ML20148E739
| ML20148E739 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 01/06/1988 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20148E744 | List: |
| References | |
| NUDOCS 8801260103 | |
| Download: ML20148E739 (25) | |
Text
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VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 116 License No. DPR-32 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Virginia Electric and Power Company (the licensee) dated October 7, 1986, as supplemented i
June 8, 1987; April 1, 1987; and May 26, 1987, 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, I
j the provisions of the Act, and the rules and regulations of the Commission; 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 Commission's regulations; 0.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 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, and paragraph 3.B of Facility Operating License No. DPR-32 is hereby amended to read as follows:
8801260103 880106 PDR ADOCK 05000280 P
(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.116., are hereby incorporated in the license.
The licensee 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 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/
H.rbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 6, 1988 I
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UNITED STATES
! ' eq,' g NUCLEAR REGULATORY COMMISSION g
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E WASHINGTON, D. C. 20555
- d s,,.a [y8 VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 116 License No. OPR-37 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Virginia Electric and Power Company (the licensee) dated October 7, 1986, as supplemented June 8, 1987; April 1, 1987; and May 26, 1987, comply with the l
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 provisions of the Act, and the rules and regulations of the i
Commission; 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 Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 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, and paragraph 3.8 of Facility Operating License No. OPR-37 is hereby amended to read as follows:
(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 116, are hereby incorporated in the license.
The licensee 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 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
,/
/
/
erbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 6, 1988 i
i l
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 116 FACILITY OPERATING LICENSE NO. DPR-32 AMENDMENT NO. 116 FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NOS. 50-280 AND 50-281 Revise Appendix A as follows:
Remove Pages Insert Pages TS 2.1-2 TS 2.1-2 TS 2.1-3 TS 2.1-3 TS 2.1-4 TS 2.1-4 TS 2.1-5 TS 2.1-5 TS 2.3-8 TS 2.3-8 TS 3.1-Sa TS 3.1-Sa TS 3.12-1 TS 3.12-1 TS 3.12-8 TS 3.12-8 TS 3.12-9 TS 3.12-9 TS 3.12-10 TS 3.12-10 TS 3.12-11 TS 3.12-11 TS 3.12-13 TS 3.12-13 TS 3.12-16 TS 3.12-16 TS Table 3.12-1 TS Figure 3.12-1A TS Figure 3.12-1A TS Figure 3.12-1B TS Figure 3.12-1B TS Figure 3.12-2 TS Figure 3.12-2 l
TS Figure 3.12-3 TS Figure 3.12-3 TS Figure 3.12-5 TS Figure 3.12-5
)
TS Figure 3.12-6 TS Figure 3.12-6 i
e 9
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TS 2.1-2 4
The reactor thermal power level shall not exceed 1187.' of rated power.
B.
The safety limit is exceeded if the combination of Reactor Coolant System average temperature and thermal power level is at any time above the appropriate pressure line in TS Figures 2.1-1, 2.1-2 or 2.1-3; or the core thermal power exceeds 118% of the rated power.
D Basis To maintain the integrity of the fuel cladding and prevent fission pro-duct release, it is necessary to prevent overheating of the cladding under all operating conditions.
This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer evefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the reactor coolant saturation temperature.
The upper boundary of the nucleate boiling regimt is termed Departure From Nucleate Boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would. result in high clad temperatures and the possibility of clad failure.
DNB is
- not, however, an observable parameter during reactor operation.
Therefore, DNB has been correlated to
- thermal power, reactor coolant temperature and reactor coolant pressure which are observable parameters. This correlation has been developed to predict the DNB flux and the location of DNB for axially Amendment Nos.116 and 116
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TS 2.1-3 uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the DNB heat flux at a partdeular core location to the local heat flux, is indicative of the margin to DNB. The DNB basis is as follows:
there must be at least a 95% probability with 95% confidence that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used. The correlation DNBR limit is based on the entire applicable experimental data set to meet this statistical criterion.(U The curves of TS Figure 2.1-1 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (three loop operation) represent limits equal to, or more conservative than, the loci of points of thermal power, coolant system average temperature, and coolant system pressure for which the calculated DNBR is not less than the design DNBR limit or the average enthalpy at the exit of the vessel is equal to the saturation value. The area where clad integrity is assured is below these lines.
The temperature limits are considerably more conservative than would be required if they vert based upon the design DNBR limit alone but are such that the plant conditions i
required to violate the limits are precluded by the self-actuated
)
I safety valves on the steam generators. The three loop operation _
l safety limit curve allows for heat flux peaking effects due to fuel densification and applies to 100% of. design flow. The effects of rod bowing are also considered in the DNBR analyses.
j The curves of TS Figure 2.1-2 and 2.1-3 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (two loop operation), represent limits equal to, or more
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Amendment Nos. 116 and 116
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. v.
TS 2.1-4 conservative, than the loci of points of thermal power, coolant system average temperature, and coolant system pressure for which either the calculated DNBR is equal te the design DNBR limit or the average enthalpy at the exit of the core is equal to the saturation value.
At low pressures or high temperatures the average enthalpy at the exit of the core reaches saturation before the calculated DNBR reaches the design DNBR limit and, thus, this arbitrary limit is conservative with respect to maintaining clad integrity.
The plant conditions required to violate these limits are precluded by the protection system and the self-actuated safety valves on the steam generator.
Upper limits of 70% power for loop stop valves open and 75% with loop stop valves closed are shown to completely bound the area where clad integrity is assured. These latter limits are arbitrary but cannot be reached due to the Permissive 8 protection system setpoint which will trip the reactor on high nuclear flux when only two reactor coolant pumps are in service.
Operation with natural circulation or with only one loop in service is not allowed since the plant is not designed for continuous operation with less than two loops in service.
TS Figures 2.1-1 through 2.1-3 are based on a F 1.55, a
~
H 1.55 cosine axial flux shape and a DNB analysis procedure including I) the fuel dent.ification power spiking as part of the generic bowing. (5) (6) TS Figure 2.1-1 is also margin to accommodate rod valid for the following limit of the enthalpy rise hot channel hn factor:
1.55 (1 + 0.3 (1-P)) where P is the fraction of rated power. TS Figures 2.1-2 and 2.1-3 include a 0.2 rather than 0.3 part power cultiplier for the enthalpy rise hot channel factor.
These hot channel factors are higher than those calculated at full power over the range between that of all control rod assemblies fully withdrawn to Amendment Mos. 116 and 116
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TS 2.1-5 maximum allowable control red assembly insertion.
The control red assembly insertion limits are covered by Specification 3.12.
Adverse power distribution factors could occur at lower power levels because additional control rod assemblies are in the core; however, the control rod assembly insertion limits dictated by TS Figures 3.12-1A (Unit 1) and 3.12-1B (Unit 2) ensure that the DNBR is always greater at partial power than at full power.
The Reactor Control and Protection System is designed to prevent any anticipated combination of transient conditions for Reactor Coolant System temperature, pressure and thermal power level that would result in a DNBR less than the design DNBR limit ( } based on steady state nominal operating power levels less th.an or equal to 100%, steady state nominal operating Reactor Coolant System average temperatures less than or equal to 574.4'T and a steady state nominal operating pressure of 2235 psig. Allowances are made in initial conditions assumed for transient analyses for steady state errors of +2% in power,
+4*F in Reactor Coolant System average temperature and !30 psi in pressure. The combined steady state errors result in the DNB ratio at the start of a transient being 10 per cent less than the value at nominal full power operating conditions.
The steady state nominal operating parameters and allowances for steady state errors given above are also applicable for two loop operation except that the steady state nominal operating power level is less than or equal to 60%.
The fuel overpower design limit is 118% of rated power.
The over-power limit criterion is that core power be prevented from reaching a value at which fuel pellet melting would occur. The value of i
118% power allows substantial margin t
I 1
Amendnent tios.116 and 116
TS 2.3-8 will prevent the minimum value of the DNBR from going below the applicable design limit during nomal operational transients and anticipated transients when only two loops are in operation and the overtemperature AT trip setpoint is adjusted to the value specified l
1 for three-loop operation. During two-loop operation with the loop stop valves in the inactive loop open, and the overtemperature AT trip setpoint is adjusted to the value specified for this conditien, a reactor trip at 60% power will prevent the minimum value of DNBR
{
from going below the applicable design limit during normal operational transients and anticipated transients when only two loops are in operation. During two-loop operation with the inactive loop stop valves closed and the overtemperature aT trip setpoint is adjusted to the value specified for this condition, a reactor trip at 65% power will prevent the minimum DNBR from going below the applicable design limit during normal operational transients and anticipated transients.
Although not necessary for core protection, other reactor trips l
provide additional protection.
The steam /feedwater flow mismatch which is coincident with a low steam generator water level is designed for t.nd provides protection from a sudden loss of the reactor's heat sink. Upon the actuation of the safety injection circuitry, the reactor is tripped to decrease the severity of the accident condition. Upon turbine trip, at greater than 10% power, the reactor is tripped to reduce the severity of the ensuing transient.
References (1) FSAR Section 14.2.1 (2) FSAR Section 14.2 (3) FSAR Section 14.5 (4) FSAR Section 7.2 (5) FSAR Section 3.2.2 (6) FSAR Section 14.2.9 (7) FSAR Section 7.2 Amendment 'los.116 and 116
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TS 3.1-Sa b.
With one Reactor Vessel Head vent path inoperable; startup and/or power operation may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of both isolation valves in the inoperable vent path, c.
With two Reactor Vessel Head vent paths inoperable; maintain the inoperable vent path closed with power removed from the valve actuator of all isolation valves in the inoperable vent paths, and restore at least one of the vent paths to operable status within 30 l
days or be in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
1 e
Basis l
Specification 3.1.A-1 requires that a sufficient number of reactor coolant pumps be operating to provide coastdown core cooling flow in the event of a loss of reactor coolant flow accident.
This provided flow will maintain the DNBR above the applicable,_ design limit.
Heat transfer analyses also show that reactor heat equivalent to approximately 10% of rated power can be removed with natural circulation; however, the plant is not designed for critical operation with natural circulation or one loop operation and will not be operated under these conditions.
When the boron concentration of the Reactor Coolant System is to be reduced, the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uniform concentration if at least one reactor coolant pump or
- one residual heat removal pump is running while the change is taking i
place.
The residual heat removal pump will circulate the equivalent of the reactor coolant system volume in approximately one half hour.
Amendment Nos.116 and 116
,i,....,...,...,,
s TS 3.12-1 T
3.12 CONTROL ROD ASTEMBLIES AND POWER DISTRIBUTION LIMITS l
t Applicability Applies to the operation of the control rod assemblies and power distri-bution limits.
gjactive
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Nens,urecoresuberiticalityafterareactortrip,alimitonpotential i
6 5reactir.'.ty insertions from hypothetical control rod assembly ejection, and an acceptable core power distribution during power operation.
Specification h
A.
ControlBanyInsertionLimits 1.
Whenever the reactor is critical, except for physics tests and control rod assembly exercises, the shutdown control rods
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shall be fully withdrawn.
2.
Whenever the reactor is critical, except for physics tests and control rod assembly exercises, the full. length control rod banks shall be inserted no further than the appropriate lic.it 1
determined by core burnup shown on TS Figures 3.12-1A or 3.12-1B for three-loop operation asi TS Tigures 3.12-4A or,
3.12-4B for two-loop operation.
l s
3.
The limits shown on TS Figures 3.12-1A through 3.12-6 may be 0
\\ 's revised on the basis of physics calculations and physics data
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1 1
obtained during unit s tar':up and subsequint operation, f.n
' i,I \\
accordance with the follcwing':
l The sequence of withdrawal of the controlling banks, when a.
1) going from zero to 100% power, is A, B, C, D.
j b.
An overlap of control banks, ccasis he with physics cal-hnendment 'los.116 and 116
TS 3.12-8 AT and Overtemperature AT trip settings shall be reduced by l the equivalent of 2% power for every 1% quadrant to average powwr tilt.
C.
A control rod assembly shall be considered inoperable if the assembly cannot be moved by the drive mechanism or the assembly remains misaligned from its group step demand position by more than 212 steps. Additionally, a full-length control rod shall be considered inoperable if its rod drop time is greater than 2.4 seconds to dashpot entry.
]
2.
No more than one inoperable control rod assembly shall be permitted when the reactor is critical.
3.
If more than one control rod assembly in a given bank is out of service because of a single failure external to the individual rod drive mechanism, (i.e. programming circuitry),
l the provisio'ns of Specifications 3.12.C.1 and 3.12.C.2 shall not apply and the reactor may remain critical for a period not to exceed two hours provided immediate attention is directed toward making the necessary repairs.
In the event the affected assemblies cannot be returned to service within this specified period,the reactor will be brought to hot shutdown l
conditions.
4.
The provisions of Specifications 3.12.C.1 and 3.12.C.2 shall not apply during physics tests in which the assemblies are intentionally misaligned, l
5.
Power operation may continue with one rod inoperable provided that within one hour either:
a.
the rod is no longer inoperable as defined in Specification 3.12.C.1, or Amendment flos.116 and 116
l TS 3.12-9 b.
the rod is declared inoperable and the shutdown nargin requirement of Specification 3.12.A.3.c is satisfied.
Operation at power may then continue provided that:
1) either:
(a) power shall be reduced to less than 75% of rated power within one (1) hour, and the High Neutron Flux trip setpoint shall be reduced to less than or equal to 85% of rated power within the next four (4) hours, or (b) the remainder of the rods in the group with the inoperable rod are aligned to within 12 steps of the inoperable rod within one (1) hour while maint aining the rod sequence and insertion limits of Figure 3.12-1; the thermal power level shall be restricted pursuant to Specification 3.12.A during subsequent operation.
2) the shutdown margin requirement of Specification 3.12.A.3.c is determined to be met within one hour 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 /> thereafter.
3) the hot channel factors are shown to be within the design limits of Specification 3.12.B.1 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- Further, it shall be demonstrated that the value of Fxy(Z) used in the Constant Axial Offset Control analysis is still valid.
4) a reevaluation of each accident analysis of Table 3.12-1 is performed within 5 days.
This reevaluation shall confirm that the previous analyzed results of these accidents remain valid for the duration of operation under these
?
conditions.
Amendment Hos.116 and 116
e TS 3.12-10 6.
If power has been reduced in accordance with Specifica-tion 3.12.C.S.b power may be increased above 75% power provided that:
a) an analysis has been performed to determine the hot channel factors and the resulting allowable power level based on the limits of Specification 3.12.B.1, and b) an evaluation of the effects of operating at the increased power level on the accident analyses of Table 3.12-1 has been completed.
D.
Core Quadrant Power Balance:
1.
If the reactor is operating above 75% of rated power with one excore nuclear channel out of service, the core quadrant power balance shall be determined:
a.
Once per day, and b.
After a change in power level greater than 10% or more than 30 inches of control rod motion.
2.
The core quadrant power balance shall be determined by one of the following methods:
a.
Movable detectors (at least two per quadrant) 1 b.
Core exit thermocouples (at least four per quadrant)
E.
Rod position Indicator Channels 1.
The rod position indication system shall be operable and capable of determining the control rod positions within 212 steps.
2.
If a rod position indicator channel is out of service, then:
a.
For operation above 50% of rated power, the position of the RCC shall be checked indirectly using core instrumentation (excore detectors and/or incore thermo-couples and/or movable incore deteceors) at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and immediately af ter any motion of the non-indicating rod exceeding 24 steps, or i
Amendment Nos.116and 116
TS 3.12-11 I
b.
Reduce Power to less than 50% of rated power within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During operations below 50% of rated power, no special monitoring is required.
3.
If more than one rod position (RPI) indicator channel per group or two RFI channels per bank are inope'rable, then the requirements of Specification 3.0.1 will be followed.
Ba s i.s.
The reactivity control concept assumed for operation is that reactivity,
changes accompanying changes in reactor power are compensated by control rod j
assembly motion.
Reactivity changes associated with xenon, samarium, fuel depletion, and large changes in reactor coolant te=perature (operating temperature to cold shutdown) are compensated for by changes in the soluble boron concentration.
During power operation, the shutdown groups are fully withdrawn and control of power is by the control groups.
A reactor trip occurring during power operation will place the reactor into the hot shutdown condition.
The control rod assembly insertion limits provide for acheiving hot shutdown by reactor trip at any time, assuming the highest worth control rod assembly remains fully withdrawn, with sufficient margins to meet the assumptions used in the accident analysis.
In addition, they provide a limit l
I Amendment Nos.116 and il6
TS 3.12-13 in service, the effects of malpositioned control rod assemblies are observable from nuclear and process information displayed in the Main Control Room and by core thermocouples and in-core movable detectors. Below 50% power, no special monitoring is required for malpositioned control rod assemblies with inoperable rod position indicators because, even with an unnoticed complete assembly misalignment (full length control rod assembly 12 feet out of alignment with its bank), operation at 50% steady state power does not result in exceeding core limits.
The specified control rod assembly drop time is consistent with safety analyses that have been performed.
1 An inoperable control rod assembly imposes additional demands on the operators. The permissible number of inoperable control rod assemblies is limited to one in order to limit the magnitude of the operating burden, but such a failure would not prevent dropping of the operable control rod assemblies upon reactor trip.
Two criteria have been chosen as a design basis for fuel performance related to fission gas release, pellet temperature, and
]
cladding mechanical properties.
- First, the peak value of fuel centerline temperature must not exceed 4700*F. Second, the minimum DNBR in the core must not be less than the applicable design limit in normal operation or in short term trans$ents.
Amendment Nos,116 and 116 i
1
TS 3.12-16 be coepensated for by tighter axial control.
Four percent is the appropriate allowance for measurement uncertainty for F obtained from a full core map (>
38 thimbles, including a minimum of 2 detectors per core quadrant, monitored) taken with the movable incore detector flux mapping system.
Measurement of the hot channel factors are required as part of startup physics tests, during each effective full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors.
The incore map taken following core loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns.
The periodic incore mapping provides additional assurance '
that the nuclear design bases remain inviolate and identify operational anomalies which would, otherwise, affect these bases.
For nor=al operation, it has been determined that, provided certain conditions are observed, the enthalpy rise hot channel f actor F limit will be met.
These conditions are as follows:
1.
Control rods in a single bank move together with no individual rod insertion differing by more than 15 inches from the bank demand position. An indicated misalignment 11mit of 13 steps precludes a rod misalignment no greater than 15 inches with consideration of maximum instrumentation error.
2.
Control rod banks are sequenced with overlapping banks as shown in TS Fipres 3.12-1A, 3.12-1B.
3.
The full length control bank insertion limits are not violated.
4.
Axial power distribution control procedures, which are given in terms of flux difference control and control bank insertion limits are observed.
Flux difference refers to the difference l
Amendment flos.116 and 116
I TS TABLE 3.12-1 i
TABLE 3.12-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE ROD 1
Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Large and Small Break Loss of Coolant Accidents Single Reactor Coolant Pump Locked Rotor Major Secondary Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) l Amendment Nos. 116 and 116
TS FIGURI 3.12-1A FULLY WITHDRAWii 225
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_Z 200
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_ (0.0, 23) 0.0 0.2 0.4 0.6 0.8 1.0 FULLY IkSERTED FRACTION OF RATED POWER FIGURE 3.12-1 A C0'! TROL BAfiK INSERTION LIll!TS FOR 3-LOOP NORfiAL OPERATI0ff-VillT 1 Amendment Nos,116 and J16
TS FIGURE 3.12-1B
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FULLY WITHDRAWN 225
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- (0.0,23) 0 0.0 0.2 0.4 0.6 0.0 1.0 FULLY INSERTED FRACTION OF RATED POWER FIGURE 3.12-1B CONTROL BANK lilSERTION LIMITS FOR NORMAL 3 LOOP OPERATION - UNIT 2 Amendment Nos.116 and116
IS FfGURE 3.12-2 i
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4 TS FIGURE 3.12-3 DELETE Amendment Hos.
and 2
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i DELETE Amendment flos,116 andl16
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TS FIGURE 3.12-6 Se 9
DELETE Amendment Nos. 116 and 116