Text

Proposed Tech Specs Re Reactor Coolant Sys Pressure Isolation Valves
	ML20076M965
Person / Time
Site:	Callaway
Issue date:	03/15/1991
From:	; UNION ELECTRIC CO.
To:	;
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ML20076M963	List: ML20076M959; ML20076M965;
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	NUDOCS 9103260242
	Download: ML20076M965 (11)
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ULNRC-2 37 7 -

ATTACHMENT 1 TECIINICAL SPECIFICATION CHANGES 9103260242 910315 DR ADOCK 05000483 PDR 4

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~RFACTOR C0013 1_5YSTEM

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-OPERAll0NAt itAKAGL JMillHGCON0lfl0NTOROPERA110N

3. 4. 6. 2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

I gpm UNIDENTIFIED LEAKAGE, 1 gpm total reactor-to-secondary leakage through all steam

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c.

ll generators not isolated f rom the Reactor Coolant System and 500 gallons-per day: through any one steam generator, n

d.

10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, 8 g'pm per RC pump CONTROLLED LEAKAGE at a Reactor Coolant System y

e.

. pressure of 2235 1 20 psig, and p

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Ieekege et : ";;. t:r 0;;hnt Sy:t= pr :eter; cf 2235 : -70 ;;f;;

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i APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

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With any PRESSURE ROUNDARY LEAKAGE, be in at least HOT STANDBY a,

l within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , h-

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b.

.With any Reactor Coolant System leakage greater than any one of the.

.above Iimits excluding PRESSURE 80VHDARY LEAKAGE and leakage from

l Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits:within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or.be'in at least HOT STANOBY-g

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

-30 hours.

With any Reactor Coolant System Pressure Isolation Valve leakage l

.c.

F greater than the above limit, reduce the leakage rate to within-limits within'4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or be in It least HOT STANOBY within the 1:

next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s-and in HOT SHUTOOWN;within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with an RCS pressure-of less than 600 psig.

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"Te'.t pressures less than 2235 psig but greater than 150 psig are allowed.

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' Observed leakage shall be adjusted for the actual test pressure up to j

2235 psig assuming the leakage to be directly proportional to pressure ciliicrenttal'to the one-half power.

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CAllAWAY~- UNIT 1 3/4 4-19 e

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e Insert 1 to T/S 3.4.6.2.f I

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The leakage from each Reactor Cooia t System Pressure l

Isolation Valve specified in Table 3.4-1 shall be i

limited to 0.5 gpm per nominal inch of valve size up to I

a maximum of 5 gpm, at a Reactor Coolant System pressure of 2235 1 20 psig.

Valves which are 2 inch or smaller (nominal size) shall be limited to 1 gpm

. leakage.*

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[HFACTOR-C001ANijSYSTEM Y/Sj

-0PtHAll0NAL tlAKAGL i

LIMillNG CONDIIION fOR OPlRA110N

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f, 3.4.6.2 Reactor _ Coolant System leakage shall be limited to:

t a,

No ' PRESSURE BOUNDARY LEAKAGE,

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,b.

1 gpm UNIDENTIFIED LEAKAGE,.

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1 gpe: total reactor-to-secondary leakage through a11 steam f-'

L generators not isolated from the' Reactor Coolant System and 500igallons per' day through any,one steam generator.

d.

10 gpm.IC.NTIFIED LEAKAGE from the Reactor Coolant System.

.e.

8'g'pm per RC pump' CONTROLLED LEAKAGE at a Reactor Coolant System u

pressure of-2235 2 20 psig, and t

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._The. leakage from each Reactor-Coolant System Pressure Isolation ~ Valve specified in Table 3.4-1 chall be-limited to o.5 gpm per nominal inch-

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- of-valve site up to a-maximum of s gpm, at a Reactor Coolant Systen pressure of-2235 t'20 psig. Valves which are 2 -inch or smaller L

(nominal site) shall be-limited to 1 gpa leakage.*

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'p, APPLICABILITY:

H0 DES 1, 2, 3, and 4.

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ACTION:

a, With any PRESSURE-BOUNDARY LEAKAGE be in at least HOT STANDBY a<ithin 6--hours'and in COLD SHUTOOWN within-the'following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

'With-any Reactor Coolant System leakage greater-than any one of the J!

aboveLitmits.cexcluding PRESSURE BOUNDARY LEAKAGE and leakage from

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Reactor Coolant System Pressure Isolation Valves,. reduce the leakage

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-ratentoLwithin limits within 4=hoursnor be-in at least HOT STANOBY^

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within the next 6_ hours and in COLO-SHUT 00WN within-the following.

pp J30 hours.

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"c.

With-any Reactor Coolant: System PressureLIsolation= Valve leakage

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< greater.than the above-limit, reduce-the leakage rate to within fa limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,-or be in at?least HOTESTANOBY within_the l.

next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOTESHUT00WN-within the :following 12 ' hours with

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'an RCS pressure of less than;600 psig.

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. " lent. pressures less=than-2235'psig but-! greater'than-150.psig are allowed, pi j,

0bserved yeakage shall'be-adjusted forcthe actual test pressure up to j.i i2235-psin assuming the leakage to ce directly proportional to pressure Jtitlere'ntial-to the one-half power.

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I, TABLE 3.4-1 r.

REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION 948 A, B C, O SI/RHR/Accum C'old leg Inj 2dtIn BBVB.

B, C. O SI/RHR Hot Leg Inj BBV001, 02,

40, 059 BIT Cold Leg I on BBPV8702 A, B RHR Horma etion EJV8841 A, B RHR Leg Recire Ctat 150 EJHVB701 A, B Hormal Suction EMV001, 002, 003, 004 5

Leg Inj Ctat ISO EM8815 BIT In t Isolation EPV010, 02

, 040 51 Cold leg tmt !$0 EPV88

,B,C,O RHR Cold leg Inj 150 956 A, B, C, O Accum inj ! solation

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'W CALLAWAY - UN!T 1 3/4 4-21 1

TABLE 3.4-1 1

g REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES f

MAXIMUM VALVE VALVE -

ALLOWABLE NUMBER SIZE (in.)

FUNCTION LEAKAGE (gpm),

BB8948A 10 RCS Loop 1 Cold Leg SI Accu Chck 5.0 BB8948B 10 RCS Loop 2 Cold Log SI Accu Chek 5.0 BB8948C 10 RCS Loop 3 Cold Leg SI Accu Chek 5.0 H

BB8948D 10 RCS Loop 4 Cold Leg SI Accu Chek 5.0 BB8949A 6

RCS Loop 1 Hot Leg SI/RHR Pump Chek 30-i.

BB8949B 6

RCS Loop 2 Hot Leg SI/RHR Pump Chck 3.0 BBB949C 6

RCS Loop.3 Hot Leg SI/RHR Pump Chck 3.0 BB8949D 6

.RCS Loop 4 Hot Leg SI/RHR Pump Chck 3.0 BBV0001 1.5 RCS Loop 1 Cold Leg. SI/ BIT Chck 1.0 BBV0022 1,5 RCS Loop 2 Cold Leg SI/ BIT Chek 1.0 L

BBV0040 1.5 RCS Loop 3 Cold Leg _SI/ BIT Chek 1.0 BBV0059 1.5 RCS Loop 4 Cold Leg SI/ BIT Chck 1.0 BBPV8702A.

12 RCS Loop 1 Hot Leg to RHR Pumps ISO 5.0 BBPV8702B 12 RCS Loop 4 Hot Leg to RHR Pumps ISO 5.0 EJ8841A 6

- RHR TRNS SIS Hot Leg Loop 2 Recirc.

3.0 EJ8841B 6

RHR TRNS SIS Hot Leg Loop 3 Recirc 3.0 EJHV8701A 12 RHR Pump A Suction ISO 5.0 EJHV8701B 12 RHR Pump B Suction ISO.

5.0 ENV0001 2

SI Pump A Disch to Hot Leg Loop 2 Chck 1.0 EMV0002 2-SI Pump A Disch to Hot Log Loop 3 Chck 1.0 EMV0003 2-SI Pump B Disch to Hot Leg Loop 1 Chck 1.0 EMV0004 2

SI Pump B Disch to Hot Leg Loop 4 Chck 1.0 l

-EM8815 3

BIT CVCS.Out Check 1.5 EPV0010 2

SI Pumps tc RCS Cold Leg Loop 1 Chek 1.0 EPV0020 2

SI Pumps to ' RCS Cold Leg Loop 2 Chck 1.0 EPV0030.

2 SI Pumps to RCS Cold Leg Loop 3 ChcP 1.0 4

EPV0040 2

SI Pumps to RCS Cold _ Leg Loop 4 Chck 1.0 EP8818A 6

RHR Pumps _to RCS Cold Leg Icop 1 Chck 3.0 l

EP8818B_

6 RHR Pumps to RCS Cold Leg Loop 2 Chck 3.0

'EP8818C 6

RHR Pumps to RCS Cold Leg Loop 3 Chek 3.0 EP8818D 6

RHR Pumps to RCS Cold Log Loop 4 Chek 3.0 EP8956A 10 SI Accu TK A-Out Upstream Chek 5.0 EP8956B 10 SI Accu Tk B Out Upstream Chck 5.0 EP8956C 10 SI Accu TK C Out Upstream Chck 5.0 l

l EP8956D 10 SI Accu TK D Out Upstream Chck 5.0-

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RE ACTOR COOLANT SYSTEM BASES

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OPERATIONAL LCAKAGE (Continued)

The eakage from any RCS pressure isolation valve is suffigiently low to ensure early detection of possible in-series check valve failure.

It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a sub-stantial length of time, verification of valve integrity is required. $1nca these valves are important in preventing overprossurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses cen-tainment, these valves should be tested periodically to ensure low probability of gross failure.

The Surveillance Requirements for RCS pressure isolation valves prbvide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage free the RCS pressure isolation valves is IDFNTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4.4.7 CHEMISTRY The limitat' ions on Reactor Coolant System chemistry ensure that corresion of the Reactor Coolant System is minimized and reduces "'e potential for Reactor Coolant System leakage or f ailure due to striass scrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural ihtegrity of the Reactor Coolant System uver the life of the plant. The associated effects of exceeding the oxygen.

chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration

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1evels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of tne Reactor Coolant System. The time interval permitting continued operation within the restrict. ions of the Transient Limits provides time for taking corrective actions to restore the contaminant l

concentrations to within the Steady-State Limits.

The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in suf ficient time to take corrective action.

1 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumett steady state reactor-to-secondary steam generator leakage rate of 1 gpm. The ralues 1

for the limits on specific activity represent limits based upon a parametric evaluation by-the NRC of typical site locations.

These values are conservative 3

in that specific site parameters of the Callaway site, such as SITE BOUNDARY j-location and meteorological cond;tions, were not considered in this evaluation.

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REVISjob

.f-REACTOR COOLANT-SYSTEM 4

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-OPERATIONAL LEAKAGE (Continued)

The' leakage from any RCS pressure isolation valve is sufficiently h

low to.,-*ure early detection of possible. in-series check valve f ailure.

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"a-4 is' apparent that when pressure isolation is provided by two in-series chec,k

'k valves and when faflure of one valve in the pair can go undetected for a sub-stantial length of time, verification of valve integrity is required.

Since b

.these valves are important in preventing overpressurization and rupture of q

the ECCS low pressure piping which 3uld result in a LOCA that bypasses con-is tainment, these valvet should be *.ts.'d peric '4ca11y to ensure low probability n

of gross f ailure.

i The Surveillave Requirement.s for RCS prer,ure isolation valves provide h

added assurance of valve integrity thereby rencing the probability of gross-p valve f ailues and consequent intenystem LOCA.. Leakage from the RCS pressure p'

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-isolation valves is IDENT2FIED LEAKAGE and will be considered as a portion of l:

the allowed limit.

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h 3/4.4.7 CHEMISTRY l

The limitations on Reactor Coolant System chemistry ensure that corrosion i

. of:the Reactor Coolant System is minimized and reduces the potential for i

Reactor Coolant System leakage or f ailure due to stress corrosion. Maintaining

il the chemistry within the St=ady-State Limits provi.ies adequate corrosion j$

protection to ensure the structural integrity of t0e Reactor Coolant System over the. life of-the plant.

The associated offects c' exceeding the oxygen,.

chloride,- and fluorido limits are time and temperature dependent..

Corrosion studies show that operation may be continued with contaminant concentration.

l 1evels in-excess of the Steady-State Limits, up to the Transient Limits, for-

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the specified limited time intervals without having a significant ef fect on p

~ the. structural-integrity of the Reactor Coolontilystein.

The time interval

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permittinc.antinued operation within the restrictions of the Transient Limits q

provides:tirs for taking corrective actions to restore the contaminant 5

- concentrations to 'within the _ Steady-State Limits.

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-The Surveillance Requirements provide adequate assurance that concentrations 1

u in. excess of the' limits will be detected in sufficient time to take corrective c

- action.

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'3/4 a;8 ' SPECIF fC ACTfVITY

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-Thh. limitations on the specific activity of the reactor coolant ensui e that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an' appropriately small fraction.of 10 CFR Part 100 dose guideline values fo11Mng.

[i i) a steam generator tube rupture accident -in conjunction with an assumed ste ady i

state reactor-to-secondary' steam generator leakage rate of 1 gpm.

The VC es for the limits on specific activity represent limits based upon a parametric

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evaluation by the NRC of typical site locations.

These values are consersative Q

in that specific site parameters of the Callaway site, such as SITE BOUNDARY l

l' location and rueteorological conditions, weresiet considered in this evalua. ion.

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Page 1 of'2

- ULNRC-2377 SAFETY _ EVALUATION This amendment application requests a revision to Technical I

Specification LCO-3.4.6.2 to change the allowed leakage limit for reactor coolant system pressure inolation valves (RCS PIVs) and j

to correct valve numbers and descriptions as shown in Table 3.4-1.

The RCS PIV LCO permits system operation in the presence of leakage through valves in amounts which do not compromise safety.

The-RCS is -isolated from other systems by valves.

During plant life these interfaces can produce varying amounts of reactor j

coolant leakage through either normal operational wear or

mechanical-deterioration.

The RCS PIV LCO permits system operation in the presence of leakage through these valves in-amounts which do not compromise safety.

PIV leakage limits apply to. leakage 1 rates for individual valves, l

The basis for this-LCO is the 1975 Reactor Safety Study (Ref. 1) which identified potential intersystem Loss-Of Coolant Accidents

-(LOCAs) as a significant contributor to the risk of core melt.

A H

subsequent study (Ref-. 2) evaluated various PIV configurations to determine the' probability of intersystem LOCAs, This study concluded that periodic leak testing of the PIVs can e

l substantially reduce intersystem LOCA probability, The proposed LCO leakage-limit is based on permitting 0,5 gpm per

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' nominal inch of valve size for valves larger than 2 inches with a maximum upper-limit of 5 gpm (Ref. 4).

Valves which are 2 inch or caaller (nominal size) shall be limited to 1 gpm leakage.

The

-previous criterion of 1 gpm for all valve-sizes was considered

.nrbitrary and was not an indicator of imminent accelerated deterioration or potential valve' failure.

A study (Ref. 3)

-concluded. allowable' leak rates based on' valve size was superior to;a single allowable velue.

The single value imposes an unjusti-fled penalty on the larger valves without providing-information on potential valve degradation.

In addition, enforcing the single; value criteria resulted in higher personnzl radiation exposures because-larger valves must be~ repaired in-place.

Updating Table 3.4-1 to correct; valve numbers and descriptions-will make this table consistent with controlled drawings and the Callaway Master Equipment List.- This 10 an editorial change only.

The ',.roposed change -to Technical Speci fication 3/4. 4.6 (LCO 3.4.6.2) does not involve an unreviewed' safety question because operation of Callaway. Plant with this change would not.

1.

Increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety enalysis report.

This change does not affect the operability requirements of the RCS PIVs or the ability of these valves to perform their.

intended safety functions.

The change revises the acceptable leakage criteria of the PIVs to values baced on valve size.

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h Attoel.msnt 2 Pogs 2 of 2 l.

ULNRC-2377 2.-

Create a possibility ?>r an accident or malfunction of a different type than any Treviously evaluated in the safety analysis report.

There is no new type of accident or malfunction being created and the method and manner of plant operation reio: ins unchanged.

The change results in individual valve leakage limits based on valve rire with total identified leakage limited to 10 gpm as currently specified in LCO 3.4.6.2.

'3.-

Reduce the margin of safety as defined in the basis for any technical specificaticn.

This is based on the fact that no plant design changes are involved and the current practices O

and procedures for. monitoring-valve leakage will not change.

Given the above discussions as well as those presented in the Significant Hazards Evaluation, the proposed change does not adversely affect or endanger the health or safety of the genera) public or involve a significant safety hazard.

Reference _o 1)

U.S. Nuclear Regulatory Commission, " Reactor Safety Study-An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants", Appendix V, WASH-1400 (NUREG-75/014),

Octots-1975.

2)

U.S.

AC, "The-probability of Intersystem LOCA: Impact Due

[

to: Leak Testing and Operational Changes", NUREG-0677, May 1980.

3)

EG & O_ Report, EOG-NTAP-6175, "In Service Leak Teoting of I

Primary-Pressure Isolation Valves", R. A. Livingston, February 1983.

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4)

-Safety Evaluation by the Office of Nuclear. Reactor L

Regulation Related to Amendment No. 50 to Facility Operating License No. HPF-2.and Amendment No. 41 to Facility Operating

-License No. NPF-8,: Alabama Power O_Mpany, Joseph M.

Farley Plant, Units Nos. I and 2, Docket Nos. 50-348 and 50-364, October 15, 1984.

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A t t.a chmen t 3 I'n y e 1 of 1 ULNRC-2377

"~.GNI#1 CANT _ HAZARDS EVALUATION This amendnw nt appliention requents a revision to Technien1 Specificat1<,n 3/4.4.6 to change the allowed lonkage limit for reactor coo.nnt nyntem prennure isolation valven to a value bnned on valve nir.e.

The proposed change to Technical Specification 3/4.4.6 does not involve a significant hazards consideration beenuse operation of Callawny Plant with this change would not-1.

Involve a sign ficant increase in the probability or concequences of an accident previounly o nlunted.

The proposed amendment still requiren exactly the name actione to be taken when or if valve leakage limite 2e oy.ceeded an in required oy current Technical Specificattons.

2.

Create the ponsibility of a new or difforent klud of nccide,.t from any previously evnlunted.

There in no new type of accident or mnifunction being crented and the method and manner of p1eut operation remains unchanged.

The change applion a lenknge limit baned on valve nize.

The total amount of identified reactor coolant nyntem lenhage remninn at 10 gpm.

3.

Involve a significant reduct.lon in a mntgin of. nnfety.

The m9rgin of safety remainn unaffected nince no design change in being mnO and total identified leakage limite remain the onme no dis:,

ned in Technical Epeci fiention 3/4. 4. 6, An discussed above, the proposed enanga doen not involve a nignificant increnne in the probability or connequencen of an accident previously evaluated or create the poncibili ty of a new or different hind of accident from any previously evaluated.

This change doec not result in a significant reduction in n margin of safety.

Therefort, it han been deteimined that the proponed change doom not involve a nianifiennt hnzardn considorntion.

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ML20076M965

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