ML20077Q366
| ML20077Q366 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/10/1995 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20077Q358 | List: |
| References | |
| NUDOCS 9501190227 | |
| Download: ML20077Q366 (14) | |
Text
m _r
- March 1 1991.
TABLE 2.2-1 ~,
2 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS -
P '
TOTAL SENSOR h ALLOWANCE ERROR M FUNCTIONAL UNIT fTA) I fS) TRIP SETPOINT ALLOWABLE VALUE
- l. Nanual Reactor Trip N.A. N.A. N.A. N.A. N.A.
$$ z mo p 2. Power Range, Neutron Flux ,
@$ w a. High Setpoint on EU 1) Four Loops Operating 7.5 4.56 0. 1 109% of RTP** 1 111.1% of RTP**
Sj 2) Three Loops Operating 7.5 4.56 0 s 80% of RTP** s 82.1% of RTP**
oo
- b. Low Setpoint Q 8.3 4.56 0 s 25% of RTP** s 27.1% of RTP**
"U 3. Power Range, Neutron Flux, 1.6 0.5 0 s 5% of RTP** with i 6.3% of RTP** with Nigh Positive Rate
- a time constant a time constant
,m 1 2 seconds 2 2 seconds 4, r Range, Neutron Flux, 1.6 0.5 0 $ 5% of RTP** with 5 6.3% of RTP** w15
\)ligh Negative Rate -
a time constant a time constant
_ Q 2 seconds 2 2 seconds j
- 5. Intermediate Range, 17.0 8.41 0 5 25% of RTP** 5 30.9% of RTP**
Neutron Flux
- 6. Source Range, Neutron Flux 17.0 10.01 0 $ 10+5 cps 5 1.4 x 10+5 cps
$g y.
- 7. Overtemperature AT - -
E
" a. Four Loops Operating z
F 1) Channels 1, II 10.0 6.80 1.71 + 1.33 See Note 1 ' See Note 2 h (Temp + Press)
.h 2) Channels III, IV 10.0 5.83 1.71 + 2.60 See Note 1 See Note 2 (Temp + Press) l@ **RTP = RATED THERMAL POWER l
l l
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i.IMITINGSAFETYSYSTEMSETTINGS JAN 31 1986 BASES '
)
REACTOR TRIP SYSTEN INSTRUMENTATION SETPOINTS (Continued)
The various Reactor trip circuits automatically open the Reacto,r trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to redundant channels and trains, the
- desjgn approach provides a Reactor Trip System which monitors numerous system variables, therefore providing Trip System functional diversity. The functional capability at the specified trip setting is required for those anticipatory or diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.
1 Manual Reactor Trip l l
The Reactor Trip System includes manual Reactor trip capability. I 4
Power Rance Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip
) '
setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power i operations to mitigate the consequences of a reactivity excursion from all !
power levels. The High Setpoint trip is reduced during three loop operation to a value consistent with the safety analysis.
The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.
Powhh Power Rance, Neutron Flux, High, Rat 5
The Power Range Positive Rate trip provides protection against rapid flux -
increases which are characteristic of a rupture of a control rod drive housing.
~
Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for all rod ejection accidents.
The Power Range Negative Rate trip provides protection for control rod
. drop accidents. At high power a multiple rod drop accident could cause local -
flux peaking which could cause an unconservative local DNBR to exist. The Power !
Range Negative Rate trip will prevent this from occurring by tripping the reac-tor. No credit is taken for operation of the Power Range Negative Rate trip )
for those control rod drop accidents for which DNBRs will be greater than 1.30.
MILLSTONE - UNIT 3 B 2-4 ,
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1 A.
March'11, 1991 -
4 TABLE 3.3-1 a .
y u, REACTOR TRIP SYSTEM INSTRUMENTATION y MINIMUM .
m TOTAL NO. CHANNELS CHANNELS APPLICABLE :
i . FUNCTIONAL UNIT .0F_ CHANNELS TO TRIP OPERABLE MODES ACTION e
5
- 1. Manual -Reactor Trip 2 1 2 1,.2 1 2 1 2 3*, 4*, 5* 11 w
- 2. Power Range, Neutron Flux ,
- a. High Setpoint 4 2 3 1, 2 2
- b. Low Setpoint 4 2 3 l#ff, 2 2
- 3. Power Range, Neutron Flux 4 2 3 1, 2 2 High Positive Rate '
U$&N 4.Q Power Range, Neutron Flux, 4 2 3 1, 2 2 g High Negative Rate *
- 5. Intermediate Range, Neutron Flux 2 1 2 Ifff, 2 3
- 6. Source Range, Neutron Flux
- a. Startup 2 1 2 2ff 4 b ~. Shutdem 2 1 2 3*, 4*, 5* 11 ,
- 7. Overtemperature AT
- a. Four Loop Operation 4 2 3 1, 2 6 "
g b. Three Loop Operation 3 2 2 1, 2 6 a
g , .s' . 8. Overpower AT m..- a. Four Loop Operation 4 2 3 1, 2 6 t
" Three Loop Operation 1, 2
- b. 3 2 2 6 ,
o
- 9. Pressurizer Pressure--Low 4 2 3 1** e 6 (1)
D g 10. Pressurizer Pressure--High 4 2 3 1, 2 6 (1)
(d) 11. Pressurizer Water Level--High 3 2 2 1** 6
, L
c TABLE 4.3-1 33 REACTOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE RE0UIRDENTS
- P Q
g ANALOG TRIP m . ,
ACTUATING MODES FOR
- CHANNEL DEVICE CHANNEL NNICH CH4NNEL OPERATIONAL OPERATIONAL ACTURTION E FUNCTIONAL UNII CHECK CALIBRATION IEST TEST SURVEILLANCE
- LOGIC TEST IS REOUIRED
[ 1. Manual Reactor Trip N.A. N.A. N.A. R(14,20) N.A. 1, 2, 3*, 4*,
5*
- 2. Power Range, Neutron Flux '
- a. High Setpoint S D(2,4), Q N.A. N.A. 1, 2 N(3,4),
$d4g Q(4,6),
R(4,5,20)
- b. Low Setpoint S R(4,20) S/U(1) N.A. N.A. 1***, 2
$ 3. Power Range, Neutron Flux, M.A. R(4,20) Q N.A. N.4. 1, 2 High Positive Rate _
o 4. Power Range, Neutron Flux, M.A. R(4,20) Q N.A. M.A. 1, 2 gh Negative Rate
- 5. Intermed ate Range 5 R(4,5) S/U(l) N.A. N.A. 1***, 2
- 6. Source Range, Neutron Flux S R(4,5) S/U(1), M.A.
M.A. 2**, 3, 4, Q(9) 5 5 '
- g. 7. Overtemperature AT S R Q N.A. N.A. 1, 2 2
- y. 8. Overpower AT S R Q N.A. N.A. 1, 2 l .I 9. Pressurizer Pressure--Low S R(20) Q(18) N.A. N.A.
, 1
$co N^ 10.' Pressurizer Pressure--High 5 R(20) Q(18) N.A. N.A. 1, 2 D
Y , 11. Pressurizer Water level--High' S R(20) . N.A.
t u Q. N.A. -1 -
12 teactor Coolant Flow--Low S R (20' Q N.A. N.A.
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l Docket No. 50-423 B14931 !
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Attachment 2 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications Reactor Trip System Retyped Up Pages N
January 1995
TABLE 2.2-1 ,
REACTOR TRIP SYSTEN INSTRUNENTATION TRIP SETPOINTS
@5 "F TOTAL SENSOR M ALLOWANCE ERROR .
E FUNCTIONAL UNIT (TA) I (Si TRIP SETPOINT ALLOWABLE VALUE
- l. Nantal Reactor Trip N.A. N.A. N.A. N.A. N.A.
E Q 2. Power Range, Neutron Flux w
- a. High Setpoint
- 1) Four Loops Operating 7.5 4.56 0 5; of RTP** s 111.1% of RTP**
- 2) Three Loops Operating 7.5 4.56 0 s 80% of RTP** s 82.1% of RTP**
- b. Low Setpoint 8.3 4.56 0 s 25% of RTP** 1 27.1% of RTP**
- 3. Power Range, Neutron Flux, 1.6 0.5 0 $ 5% of RTP** with 16.3% of RTPa* with ,
High Positive Rate a time constant a time constant y 1 2 seconds 1 2 seconds an
- 4. Deleted
- 5. Intemediate Range, 17.0 8.41 0 s 25% of RTP** 1 30.9% of RTP**
Neutron Flux
- 6. Source Pange, Neutron Flux 17.0 10.01 0 s 10+' cps s 1.4 x 10+' cps
- 7. Overtcresrature AT f
s a. Frnr Loops Operating a
- 1) Channels I, II 10.0 6.80 1.71 + 1.33 See Note 1 See Note 2 (Temp + Press)
. 2) Channels III, IV 10.0 5.83 1.71 + 2.60 See Note 1 See Note 2 (Temp + Press)
=
- **RTP = RATED THERNAL POWER
LINITING SAFETY SYSTEN SETTINGS
. i BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)
The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore providing Trip System functional diversity. The functional capability at the specified trip setting is required for those anticipatory or diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.
Manual Reactor Trio The Reactor Trip System includes manual Reactor trip capability.
Power Rance. Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursio'n beginning from low power, and the High Setpoint trip prcvides protection during power operations to mtU gate the consequences of a reactivity excursion from all power levels. The High Setpoint trip is reduced during three loop operation to a value consistent with the safety analysis.
The Low Setpoint trip may be manually blocked above P-10 (a power level !
of approximately 10% of RATED THERMAL POWER) and is automatically reinstated ,
below the P-10 Setpoint. l l
Power Rance. Neutron Flux. Hiah Positive Rate l The Power Range Positive Rate trip provides protection against rapid flux l increases which are characteristic of a rupture of a control rod drive housing.
Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for all rod ejection accidents.
NILLSTONE - UNIT 3 B 2-4 Amendment No.
0208
TABLE 3.3-1 .
! REACTOR TRIP SYSTEM INSTRUMENTATION N MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE 7 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E 1. Manual Reactor Trip 2 1 2 1, 2 1 Z 2 1 2 3*, 4*, 5* 11 w
- 2. Power Range, Neutron Flux
- a. High Setpoint 4 2 3 1, 2 2
- b. Low Setpoint 4 2 3 liff, 2 2
- 3. Power Range, Neutron Flux 4 2 3 1, 2 2 High Positive Rate
- 4. Deleted
- 5. Intermediate Range, Neutron Flux 2 1 2 1#ff, 2 3 s
- Source Range, Neutron Flux 6.
Y a. Startup 2 1 2 2ff 4 N b. Shutdown 2 1 2 3*, 4*, 5* 11
- 7. Overtemperature AT
- a. Four Loop Operation 4 2 3 1, 2 6
- b. Three Loop Operation 3 2 2 1, 2 6
- 8. Overpower AT
- a. Four Loop Operation 4 2 3 1, 2 6
- b. Three Loop Operation 3 2 2 1, 2 6
$3 9. Pressurizer Pressure--Low 4 2 3 1** 6 (1)
- 10. Pressurizer Pressure--High 4 2 3 1, 2 6 (1) g: 11. Pressurizer Water Level--High 3 2 2 1** 6 4
- , -_w- . . . . - - . . - _ .-_-y, - - -- . , _ _ _ , . - _y
25 .
- F TABLE 4.3-1 M
g REACTOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE REQUIRENENTS TRIP g ANALOG ACTUATING N0 DES FOR CHANNEL DEVICE WHICH
- CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE w FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
- 1. Nanual Reactor Trip N.A. N.A. N.A. R(14,20) N.A.
g 2, 3*, 4*,
- 2. Power Range Neutron Flux Highfetpoint a.
( ,, ), N.A. N.A. 1, 2 S D Q NL J,
- b. Low Setpoint S b'6)20)
R[ , S/U(1) N.A. N.A. 1***, 2
- 3. Power Range, Neutron Flux, N.A. R(4,20) Q N.A. N.A. 1, 2 High Positive Rate g 4. Deleted
[5.
Intermediate Range S R(4,5) S/U(1) N.A. N.A. 1***, 2 5 6. Source Range, Neutron Flux S R(4,5)
S/(U[1),
M.A. N.A. 2**, 3, 4, Q 9) 5
- 7. Overtemperature AT S R Q N.A. N.A. 1, 2
- 8. Overpower AT S R Q N.A. N.A. 1, 2
- 9. Pressurizer Pressure--Low S R (20) Q(18) N.A. N.A. I
- 10. Pressurizer Pressure--High S R (20) Q(18) N.A. N.A. 1, 2 Pressurizer Water Level--High S R (20) N.A. N.A. I f11. Q g12. Reactor Coolant Flow--Low S R (20) Q N.A. N.A. 1 5
.ilf
, . _ _ - ,m - _ _ _ _ __
r- p
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.. i e
Docket No. 50-423 l B14931 )
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l Attachment 3 Millstone Nuclear Power Station, Unit No. 3 I Plant Safety Evaluation for Millstone Generating !
Station, Unit 3 VANTAGE 5H Fuel Upgrada-Section 5.1.6.3 l l
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January 1995 l
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'5.1.6.3 Rod Cluster Control Assembly Misalionment 5.1.6.3.1 Introduction FSAR Section 15.4.3.1 contains a general discussion of the RCCA misalignment accidents that identifies causes and provides descriptions of the specific events involved.
As presented there, the RCCA misalignment accidents include:
- 1. One or more dropped RCCAs within the same group.
- 2. A dropped RCCA bank.
- 3. Statically misaligned RCCA.
- 4. Withdrawal of a single RCCA.
The dropped RCCA, dropped RCCA bank, and statically misaligned RCCA events are classified as ANS Condition 11 accidents,i.e., faults of moderate frequency. The single RCCA withdrawal is classified as an ANS Condition til event, i.e., an infrequent fauft. The general acceptance criteria for these event categories are discussed in FSAR Section 15.0.1.
For the three Condition ll events, the analysis is performed to verify that the DNBR design basis is met. As explained in FSAR Section 15.4.2.4, since the single RCCA withdrawal event is a Condition til accident. the applicable general criteria allow a small fraction of the fuel to experience damage. For Millstone Unit 3, the current licensing basis analysis reports that an upper bound on the number of fuel rods predicted to experience DNB is 5% of the total fuel rods in the core. The single RCCA withdrawal analysis for the VANTAGE SH transition uses the same 5% limit as a measure of acceptability for this event.
The transition to VANTAGE SH fuel, itself, has almost no impact on the discussion in FSAR Section 15.4.3 with regard to the various RCCA misalignment accidents. For example, the means of detecting a dropped RCCA bank, misaligned RCCA, or single RCCA withdrawal remain just as described in FSAR Section 15.4.3.1. However, certain of the modified safety analysis assumptions, while not impacting the general methodology, do alter the explicit DNB related calculations for a given case. Among the assumptions falling into this category are the use of the Revised Thermal Design Procedure, the increased RCCA drop time for the VANTAGE SH fuel, and the increased F3 s power distribution peaking factor.
There is also one modified assumption that has a more direct impact on RCCA misalignment analysis. With regard to the dropped RCCA transient (one or more dropped RCCAs), the VANTAGE SH transition analysis does not take credit for any direct reactor trip or for an automatic power reduction due to the dropped RCCA(s). Specifically, no credit is taken for the negative flux rate trip which, in the current FSAR licensing basis analysis, is assumed to produce a reactor trip for dropped RCCA(s) that produce a ]
reactivity insertion greater than 400 pcm. The methodology represented by this approach to the dropped RCCA event is fully documented in Reference 29.
o0309-10 10 4 80890 5-72 j l
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f 1 V 5.1.6.3.2 Method of Analysis As in the current licensing basis analyses for the RCCA misalignment events, the cases considered are bounding for both four and three loop operating conditions. For evaluation of DNB during the drcpped RCCA event (one or more dropped RCCAs), the general analytical methods described in FSAR Section 15.4.3.2 are modified to be consistent with no credit being taken for the negative flux rate trip. . Additionally, the statepoints calculated for evaluating the DNB design basis are based on plant conditions as defined by RTDP and a hot channel peaking factor consistent with the increased FAH assumed for the VANTAGE SH transition. The transient response, nuclear peaking factor analysis, and DNB design basis confirmation are performed in accordance with the methodology of Reference 29. For Millstone Unit 3, the use of Reference 29 as the basis for the analysis ,
represer.ts the replacement of the current licensing basis methodology of Reference 30 which is discussed in the FSAR.
A dropped RCCA bank results in a symmetric power change in the core. As discussed in Reference 29. assumptions made for the dropped RCCA(s) analysis provide a bounding analysis fer the dropped RCCA bank.
For the statically misaligned RCCA, the DNB evaluation includes consideration of steady-state power distributions analyzed using the computer codes and general methods as l p described in FSAR Section 15.4.3.2. However, the initial condition assumptions used are V nominal power, nominal RCS pressure, nominal RCS average temperature, and minimum measured flow as dictated by RTDP. Uncertainties in these initial conditions are included in the limit DNBR as discussed in Section 4.0. In contrast, the current FSAR analysis for this event uses thermal design flow and is based on power, pressure, and temperatures which directly reflect the uncertainties as required by STDP.
For the single RCCA withdrawal event, the calculation of power distributions and the associated peaking factors within the core, is performed using the methods and computer codes defined in FSAR Section 15.4.3.2. For the VANTAGE SH transition, the determination of the minimum DNBR includes consideration of initial conditions that are defined consistent with the RTDP methodology.
5.1.6.3.3 Results Single or multiple dropped RCCAs within the same group result in a negative reactivity i insertion. The core is not adversely affected during this period, since power is decreasing I rapidly. Power may be re-established either by reactivity feedback or control bank l
< withdrawal.
l Following a dropped RCCA event initiated from manual rod control, the plant will establish a niew equilibrium condition. The equilibrium process without control system interaction is monotonic, thus removing power overshoot as a concern and establishing the automatic l v
] rod control mode of operation as the limiting case.
1 00309-t 0.1 D 000890 . 5 73
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o For a dropped RCCA event in the automatic rod control mode, the rod control system detects the drop in power and initiates control bank withdrswal. Power overshoot may occur due to this action by the automatic rod controller after which the control system will insert the control bank to restore nominal power. In all cases, the minimum DNBR remains above the limit value.
Following plant stabilization normal rod retrieval or shutdown procedures are followed.
The operator may manually retrieve the RCCA by following approved operating procedures.
A dropped RCCA bank typically results in a negative reactivity insertion greater than 400 pcm. The core is not adversely affected during the insertion period, since power is decreasing rapidly. The transient will proceed as described above, however, the return to power will be less due to the greater worth of an entire bank. The analysis demonstrates that a reactor trip is not required to mitigate the consequences of the transient. Following plant stabilization, normal rod retrieval or shutdown procedures are followed to further cool down the plant.
Consistent with the description for statically misaligned RCCA found in FSAR Section 15.4.3.2, the most limiting misalignment case is considered. As noted above, the analysis for the VANTAGE SH transition considers initial conditions based on RTDP The DNB design is confirmed to be met for the RCCA rr isalignment incident, and thus the ability of the primary coolant to remove heat from ti e fuel rod is not reduced.
Following the identification of an RCCA group misalignment condition by the operator, the operator is required to take action as required by the plant Technical Specifications and operating instructions.
I For the single rod withdrawal event, the two cases found in the FSAR have been considered. Those are single RCCA withdrawal with the reactor in automatic and manual ;
control. For both of these cases, the analysis demonstrates that although a reactor trip on !
the overtemperature AT trip would actually be expected, that trip does not occur sufficiently fast in all instances to prevent the minimium DNBR in the core from falling below the safety analysis limit. The evaluation for this event at the power and coolant !
conditions at which the overtemperature aT trip would be expected to trip the plant shows l
that an upper limit for the number of rods with a DNBR less than the limit value is 5%. I 5.1.6.3.4 Conclusions For cases of dropped RCCAs or dropped RCCA banks, the DNBR remains greater than the limit value (see Table 4-2). Therefore, the DNB design basis is met and the conclusions of the FSAR for this event remain valid.
O' 00309-10.10.000890 5-74
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. For the limiting cases associated with the statically misaligned RCCA event the ONB ,
. remains above the safety analysis limit value (see Table 4-2). Therefore, the DNB design i basis is met and the conclusions of the FSAR remain valid for this event. j For the accidental withdrawal of a single RCCA event, with the reactor in either automatic ;
or manual control mode and initially operating at full power, an upper bound on the number of fuel rods experiencing DNBR is 5 percent of the total fuel rods in the core. ,
tha R su n an in reas d Re c or C la Flo te This subsection is not applicable to Millstone Unit 3. t 5.1.6.5 Inadvertent Loadina and Operation of a Fuel Assembly in an Imorocer Position ,
The in-core instrumentation's ability to detect gross differences between measured and . I predicted thimble reaction rates is unaffected by fuel typei therefore, the conclusions of the FSAR remain valid. [
5.1.6.6 Spectrum of Rod Cluster Control Assembly Election Accidents 5.1.6.6.1 Introduction f This accident, as discussed in FSAR Section 15.4.8, is defined as the mechanical failure of a control rod mechanism pressure housing resulting in the ejection of a RCCA and !
drive shaft. The consequence of this mechanical failure is a rapid positive reactivity i insertion together with an adverse core power distribution, pbssibly leading to localized l fuel rod damage, j This event W classified as an ANS' Condition IV event, i.e. a limiting fault. The general ,
acceptance criteria for this event category are discussed in FSAR Section 15.0.1,4. The [
specific limiting criteria evaluated in the analysis for this event are summarized as follows: I
- 1. Average fuel pellet enthalpy at hot spot below 225 cal /g for unirradiated fuel and 200 cal /g for irradiated fuel. f l
- 2. Peak reactor coolant pressure less than that which could cause stresses to :
exceed the faulted condition stress limits. l t
- 3. Fuel melting will be limited to less than ten percent 10% of the fuel volume at !
the hot spot even if the average fuel pellet enthalpy is below the limits of l criterion 1 above. !
It should be noted that the current FSAR includes an additional criterion that the average clad temperature at the hot spot must remain below 2700 'F. The elimination of the clad O .
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00309-10 10 080890 5-75
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