ML19322B363
| ML19322B363 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 12/22/1975 |
| From: | Purple R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19322B344 | List: |
| References | |
| NUDOCS 7912020208 | |
| Download: ML19322B363 (32) | |
Text
-
e
. UNITED ST ATEs NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 205$5 s
h DUKEPOWERCOMPANYk DOCKET NO. 50-287 OCONEE NUCLEAR STATION, NIT 3 i
AMENDMENT TO FACILITY OPEl% TING LICENSE
~
Amendment No.I3 i
License No. DPR-55 The Nuclear Regulatory Commission (the Commission) has found that:
1.
The application for amendment by Duke Power Company (the A.
licensec) dated January 15, 1975, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissis s rules and regulations set forth in 10 CFR Chapter I; -
The facility will operate in conformity with the application, B.
the provisions of the Act, and the rules and regulations of the Commission; There is reasonabic assurance (i) that the activities authori:cd C.
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; j
and The issuance of this amendment will not be inimical to the common D.
defense and security or to the health and safety of the public.
Accordingly, the license is amended by a change to the Technical 2.
Specifications as indicated in the attachment to this license amendment and Paragraph 3.B of Facility License No. DPR-55 is hereby amended to read as follows:
OM%
e a s p,r 1
s e.e-4 7912020
5+
l l
2-
"B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications, as revised by issued changes thereto through Change 'No.13 "
3.
'shis license amendment is effective January 1,1976.
FOR THE NUCLEAR REGULATORY COM4ISSION Original signed by R. A. Purple
,i Robert A. Purple, Chief Operating Reactors Branch #1 Division of Reactor Licensing
Attachment:
l Change No.13 to the Technical Specifications
\\
Date of Issuance: DEC 2 21975 f
?
?
y l
[.
ye.
"x svaname >
.J Yf A --
\\
2*
~
' 2 c
Form ABC)l0 (Rev. 9 53) ABO 4 0240 W m sa sovemamewv remvine ormsom sete.ase.ves -
l
]<.
c,
.s.-
o A'ITACIDIENT TO LICENSE AMENDMENTS AMENDMENT NO.1 G T0 ' FACILITY LICENSE NO. DPR-38 CilANGE NO. p '; 'IO.TliCilNICAL SPECIFICATIONS; AMENDMENT NO.1 6 TO' FACILITY LICENSE NO. DPR-47 CilANGE NO.9 i TO 'IECilNICAL SPECIFICATIONS; AMENDMENT NO. 1:sTO FACILITY LICENSE NO. DPR-55 CilANGE NO.13 TO TECil JICAL SPECIFICATIONS DOCKET NOS. 50-269, 50-270, AND 50 287 Revise Appendix A as follows:
Reinovo Pages Insert New Pages i
it ii ii l
iii iii
- V iv y
vi vi 1-5 1-5 (blank) 3.1-19 3.1-19 3.1-19a 3.1-20 3.1-20 4.2-1 4.2-1 4.2-2 4.2-2 4.2-3 4.2-3 4.4-1 4.4-1 4.4-2 4.4-2 4.4-3 4.4-3 4.4-4
~4.4-4 4.4-7 4.4-7 4.4-8 4.4-8 4.4-9 4.4-9 4.4-10 4.4-10 4.13-1 4.13-1 6.1-2 6.1-2 6.1-4 6.1-4 6.2-1
)
6.2-1 6.6-1 6.6-1 thru 6.6-12 thru 6.6-9 pueo
7.
(
Section Panc
' 1. 5.,,4
,1nntrument channel Calibration 1-3 1.5.5 Heat Balance Check 1-4 1.5.6 Hea-Balance Calibration 1-4 1.6 QUADRANT POWER TILT 1-4 1.7 C0lTIAllCIEhT Ih7EGRITI 1-4 3G 21 2
_ SAFETY LIMITS AND LDfITING SAFETY SYSTE!! SETTINGS 2.1-1 13 2.1 SAFETY LIMITS, REACTOR CORE
~
2.1-1 2.2 SAFETY LD1IT, REACTOR COOLANT SYSTEM PRESSURE 2.2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE 2.3-1 INSTRUMENTATION 3
LIMITING CONDITIONS FOR OPERATION 3.1-1 3.1 REACTOR COOLANT SYSTEM 3.1-1 3.1.1 Operational Components 3.1-1 3.1.2 Pressurization. Heatup, and Cooldown Linitations 3.1-3 3.1.3 Minimum conditions for Criticality 3.1-8 3.1.4 Reactor Coolant System Activity 3.1-10 3.1.5 Chemistry 3.1-12 3.1.6 Leakane 3.1-14 3.1.7 Moderator Temperature Coefficient of Reactivity 3.1-17 3.1.8 Sinnie Loop Restrictions 3.1-19 3.1.9 Low Power Physics Testing Restrictions 3.1-20 i
3.1.10 Control Rod Operation i
3.1-21 3.2 HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS
. 3.2-1 3.3 DIERGENCY CORE COOLING, REACTOR BUILDING COOLING, REACTOR 3.3-1 BUILDING SPRAY, AND PENElaATION ROOM VENTILATION SYSTEMS 11 DEC I2 DN.
4 L
Page Section 4.5-6 4.5.2 Reactor Building Coolinn Systems 4.5-10 4.5.3 Penetration Room Ventilation System 4.5 12 4.5.4 Low Pressure Iniection System Lenkane 4.6-1 DIERGENCY POWER SYSTai PERIODIC TESTING 4.6 4.7-1 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4.7-1 4.7.1 Control Rod Drive System Functional Tests 4.7-2 4.7.2 Control Rod Pronram Verification.
4.8-1 4.8 MAIN STEAM STOP' VALVES 4.9-1 DiERGENCY FEEDWATER PUMP PERIODIC TESTING 4.9 4.10-1 4.10 REACTIVITY ANOMALIES 4'.11-1 4.11 ENVIRONMENTAL SURVEILLANCE 4.12-1 4.12 CONTROL ROOM FILTERING SYSTDI 4.13-1 4.13 FUEL SURVEILLANCE 4.14-1 4.14 REACTOR BUILDING PURGE FILTERING SYSTEM 4.15-1 4.15 IODINE RADIATION MONITORING FILTERS 4.16-1 4.16 RADIOACTIVE MATERIALS SOURCES 5 1-1 5
DESIGN FEATURES 5.1-1 5.1 SITE 5.2-1 5.2 CONTAINMENT 5.3-1 5.3 REACTOR 5.4-1 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 6.1-1 6
ADMINISTRATIVE CONTROLS 6.1-1 6.1 ORGANIZATION, REVIEW, AND AUDIT 6.1-1 6.1.1 Organization 6.1-2 6.1.2 Review and Audit
{S 6 )I 'f 6.2-1 ACTION TO 3E TAK'EN IN THE EVENT OF AN' INCIDENT 2
6.2 REPORTABLE TO THE COMMISSION 1
^k iv
. ? 197 5
r 3.1. 8,
, Single Loop Restrictions-Sp'ecif ica tion The following special limitations are placed on sing'le loop operation in addition to the limitations set forth in Specification 2.3.
Single loop operation is authorized for test purposes only.
3.1.8.1 At least 23 incore detectors meeting the requirements of Technical Specification 3.5.4.1 and 3.5.4.2 shall be available throughout 3.1.8.2 this test to check gross core power distribution.
The pump monitor trip setpoint shall be set at no greater than 3.1.8.3 50 percent of rated power.
The outlet reactor coolant temperature trip setpoint shall be set 3.1.8.4 at no greater than 610 F.
At 15 percent of rated power and every 10 percent of rated power above 15 percent, measurements shall be taken of each operabic 3.1.8.5 incore neutron detector and each operabic incore thermocouple, reactor coolant loop flow rates and vessel inlet and outlet temperature, and evaluation of this data de'termined to be at-ceptable before proceeding to higher power levels.
A report covering single loop operation, permitted by Specification 3.1.8, shall be submitted within 90 days af ter completion of testing.
3.1.8.6 This report shall: include the data obtained together with analyses and interpretations of these data which demonstrate:
26 (1) Coolant flows in the idle loop and operating loop are as predicted.
21 (2) Relative incore flux and temperature profiles remain es-93 sentia11y the same as for four pump operation at each power level taking into account the reduced flow in single loop operation.
(3) Operating loop temperatures and flows are obtained which justify the revised safety system setting prescribed for the temperature and flow instruments located in the operating loop (which must sense the combined core flow plus the cooler bypass flow of the idle lono).
Subecquent single loop operation shall be contingent upon Commission approval.
Bases (1) supplement the 1/6 scale model,
The purpose of single loop testing is to(2) verify predicted flow through the idle lo test information, j
that changes in power level do not affect flow distribution or core power a
3.1-19 DEC 2 21975
. distribution, and (4) demonstrate that limiting safety system settings (pump monitor trip cetpoint and reactor coolant outlet temperature trip setpoint) can be conservatively adjusted taking into account instrument errors.
Limiting the pump monitor trip setpoint to 50 percent of rated power and the reactor coolant outlet temperature trip setpoint to 610oF to perform this con-firmatory testing assures operation well within the core piotective safety limits shown in Figure 2.1-3, Curve 2.
Incore thermocouples will be installed and data will he taken to check outlet core temperature profiles.
These data will be used in evaluating test results.
h i
I 3
o S
4 0
4 d
4 3.1-19a j
DEC 2 21975
3.1.9
_1.ow Power Physics Testing Restrictions
_ Specification
(
The following special limitations are placed on low power physics testing.
3.1.9.1 Reactor Protective System Requirements Below 1720 psig shutdown bypass trip setting limits shall apply in u.
_ accordance with Table 2.3-1A - Unit 1.
2.3-1B - Unit 2.
2.3-lc - Unit 3.
b.
Above 1800 psig nuclear overpower trip shall be set at less than 5.0 percent.
Other settings shall be in accordance with Table 2.3-1A - Unit 1.
2.3-1B - Unit 2.
2.3-1C - Unit 3.
3.1.9.2 Startup rate rod withdrawal hold shall be in effect at all cimes.
This applieh to both the source and intermediate ranges.
Bases Technical Specification 3.1.9.2 will apply to both the source and intermediate ranges.
The above specification provides additional safety nargins during low power physics testing.
{3/
e s
s O
3.1-20 e, j l8h',YO]p,;,
t DEC 2 21975 i ?-
a,. x..,3 ;.,g
o 4.2 REACTOR COOLANT SYSTEM SURVE1LLANCE Applicability Applies to the surveillance of the Reactor Coolant Systua pressure boundary.
Obiective To assure the continued integrity of the Reactor Coolant System pressure boundary.
Specification Prior to initial unit operation, an ultrasonic test survey shall 4.2.1 be made of Reactor Coolant System pressure boundary welds as required to establish preoperational integrity and baseline data
~
for future inspections.
4.2.2 Post-operational inspections of components shall be made in ac-cordance with the methods and intervals indicated in IS-242 and 15-261 of Section X1 of the ASME Boiler and Pressure Vessel Code, 1970, including 1970 winter addenda,.except as follows:
IS-261 Item Component Exception.
1.4 Primary Nozzle to dessel 1 RC outlet nozzle to be Welds inspected after approxi-mately 3 1/3 years operation.
2nd RC outlet nozzle to be inspected after approx. 6 2/3 yrs.
operation.
4 RC inlet nozzles and 2 core flooding nozzles to be in-spected at or near end of interval 3.3 Primary Nozzle to Safe End Not Applicable Welds 4.3 Valve Pressure Retaining Not Applicable Bolting Larger than 2" 6.1 Valve Body W, elds Not Applicable 6.3 Valve to Safe End Welds Not Applicable 6.6 Integrally Welded Valve Not Applicable Supports 6.7 Valve Supports & Hangers Not Applicable O
DEC 2 21505
'M 4.2-1
4.2.3 shall be maintained at the icvel required by the ori ceptance standards throughout the life of the station.
cvidence, as a result of the tests outlined in Table IS-261 of Any Section XI of the code, that defects have' developed or grown, shall be investigated, including evaluation of comparable areas of the Reactor Coolant System.
4.2.4 The results of the Inservice Inspections performed pursuant to Specifica tions Commission within 90 days of completion'.4.2.1, 4.2.2, and 4.2.3 3g p1 4.2.5 13 To assure the structual integrity of the reactor internals through-the life of the unit, the two sets of main internals bolts out (connecting the core barrel to the core support shield and to th lower grid cylinder) shall remain in place and under tension e
will be verified by visual inspection to determine that This bolt locking caps remain.in place.
the welded All locking caps will be inspected after hot functional testing and whenever the internals are removed from the vessel during a refueling or maintenance shutdown.
inspected each refueling shutdown.The core barrel to core s6pport s 4.2.6 Sufficient parison and evaluation oJ future inspections. records of each in 4.2.7 The inservice inspection program shall be reviewed at five years to consider incorporation of new inspection techniq the end of and equipment which have been proved practical and the conclusion ues of this review and evaluation shall be discussed with the s
4.2.8 At approximately three-year intervals, the bore and keyway reactor coolant of each volumetric examinatian. pump flywheel shall be subjected to an in-place, necessitate flywheel removal, a surface examination of e surfaces and a complete volumetric examination shall be performed if the interval measured from the previous such inspecti greater than 6 2/3 years.
on is 4.2.9 For Unit 1 and Unit 2, a B Type vessel specimen capsule shall be withdrawn af ter one year of operation and an A Type capsule shall be withdrawn af ter 11,17, and 22 years of operation drawal schedules may be modified to coincide with those refueli The with-outages or unit shutdowns most closely approaching the withdrawal ng schedule.
with ASTM-E-185-70. Specimens thus withdrawn shall be tested in accordance shall be withdrawn after one year of operation and an capsule shall be withdrawn af ter 7,14, and 17 years of operati The withdrawal schedules may be modified to coincide with th on.
refueling outages or unit shutdowns most closely cpproaching the ose withdrawc1 schedule.
accordance with ASTM-E-185-72. Specimens thus withdrawn shall be tested in shall be reported to the Commission within 90 days of comple
%g of testing.
g ji i
L V
1g 6e 4.2-2 DEC 2 21975
9
)
p 4.2.10 During the first two refueling periods, two reactor coolant
\\
system piping elbows shall be ultrasonically inspected along their longitudinal welds (4 inches beyond each side) for clad bonding and for cracks in both the clad and base metal.
The elbows to be inspected are identified in B&W Report 1364 dated December 1970.
Bases The surveillance program has been developed to comply with Section XI of the ASME Boiler and Pressure Vessel Code. Inservice Inspection of..? clear Reactor Coolant Systems, 1970, including 1970 winter addenda, edition.
.'ie program places major emphasis on the area of highest stress concentrations and on areas where fast neutron irradiation might be sufficient to change material properties.
The reactor vessel specimen sbrveillance program for Unit 1 and Unit 2 is
~
based on equivalent exposure times of 1.8, 19.8, 30.6 and 39.6 years.
The contents of the different type of capsules are defined below.
A Type B Type Weld Material llAZ Material RAZ Material Baseline Material Baseline Material
(\\
For Unit 3 the Reactor Vessel Surveillance Program is based on equivalent exposure times of 1.8, 13.3, 26.7, and 30.0 years.
selected and fabricated as specified in ASTM-E-185-72.The specimens have been Early inspection of Reactor Coolant System piping elbows is considered desirable in order to reconfirm the integrity of the carbon steel base metal when explosively clad with sensitized stainless steel.
If no degradation is observed during the two annual inspections, surveillance requirements will revert to Section XI of the ASME Boiler and Pressure Vessel Code.
e 4
4.2-3 DEC 2 21975
I 4.4 HEACTOR BUILDING
~
4.4.1 Containment Leakage Tents Applicability Applies to containment Icakage.
Objective To verify that Icakage f rom the Reactor Building is mai tained within allowabic limits.
Specification 4.4.1.1 Integrated Leak Rate Tests 4.4.1.1.1 Design Pressure Leak Rate The maximum allowabic integrated Icak rate, La, from the Reactor Building at the 59 psig design pressure, P, shall not exceed 0.25 weight percent of the p
building atmosphere at that pressure per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.1.1.2 Testing at Reduced Pressure The periodic integrated Icak rate test may be performed at a test pressure, Pg, of not 1 css than 29.5 psig provided the esultant Icakage rate, Lt, dpes not exceed a pre-established fraction of La determined as follows:
n.
Prior to reactor operation the initial value of the integrated leak rate of the Reactor Building
' '. b : measured at design pressure and at the reduced periodic integrated leak rate tests.
The leak pressure to be used L
..i rates thus measured shall be identified as Lpm and Ltm respectively.
Lt shall not exceed la(Ltm/ Lpm) for va}ues of (Ltm/ Lpm) not greater than 0.7.
b.
Lt shall not exceed La(Pt/Pp) for values of (Ltm/ Lpm) above 0.7.
c.
d.
If Ltm/ Lpm is less than 0.3, the initial integrated test results shall be subject to review by the URC to establish an acceptable value of Lt.
l 4.4.1.1.3 Conduct of Tests The test duration shall be at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, except that if both the a.
following conditions are met, the test duration shall be at least 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />s:
(1) All test conditions, including the test procedure, shall be similar to the initial integrated leak rate tests.
(2) When the test is terminated, building pressure shall have stabilized and shall not be increasing.
i b'
4.4-1 y
DEC 2 21975 3
~
~
7 -
1
b.
Test accuracy shall be verified by' supplementary means the quantity of air required to return to the starting pointsuch as measuring posing a known Icak rate to demonstrate the validity of measurements or by im-
~
Closure of containment isolation valves for the'purpase of the te t c.
be accomplished by the means provided for normal operation of the v l s shall without preliminary exercises or adjustment.
a ves 4.4.1.1.4 Frequency of Test After the initial preoperational leak rate test, shutdown for inservice inspection, tests shall be performed at app ween each major addition, an integrated leak rate test shall be performed at each 10 y In interval, coinciding with the inservice inspection shutd ear own.
4.4.1.1.5 Conditions for Return to Critica.'
a.
If Lt local Icak rate testing need not be completed prior to retu cality following a periodic integrated leak rate test.
o criti-b.
If Lt is greater than 50 percent and not greater than 100 percent value permitted in 4.4.1.1.2, return to criticality will be perform'ed of the conditioned upon demonstration taat local Icakage into the penetration room, measured at full design pressure-50 percent of that permitted by 4.4.1.1.2., accounts for all leakage above If this cannot in demon-strated within 30 days of returning to criticality, the reactor shall b j
shut down.
e the unit shall not be made critical.If Lt is greater than 100 percent of th c.
i 4.4.1.1.6 Corrective Action and Retest i
integrated leak rate test need not be repeated, provide
.., the measurements are made before and af ter repair to demonstrate that ate leak rate to an acceptable value. rate reduction achieved by repairs reduces the
)
the leak ae 4.4.1.1.7 Report of Test Results The results of the initial Containment integrated leak rate t be submitted to the Commission within 90 days. of comp est and subsequent which shall
,~ g
'g
... test.
4.4.1.2 Local Lenk Rate Tests 13 4.4.1.2.1 Scope of Tasting The loct.'
icak rate shall ba measured for each of the followi ng components:
I tgw 4.4-2 DEc 2E I970
I r
a.
Personnel hatch b.
Emergency hatch Equipm'nt hatch seals c.
e d.
Fuel transfer tube seals c.
Reactor Building normal sump drain line f.
Reactor coolant pump seal outlet line g.
Reactor coolant pump seal inlet line h., Quench tank drain line 1.
Quench tank return line
- j. Quench tank veist line k.
Normal makeup to Reactor Coolant System 1.
High pressure injection line m.
Electrical penetrations n.
Reactor Building purge inlet line o.
Reactor Building purge outlet line I
p.
Reactor Building sample lines q.
Reactor coolant letdown line 4.4.1.2.2 Conduct of Tests a.' Local leak rate tests shall be performed at.a pressure of not less than 59 psig.
b.
Acceptable methods of testing are halogen gas detection, soap bubbles, pressure decay, hydrostatic flos or equivalent.
1 4.4.1.2.3 Acceptance Criter'a The total 1cakage from all penetraticas and isolation valves shall not exceed 0.125 weight percent of the Reactor Building atmosphere per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.1.2.4 Corrective Action and Retest j
a.
If at any time it is determined that the criterion of 4.4.1.2.3 above is i
exceeded, repairs shall be initiated immediately, b,
If conformance to the criterion of 4.4.1.2.3 is not demonstrated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following detection of excessive local leakage, the reactor shall be shut down and depressurized until repairs are effected and the local leakege meets the acceptance criterion as demonstrated by retest.
4.4.1.2.5 Tes t - Frequency Local leak detection tests shall be performed annually, except that:.
a.
The equipment hatch and fuel transfer tube seals shall be additionally tested after each opening.
b.
The personnel hatch and emergency hatch outer door seals'shall be tested at four-month intervals, except when the hatches are not opened during that interval.
In no case shall the test interval be longer than 12 months.
4.4-3 O
DEC 2 01975.
l Isolati.a Valve Functional Tests 4.4.1.3 Quarterly, remotely-operated Reactor Building isolation valves sh function unless auch f
stroked to the position required to fulfill their sa et -The latter valves shall be operation is not practical during unit operation.
tested during each refueling shutdown.
Annual Inspection 4.4.1.4 orior surfaces of the A vinual examination of the accessibic interior and.evfdrmedannuallyand containment structure and it s components shall be pertouncoverany{cvidenceofdeterior prior to any integrated leak rate test, l integrity or leak-tightness.
which may af fect either the containment's structuraThe disc ied by cor-destructive tests rective actions in accord with acceptable procedures, non-ior to the conduct of l
and inspections, and local testing where practica, prResults of the inspection shal og
,[ 3 19 i
any integrated Icak rate test.
to the Commission within 90 days of completion.
Reactor Building Modifications h Reactor 4.4.1.5 Any major modification or replacement of components affecting t e d leak r. ate test Building integrity shall be followed by either an integrateas appropriate or a local Icak rate test, criteria of 4.4.1.1.4 and 4.4.1.2.3, respectiv.ely.
Bases Building is designed for an internal pressure of 59 psig and a The Reacto steam-air mixture temperature of 286*F.
design pressure and leak rate tainment is strength tested at 115 percent ofThe containment is also leak tested pri re.
These tested at the design pressure. initial operation at approximately 50 perce tests verify that the leak rate from Reactor Bu the relationships given in the specification.
it life The performance of a periodic integrated Icak rate test during un in frem the containment, provides a current assessment of potential, leakasc i
f the containment.
case of an accident that would pressurize the inter or oI f the containment under accident conditions, this periodic. test is toliminary isolation valves are to be closed in the normal manner.for the periodic ide an accurate measurement of the leak rate and it dupThe specification pro tial Icakage at leak rate test at 29.5 psig.
relating the measured leakage of air at 29.5 psig to the po lly keyed to the refueling schedule for the reactor, because 39 psig.
be performed during refueling shutdowns.
is based on The specified frequency of. periodic integrated leak rate testsFirst is three major considerations.
t DEC 2 21975 4.4-4
I s
y R
its sir,nificance to the load-carrying. capability of the structure. The h
cheathing filler will be sampled and inspected for changes in physical
[
eppearance.
Wire sampics shall be selected in such a manner that kith the third inspection.
wires '! rom all nine surveillance tendons shall have been inspected and tested.
4.4.2,2 Inspection Intervals and Reports C
For finit 1, the initial inspection shall be within 18 months of the initial
)
t Rear. tor Building Structural Integrity Test. The intpection intervals, nensured i
l fr';m the date of the initial inspection, shall be two years, four years and every five years thereafter or as modified based on experience.
For Units 2 and 3 the inspection intervals measured from the date of the initial structural test shall be one year, three years and every five years thercafter or as
,7 modified based on experience.
Tendon surveillance may be conducted during reactor operation provided design conditions regarding loss of adjacent g-tendons are satisfied at all times..
g A quantitative analytical report covering results of each inspection shall be
- submitted to the Conmission within 90 days of completion, and shall especially
,," g
3 address the following conditions, should they develop:
13 a.
Broken wires.
b.
The force-time trend line for any tendon, when extrapolated, that extends beyond either the upper or lower bounds of the predicted design band.
Unexpected changes in corrosion conditions or sheathing filler properties.
c.
4.4.2.3 End Anchorage Concrete Surveillance a.
The end anchorages and adjacent concrete surfaces of the surveillance tendons will be inspected.
In addition, other locations for surveillance
' k will be determined by information obtained from design calculations, pre-E stressing records, observations, and deformation measurements made during f,
prestressing.
i b.
The inspection interval will be approximately one-half year and one year 4
after the operation of the unit and will occur during the warmest and j
coldest part of the year.
t t..
c.
The inspections made shall include:
E 5
(1) Visual inspection of the end anchorage concrete exterior surfaces.
S (2) A determination _of the temperatures of the liner plate area or con-tainment interior surface in locations near the end anchorage concrete under surveillance.
(3) Measurement of concrete temperatures at specific end anchorage 7
concrete surfaces being inspected.
~
^
c 4.4-7 DEC 2 21975 i
M
('
i (4) The mapping of the predominant visible concrete crack patterns.
(5) The. measurement of the crack widths, by use of optical comparators or wire feeler gauges.
(6) The measurement of movements, if any, by use of denountable mechanical extensometers.
d.
me measurements and observations shall be compared with those to which prestressed structures have been subjected in normal and abnormal load conditions and with those of preceding measuremchts and observations at the same location on the reactor containment.
c.
The acceptance criteria shall be as follows:
If the inspections determine that the conditions are favorable in compari-son with experience and predictions, the close inspections will be termi-nated by the last of the inspections stated in the schedule.
If the inspections detect symptoms of greater than normal cracking or movements.
.an immediate investigation will be made to determine the cause, f.
Results of the inspection shall be reported to the Commission within 90 30 days of completion.
31 5
13 4.4.2.4 Liner Plate Surveillance 4.4.2.4.1 The liner plate will be examined prior to the initial pressure test in accessibic areas to determine the following:
Location of areas which have inward deformations. The a.
magnitude of the inward deformations shall be measured and recorded.
These areas shall be permanently marked for future reference and the inward deformations shall be measured between the angle stiffeners which are on 15-inch The measurements shall be accurate to i 0.01 centers.
inch. Temperature readings shall be obtained on both the liner plate and outside containment vall at the locations where inward deformations occur, b.
Locations of areas having strain concentrations by visual examination with emphasis on the condition of the liner surface.
The location of these areas shall be recorded.
4.4.2.4.2 Shortly after the initial pressure test and approximately one year after initial startup, a re-examination of the areas located in Section 4.4.2.4.1 shall be made. Measurements of the inward deformations and observations of any strain con-centrations shall be made.
4.4.2.4.3 If the difference in the measured inward deformations exceeds 0.25 inch (for a particular location) and/or changes in strain concentration exist, an investigation shall be made.
The investigation will determine any necessary corrective action.
4.4-8' DL.C 2 21975
4.4.2.4.4 The surveillance program shall be dincontinued after the one year after initial startup inspection if no corrective action was needed.
If corrective action is required, the frequency of inspection for a continued surveill.nce program shall be determined.
'9 4.4.2.4.5 Results of the surveillance shall be reported to the Com-o p
l e3 3l mission within 90 days of completion.
Bases Provisions have been made for an in-service surveillan e program, covering the first several years of the life of the unit, intended to provide suf-ficient evidence to meintain confidence that the intendity of the Reactor Building is being preserved.
This program consists of tendon, tendon 1
anchorage and liner plate surveillance.
To accomplish these programs, the following representative tendou groups have been selected for surveillance:
Horizontal - Three 120 tendoria comprising one complete hoop system below grade.
Verticel - Three tendons spaced approximately 120 apart.
Dome - Three tendons spaced approximately 120 apart.
The inspection during this initial period of at least one wire from each of the nine surveillance tendons (one wire per group per inspection) is con-sidered sufficient representation to detect the presence of any wide spread tendon corrosion or pitting conditions in the structure.
This program will be subject to review and revision as warranted based on studies and on results obtained for this and other prestressed concrete reactor buildings during this period of time.
4 REFERENCES (1) FSAR Section 5.6.2.2 s
N 4.4-9 l
DEC 2 21975 E.
P
[
t 9
4.4.3 Hydronen Purne System Applicability
()
Applies to testing Reactor Building Perge System.
Objective To verify that this system and components are operable'.
I Specification 4.4.3.1 Operating Tests t
An in-place system test shall be performed annually.
This test shall consist of a visual inspection, hook-up of the systeu to one of the three reactor buildings, a flow measurement using flow instruments in the portable purging st'ation and pressure drop measurements across the filter banks.
Flow shall be design flow or higher, and pressure drops across the filter bank shall not exceed two times the pressure drop when new.
Fan motors shall be operated continuously for at 1 cast one hour, and valves shall be proven operable.
This test shall demonstrate that under simulated emergency conditions the system can be taken f rom storage and placed into operation within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
4.4.3.2 Filter Tests Annually, leakage tests using DOP on HEPA units and Freon-112 (or
( ')
equivalent) on charcoal units shall be performed at design flow on the filter.
Removal of 99.5% DOP by each entire HEPA filter uni t and removal of 99.0% Freon-112 (or equivalent) by each entire charcoal absorber unit shall constitute acceptable performance.
These tests must also be performed after any maintenance which may affect the structural integrity of either the filtration system units or of the housing.
4.4.3.3 H2 Detector Test Hydrogen concentration instruments shall be calibrated annually with proper consideration to moisture effect.
Bases The purge system is composed of a portable purging station and a portion of the Penetration Room Ventilation System.
The purge system is operated as necessary to maintain the hydrogen concentration below the control linit.
The purge discharge from the Reactor Building is taken from one of the Penetration Room Ventilation System penetrations and discharged to the unit A suction may be taken on the Reactor Building via isolation valve vent.
PR-7 (Figure 6-5 of the FSAR) using the existing vent and pressurization connections.
g 4s 4.4-10 DEC 2 21975
4.13 PUEL SURVEILLANCE 8
s Applicability
~
Applies to the fuel surveillance program for fuel rods of Unit 1.
Obiective To specify the fuel surveillance program for fuel rods.
Specification 4.13.1 Visual Inspection Two (2) Oconce Unit 1 fuel assemblics will be designated for visual inspection. These same assemblies will be inspected during each of the first three refuelings of Unit 1.
Underwater viewing devices will be used to determine that the fuel rods have maintained their structural integrity.
4'.13.2 Dimensional Exaninstion Measurements of the length and outside diameter will be mad,e on select ed peripheral rods of the following fuel assemblics of the first core of Unit 1 both prior to operation and at the times
~
specified:
One assembly after the first cycle.
n.
b.
Four assemblies after the second cycle.
c.
Two assemblies af ter the third cycle.
4.13.3 Results of the fuel surveillance program shall be submitted to the 2G
.'2' Commission within 90 days of completion of the program.
t g1 Bases This fuel surveillance program provides substantiating information for the first core in,the present generation of BLW reactors.
It provides for examination of fuel rods at the end of the first, second, and third cycles of Unit 1 to determine if fuel rods have maintained their integrity and to determine the extent, if any, of dimensional changes in diameter and length.
\\
4.13-1 DE: 2 21975 A
i,
+
c.
tjuorum The chairman plus two members shall constitute a quorum.
d.
Rcaponsibilitics The committee shall have the following responsibilities:
1.
Review all new procedures or changes to existing proc dures determined by the station Manager or his designate to affect o,pe ational safety.
2.
Review station operation and safety considerations.
3.
Review reportabic occurrences and violations of Techni al Specifica-l2G 31'/417 tions and make recommendations to prevent recurrence.
4.
Review all proposed tests that affect nuclear safety or radiation safety.
5.
R2 view proposed changes to Technical Specifications and safety-related changes or modifications to the station design.
c.
Authority
'The Station Review Committee shall make recommendations to the station Manager regarding Specification 6.1.2.1-d.
f.
Records Minutes of all meetings of the committee shall be.kept at the station, and copics shall be sent to the station Manager, Vice President, 12G 2 '1[ I 3 Steam Production, and the chairman of the Nuclear Safety Review Committee.
6.1.2.2 Nuclear Safety Review Committee The Executive Vice President and General Manager shall appoint a Nuclear a.
Safety Review Committee having responsibility to verify that operation of the station is consistent with company policy and rules, approved operating procedures, and license provisions; to review important pro-posed station changes, and tests; to verify that abnormal occurrences and unusual events are promptly investigated.and corrected in a manner which reduces the probability of recurrence of such events; and to detect trends which may not be apparent to a day-to-day observer.
b.
The activities of the Nuclear Safety Review Committee shall be guided by a written charter that contains the following:
Subjects within the purview of the committee Responsibility and authority Mechanisms for convening meetings i
Provisions for use of specialists or subgroups 6.1-2 DEC 2 91975 l
f s
f.
Hecting Frcquency:
intervals not The committee shall meet at Icast three times per year at During to exceed iive months and as required on call by the chai,rman.
the period of initial operation, this committee shall meet at least once par calendar quarter.
g.
Quorum:
The chairman or vice-chairman plus three members, or appointed alternates, No more than a minority of the quorum shall shall constitute a quorum.
have direct line responsibility for station operation.
h.
Meeting Hinutes:
Minutes of all scheduled meetings of the cotznittee shall be prepared and These minutes shall shall identify all documentary materials reviewed.
be formally approved, retained, and also promptly distributed to the Executive Vice President and General Manager; Senior Vice President, Engineering and Construction; Senior Vice President, Production and Trans-g 36[ 2 1g 5
mission; Vice President, Design Engineering; Vice President, Steam Production; and station Manager.
A copy of these minutes shall be kept on file at the station.
As a safety review to the normal operating organization, the committee i.
shall review the following:
Proposed tests and experimentn, and results thereof, when these con-1.
stitute an unreviewed safety question defined in 10CFR50.59.
Proposed changes in equipment or systems which constitute an unreviewed 2.
safety question defined in 10CFR50.59, or which are referred by the operating organization.
All requests to the NRC/DRL for changes in Technical Specifications l 3 G[2 '1 13
~
3.
or license that involve unreviewed safety questions as defined in 10CFR50.59.
Violations of statutes, regulations, orders, Technical Specifications, 4.
license requirements, or internal procedures, or instructions having
' safety significance as determined by the NSRC.
2 O!I f5 Reportable Occurrences as defined in 6.6.2.1 of these specifications.
5.
l2G[2I[13 Vice President Special reviews or investigations as required by the 6.
Steam Production, or the station Manager, President, 6.1-4 DEC 2 21975
6.2 ACTION TO BE TAKEN IN THE EVENT OF.A REPORTABLE OCCURRENCE I Yp th/13 6.2.1
'Any reportable occurrence shall be investigated promptly by the station Manager.
6.2.2 The station M nager shall promptly notify the Vice a
President, Steam Production, of any reportabic occurrence.
,,,I /,3
/*
j The Station Review Committee shall review a written report iwhich shall describe the circumstances Icading up to and resulting from the occurrence and shall recommend appropriate action to prevent or minimize the probability'of a recurrence.
6.2.'3 The Station Review Committee report shall be submitted to the Nuclear Safety Review Committee for review of any recommendations.
Copies shall also be sent to the station Manager and the Vice President, Steam Production.
O t
e S
J e
9 4
6.2-1 DLC 2 21975 i
-l b
~ua
6.6 STATION REPORTING REQUIREMENTS 6.6.1 Rput_ine Report,n The fulluwing reports shall he submitted to the Director, Office of inspection and Enforcement Region II, Atlanta, Georgia.
6.6.1.1 Startup Report A summary report of unit startup and power escalation 1esting shall be submitted following (1) receipt of an operating licensq, (2) amendment to the facility license involving a planned increase in power level, (3) installation of fuel that has a different design or has{ been manufactured by a dif ferent fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the unit.
Startup reports shall be submitted (1) within 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) nine months following initial criticality, whichever occurs first.
If a startu'p' report does not cover all three events, i.e., initial criticality, completion of the startup test program and re-sumption or commencement of commercial power operation, supplementary reports shall be submitted at least every three months until all three events are completed.
6.6.1.2 Annual Operating Repor.
Routine operating reports covering the operation of the unit during the a
previous calendar year shall be submitted prior to April 1 of
,_,ch year.
The initial report shall be submitted prior to April 1 of the year following init_a1 criticality.
Each annual operating report shall provide the following:
a.
Operations Summary (1) A narrative summary of operating experience during the report period relating to safe operation of the facility, including safety-related maintenance not covered in 6.6.1.2.a(2e)
(2) For cach outage or forced reduction in power 1 of over 20 percent of
~
design power 1cvel where the reduction extends for greater than four hours.
1/The term " forced reduction in power" is defined as the occurrence of a component failure or other condition which requires that the load on the unit be reduced for corrective action immediately or up to and including the very next wcckend.
Note that routine preventive maintenance, sur-
' veillance and calibration activitics requiring power reductions are not covered by this section.
Entire Page Revised 6.6-1 DEC 2 21975
e v
e l.
(a) the proximate cause and the system and major component involved (if the outage or forced reduction in power involved equipment
. malfunction);
(b) a brief diocussion of (or reference to icpo ca of) any reportable occurrences pertaining to the outage or power reduction; (c) corrective action taken to reduce the probability of recurrence, if appropriate; (d) operating time lost as a result of the dutage or power reduction (for scheduled or forced outages,2/ use the generator off-line hours; for forced reductions in power, use the approximate duration of operation at reduced power);
(c) a description of major cafety-related corrective maintenance performed during the outage or power reduction, including the system and component involved and identification of the critical path activity dictating the length of the outage or power reduc-tion'; and (f) a report of any single release of radioactivity or unusual radiation exposure specifically associated with the outage which accounts for more than 10 percent of the allowable arnual values.
b.
Changes, Tests and Experiments A brief description and the summary of the safety evaluation for those
~
changes, tests, and experiments carried out without prior Commission approval pursuant to the provisions of 10CFR50.59.
Reporting of Radioactive Effluent Releases 2/
c.
i
)
Data shall be reported to the Commission in a form similar to that shown in Table 6.6-1 and shall include the following:
(1) Caseous Releases (a) Total radioactivity (in curies) releases of noble and activation gases.
(b) Maximum noble gas release rate during any one-hour period.
(c) Total radioactivity (in curies) released, by nuclide, based on representative isotopic analyses performed.
l 2/The term " forced outage" is defined as the occurrence of a component failure or other condition which requires that the unit be removed from service for corrective action immediately or up to and including-the very next veckend.
3/ Shall be reported on a semi-annual basis.
6.6-2 Entire Page Revised DEC 2 ' 1975 e
~
(d) Percentage applicable limit n released.
('/) lo line lh leasen (a) Total 1-131, 1-133, 1-135 radioactivity. (in curies) released.
(b) Total radioactivity (in curies) released, lby nuelide, based on representative isotopic analyses performe.
(c) Percentage of limit.
(3) Particulate Releases (a) Cross radioactivity (S-y) released (in curi,es) excluding back-ground radioactivity.
(b) Cross alpha radioactivity released (in curies) excluding back-ground radioactivity.
(c) Total radioactivity released (in curies) of nuclides with half-lives greater than eight days.
(d) Percentage of limit.
(4) Liquid Releases (a) Gross radioactivity (8-y) released (in curies) excluding tritium "and average concentration released to the unrestricted area at the Ecowee llydro unit.
(b) The maximum concentration of gross radioactivity (B-y) released to the unrestricted area (averaged over the period of release).
(c) Total tritium and alpha radioactivity (in curies) released and average concentration released to the unrestricted area at the
~Keowce Hydro unit.
(d) Total dissolved gas radioactivity (in curies) and average con-centration released to the unrestricted area at the Keowae Hydro unit.
(c) Total volume (in liters) of Keowee Hydro liquid waste released.
(f) Total volume (in liters) of dilution water used prior to release from the restricted area.
(g) Total radioactivity (in curies) released, by nuclide, based on representative isotopic analyses performed.
(h) Percentage of limit for total activity released.
L
[
6.6-3 Entire Page Revised DEC 2 21975
~
(5) Solid Waste of solid waste packaged (in cubic f eet).
(a) The total anount (b) Estimated total radioactivity (in curies):
(c) Disposition including date and destination if sh4pged off site.
~
(6) Environmental Monitoring (a) For each medium sampled during the reporting period, the following information shall be provided.
1.
Number of sampling locations.
2.
Total number of sampics.
Number of locations at which levels are found to be sig-3.
nificantly greater than local backgrounds.
4.
Highest, lowest, and the average concentrations or levels of radiation for the sampling point with the highest average and description of the location of that point with respect to the site.
(b) If levels of station-:ontributed radioactive materials in en-vironmental media indicate the likelihood of public intakes in excess of 3 percent of those that could result from continuous exposure to the concentration values listed in Appendix U.
Table 11, Part 20, estimates the likely resultant exposure to individuals and to population groups, and assumptions upon which estimates are based shall be provided.
(These values are com-parable to the top of Range 1, as defined in FRC Report No. 2.)
(c) If statistically significant variations in off-site environmental concentrations with time are observed and are attributed to station releases, correlation of these results with effluent releases shall be provided.
d.
Personnel Exposure and Monitoring A tabulation (supplementing the requirements of 10 CFR 20.407) of the number of personnel receiving exposures greater than 100 mrcm in the reporting period and their associated man-rem exposure, according to duty function, e.g., routine plant surveillance and inspection (regular duty), routine plant maintenance, special plant maintenance (describe maintenance), routine fueling operation, special refueling operation (describe operation), and other job-related exposures.
e.
Fuel Examinations
/
- Indication of failed fuel resulting from irradiated fuel examinations, includ-ing results. of eddy current tests, ultrasonic tests, or visual examinations 1
l :;
completed during the report period.
\\~
6.6-4 Entire Page Revised DEC 2 0975
6.6.2 Non-houtine R ports 3
6.6.2.3 Reportable Occurrences Prompt Notification with Uritten Followup a.
The types of events listed below shall be reportedi with n 24. hours of discovery (by telephone, telegraph, r.:ailgran, or fhesimile transmission to the Director, Office of Inspection and Enforcer.fnt, Region II, or his designate) uith a writ. ten followup report within tuo wechs' to the Director, Office of Inspcction and Enforcement, Regica 71 (cc py to the Director, Office of Management Information and Program Control, USNR -).
(1) railure of the Reactor Protective System t.o trip,.as required, wher, a monitored parameter reaches the setpoint z.pecified as the limiting safety syr. ten setting in the Technical Specifications.
(2) Operation of the unit or af fected systems vhen any parameter or operation r.ubject to a limitinp. condition for operation is less l
conservative than the 3 cast conservative aspect of the liuiting j
condition for operation established in the Technical Specifications.
r (3) Abnormal degradation discovered in fuci c3 adding, reactor c*oolant pressure boundary or primary centainment..
(4) Reactivity anomalies involving dir.ngreement. with predicted value of resc;1vity bc3cnce under st eady-state conditions great.cr than or equ:11 to W. AK/h; a caltu3ated s cact.!vity balance indicnting shut.dcun margin 3ess conservative than specified in the technical specifications; short-terrn reactivity increases t hat correspond to a reac tor period of less thah 5 seconds, or if suheritical, an unplanned reactivity insertion of more than 0.5% Ak/h; or any unplanned criticality.
(5) railure or mlfunction of one or mere componcnts which prevents or could prevent, by itself, the fulfillment of the functional require-ments of systeins required to cope with accidents analyzed in the Safety Analysis Report.
(6) Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillipent of the funct.Jonal requirements of systems required to cope with achidents analyzed in the Safety Analysis Report.
(7) Conditions arising from natural or man-made events that, as a direct result of the event, require unit shutdown, operation of safety systems, or other protective measures required by Technical Specifi-Cations.
(S) Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the Safety Analysis Report or in the bases for the Technical Specifications that have or could have pernitted reactor operation in a manner less conservative than assuned in the analyses.
y fI,'
k
' DEC 2 21975_
n 6.6-5 Entire Page Revised
~
b.
Thirty,, Day _Urf tten Reports The types of events listed below shall be the subject of written reports to the Director, Office of Inspection and Enforcement, Region II, within 30 days of discovery of the event.. (Copy to the Director, Office of Manage-ment Information and Program Control, USNRC).
(1) Reactor protection system or engineered safety f eature instrunent settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfill-ment of the functional requirements of affected systems.
I (2) Conditions leading to operation in a degraded upde permit ted by a limiting ccndition for operation or shutdown required by a limiting condition for operation.
(3) Observed inadequacies in the impicnentation of aduinistrative or operation of a unit which could cause procedural controls durina.
reduction of degree of redundancy provided in the Reactor Protective System or Engineered Safety Feature Systems.
6.6.2.2 Environmental Monitorii.g If individual milk sampics show 1-131 concentrations of 1 picoeuries per liter or greater, a plan shall be submitt ed within one ue. k advising the a.
related annual doses will NRC of the proposed action to ensure the plant be within the design objective of 15 nrem/yr to the thyroid of any indi-vidual.
If allk sampics collected over a calendar quarter shou average concentrations b.
of 4.8 picocuries per liter or greater, a plan shall he submit t ed within 30 related days advising the URC of the proposed action to ensure the plant annual doses will be within the design objective of 15 mrem /yr to the thyroid of any individual.
annual report period, a measured 1cyc1 of radioactivity If, during any c.
in any environmental medium other than those, associated with gaseous radiciodine releases exceeds ten times the control station value, a written notification will be submitted uithin one veck advising the NRC This notification should include an evaluation of any of this condition.
release conditions, environmental factors, or other aspects necessary to ex' lain the anomalous result.
p d.
If, during any annual report period, n~ measured lbvel of radioactivity in any environmental medium other than those associated with gaseous radictodine releases exceeds four times the control station value, a written notification will be submitted within 30 days advising the.NRC of This notification should' include au cvaluation of any this condition.
reletse conditions, environmental factors, or other aspects necessary to explain the anomalous result.
_ nam
_%4 DEC 2 21975 Wes Entire Page Revised 6.6-6
6.6.3 Special Reports Special reports shall be submitted to the Director, Office.of Inspection and En-forcement, Region 11, within the time period specified 'or each report. These re-ports shall be submitted covering the activities identil Jed below pursuant to the requirements of the applicahic reference specification; a.
Electrical System Degradation, Specification 3.7.
b.
Excessive Liquid Waste Releases, Specification 3.9 Excessive Gaseous Waste' Releases, Specification 3. O.
c.
d.
Inservice Inspection, Specification 4.2.4.
c.
Reactor Vessel Specimen Surveillance, Specification;4.2.8.
f.
Containment Integrated Leak Rate 12st, Specificatioty 4.4.1.1.7.
g.
Reactor Building Annual luspection Report, Specification 4.4.1.4.
h.
Tendon Stress Surveillance, Specification 4.4.2.2.
i.
End Anchorage Concrete Surveillance, Specification 4.4.2.3.
j.
Liner Plate Surveillance, Specification 4.4.2.4.
k.
Single Loop Operation, Specification 3.1.8.
1.
Fuel Surveillance Program, Specification 4.13.
P00R ORENAL 6.6-7 Entire Page Revised DEC 2 21975 o
o L-Dg NE3r-
= -
=
L A
TO T
ll' i
l !s II!
I
- lII i s
1)l!
1
,j
}
i i
I lY I i
!i '
i '
Il li ;l ia ili t
ce D
i '
l
! t tI3
,6
!i} +
, )e vo N
l 8
l I
3
- I i
~
6 i tcO li i 4 '
i t
1
- (
t ce r
S ae Y
li i i ?!
?
f I!i II !
I i '
i tf }4i ru A
I i
!!i Iil l
8 I
i ll}
- il
.!'f-v lu J
1 II 8
I
!i I
e t
1 n
u S
J TY E
lii
'J 3
?
8 lliI
( !
s
.*Il LF y
T a
E M
E V
I II II I!
i t
~
I 3
l t
1 i
- iL.
C 1
r A
e D
A 6
I D
6 A i I I
!I 6 i I lii I
!I 4!'
R E
r 6
(t I t
6 i
f*
F a
L O M
B T
A R
T O
!4 f
I
!l 14 8
i!!I i
!?6 4
i i i i 1}l
- !l.'.t P
E R
b e
T Ili ii1
!I $
I 1 iiI '
n.
,I a
J il iiIi
!I I8 i
.M l
s 1 s
e - 'e s I a 1 s 1
= e e t
e ~
c ~
e-r -.
i ff i /
i/
i/
in r ii
- c rf r1 tf rf t - -
U u cc u C uc ue
.1 e
C u
u c
r c'
c u
. 4 < n I
HIf3 IIYi
!I l
!III' ltl l
I sd
!YI1,t 4* IIt i.
(
e e f
e e
1 l
ms s
J s
s J
a a
e e f
e e
s l, l l.
l l
i a
ee f
e e
d cy lr J
r r
it
)y Dn 0
n t
n t
r iv y
o fi Co 9
o i
a e
U ci (i
1 i
y i
e tt et N
l s
t i
l t
pc O
aa a
r a
c y
r t.
T 3
r G
a s
sa I
T t
- t. t.
g t
r.
a t-w m
g e a
- z. g :
u
- n o1 ll TA i
n3 g
n NT v
r> g3 aa AS ie
=
ct
'd a
d ip fmg.
anri i
t-io 1
m g
e t
a1 an TR r
sl l
n e p0 nt g
ka u
go e
n R ra q
h OA e
g C.
ru lc l
C F.
s oe cr I. c.
i1il e
r irtn 2
c eo L
a l 1 d-tf r e y
r n
i itr l
RC7 e
d p
1_ g F e
R Mg aLa r - n b
r E L'0 l
f ei 0
ft 1
3 r_ t s.
WN-e il
- d. ta r.
o e c. e1 s4 n
r ga 3
oi O
A R
t x
e lt e
PE/
s e ma.
1v oe v 3ts m r. c p -
e:
(1,7 A 3
57 4 ti ES d
s T A_M t
5 I. /
u t & *qL391 ca9 13J33 081 45 0
e sT A_
o nl EN -
t i
- 4 O S u
g))
i r )
)
i) l q
1 1 1
65 5 56 9
cn E_ lL l 3 o b s t+s r-t Nab
.4
.l t CN e
e T
a b Dab rot o r-S
- ccs s o DOC a
z V gUdB S i
X;c C C CC
- n1 r l_
I P. teiu L
'.Z 2{Il 5 8.
p 7
8 l.{
I m*
I.
oW t to Io.%U 1
0 0
t l
g L
ATOT lI!l' l l I
I
!i l
e I
l Iil 6
8 i
i ii
!lI I
!ll c
e D
v.
)
N
,1i l
tcO ll tp e
r S
ae Y
i t
!I l
.ru A
v lu J
1l l
)
e D
n E
u S
J T
m N
I F
ilI T
U N
LT w
O F
a C
E M
(
E V
I 1
T C
r A
p 6
0 A
I D
6 A
I l
1I P
E r
F a
L O
M BA T
I I!
,ii l
R T
O PE R
be F
Ij na J
iI~
IiIi c
s. s s
9 e
s i n e e
c e
s e
f e l i
f i
/
i n r, r r
r i
r u
u u
u C
u U.
C C
c c u
C c
.i 16 J
- r e
o o
a t
f i
h a
d o
r t
a l
i r
a e
m s
i s
s a
l s
s e
o o
l e
s r
r e
l e
g g-r b
s a
a f
e e
s e
N e
e t
t at t
d O
l r
a a
I e
a s l1 l
s i
s e
p s
e s
T R
c n
, uy e
n e t
a c'
u
- c. t l
a s a
e s
iLiii.
YA e
NT e
e t
b a s l
l a
0 AS n
o t(
t t y o
f gn u
e t
4 1
Q_h P
r l
r irl n
o
,c e
r a 1
MR o
a a y r a t.
bg.gr yt e
l 7
s 3
3 8 OA b
s pt
- t. p q m
t o2 a039 n
1 3
5 5
l D37 M1i CE r
5 3t 1
i a
u n
e1 L91 3 e3 33 8 gd8 6
L i
ll l v ll n m
e L.
o a
g1 1
1 a r-RC8 A
aa ai aai i
c n h p
t ;s r
s r
o-MS a
r r ca v
U0 t t t t ttd x
r orc WN C S l
I 1
I 3K XKK K
X v r A
o o oe noa a
e s a a
q O
A I
T T Tn T T.r M
P ab c
I P
H a
PE/
I E5 EN -
1 2 1 4
5 6
7 8
ROS UCN DOO OnO o 5t h dNe i
QD~W** -g% MF
)
s f
.pe