ML20151D537
| ML20151D537 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/08/1988 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20151D540 | List: |
| References | |
| NUDOCS 8804140302 | |
| Download: ML20151D537 (19) | |
Text
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CAROLINA POWER & LIGHT COMPANY. et al.
DOCKET N0. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 149 License No. OPR-62 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Carolina Power & Light Company (the licensee), dated September 4, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issumce of this amendment is in accordance with 10 CFR Part 51 of the Comission'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; and paragraph 2.C.(2) of Facility Operating License No. DPR-62 is hereby Amended to read as follows:
9804140302 800400 PDR ADOCK 0500 4
P (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.149, are hereby incorporated in the license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
\\5\\
Elinor G. Adensam, Director Project Directorate II-1 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance,
April 9,19R8 s
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t ATTACHMENT TO LICENSE AMENDMENT NO. 149
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FACILITY 0PERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by marginal lines.
Remove Pages Insert Pages j
IV IV 1-2 1-2 1-5 1-5 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 l
3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 l
4 3/4 2-10 3/4 2-10 j
3/4 2-12 3/4 2-12 4
)
3/4 2-14 3/4 2-14 3/4 3-42 3/4 3-42 8 3/4 2-2 B 3/4 2-2 1
)
B 3/4 2-3 8 3/4 2-3 1
I 1
4 41 1
a
,n
1 l
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i
i l
SECTION PACE 3/4.0 APPLICABILITY..............................................
3/4 0-1 l
3/4.1 NEACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN KARCIN..........................................
3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES.....................................
3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operab(11ty..................................
3/4 1-3 Control Rod Maximum Scram Insertion Times................
3/4 1-5 Control Rod Average Scram Insertion Times................
3/4 1-6 Four Control Rod Group Insertion Times...................
3/4 1-7 Control Rod Scram Accumulators...........................
3/4 1-8 Control Rod Drive Coupling...............................
3/4 1-9 Control Rod Position Indication..........................
3/4 1-11 Control Rod Drive Housing Support........................
3/4 1 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimiser......................................
3/4 1-14 Rod Sequence Control System..............................
3/4 1-15 Rod Block Monitor........................................
3/4 1-17 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................
3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE...............
3/4 2-1 3/4.2.2 APRM SETP0!NTS...........................................
3/4 2-7 3/4.2.3 MINIMUM CRITICAL POWER RATI0.............................
3/4 2-8 3/4.2.4 LINEAR HEAT CENERATION RATE..............................
3/4 2-14 l
l BRUNSWICK - UNIT 1 IV Amendment No. 33.149
DEFINITIONS CHANNEL FUNCTIONAL TEST (Continued) b.
Bistable channels - the injection of a simulated signal it.to the channel sensor to verify OPERABILITY including alarm and/or trip functions.
j 1
CORE ALTERATION j
CORE ALTERATION shall be the addition, removal, relocation, or movement of fuel, sources, incore instruments, or reactivity controls in the reactor core with the vessel head removed and fuel in the vessel.
Suspension of CORE i
ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative location.
CRITICAL POWER RATIO The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in an assembly which is calculated, by application of an NRC approved CPR correlation, to i
cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.
DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be concentration of I-131, y Ci/ gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, 1-133, I-134, and 1-135 actually present. The following is defined equivalent to 1 uCi of I-131 as determined from Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites":
I-132, 28 uCil I-133, 3.7 uCil I-134, 59 uCil I-135, 12 uCi.
E -AVERACE DISINTECRATION ENERCY E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta an' gamma energies per disintegration (in MeV) for isotopes with half lives greater than 15 minutes making up at least 95% of the total non-iodine activity in the coolant.
EMERCENCY CORE COOLING SYSTEM (ECCS) RESP 0NSE TIME The EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval f rom when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).
Times shall include diesel generator starting and sequence loading delays where applicable.
RRUNSWICK - UNIT 2 1-2 Amendment No. P3,149
DEFINITIONS OFFSITE DOSE CALCULATION MANUAL (ODCM)
The OFFSITE DOSE CALCULATIONAL KANUAL (00CM) is a manual which contains the current methodology and parameters to be used to calculate',offsite doses resulting from the release of radioactive gaseous and liquid effluents; the methodology to calculate gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints; and, the requirements of the environmental radiological monitoring program.
OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).
Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal ard emergency electric power sources, cooling or saal water, tubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).
OPERATIONAL CONDITION i
An OPERATIONAL CONDITION shall be any one inclusive combination of mode switch position and average reactor coolant temperature as indicated in Table 1.2.
PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related' instrumentation and are 1) described in Section 14 of the Updated FSAR, 2) authorized under the l
provisions of 10 CFR 50.59, or 3) otherwise. approved by the Commission.
PRESSURE BOUNDARY LEAKACE PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable f ault in a 2
reactor coolant system component body, pipe wall, or vessel wall.
l PRIMARY CONTAINMENT INTECRITY PRIMARY CONTAINMENT INTEGRITY shall exist whent a.
All penetrations required to be closed during accident conditions are eithert 1.
Capable of being closed by an OPERABLE containment automatic l
isolation valve system, or l
2.
Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.1, or 3
4 BRUNSWICK - UNIT 2 1-$
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- FUEL TYPE P90R823rdi (P8X8R) f, MIHLM AVERAGE P1ANAR 1.ItEAR lEAT GENERATION RATE (MAPtJIGR) z*
VERSUS AVERAGE Pl>HAR EXI%tRE 23 oy Figure 3.2.1 2 authorizes operation only up to an average fuel bundle burnup of 33,000 lifD/ tit l
- Amendment 149 1
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- Amendment 149 authorizes operation only up to an average fuel bundle buraup of 33,000 islD/ tit l 0
POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS L_fMITING CONDITION FOR OPERATION 3.2.2 The flow-blased APRM scram trip setpoint (S) and rod block trip set point (Sgg) shall be established according to the following relationships i
S $ (0.66W + 54%) T SRB $ (0.66W + 42%) T wheret S and SRB are in percent of RATED THERMAL POWER.
W = Loop recirculation flow in percent of rated flow, T = Lowest value of the ratio of design TPF divided by the MTPF obtained for any class of fuel in the core (T 5 1.0), and Design TPF for P8 X 8R. fuel = 2.39 l
l BP8 x 8R fuel = 2.39 CE8 fuel = 2.48 l
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTIONt With 5 or SRB exceeding the allowable value, initiate corrective action within 15 minutes and continue corrective action so that S and S ' are within the gg required limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.2 The MTPF for each class of fuel shall be determined, the value of T calculated, and the flow biased APRM trip setpoint adjusted, as requiredt a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is i
c.
c'perating with a LIMITINC CONTROL ROD PATTERN for MTPF.
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BRUNSWICK - UWIT 2 3/4 2-7 Amendment No. 107, 123 149
l POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core flow, shall be equal to or greater than the MCPR limit times the Kg shown'in Figure 3.2.3-1 with the following MCPR limit adjustmentst Beginning-of-cycle (BOC) to end-of-cycle (EOC) minus 2000 MWD /t with a.
ODYN OPTION A analyses in ef fect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed belovt 1.
MCPR for P8 x 8R fuel = 1.34 2.
MCPR for BP8 x 81 fuel = 1.34 3.
MCPR for CE8 fuel = 1.34 b.
EOC minus 2000 MVD/t to EOC with ODYN OPTION A analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below:
1.
MCPR for P8 x 8R fuel = 1.35 2.
MCPR for BP8 x 8R fuel = 1.35 3.
MCPR for CE8 fuel = 1.35 c.
BCC to EOC minus 2000 MWD /t with ODYN OPTION 8 analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below 1.
MCPR for P8 x 8R fuel = 1.27 2.
MCPR for BP8 x 8R fuel = 1.27 3.
MCPR for CE8 fuel = 1.27 d.
EOC minus 2000 MWD /t to EOC with ODYN OPTION 5 analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below 1.
MCPR for P8 x SR fuel = 1.31 2.
MCPR for BP8 x 8R fuel = 1.31 3.
MCPR for CE8 fuel = 1.31 APPLICABILITYt OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER 1
BRUNSWICK - UNIT 2 3/4 2-8 Amendment No. 137. 173 149
POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B)
LIMITING CONDITION FOR OPERATION 3.2.3.2 For the OPTION 8 MCPR Limits listed in specification 3.2.3.1 to be shall be less than or l
used, the cycle average 20% (notch 36) scram time (t g),wheretfy*)andt equal to the Option 8 scram time limit (t are determined g
as follows:
n I
t,y,
, where
=
t i=1 i = Surveillance test number, n = Number of surveillance tests performed to date in the cycle (including BOC),
Wg = Number of rods tested in the ich surveillance test, and t = Average scram time to notch 36 for surveillance test i t
(' ' " '
g = u + 1.65 (,
y) t I
i=1 i = Surveillance test number n = Number of surveillance tests performed to date in the cycle (including BOC),
th Ng = Number of rods tested in the i surveillance test i
Ng = Number of rods tested at BOC, u = 0.813 seconds l
(mean value for statistical scram time distribution from de-energitation of scram pilot valve solenoid to pickup on notch 36),
o = 0.018 seconds l
(standard deviation of the above statistical distribution).
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or
]
equal to 25% RATED THERMAL POWER.
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BauuSw!CK - UNIT 2 3/4 2-10 Amendment No. 73. m,
149
Y si TABLE 3.2.3.2-1 v.
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TRANSIENT OPERATING LIMIT MCPR VALUES 1
e h 1RANSIENT FIEL TYPE P8m8A BP8m8R CES u
NONPRESSURIZATION TRANSIENTS 1.27 1.27 1.27
'gDC + EDC I
PRESSURIZATION TRANSIENTS MCPR N
M MD "A
E A
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5 noC. soc - 200o 1.x 1.27 1.x 1.27 i.x t.27 1.35 1.31 1.35 1.31 1.35 1.31 gDC - 2000 + EDc 8
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EE sawswIcx - WIT 2 3/4 2 Amendment No. 101, 72), 149
___--_.-_-_---__-_-.-12.,
POWER DISTRIBUTION LIMITS I
3/4.2.4 LINEAR HEAT CENEilATION RATE LIMITINC CONDITION FOR OPERATION
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l 3.2.4 The LINEAR HEAT CENERATION RATE (LHCR) shall not exceed 13.4 kw/ft for P8 8R and BP8xtR fuel assemblias and 14.4 kw/ft for CE8 fuel as'semblies.
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% cf RATED THERMAL POWER.
ACTIOWs With the LHCR of any fuel rod exceeding the above limit, initiate corrective action within 15 minutes and continue corrective action so that the LHCR is within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.4 LHCRs shall be determined to be equal to or less than the limit:
s.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is c.
operating on a LIMITINC CONTROL ROD PATTERN for LHCR.
j BRUNSWICK - UNIT 2 3/4 2 14 Amendmens No. J@l.
7 D.149
5
-2 TABLE 3.3.4-2 Q
CONTROL ROD WITHDRAWAL BIDCK INSTRtmENTATION SETPOINTS TRIP F120CTION AND INSTRUMENT NtMBER TRIP SETPOINT ALIDWABLE VALUE g
1.
APRM (CSI-APRM-CH. A B.C.D E.F)
Upscale (Flow Biased)
$ (0.66W + 421) T *)
$ (0.66W + 42%) T *)
I I
a.
b.
Inoperative NA NA c.
Downscale
> 3/125 of full scale
> 3/125 of full scale d.
Upecale (Fised) 312%ofRATEDTHERMALPOWER 312%ofRATEDTHERMALPOWER 2.
BOD BLACK MONITOR (C51-RSM-CN.A,B) a.
Upacale
$ (0.66W + 39Z) T *)
$ (0.66W + 39%) T *)
I I
b.
Inoperative NA NA c.
Downscale 3 3/125 of full scale 3 3/125 of full scale 3.
SOURCE RANCE MONITORS (C51-SRM-K600A,B,C,D) a.
Detector not full in MA NA 5
5 R
b.
Upscale
$1x 10 cps
<1x 10 cps c.
Inoperative NA NA d.
Downscale 3 3 cps 3 3 cps 4.
INTERMEDIATE RANCE MONITORS (C51-IRM-K601A,B,C,D,E,F,C,H) a.
Detector not full in M4 tiA b.
Upscale
$ 108/125 of full scale
$ 108/125 of full scale c.
Inoperative NA NA d.
Downscale 3 3/125 of full scale 3 3/125 of full scale 5.
SCRAM DISCHARCE VOLUME (C12-LSH-N013E) a.
Water Level High
< 73 gallons
$ 73 gallons g
8 (a) r as defined in specification 3.2.2.
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Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLART ACCIDENT ANALYSIS FOR BRUNSWICK - UNIT 2 Plant Parameterst Core Thermal Power 2531 Hvt which corresponds to 105% of rated steam flow 6
Vessel Steam Output 10.96 x 10 Lbs/h which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure 1055 psia Recirculation Line Break Area for Large Breaks a.
Discharge 2.4 ft2 (DBA) 1.9 ft2 (80% DBA) 2 b.
Suction 4.2 ft Number of Dellied Bundles 520 Fuel Parameters t' PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL SUNDLE CENELATION LATE PEAXINC POWER **
FUEL TYPES CEONETRY (kv/ft)
FACTOR LATI0' Reload Core BP/P8x8R 13.4 1.4 1.20 CE8x8EB 14.4 1.4 1.20 j
t A more detailed list of input to each model and its source is presented in Section II of Reference 1.
'This power level meets the Appendix K requirement of 102%.
To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation is performed with an MCPR of j
1.18 (i.e., 1.2 divided by 1.02) for a bundle with an initial MCPR of 1.20.
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BRUNSWICK - UWIT 2-5 3/4 2-2 Amendment No. 73.149
POWER DISTRIBUTION LIMITS BASES I
3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a TOTAL PEAXING FACTOR of 2.39 for P8x8R and 8P8x8R fuel and 2.48 for CE8 fuel. The scram setting and rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than 1.0 in the degraded situation. The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the combtnation of THERMAL POWER and peak flux indicates a TOTAL PEAXING FACTOR greater than 2.39 for P8x8R and BP8x8R fuel and 2.48 for CE8 fuel. This adjustment may be l
accomplished by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced APRM high flux scram curve by the reciprocal of the APRM gain change. The method used to determine the design TPF shall be consistent with the method used to determine the HTPF.
3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel claddingintegritySafegLimitMCPRof1.07,andananalysisofabnormal operational transients.
For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient, assuming an
" instrument trip setting as given in Specification 2.2.1.
To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result-in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of j
flow, increase in pressuri and power, positive reactivity insertion, and coolant temperature decrease.
Unless otherwise stated in cycle specific reload analyses, the limiting I
transient which determines the required steady state MCPR limit is the turbine trip with failure of the turbine bypass.
This transient yields the largest a MCPR.
Prior to the analysis of abnormal operational transients an initial fuel bundle MCPR was determined. This parameter is based on the bundle flow describedinSection4.4ofNEDO-20360gateflowdistributionmodelas calculated by a CE multichannel steady and on core parameters shown in I
Reference,3, response to Items 2 and 9.
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i BRUNSWICK - UNIT 2 3 3/4 2-3 Amendment No. 107, ??3.
I 149 J
- _ - -