ML20214S922
| ML20214S922 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 09/23/1986 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214S924 | List: |
| References | |
| TAC-61264, TAC-61518, NUDOCS 8609300139 | |
| Download: ML20214S922 (17) | |
Text
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UNITED STATES p,
NUCLEAR REGULATORY COMMISSION t
j WASHINGTON, D. C. 20555
%...../
MISSISSIPPI POWER & LIGHT COMPANY MIDDLE SOUTH ENERGY, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 19 License No. NPF-29 1.
The Nuclear Regulatory Commission _(the Commission) has found that A.
The applications for amendment by Mississippi Power & Light Company.
Middle South Energy, Inc., and South Mississippi Electric Power Association, (the licensees) dated April 14, 1986 and May 12, 1986, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application. the i
provisions of the Act, and the rules and regulations of the Commission; There is reasorable assurance (i) that the activities authorized by C.
this amendtr,ent can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Coninission's regulations; a
D.
The issuance of this amendment will not be inimical to the common ly defense and security or to the health and safety of the public; and
?
j~
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, l3 and paragraph 2.C.(2) of Facility Operating License No. NPF-29 is hereby l
amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 19, are hereby incorporated into this
. license. Mississippi Power & Light Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
8609300139 860923 DR ADOCK 0500 6
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4
. ' 3.
. Technical' Specification Pages 3/4 3-18, 3/4 3-88, B 3/4 3-2, 5-2 and 5-6 in this amendment are effective upon issuance of this amendment. The change made on Technical Specification Page 3/4 3-71 is effective when the equipment necessitating the change on this page is installed and made operable.
FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:
Walter R. Butler, Director BWft Project Directorate No. 4 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 23, 1986 Previously concurred *:
PD#4/LA*
PDf4/PM*
OGC*
PD#4/D l
M0'Brien LKintner:lb MYoung WButler p 09/10/86 09/10/86 09/11/86
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ATTACHMENT TO LICENSE AMENDMENT NO. 19 FACILITY OPERATING LICENSE NO. NPF-29 DOCKET NO. 50-416 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleafpage(s) provided to maintain document completeness.*
Remove Insert 3/4 3-17 3/4 3-17*
3/4 3-18 3/4 3-18 3/4 3-71 3/4 3-71 3/4 3-72 3/4 3-72*
3/4 3-87 3/4 3-87*
3/4 3-88 3/4 3-88 B 3/4 3-1 B 3/4 3-1*
B 3/4 3-2 B 3/4 3-2 5-1 5-1*
5-2 5-2 5-2a 5-5 5-5*
5-6 5-6 l
l
a l
1
!{
-44 l
TABLE 3.3.2-2 (Continued)
ISOLATION ACTUATION INSTRUNENTATION SETPOINTS c,
E ALLOWA8LE T
TRIP FUNCTION TRIP SETPOINT VALUE E
e Q
2.
MAIN STEAM LINE ISOLATION (Continued) g.
Main Steam Line Tunnel a Temp. - High 1 101'F**
1 104'F**
h.
Manual Initiation NA NA 3.
SECONDARY CONTAINMENT ISOLATION a.
Low Low, Level 2 1 -41.6 inches
- 1 -43.8 inches b.
Drywell Pressure - High 5 1.23 psig i 1.43 psig c.
Fuel Handling Area Ventilation
{
Exhaust Radition - High High
< 3.6 mR/hr**
1 4.0 mR/hr**
d.
Fuel Handling Area Pool Sweep wy Exhaust. Radiation - High High 1 30 mR/hr**
1 35 mR/hr**
e.
Manual Initiation NA NA
{
4.
REACTOR WATER CLEANUP SYSTEM ISOLATION j
A, Flow - High 5 79 gpm 5 89** gpa j
a.
b.
A Flow Timer 5 45 seconds 1 57 seconds c.
Equipment Area Temperature - High i
1.
RWCU Hx Room
< 120*F
< 126'F 2.
RWCU Pump Rooms I 170*F E 176*F'
~~
3.
RWCU Valve Nest Room 5135'F 5141*F d.
Equipment Area a Temp. - High 1.
RWCU Hx Room
< 65'F
< 66*F i
2.
RWCU Pump Rooms 7 115'F I 118'F 3.
RWCU Valve Nest Room 570*F 573*F e.
Reactor Vessel Water Level - Low Low, Level 2 1 -41.6 inches
- 1 -43.8 inches t
4
h 3,fi il l
n TABLE 3.3.2-2 (Continued)
ISCLATION ACTUATION INSTRUNENTATION SETPOINTS 8
ALLOWA8LE g;
TRIP FUNCTION TRIP SETPOINT VALUE 4.
REACTOR WATER CLEANUP SYSTEM ISOLATION (Continued)
[
f.
Main Steam Line Tunnel Ambient Temperature - High 1 185'F**
1 l')1*F**
g.
Main Steam Line Tunnel A Temp. - High 1 101*F**
1 104*F**
h.
SLCS Initiation MA NA 1.
Manual Initiation MA NA 5.
REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.
RCIC Steam Line Flow - High I
R 1.
Pressure
< 56" H O
< 64" H O l
2 2
2.
Time Delay 5 2 seconds 512 seconds i.
l b.
RCIC Steam Supply Pressure - Low 1 60 psig 1 53 psig c.
RCIC Turbine Exhaust Diaphragm Pressure - High 1 10 psig i 20 psig d.
RCIC Equipment Room Ambient Temperature - High 1 185'F**
1 191*F**
e.
RCIC Equipment Room A Temp. - High
< 125'F**
< 128'F**
f.
Main Steam Line Tunnel Ambient t
Temperature - High i 185'F**
1 191*F**
i g.
Main Steam Line Tunnel a Temp. - High 1 101*F**
1 104*F**
I h.
Main Steam Line Tunnel Temperature Timer 1 30' minutes 1 30 minutes i.
RHR Equipment Room Ambient Temperature -
j g
High i 165'F**
1 171*F**
1 g
l j.
RHR Equipment Room a Temperature -
g High 1 99*F**
1 102*F**
k.
RHR/RCIC Steam Line Flow - High 5 145" H O
$ 151" H 0 2
p L
m..
TABLE 3.3.7.4-1 (Continued)
I REMOTE SHUTDOWN SYSTEM CONTROLS MINIMUM CHANNELS OPERABLE CONTROL Div 1 Div 2 b
b 12.
RHRInjectionValves 2
2 13.
RHR Test Line Valve 1
1 14.
1 15.
RHR HX Flow to Suppression Pool Valve 1
1 16.
RHR Discharge to Radwaste Valve 1
1 b
b 17.
2 18.
Diesel Generator HX Inlet Valve 1
1 b
b 19.
Safety / Relief Valves 6
6 20.
Control Room to Shutdown Panel Transfer 1
NA Switch 21.
RCIC Turbine Flow Controller 1
NA 22.
RCIC Suction Flow Suppression Pool Valve 1
NA 23.
RCIC Injection Shutoff Valve 1
NA 24.
NA 25.
RCIC Recire. Main Flow Bypass Valve 1
NA l
l 26.
RCIC Test RTN to CST IB Valve 1
NA 27.
RCIC Test RTN to CST OB Valve 1
NA 28.
Steam to RCIC Turbine Valve 1
NA 29.
RCIC Turbine Trip & Throttle Valve 1
NA 30.
RCIC Turbine Cooling Water Valve 1
NA 31.
RCIC Turbine Local Control Select Switch 1
NA 32.
RCIC Gland Seal Compressor 1
NA 33.
Shutdgwn Cooling Isolation Valve Reset Switch 1 1
NOTE:
a.
1 per cooling tower fan b.
1 per valve GRAND GULF-UNIT 1 3/4 3-71 Amendment No.19 l
Effective date:
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INSTRUMENTATION TABLE 4.3.7.4-1 RfMOTESHUTDOWNMONITORINGINSTRUMENTATIONSURVEILLANCEREQUIREMENTS CHANNEL CHANNEL INSTRUMENT CNECK CALIBRATION 1.
Reactor Vessel Pressure M
R 2.
R 3.
Suppression Pool Water Level M
R 4.
Suppression Pool Water Temperature M
R
~
5.
RHR System Flow M
R 6.
Standby Service Water System Flow M
R 7.
RCIC Turbine Speed M
R 8.
Condensate Storage Tank Level M
R I
t i
l t
l l
O GRAND GULF-UNIT 1 3/4 3-72
.. c,,.. ~........
TABLE 3.3.7.9-1 (Continued)
FI N STRUMENTATION MINIMUM INSTRUMENTS OPERABLE
- ll)
SM0 Kell):
R00M ELEV ROOM NAME NEAT FLAME d
W)
M M
- 12. Zone 2-17 16/0 F
1A101 93' Passage 1A109 93' NPCS Pump Rm.
IA111 93' Piping Penetration Rs.
1A114 93' Fan Coil Area (Partial) 1A117 93' Misc. Equip. Area (Partial) 1A121 103' East Corridor 1A122 103.. South Corridor (Partial) 1A123 103' North Corridor (Partial)
'13. Zone 2-18 20/0 1A201 119' East Corridor IA211 119' North Corridor (Partial) 1A215 119' South Corridor (Partial)
- 14. Zone 2-19 13/0 1A314 139' So'uth Corridor (Partial) 1A316 139' North Corridor (Partial)
IA321 139' MCC Area IA322 139' Centrifugal Chiller Area IA323 139' SGTS Area IA324 139' HVAC Equip. Area IA326 139' SGTS Area
- 15. Zone 2-20 2/0 IA305 139' Steam Tunnel
~
- 16. Zone 2-21 4/0 IA12 185' Stairwell
~
i IA12 208' Stairwell IA12 245' Stairwell d.
DIESEL GENERATOR BUILDING l
1.
Zone 2-10 9/0 1D301 133' Cctridor 0/3 (Deluge)
~
1D304 133' Day Tank Area ID306.
133'..*Div. III Diesel Gen. Room 3D401 158' Div. III Diesel Gen.
Room 0/7 (Deluge) 2.
2-11 6/0 ID3b3 133' D e Tank Area 1D308 133'.
Div. II Diesel Gen. Room 1D402 -
158' Div. II Diesel Gen.
Room O/7 (Deluge)
GRAND GULF-UNIT 1 3/4 3-87 Amendment No. 4 Effective Date:S A pril S 6
TABLE 3.3.7.9-1 (Continued)
FINE DETETEFIRSTRUMENTATION MINIMUM INSTRUMENTS OPERABLE
- ll) ll)
SM0KE ROOM ELEV ROOM NAME HEAT FLAME II7Y)
II777 II7YI 3.
Zone 2-12 6/0 1D302 133' Day Tank Area 1D310 133' Div. I Diesel Gen. Room 1D403 158' Div. I Diesel Gen. Room 0/7 (Deluge) e.
STANDBY SERVICE WATER PUMP HOUSE 1.
Zone 2-1 4/0 1M110 133' SSW Pump Rm. A 1M112 133' SSW Valve Rm. A 2H110 133' SSW Pump Rm. B 2M112 133' SSW Valve Rm. B f.
CHARCOAL FILTER TRAINS 1.
Standby Gas Treatment System Filter Trains A & B 2/0 (Allison Thermistor Wire)
Auxiliary Building E1. 139' 2.
Control Room Standby Fresh Air System Filter Trains A & B 2/0 (Allison Thermistor Wire)
Control Building E1. 133' g.
CONTROL BUILDING (PGCC HALON SYSTEMS)
OC503 166' Control Room (Unit 1 side)
Module /Halon Panel 1H13-U700/1H13-P900 0/10 10/0 1H13-U701/1H13-P901 0/10 15/0 1H13-U702/1H13-P902 0/9 14/0 1H13-U703/1H13-P903 0/11 17/0 1H13-U720/1H13-P920 0/7 13/0 SH13-U730/SH13-P930 0/11 12/0 1H13-U738/1H13-P938 0/10 12/0 SH13-U739/SH13-P939 0/5 14/0 GRAND GULF-UNIT 1 3/4 3-88 Amendment'No. 19 l
l l
o I
3/4 L INSTRUMENTATION I
~
BASES 3/4.3.1. REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:
a.
Preserve the integrity of the fuel cladding.
[
b.
Preserve the integrity of the reactor coolant system.
Minimize the energy which must be absorbed following a loss-of-coolant j
c.
accident, and d.
Prevent inadvertent criticality.
This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of main-l tenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.
The reactor protection system is 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 bases for the trip 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 that the protective functions associated with each channel are com-plated witnin 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 sequeatial, 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, or (2) utilizing replacement sensors with certified response times.
3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip set-points and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance.
Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety.
Negative barometric pressure fluctuations are accounted for in the trip setpoints and allowable values specified for drywell pressure-high. The setpoints of other
' instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the nomal operating range to prevent inadvertent actuation of the systems involved.
Except:for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected.
For D.C. operated valves, a 3 second delay is assumed before the valve starts to move.
For A.C. operated valves, it is_ assumed that GRAND GULF-UNIT 1 B 3/4 3-1
INSTRUMENTATION BASES ISOLATION ACTUATION INSTRUMENTATION (continued) the A.C. power supply is lost and is restored by startup of the emergency diesel generators. In this event, a time ^of 10 seconds is assumed before the valve starts to move.
In addition to the pipe break, the failure of the'D.C. operated valve is assumed; thus the signal delay (sensor response) is concurrent with the 10 second diesel startup. The safety analysis considers an allowable l
inventory loss in each case which in turn determines the valve speed in conjunc-tion with the 10 second delay.
It follows that checking the valve speeds and the 10 second time for emergency power establishment will establish the. response time for the isolation functions.
However, to enhance overall system relia-bility and to monitor instrument channel response time trends, the isolation actuation instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the i
difference between each Trip Setpoint and the Allowable Value is equal to or greater than the drift allowance assumed for each trip in the safety analyses.
3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control.
This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the_ design protection.
Negative barometric pressure fluctuations are accounted for in the trip setpoints and allowable values specified for drywell pressure-high.
Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.
Operation with a trip set less conservative than its Trip Setpoint but 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 greater than the drift allowance assumed for each trip in the safety analyses.
3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences ~of the unlikely occurrence of a failure to scram during an anticipated transient.
The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, dated March 1971 and NE00-24222, dated December 1979, and Section 15.8 Appendix 15A of the FSAR.
The end-of cycle recirculation pump trip (E0C-RPT) system is a part of the Reactor Protection System and is an essential safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle.
The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity., Each EOC-RPT system trips both recirculation pumps, reducing coolant '
flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the E0C-RPT protective i
i GRAND GULF-UNIT 1 B 3/4 3-2 Amendment No. 19 l i
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5.0 DESIGN FEATURES 5.1 SITE EXCLUSION' AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.
LOW POPULATION ZONE The low population zone shall be as shown in Figure 5.k.2-1.
5.1.2 UNRESTRICTED AREA AND SITE BOUNDARY FOR GASE0US EFFLUENTS AND FOR LIQUID EFFLUENTS 5.1.3 The UNRESTRICTED AREA AND SITE BOUNDARY for gaseous effluents and for liquid effluents shall be as shown in Figure 5.1.3-1.
The gaseous effluent release points are shown in Figure 5.1.1-1.
- 5. 2 CONTAINMENT CONFIGURATION 5.2.1 The containment is a steel lined, reinforced concrete structure composed-of a vertical right cylinder and a hemispherical done.
Inside and at the bottom of the containment is a reinforced concrete drywell composed of a vertical right cylinder and a steel head which contains an approximately eighteen to nineteen foot deep water filled suppression pool connected to the drywell through a series of horizontal vents. The containment has a minimum net free air volume of 1,400,000 cubic feet.
The drywell has a minimum net free air volume of 270,000 cubic feet.
DESIGN TEMPERATURE AND PRESSURE 5.2.2 The containment and drywell are designed and shall be maintained for:
a.
Maximum internal pressure:
1.
Drywell 30 psig.-
2.
Containment 15 psig.
b.
Maximum internal temperature:
1.
Drywell 330*F.
2.
~ Suppression pool 185'F.
c.
Maximum external-to-internal differential pressure:
1.
Drywell 21 psid.
2.
Containment 3 psid.
~
5.2.3 The gecondary containment consists of the Auxiliary Building and the Enclosure Building, and has a minimum free volume of 3,640,000 cu5ic feet.
GRAND GULF-UNIT 1 5-1
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1 GRAND GULF-UNIT 1 5-2 Amendment No.19
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1 FIGURE 5.1.1-1 EXCLUSION AREA AND GASEOUS EFFLUENT RELEASE POINTS 5-2a Amendment No. 19
~
DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 800 fuel assemblies with each fuel assembly containing 62 fuel rods and two water rods clad with Zircaloy -2.
Each fuel rod shall have a design nominal active fuel length of 150 inches.
The initial core loading shall have a design nominal enrichment of 1.708 weight 1
1 percent U-235.
Reload fuel shall be similar in physical design to the initial core loading.
CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 193 control rod assemblies, each consisting of a cruciform array of stainless steel tubes containing a design nominal 143.7 inches of boron carbide, B C, powder surrounded by a cruciform 4
shaped stainless steel sheath.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
a.
In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, 7
b.
For a pressure of:
1.
1250 psig on the suction side of the recirculation pump.
2.
1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.
3.
1550 psig from the discharge shutoff valve to the jet pumps.
c.
For a temperature of 575'F.
VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,000 cubic feet at a nominal T,,, of 533'F.
l l
GRAND GULF-UNIT 1 5-5
DESIGN FEATURES 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.2-1.
5.6 FUEL STORAGE CRITICALITY 5.6.1 The spent fuel storage racks are designed and shall be maintained with:
A k,ff equivalent to less than or equal to 0.95 when flooded with a.
unborated water, including all calculational uncertainties and biases as described in Section 9.1 of the FSAR.
l b.
A nominal 6.26-inch center-to-center distance between fuel assemblies l
placed in the storage racks.
5.6.1.2 The k,ff for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.
DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 202'5 1/4".
CAPACITY 5.6.3 The spent fuel storage capacity is designed and shall be maintained with a storage capacity limited to:
No more than 2324* spent fuel assemblies in the spent fuel pool, and a.
b.
No more than 800 spent fuel assemblies in the upper containment pool.
Placement of fuel in the upper containment pool is limited to temporary storage of fuel during refueling operations.
Prior to return to reactor criticality, all spent fuel shall be removed from the upper containment pool.
5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7.1-1.
- The physical limit is 4348. The 2324 limit reflects the number of spent fuel assemblies that can be stored in the spent fuel pool without excessive reliance on RHR supplement cooling; i.e., for a time period in excess of a normal refueling duration.
GRAND GULF-UNIT 1 5-6 Amendment No.19 l
.