the possibility of a new and different kind of accident from any previously evaluated. The proposed change does not involve any changes to the plant as-built design. The accident monitoring

)

instrumentation is designed and intended for monitoring purposes in

}

a wide variety of accident conditions. The impact of the proposed l

f to the operability (availability) of this change with respect instrumentation was addressed previously.

m l

to U-601048 Page 60 of 157 (3) The proposed change does not involve a significant reduction in c.

margin of safety. Since the accident monitoring instrumentation does not perform any automatic isolation er initiation trip functions, no automatic initiation / isolation trip setpoints are l

I involved. No changes.to any alarm points (as applicable) are If the instrument availability is considered a margin of involved.

safety, the impact on that aspect has been previously evaluated.

Description and Justification for Change (2)

The purpose of this change is to correct a typographical error identified on p. 3/4 3-87 (Table 3.3.7.5-1) of the CPS Technical Specifications. The "t" note for the suppression pool water temperature sensors refers to Specification "3.5.3.1."

It should refer instead to Specification "3.6.3.1" where requirements for the other suppression pool temperature sensors are specified.

Basis for No Significant Hazards consideration for Change (2)

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 tn accident previously evaluated.

This is an administrative change to correct a typographical error and does not affect any accident analysis.

(2) The preposed change does not creata the possibility of a new or different kind of accident from any accident previously evaluated.

This is an administrative change to correct a typographical error and does not affect plant design.

(3) The proposed change does not involve a significant reduction in a margin of safety. The correction of this typographical error does not affect a margin of safety.

Description and Justification for Change (3)

The Limiting condition for Operation under Specification 3.3.7.5 states that the channels shown on the table shall be OPERABLE. The general ACTION statement within the LCO states, "with one or more accident monitoring instrumentation channels inoperable, take the ACTION required by Table 3.3.7.5-1."

ACTION 81 on Table 3.3.7.5-1 however, does not explicitly state what action should be taken if one drywell and/or containment high range gross ganma radiation monitor channel is inoperable. This is explained as follows.

i l

'4 j

'I Attachment'3

)

'4 to U-601048 1

o.

-Page-61~ of 157

{

' ACTIONS 80 and 82 each contain two levels of ACTION (parts "c." and "b"

)

t of each ACTION). The first level (pert "a") addresses the r.cndition l

vith the number of O'PERABLE chacnels less than the Required Number of Channels requirement; the seccnd level addresses the condition with the j

'l number of OPERA 3LE channels less than the Minimum-Channels' Operable requirement. ACTION 81 however only (currently) addresses the latter condition even though item 11 on* Table 3.L 7.5-1 (Containment /Dryvell 1

High Range Gross Gamm.2 Radiation Monitors) lists a greater number of i

channels in the.7.EQUIRFD NUH.BER OF CHANNELS column than in the MINIMUM CHAhNELS OPERABLE column '(as appropriate). Thus no action is currently i

d required if,.for example, one containment high range gross gamma radiation monitor is out of service cince the number of channels OPERABLE in that case vould be less than the' REQUIRED NUMBER OF CHANNELS requirement but greater than the MINIMUM CHANNELS OPERABLE requirement.

The propo wd change would revise the apparent inconsistency between the ACTION associated with the Limiting Condition for Operation (3.3.7.5) and the ACTION (81) specified on the associated cable-(Table 3.3.7.5-1),

IP believes the proposed ACTION resolves the inconsistency without changing the intant of the Specification.

1 Basis for No Significant Hazards Consideration for Chance (3)

According to 10CFR50.92, a proposed change to the license.(Technical 1

Specifications) involves no significant hatards consideration if 3

operation of the facility in accordance with the proposed change would (1) involve a significant increase in the probability or not consequences of an accident previously evaluated; et (2) create the possibility of a new or different kind of accident from any accident' previously evaluated; or (3) involve a significant reduction in a margin of safety.

]

The proposed change does not invcive a significant increase in the 1

(1) probability or consequences of an accident previously evaluated.

The change provides clarification of the existing Specification and

{

s does not change the intent of the Specification. There is no effec ~ on existing accident analyses.

J j

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

i the existing Specification without changing the intent of the Specification does not affect plant design, and therefore does not j

{

create the possibility of a new accident scenario.

j (3) The proposed change does not involve a significant reduction in a This is an administrative change that phovides margin of safety.

clarification without changing intent and therefore does not affect

)

a margin of safety.

l 1

4 I - J 1

Attach.nent 3

'l to U-601048 Page 62 of.157 j

i Description and Justification fcr Change (4)

Requirements for the Safety / Relief Valve (SRV) Acoustic Monitors appear in two secticus of the CPS Technical Specifications:

(1) The Accident Monitoring Instrumentation.section (Technical Specifications 3.3s7.5 and.

4.3.7.5) and - (2) The Saf ety/ Relief Valves seccion (Technical Specifications 3.4.2.1 and 4.4.2. z.1). Due'to redundancy between the two sections, the SRV Acoustic Monitors should be deleted from the Accifent Monitoring Inst: a yntation section of the Technical Specifications. The SRV seu.ica (3.4.2.1, 4.4.2.1.1) contains the me.t precise requirements regarding the ACTION and surveillance requireman o for the monitors, to that s6ction should be retained. IP had considered.

proposins changes that had been identified to resolve seme minor technical administrative differences between the two sections, but it was believed that regt.esting a change to delete the conitors from the i

The Accident Monitoring Instrumentation section was more appropriate.

deletion as requested (and as shewa on the marked-up pages from the Technical Specifications) would not remove any requirements for maintaining and demonstrating OPERLSILITY of the nonitors since these i

requirements will remain unchanged.in Technical Specifications 3.4.2.1 and 4.4.2.1.1.

l The proposed change is consistent with the Technical Specifications approved for the River Lend Station.

Basic for No Significant Hazards Consideration'for Change (4)_

According'to 10CFR50.92, a nroposed change to the license (Technical Specifications) involves no significant hazards consideration if operation of the facility in accordance with the proposed change would l

(1) involve a significant increase in the probability or not consequences of an accident'previously evaluated; or (2) create the l

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) Deleting the SRV acoustic monitors from Specification 3.3.7.5 (and 4.3.7.5) does not involve a significant increase in the probability or consequences of an accident previously evaluated because OPERABILITY of the monitors will be required and demonstrated under i

The OPERABILITY, ACTION and Specifications 3.4.2.1 and 4.4.2.1.1.

Surveillance requirements specified in Specification 3.3.7.3 (and 4.3.7.5) are equivalent to those specified in Specification 3.4.2.1 (and 4.4.2.1.1).

The proposed change does not impact Clinton's to Regulatory Guide 1.97 in that the requirements of commitment Specifications 3.4.2.1 and 4.4.2.1.1 will require OPERABILITY of the acoustic monitors to be maintained. This will help to ensure that in case of an accident, SRV position status will be provided to the operators to allow them to respond appropr1ately to accident conditions.

(r; 1-i i ~

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-. g g'

.Page 63 of'157

.j i

(2)' The proposed change does not create the poss1bility of a'new or different' accident.from any' accident previously evaluated because

-l

'nt. design changes or new modes of operation are introduced by the

'j change.

l (3) The proposed change does not involve a'significant reduction in a 1

margin of safety as far as a margin of' safety applies'to the acoustic monitors.. The Technical Specification requirements remain

.i essentially. unchanged.

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Attachnent 3

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to U-6n!nt$

Page 64 of 157 1

INSTRUMENTATION ACCIDENT MONII'ORING INSTRUMENTATION i

.L..IMITING CONDITION FOR OPERATION' 3.3.7.5 The accident monitoring instrumentation channels shown in Table 3.3.7.5-1 shall be OPERABLE.

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.

SURVEILLANCE REQUIREMENTS 4.3.7.5 Each of the above required accident monitoring instrumentation f

J channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK l

and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.5-1.

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Page 66 of 157 j

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- TABLE 3.3.7.5-1 (Continued)

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ACCIDENT MONITORING INSTRUMENTATION

)

ACTION 1

With the number of OPERABLE accident monitoring instruments-ACTION 80 -

a.

1 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 SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . %. prems,en: J SMoon 3 c4 m

a n u w e..

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 SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

F 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:

Initiate the preplanned alternate method of monitoring the

'P@

a.

g"'f]

appropriate parameter (s), and b.

Prepare and submit & 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 1

to OPERABLE status.

ACTION 82 -

With the number of OPERABLE accident monitoring instrumentation 4

a.

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 (s) to OPERABLE status within 30 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 and in COLD SHUTOOWN within the fol-lowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , With the number of OPERABLE accident monitoring instrumentation b.

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 COLD SHUTOOWN within the i

following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

}

1 I

c. Th. provisiee of Spec /tcotco 3.6 4 are nor.apaWe..

I

)

CLINTON - UNIT 1 3/4 3-88

, Attachment 3 to U-601048 Page 67 of 157 x 4

-\\

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ACTION 81 -

a '.

Wiih the number of OPERABLE channels less fhan ibe Reguired Number eji Channels shown in Table t 3.3.7.5 - 1, operation. rnay continue provided fhe Minimum Channels OPER ABLE requirernent is rnet'..

l b.WiththenumberofOPERABLEChannelslessthanrequiredbythe 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:

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

2. Prepare and submit a Special Report to'the Commission pur.

f 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.

C. Tht. provisions f $pecikicafien 3.0.4 are not app caWe.

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l to U-601048 Page 69 of 157 q

l PACKAGE NUMBER 7 g

Description of Change 1l The purpose of this proposed change is to delete CPS Technical Specification 3/4.3.8, " Turbine Overspeed Protection System": based on the ' justification provided below. Deletion of the Specification would relieve Clinton of its commitments to continue inspecting and not testing the applicable valves and instrumentation as discussed in FSAR Section 10.2.3.6.

The intent is to remove these requirements from the plant Technical Specifications based on the following.

From a safety point of view, the chief concern for turbine overspeed from protection is the possible hazard to safety-related equipment

" missiles" that could possibly be generated due to a turbine overspeed condition. The turbine-generator orientation at CPS is favorable for from reducing,the probability of damage to safety-related equipment turbine missiles since all safety-related components and structures are located in the axial direction from the turbine-generator. Because of the turbine-generator location and orientation, safety-related equiprent lies outside the low-trajectory missile strike zone which is defined as the zone where objects may be subject to a direct hit from a turbine

' missile. Therefore, the only' type of turbine missile that presents a hazard to CPS safety-related equipment is a high trajectory missile, defined as a missile that initially travels upward and then can cause

]

damage when falling.

CPS FSAR'Section 3.5.1.3, using GE and NUREG-0800 (NRC Standard Review Plan) turbine failure data, indicates that the probability of damage The from high trajectory turbine missiles is acceptably low.

probabilityofdamagetosafety,{qlatedequipmentbasedonGE per year and based on NUREG-0800 data

~

turbinefagluredatais1.0Ex10Both values are less than the NRC acceptable risk is 7.5x10 per year.7 per year for the loss of an essential system from

~

rate of less than 10 a single event.

The above analysis was provided to NRC in Letter dated June 30, 1986 (U-600619).

Furthermore, deletion of the Turbine overspeed Protection System Technical Specification is supported by the fact that the Specification the criteria described by the NRC Interim Policy Statement does not meet on Technical Specification Improvements noticed in the Federal Register on February 6, 1987.

L l

i i

l'

1 to U-601048 Page 70 of 157 Basis-For No Significant Hazards Consideration According-co 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 j

I 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 (1) The proposed change does not involve a significant increase in the j

probability or consequences of an accident previously evaluated l'

because it has been determined that, although the turbine overspeed protection syste-is necessary for protection of the turbine from both an operational and economic point of view, the system (and therefore the Technical Specification) is not essential to j

mitigating the consequences of an accident, and that the probability of damage to safety-related equipment is acceptably low l

based on the probability of turbine failure and the orientation of j

~

the turbine-generator at Clinton.

(2) The proposed change does not create the possibility of a new or different kind of accident from any previously analy::ed. It was.

determined that the turbine missile accident was the applicable accident to be considered for analysis in providing a basis for the i

proposed change. No new kind of accident is created by the deletion of specification 3/4.3.8.

(3) The proposed' change does not involve a significant reduction in a margin of safety. The favorable orientation of the turbine i

provides a margin of safery such that the possibility of missile

,l damage to safety-related equipment is acceptably low so that a Technical Specification for turbine overspeed protection is not

]

required, l

l l

l 1

l l

1 l

to U-601048 Page 71 of 157 i

_ INSTRUMENTATION 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM y

i LIMITINGCONDITIONFOROPERATIO$

q

~

y At least one turbine overspeed protection system shall be OPERABLE.

l

~~

e APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

I With one turbine control valve, one turbine stop valve and/or with one 2

a.

turbine combined intermediate valve inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or close at least one valve in the affected steam lead or isolate the turbine from the steam supply j

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With the above required turbine overspeed protection system otherwise i

j b.

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

The provisions of Specification 3.0.4 are not applicable for entry into c.

OPERATIONAL CONDITION 2 provided the turbine is tripped and the turbine i

I is isolated from the main steam supply by closed turbine stop and control

{

valves.

I

[

fSURVEILLANCEREQUIREMENTS l

l 1

l l

4.3.8.1 The provisions of Specification 4.0.4 are not applicable.

4.3.8.2 The above required turbine overspeed protection system shall be 1

l l

demonstrated OPERABLE:

l 3

)

At least once per 7 days by cycling each of the following valves from the a.

l running position ana observing valve closure:

i f

1.

Four low pressure turbine combined intermediate valves, and J

i 2.

Four high pressura turbine stop valves.

I At least once per 31 davs by cycling each of the four high pressure turbine b.

control valves from the running position and cbserving valve closure.

At least once per 18 months by performance of a CHANNEL CALIBRATION of I

{

r.

the turbine overspeed protection instrumentation.

i At least once per 40 months by disass abling at least one of 'each of the

,/

t l

d.

above valves and performing a visual and sur face inspection of all valve seats, disks and stems and verifying no unacceptable flaws or excessive If unacceptable flaws or excessive corrosion are found, all corrosion.

other valvw uf that type shall be inspected.

x t

3/4 3-108 CLINTON - UNIT 1

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PACKAGE'NUM3ER 8

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s Description of Change J

I Technical Specification 3.4.2.2 currently specifies 'a tolerance of 22k 6

/

for the Low-Low Set Function.Setpoint associated with the safe,ty/reliep

' valves (SRVs). The General Electric (GE) design specificacMns -

1

,'h

) ;

(22A4622AV), however, specify a tolerance of g 15 psi. The-lattet is H

4f consistent with. th tolerance specified for the' relief functiop ;h[the j

SRVs (specified in Technical Specification 3.4.2.1) since *.he ?rel d V.

function and the low-low set functions are botb effected by pressure [ '

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r

. sensor - crip unit 1ustrument loops and since L5 psi is the toleran'ct.

A' specified for these instrument loops.

^Q*

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j

,d; A review of the curec documentation used t

.i "f

Specifications for. Clinton revealed that 'g o, develop the draf t Te

12%" was the tolerance originally recommended by GE.

The source document, however, refers to the above-mentioned design specification (22A4622AV)t. The problem was i

discussed with GE, and although the exact oridin of th2 originally

~

opecified tolerance was not determined, it was agreed that the value should be consistent with the current design specification.

y l

b i

Therefore, the tdlerance specified in the "Lowdok set Function l

2 Setpoint"columninTechnicalSpecification3.4.2.1'shouldbechanfed[

f' -

f rom "12%" to "t15 psi".

Itoshould be noted that the latter reprei dr.y.

i a tighter tolerance and in. the value that has been specified in dad plant calibration procedure.

['

l

/

By is for No Significant Hazards Consideratit3 j

f l

According to 10CFR50.92, a proposed change to the license di Technical

. Specifications) involves no significant hazakds consideration /;.f a

e would operation of the facility in accordance with the propMgd chat s

d not (1) involve a significant increase fn',rhe probability or;3 consequences of an accident previously evalpned;.et (2)jcreate he j

possibility of a new or different kind of accident from any accident previously eveluated; or (3) involjt a.sig,nqficantreductioninamargin

' l of safety.

The proposed change does not involve a significant increase in the (1) probability or consequeacns of an accident previously evaluated.

k This change establishes konformance with the General Electric design specification values specified for the setpoint tolerance of

}

the Safety / Relief Valve Lov-Low Set function. This change does not j

affect any previous accident analyses.

j

.J The proposed change does not create the possibility of a new or-1 (2) different kind of accident ~from any accident previously evaluated.

l A new accident scenario has act been created by changing this j

tolerance to match the intended design value.

(3) The proposed change does not involve a significant reduction in a margin of safety. Changing the setpoint tolerance to match the affect c design value is en administrative change that does not

- margin of safety.

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3 Page 73 of 15 7 RitV. TOR CODLANT f/i$Tp L

SAbhREthi ES' LOW OW SET FUNCTION i

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LIMITyg}CONDITIONFOROPER,ATION

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3.4ip.f The low-low set function of the following re ktor coolant system i'

safety /re'ief valves shall be OPERABLE with the following settings *:

d Low-Low Set Functio Ii

/6 Setecint* (nsig) A Q 15 p.t.

y/

Valve No.<

Open Cicae j-t,

,1033 926 Y

F051D'['I/

/ 1073 936

F051C, 1113 946 F

F047F?'

g F051Ef N':

U13 946 1

4 F051G M

1113 946 c

j j APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With the low-low set functi n of one of the abov.e required reactor coolant a.

system safety / relief valves inoperable, restore the inoperable low-low set function to OPERABLE status within 14 days or te in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWy Wthin the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1.

.j.

b.

With the low-low set function of more tNn one of the above required reactor coolant system safety / relief %Nes' hoprable, 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 SHUTDOWH within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With either low-low set function pressure actuntion trip system "A" or c.

< B" inoperable, restore the inoperable trip spstem to OPERABLE status c.

T within 7 days; otherwise, be in at.liast HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

}e yx andinCOLDSHUT,DOWNwithinthefoyqding24 hours.

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x SURVEILLANCEREQUIREMEKTS j

(

)

4'.4.2.2 The low-low ut function pressure actuatiod instrumentation shall be demonstrated OPERABLE by performance b' a:

CHANNEL FUNCTIONAL TEST, includfng calibration of. the trip unit, at least l

a.

once per 31 days.

i CHANNEL CALIBRATION and LOGIC SYSTEM FUNCTIONAL TES7 e4 leest once per 18 b.

Each of the two trip systems or divisions of the low-low set months.

function actuation logic associated # th the Nuclear System Protection

{

System shall be manually testecf independent of the SELF TEST SYSTEM during e

1 separate refueling outages such that both divisions and all channel trips j

aretestedatleastonceeveryfourfuelcycles.t f,

0ne channel may be placed in an incterable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for the pur-pose of,perf orming surveillance teting in accordance with Specification 4.4.2.2.

- The lif t setting pressure shall correspord to ambient conditions of the valves at noninal operating temperatures and bressures, tManual testing for the purpose of satisfying Specification 4.4.2.2.b. is not re-quired until after shutdown durhg the firret regularly scheduled refueling outage.

i CLINTON - UNIT 1 3/4 4-11

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! PACKAGE NUMBER __9_

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'. Twoi changes are" proposed for the CPS Technical Specifications regarding

' reactor coolant system leakage. The dirst proposed change would modify g$.

Specification 4.4.3.2.1.b to reflect the fact that ths drywell floor and y'

.aquipment drain sump leak detection ryacem instrumentation does not

'j inclu6e direct quantitative indicatim of sumNaval as the. current x.

3 41' Technical Specification implies. The second pr3 posed change would add'a, tote attached to Specification 4.4.3.2.1.a to'indleate that the drywell arsospheric particulate and gaseous radioactivity monitoring system does M 1 not provide a means of quantifying' leakage for determining that" leakage ag' t'

is within the. limits specified :in the Limiting Condition 'for OperatSn 4

-J (3.4.3.2).

<a A more, detailed 3e.ocription $tud justifica$ ion as well'as a basis for no significant trazards consideration is provided for,each of thesa two l'

. changes below.

_ 'scription and Justiff-:stion for Change (1)

De The drywell floor -drain sump and equipuent drais sump leak de:ection

' system (s) is designed (t the requirements of Regulatory Guide 1.45.

A y

system of timers monttors sump fill rate and pump-out frequency from

'tU which average leakage rates are determined. No direct quantitative indiention of sump level (other than an alarm) is provided nor is it W

j, required. Monitoring'of the aump flow rate meets the intent of the Technical Specification (3.4.3.2/4.4.3.2.1.b). Therefore, the rechnical s_

y Specification should be revised as showr. on the marked-up page,to resove-the implication that sump level will be directly monitored.

{3 I

j Basis for No Signifidact Euards Consideration Change (1)

Accordingto10CFRSO92,apropbsedchangetothelicense(Technical Specifications) involvec no significant ha:crds consideration if operation of tha facility in accordance with the proposed change would (1) involve al significant increase in the probability or not consequences of arl 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 l

t i

of safety.

(1) The' proposed change'does neh involve a significant increase in the pobability or consequences of an accident previously evaluated because the drywell floor and equipment drain sump leak detection system does not provide any automatic trip functions for effecting l

1solations or actuating safety systems required to mitigate the cctssequences of an accident. However, in accordance with General Design driterion 30, the system does provide a means for detecting, and to ch.etextent practical, identifying, the location of the source of reactor coolant leakage that could lead to an accident or

}

jeoparlize safety. In this respect, the proposed change does not The the system's capability of performing its function.

affect only purpose of the proposed change is to make the noted Technical j

Specification (4.4.3.2.l.b) consistent with system design.

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m to U-601048 Page 75 of 157

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(2) The proposed chang;s does not create the possibility of a new or j

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different kind of accident from any accident previously evaluated because the proposed change does not involve any changes to the as-built plant design nor does it introduce any new mode of operation such that a new or different accident must be considered.

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(3) The proposed change does not involve a significant reduction in a

]

margin of safety because the proposed change has no impact on any j

eetpoint, analytical or design limit assumed or required by any 4

accident analysis. Sump level indication is not relevant to any 1

margin of safety or the-system's capability of determining sump flow rates and hence leakage rates.

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+

Description and Justification for Change (2)

O

[h'

.Clinton's as-built design for reactor coolant leakage detection employs g

more than the minimum required de:ection method 9 prescribed by 5

Regulatory Position C.3 of Regulatory Guide 1.45, " Reactor Coolant 3.

Presoure Boundary Leakage Detection Systems." In tha Clinton Safety L' :

(SER), Section 5.7. 5, NRC concluded that the leakage Evaluatica Report l

detection systems employed at Clinton are in compliance with the guidance found in Regulatory Guide 1.45 and satisfy the requirements of General Design Criterion 30.

NRC also noted that the Technical i

Specifications snall limit unidentified leakage to 5 gpm and identified

~

\\'

Icakage to 25 spm. Technical Specification 4.4.3.2.1 implies, however, that each and every detection system listed in the Specification can be used to quantify leakage for the purpose of determining that leakage is-within the limits specified in the Limiting. Condition for Operation n

j

[j (3.4.3.2).

1 In a letter (U-600809), dated January 12, 1987, Illinois Power informed theihRC of a revision to tho' CPS FSAR which revised IP's coc:mitments to IP intends instead to commit to compliance with Regulatory Guide 1.45.

ANSI /ISA Standard S67.03-1982, "Standatd for Light Water Reactor Coolant

?ressure Bcundary Leak Detection." The letter stated, "This is being done because Regulatory Guide 1.45 is technically out-of-date, ANSI /ISA S67.'03-1982 is new recognized as the more appropriate requirements y

document for leak detection systems, and the air particulate monitoring system is not capable of detecting a leakage rate of 1 gptt within cn (A new FSAR hour as required by Position C.5 ef Regulatory Guide 1.45."

revision wil.1 be implemented in the next update of the FSAR.)

To be consistent with this position, a change en the Technical Specification is proposed to remove the existing implication in the Technical Specifications which suggests that monitoring the drywell atmospheric particulate and gaseous radioactivity will demonstrate that reactor coolant system leakage is within the limits specified in the Drywell particulate and gaseous Limiting Condition for Operation.

rad.ioactivity is associated with unidentified leakage, and although the O-drywell radioactivity monitoring instrumentation meets the instrument g

sensitivities (of 10-6 pCi/cc radioactivity for air particulate monitoring and of 10 yC1/cc radioactivity for radiogas monitoring) as L

well as the seismic qualification requirements specified by Regulatory i

Cuide 1.45, no means currently exists to precisely relate the monitored r%

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to U-601048 Page 76 of 157 j

variables to a reactor coolant leakage rate..The proposed change to the

.(

Technical Specifications would incert a note attached to' Specification-4.4.3.2.1.a for the drywell particulate and gaseous rautoactivity monitors which states (in parenthesis), "not a means of quantifying leakage." Appropriate changes to the BASES are also requested.

(See attached marked-up pages from the Technical Specifications and BASES.)

The proposed change is consistent with the same Specification approved i

for the Perry Nuclear Power Plant,

.f Basis for No Significant Hazsrds Consideration j

According to 10CFR50.92, a prctosed change to the license (Technical Specifications) involves no significant hazards consideration if operation of the facility in accordance with the proposed change would (1) inv,olve a significant increase in the probability or

.not consequences of an accident previously evaluated; or (2), create the

.I possibility of a new or different kind of accident from any accident previously. evaluated; or (3) involve a significant reduction in a margin

]

of sa#ety.

(1) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the drywe'll atmospheric particulate and gaseous radioactivity leak detection system does not provide any automatic trip functions for effecting isolations or actuating safety systems required to mitigate the consequences of an accident. However (in accordance with General Design Criterion 30), the system does provide a means for detecting and to the extent practical, identifying the location of the source of reactbr coolant leakage that could lead to an accident or jeopardize safety. In this j

respect, the proposed change does not involve a change to the cystem's as-built design or the built-in capabilities of the system. The only purpose of the proposed change is to make the noted Technical Specification (4.4.3.2.1.a) consistent with system design.

It should be noted that no changes are proposed which would affect i

che requirements for OPERABILITY. Monitoring of the drywell

]

parti 2ulate and gaseous radioactivity will continue to be performed as required by Surveillance 4.4.3.2.1.a to note any changes or sudden departures from trends that may be established. The l

redundant systems identified in the Technical Specification will continue to provide the capability for quantifying leakage needed to ensure that leakage is within the specified limits.

(2) The proposed change does not create the possibility of a new or different kind of accident from any previously e aluated because no changes to plant design are involved. Plant / system operation will continue to be perfortaed in the same manner so that no new or different accident must be considered.

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b.

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Attachment.3 t.

to U-601048 Page 77 of 157 1

- (3).The~ proposed change does not involve a significant reduction in a.

margin of safety because the proposed change has no. impact on any:

setpoint, analytical or design limit assumed or required by any accident analysis. The alarm points and operating characteristics

.of the monitoring system are unaffected by the' proposed change.

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, Attachment 3'

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Page 78 of 157 f

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~ REACTOR C00LANT SYSTEM i

' OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2' Reactor coolant system leakage.shall be_ limited to:

' a.

No PRESSURE BOUNDARY LEAKAGE.

b.

5 gpm UNIDENTIFIED LEAKAGE.

4 I

' 25 gpm IDENTIFIED LEAKAGE (averaged over any 24-hour period) 3 c.

d.

0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm from any reactor coolant system pressure isolation valve specified-in

~l Table 3.4.3.2-1,'at rated reactor pressure.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

l ACTION _:

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within' a.

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

b.

With any reactor coolant system leakage greater than the limits in b and/or c, above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-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 , With any reactor coolant system pressure isolation valve leakage greater c.

than the above limit, isolat'e the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two other closed manual or deactivated automatic valves, or be in.at least

'l 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.

SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:

l Monitoring the drywell atmospheric particulate and gaseous radio-lake) activity at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sg(not a mu^5o+ guo hi$

a.

L t

.I MonitoringthedrywellfloorandequipmentdrainiumplevelandJsumpflow

.b.

rate at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , Monitoring the drywell air coolers condensate flow rate at least once per c.

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

l CLINTON - UNIT 1 3/4 4-13 t

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REACTOR COOLANT SYSTEM 1

4 BASES 3/4.4.1 RECIRCULATION SYSTEM (Co'ntinued) one of several hydraulic power unit or analog control circuit failure signals.

The " motion inhibit" signal causes hydraulic power unit shutdown and hydraulic isolation such that the flow control valve fails "as is."

This design feature insures that the flow control valves do not respond to potentially erroneous control signals.

q 1

Electronic limiters exist in the position control loop of each flow control

{

valve to limit the flow control valve stroking rate to 10 1% per second in opening and closing directions on a control signal failure. The analysis of the recirculation flow control failures on increasing and decreasing flow are presented in Sections 15.3 and 15.4 of the FSAR respectively.

The required surveillance interval is adequate to ensure that the flow control 1

valves remain OPERABLE and not so frequent as to cause excessive wear on the system components.

3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves (SRV) operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1375 psig in accordance with the ASME Code. A total of 11 OPERABLE safety-relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.

Any combination of 5 SRVs operating in the relief mode and 6 SRVs operating in the safety mode is acceptable.

Demonstration of the safety-relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specifica-tion 4.0.5.

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1 The low-low set system ensures that safety / relief valve discharges are minimized for a second opening of these valves, following any overpressure transient.

This is achieved by automatically lowering the closing setpoint of 5 valves and lowering the opening setpoint of 2 valves following the initial opening.

In this way, the frequency and magnitude of the containment blowdown duty cycle is substantially reduced.

Sufficient redundancy is provided for the low-low set system such that failure of any one valve to open or close at its reduced set-point does not violate the design basis.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE gy pc.pha. W '

3/4.4.3.1 LEAKAGE DETECTION SYSTEMS Ni i e The RCS leakage detection systems required by this specification are provided l

j to monitor and detect leakage from the reactor coolant pressure boundary. These l l detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

]!

t They provide the ability to measure leakage from fluid systems in the drywe jl.

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B; a/4 4 - 3 1

The RCS leakage defection sys4 cms reguired by fhis specihcab 'are provided do moni4or. and detect leakage { rom' fhe.reac4c boundary.

~T hese defection syslems meef +ke infent ch Regula4ery^

1.45

Reacier Coolant Pressure Boundary Leakage Defe'ction Sys4 ems.May j 3 'and are consistent wifh the recommendations ofANSI S67.03, teJ7 3

Standar[ for ' Light Water Reacier Coolant Pressure. Boundary Leak' D Icakage krom They provide fLe' abildy o detect and/or 4

measure 1982

}luid 'sysktms in +he drywell.

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Page 81 of 157 i

ACKAGE NUMBER 10 Summarv of Proposed Changes This change request consists of three proposed changes. The first i

proposed change consists of those changes required to' delete the OPERABILITY and surveil]ance requirements associated with 50' stops installed for the -VR/VQ* system containment isolation valvec on the basis that the 50' stops will now be considered a part of the permanent design for these valves. The second proposed change would insert footnotes into the Limiting Conditions for Operation and applicable surveillance requirements associated with Specifications 3.6.1.8 and 3.6.2.7 to exclude the time when valves are opened for performing stroke-time testing (required by the ISI program and/or Specification 4.6.4.3) from the cumulative system operation time limited by the Limiting Conditions for Operation. The third chcnge proposed would extend the application of Note "(a)" ("May be opened under administrative control") in Table. 3.6.4-1 of the Technical isolation valves Specifications'to include specific VR/VQ containment which need to be opened while conducting certain local leak rate tests.

All of these changes are described in more detail on the following A basis for no significant hazards consideration is also pages.

provided for each of the three changes.

Description and Justification For Change (1)

As noted on p. 3-16 of SSER 5, IP committed by letter dated November 17, 1983 to NRC (U-0678) to utilize mechanical stops to limit applicable valves from' opening more than 50*.

The stops would be in place during operational modes 1, 2 and 3 and would be removed during modes 4 and 5 to provide full purge flow if required during maintenance activities.

The NRC_ acknowledged in the,SER that "the staff has determined that with the confirmation of the limitation of the 24-inch and 36-inch valves to a 50' maximum opening angle...the applicant has addressed the staff's concerns in [this area)...".

Technical Specifications were subsequently developed to provide OPERABILITY and surveillance requirements for the 50* stops.

CPS has since resolved its position with respect to this is. sue and has deternined that' adequate purge flow will be obtained during modes 4 and That is, the 50* stops will not be y

5 with the 50' stcps in place.

periodically removed as originally suggested and vill chus be considered This a part of the permanent as-built design for the affected valves.**

obviates the need for those Technical Specifications originally provided to periodically confirm the 50' maximum opening angle limitation.

VR = Containment Building Ventilation System (including the low-volume ventilation system)

VQ = Drywell Purge System (Overlaps VR System) f

- This would not preclude removal of the 50* stops; but such removal to the controls and requirements of the plant would be subject (A modification requires a Safety Evaluation modification program.

per 10CFRSO.59.)

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to U-601048 Page 82 of 157 The proposed changes are'shown on the attached marked-up pages from the CPS Technical Specifications, 3.6.1.8, 4.6.1.8.2,

-3,

-4, -5 and 3.6.2.7, 4.6.-2.7.4.

Corresponding changes to the BASES are also included.

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 E

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 accida.ut 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 for the following reason:

IP's commitment remains unchanged with respect to the original requirements for limiting the amount the valves can be open (determined to be a solution to an NRC concern resulting from the POSI-SEAL analyses). Compliance with this commitment (installing 50* stops) ensures the valves will function as required for containment isolation during design basis accident conditions.

(2) The proposed change does not create the possibility of a new or different kird of accident from any accident previously evaluated because (1) the effecsive plant configuration (with respect to the-50* stops being in place) during OPERATIONAL CONDITIONS 1, 2, 3 remains unchanged; (2) it.has been determined that keeping the 50*

stops ir place during OPERATIONAL CONDITIONS 4 and 5 will still provide adequate purge flow which is the only concern of interest with respect to VR/VQ operation, and (3) the operability of these isolation valves, as also required in modes 4 and "#"

containment according to Table 3.6.4-1, will be unaffected. Thus, the impact of the change has been considered with respect to the plant configuration or conditions other than those associated with maintaining containment integrity The possibility of a new or different accident is not created by the change.

If the 50* limitation to opening the affected valves is considered (3) a margin of safety with respect to ensuring containment isolation capability and minimizing dose in accordance with 10CFR100, then t

tt.c proposed change does not involve a significant reduction in a j

Otherwise the

]

margin of safety for the reascns noted above.

proposed change does not require a change to any trip setpoints or design limits that could affect a margin of safety assumed in any accident analysis, i

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Description and Justification for Change (2)

Specification 3.6.1.8 states that the containment building ventilation 36-inch supply (1VR001A, B) and'36-inch exhaust (1VQ004A, B) isolation valve (s) may be open for containment ventilation system operation with such operation limited to less than or equal to 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months per year.

Containment vent 11atton system operation is conservatively defined as j

any time the 36-inch supply and/or exhaust isolation valves are open.

Similarly, Specification 3.6.2.7 states that the dryvell vent and purge i

system 24-inch (1VQ002) or the 10-inch (lvQ005) exhaust isolation valve may be open for drywell vent system operation with such operation limited to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months per year. Dryvell vent system operation is defined as any time either the 10-inch or the 24-inch inboard exhaust valve is open (concurrent with all valves of Specification 3.6.1.8 closed).

The above-mentioned valves, however, must be tested according to the ISI program which requires the stroke times to be determined every 92 days.

The testing requires the valves to be opened, one at a time, for a short period of time. Although system operation is defined as any time the noted valves are open, the purpose of the cumulative system operation time limit is to limit the release of radioactivity present in the drywell/ containment atmosphere to the environs during containment /drywell ventilation system operation under normal operating conditions.

(The system is usually used for reducing containment j

airborne radiation levels for personnel access in order to perform i

maintenance and testing.) Opening the valves for the purpose of testing does not require the system to be operated (fans-run, etc.).

Furthermore, the stroke timing performed to implement ISI requirements is performed such that the valves are stroked one at a time thus keeping the other valve in series closed. Although it was not the intent, the

.aurrent Technical Specification implies that the time the valves are 1

opened for the performance of ISI testing should be noted and taken into accounc with respect to the cumulative time allowed under the Limiting Conditions for Operation. This accountability problem was not apparently anticipated at the time the current Technical Spec 1fication was prepared.

The proposed change.would exempt the time when valves are opened to complete surveillance (stroke-time) testing required by the ISI Program (and/or Technical Specification 4.6.4.3) from the cumulative system i

operation time required to be routinely determined to ensure compliance (This is required due to the with Specification 3.6.1.8 and 3.6.2.7.

f.

conservative definition of " system operation" as discussed above.)

Existing footnotes attached to the Limiting Conditions for Operation would be revised to reflect this exemption.

(See marked-up pages attached.)

Basis.For No Significant Hazards Consideration for Change (2)

According to 10CFRSO.92, a proposed change to the license (Technical involves no significant hazards consideration if Specifications) operation of the facility in accordance with the proposed change would f

1 not (1) involve a significant increase in the probability or 4

I..

I l

to U-601048 Page 84 of 157 consequences of an accident previously evaluated; or (2) create the l

possibility of a new or different kind of accident from any accident q

previously evaluated; or (3) involve a significant reduction in a margin l

of safety.

l (1) The proposed change does not involve a significant increase in the 1

probability or consequences of an accident previously evaluated for j

)

the following reasons.

l The basis for limiting the amount of time the valves can be opened j

for drywell/ containment ventilation system operation is to limit I

the release of radioactivity to the environs during normal l

operating conditions and is thus not associated with an accident.

In consideration of this basis however, the proposed change will j

~

l not increase the amount of radioactivity released to the environs because the system will not be operated to perform the ISI stroke-

)

time tests for the applicable valves and because the drywell/

containment penetration will remain effectively closed (by at least j

one valve) during the performance of the testing.

I Additionally, with respect to containment integrity, the proposed change does not impact the containment isolation capability of the applicable valves (1VR001A, B and IVQ004A, B) and supports testing required to verify operability of these valves. It therefore does not involve a significant increase in the consequences of design j

basis accidents (DBA LOCA) associated with or requiring containment j

integrity.

(2) It has been determined that the proposed change does not create the l

possibility of a new or different kind of accident from any j

previously evaluated because no design changes are involved and no j

new modes of operation,ar,e introduced such that a different kind of l

accident could be created or have to be considered.

l l

(3) The proposed change does not involve a significant reduction in a l

margin of safety because the change does not involve any changes to setpoints or limits associated with any margin of safety assumed or l

required by a safety analysis.

Description and Justification for Change (3) i I

As noted in Appendix D of the CPS FSAR (p. D-47 6 -48), the VQOO6A, -B and VR002A, -B valves are 4-inch bypass isolation valves whf ch are I

keylocked at the handswitch in the closed position.

(Key removable in closed.pocition and kept under administrative controls.)

Under this propcsed change, a note ["(a)"] would be inserted into Table

'3.6.4-1 for valves IVR002A, -3 and IVQOO6A, -B to allow the valves to be opened under administrative control during applicable OPERATIONAL The valves must be opened during the performance of leak CONDITIONS, containment testing associated with the 36-inch supply and exhaust ventilation isolation valves every 92 days as required by Technical I

Specification 4.6.1.8.3 (The VR002A, and VQ006A, B valves actually need to be opened to support leak r

.g of the valves ;1VR003 and IVQ007) in the test lines associatt

.th the 36" valves since these to U-601048 Page 85 of 157 valves (1VR003 and IVQOO7) are part of the test boundary for the 36" valves.) It should be noted that for performance of the leak testing of the IVR003 and VQ007 valves, the 36-inch valves must be closed. That is. the 36-inch valves will be closed when the VQ006A -B and VR002A,-B valves are open so that that the penetration will remain effectively closed during the test.

The requested change is consistent with the " Note (a)" p ovisions allowed for other valves within Table 3.6.4-1 and with the allowance provided by the "*" note attached to ACTION "a" under the Limiting Condition for Operation (3.6.4).

The requirement under Specification 3.6.1.8 which states that the 36-inch supply and exhaust isolation valves must be closed when the 12-inch valves (1VR006A, B and IVR007A, B)_are open will still be imposed even though the 12-inch system will not be placed into operation when the IVR006A, B valves are opened for the purpose of leak testing.

Basis For No Significant Hazards Consideration Change-(3)

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 (1) involve a significant increase in the probability or not 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 j

of safety.

(1) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the penetration will remain effectively closed during performance of the leak test. The proposed change is requested to support leak testing which is required to verify.and ensure containment integrity. In addition, the provisions of note (a) would require administrative control of the valve. The proposed change would therefore have no significant adverse impact on containment integrity which is required to limit the release of radioactivity to the environs resulting from a design basis accident (DBA-LOCA).

(2) It has been determined that the possibility of a new or different kind of accident from any previously evaluated is not created by the proposed change. The impact of the change in mitigating the consequences of the applicable accident has been evaluated (above).

The proposed change does not involve any design changes nor does it j

introduce a new mode of operation such that another accident l

l L

scenario has to be considered.

l

)

(3) The proposed change does not involve a significant reduction in a margin of safety since the change does not involve any changes to i

setpoints or limits associated with any margin of safety assumed or l

required by a safety analysis. The proposed change would support leak testing to verify that the leak rate of the 36-inch valves is within the allowed limit.

L L-1

i l-to U-601M8

(

Page 86 of 157 CONTAINMENT SYSTEMS CONTAINMENT BUILDING VENTILATION AND PURGE SYSTEMS i

LIMITING CONDITION FOR OPERATION l

3.6.1.8 The primary containment building ventilation 36-inch supply (1VR001A, IVR0018)Dnd 36-inch exhaust (1VQ004A, IVQ0048% solation valves and the i

containment purge 12-inch supply (1VR006A, IVR0068) and 12-inch exhaust (1VR007A, IVR0078) isolation valves shall be OPERABLE, and a.

Primary containment building ventilation 36-inch supply and exhaust isolation valve (s) may be open for containment ventilation system operation % with 1

i such operation limited to $500 hours # per year for reducing airborne activity and atmosphere control for personnel safety, b.

Primary containment building ventilation 36-inch supply and exhaust isolation I

valves shall be closed when the 12-inch containment purge isolation valve (s) are open.

The 12-inch containment purg'e valves may be opened for reducing airborne activity and for atmospheric control to support containment access requirements to perform surveillance in accordance with these Technical Specifications.## When the 12-inch containment purge system is not required to support these access needs, the 12-inch valves shall be closed.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a.

With the containment building ventilation 36-inch supply and/or exhaust isolation valve (s) inoperable or open for more than 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months per year l

for containment ventilation ' system operation, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months close the open 36-inch isolation valve (s) or be in at least HOT SHUTDOWN within I

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.

With, containment purge 12-inch supply and/or exhaust isolation valve (s) in-operable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months close the inoperable valve (s) 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 .

l

These valves shall be-blocked--to--restrict them fron-opening mere than 50 I.

t ontainment ventilation system operation shall be defined as any time 36-inch I

supply *and/or exhaus,t is.olation valves are opendemer + +tud t e-t sm3 l1l w -

purw oe e. sgu. w n.

So.s.

Applicable for the period from initial fuel load to 3 months after completion j

i of the first refueling outage, otherwise a 250 hours0.00289 days 0.0694 hours 4.133598e-4 weeks 9.5125e-5 months per 365 days limit shall be imposed.

1. The 12-inch containment purge valves may be maintained open when required I

to support multiple daily access to the containment to perform required surveillance.

CLINTON - UNIT 1 3/4 6-12 i

to U-601048 Page 87 of 157 CONTAINMENT SYSTEMS CONTAINMENT BUILDING VENTILATION AND PURGE SYSTEMS LIMITING CONDITIONS FOR OPERATION (Continued) 3.6.1.8 ACTION (Continued)

With the containment purge supply and/or exhaust isolation valves with c.

resilient material seals having a measured leakage rate exceeding the limit of Surveillance Requirement 4.6.1.8.3, restore the inoperable valves to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 or in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.1.8.1 The cumulative time that the 36-inch supply and exhaust containment

. ventilation isolation valves have been open during the past 365 days for containment ventilation system operation shall be determined at least once per 7 days, g

inxer

_ 4.6.lgAt least once per 31 days, verify that each 36-inch supply and

" cere hd-xhaust containment ventilation isolation valve is blocked to restrict each valve from opening more than 50'.

At least once per 92 days each 36-inch supply and exhaust contain-4.6.1.8.3 ment ventilation isolation valve (with resilient material seals) shall be demonstrated OPERABLE by verifying that the measured rate is <0.01 La when pressurized to Pa.

)

Prior to opening the containment building ventilation system 36-inch 4.6.1.8.4 supply and/or exhaust valve (s), verify that each containment purge 12-inch supply and exhaust isolation valve is closed.

Prior to opening the 12-inch valve (s), verify that the 36-inch con-4.6.1.8.5 Once tainment building ventilation supply exhaust isolation valves are closed.

the requirement for reducing airborne activity and atmospheric control is com-pleted, the 12-inch valves shall be closed.

l e

CLINTON - UNIT 1 3/4 6-13

i fAttachment 3~

to U-601048:

Page 88 of 157

' CONTAINMENT ~ SYSTEM 5' ORYWELL' VENT AND PURGE SYSTEM LIMITING CONDITION FOR~0PERATION L

3.6.2.7 The drywell vent and purge system 24-inch supply isolatoion valves tion valves, and the 36-inch outboard isolation valve (1VQ003 M)&'6xhau (1VQ001A, IVQ001BJs, the'10-inch (1VQ005) and 24-inch (1VQ002 l

l shall be OPERABLE.

l Each 24-inch. supply isolation valve shall be sealed. closed.

'a.

b.

Eitherthe10-inch (1VQ005)orthe24-inch (1VQ002)$Iphaust. isolation-valve may.be opp for drywell vent system operation

D ith such operation limited to 5* w hours per 365 days for pressure. control.

APPLICABILITY: OPERATIO.NAL CONDITIONS 1, 2, and 3.

ACTION:

With a 24-inch drywell vent and purge supply isolation valve (s) (IVQ001A, a.

IVQ001B) open,.not sealed closed or otherwise inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months close and seal the valve (s) 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 nours.

b.

With-a 10-inch (1VQ005) or 24-inch (1VQ002) drywell vent and purge exhaust ~

isolation valve (s) inoperable or open for more than 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months per 365' days, for drywell vent system operation @, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months close the open L

10-inch and 24-inch exhaust isolation valve (s) 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 cnd in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The provisions of Specification 3.0.4 are not applicable provided the c,

affected penetration is isolated in accordance with ACTION 'a, above, and provided that the associated system, if applicable, is declared inoperable and appropriate ACTION statements for that system are performed.

SURVEILLANCE REQUIREMENTS 4.6.2.7.1 Each 24-inch drywell vent and purge supply isolation valve (1VQ001A, I

IVQ001B) shall be verified to be sealed closed at least once per 31 days.

j l

This velve(s) shall be blocked to restrict it frcm opening =cre thcn 50.

I Il

f ~rywell vent system operation shall be defined as any time either the 10-inch Dor the 24-inch inboard exhaust valves are open concurrent with all' valves of l

Specification 3.6.1.8 closed, msee cracsep

- f7pplicable for the period for initial fuel load date to 3 months after comple-L tion of'the first refueling outage, otherwise these valves should be locked I

closed.

CLINTON - UNIT 1 3/4 6-21 l

I 1

)

1 l

to U-601048 Page 89 of 157

)

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INSEPT TO

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when e,6cc of -m2 elas opnel $r inservice b+in3 pecfemed pursuant %

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of s cif a h n

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to.U-601048-Pagei 90 of.157-m

~ CONTAINMENT SYSTEMS DRYWELL VENT AND PURGE SYSTEM

~

SURVEILLANCE REQUIREMENTS (Continued) 4.6.2.7.2 ;The. cumulative time that the 10-inch and 24-inch drywell vent and-purge exhaust isolation va1ves.(1VQ005 IVQ002) have'been open during the past

'365' days;for purge ~ system operation *F s, hall be determined <at least'once.per l

7 days.

Prior to. opening a-10-inch'(IVQ005) or 24-inch drywell vent and-

.4.6.2.7.3 purge exhaust -isolation valve-(1VQ002) demonstrate that the 12-inch' containment Lcontinuous purge isolation supply' valves'and the 36-inch. containment ventila-

~

.' tion isolation supply valves of Specification 3.6.1.8 are closed.

and 4.6.2.7.4-At least once per 31 days valves'.1VQ001A, 1VQ001B, 1VQ002 L

IVQOO3 shall be verified to'be blocked to restrict valve opening' to less than l

'or equal'to 50 L _

l u'

The blocking device Valves IVQ0018 and IVQ002 are located inside the drywell.

shall be verified to be installed prior to drywell closing and during each L

COLD. SHUTDOWN except that such verification need not be performed more of ten -

l' than once pens 2 days.

l, CLINTON - UNIT 1 3/4 6-22

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' Attachment 3 f

to U401n48 Page 92 of. lS7 -;

1 CONTAINMENT SYSTEMS 8ASES~

~~

3/4.6.2.4 DRWELL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the drywell will be maintained comparable to the original design specification for the life of the unit.

A visual inspection in conjunction with Type A leakage tests is suffi--

cient to demonstrate this capability.

3/4.6.2.5 ORWELL INTERNAL PRESSURE l

l The limitations on drywell-to-containment dif ferential. pressure ensure that the -

drywell peak calculated pressure of 18.9 ps'g does not exceed the design pressure.

of 30.0 psig and that the containment peak pressure of 9.0 psig does not exceed i

the design pressure of 15.0 psig during steam line break conditions.. The maxi-

~

mum external drywell pressure differential is limited to'O.2 psid, well below the pressure at which suppression pool water will be forced over the wier wall

.!3 The' limit of 1.0 psid for initial positive drywell to end into the drywell.

containment pressure will limit the drywell pressure to 18.9 psid which is'less than the design pressure and is consistent with the safety analysis to limit drywell internal pressure.

3/4.6.2.o ORWELL AVERACE AIR TEMPERATURE The limitation on drywe'll average air, temperature ensures that peak drywell 1

temperature does not exceed the design. temperature of 330 F during LOCA condi-tions and is consistent with the safety analysis.

i 3/4.6 2.7 DRWELL VENT AND PURGE The drywell purge system must be normally maintained closed to eliminate a

. potential challenge to containment structural integrity due to a steam bypass Intermittent venting of the drywell is allowed for g

of the suppression pool pressure control during OPERATIONAL CONDITIONS 1, 2, and 3, but the cumulative time.of venting is limited to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months per 365 days. Venting of.the drywell is prohibited when the 12-inch continuous containment purge system or the

/

36-inch containment building ventilation system supply or' exhaust valves are This eliminates any resultant direct leakage path from the drywell to l

open.

the environment.

and 3, the dryw l

i l

(IVQOO2, they we bicckedp1 iso at on va vesso as not to open

[

In OPERATIONAL CONDITIONS I IV0003L-can ba opened only-

'de-T This assures that the valve ould be able to close against drywell pressure P e % N buildup resulting from a LOCA.

uphJ '

Operation of the drywell vent and purge 24-inch supply and ex y"y'3 mcy bc removed to-aMow-fuM-opening-of the vc4v%and the cumulative time l

i for vent and purge operation is unlimited.

l l

CLINTON - UNIT I B 3/4 6-5

)

to U-601048

?

Page 93 of 157 I

CONTAINMENT SYSTEMS

[

8ASES t

3/4.6.1.8 CONTAINMENT BUILDING ENTILATION AND PURGE SYSTEMS/ Y*deh d.m ts The 36-inch containment purge supply and exhaust isolation valves'ere re@ ired te-tc blocke#to restrict their opening to 50* during plant OPERATIONAL CONDI-TIONS 1, 2, and 3, since these valves have not been demonstrated capable of clos-ing from the full open position during an accident. Maintaining these valves blocked during plant operations ensures that excessive quantities of. radioactive To provide.

materials will not be released via the containment purge system.

assurance that the 36-inch valves cannot be inadvertently fully opened, they are blocked in accordance with staff's recommendations accepted in SSER 5, paragraph 6.2.4.1.

The use of the containment purge lines is restricted to the 12-inch purge supply and exhaust isolation valves since, unlike the 36-inch valves, the 12-inch valves close during accident conditions and therefore the site boundary dose guidelines of 10 CFR Part 100 would not be exceeded in the event of an accident during purg-ing operations. The design of the 12-inch purge supply and exhaust isolation valves meets the requirements of Branch Technical Position CSB 6-4, " Containment Purging During Normal Plant Operations."

The use of the 12-inch containment purge exhaust and supply lines shall be in accordance with the Clinton Power Station (CPS) " Interim Guidelines for Con-tainment Purge Operation" provided in Illinois Power (IP) Letter U-0731, dated September 10, 1984. These guidelines establish a mechanism for minimizing operation of this system as used only for purposes of reducing containment air-borne radioactivity and atmospheric control. To support these interim guidelines, a " Containment Access Management Program," in accordance with the referenced letter and IP Letter U-0716, dated June 29, 1985, is implemented to coordinate all access requirements, hereby minimizing containment occupancy times and thus This minimizing the required operation of the 12-inch containment purge system.

is in accordance with the staff's recommendations accepted in SSER 5, para-Once the 12-inch containment purge system is initiated, it will j

graph 6.2.4.1.

{

remain operating to support multiple daily containment access requirements in accordance with referenced guidelines.

Otherwise, the 12-inch containment purge j

i exhaust and supply lines will be isolated.

Continuous containment purge using the 36-inch containment building ventilation Intermittent use of system is limited to only OPERATIONAL CONDITIONS 4 and 5.th I

for the purpose of reducing airborne activity levels, or containment pressure, and atmosphere control (excluding temperature and humidity), and shall not ex-ceed 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months of use per 365 days.

1 j

Leakage integrity tests with a maximum allowable leakege rate for 36-inch supplyi and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow the opportunity for repair before gross leakage The 0.60 La leaking limit should not be exceeded when the failures develop.

8 3/4 6-3 CLINTON - UNIT 1

l I

r.

i L

to U-601048 Page 94 of 157 i

PACKAGE NUMBER 11 Section 3/4.5, EXERGENCY CORE COOLING SYSTEM, of the CPS Technical 1

Spe'ifications contains Surveillance Requirements (4.5.1.e.1 and c

4,5.1.e.4) for the Automatic Depressurization System (ADS) accumulator low pressure alarm system. However, there is no, ACTION included in the Limiting Condition for Operation section of the Specification (3.5.1) that addresses the alarm (s). The ACTIONS currently specified in Specification 3.5.1 only address inoperability of the ADS valves themselves. Inoperability of an ADS accumulator low pressure alarm system does not necessarily constitute inoperability of the associated ADS valves, especially since the Clinton as-built design includes other redundant instrumentation that can be used to monitor accumulator pressure.

i

?

A change to CPS Technical Specification 3.5.1 is therefore proposed that would include an ACTION to be taken when an ADS accumulator low pressure alarm system instrumentation channel has been declared inoperable.

(See attached marked-up page from the CPS Technical Specifications.)

Additional details and bases for No Significant Hazards Consideration j

are provided below.

System Description

As noted.in the Clinton Safety Evaluation Report (Supplement 5), the safety-grade ADS pneumatic supply is separate for the two divisions of ADS valves. The loss of air supply to one division of ADS valves will not prevent.the safe shutdown of the plant, and the failure of any one component will not result in the loss of air supply to more than one division of ADS valves. A simplified diagram of one division of the ADS accumulator-air supply system is attached.

(See Fig. 1.)

1 The following instrumentation may be noted on the drawing.

(1) Two pressure instruments are located on each divisional supply header.

(a) One instrument (1 PSIA 084 for Div I/IPSIA085 for Div II) provides a signal for a pressure alarm annunciator on panel P800* in the main control room. The pressure switch is powered from a divisional safety-related power supply.

This is the ADS accumulator low pressure alarm system instrument identified in the Technical Specifications.

It should be noted that the pressure switch also has a l

locally mounted pressure gauge that can be read from inside the containment (outside the drywell) thus providing local indication if needed.

(b) Another instrument (1PTIA078 for Div. I/IPTIA079 for Div.

II) provides direct pressure indication (LPIA078/079) on panel P601* in the main control room. This instrument is j

also powered from a divisional safety-related power

~

i supply and was installed to the requirements of Regulatory Guide 1.97.

See FSAR Fig. 7.5-1 for location of these panels.

I

)

3 Attachment to U-601048 Page 95 of 157 I

(2) A pressure switch (1B21-N542B,

-H,

-C,

-S,

-K,

-D, -E).is a

i located on each ADS valve accumulator within the drywell, Each pressure switch provides a signal to a common alarm on panel P601* in the main control room.- Although, due to their location, the switches cannot be periodically functionally tested like the Technical-Specification-related instruments, they are calibrated in accordance with the plant maintenance program and'are also used to support leak testing of the accumulator air-supply check valves. It should be noted that, for the purposes of troubleshooting, it is also possible to attach voltmeter instrument leads'to points provided within the control room cabinet containing the common alarm circuitry and determine if or which accumulator pressure switch is providing an alarm signal.

Description and Justification for Change si Although inoperability of an accumulator may constitute inoperability of the associated ADS valve, inoperability of the low pressure alarm (s) referred to in the Technical Specifications does not necessarily constitute inoperability of the associated accumulators. The as-built

' design at Clinton includes redundant instrumentation such that sufficient capability exists to determine the status of ADS accumulator pressure even with the " ADS low pressure alarm system" instrument (s) inoperable. This is the basis for the proposed ACTION.

Upon a loss of the " primary" alarm (i.e., the instrument associated with Technical Specification 3/4.5), operators can monitor ADS accumulator pressure by other means using the backup instrumentation described above. For example, ADS accumulator header pressure (the process variable monitored by the primary instrument) can be periodically verified using the indication provided by the instrument (s) described in the paragraph (1)(b) on the previous page. The locally mounted gauge described in paragraph (1)(a) could also be utilized if the failure of the primary alarm was not due to the sensor. And although the common alarm function provided by the individual accumulator switches described substitute for the in paragraph (2) is not considered an equivalent primary alarm, the system can be used to assist in determining the status of the sssociated ADS accumulator (s) once the provisions of the proposed ACTION are in effect.

The proposed ACTION vould ensure that accumulator pressure is periodically verified and evaluated using the available backup the instrumentation described above on a frequent enough basis such that measures would provide an equivalent degree of confidence in accumulator integrity as would be provided by the continuous monitoring associated On this basis, continued with the ADS low pressure alarm instrument.

plant operation ought to be allowed, at least for an extended period of time, with the primary ADS low pressure alarm instrument inoperable. The out-of-service time allowed under the proposed ACTION is consistent with this conclusion.

See FSAR Fig. 7.5-1 for location of these panels.

i 1

4 y

'S e

i m

b C

).

i*

li to U-601048 Page:96 of 157 L

.This is also.the basis for the exemption to Specification 3,0.4 to be I

included with the-ACTION.: That is.-in view of the availability of backup instrumentation for determining ADS accumulator pressure status,'

'inoperability of the primary ADS' low pressure alarm system instrument (s)

, referred to'in the' Technical Specifications should'not prohibit entry

]

into the applicable OPERATIONAL CONDITION 1, 2, or 3.-

y Basis'For No Significant Hazards Consideration'

]

L 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 prev'iously evaluated:~or.(2) create the possibility of a new or different kind of accident from any' accident q

, previously evaluated; or (3) involve a significant reduction:in.a margin-l j

.of safety.:

l Ls

' (1) The proposed change does not involve a significant' increase in the l

probability or consequences of'an accident previously evaluated-because the proposed ACTION ensures that the ADS header pressure l

will be appropriately monitored utilizing backup instrumentation.

-l that has'been'determinet to be adequate'for.this purpose.

A periodic check of the.noted backup indication may be considered.

I equivalent to a continuous monitor with alarm-only capability.

This is especially true in view of-the. fact-that the; indicator (s),

as'long as it is periodically checked, can potentially indicate a degraded condition before'the alarm setpoint would be reached on the primary instrument.- Also the periodic checks will be performed frequently enough such that the probability of-occurrence of a-gross. leakage condition. suddenly occurring between checks is j

acceptably low.

j In accordance with NUREG 0737, Item II.K.3.28, leak testing of the accumulator check valves to ensure a leak-tight system will continue'to be performed as required according to commitments--

identified in the Clinton Safety Evaluation Report.

The proposed change does not impact system design nor does it

.significantly impact system operation. No changes are made to the ADS system such that system performance as required in the Safety Analysis would be affected in responding to design basis accidents.

(2) The proposed change does not create the possibility of a new or-different kind of accident from any accident previously evaluated because no design changes or new oridifferent modes of operation are proposed for the plant. 0peration under the proposed ACTION

(' determined to be acceptable on the basis discussed statement above) does not constitute a different mode of operation since adequate monitoring of the ADS accumulator header pressure and adequate verification of instrument operability will be maintained.

the same accident scenario The ACTION proposed takes into account assumed for the design and operational requirements established for the ADS Header Pressure Alarm System.

i

I to U-601048' l

Page 97 of 157 (3) It has been determined that the proposed change does'not involve a significant reduction in a margin of safety. As noted in (1).

~ bove, ADS system operation is not significantly. impacted by the a

proposed change, and no margin of safety is impac::ed with respect

.co supplying adequate pneumatic pressure to the ADS valves.

~1 l

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- -__--____._.________j

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8

. Attachments 3 to U-601048 Page 98 of 157 I

3/4.5 EMERGENCY CORE COOLING SYSTEMS-3/4.5.1 ECCS - OPERATING q

1 LIMITING CONDITION FOR OPERATION '

3.5.1 ECCS Divisions I, II and III shall be OPERABLE with-s a.

ECCS Division I consisting of:

1.

The OPERABLE low pressure core spray (LPCS) system with.a-flow path capable of taking suction,from the' suppression pool' and transferring the water through the spray.sparger t'o the reactor vessel.

.]

2.

'The OPERABLE low pressure; coolant' injection (LPCI) subsystem "A" of the RHR system with a flow path' capable of taking Buction from the suppression pool and transferring the water to the reactor.vs.sel.

3.

Seven OPERABLE ADS valves.

i b.

ECCS Division II consisting of:

1.

The OPERABLE low pressure coolant injection (LPCI) subsystems "B" and "C" of.the RHR system, each with a flow path capab1_e of taking l

suction from the suppression pool and transferring the water to the reactor vessel.

2.

Seven OPERABLE ADS valves.

ECCS Division III consisting of'the OPERABLE high pressure core spray (HPCS) c.

system with a flow path capable of taking suction from the suppression pool and transferring the. water through the spray sparger to the reactor vessel.

OPERATIONAL CONDITIONS 1, 2**#, and 3**##

APPLICABILITY:

ACTION:

For ECCS Division I, provided that ECCS Divisions-II and III are OPERABLE:

a.

With the LPCS system inoperable, restore the inoperable LPCS system 1.

to OPERABLE status within 7 days.

The ADS is not required to be OPERABLE when reactor steam dome pressure is f

less than or equal.to 100 psig.

i

See Special Test Exception 3.10.5.

1. 0ne LPCI subsystem of the RHR system may be aligned in the shutdown cooling f.

mode when reactor vessel pressure is less than the LPCI cut-in permissive t

f' setpoint.

N. % y, w % y g,y, <,

l

)

CLINTON - UNIT 1 3/4 5-1

j to U-601048 Page 99 of 157 EMERGENCY CORE COOLING SYSTEMS ECCS - OPERATING LIMITINGCONDITIONFOROPERATION_(Continued)

)

3.5.1 ACTION (Continued):

2.

With LPCI subsystem "A" inoperable, restore.the inoperable LPCI sub-system "A".to OPERABLE status within 7 days.

3.

With the LPCS system inoperable and LPCI subsystem "A" inoperable, restore at least the inoperable LPCI subsystem "A" or the inoperable LPCS systein to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

l 4.

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 .

For ECCS Division II, provided that ECCS Divisions I and III are OPERAB'.E:

b.

1.

With either LPCI subsystem "B" or "C" inoperable, restore the inoper-able LPCI subsystem "B" or "C" to OPERABLE status within 7 days, With both LPCI subsystems "B" and "C" inoperable, restore at least 2.

the inoperable LPCI subsystem "B" or "C" to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3.

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 ",

For ECCS Division III, provided that ECCS Divisions I and II and the RCIC

)

c.

system are OPERABLE:

1 With ECCS Division III inoperable, restore the inoperable division to j

1.

OPERABLE status within 14 days.

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 2.

in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

For ECCS Divisions I and II, provided that ECCS Division III is OPERABLE:

d.

With LPCI subsystem "A" and either LPCI subsystem "B" or "C" inoper-

.1,.

able, restore at least the inoperable LPCI subsystem "A" or inoper--

i s

able LPCI subsyster B" or "C" to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

With the LPCS system inoperable and either LPCI subsystems "B" or "C" 2.

inoperable, restore at least the inoperable LPCS system or inoperable LPCI subsystem "B" or "C" to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

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 3.

in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *.

Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

CLINTON - UNIT 1 3/4 5-2 acbp s N ge pmde e,q fei c,%q.

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Attachmetic 3

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to U-601048 Page 100 of 157 t

EMiRGENCY CORE COOLING SYSTEMS _

ECCS -0PERATING LIMITING CONDITION FOR OPERATION (Co Qnued) 0 g

i 3.5.1 ACTION (Continued):

i f

/

k' For ECCS Divisions I and II, provided that ECCS Divisien III is OPERABLE l

e.

l and Divisions I and II are otherwise OPERABLE:

1.

With one of the above required A05 valves inopera5le, restore the inoperable ADS valve to OPERABLE status within 14 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 reduce reactor staam dome pressure to { 100 p;ig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i 2.

With two or more of the above required ADS valves inoperable, be in l

at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to f 100 psig wittip the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l f.

insect otruked 3.[ Coolant System, a Special Report shall be prepared and submitted t j i l

Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and tha *.otal accumulated actuation cycles to date. The current value of the usesge factor for each affected safety' l

injection nozzle shall be provided in this Special Report whenever its j

value exceeds 0.70, l

l SURVEILLANCE RE0VIREMENTS t

(

.i 4.5.1 ECCS Divisions I, II, and III shall be duonstrated OPERABLE by:

l I

Atleastonceper31daysfortheLPCS,LbI,andHPCSsystems:

,/

l a.

1.

Verifying by venting at the high point vents that the system piping

{

from the pump discharge valve to the system isolation valve is filled l

with water.

Verifying that each valve (manui;, power operated, or automatic) in 2.

the flow path that is not locked, sealed, or otherwise secured in l

J position, is in its correct

position.

Verifying that when tested pursuant to Specification 4.0.5 each:

b.

1.

LPCS pump develops a flow of at least 5010 gpm with a pump l

differential pressure greater than or equal to 276 psid.-

/

l 2.

LPCI pump develops a flow of at least 5050'gom with a pump l

differential pressure greater thm or equal to 113 psid.

{

i

/

t l

"Except that an automatic valve capable of automatic return to its ECCS p k -

tion when an ECCS signal is present may be in position for 7.nother mode o.

operation.

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1.

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6 PACKAGE NUMBER 12 j

D

,p Ifh Background / Reason for Change The NRC' issued Generic Letter 84-15, PROPOSED STAFF ACTIONS TO IMPROVE l

.AND MAINTAIN DIESEL GENERATOR RELIABILITY, which described an approach j

for enhancing diesel generator reliability. One item of the letter was l

.g.

v directed towards reducing the number of cold fast-start surveillance

-tests co prevent premature diesel engine' degradation. An example of an acceptable. Technical Specification was provided in Generic Letter 84-15 i

which would reduce the frequency of fast-start tests of diesel y

generators from ambient conditions. The. example Inserted an asterisk j

(*) in those surveillance requirements involving a diesel start and was l

accompanied by a footnote which limited the number of diesel generator

]

starts from ambient conditions to at least once per 184 days and allowed the remaining engine starts to be preceded by an engine prelube/ warmup,

,j

.p.

('

and to include gradual loading as recommended by the manufacturer.

The asterisk (*) was not inserted in CPS Technical Specification j

t 4.8.1.1.2.a.5 at the time the recommendations of the generic letter were l

1 incorporated into the CPS Technical Specifications.,This was because it was assumed that Surveillance Requirement 4.8.1.1.2.a.5 would be executed as testing was performed for other surveillance requirements i

which would be preceded with a diesel engine prelube/ warmup procedure, and gradual loading.

(For example, a diesel generator might be started 1

to satisfy Specification 4.8.1.1.2.a.4 and then allowed to run for 60 minutes to satisfy Specification 4.8.1.1.2.a.5.

This series of tests would be' preceded by a warming /prelube and gradual loading used to i

advantage for satisfying both Specifications.) It was thus decided, for maximum flexibility, that the asterisk (*) should also be inserted into Specification 4.8.1.1.2.a.5.

Description and Justification I

Under the proposed change, a superscripted asterisk (*) should be inserted just after "90 seconds" in Specification 4.8.1.1.2.a.5, Additionally, the note that is associated with the asterisk should also a

be revised because the note would now also apply to Specification 4.8.1.1.2.a.5.

This Specification requires loading and synchronization The note of the generator in less than or equal to 90 seconds.

currently only refers to the 12-second generator start (as applicable to Specification 4.8.1.1.2.a.4), and should be revised to also refer to the j

90-second loading and synchronization required by Specification j

4.8.1.1.2.a.5.

Therefore " synchronization and loading (90 sec)" should I

be inserted into the note as shown. This will bring the Specification into compliance with the Generic Letter 84-15.

I

4 t

to U-601048 Page 104 of 157 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 margin.

of safety.

(1) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluaced because the intent of the change, as prescribed by Generic Letter 84-15, is to prevent premature diesel engine degradation and increase diesel generator reliability. Improved diesel generator reliability can have significant safety benefit, The proposed change will reduce the number of diesel generator starts from ambient conditions, but it does not propose to eliminate such starts from teeting. The NRC has acknowledged that some testing to demonstrate fast start capability from ambient conditions is required for the diesel generators because the design basis (Loss of Coolant Accident with loss of offsite power) requires such a capability. The proposed change provides o cold-start test frequency that is consistent with Generic Letter 84-15.

(2) It has been determined that the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change involves no design changes nor does it introduce any new mode of cperation such that another type of basis accident would have to be censidered.

The proposed change should improve diesel generator reliability in responding to the basis accident that applies, i.e., 'LOCA with loss of offsite power.

(3) The proposed change does not involve a significant reduction in a margin of safety. No design or setpoint changes are involved that would affect a safety margin assumed in the accident analysis.

Increased reliability of the diesel generator provides an increased margin of safety in ensuring that the diesel generator will not fail on demand.

L l

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toiU-601048 g" Y'

' Page' 105 'o f 157 '

,i ELECTRICAL' POWER SYSTEMS-1 AC: SOURCES ' OPERATING

$dRVEILLANCEREQUIREMENTS

~

q t

4.8.1.1.1-Each of the abevel required independent circuits between the offsite'

. l transmission network.and the onsite Class 1E distribution system shall be:

o.

a.' -Detarmined OPERABLE'at-least once per 7 days by. verifying correct breaker 1

alignments-and indicated power availability, and-1 b.'

Demonstrated OPERABLE at least once per 18 months during shutdown by trans-l ferring, manually and automatically, unit power supply from the normal

+

circuit to the alternate circuit.

4.8.1.1.2 Each of'the above required diesel generators shall.be demonstrated OPERABLE:

a.

In accordance with-the frequency lspecified in Table 4.8.1.1.2-1 on'a~ STAG -

]j

-GERED TEST BASIS by:

1.

Verifying the fuel level in the day fuel tank.

2.-

Verifying the fuel level in the fuel storage tank.

~

3.

Verifying the fuel transfer pump starts and transfers fuel from the 4

. storage system to the' day fuel tank.-

g 4.

Verifying the diesel starts from ambient condition and accelerates to at-least 900 t.18 rpm in less than or equal to 12 seconds." The ganerator voltage and frequency shall be 4160 2 420 volts and 60 2 1.2 H2'within 12 seconds

after the start signal. Thedieselgeneratorl shall be started for. this test by.using'one of the following signals:

1 a) Manual.

b) Simulated loss of offsite power by itself.

c) Simulated loss of offsite' power in conjunction with an ESF actua-tion te.st signal.

d) An ECCS actuation test signal by itself.

5.

Verifying the diesel generator is synchronized, loaded to greater than or equal to 3869 kW for diesel generator 1A, 3875 kW for diesel generator 18 and 2200 kW for diesel generator 1C in less than or' equal to 90 seconds,* and operates with this load for at least 60 minutes.

I!

a4 syndiconipen ed to 3(soseceJs)

)

"The diesel generator start (12 seconds) from ambient conditions shall be per-

!)

p' formed at least once per 184 days in these surveillance tests. All other - - 6 an 7

l!

engine starts for the purpose of this surveillance testing may be precedede engine prelube period and/or by warmup proceduras and may also include gradual j

loading as recommended by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized.

I CLINTON - UNIT 1 3/4 8-4 L

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3 l

Attachment.3 1

-to U-601048 Page 106.of'157 4

PAC 7. AGE iCMBER 13-l b

I M'

i BacEtircund..

t i.

During the. review;ofla Plant Modification it was determined that the q

. emergency DC battery loads'_had changed. Changes-to,the' CPS low power-l operating' license Technical Specifications were rade to reflect the

"{

revised: loading.1 Subsequent review has determined,that a Division'.II-1 m

m

. load.(Fire Protection distribution panel);had not' originally beenLeaken

{

~

into account..

A

.)

I Description and Reason for Change As a' result of:the change-to:the Division II. load requirements, CPS j

?

Technical Specification 4.8.2.1.d.2.b should be revised to accurately-reflect'the 4-hour Division II battery emergency. loading profile, q

l Basis for No Significant Hazards Consideration 1

According to 10CFR50.92,.a proposed change to the license (Technical-l Specifications) involves-no' significant hazards ' consideration'if

. operation of the facility 'in accordance with.the proposed change.would (1) involve - a significant increase' in the probability or.

not consequences'of an accident'previously evaluated; or (2) create the possibility'of a new oridifferene kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin

..)

j of safety.

W (1), The proposed change does not involve a significant increase in the

"]

probability or consequences of an accident previously evaluated.for j

q the following reasons-(a)' Incorporation.of the proposed Technical Specification change

.j will provide a Technical Specification that more' accurately

']

reflects plant design. The new loading profile is_a design

.q consideration that has been taken into account for determining

.)

I adequate battery capacity that ensures the battery is capable of supporting loads essential to mitigating the consequences o

J of a design basis accident.

)

(b) It has been determined that'the change in battery load is not j

significant in comparison to the battery capacity and analysis j

has shown that the batteries are capable of supplying the new loads now and at the end of their twenty year life per.IEEE 485. A comparison of the new loads to battery capacity shows that for the 4-hour rating the bactery is at 64.9% of rated capacity. For the maximum current rating, the new load is only at 42.7% of rated. Battery testing requirements per IEEE 450 do not require retesting for every small load addition.

l:

In that this is not a substantial change in load no retesting is required and no new accident scenarie has been created.

I-l L

J

i:

b to U-601048 Page 107,of:157 (c) The change to the Technical' Specification vill ensure that the battery is tested to' the correct and more stringent acceptance cr1teri'a based on the. revised loading..(It has been

' determined that.no testing to the new. load profile is required j

at this time based on the fact that-the change in load profile is insignificant relative to the battery capacity and the fact

~

that.past testing has' demonstrated that the battery had ample capacity to support the load. profile).

i (2) The proposed change does not create the' possibility of a new or l

different kind of accident from any accident previously evaluated j

because no change to the as-built configuration of the plant is proposed for the change. The proposed change brings the Technical Specifications and testing acceptance criteria into conformance j

with plant as-built design. The impact of the proposed change has f

been assessed with respect to the design basis accident that l

applies and no new or different type of accident is applicable.

1 (3) The" proposed change does not involve a significant reduction in a margin of safety. If the difference between battery capacity and i

battery load is considered a margin of safety, it has been determined that the increase in battery load reflected by the

. proposed change is insignificant relativa to the overall battery load. ' Adequate margin has already been demonstrated by the successful performance of previous battery capacity tests.

i i

4 4

l l

l 1

L to U-601048 L

Page 108 of 157 ELECTRICAL POWER SYSTEMS i

l DC SOURCES - OPERATING J

SURVEILLANCE REQUIREMENTS (Continued) 4.8.2.1 (Continued) l b) livisionII 450 amperes for the first 60 seconds 1%7486 amperes for the next 59 minutes ico 7 N amperes for the next 180 minutes c)

Division III

> 112 amperes for the first 60 seconds E52amperesforthenext239 minutes d)

Division IV

> 127 amperes for the first 60 seconds s 117 amperes for the next 59 minutes

[44amperesforthenext180 minutes e.

At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a l

performance discharge test. Once per 60 month interval, this performance discharge test may be performed in lieu of the battery service test.

f.

At least once per 18 months, during shutdown, performance discharge tests of battery capacity shall be given to any battery that shows signs of degra-dation or has reached 85% of the service life expected for the application.

Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.

1 e

CLINTON - UNIT 1 3/4 8-14

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L to U-601048

.l Page 109 of 157-

-l PACKAGE NLHBER 14

]

Description of Change 1

This change raquest for Specification 3.9.12 is to updace the Specification to match the wording in Full Power Operating License (NPF-62) condition 2.D.(b) regarding the transfer of irradiated fuel j

from the reactor vessel.

The footnote on pages 3/4 9-19 and 3/4 9-20 contains wording from the Low Power Operating license. This wording was revised by the NRC to the wording now contained in License Condition 2.D.(b) of the Full Power Operating License.

Basis for No Significant Hazards Consideration According to 10CFR$0.92, a proposed change-to the license (Technical l

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.

This is an administrative change to reflect a change in the wording of this license condition from that in the Low Power Operating-License (NPF-55). The change does not modify the ' intent of the j

license condition. No ' previous accident analysis 1s. impacted by

{

this change.

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

This is an administrative change that does not create a new accident scenario.

(3) The proposed change does not involve a significant reduction in a margin of safety. This is an administrative change that does not affect a margin of safety.

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to U-601048

_j j t Page 110 o f 23 7. j REFUELING' OPERATIONS

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L.

3/4.9.12' INCLINED FUEL TRANSFER SYSTEM d

i LIMITING CONDITION FOR OPERATION l

3.9.12-The inclined fuel transfer' system (IFTS) may be in operation provided that:

.a.-

The access doors

of all rooms through which the transfer system penetrates I

are' closed.and locked.

b.

All access door

interlocks are OPERABLE.

y c;

The' blocking valve located in the fuel building IFTS hydraulic power unit

.is OPERABLE.**

d.-

At least one-IFTS carriage position indicator is OPERABLE at each carriage position and at least~one liquid level sensor is OPERABLE.**

Any keylock switch that provides IFTS access control-transfer system lock-e.

- out' is OPERABLE.

APPLICABILITY: When the IFTS containment blank. flange 'is removed.

-ACTION:

.With the requirements of the above specification not satisfied, suspend IFTS eperation with the _IFTS at' either terminal point'. The provisions of Specifi-cation 3.0.3 are not applicable.

SilRVEILLANCE REQUIREMENTS 4.9.12.1 Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to.the startup and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months y

during operation of the IFTS, verify that no personnel are in areas immediately

.i adjacent to the IFTS and that all access doors

to rooms through which the IFTS penetrates are closed and locked.

4.9.12.2 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the operation of IFTS and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, verify that at least one IFTS carriage position indicator is OPERABLE at'each carriage position and at least ene liquid level indicator is OPERABLE.**

'* Includes removable shields.

- The blocking valve in the fuel building IFTS hydraulic power unit and the p

l<

liquid level indicator are not required to be OPERABLE for the purposes of l

these specificationsptil af ter ftte4-4eading7-but-before-exceed +ng-5%-of 4ATCO T" R"AL POWCR sqbefor; removal of-the-reactor-pressure-vessel-head-

-aMeethe ini-t4a44r4tdsaMty,

\\pnoe to otTieadic3 temdisted h) b h reW vuset CLINTON - UNIT 1 3/4 9-19 m

l.

L ~

to U-601048

' REFUELING OPERATIONS.

Page lit of 157 INCLINED FUEL TRANSFER SYSTEM l

SURVEILLANCE REQUIREMENTS (Continued) 4.9.12.3 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the operation of the IFTS and at least once i;

per 7 days thereafter, verify that-a.

All access door

interlocks are OPERABLE.

b.

The blocking valve in the Fuel Building IFTS hydraulic power unit is OPERABLE.**

The keylock switches which provide IFTS access or control-transfer system c.

lockout are OPERABLE.

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(49 Ace-w dk" CFERA G LG"

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Includes removable shields.

- The blocking valve in the fuel building IFTS hydrauli power unit and the i

l liquid level -indicator are not required to b@BL for the purposes of these specifications gntil af ter fuel iceding, bu; bcfore eseeding SYr-ef l

RAT O TilG,"AL-POW C ar befert remova4-+f-the recetoa-ye+:urc "c::ci hed 41ter the-4ft41444-c aMty j

l pcioc s n h v,s v r 4 +,2 p g 4,, +, e,,33,j, CLINTON - UNIT 1 3/4 9-20"

2 :

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to U-601048 Page 112 of 157-PACK _AG.E NUMBER 15 Description of Change The propo%d change affects Figure 6.2.2-1 " Unit Onsite Organization."

The asterisks are used to indicate qualification requirements for various unit staff personnel. Qualifications are addressed in Specification 6.3, " Unit Str.ff Qualificat ons," and in Chapter 13 of the d

FSAR; therefor.e the arterisks should be deleted from Figure 6.2.2-1.

.l i

Additionally, the positions of Director - Plant Maintenance'and Assistant Manager - Startup have been deleted due to company i

organizational restructuring.

Basis For N6 Significant Hazards Consideration

~

According to 10CFR50.92, a proposed change to the license'(Technical Specifications) involves no significant hazards consideration if c.

operation of the facility in accordance with the proposed change would (1) involve a significant increase in the probability or not 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) favolve a significant reduction in a margin of safety.

o t

(1) The proposed change does not involve;a significant' increase in the probability or consequences of an accident previously evaluated.-

This change is an administrative change that eliminates the j

identification'of_ unit staff qualification requirements on the

)

organizational chart. The proposed change does not affect IP's commitments regarding the qualifica::1ons and training of unit staff

(

personnel.

(2) The proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. These changes are only administrative changes regarding unit staff qualification requirements. Plant' design is not affected; therefore no new accident scenario has been created.

(3) The proposed change does not involve a significant reduction in a l

margin of safety. These are administrative changes that do not impact plant design and therefore do not af fect a margin cf safety.

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l to U-601048 Page 114 of 157 PACKAGE NLHBE L16

.This Technical Specification change package contains three separate 3

.changee to page 6-7 and two changes to page 6-3.

Description of Channe (1)

This change is to Specification 6.3 " Unit Staff Qualifications". The exemption to ANBl/ANS 3.1-1978 Section 4.4.2 " Instrumentation and

]

. Control" qualification requirements for the Clinton Supervisor - Control j

.and Instrumentation should be deleted.

Mr. J. Palmer now fills the

.]

position of Supervisor - Control and Instrumentation.and fully meets the

)

i ANSI /ANS 3.1-1978 qualification requirements.

1 Basis For No Significant Hazards Consideration For Change (1)

J 1

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

This is an administrative change that eliminates an exemption to a qualification requirement. No accident analyses are affected.

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

This is an administrative change and does not affect accident analyses.

.q 3)

The proposed change does not involve a significant reduction in a margin of safety. This is an administrative change that removes an exemption to'a qualification requirement. No margin of safety is i

reduced.

1 t

Description of Change (2)

This change is to Specification 6.3 " Unit Staff Qualifications" and 6.4

{

" Training." In Specification 6.3, the last sentence should be de.leted

{

as it has been superseded by the 1987 v=.31on of 10CFR55, " Operators Licences". Additionally, the phrase "and the Radiation Protection Supervisor" should be deleted. During the licensing process, the title of this individual was changed from Radiation Protection Supervisor to Director - Radiation Protection. This change was forwarded to the NRC but deletion of the old title was overlooked. Specification 6.4 should be modified as indicated on the attached marked page to delete material l

that has also been superseded by the 1987 version of 10CFR55.

I

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to U-601048 1

Page 115 of 157

.The requirements of the March 28, 1983 letter, were incorporated as appropriate by the NRC into the 1987 version of 10CFR55.

j Description of Change (3)

I This change is to Specification 6.4 and to Figure 6.2.1-1.

In January C

1988, the' position of Director - Nuclear Training will be upgraded to

)

Manager.- Nucicar Training..The two referenced Specifications should be j

. updated acecrdingly.

Additionally, the word " operations!' under the Manager - Scheduling and.

l Outage Management should be replaced with " staff"'in order to be consistent with the rest of the organi::ation. chart, i

Basis For No Significant Hazards Consideration For Changes (2) and (3)

According to 10CFR50.92, a proposed change to the license (Technical Specification) involves'no significant hazards considerate'on if operation >>f 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 3

previously evaluated; cc (3) involve a significant reduction in a margin-7 of safety.

(1h The proposed change does not involve a significant increase in the possibility or consequences'of an accident previously evaluated because these changes are merely administrative in nature and do' not affect any accident analyses.

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

-l these changes are administrative and do not affect plant design.

(3) The proposed change does not involve a significant reduction in a f

margin of safety because these are administrative enhancements to operation and do not af fect the margin of safety.

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Attachment d' to U-601048-Page.117'of 157L i

ADMINISTRATIVE' CONTROLS' f

.3' UNIT STAFF QUALIFICATIONS' 6.3.1 Eachmember'oftheunit'5taffshallmeetorexceedthe' minimum.qualifi-

. cations of ANSI /ANS 3.1-19784*fexcept for the Director-Plant Radiation' Protec- ~

s tion and-the-Rad 4et4en4roteet4an-Superv4+oe who shall. meet or exceed the.quali-ficationsLof Regulatory Guide 1.8, September 1975.. The 'icenseri-Operator: cad-Senior Operators shall. c.lso sect or cxceed the minimum qualific+t4cn; Of-the

-!trch 28, 1980 " C letter tel:1' 'icen:000.

6.4 : TRAINING g

N 6.4.1 A retraining and replacement trainingp rogram'for the' unit staff shall be maintained under the direction of the' Oirector-Nuclear Trainingd hall. meet or exceed the' requirements and-recc=cadations of Secti;n S.5 cf /SSI/AFr

&-14978--and Appendix "AL of 10 CFR Part 55.end-the*upp4ementei-requ+rements-speciflad in Sections-A-and-C-of-Enc 40sure-1+f-the-March 2C,1980-NRC letter te all licenseer, and-sha44-4nc4ude f=i4faeirat4en Oith reicvant-4ndusey cperatiron:1' experiences

' 6.' 5 REVIEW AND AUDIT 6.5.1 FACILITY REVIEW-GROUP (FRG) 1

'l

FUNCTION i

6.5.1.1' The FRG shall function to advise the Manager-Clinton Power Station on -

y all matters 'related to nuclear safety.

COMPOSITION 6.5.1.2 The FRG shall be composed of the:

Chairman:

Assistant Manager-Clinton Power Station l

Member:

. Assistant Manager-Plant Maintenance Memberi Assistant Manager-Plant Operation.

Member:

Director Plant Technical 3

Member:

Supervisor-C&I Maintenance Member:

Director Plant Radiation Protection Member:

Supervisor-Huclear Member:

Supervisor-Chemistry a.

A N

h

.A, 7

i

- **The staff reported its conditional acceptance of the applicant's request for exception to ANSI /ANS 3.1-1978 qualification requirements for the Clinton j

N Jupervisor of Control and Instrumentation in NUREG-0853, Supplement No. {/

j CLtNTON - UNIT 1 6-7 j7e I

l

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to U-601048 Page 118 of 157 PACKAGE NUMBER 17 Description of Chaege

'This proposed change to Technical Specification 6.8.2 and 6.8.3.c.would clarify the requirements for approval of those procedure: listed in Specification 6.8.1.

Specification 6.8.2 currently states that each procedure of Specification 6.8.1, and changes thereto (including temporary changes) shall be " approved by the Power Plant Manager."

However, not all of the procedures that meet the Specification 6.8.1 critieria are Plant Staff procedures. Such procedures would be associated with support organizations or programs for which managers other than the Power Plant Manager are directly responsible. But because the Power Plant Manager has overall responsibility for operation of the facility (as stated in the FSAR), his signature for concurrence, as a minimum, would still be required. The approval signature, however, would'be the responsible manager's.

IP believes this would meet the intent of the existing Specification.

However, to ensure conformance with the exact words of Specifications 6.8.2 and 6.8.3.c. a change to thz wording is proposed in support of the above described and proposed approval process. Specifically, the proposed change would replace the words " Power Plant Manager" where they appear in. Specifications 6.8.2 and 6.8.3.c vi'th " appropriate responsible manager".

Basis for No Significant Hazards Consideration According to'10CFR50.92, a proposed change to the license (Technical Specifications) involves no s1gnificant hazards consideration if-operation of the facility in accordance with the proposed change would (1) involve a significant increase in the probability or not consequences of an accident previously evaluated: or (2) create the possibility of a new or dif ferent kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

increase in the The proposed change does not involve a significant (1) probability or' consequences of an accident previously evaluated because this is an administrative change clarifying procedure approval. No accident analyses are impacted.

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

This is an administrative clarification of procedure approval and does not affect plant design. Therefore, the possibility of a new type of accident has not been created.

reduction in a I

(3) The proposed change does not involve a significant margin of safety. Administrative : clarification of procedure approval does not affect a margin of safety.

II

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Page 119'of 117.

=g

! ADMINISTRATIVE CONTROLS i

n l

,-6.8 DPROCEDURES ANO PROGRAMS,

't

-PROCEDURES 6.8.1' Written procedures.shall be established, implemented, and maintained

, covering the activities referenced below:

The applicable procedures recommended in Appendix A of Regulatory i

a.

. Guide 1.33,-Revision.2, February 1978.

,The' applicable procedures required to implement the requirements of NUREG-0737 b.

andsuppjementthereto.

c.

Ref uelitig ' operations.

I d.

Surveillance and test activities of safety-related equipment, j.

Security Plan implementation.

-e.

L, f.

Emergancy Plan implementation, Fire Protection Program imp' lamentation.

.l g.

j h.

PROCESS CONTROL PROGRAM implementation.

OFFSITE-OSE CALCULATION MANUAL implementation.

0 i.

j.

Quality Assurance Prcgram for. effluent and environmental monitoring.

l REVIEW AND AFPROVAL Each procedure of Specification 6.8.1, and changes thereto, shall'be i

6.8.2 jl Q._. reviewed in accordanca'with 6.5'1.6 and 6.5.3 as appropriate an r

ically as set forth in administrative procedures,

.m

r. -

i.

-.PORARY CHANGES-TEM - - -

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

The intent of.the original procedure. is fiot altered; j

a.

The change is approved by two members of the unit management staff,, at i

b.

least one of whom holds a Shnior Operater license on the unit affected; and I,

The change is documented, reviewed in accordance with 6.5.1.6 and 6.5.3 c.

as appropriate, snd approved by the Power-Plant Mencges within 14 days of

.l D

implementation.

ggns resgus% wqtr i

CLINTON - UNIT 1 6-15 L__3 L _ _

l to U-601048 l

Page 120 of 157 PACKAGE NUMBER 18

.j i

j Deceription of Change

)

This proposed change is to. Specification 6.9.1.8, " Monthly Operating Reports." The 1987 update to 10CFR50.4 "Uritten Communications" changed the address to which reports are sent.

Basis For No Significant Hazards Consideration According to 10CFR50.92, a proposed change to the license (Technical I

Specifications)' involves no significant harards consideration if operation of the facility in accordance with the proposed change would not1(1) involve a significant increase in the probability or j~

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.

This is strictly an administrative change and does not impact 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 The changing of an address where monthly repo.*ts are sent has no i

impact on accident analyses.

(3) The proposed change does not involve a significant rseduction in a margin of safety. This is en administrative change that does not affect a margin of safety a

l 1

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s tWi-

q to U-601048 Page 121 of 157 ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued)

The Semiannual Radioactive Effluent Release Reports shall irclude a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radio-active materials in gaseous and liquid effluents made during the reporting period.

The Semiannual Radioactive Ef fluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), pursuant to Specifications 6.13 and 6.14, i

i respectively, as well as any major changes to liquid,* gaseous, or Solid Radwarte Treatment Systems pursuant to Specification 6.15.

It will also include a list-ing of new locations for dose calculations and/or environmental monitoring iden-tified by the land use census pursuant to Specification 3.12.2.

MONTHLY OPERATING REPORTS

/ BocumedCchel hsk 6,9.1.8 Routine reports of operating statistics and shutdow experience, including documentation of all challenges to the main steam ystem safety /.

relief valves, shall be submitted on a monthly basis to the c40@ffice

~,

of ",ezcuree-Management, U.S. Nuclear Regulatory Commission, Washington, D.C.

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

(

SPECIAL REPORTS 6.9.2 Special reports shall ue submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.2 The following records shall be retained for at least 5 years:

Records and logs of unit operation covering time interval at each power a.

level.

b.

Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety.

c.

All REPORTABLf. EVENTS.

Records of surveillance activities, inspections, and calibrations required d.

by these Technical Specifications and the Fire Protection Program.

CLINTON - UNIT 1 6-21 o-

s Attacnment 3 to U-601048 Page 122 of 157.

> PACKAGE NUMBER 19 Justification and Description of Change CPS has proposed a plant modification.to add test connections upstream of certain excess flow check valves (1CM002B, ISM 008, ISM 011, 1E22-F332, 1E51-F377B) to facilitate testing of these valves as required by Technical Specification 4.6.4 4.

The new test connections to be added will themselves contain isolation valves which must be added.to the Test Connections, Vents and Drains section of Table 3.6.4-1 in the CPS Technical Specifications. Marked-up pages from the Technical Specifications are attached.

In order to ensure compliance with the Technical Specifications (once-the amendment is approved) and yet provide some flexibility in the schedule for completing the modification, IP proposes that notes be included on the applicable pages allowing the proposed Technical Specification change to become effective once the modification is complete. At that time, the OPERABILITY requirements and provisions of the Technical Specifications shall be in effect for the subject.

valves.

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

-l previously evaluated; or (3) involve a significant reduction. in a margin of safety.

The proposed change does not involve a significant increase in the (1) l probability or consequences of an accident previously evaluated for the following reasons, a)

With respect to maintaining containment integrity, the l

addition of the test connections supports testing required to verify the capability of the excess flow check valves to check flow at a particular differential pressure assumed for accident conditions, b)

The added test connections are double-valved, capped, and meet the applicable design / safety requirements to ensure containment integrity.

l Adding the valves to Table 3.6.4-1 also makes all of the

^

c) applicable Technical Specification requirements associated with containment integrity applicable to the new valves as well.

d)

Finally, the addition of the test connections does not affect l

the functional characteristics of the excess flow check valves and therefore does not affect operation of the associated systems.

l 1-

l. ]

[

to U-601048 Page 123 of 157 (2) It has been determined that the proposed change does not create the possibility-of new or different kind of accident from any accident previously evaluated. It was determined:that the addition of the test connections should be evaluated with respect to containment integrity and the operability of the excess flow check valves including their impact on the associated systems (Containment Monitoring, Suppression Pool Makeup,.High Pressure Core Spray and Reactor Core Isolation Cooling).. This impact hasLa1 ready been evaluated, and no further consideration of a new or different kind of accident is required.

(3) The proposed change does not involve a significant reduction in a

, (

margin of safety because the proposed change does not invclve,a change co any trip setpoints, analytical values, or design limits required or assumed in any safety analysis.

)

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Ateachment-3 to U-601048 Page 124 of 15-

~ CONTAINMENT SYSTEMS 3/4.6.4 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.4 The containment isolation valves and the instrumentation line excess flowcheckvalvesshowninTable3.6.4-1shallbeOPERABLE{withisolationtimes less than or equal.to those shown in. Table 3.6.4-1

' APPLICABILITY: As shown in Table 3.6.4-1.-

ACTION:

4 With one or more of the containment isolation valves shown in Table 3.6.4-1

]

a.

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 I

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.

OtherwisebeinatleastHOTSHbTDOWNwithinthenext12 hours'andinCOLD

-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 a

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.

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 d'uring CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

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

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TABLE 3.6.4-1 (Continued) j i

CONTAINMENT ISOLATION VALVES l

k s

TABLE NOTATIONS (a) Hay be opened on an intermittent basis under administrative control.

l (b) Excess flow check valve actuation differential pressure.

't 1

(c) Isolation valving for instrument lines which penetrate the containment con-form to the requirements of NRC Regulatory Guide 1.11.

The in-service in-spection program will provide assurance of the operability and integrity j

of these-isolation provisions.. Type "C" testing will not be parforced on the instrument line isolation valves. The instrument lines will be within I

the boundaries of the Type "A" test, open to the media (containment atmo-l sphere or suppression pool water) to which they will be exposed under pos~

f i

tulated accident conditions.

Instrument taps from the process line located betwecn the process isolation valves and the penetration, and not them-selves penetrating containment, will be Type "A" and/or "C" tested along

-l with the process line isolation valves.

l i

(d) Excess flow check valve.

i i

(e) The RHR tystem may be operating in the shutdown cooling mode during the Type A test. These valves are tested using water but the results are not i

required to be added to the Type A test results. The LPCS, HPCS, and RHR

-may be aligned in the normal standby or injection mode during the Type A test. This will expose the closed loop outside containment to containment

-l pressure through the suppression pool. This is the closest valve alignment to the post-LOCA alignment possible.

Type C water test results on these suction valves will not be added to the Type A test results.

(f) Valves shall-be closed in accordance with SECONDARY CONTAINMENT INTEGRITY.

s (g) Valves shall be " sealed closed" by utilizing mechanical devices to seal or lock the valve closed or to prevent power from being supplied to the valve operator.

l When handling irradiated fuel in secondary containment and during CORE ALTERATIONS and operations with a potenti:.1 for draining the reactor vessel.

Isolates on RCIC low steam line pressure only, t

Isolation signal descriptions are provided in Table 3.6.4-2.

For test pressure = 9.0 psig, the valve (s) shall be pressurized using air l

or nitrogen, and for test pressure = 9.9 psig, the valve (s) shall be l

4 pressurized using water.

With any control rod withdrawn. Not applicable to any control rods

(

removed per Specification 3.9.10.1 or 3.9.10.2.

I

\\ (b) d these valves is not required until comp ehon of correspenclinf l

l OPERAB!UTY plant moefication.

CLINTON - UNIT 1 3/4 6-61 x

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to U-601048 i

Page 128 of 157 f

PACKAGE NUMBER 20 y

Description of Change l

l Under the proposed change a provisional footnote (*) will be added to the'0PERABILITY requirements.specified in the Limiting Condition for

. Operation for the MSIV Leakage Control System (MSIV-LCS), CPS Technical 3

i Specification 3.6.1.4 (p. 3/4 6-7).

The note is applicable only to the I

MSIV-LCS instrumentation, and it will state, "An MSIV leakage control i

. system instrumentation channel may be placed jn 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 channel in I;

the tripped condition provided the otbu channel or channels monitoring that parameter are OPERABLE."

(

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1 i

_ Justification.

As noted in FSAR section 6.7.1.I the MSIV-LCS is darigned to be capable of performing its safety function following a LOCA and an assumed single

)

v 7 active failure. Section 7.3.2.3.1 notes that the MSIV-LCS has redundant.

and separate instrumentation and controls to ensuke that the system will be able to maintain its functional capability a guming a single active failure._'The use of different divisional power'for each subsystem ensures that the system will be' operable upon loss of a single power

' 'cq division. In addition, the FSAR notes that the system is dojkgned to l

conformtoRegulatoryGuide1.22andcanthusbefunctionaljt rested during reactor operation. In accordance with IEEE 279, th6 sensors can be checked one at a time by, application of simuleted signals during/

normal plant operations. Channel independence as well as sufficient g

separation between subsystems is also ensured by conformance to thisj standard..Thus the instrumentation configuration is such that ringle failure criteria have been app. lied to its design and such that testing performed during plant operation on one subsystem will not affect the operability of the other subsystem.

s.

The NRC has acknowledged that single f ailure criteria may / e, violated 6

]

duringtheperformanceoftestingforshortperiodsoftime.krhir j

i short-term exception is acknowledged in Pircraph 4.11 of,IEEE 279.

4 Accordingly, provisions for' having a channal(s) under test fur. shore periods of time (typically 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) without having to enter an ACTION statement appear throughout the instrumentation OPERABILITY requirements in the instrumentation section of the Technical Specifications as appropriate. The propoced change is consistent with those provisicus.

Basis For No Significant hzaras Consideration-

,e, According to 10CFR50.92, a proposed change to the license (Technical i

Specifications).ir.volves no significant hazards consideration if

)

operation of the:?acility in accordance with the proposed change would (1) involve a significant increase in the probability or not consequences of'an 2ccident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident I

previously evaluateds or (3) involve a significeit re, duction in a margin

/ (

of safety.

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s' (1) ' The proposed c71nge does not invgive a significant increa d inlihe

.e

{ '

-l probabilithod knsequences of an accident previously evaluated j

4 because +

. Ismig.e to the OPERABILITY requirements f or the MSIV-LCS 1

. innram$n@tatib +.hs negligible impact on.the availability of the

{

I system based onk f.k _ preta'ise, as stated in IEEE 279, that the time q

allowed for the ' channel. under test to be inoperable is so short.

that the probability of f ailure of the active channel (s) would b'gl commensuratejui,ththeprobabilityoffailureofallredundantBecaurh/

k.

dyc em cha,nhels duritrgthe normal interval between tests.

the propostI change he no significant impact on the design or

't q

opp AEILIT*'of the lystem, the system will remain fully capable-q? l i

j y performing.tsfunctdbntomitigatetheconsequencesofthe g

f

, accident et of v'dich j t' was. designed.

(The MSIV-LCS is designed to pinimi s fie' release of fika, ion products which could bypass.the

/

seat leakage through the MSIVs g /j(tandby gm Trea'hput syst!em tip\\

i p sh ter the pg stplated LOCA.)

J,

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-,t

,q lI (2)'- The. proposed ebphge does not create the poss1.bility of a new or different kind of accident from any previouM y evaluated because it i

,)

.does not chaTja the design of the facility Ed it does not apply to l

f'

/

any system other than the MSIV-I.CS.

(As noted above, the propos,td change does not significantly impact the OPERABILITY of the 5

MSIV-LCS.)~

\\-

/

(3) The proposed change des not involve a significant reduction in the margin ofvafety bedg se the actuation instrumentation trip setpointsar.noYaffacted}"thechangeandbecauseofthe l

f ollowing.,

,I j

j

/

Under the current ACTION statement, an inoperable MSIV-LCS If subsystem must be returned tp 0FERABLE status within 30 days.

the plant must be in. ET SUUTDOWN within the next 312 hours0.00361 days 0.0867 hours 5.15873e-4 weeks 1.18716e-4 months and

not, COLD SHUTDOWN within the'following 24, hours. Thus, tile time j

'{

-allowed for the channel to be dnoperable for testing (subject to the provisions of tha note) is bli A.ficant relative to the ot[t-of-service ' time.,ellowed for,the/

s snbsystem under the ACTION s12tament.

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Page 13n of 157 CONTAINMEtiT - SYSTEMS NSIV LEAKA$f CONTROL SYSTEM u

L li 4

LIMITINGCONDITIONFOROPENATION' v,

3.6.1.4 Two in spcndent HH V leakage control system (LCS) subsystems shall be l

4 OPERABLE?.4)

APPLICABRITY: OPERATIONAL CONDITIGHS 1, 2, and 3.

e>3 i g u",

\\

ACTION:

4,/

)

With one MSN leakage cchtrol system subsystem inoperable, restore.the inoper-1 able. subsystem 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 following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ;

'\\

SURVEILLANCE REQUIREMENTS 4.6.1.4 Each MSIV leakage control system subsystem shall be demonstrated OPERABLE:

At leart once per 31 days by verifying:

a.

1.

Blower OPERABILITY by starting the blowers from the control room and operating.the blowers for at least 15 minutes.

2.

Heater OPERABILITY by demonstrating electrical continuity of the beating element circuitry.

I b.

During each COLD SHUTDOWN, if not performed within the previous 92 days, ;

i by'<:ycling each remote, manual and automatic motor operated valve through 1-at least one complete cycle of full trevel.

c.

At least once per 18 months by:

o

-Performanceofafunctionaltestwhichincludessimulatidactuation 1.

of the subsystem throughout its operating sequence, and verifying

that a ch automatic valve actuates to its correct position, the

' blower (s) start.

2.

Verifying that the blowers develop at least the required vacuum at the rated c3pacity and each heater unit draws 7.4 te 5.2 amperes per phase. <

v' a)

Inboard subsystem, 15" H O vacuum at g 100 scfm.

2 ti)

Outboard subsystem,15" H O vacuum at 1200 scfm.

2 s

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$6 NN { ta36. G in m i n o p v.cd4c s tO% be up to 2,. how cs ke r9f ed Sucye; % cc w>%eweph y A_ da.3nd M og J Q ed gf. We 6 proped all CNc chun eh w Nep.q q p.n er g( Q@MAgg CLINTON - UNIT 1 3/4 6-7

1 "rl,(

to U-601048 j

Page 131 of 157 1

,00NTAINMENT SYSTEMS MSIV LEAVAGE CONTROL SYSTEM l

SURVEILLANCE REQUIREMENTS (Continued) 4.6.1.4 (Continued) d.

By verifying the flow, pressure, temperature and level' instrumentation to be OPERABLE by performance of a:

i 1.

CHANNEL FUNCTIONAL TEST at least once per 31 days, and a 2.

CHANNEL CALIBRATION at least once per 18 months.

9.

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k CLINTON - UNIT 1 3/4 6-8 i

l

to U-601048 Page 132 of 157 PACKAGE NUM3ER 21 Description and Justification for Change l

Specification 4.0.4 states.that entry into an OPERATIONAL CONDITION or other specified applicable condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting, Condition for 1

Operation have been performed within the applicable surveillance interval or as otherwice specified.

Specification 4.11.2.7.2 requires the radioactivity rate of nobie gases from the off-gas recombiner effluent to be determined (to be within the limits specified in the Limiting Condition for Operation) at two specified frequencies:

(1) At least once per 31 days, and (2) Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase of 30% in'the indicated nominal steady-state fission gas release from the primary coolant (with certain provisions).

The APPLICABILITY of this Specification (i.e., the applicable OPERATIONAL CONDITION) is "during operation of the main condenser air ejector."

Although it is readily apparent that Surveillance 4.11.2.7.2 (associated with the Limiting Condition for Operation for Specification 3.11.7.2) cannot be performed until after entering the special applicable OPERATIONAL CONDITION, an exemption to Specification 4.0.4 should be inserted as shown (see marked-up page 3/4 11-17) to ensure compatibility between Specifications 4.0.4 and 4.11.2.7.2.

The exemption formally allows the plant to enter the applicable OPERATIONAL CONDITION without having first performed the required surveillance.

The proposed change is consistent with the specification approved for the Perry Nuclear Power Station.

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 margin of safety.

1)

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

This is an administrative change to achieve consistency in the format and structure of the Technical Specifications involved. No accident analyses are affected.

2)

The preposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

This is an administrative formality and does not create a new accident scenario or affect plant design.

m, 1

to U-601048 Page 133 of 157 3)

The proposed change does not involve a significant reduction in a margin of safety. This is an administrative change that is-consistent with the Specifications involved. No margin of safety is affected. The Surveillance Requ1rements will continue to be satisfied in the same' manner.

i e

1 i

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At ta climen t 3 to U-601048

- Page 134 of 157 '

RA010 ACTIVE EFFLUENTS MAIN CONDENSER LIMITING CONDITION FOR OPERATION l

i The radioactivity rate of noble gases measured at the offgas recom-3.11.2.7 biner effluent shall be limited to less than or equal to 289 millicuries /sec after 30 minutes' decay.

APPLICABILITY:

During operation of the main condenser air ejector.

ACTION With the radioactivity rate of noble gases at the offgas recombiner effluent l

exceeding 289 millicur.ies per second after 30 minutes decay, restore the gross l

radioactivity rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 .

l l

SURVEILLANCE REQUIREMENTS

' 11.2.7.1 The radioactivity rate of noble gases at the offgas recombiner effluent shall be continuously monitored.

.]

{

The radioactivity rate of noble gases from the offgas recombiner 4.11.2.7.2 l

effluent shall be determined to be within the limits of Specification 3.11.2.7 l

l at the following frequencies *by performing an isotopic analysis of a represen-tative sample of gases taken at the discharge (prior to dilution and/or dis-charge) of the of fgas recombiner:

l At least once per 31 days.

l Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase, as indicated by the Retreatment a.

b.

Off-Gas process radiation monitor listed in Table 3.3.7.1-1, of greater than 50%, after factoring out increases due to changes in THERMAL POWER level, in the nominal steady state fission gas release from the primary coolant.

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l 3/4 11-17 CLINTON - UNIT 1 l

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_______._a

to U-601048 Page 135 of 157' 1

PACKAGE NDfBER 22 Description and Justification of Change 1

Two changes are proposed to revise the wording appearing in ACTION 70 on i

p. 3/4 3-72 of the CPS Technical Specifications.

l I

The first change applies to part "a" of the ACTION which states, "with

)

one of the required monitors inoperable, place the inoperable channel in the (downscale) tripped condition within I hour..."

The word l

"downscale" and the associated parenthesis should be deleted because f

they do not add meaning to the intent of the ACTION. The intent is to l

place the monitor in the same tripped state that would result if the l

monitor actually sensed high radioactivity at the Main Control Room air intake. The monitor is designed to trip in response to a number of conditions including the downscale condition. A trip therefore can be effected a number of ways including disconnecting the detector (which is effectively equivalent to a downscale trip) or interrupting power to the monitor. The Technical Specification does not need to specify how the monitor should be tripped. Clinton should be allowed to place the moniMr in the tripped condition as required and in an appropriate j

manner that will maintain it tripped (as long as it is required) with or 1

without the words "(downscale)" appearing in this Technical l

Specification. To avoid possible confusion, IP recommends deleting

"(downscale)".

The second change applies to part "b" of ACTION 70 and concerns use of the term " recirculation" in describing the mode of operation initiated by an air intake high radiation condition. The proposed change would make the wording consistent with the rest of the Technical Specifications by removing Standard Technical Specification phraseology and inserting plant specific wording. The term " recirculation" is misleading because several of the modes of operation specified in Technical Specification 3/4.7.2 involve a recirculation path. The high radiation mode, as referred to in Specification 4.7.2.e.5, is the mode of operation initiated by a high radiation condition sensed by the air intake radiation monitors. Therefore, " recirculation" should be replaced by "high radiation" to be more precise and avoid confusion.

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 (1) involve a significant increase in the probability or not 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.

involve a significant increase in the The proposed change does not 1) probability or consequences of an accident previously evaluated.

This is an administrative change that clarifies the Specification.

The clarification does not change the intent of the Specification.

u

j

,co U-601048'

.Page-136'of 157 '

i 2 )'

The proposed change does not-create the possibility of a new'or different kind of accident from any accident previously evaluated.

This is an administrative clarification that does not affect plant

~ design or accident analyses.

3)

' The proposed change does not. involve a significant reduction in a

~

margin of safety.~ This is an administrative clarification that l

does not= affect a margin of~ safety.

H l

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I i

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to U-601048 i

Page 137 of 157 !

j t

TABL_E3.3_.].2-1'(Continued)

RADIATION MONITORING INSTRUMENTATION TABLE NOTATIONS i

When irradiated fuel is being handled in the secondary containment.

j Alarm only.

During operation of the main condenser air ejector.

With fuel in the new fuel storage vault.

1. With irradiated fuel in the spent fuel storage pool.

i t Reactivity concentration expected at the monitor location is a noble gas mix with a 2.9 minute decay.

l tt Radioactivity concentration expected at the monitor location is a noble gas mix released from the off-gas treatment System.

t (a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for re-j quired surveillance without placing the trip system in the tripped condi-i tion provided at least one other OPERABLE channel in the same trip system f

is monitoring that parameter.

ACTION Witn one of the required monitors inoperable, place the ACTION 70 -

a.

inoperable channel in the %wnscsitt tripped condition I

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; restore the inoperable channel to OPERABLE i

status within 7 days, or, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , initiate f

and maintain operation of the control room emergency l

filtration system in the(recirculationfiEode of operation, i

o par.n.

I b.

With both of the required monitors inoperable, initiate and maintain operation of the control rcom emergency filtration system in the (recirculationjYode of operation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

i

% c AA,

ACTION 71 -

With the required munitor inoperable, perform area surveys of the monitored area with portable monitoring instrumentation at

)

J least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 72 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requ1rement, gases from the esin con-denser off gas treatment system may be released to environment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided:

The off gas treatment system is not bypassed, and b.

The post-treatment air ejector off gas PRM high range noble f

a.

gas activity monitor is OPERABLE.

With the number of channels OPERABLE less than required by the ACTION 73 -

Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are.taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed for gross noble gas activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the number of channels OPERABLE less than required by the ACTION 74 -

Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is esticated

~

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

CLINTON - UNIT 1 3/4 3-72 l

m-i t

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1 I

e to U-601048 Page 138 of 157 PACKAGE NUMEER 23 Description and Justification The' purpose of this proposed change to the CPS Technical Specifications l

1s to simplify the CHANNEL' FUNCTIONAL TEST requirements for the Low and' High Power Setpoint (turbine first-stage pressure channels) associated L

with the Rod Pattern Control System (RPCS) in Table 4.3.6-1 of the CPS -

Technical Specifications. The proposed change would delete notes "(d)"

as well'as note "(c)" and the corresponding "D" such that only and 'g)"M" would remain under the CHANNEL FUNCTIONAL TEST column for "S/U both the low and high power setpoint functions. Justification fcr these changes is provided below, i

Test requirements associated with the RPCS high and low power setpoints are identified not only in Specification 4.3.6, but also in Specification 4.1.4.2. The changes proposed for the CHANNEL FUNCTIONAL TEST requirements specified in Table 4.3.6-1 help'to resolve the scope of testing that should be performed under the two Specifications with no loss in adequacy;of the overall testing that is intended.

The high and low power trip inputs to the RPCS are derived from turbine first-stage. pressure. These channels may thus be regarded as turbine first-stage pressure instrumentation channels. The scope of the CEANNEL FUNCTIONAL TEST performed under Specification 4.3.6 (Table 4.3.6-1) for each of these channels is determined by the definition of a CHANNEL FUNCTIONAL TEST as defined by the Technical Specifications (DEFINITION 1.6).

The CHANNEL FUNCTIONAL TEST, as performed to verify OPERABILITY of the turbine first-stage pressure instrumentation, however, should not be confused with nor performed at the same frequency as the testing performed to verify proper operation of the RPCS with respect to the low and high power setpoints. C,onfusion results from the fact that a t rbine first-stage pressure trip module (s) has to be tripped in order uto test operation of the RFCS enforced at the low or high power setpoint. The trips will only be in place when plant conditions are such that the trip setpoints are actually exceeded or during the performance of a CHANNEL FUNCTIONAL TEST of the turbine first-stage Although the turbine first-stage pressure low pressure instrumentation.

and high power trips are required and verified as part of the testing performed to verify OPERABILITY of the RFCS in accordance with Specification 4.1.4.2, it is inappropriate to identify this testing as testing performed to satisfy CHANNEL FUNCTIONAL TEST requirements specified within Table 4.3.6-1, especially when the CHANNEL FUNCTIONAL TEST is defined as stated. For this reason and the reason discussed below, there is no need to make the CHANNEL TUNCTIONAL TEST frequencies for the turbine first-stage pressure instruments agree with the frequencies specified for the testing perfarmed for the RFCS nnder

{

Specification 4.1.4.2.

k

.At Clinton, the trip' modules associated with the turbine first-stage pressure instrumentation fall within the test scope of the Self-Test j

System (because they are located within the Nuclear System Protection l

(See j

System [NSPS}) and are therefore continuously monitored / tested.

FSAR Section 7.2.1.)

A high degree of reliability may therefore be

{

It is not j

c.ssumed for these inputs to the rod pattern control system.

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Page 139 of 157-3 0

-ne'cessary to prove:that thb RPCS is' "seeing" these inpbts every time the'

,)

turbine: first-stage pressure channels are tested.

It may be assumed

~

that'ence a CHANNEL FUNCTIONAL TEST is performed on the-turbine-E,ll first-atage pressure' instrumentation _just prior to startup (in

'accordance wich note."(b)".of Table'4 2 6-1] the high and low power trip t

P*

channels will be'0PERABLE astinputs to the Rod. Pattern.Concrol System'so that the CHANNEL' FUNCTIONAL TEST for.these channels may theh be) 4 CW performed at,the normal ~once-per-31-day frequency.: Verification that 1

the RPCS properly uinhibits or restricts control' rod movement. verifies l

that the.RPCS is,"seeing" theseichannels/ trips as.-it is tested according-

i h

S.

(,yJ to, Specification 4.1;4' 2 during or prior, to thesvarious conditions

-listed within that Specification. Such testing should not require j

overtesting the turbine first-stage pressure trip modules since those

{

@h trip signals are already adequately tested and monitored as described-above.

p Note "(c)", as attached to."D" in the CHANNEL FUNCTIONAL TEST' column of-f'

. Table 4.3.6-1, suggests that.a CHANNEL FUNCTIONAL TEST of ths high and low power setpoints/ turbine first-stage pressure instruments should 'be -

o performed'"within one hour prior to control rod movement (unless j

L,#

g J

w h,

_ performed within' the previous -24' hours) and as nach power range above the RPCS low power setpoin't is' entered for the first. time during any-L$

On the basis of the' 24'-hourperiodduringpog"increaseordecrease.

testing should be deleted free Table

,1 discussion above,.the "D 4.3.6-1 since the. rod pattern controller will be adequately tested K"c

- pursuant to Specification 4.1.4.2 and the operability,of the'curbine self does not-need :to.be verified as

[L first-stageinstrumentationgd 4~

frequently. Deletion of "D is consistent with the Technical

[,1 '

Specification approved for the Perry Nuclear Power Plant.

Note- (e) is a carry over. from earlier versions of the BWR Technical Specifications. ' - The note refers to the' " reactor mant.al control" multiplexing input _which literally refers to earlier designs. As it may

_l be applied to Clinton (BWR6-Solid State) it may be-taken to maan that.

i the associated CHANNEL FUNCTIONAL TEST'should include verification that h

the RPCS is seeing, besides-the correct inputs from the turbine first j

stage pressure (high/ low power trip) senscrs, tha multiplexed control

.g 2 rod. position informat1on such that'the rod movement is correspondingly q

' L J

inhibited or restricted as appropriate. As noted in TSAR section 7.6.1.7.3, rod position is the' primary data input for the RPCS:

"There is a dual rod position probe for each drive. Each probe has

{

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two sets of reed switches for rod position information and will I

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' provide, through different connectors, inputs ^to different rod

_ position multiplexer. Two rod position multiplexer are provided,

]

!y W one for each channel. These multiplexer transmit rod position j

~

'l data to the rod action controls. These controls will decode the multiplexed data and provide rod position data to the RPCS controller for all rods. The rod position multiplexer ans controls are arranged in two divisions."

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Page 140 of 157 f

The RPCS is a subsystem of the Rod Control and Information System

~(RCIS). The RCIS contains its own self-test system which, with respect to the RPCS, provides "a means of comparing' the outputs of the RPCS logic devices as a way of monitoring the performance of the two channels. Both chaanels must be operable and have identical outputs before rod motion is permitted. Failed comparison and circuit failures or inoperative conditions will be indicated in the main control room."

The RCIS/RPCS for Clinton is an improvement.over earlier designs since it automatically performs test 1ng which satisfies.the intent of note

"(e)".-

In addition, the intent of note "(e)" is also satisfied when testing is performed in accordance with Specification 4.1.4.2.

The note is misleading, unnecessary, adds confusion to the existing requirements j

in Table 4.3.6 1, and should thus be deleted.

l It should be noted that note "(e)" has been deleted from Table 4.3.6-1 in both Grand Gulf Nuclear Station and the Perry Nuclear Power Plant Technical Specifications. The proposed change is therefore consistent with previously approved Technical Specifications.

I Mote (d) attached to the "M" in the CHANNEL FUNCTIONAL TEST column of Table 4.3.6-1 states that the CHANNEL FUNCTIONAL TEST for the high and L

low power setpoint/ turbine first-stage pressure instruments shall be performed "at least once per 31 days while operation continues within a given powe.r range above the RPCS low power setpoint." For the sake of simplicity, the note should be deleted. Without "(d)" attached to the "M", the GANNEL FUNCTIONAL TEST will be performed at least once every J

31 days for the applicable OPERATIONAL CONDITONS. The OPERATIONAL

/

I CONDITIOUS specified encompass the core specific conditions of

" operation...within a given power range of the RPCS low power setpoint."

Although deleting the note cculd require the CHANNEL FUNCTIOVAL !E3I(3' j

(d'^

W u.

to be performed more often than required with the ncte for example, was in operation for an extended period of time in mode 1 or 2 but at c power level below that corresponding to the low power i

setpoint), Plant Operations feels that simplifying the requirement is Therefore the proposed change would delete note "(d)".

l more important.

It should be noted that note "(d)" has been removed from the Perry Nuclear Power Plant Technical Specifications.

Finally, note "(b)".should be reworded by replacing the existing words, j

"within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if net performed within the previous

]

7 days," with "within 7 days prior to startup." IP feels that the J

existing words are potentially confusing, and that the proposed note j

will ensure that the instrumentation is OPERABLE for'its intended 1

function to the same degree of confidence as with the current note.

)

Although the intent of the note is to have the CHANNEL FITNCTIONAL TEST performed as close to startup as possible (within 24 hourh,), the l

efficacy of the note is undermined somewhat by the words "if not performed within the previous 7 days." The note requires anticipating time of a startup so as to implement the first part of the the exact note. This is not always possible. That is, plant startup could be l

I delayed for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after test completion. As long as the is still delay after test completion is less than 7 days, the test acceptable even though the intent of the first part of the current nots (within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) is not fulfilled. The proposed note would provide greater operational flexibility, I

i a

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s/

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to U-601048 q

p Page 141 of 157 j

4 l

Although the proposed note is not an exact equivalent of the current I

note', IP feels-that the proposed change is justified in view of the t

L reliability'of the instrumentation as discussed above, the fact that the surveillance test interval for the applicable OPERATIONAL CONDITIONS is much longer than the time constraints of the note, and that removal of the note will make the Technical Specification easier to implement with no significant difference in the effectiveness of the r

proposed note compared to the. current one. The wording of the proposed l

i note is identical to that of note "(b)" appearing in the Technical Specifications approved for the Perry Nuclear Power Plant.

l l

Basis For No'Significant Consideration j

1 According to 10CFR50.92, a proposed chan'ge to the license (Technical Specifications) involves no significant hazards. consideration if -

. operation of the facility in accordance with the proposed change would (1) involve a significant increase in the probability 'or not

, consequences of an accident previous y eva uate ; or (2) create the l

l d

possibility of a new or different kind of accident from any accident

{

l 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 of the following reasons.

(a) The purpose of the rod pattern control system is to reduce the consequences of the postulated rod drop accident to an acceptable level by restricting the patterns of control rods that can be established to predetermined sets. This condition is enforced at low power levels in the range below the low power setpoint. The. proposed change is not applicable to this range except at,the upper bound, i.e., the low power setpoint (LPSP). It has been determined, however, that under the proposed change, adequate testing will be performed (both manually and automatically) to ensure that the change in mode of operation occurs at the LPSP in accordance with the intent of the Technical Specifications; i.e., the rod withdrawal limiter mode of operation will yield to the rod pattern control system mode of operation when sensed power decreases below the LPSP (and vice versa when sensed power increases above the LFSP).

(b) The purpose of the rod withdrawal limiter (RWL) is to restrict continuous rod withdrawals to prevent excessive changes in the heat flux at power levels above the LPSP (when no further restrictions on red pattern are imposed). This mode of operation (including che range between the LPSP and the high power setpoint [HPSP] and the range above the HPSP) will not be affected by the proposed change. It has been determined j

that under the proposed change, adequate testing will be the

]

performed (both manually and automatically) to ensure thateach rod withdrawal limiter performs its intended function at povar range above the LPSP in accordance with the intent of the Technical Specifications.

O

,s iP' j

1

~ * ~

to U-601048 Page 142.of 157

' j.

(2)- 'The proposed change does not involve a significant incroace in the probability or consequences of an; accident previously evaluated _

]j because'the proposed change involves no design changes or new modes-of operation.; No further consideration of any accident.outside of s

'those already evaluated is required.

- (3) It has.been determined that the proposed change does,not involve a:

significant reduction in aamargin'of safety. No changes to the i'

s'etpoints.or any analytical limits associated with'the accidents or considerations discussed in part (1) are 'affected by the' proposed change., If reliability of the applicable instrumentation is considered.a margin of safety, it has already been determined that this margin of safety is not significantly. reduced.

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TABLE 4.3.6-1 (Cont:nued)-

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CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE-RE0VIREMENTS-1.

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' TABLE' NOTATIONS e

-(a) : Neutron' detectors may'be excluded frbm CHANNEL CALIBRA IT ON.

1

-- a

. ithin' 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior,to startup, if not performed within the; previous ; \\'.

(b)

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7. days /

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(c) Within one hour prior to control rod movement, unless performed withia.

Lthe previous 24. hours, and.as each power, range above theLRPCS' low powerL i

i during any 24. hour period during -

.t 4

b.;setpoint is entered for theif rst t me

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power increaseg.destease.

(d) At least once per.31 days while operation continues within..a given powerpe-

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m range =above the.RPCS-low-power setpoint.-

((e)-:Includesreactormanualcontrol;maltiplexingsystemin 7

-(f),. Calibrate the analog; trip module at leastionce per 31' days.

With'any control rod withdrawn. Not applicable-to control rods removed

~

per Specification 3.9.10.1.or 3.9,10.2.

s J

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p CLINTON -' UNIT 1 3/4 3-69

l e

to U-601048 Page 145 of 157

_ PACKAGE NUMBER 24 Description and Justification for Proposed Change l

The purpose of this proposed change (s) is to correct typographical errors and clarify existing requirements regarding Note (a) in Table 3.6.4-1 of the Technical Specifications. Under this proposed change:

l (1) 'Nute "(a)" as it appears in the " APPLICABLE OPERATIONAL CONDITIONS" column of Table 3.6 4-1 (Containment Isolation Valves) would be relocated as shown on'the attached marked-up pages (3/4 6-42 through 3/4 6-51),

(2) the note itself, as it is written on p. 3/4 6-61 under " TABLE NOTATION 3", would be modified as a result of the relocation, (3) a typographical error identified on page 3/4 6-42 would also be corrected.

These are discussed in more detail below.

The way that tha "(a)" is located now makes it appear as if note (a) is applicable only to OPERATIONAL CONDITION 3 (or, for the 1B21-F063 valves, in the OPERATIONAL CONDITION designated "#"). This is a typographical error carried over from corrections that had been made to the draft Technical Specifications. At one time, the APPLICABLE OPERA {IfRAL CONDITIONS had erroneously been specified as "At All Times

". When this was corrected to show the correct applicabic OPERATIONAL CONDITIONS, i.e., "1,2,3" (or "1,2,3,#"), "(a)" was left in place giving the appearance that "(a)" is attached only to "3" (or "#").

The intent of note "(a)" is'to. allow the applicable valves to be opened on an intermittent basis under administrative control during the applicable operational conditions.

(For most valves that are not secondary containment bypass paths, the applicable OPERATIONAL CONDITIONS are 1,2,3 because, as specified by Technical' Specification 3.6.1.1, PRIMARY CONTAIN>ENT INTEGRITY is required for OPERATIONAL CONDITIONS 1,2,3.

Other valves may have other OPERATIONAL CONDITIONS as required, but the intent of the note still applies unless specially excepted.) Rather than attach "(a)" to every applicable OPERATIONAL CONDITION number or symbol listed for every applicable valve, "(a)" may be relocated as shown. Since note "(a)" is applicable to all of the test connection, vent and drain valves and their corresponding applicable OPERATIONAL CONDITIONS listed in Table 3.6.4-1, note "(a)"

can be attached to the " Test Connections, Vents, and Drains" heading appearing on pages 3/4 6-42 through 3/4 6-51.

To ensure that there is no doubt that the note is applicable during any and all of the applicable OPERATIONAL CONDITIONS for the applicable valves, the wording of,the note itself on p. 3/4 6-61 vill be modified by adding "during applicable OPERATIONAL CONDITIONS." (It is understood that the note is not needed for OPERATIONAL CONDITIONS other then the applicable OPERATIONAL CONDITIONS.)

See DEFINITION 1.31 for PRIMARY CONTAIN>ENT INTEGRITY on p. 1-5 of the CPS Technical Specifications. The part of the definition that says,

"...except as provided in Table 3.6.4-3 of Specification 3.6.4" corresponds to note "(a)".

h o

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l to U-601048

_q

(:

Page 146 of 157 l

k While reviewing ' Table 3.6.4-1 for applicability of the changes described above, a typographical error was noted on page 3/4 6-42: A "(b)" is attached to the " Test, Connections, Vants, and Drains" heading where it

{

is intended to insert "(a)"...This is clearly a typc, graphical error

<4 ;

. since note "(b)" does not apply to the test connection, vent and drain valves.

(Note "(b)" appli'es to the excess-flow check valves listed in Whe "Other Isolation Valves" section of the table ) No other notes j

i listed on p.'3/4.6-51 categorically apply to the test connection, vents, and drains other.than note "(a)".

Under the proposed change, "b" would j

1 be appropriately replaced by "a".

d Basis'For No Significant Hazards Consideration

.a According to 10CFR50.92, a proposed chango to the license (Technical Specifications) involves no significant hazards consideration if operation of the facility in accordance vitb the proposed change would'

.l (1) involve a significant increase in the probability or j

not consequences of an accident previously evaluated; or'(2) create che l

possibility of a new or different kind of accident from any accident l

previous?y evaluated; or (3) involve a significant reduction in a margin i

j of safety.

(1) The proposed change involves no significant increase in the probability or consequences of an accident previously evaluated because the proposed change does not aff ect the intent or j

The

]

implementation.of the applicable Techn1 cal Specifications.

purpose of the change is to clarify existing requirements by t

correcting existing typographical errors that have been identified.

(2)

It has been determined that the proposed change does not create the possibility of a new or different kind of accident'from any i

previously evaluated since the proposed change does not involve any

)

The design changes, new requirements, or new modes of operation.

scope of tha proposed change is confined only to the correction of

]

typographical errors and.the clarification of existing.

f

,1 requirements.

(3) The proposed change does not involve a significant reduction in a margin of safety because no setpoint, design or analytical limit assumed or required by any analysis is affecced by the proposed q

change. 'The intent of the existing Technical Specification

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requirements would remain unchanged since the change is proposed only to correct typographical errors.

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to U-601048 q

Page 157.of 157 5

TABLE 3.6.4-1 (Co,ntinued)

!s',.l CONTAINMENT 150LAT10N VALVES

,j f

/*

TABLE NOTATIONS t-c' l)

/

(a) May be opened en an intermitt'ent basis under apinistrative. control du3epat g OP*mem.

CC@mcNS.

p (b) Excess flow check valve actuation differential pressure.

p (c) Isolation valving for instrument lines which penetrate the containment con-form to the requirements of NRC Regulatory Guide 1.11.

The in-service in-spection program will provide assurance of the operability and integrity o i 4

of these isolation provisions. Type "C" testing will not be performed on

' h' the instrument line isolation valves. The. instrument lines will be within the' boundaries of the Type "A" test, cpen to the media (containment atme-sphere or suppression pool water) to which they will be exposed under pos-tulated accident conditions.

Instrun nt taps from the process line located between the process isolation valves and the penetration, and not them-i selves penetrating containment, will be Type "A" and/or "C" tested along with the process line isolation valves.

j,,

M (d) Excess flow check valve.

(e)' The RHR system may be operating in the shutdown cooling mcde during the Type A test. These valves are tested using water but'the results are not

) required to be added to the Type A test results. Ttn LPCS, HPCS, and RHR i

/ ),f may be aligned in the normal standby or injectipi mode during the Type A-M' J:'

test. This will expose the closed loop outsids containment to containment f

pressure through the suppression pool. This is the closest valve alignment to the post-LOCA alignment possible. Type C water test resu1Ls on there 7

' mtion valve.s will not be added to the Type A tast results.

(f)( Valves shall be, closed in accordance with SECONDARY CONTAINMENT INTEGRITY.

alves shall be " sealed closed" by utilizing mechanical devices to seal (g) j\\or lock the valve closed or to prevent power fron being supplied to the i - "

p1veoperator.

When handling irradiated fuel in setdedary containment and during CORE ALTERATIONS and operations yith a,' potential for draining the reactor vessel.

y 7,

Isolates on RCIC low steam line pressure only.

t Isolation signal descriptions are provided in Table 3,6,4-2.

For test prescure = 9.0 asig, the valve (s) shall be pressurized using air or nitrogen, and for ten pressure = 9.9 ps'ig., the valve (s) shall be i pressurized using ' water, With any control-rod withdrawn.

Not appHcable to any control rods removed per Specification 3.9.10.1 or 3.9.10.2.

t CLINTON - UNITr1 3/4 6-61 i

l

(

_1

ML20236K708

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