Text

Model Tech Specs Re Safety Limit Setpoints & Refueling Operations in Response to Commitment to Install Permanent Scram Discharge Sys Mods
	ML20024B407
Person / Time
Site:	Peach Bottom
Issue date:	06/24/1983
From:	; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20024B404	List: ML20024B403; ML20024B407;
References
	NUDOCS 8307080663
	Download: ML20024B407 (30)
	v • d • e

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SAFETY LIMITS AND LIMITING SAFE ~( SYSTEM SETTINGS

2. 2 LIMITING SAFETY SYSTEM SETTINGS l

REACTOR PROTECTION SYSTEM INSTRUM6NTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1.

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

With a reactor protection system instrumentation setpoint less conservative than the salue shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Satpoint value.

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LIMITING SAFETY SYSTEM SETTING BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 8.

Scram Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion of the control rod drive ~ pistons during a reactor scram.

Should this volume fill up to a point where there is insufficient volume to accept the displaced water et pcessunes belemt 55 peig, -conteci red insertica~culd be hinde ed.

TM renc-tor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water frcus the movement of the rods at pressures below 65 psig when they are tripped.

The trip setpoint for each scram discharge volumeisequivalenttoacontaingvolumeof(

) gallons of water.

9.

Turbine Stoo Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stoo valves. With a trip setting of (5)% of valve closure from full open, the l

resultant increase in heat flux is such that adequate thermal margins are maintained during the worst case transient (assuming the turbine bypass valves (fail to) operate).

l 10.

Turbine Control Valve Fast Closure, Trio Gil Pressure-Low The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection coincident with failure of the turoine bypass valves.

The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by the fast acting sole-noid valves and in less than (30) milliseconds after the start of control valve fast closure.

This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main tureine control valve actuator disc dump valves.

This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to tne Reactor Prctection System.

This trip setting, a faster closure time, and a different valve characteristic from that of the turbine stop valve, combine to produce i

transients which are very similar to that for the stop valve.

Relevant tran-sient analyses are oiscussed in Section (15.1.0) of the Final Safety Analysis Report.

11.

Reactor Mede Switch Shutdown Position l

The reactor mode switch Shutdown position is a redundant channel to the l

autcmatic protective instrumentation channels anc provides adcitional manuai l

reactor trip capability.

'12.

Manual Scram The Manual Scram is a redundant chan.el to the aut:matic protactive instrumentation channals and provides manual reactor trip capability.

SE-STS (EWR/~)

E 2-9 l

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4.

REACTIVITY CONTROL SYSTEMS 3/4.1.3 CONTROL R005 CONTROL R00 OPERABILITY LIMITING CONDITION FOR OPERATION l

3.1.3.1 All control rods shall be OPERABLE.

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APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With one control rod inoperable due to being immovable, as a result of a.

excessive friction or mechanical interference, or known to be untrippable: '

l.

Within one hour:

a)

Verify that the inoperable control rod, if withdrawn, is separated from all"dther inoperable control rods by at least two control cells in all directions.

'b)

Disarm the associated directional control valves ** either:

l i

1) Electrically, or l

2) Hydraulically by closing the drive water and exhaust l

water isolation valves.

l l

c)

Comply with Surveillance Requirement 4.1.1.c.

Otherwise, be in at least HOT SHUTOOWN 9ithin the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l 2.

Restore the inoperable control rod to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months l

or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With one or more control rods trippable but inoperable for causes other l

than addressed in ACTION a, above:

1.

If the inoperable control rod (s) is withdrawn, within one hour:

a)

Verify that the inoperable withdrawn control rod (s) is separated fraa all other inoperable control rods by at least two control cells in all directions, and b)

Demonstrate the insertion capability of the inoperable withdrawn l

control rod (s) by inserting the control rod (s) at least one notch by drive water pressure within the nor al operating range *.

Otherwise, insert the inoperable withdrawn control rod (s) and disarm the associated directional control valves ** either:

a)

Electrically, or i

b)

Hydraulically by closing the drive water and exhaust water

[

isolation valves.

"The inoperable control rod may then be withcrawn to a cosition no furtner withdrawn than its position when found to be inoperable.

- May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to CPERAELE status.

GE-STS (EWR/4) 3/4 1-3

[f REACTIVITY CONTROL SYSTEMS

\\

LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued) 2.

If the inoperable control rod (s) is inserted, within ene hour disarm the associated directional control valves ** either:

a)

Electrically, or b)

Hydraulically by closing the drive water and exhaust water l

isolation valves.

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 .

l

'dith more than 8 control rods inoperable, be in at least HOT SHUTDC'4N c.

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

SURVEILLANCE REOUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be i

demonstrated OPERABLE by:

I At least once per 31 days verifying each valve to be open,* and l

a.

b.

At least once per 92 days cycling each valve through at least one l

complete cycle of full travel.

4.1.3.1.2 When above the (preset power level) (low power setpoint) of the RWM l

and RSCS, all withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch:

At least once per 7 days, and a.

b.

At least oni:e per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when any control rod is immovable as a result of excessive friction or mechanical interference.

4.1.3.1.3 All control rods shall be demonstrated OPERASLE by performance of Surveillance Recuirements 4.1. 3. 2, 4.1. 3. 4, 4.1. 3. 5, 4.1. 3. 6 and 4.1. 3. 7.

"These valves may be closed intermittently fer testing under administrative controls.

""May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.

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GE-STS (EWR/4) 3/4 1-4

n

REACTIVITY CONTROL SYSTEMS l

SURVEILLANCE REQUIREMENTS (Continued) 4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by i

demonstrating:

a.

The scram discharge volume drain and vent valves OPERABLE, when 1

control rods are scram tested from a normal control rod configura-tion of less than or equal to (50)% ROD DENSITY at least once per 18 months, by verifying that the drain and vent valves:

1.

Close within (30) seconds after receipt of a signal for control rods to scram, and 2.

Open when the scram signal is reset.

l b.

Proper (float) (level sensor) response by performance of a CHANNEL l

FUNCTIONAL TEST of the scram discharge volume scram and control rod I

block level instrumentation (AT level measuring system) (after each scram from a pressurized condition) (at least once per 31 days).

e 4

GE-575 (3WR/4) 3/4 1-5

': 3.

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e REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR LIMITING CONDITION FOR OPERATION 3.1. 4. 3 Both rod block monitor (RBM) channels shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to (30)% of RATED THERMAL POWER.

l

ACTION:

With one RBM channel inocerable, restore the inoperable RBM channel a.

to OPERABLE status withi'n 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and verify that the reactor is not operating on a LIMITING CONTROL ROD PATTERN; otherwise, place the incperbie rod block monitor channel in the tripped condition within the next hour.

~ b.

With both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within one hour.

SURVEILLANCE REOUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE by performance of a:

CHANNEL FUNCTIONAL TEST and CHANNEL CALIERATION at the frequencies a.

and for tne OPERATIONAL CONDITIONS specified in Table 4.3.6-1.

b.

CHANNEL FUNCTIONAL TEST prior to control red withdrawal when the reactor is operating on a LIMITING CONTROL ROD PATTERN.

GE-ST5 (SWR /4) 3/4 1-18 t

REACTIVITY CONTROL SYSTEMS

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BASES 3/4.1.3 CONTROL RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) limit the potential effects of the rod drop accident.

The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued

' operation.

A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic, problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a ganeric problem, therefore with a control red immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-insertaa position are

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consistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.

The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than (1.06) during the limiting power transient analyzed in Section (15. _ ) of the FSAR.

This analysis shows that the negative reactivity rates resulting from the scram with the acerage response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than (1.05).

The occurrenca of scram times longer then those specified should be viewed as an indication of a-systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periccs of time with a potentially serious problem.

The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a l

reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators are declared inocarable and Specification 3.1.3.1 then applies.

This prevents a pattern of inoperable accumulators that would result in less rea:tivity insertion on a s:ra than has been analyzed even though control rods with inocerable accumulators may still be inserted with normal drive water pressure.

Operability of the accumulator ensures that there is a means available to insert tne control rods even under the most unfavorable depressuri:ation of the reactor.

l GE-STS (5WR/a) 3 3/4 1-2

'?

j

.e 1

REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued) i' Control rod coupling integrity is required to ensure compliance with the 3

analysis of the rod drop accident in the FSAR.

The overtravel position feature J

provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after l

completing CORE ALTERATIONS that could have affected the control rod coupling integrity.

The subsequent check is performed as a backup to the initial demon-stration.

In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control red position indication system must be OPERABLE.

The control rod housing support restricts the outward movement of a control rod to less than (3) inches in the event of a housing failure.

The amount of rod reactivity which could be added by this small amount of rod withdrawal is less'than a normal withdrawal increment and will not contribute to any damage to the primary coolant system.

The support is not required when there is no 4

pressure to act as a driving force to rapidly eject a drive housing.

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

I 3/A.l.4 CONTROL RCD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual contrsi rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to i

result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident.

The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than (20)% of RATED THERMAL POWER, there is no possible rod worth which, if dropced at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm.

Thus requiring the RSCS and RhM to be OPERABLE when iHERMAL PCWER is less than or equal to (20)% of RATED THERMAL PCWER provides l

acequate control.

4 The RSCS and RhM provide automatic sucervision to assure that out-of-sequence rods will not be withdrawn or inserted.

TheanalysihoftheroddropaccidentispresentedinSection(15.__)of the FSAR ai d the techniques of the analysis are presented in a topical report, Reference 1, and two supolements, References 2 and 3.

The R2M is designed to automatically prevent fuel damage in the evert of erroneous rod withdrawal from locations of nign cower density during hign cower operation.

Two channels are provided.

Tripping one of the enannels will cloc<

erroneous rod withdrawal soon enougn to prevent fuel damage.

This system backs uo the written secuence used by tne operator for witndrawal of control rods.

GE-STS (SWR /A) 3 3/4 1-3 w

G" I

3/4.3 INSTRUMENTATION i

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

APPLICAEILITY:

As shown in Table 3.3.1-1.

ACTION:

With the number of OPERABLE channels less than required by the Minimum a.

OPERABLE Channels per Trip System recuirement for one trip system, place the inoperable channel (s) and/or that trip system in tne tripped condi-tion" within one hour.

The provisions of Specification 3.0.4 are not acplicable.

b.

With the number of OPERABLE channels less than recuired by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.

SURVEILLANCE REOUIREMENTS 4.3.1.1

+ h reactor protection system instrumentation channel shall be cemonstratea OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstratec to be within its limit at least once per 18 months.

Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a scecific reactor trip system.

l "An inoperaole cnannel neec not be placed in the trippec condition where this l

would cause the Trip Function to occur.

In these cases, the ir.::erable cnannei shall be restored to GPER ELE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACT.'"*. recuired bv Tabie 3.3.1-1 for tnat Tric Function snali ce taken.

l

~

'"If more channels are incpera:1e in one tric sys;em nan in the other, piace

{

the trip system with more inoceracle cnannels in :ne tri;;ed c:r.di:icn, t

except when this would cause the Trip Function to occur.

GE-STS (SWR /*)

3/4 3-1

TABLE 3.3.1-1

.n

<, l 111 AC10R PROTECTION SYSTEM INSIRllHENTATION ri APPLICABLE

MINIMON, 2.

OPERATIONAL OPERAulE CilANNELS l'lillCI~10flAL lilllT CONDITIONS PER TRIP SYSTEM (a)

ACTION 1.

InteYmesliate Range lionitors(I'):

a.

lieutron Flux - liigh 2

3 1

3,4(c) 3(d) 3 2

2 5

li.

Inoperative 2

3 1

f.

3, 4 2(d) 2 S

3 3.

Average Power Range Monitor ("):

I.b' 2.

a.

lieutron Flux - Upscale, Setdown 2

2 1

{

3(c) 2(d) 2 li.

I' low Iliaseil Simulated Thermal l

Power - Upscale 1

2 4

c.

Iixed lieutron I' lux - Upscale.

1 2

4

<t.

Inoperative 1, 2 2

1 3

5(c) 2(d) 3 l

2 (e.

Ilounscale -

1(U) 2

4) l

~

3.

Iteactor Vessel Steam Dome l'ressure - liigh 1, 2(I) 2 1

e, 4.

Iteactor Vessel Water Level - Low, level 3 1, 2 2

1 S.

11ain Steam Line Isolation Valve -

IU)

Closure I

4 4

O s

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION h

APPLICABLE MINIMUM R

OPERATIONAL OPERABLE ClfklINELS.

O futiCTI0llAL Uilli CONOTTIONS.

PERTRIPSY51(M(g)

ACTION 6.

Main Steam Line Radiation -

liigli 1, 2(g) 2 5

7.

(Primary Containment) (Orywell)

Pressure - liigh 1, 2(,,)

2 1

8.

Scram Discharge Volume Water 2

1 level - liigh 1, 2(j) 5 2

3 9.

Turbine Stop Valve - Closure 1(3) 4(k) 6

{

't' 10.

Turbine Control Valve Fast Closure I)

Valve Trip System Dil Pressure - Low 1((D))(3) 2 6

l 11.

Reactor Mode Switch Shutdown 1

Position 1, 2 -

1 1

3, 4 1

7 5

1 3

l 12.

Manual Scram 1, 2 2

1 3, 4 2

8 l

5 2

9 4

b,*

o i

3

<o TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION ACTION 1 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position d

within one hour.

ACTION 3 Suspend all operations involving CORE ALTERATIONS

and insert all insertable control rocs within one' hour.

ACTION 4 Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 5 Be in STARTUP with the main steam line isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in at itast HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 5 Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to < (250) psig, equivalent to THERMAL POWER less than (30)% of RATED THERMAL POWER, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ACTION 7 Verify all insertable control rods to be inserted within one hour.

ACTION 8 Lock the reactor mode switch in the Shutdown position within one hour.

ACTION 9 Suspend all operations involving CORE ALTERATIONS", and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within one hour.

"Excep movement of IRM. SRM or spe:ial movable dete:t:rs, or replacement of LPRM strings provided SRM instrumentation is OPERAELE cer 3:ecifica-icn 3.9.2.

GE-STS (EWR/4) 3/4 3-4

r

.?

1.'

~

TABLE 3.3.1-1 (Continued)

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

(c) This function shall be automatically bypassed when the reactor mode switch is in the Run position.

(c) The " shorting ' inks" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn

and shutdown margin demonstrations performed per Specification 3.10.3.

(d) The non-coincident NMS reactor trip function logic is such that all channels go to both trip systems.

Therefore, when the " shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRMS and 6 IRMS.

(e) An APRM channel is inoperaole if there are less than 2 LPRM inputs per level or less than (11) LPRM inputs to an APRM channel.

(f) This function is nct required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.

(g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i) With any control rod withdrawn.

Not applicable to control rods removed per Speci fication 3.9.10.1 or 3.9.10. 2.

(j) This function shall be automatically bypassed when turbine first stage pressure is < (250) psig, equivalent to THERMAL POWER less than (30)%

of RATED THERMAL POWER.

(k) Also actuates the EOC-RPT system.

" :: requirec for control rods removet oer Specification 2.9.10.1 cr 3.5.10.2.

GE-STS (EWR/4) 3/4 3-5

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itLAC10llPR0ll.CTIONSYSTEMINSTRUMENTATI0llSURVEILLANCEREQUIlkMENTS.

.n G,

CilANNEL OPERATIONAL ~

CllANNEL FUNCTIONAL CllANNEL CON 0lTIONS FOR WillCil J.

f ilt CilotlAl. tiillT CllECK TEST CALIDRATIOH(a) fkJRVEILLANCEREQUIRED l.

Intermediate Range Hiinitors:

IC) a.

fleutron Flux - liigh S/U,S,(la)

S/U

,W R

2 l

5 W

R 3,4,5 h.

Inoperative NA W

NA 2,3,4,5 Average Power Range Monitor (I):

2.

a.

lieutron Flux -

5/U.S,(b)

S/U(c),W SA 2

Upscale, Setdown S

.W SA 3, 5 b.

Flow liiased Simulated S/U

,W W(d)(e) SA,(RINI) 1 IIUI)

IC) lhermal Power - Upscale 5D 3

c.

Fixed ticutron Flux -

7l Upscale S

S/U(c),y y(<!), SA 1

NA 1, 2, 3, 5 l d.

Inoperative' NA W

j (e.

Ilownscale S

W' SA 1)

-l I

3.

Reactor Vessel Steam Home Pressure - liigh (S)

H (R) 1, 2 4.

Reactor Vessel Water Level -

Low, tevel 3 (S)

M (R) 1, 2 S.

Main Steam Line Isolation Valve - Closure NA H

R 1

6.

flain Steam Line Radiation -

II) liiuh S

H R

1, 2 7.

(Primary Containment) (Drywell) l Pressnre - liigh (S)

M (R) 1, E 5:.

~_

' 'J TABLE 4.3.1.1-1 (Continued) g REACIORPR0ffCIlottSYSTEMINSTRUMI:NTATIONSURVEILLANCEREQUllNMENTS TA OPERATIONAL CllANNEL

^

E CllANNEL FUNCTIONAL CilANNEL ONDITIONS FOR WilICil R

fullCTIONAI 0111I CllECK TEST CALIBRATION SiRVEILLANCE REQUIRED 3

11.

Scram Discharge Volume. Water Level - liigh (S)

M (R) 1, 2, 5(3) l

~

9.

Turbine Stop Valve - Closure (S)

M (R) 1 10.

Turbine Control Valve Fast Closure Valve Trip System Oil Pressure - Low (S)

M (R) 1 11.

Reactor Mode Switch Shutdown Position NA R

NA e'

1,2,3,4,5 12.

Manual Scram NA M

NA 1,2,3,4,5 T

co

..p Ya) Neutron detectors may be excluded from CilANNEL CAllDRATION.

(h) The IRM and SRM channels shall be determined to overlap for at least (I) decades during each startup s

after entering OPERAll0NAL CON 0lT10N 2 and the I[tM and APRM channels shall be determined to overlap for at least (h) decades during each controlled shutdown, if not performed within the previous 7 days.

(c) Within 24 liours prior to startup, if not. performed within the previous 7 days.

(d) lhis calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat halance during OPERATIONAL CONDITION 1 when TilEllMAL POWER > 25% of RATED lilERMAL POWER.

Adjust the APRM channel if the absolute difference is greater than 2% of RATED TilERMAL POWI.R.

Any APRM channel gain adjustment made in compliance with Specificatio'n 3.2.2 shall not be included in determining the absolute difference.

(c) This calibrat. inn shall consist 6f ti,ie adjustment of the APRM flow biased chanhel to conform to a z,

calibrated flew signal.

(f) The iPitth shall be calibrated at Icast once per 1000 ef fective full power hours (EFPil) usity; the TIP system.

(g) Verify measured core flow to be greater than or equal to established core flok at the existing pump speed.

((h) This calibration shall consist of (the adjustment, as requirc:d, of) (verifyinD) the 6 i I second simulated thermal power Lime constant.)

(i)

This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.

^

(j) tiith any control rod withdrawn.

Not applicable to control rods rem 6ved per Specification 3.9.10.1 or 3.9.10.2.

}.

t a

INSTRUMENTATION 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERdTION 3.3.6.

The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE wi.th their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.

APPLICABILITY: As shown in Table 3.3.6-1.

ACTION:

a.

With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent

, ith the Trip Setpoint value.

w b.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, take the ACTION required by Table 3.3.6-1.

SURVEILLANCE REQUIREMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.

l GE-STS (EWR/a) 3/4 3-47

t

'lABLE 3.3.6-1 y

CONTROL ROD BLOCK INSTRUMENTATION Yi MINIMUM APPL.ICABLE OPERABLE CHANNELS OPERATIONAL O

1 RIP I tillCl10ll PER TRIP FDHCTION CONDITIONS ACTION 1.

ROD !!!OCK 110til TOR ("I U

a.

Upscale 2

1*

60 13.

Inoperat.ive 2

1*

60 c.

Downscale 2

1*

60 2.

APRM a.

F low liiased Neutron Flux -

Upscale 4

1 61 li.

Inoperat.ive 4

1, 2, S 61 c.

Downscale 4

1 61 d.

Heut. con flux - Upscale, Startup 4

2, S 61 3.

50llRCE HAllGE MON 110RS in 'I 3'

2 61 II E'

a.

Detector not fulI 2

5 61 3

2 61 tipscale(c) b.

g 2

5 61 3

2 61 Inoperative (c) c.

2 5

61 d.

Downscale(d) 3-2 61 2

5 61 4.

INIERMEDIATE RANGE MONITORS Detector not full in (("))

6 2S 61 a.

li.

Upscale 6

2, S 61 Inoperati 6

2, S 61 c.

Downscale{g) 6 2, S 61 d.

S.

SCRAl1 DISCllARGE V01 time a.

Water Level-liigli (2) 1, 2, S**

62 ii.

Scram Trip Bypass (2)

(1, 2,) S**

62 6.

HEACIOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale 2

1 62 li.

Inoperative 2

1 62 c.

(Comparator) (Downscale) 2 1

62

l,;)

.?

TABLE 3.3.6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION ACTION ACTION 60 Declare the R8H inoperable and take the ACTION required by Specification 3.1.4.3.

ACTION 61 With the number of OPERABLE Channels:

a.

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channelto OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.

b.

Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.

ACTION 62 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.

NOTES With THERMAL POWER > (30)% of RATED THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

a.

The REM shall be automatically bypassed when a peripneral control roa is selected (or the reference APRM channel indicates less than (30)% of RATED THERMAL POWER).

b.

This function shall be automatically bypassed if detector count rate is

> 100 cps or the IRM channels are on range (3) or higher.

c.

This function shall be automatically bypassed when the associated IRM cnannels are on range 8 or higher.

d.

This function shall be autcmatically bypassed when the IRM channels are i

on range 3 or higher.

l e.

This function shall be attcmatically bypassed when the :RM channels are l

on range 1.

l GE-5TS (SkR/a) 3/4 3-49

TAllLE 3.3.6-2 rTi CONIR01. R00 BLOCK INSTRUMENTAll0N SETPOINTS

.~2

_l RIP l~tlNCI ION TRIP SETPolN1 ALLOWARLE VALUE I.

HOI) 111 OCK HONITOR 5 0.66 W'+ (40)%

$ 0.66 W + (43)%

R a.

Upscale 8

b.

Inoperative NA NA c.

Downscale 1 (5)% of RAlED TilERMAL POWER

> (3)% of RATED TilERMAL POWER 2.

-APRM a.

Flow liiased Neutron Flux -

tipscale 5 0.66 W + (42)%*

$ 0.66 W + (45)%"

b.

Inoperative NA NA c.

Downscale

> (S)% of RA1ED TilERMAL POWER

> (3)% of RATED TiiERMAL POWER d.

Neutron Flux - Upscale, Startup 5 (12)% of RATED TilERMAL POWER 2 (14)% of RATED TilERMAL POWER 3.

SOURC Q ANGE MON 110RS a.

Detes. tor not full in NA NA 0

b.

Upscale 5.(2 x 10 ) cps 5 (5 x 10 ) cps t'

c.

Inoperative NA NA

'i'

> (2) cps d.

Downscale

> (3) cps 4.

.lNIERMEDIAlE RANGE MONITORS a.

Detector not full in NA NA b.

Upscale 5 (108/125) divisions of 5 (110/125) divisions of full scale full scale c.

Inoperative NA NA d.

Downscale 1 (S/125) divisions of

> (3/125) divisions of full scale full scale 5.

SCRAM DISCllARGE VOLUME a.

W. iter l evel-liigh 5(

) inches 5( ) inches b.

Scram Trip Bypass NA NA 6.

REAC10R C001 ANT SYSTEM RECIRCULATION FLOW a.

tipscale

$ (108/125) divisions of 5 (111/125) divisions of

~

full scale full scale NA b.

Inoper ative NA 5 (11)% flow deviation c.

(Compar. Lor) (Downscale)

$L(10)% flow deviation lbe Avera0e Power Range Monit.or rod block function is varied as a funct. ion of recisculation loop flow E

The. trip setting of th.is ft,mction must be maintained in accordance with Spec 9fication 3.2.2.

(W).

y C

{5I TABLE 4.3.6-1 '

CONTROLRODBLOCKINSTRUMENIATIONSURVEILLANCEREQUIREMENT!!

m

',[ '

CilANNEL OPERATIONAL M

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR MIICN CHECK TEST CALIBRATION ")

SURVEILLANCE REQUIRED I

O' TRIP FUNCIION 1.

R0D Bl0CK MONITOR S/U((b)(c)

(c) ja q

a.

Upscale NA S/U(b)(c),(c)

NA 1*

b.

Inoperat.ive NA b)(c) (c) c.

Downscale NA S/U q

ja 2.

APRM a.

Flow Dias'ed Neutron Flux'-

1 (NA)

S/U

,M (Q) 1 Upscale.

.s.,

NA 1, 2, 5 b.

Intmarative '

NA S/U(b),M (NA)

S/U H

IO)

I c.

Downscale c

' d.

Neutron Flux -IJpscale, Startup (NA)

,S/U(b)'M (Q) 2, 5 D

3.

SOURCE RANGE MONITORS f

u w

a.

Detector not. full in NA 5/U(b) W NA 2, S

^

/

ki b.

Upscale NA S/U

,W Q

2, 5 S/U(b),W S/U NA -

2, 5 l

c.

Inoperative, NA

,W Q

2, 5 d.

Downscale

. NA j

4.

INIERNEDIATE RANGE MONIl0RS a.

Detector not full in NA S/U )W NA 2, 5 b.

Upscale NA S/U

,W Q

2, 5 c.

Inoperative NA S/Ug),W NA 2, S d.

Downscale NA S/U

,W Q

2, S S.

SCRAM OISCllARGE VOLUME a.

Water Level-liigh NA (H) (Q)

R 1, 2, S**

b.

Scrans Trip Bypass NA M

NA (1, 2.) S**

6.

REAC10R COOLANT SYSIEM RECIRCULATION FLOW a.

' Upscale NA S/U(b) M q

1 U.

b.

Inoperative NA G/U

,M NA 1

c.

(Comparator) (Downscale)

NA S/U

,H Q

1

f. '.

~*

t TABLE 4.3.6-1 (Continued)

CONTROL R00 BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTES:

a.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

b.

Within 24,, hours prior to startup, if not performed within the previous't oays..;

c.

Includes reactor manual control multiplexing system input.

With THERMAL POWER > (30)% of RATED THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

GE-STS (BWR/4) 3/a 3-52

t.,';

nz.

Lk

-l REFUELING OPERATIONS BASES 3/4.9.6 REFUELING PLATFORM The OPERABILITY requirements ensure that (1) the refueling platform will be used for handling control rods and fuel assemblies within the reactor pressure vessel, (2) each crane and hoist has sufficient load capacity for handling fuel assemblies and control rods, and (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL The restriction on movement of loads in excess of the nominal weight of a fuel assembly over other fuel assemblies in the storage pool ensures that in the event this load is dropped 1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible dictortion of fuel in the storage racks will not result in a critical. array.

This assumption is consistent with the activity release assumed in the safety analyses.

l 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL -SPENT FUEL STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove (99)% of 'the assumed (10)% fodine gap activity released from the rupture of an irradiated fuel assembly.

This minimum water depth is consistent with the assumptions of the accident analysis.

3/4.9.10 CONTROL ROD REMOVAL These specifications ensure that maintenance or repair of control rods or control rod drives will be performed under conditions that limit the probability

~

of inadvertent criticality.

The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.

3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal loop be OPERABLE or that an alternate method capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation ensures that 1) suf-ficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required during REFUELING, and 2) sufficient coolant circulation would be available through the reactor core to assure accurate temperature inoication and to distribute and prevent l

stratification of the poison in the event it becomes necessary to actuate the standby liquid control system.

The requirement to have two shutdown cooling mode loops OPERASLE when there is less than (23) feet of water above tne reactor vessel flange ensures that a single failure of the operating loco will not result in a cceplate loss of resic-ual heat removal capability.

With the reactor vessel head removed and (23) feet of water above the reactor vessel flange, a large heat sink is availacle for core cooling.

Thus, in the event a failure of the coerating RHR loop, adequate time is provided to initiate alternate methods capable of decay heat removal l

or emergency procedures to cool the core.

GE-STS (SWR /4)

S 3/* 9-2

l

~

l

}..

)

l.-

REFUELING OPERATIONS 3/4.9.10 CONTROL ROD REMOVAL SINGLE CONTROL R0D REMOVAL LIMITING CONDITION FOR OPERATION 3.9.10.1 One control rod and/or the associated control rod drive mechanism cay be removed from the core and/or reactor prest.ure vessel provided that at least the following requirements are satisfied until a control rod and associ-sttd tontrol md t.-iv-ru m aiieu are w e controi roa 1s.iuT1y inserted in the core.

The reactor mode switch is OPERABLE and locked in the Shutdown position a.

or in the Refuel position per Table 1.2 and Specification 3.9.1.

b.

The source range monitors (SRM) are OPERABLE per Specification 3.9.2.

The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied, c.

except that the control rod selected to be removed; 1.

May be assumed to be the highest worth control rod required to be assumed to be fully withdrawn by the SHUTDOWN MARGIN test, and 2.

Need not be assumed to be immovable or untrippable.

d.

All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.

e.

All other control rods are inserted.

APPLICABILITY:

OPERATIONAL CONDITIONS 4 and S.

ACTION:

With the requirements of the above specification not satisfied, suspend removal of the control rod and/or associated control rod drive mechanism frem the core and/or reactor pressure vessel and initiate action to satisfy the above requirements.

GE-5TS (EWR/4) 5/4 9-12

-s REFUELING OPERA 710NS SURVEILLANCE REOUIREMENTS 4.9.10.1 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start of removal of a control rod and/or the associated control rod drive mechanism from the core and/or reactor pressure vessel and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter until a control rod and associ-ated control rod drive mechanism are reinstalled and the control rod is inserted in the core, verify that; The reactor mode switch is OPERABLE and locked in the Shutdown position a.

or in the Refuel position with the "one rod out" Refuel positian interlock OPERABLE per Specification 3.9.1.

b.

The SRM channels are OPERABLE per Specification 3.9.2.

l The SHUTOOWN MARGIN requirements of Specification 3.1.1 are satisfied c.

per Specification 3.9.10.1.c.

l d.-

All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod or control rod drive mechanism to be removec from the core and/or reactor vessel are removed from the core cell.

e.

All other control rods'are inserted.

1 I

GE-STS (EWR/4) 3/4 9-13

.s

p. '.

d REFUELING OPERATIONS MULTIPLE CONTROL ROD REMOVAL LIMITINGCONDITIONFOROPERdTION 3.9.10.2 Any number of control rods and/or control rod drive mechanisms may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core.

The reactor mode switch is OPERABLE and locked in the Shutdown position a.

or in the Refuel Refuel position " position per Specification 3.9.1, except that the one-rod-out" interlock may be bypassed, as required, for those control rods and/or control rod drive mechanisms to be removed below., after the fuel assemblies have been removed as specified b.

The source range monitors (SRM) are OPERABLE per Specification 3.9.2.

The. SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.

c.

d.

All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.

The four fuel assemblies surrounding each control rod or control rod e.

drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.

l APPLICABILITY: OPERATIONAL CONDITION 5.

ACTION:

With the requirements of the above specification nct satisfied, suspend removal of control rods and/or control rod drive mechanisms frca the core and/or reactor pressure vessel and initiate action to satisfy the above requirements.

GE-STS (5WR/4) 2/4 9-14

REFUELING OPERATIONS SURVEILLANCE REOUIREMENTS i

4.9.10.2.1 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start of removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core, i

verify that:

i a.

The reactor mode switch is OPERABLE and locked in the Shutdown oosition I or in the Refuel position per Specification 3.9.1.

l b.

The SRM channels are OPERABLE per Specification 3.9.2.

c.

The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.

i d.

All other control rods are either inserted or have the surrounding 1

four fuel assemblies removed from the core cell.

4 e.

The four fuel assemblies surrounding each control rod and/or control rod drive mechanism to be removed from the core and/or reactor ressel 3

are removed from the core cell.

i 4.9.10.2.2 Following replacement of all control rods and/or control red drive mechanisms removed in accordance with this specification, perform a functional test of the "one-rod-out" Refuel position interlock, if this function had been bypassed.

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l GE-STS (EWR/4) 3/4 9-15

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INSTRUMENTATION l

BASES l

3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is l

provided to initiate actions to assure adequate core cooling in the event of I

reactor isolation from its primary heat sink and the loss of feedwater flow to

(

the reactor vessel without providing actuation of any of the emergency core

(

cooling equipment.

I Daaratinn with a trin sat Ms maavutius than 31.s Trin A a m 4e hu+

within its specified Allowable Value is acceptable.on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION l

The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits.

The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the l

difference 'between each Trip Setpoint and the Allowable Value is equal to or I

less than the drift allowance assumed for each trip in the safety analyses.

3/4.3.7 MONITORING INSTRUMENTATION

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3/4.3.7.1 RADIATION MONITORING INSTRUMENTATICN The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the l

individual channels; (2) the alarm or automatic action is initiated when the radi-ation level trip setpoint is exceeded; and (3) sufficient information is avail-able on selected plant parameters to monitor and assess these variables follow-ing an accident.

This capability is consistent with the recommendations of (NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980).

3.4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures that suf-ficient capability is available to promptly determine the magnitude of a seismic event and evaluate the. response of those features important to safety.

This cacability is required to permit comparison of the measured response to that used in the design basis for the unit.

(This instrumentation is consistent with the recommendations of Regulatory Guide 1.12 " Instrumentation for Earthquakes",

April 1974.)

3/4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION f

The OPERASILITY of the meteorological monitoring instrumentation ensures that sufficient meteorological data is available for estimating potential radia-tion cases to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.

(This instrumentation is consistent with the recommenda-tions of Regulatory Guide 1.23 "Onsite Mateoroicgical Programs," February 1972.)

GE-3TS (SWR /4)

B 3/4 3-4

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3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

Preserve the integrity of the fuel cladding.

a.

b.

Preserve the integrity of the reactor coolant system.

Minimize the energy which must be adsorbed following a loss of-coolant c.

accident, and d.

Prevent inadvertent criticality.

This specification orovides the limitina conditions for ooeration necessary to preserve'the ability 'f the system to perform its intended ' function evan

~

o during periods when instrument channels may be out of service because of main-tenance.

When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system i's made up of 'two independent trip systems.

There are usually four channels to monitor each parameter with two channels in each trip system.

The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram.

The system meets the intent of.IEEE-279 for nuclear power plant protection systems. The bases for the trip j

settings of the RPS are discussed in the bases for Specification 2.2.1.

The measurement of response time at the specified frequencies provides assurance thct tua protective functions associated with each cnannel are com-pleted within the time limit assumed in the accident analysis.

No credit was taken for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, er (2) utilizing replacement sensors with certified response times.

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GE-STS (SWR /4)

B 3/4 3-1

ML20024B407

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