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Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20217D7961999-10-12012 October 1999 Proposed Tech Specs Pages,Removing Turbine EHC Low Oil Pressure Trip from RPS Trip Function Requirements in TS Sections 2.2 & 3/4.1.A ML20210R8281999-08-13013 August 1999 Revised Bases Page B.3/4.9-6 to TS Section 3/4.9,providing Clarity & Consistency with Sys Design Description in UFSAR Sections 8.3.2.1 & 8.3.2.2 ML20209J2321999-07-16016 July 1999 Proposed Tech Specs 3/4.7.D Replacing Limit for Any One Msli Valve of Less than or Equal 11.5 Sfch with Aggregate Value of Less than or Equal 46 Scfh for All MSIVs ML20196K1941999-06-30030 June 1999 Rev 2.0 to Chapter 11 of Quad Cities Offsite Dose Calculation Manual ML20209C2951999-06-29029 June 1999 Proposed Tech Specs Section 3/4.3.C, Reactivity Control - Control Rod Operability ML20211C3311999-04-30030 April 1999 Rev 2.0 to Generic ODCM for Dresden,Quad Cities,Zion, Lasalle,Byron & Braidwood ML20205L2631999-04-0505 April 1999 Tech Spec Page B 3/4.5-2 to TS Section 3/4.5, ECCS, to Clarify Requirement Discussed in ML20205J9741999-03-30030 March 1999 Proposed Tech Specs,Deleting Various License Conditions That Have Been Completed,Making Editorial Changes & Providing Clarifying Info ML20205J9321999-03-30030 March 1999 Proposed Tech Specs 3/4.6.E Changing SRs 4.6.E.2 to Allow one-time Extension of 18 Month Requirement to Pressure Test or Replace One Half of MSSVs to Interval of 24 Months ML20205J9911999-03-30030 March 1999 Proposed Tech Specs Allowing Alternative Methodology for Quantifying RCS Leakage When Normal RCS Leakage Detection Sys Is Inoperable ML20199L6921999-01-21021 January 1999 Proposed Tech Specs Section 3/4.6.I,relocating from Chemistry TS Requirements to UFSAR ML20199L7741999-01-21021 January 1999 Proposed Tech Specs Bases for Sections 3/4.10.K & 3/4.10.L, Provides Description of Design & Operation of RHR SD Cooling Subsystem ML20196H4571998-11-30030 November 1998 Proposed Tech Specs 3/4.8.J, Safe Shutdown Makeup Pump, Reducing Current AOT from 67 Days to 14 Days ML20196F6451998-11-30030 November 1998 Proposed Tech Specs 3/4.1.A,3/4.10.B & 3/4.12.B,proposing Changes to Relocate Requirement to Remove RPS Shorting Links Which Enable non-coincident Scram for Neutron Instrumentation,To Licensee Controlled Document ML20196K5861998-11-0505 November 1998 Rev 3 to Qcap 0280-01, Process Control Program for Processing of Radioactive Wet Wastes at Quad Cities Nuclear Power Station ML20155D8091998-10-29029 October 1998 Proposed Tech Specs Bases Sections 3/4.2.D & 3/4.5.D, Providing Clarity & Consistency with Sys Design Description Contained in UFSAR Section 5.4.6.2 ML20195J9041998-09-24024 September 1998 Rev 0 to TR-VQ1500-02, Clean ECCS Suction Strainer Head Loss Test Rept ML20151S7991998-08-31031 August 1998 Proposed Tech Specs,Increasing Max Allowable MSIV Leakage from 11.5 Scfh to 30 Scfh Per Valve When Tested at 25 Psig, IAW SR 4.7.D.6 ML20236W8401998-07-31031 July 1998 Proposed Tech Specs Bases 3/4.7.C & 3/4.7.12.C,clarifying Testing Requirements for Primary Containment Excess Flow Check Valves ML20247D7761998-05-0505 May 1998 Proposed Tech Specs Page B 3/4.4-1,changing Administrative Error.Bases for Net Quantity of Gallons for Solution Is Changed from 3254 (Correct Quantity) to 3245 ML20246Q3481998-04-29029 April 1998 TS Page B 3/4.5-3,reflecting Change to TS Bases for Section 3/4.5.C ML20217G1481998-03-27027 March 1998 Proposed Tech Specs Bases Section 3/4.5.A,reflecting Design Info Contained in Rev 4 to Ufsar,Dtd Apr 1997 ML20216C6381997-08-29029 August 1997 Proposed Tech Specs,Incorporating New Siemens' Methodologies That Will Enhance Operational Flexibility & Reducing Likelihood of Future Plant Derates ML20196G0271997-05-0101 May 1997 Proposed Tech Specs 4.9.A.8.b Revising Load Value for Diesel Generator to Be Equal to or Greater than Largest Single Load & Revising Frequency & Voltage Requirements During Performance of Test ML20138G3321997-04-29029 April 1997 Proposed Tech Specs,Permitting Loading of ATRIUM-9B Fuel in Plant Unit Core for Operational Modes 3,4 & 5.Modes Will Support Refueling Activities Such as Fuel Load,Vessel re- Assembly & Single Rod Timing ML20138B3231997-04-21021 April 1997 Proposed Tech Specs,Requesting That NRC Grant Exigent Amend to TS 2.1.B & 6.9.A.6.b to Support Plant Unit 2 Cycle 15 Operation Scheduled to Begin 970519 ML20137G3981997-03-26026 March 1997 Proposed Tech Specs 3/4.7.P Re Standby Gas Treatment & TS 5.2.C Re Secondary Containment ML20135F7321997-03-0303 March 1997 Proposed Tech Spec Bases 3/4.9.E,clarifying Purpose of SR 4.9.E ML20135D9461997-02-24024 February 1997 Proposed Tech Specs,Clarifying Bases for TS Surveillance 4.8.D.5.c ML20138L4011997-02-17017 February 1997 Proposed Tech Specs Section 2.1.B Re Thermal Power,Section 3/4.11 Re Power Distribution Limits,Section 3/4.6 Re Primary Sys Boundary,Section 5.3 Re Reactor Core & Section 6.9 Re Reporting Requirements ML20138L3701997-02-17017 February 1997 Proposed Tech Specs 4.9.A.8.h Re Diesel Generator Endurance Test Surveillance Requirements ML20134D2191997-01-27027 January 1997 Proposed Tech Specs Deleting marked-up Sentence from TS Bases for Section 3/4.7.K ML20129K3321996-10-18018 October 1996 Cycle 15 Startup Test Results ML20129C2391996-10-16016 October 1996 Proposed Tech Specs for Dresden 2 & 3 & Quad Cities 1 & 2, marked-up to Show Transition Verbiage ML20129D3981996-09-20020 September 1996 Proposed Tech Specs 3/4.6.K,updating Pressure-Temp Curves to 22 Effective Full Power Yrs & TS Bases ML20216H8841996-06-30030 June 1996 Revs to ODCM for Quad Cities,Including Rev 1.8 to Chapters 10,11,12 & App F ML20116F3971996-06-30030 June 1996 Rev 1.8 to ODCM, Annex,Chapters 10,11,12 & App F ML20113C3571996-06-25025 June 1996 Proposed Tech Specs Re Upgrade Program ML20113A7861996-06-10010 June 1996 Proposed Tech Specs,App A,To Reflect Transition of Fuel Supplier from General Electric to Siemens Power Corp ML20117D7121996-05-0606 May 1996 Proposed Tech Specs,Implementing New LCO & SR Re Revs to TS for 10CFR50,App J,Lrt ML20107A1881996-04-0404 April 1996 Proposed Tech Specs 3.4/4.4 Re Standby Liquid Control Sys ML20101H1381996-03-25025 March 1996 Complete Version of TS Upgrade Program Pages That Reflect Current Configuration of Plant & Specifies SRs That Will Not Be Current Upon Implementation of Tsup Project ML20097D9231996-02-0808 February 1996 Proposed Tech Specs,Upgrading Existing TS 3/4.5, Eccs ML20100C0441996-01-24024 January 1996 Secondary Containment Leak Test Summary ML20093K7721995-10-12012 October 1995 Quad-Cities Nuclear Power Station Unit 2 Cycle 14 Startup Test Results Summary ML20098A3821995-09-20020 September 1995 Proposed Tech Specs,Revising TS Upgrade Program & Improving Plant Submittals ML20086D4741995-06-30030 June 1995 Proposed Tech Specs Re TS Upgrade Program for Dresden Units 2 & 3 & Quad Cities Units 1 & 2 ML20087H8651995-05-0202 May 1995 Proposed Tech Specs Re TS Upgrade Program Section 3/4.10 ML20082H7481995-04-10010 April 1995 Proposed Tech Specs,Revising SR for HPCI & RCIC Sys ML20080K8171995-02-23023 February 1995 Proposed Tech Specs,Changing Name of Iige to Reflect Results of Merger Between Iige,Mid American Energy Co,Midwest Power Sys Inc & Midwest Resources Inc 1999-08-13
[Table view] Category:TEST REPORT
MONTHYEARML20195J9041998-09-24024 September 1998 Rev 0 to TR-VQ1500-02, Clean ECCS Suction Strainer Head Loss Test Rept ML20129K3321996-10-18018 October 1996 Cycle 15 Startup Test Results ML20100C0441996-01-24024 January 1996 Secondary Containment Leak Test Summary ML20093K7721995-10-12012 October 1995 Quad-Cities Nuclear Power Station Unit 2 Cycle 14 Startup Test Results Summary ML20078M9041994-11-23023 November 1994 Quad-Cities Nuclear Power Station Unit One Cycle Fourteen Startup Test Results Summary ML20078B9221994-10-20020 October 1994 Reactor Containment Bldg Integrated Leak Rate Test,Quad- Cities Nuclear Power Station,Unit 1,940723-24 ML20059G0031993-10-29029 October 1993 Revised Quad-Cities Nuclear Power Station Unit 2 Cycle 13 Startup Test Results ML20056F7161993-08-24024 August 1993 Quad-Cities Nuclear Power Station,Unit 2 Cycle 13 Startup Test Results Rept ML20056F0621993-08-19019 August 1993 Quad-Cities Nuclear Power Station,Unit 1 Cycle 12 Startup Test Results ML20056F0631993-08-19019 August 1993 Quad-Cities Nuclear Power Station,Unit 2 Cycle 12 Startup Test Results ML20046B5781993-05-19019 May 1993 Reactor Containment Bldg Integrated Leak Rate Test Quad- Cities Nuclear Power Station. ML20044C0631993-03-0909 March 1993 Cycle 13 Startup Test Rept Summary. W/930309 Ltr ML20102A6931992-07-23023 July 1992 Quad-Cities Nuclear Power Station,Unit 2,Cycle 12 Startup Test Results ML20099H0521992-04-0606 April 1992 Reactor Containment Bldg Integrated Leak Rate Test, Quad-Cities Nuclear Power Station Unit II for Period 920401-06 ML20085A9551991-07-19019 July 1991 Quad-Cities Nuclear Power Station Unit 1 Cycle 12 Startup Test Results ML20077G7941991-03-0202 March 1991 Reactor Containment Bldg Integrated Leak Rate Test. W/ ML20011F5141990-02-26026 February 1990 Quad-Cities Nuclear Power Station Unit 1 Cycle 11 Startup Test Results. W/900226 Ltr ML20011E7161990-02-0606 February 1990 Reactor Containment Bldg Integrated Leak Rate Test,Quad Cities Nuclear Power Station Unit 1,891114-15. ML20055C8591989-10-31031 October 1989 Special Neutron Attenuation Test for High Density Spent Fuel Racks (Wet), Final Rept ML20045H5641989-09-19019 September 1989 Equipment Qualification Rept for Moore,Isolator,Moore Signal Transmitter & Marathon Fuse Block,CWE-3840,890919. W/930713 Ltr ML20206F4641988-11-16016 November 1988 Reactor Containment Bldg Integrated Leak Rate Test, 880612-13 ML20205H1071988-10-19019 October 1988 Cycle 10 Startup Test Results ML20153G0421988-06-13013 June 1988 Reactor Containment Bldg Integrated Leak Rate Test ML20150C1541987-12-0606 December 1987 Reactor Containment Bldg Integrated Leak Rate Test ML20153G3481987-11-25025 November 1987 Rev 1 to Irradiation Study of Boraflex Neutron Absorber Interim Test Data, Interim Technical Rept ML20238E2911987-06-25025 June 1987 Rev 0 to Irradiation Study of Boraflex Neutron Absorber, Interim Test Data ML20205K1401987-03-26026 March 1987 Quad-Cities Nuclear Power Station Unit 2 Cycle 9 Startup Test Results ML20210C8591986-10-15015 October 1986 Reactor Containment Bldg Integrated Leak Rate Test, Quad-Cities Nuclear Power Station,Unit 2,861014-15 ML20199F1321986-06-17017 June 1986 Quad-Cities Nuclear Power Station Unit 1 Cycle 9 Startup Test Results ML20197A7621986-03-23023 March 1986 Reactor Containment Bldg Integrated Leak Rate Test, Quad-Cities Nuclear Power Station,Unit 1,860322-23 ML20134N7871985-08-28028 August 1985 Quad-Cities Nuclear Power Station Unit 2,Cycle 8,Startup Test Results ML20135F1211985-05-28028 May 1985 Reactor Containment Bldg Integrated Leak Rate Test, for 850526-28 ML20108A4921984-11-0808 November 1984 Cycle 8 Startup Test Results ML20100N1091984-08-31031 August 1984 Reactor Vessel Irradiation Surveillance Program Analysis of Capsule 8, Final Rept ML20107F0991984-07-24024 July 1984 Reactor Containment Bldg Integrated Leak Rate Test, Quad-Cities Nuclear Power Station,Unit One,840724-27 ML20084N0341984-05-0808 May 1984 Cycle 7 Startup Test Results ML20100N1151984-03-31031 March 1984 Reactor Vessel Irradiation Surveillance Program Analysis of Capsule 18, Final Rept ML20084F7091984-02-10010 February 1984 Reactor Containment Bldg Integrated Leak Rate Test ML20082V2381983-12-13013 December 1983 As-Found Type B & C Local Leak Rate Test Results for Unit 1 During Last Three Refueling Outages ML20071A8371983-02-17017 February 1983 Quad-Cities Nuclear Power Station,Unit 1,Cycle 7 Startup Test Results ML20028F1151983-01-20020 January 1983 Reactor Containment Bldg Integrated Leak Rate Test, Quad-Cities Nuclear Power Station,Unit One,821216-17. ML20042A6411982-03-15015 March 1982 Quad-Cities Nuclear Power Station Unit 2,Cycle 6 Startup Test Results. ML19341D6711981-04-0303 April 1981 Quad-Cities Nuclear Power Station,Unit 1,Cycle 6 Startup Test Results. ML19318B6651980-06-19019 June 1980 Cycle 5 Startup Test Results. ML19321B1711980-03-12012 March 1980 Reactor Containment Bldg Integrated Leak Rate Test. ML19257D1571979-12-31031 December 1979 Quad-Cities Nuclear Power Station Units 1 & 2 Secondary Containment Leak Rate Test Summary, Per Tech Specs Section 6.6.C.3 ML19225A5521979-07-0505 July 1979 Reactor Containment Bldg Integrated Leak Rate Test, 790219-22 ML19263D2101979-03-0606 March 1979 Secondary Containment Leak Test Rept.Concludes Standby Gas Treatment Sys Maintains 1/4 Inch of Water Vacuum in Secondary Containment Bldg Under Calm Conditions W/Flow Rate No Higher than 4,000 Cfm ML17187A8031969-05-16016 May 1969 Cavitation Test Rept 12x14x14-1/2 Cvds Pump. 1998-09-24
[Table view] |
Text
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6 QUAD-CITIES NUCLEAR POWER STATION UNIT 1 CYCLE 12 STARTUP TEST RESULTS 1
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i TABLE OF CONTENTS Test No. Title Page .
1 Shutdown Margin 1
{
2 Core Verification 2 !
3 Initial Criticality 2 ;
1 4 TIP Reproducibility and 3 Core Power Symmetry Analysis :
l 1
SDdGR'.UISTVTUP
'1. Shutdown Margin Demonstration and Control Rod Functional Checks ;
Purpose l The purpose of this test is to demonstrate for this core loading in the most reactive condition during the operating cycle, that the reactor is .
subcritical with the strongest control rod full out and all other rods ;
i fully inserted.
criteria If a shutdown margin of 0.322% oK (0.25% + R + B 4 C settling penalty) cannot be demonstrated with the strongest control rod fully withdrawn, i the core loading must be altered to achieve this margin. The core reactivity has been calculated to be at a maximum 4000 mwd /ST into the cycle and R is given as 0.032% AK. The control rod B4 C settling penalty for Unit One is 0.04% AK.
Results and Discussion On January 5, 1991 and April 19, 1991, control rod M-7 was fully withdrawn to demonstrate that the reactor would remain subcritical with the strongest rod out. This maneuver was performed twice due to a change in the cooling water alignment in the reactor between the two dates. Rod M-7 was calculated by GE to have the highest worth with the ,
core fully loaded. The strongest rod out maneuver was performed to -
allow single control rod withdrawals for CR0 testing.
Control Rod functional subcritical checks were performed as part of control rod friction testing. No unexpected reactivity insertions were observed when any of the 177 control rods were withdrawn and all control rod drives functioned properly. !
General Electric provided rod worth information for the two strongest diagonally adjacent rods L-6 and N-6 with rod M-7 full out. This method provided an adequate reactivity insertion to demonstrate the desired shutdown margin. On April 21, 1991, a diagonally adjacent shutdown margin demonstration was successfully performed. Using the G.E.
supplied rod worth for M-7 (the strongest rod) and diagonally adjacent ;
rods L-6 and N-6, it was determined that with M-7 at position 48, L-6 at position 48, and N-6 at position 08, a moderator temperature of 134cF, i and the reactor subcritical, a shutdown margin of 0.926% AK was i demonstrated. The G.E. calculated shutdown margin with M-7 withdrawns )
and 68cf reactor water temperature was 2.533% AK at the beginning of J cycle 12.
l At approximately 4000 mwd /ST into cycle 12 a minimum calculated shutdown l margin of 2.501% oK will occur with M-7 fully withdrawn.
1 l
snaontuisnerce I
3
~
G.E.'s ability to determine rod worth was demonstrated by the accuracy of their in-sequence criticality prediction. The oK difference between -
the expected critical rod pattern and the actual critical rod pattern '
was determined to be 0.065% oK after correcting for temperature and .
period. This initial critical demonstrated that the actual shutdown :
margin at the beginning of cycle 12 was 2.598% AK and 2.566% AK at 4000 :
mwd /ST into cycle 12.
- 2. Core Verification Purpose :
The purpose of this test is to verify proper core location and orientation for each core fuel assembly.
Criteria Prior to reactor startup the actual core configuration shall be verified to be identical to the planned core configuration.
Results and Discussion The Unit One Cycle 12 core was verified on January 4,1991. Fuel assembly orientation, seating, and ID serial number were verified for each assembly. Two passes were made over each assembly. The first pass to verify orientation and seating of assemblies. The second pass to verify bundle ID numbers. A video camera was used during the inspection. All assemblies were found to be properly seated and orientated in their designated locations.
On January 8, 1991, 16 fuel assemblies were reverified due to moving 4 fuel assemblies for LPRM flange work. Two passes were again made for orientation, seating and ID verification. All 16 assemblies were found to be properly seated and orientated in their designated location. !
The bundle ID numbers are shown in Figure 1. f
- 3. Initial Critical Prediction Purpose The purpose of this test is to demonstrate General Electric's ability to >
calculate control rod worths and shutdown margin by predicting the insequence critical.
Criteria General Electric's prediction for the critical rod pattern must agree i within 1% AK to actual rod pattern. A discrepancy greater than 1% oK will be cause for an On-Site Review and investigation by Nuclear Fuel Services.
STMGRiUISTRTUP
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?
Results and Discussion
{
On April 24,1991, at 0424 hours0.00491 days <br />0.118 hours <br />7.010582e-4 weeks <br />1.61332e-4 months <br /> the reactor was brought critical with reactor water temperature at the time of criticality of 170cF. The oK difference between the expected critical rod pattern at 680F and the actual critical rod pattern at 170oF was 0.00276 AK from rod worth tables supplied by General Electric. The temperature effect was
-0.00175 oX from General Electric supplied corrections. The excess reactivity yielding the 166 second positive period was 0.00036 AK. These reactivities result in a 0.00065 AK difference (0.0656% AK) between the expected critical rod pattern and the actual rod pattern. This is within the 1% AK required in the criteria of this test, and General Electric's ability to predict control rod worth is, therefore, successfully demonstrated.
i
- 4. Core Power Distribution Symmetry Analysis I
Purpose The purpose of this test is to determine the magnitude of indicated core power distribution asymmetries using data (TIP traces and OD-1) ;
collected in conjunction with the CMC update. '
Criteria A. The total TIP uncertainty (including random noise and geometric f uncertainties obtained by averaging the uncertainties for all data sets) must be less than 9%.
t B. The gross check of TIP signal symmetry should yield a maximum deviation between symmetrically located pairs of less than 25%. [
t Results and Discussion !
Core power symmetry calculations were carried out based upon a computer [
program OD-1 data run on May 9, 1991 and July 9, 1991 at 100% power. l The average total TIP uncertainty from the two TIP sets was 4.260%.
The random noise uncertainty was 0.926%. This yields a geometrical '
noise uncertainty of 4.158%. The total TIP uncertainty was well within the 9% limit. l r
Table 1 lists the symmetrical TIP pairs and their respective average -
deviations. Figure 1 shows the core location of the TIP pairs and their '
average TIP readings. The maximum deviation between the symmetrical pairs was 12.763% for pair 32-41. The maximum deviation between symmetrically located pairs for the single 0D-1 data run was well within the 25% limit.
stucasuisirnir ;
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The method used to obtain the uncertainties consisted of calculating the i
- average of the nodal ratio of TIP pairs by:
3 n 22 1 I I Rij
_R = 18n j=1 i=5 where Rij is the ratio for the ith node of TIP pair j, there being n such pairs, where n=18.
Next the standard deviation of the ratios is calculated by:
n 22 I I (Rij - R)2 1/2 o_= j=1 i=5 4
R (18n - 1) ;
o is multiplied by 100 to express o, as a percentage of the ideal ;
value of o, of 1.0.
% o, = g,x 100 l
The total TIP uncertainty is calculated by dividing % ir, byV 2 in i order to account for data being taken at 3 inch intervals and analyzed ;
on a 6 inch nodal basis.
In order to calculate random noise uncertainty the average reading at i each node for nodes 5 through 22 is calculated by:
~
MT NT
= 1 I I BASE (N, M, K)
BASE (K) NT x MT M=1 N=1
~
where NT = number of runs per machine - 5 MT = number of machines = 5 BASE (K) = average reading at nodal level K, K = 5 through 22 1
1 The random noise is derived from the average of the nodal variances by:
22 MT NT
~
1/2
%g noise = K= M= N= Bd5 (k) x 100 18 (NT x MT -1)
Finally the TIP geometric uncertainty can be calculated by:
% o geometric = (% s total' - % o noise')2/* l Table 1 STMGR\UISTRTUP l
.i, CORE SYMMETRY i Based on OD-l's From 1 05-09-91 and 07-09-91 (100% power) l
, 1 SYMMETRICAL TIP AVERAGE PAIR NUMBERS ABSOLUTE DIFFERENCE % DEVIATION a-b T= T - T, % = 100 X T/((T + T,)/2) 1-6 3.770 4.376 2-12 6.230 6.677 3-19 6.030 6.508 ,
4-26 4.600 5.043 5-33 2.785 5.448 8-13 6.930 5.560 !
4 9-20 2.645 2.270 10-27 2.550 2.239 11-34 1.875 1.807 15-21 13.750 11.159 l 16-28 11.255 9.013 ;
17-35 3.280 2.901 !
18-39 1.590 2.638 i 23-29 4.180 3.326 ;
24-36 3.040 2.797 ,
25-40 2.150 3.315 l 31-37 2.380 2.040 ;
32-41 5.760 12.763 :
t 22 Average % Deviation = 5.021 1 T,= lE T,(K) /18 l i=5 '
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a susta==u =e UNIT ONE POk'ER SYMMETRY AVERAGE BASE READINGS e asi- tamous"* (NODE 5-22)
BASSED ON OD-l's from 5-9-91 (100% PO'w'ER) 7-9-91 (100% POk'ER)
.