ML20064N980
| ML20064N980 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, McGuire  |
| Issue date: | 02/14/1983 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20064N977 | List: |
| References | |
| PROC-830214, NUDOCS 8302170047 | |
| Download: ML20064N980 (30) | |
Text
{{#Wiki_filter:_ _ - - _. __ . _ . . _ . . . _ . . . _ _ . _ _ _ _ ._ _. McGUIRE NUCLEAR STATION Calculational Methodology for Equipment Qualification Radiation Dose This report illustrates the model used for calculating dose rates.and integrated doses for equipment qualification purposes. Information pertaining to releases of radioisotopes are from NUREG 0588 and NUREG 0737. The Core isotopic inventory is calculated using the ORIGEN computer. code for maximum equilibrium concentrations at maximum fuel burnup. See Table 1 for the list of major radioisotopes important to radiation dose. 1 I. Containment - I. a. Gamma Dose Inside Containment The source terms used to calculate radiation dose rates inside con-tainment are: 100% noble gases, 50% iodines and 1% remaining fission products both in the containment atmosphere and in the liquid of the containment sump. Concentrations of radioisotopes in the atmosphere and in the sump are listed in Table 1. Source Strengths and energies
- are included in this table. It is assumed that these fission products are instantaneously released into the containment free volume and distributed throughout upper and lower containment.
'l The containment is modeled as a right-circular cylinder 55.77 feet in radius, 106.2 feet in height or a free volume of 1.038 ft.3 For conservatism, the detector point representing equipment inside of containment is chosen at the centerline. (r, z,e) = 0, 53 feet, 00).
1 A dose rate of 5.06 x 106 Rads / hour is calculated using the KAP-VI computer code ~ assuming no credit for decay, shielding, leakage, plateout or washout by sprays. This cose rate is determined to be the maximum value at any location within containment post-accident. Source terms and dose rates are generated at this contairsent location for times following the DBA to account for radioactive decay. Dose rates are normalized to the maximum time-zero value and integrated using trapezoidal methodology for one year post-accident. This integration factor (20.9) times the time zero dose rate reprgsents the one year
- integrated dose inside of containment (1.06 x 10o Rads). In this way, a time dependent fraction of the one-year accident dose can be plotted vs. time for equipment qualification duration less than one year, See Figure 6, (for 30 dags the integration fraction is. 7 or 70% of the i one-year dose = 7.4 x 10' Rads). The Reactor Building is sectioned into
- radiation zones to account for variations in the 40-year normal operating dose. The sum of this 40-year dose and the post-accident dose is added
! to give the total integrated dose to equipment. See Table 2 " Reactor Building," Example 1 and Figure 1 for radiation dose data. j I i i l t 83KX2170047 830214 i
- PDR ADOCK 05000369 P PDR
I. b. Inside Containment Compartment Doses For equipment located inside of lower containment compartments, i.e., accumulator tank rooms, credit is taken for concrete shielding from the containment atmosphere activity. Analysis performed assumed source terms the same as in Section I. a. Two independent calculations were performed using different analysis; point kernal methodology using KAP-VI as in Section I. a. and volume adjusted semi-infinite cloud methodology from Regulatory Guide 1.4. Both analyses gave results of 5.5 x 105 Rads / Hour dose rate at time = 0 hours post-accident. Since the same source of activity is used for the accumulator room, the fraction of the one-year integrated dose, Figure 6, can be used for equipment qualified for times less than one year. The one-year integrated dose is the product of the integration factor times the T = 0 hour dose rate: 5.5 x 105 x 20.9 = 1.2 x 107 Rads. See example 4 for tables corresponding to the accumulator tank room dose data. I. c. Beta Dose In-Containment A beta dose calculation was performed assuming an infinite uniform cloud containing concentrations equivalent to those used in Section I. a. above. The calculation uses the beta dose equation in Regulatory Guide 1.4 with average beta energies per isotope from ORNL/NUREG/TM-102. The one-year accident dose from beta radiation was calculated to be 8.4 x 108 Rads. Based on guidance from IE Bulletin 79-OlB and NUREG-0588, 70 mils thickness of cable insulation is sufficient to reduce the beta dose to approximately one-tenth of the gamma dose and therefore can be considered negligible. All Class lE equipment located inside containment that is required to mitigate a LOCA, MSLB, or HELB has sJfficient shielding to prevent the exposure of any organic materials associated with this equipment to a beta radiation environment. II. Reactor Building Annulus The equipment located in the annulus, i.e., between the containment shell and Reactor Building wall, will be subjected to radiation exposure from the in-containment atmosphere and from containment leakage into the Annulus. The source terms for in-containment exposure are the same as Section I above. The geometry of the source is separated into three zones: sources inside of the cranewall, sources outside of the cranewall, and sources in the containment dome. Detector points are distributed at various elevations inside of the annulus to account for the three source geometries and credit is taken for the 3 feet thick concrete cranewall and the 1" steel containment shell shielding. Annulus gaseous activity is detennined as in Section I above. Radio-isotopes are distributed equally throughout the annulus volume and a dose rate is calculated at detector point locations vertically distri-buted throughout the annulus. The total integrated dose to equipment in the Annulus is the sum of exposure from in-containment activity,
gaseous activity in the annulus and exposure from 40 years of normal operation. See example 2. III. Auxiliary Building Sources of exposure in the Auxiliary Building are from recirculating Reactor sump liquids through the safety injection, containment spray and residual heat removal systems. Fractions of core radioisotopes in the liquid containing systems are: 100% noble gases 50% iodines 1% remaining fission products (See sump concentrations at T = 0 time in Table 1) A 10-minute delay time is assumed for the start of recirculation, after which, the only method of removal is decay. Shielding models for piping and components for each room are developed for use in the computer codes SHIELD and KAP-VI. Average dose rates are , calculated for 10-minute decayed liquid activity in each room affected by the LOCA. (See example 3). A normalized dose rate decay curve from a standard geometry model is integrated over the one-year post-accident, the sum of which gives an integrating factor to be applied to the maximum dose rate in each room to give the one-year integrated dose. It was shown that the 10-minute delay time has a negligible effect on this integration factor compared to the in-containment model. Therefore, Figure 6 can be used for equipment qualification times less than one year post-accident for equipment located inside of the Auxiliary Building affected by recircu-lating sump liquids. IV. Areas Affected by In-Containment Radiation Streaming Outside of Containment The McGuire Reactor Building has two personnel access hatches which lack Reactor Building concrete shielding. Personnel access hatches were modeled to reflect radiation streaming exposure to areas adjacent to these " openings". Areas affected are the spent fuel pool area and the lower electrical penetration room. (See Table 4 and Figure 5 for radiation data.) V. Auxiliary Building Ventilation Sources of activity in Auxiliary Building results from the use of the Annulus Ventilation System. Due to containment leakage, and the need , to maintain a negative pressure in the annulus, post-accident periodic purges are necessary through the VE System filter bed. The 4" equivalent carbon bed is assumed to remove 100% of the iodine. The only method of removal is through decay of the isotopes of iodine. In the same way as Section I and III, the integrated dose to the area affected by the VE filters is plotted versus time for equipment operating for less than one-year post-accident (Figure 7).
~
The second source of activity results from containment bypass leakage into the Auxiliary Building through the Reactor Building concrete wall. A special compJter Code, BPAUX was developed to give the Auxiliary Building Ventilation System, VA, filter bed loading as a function of time post-accident. The resultant equipment exposure is the sum of the effects of annulus purge and bypass leakage (See Table 3, el. 767 and Figure 4 for radiation dose data.) An additional calculation was performed for a source of activity in ; the VA filter bed assuming leakage from ECCS canponents. This source I was considered negligible compared to bypass leakage. I RGE/sr
McGUIRE NUCLEAR STATION Examples of Radiation Data Reported in NUREG 0588 Response
- 1. Equipment ID: Valve Motor (Table 5)
Operators (Upper Containment) Manufacturer: Limitorque Model #SMB Radiation Accident environment for operation time post-accident of
~
5 minutes. Assuming i hour time duration see Figure 6 for the fraction of the 1-year dose. Read .035 from Figure 6 Next, find radiation zone on Figure 2 where valve operator is located. Read: Zone 23 El. 778+10 Next, find on Table 2, the Zone number and Reactor Building elevation. Read: Nonnal operating LOCA Dose 40-year dose (Rads) (Rads) 5.0 x 104 1.0EE Next, multiply the 1-year LOCA dose by the fraction of the 1-year integrated dose.
- 1.0E8 x .035 = 3.5 x 10 6 Rads Add 40-year normal operating dose to this value to obtain the Total Integrated Dose (TID) for this equipment
1 Hour Post-LOCA Dose + 40-Year Normal Operating Dose : TID 3.5 x 106 Rads + 5 x 104 Rads = 3.6 x 106 Rads t or 4 x 106 Rads reported in 0588 submittal. l
- 2. Equipment ID: Transmitter-Pressurizer Level (Table 8)
Manufacturer: Barton Model #764 (Lot 2) Operability required in accident environment = 2 weeks l See Figure 6 for the fraction of the one-year integrated dose l Read: - 0.6 Next, see manufacturer's file for location of equipment and radiation zone. (Table 8) \ Read: Radiation Zone 8a Elevation 738+3 ! See Table 2 for Zone 8a, elevation 738+3.
t Read: Nomal Operating LOCA Dose 40-Year Oose (Rads) (Rads) 2.0E4 2.0E7 Next, multiply the fraction of the one-year integrated dose times the one year LOCA dose to obtain the 2-week accident dose: 2.0 x 107 Rads x 0.6 = 1.2 x 107 Rads Add the 40-year normal operating dose to get the total integrated dose: 1.2 x 10 7Rads + 2.0 x 104 Rads = 1.2 x 107 Rads or 1.4 x 107 Rads as reported in NUREG 0588 response.
- 3. Equipment Located in the Annulus: Table 6 Equipment ID: Transmitter - RCS Pressure (WR)
Manufacturer: Rosemount Model # 1153 GA9 Or rability required in accident environment = 2 weeks See Figure 6 for the fraction of the LOCA dose absorbed in 2 weeks or 336 hours. Read: ' O.6 from Figure 6 Next, find location and elevation of the equipment in the Annulus. (From equipment qualification file). Zone 10, elevation 738 + 3 Read: Normal Operation LOCA Dose 40-year dose (Rads) (Rads) 1.0E4 Rads 2.0E7 Now, multiply the LOCA dose times the fraction of the one year integrated dose, to get the 2 week dose. 2.0E7 x .6 = 1.2 x 107 Rads Now, add the 40-year nomal operating dose to the post-accident dose to get the total integrated dose. 1.2 x 10 7 Rads + 1.0 x 104 Rads = 1.2 x 107 Rads as reported in NUREG 0588 response.
- 4. Equipment located in the Auxiliary Building effected by post-LOCA recirculating fluid. Table 7.
Containment spray pump motoes Manufacturer: Westinghouse, Buffalo Model # - 73F56019-1573, -2573, -3573, -4573.
See Figure 3 for radiation zone number for elevation 695 Read: Zone 3 See Table 3, Zone 3, elevation 695 for the Auxiliary Building Read: DBA-LOCA Normal Operating 40-year Dose (Rads) Dose (Rads) 5.2 x 105 1.0 x 103 This equipment is qualified for the entire one-year post-LOCA, therefore add both values to get: TID = 5.2 x 105 Rads
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l TABLE 2 REACTOR BUILDING Radiation Data Normal Oper. Normal Oper. + Dose Rate 40 Yr. Dose LOCA Dose LOCA Dosel TID 2 Elevation Zone (R/Hr.) (Rads) Rate (R/Hr.) (Rads) (Rads) 696 + 11 1 2.9E2 1.0E7 4.8E6 1.0E8 1.1E8 725 + 0 2 2.9El 2.0E7 4.8E6 1.0E8 1.2E8 3 2.9E0 1.0E6 4.8E6 1.0E8 1.0E8 4 2.9E-2 1.0E4 1.0E6 2.0E7 2.0E7 g 738 + 3 5 1.4E1 5.0E6 4.8E6 1.0E8 1.0E8 8 6 5.7El 2.0E7 4.8E6 1.0E8 1.2E8 7 2.9E-1 1.0E5 4.8E6 1.0E8 1.0E8 8 5.7E-2 2.0E4 4.8E6 1.0E8 1.0E8 8a 5.7E-2 2.0E4 6.0E5 2.0E7 2.0E7 9 5.7E-1 2.0E5 4.8E6 1.0E8 1.0E8 10 2.9F-2 1.0E4 1.0E6 2.0E7 2.0E7 767 + 11 11 1.4E1 5.0E6 4.8E6 1.0E8 1.0E8 12 5.7E0 2.0E6 4.8E6 1.0E8 1.0E8 13 5.7El 2.0E7 4.8E6 1.0E8 1.2E8 14 2.9E-1 1.0E5 4.8E6 1.0E8 1.0E8 15 5.7El 2.0E7 4.8E6 1.0E8 1.2E8 16 5.7E-2 2.0E4 4.8E6 1.0E8 1.0E8 17 2.9E-2 1.0E4 1.5E6 3.0E7 3.0E7 778 + 10 18 1.0E1 3.5E6 4.8E6 1.0E8 1.0E8 19 2.9E-1 1.0E5 4.8E6 1.0E8 1.0E8 20 1.4E1 5.0E6 4.8E6 1.0E8 1.0E8 21 5.7E0 2.0E6 4.8E6 1.0E8 1.0E8 22 5.7E-2 2.0E4 4.8E6 1.0E8 1.0E8
TABLE 2 (Sheet 2) REACTOR BUILDING Radiation Data Normal Oper. Normal Oper. Dose Rate 40 Yr. Dose LOCA Dose LOCA Dosel TID 2 Elevation Zone (R/Hr.) (Rads) Rate (R/Hr.) (Rads) (Rads) 23 1.4E-1 5.0E4 4.8E6 ) OE8 1.0E8 778+10 24 2.9E-2 1.0E4 1.5E6 3.0E7 3.0E7 1.4E1 5.0E6 4.8E6 1.0E8 1.1E8 841 + 10 25 1.0E8 26 1. 4 E-1 5.0E4 4.8E6 1.0E8 5.7E-2 2.0E4 4.8E6 1.0E8 1.0E8 27 _4 2.9E-2 1.0E4 2.2E6 4.5E7 4.5E7 g; 28 R 1.4E-1 5.0E4 4.8E6 1.0E8 1.0E8 n3 842 + 5 29 1.0E8 30 2.9E-2 1.0E4 4.8E6 1.0E8
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I TABLE 3 ;, I AUXILIARY BUILDING {! Radiation Data i I Normal Oper. , Max. Dose DBA-LOCA1,2 40 Yr. Dose TID 3 ; Elevation Zone Rate (R/Hr.) Dose (Rads) (Rads) (Rads) 695 + 0 1 2.8E4 6.0E5 1.0E5 7.0E5 2 1.5E4 3.2E5 1.4E5 4.6E5 3 2.4E4 5.2E5 1.0E3 5.2E5 4 4.1E3 8.8E4 1.0E3 9.0E4 I 5 2.8E2 5.9E3 1.0E3 7.0E3 6 1.4E5 1.4E5 y 7 1.0E5 1.0E5 - B 8 ----- ----- 4.6E5 4.6E5
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TABLE 3 (Sheet 4) AUXILIARY BUILDING Radiation Data ! Normal Oper. Max. Dose DBA-LOCA1,2 40 Yr. Dose TID 3
- Elevation Zone Rate (R/Hr.) Dose (Rads) (Rads) (Rads) 81 1.6E2 3.4E3 1.0E3 4.4E3 750 + 0 (Cont'd) 1.0E3 82 4.7E0 1.0E1 1.0E3 83 4.7El 1.0E2 1.0E3 1.1E3 84 4.7E2 1.0E3 1.0E3 2.0E3 85 1.4E4 3.0E5 5.0E3 3.1E5 86 ----- -----
3.6E5 3.6E5 87 ----- ----- 1.7E8 1.7E8 2.1E4 3.9E5 8.4E4 4.8E5 e 88 5.0E3 5.0E3
,E 89 m 90 2.9E6 2.9E6 " 91 4.4E6 4.4E6 I 92 ----- -----
1.0E3 1.0E3 93 3.5E2 3.5E2 4.2E-6 2.3E-5 1.0E3 1.0E3 767 + 02 94 95 5.6E-3 1.3E0 3.5E2 3.5E2 96 1.8E-1 2.5El 3.5E2 3.8E2 97 3.8E3 1.6E6 3.5E2 1.6E6 98 8.03 3.1E3 3.5E2 3.5E3 l 99 1.1E3 3.0E5 3.5E2 3.0E5 100 3.3E5 2.0E8 ----- 2.053 NCIES: 10se Figure 5.0-1 to determine fraction of 1 year dose except for zones on Elevation 767+0. 2for zones on Elevation 767+0 use Figure 5.0-2 to determine fraction of 1 year dose. 3 TID = Total Integrated Dose. Equal to total of 40 year normal operating dose plus 1 year DBA-LOCA dose.
TABLE 3 (Sheet 5) AUXILIARY BUILDING Radiation Data Normal Oper. Max. Dose DBA-LOCAL,2 40 Yr. Dose TID 3 Elevation Zone Rate (R/lir.) Dose (Rads) (Rads) (Rads) 767 + 02 (Cont'd) 101 1.2E3 6.0E5 1.0E3 6.0E5 102 1.0E2 5.2E4 1.0E3 5.3E4 , 103 6.5E3 3.3E6 1.0E3 3.3E6 104 1.1E3 5.4E5 1.0E3 5.4E5 105 6.5E3 3.3E6 1.0E3 3.3E6 106 3.8E3 2.0E6 1.0E3 2.0E6 _4 107 5.2E2 2.7ES 1.0E3 2.7E5 g 108 6.2E-2 1.2E1 1.0E3 1.0E3
; 109 4.6E4 2.4E7 1.0E3 2.4E7 m, 110 8.5E5 4.7E-3 1.0E3 1.0E3 111 3.0E-4 5.5E-1 1.0E3 1.0E3 112 3.5E-3 4.9 3.5E2 3.5E2 113 2.5E-2 4.6 2.0E2 2.0E2 114 ----- -----
3.5E2 3.5E2 NOTES: 10se Figure 5.0-1 to determine fraction of 1 year dose except for zones on Elevation 767+0. 2for zones on Elevation 767+0 use Figure 5.0-2 to determine fraction of 1 year dose. 3 TID = Total Integrated Dose. Equal to total of 40 year normal operating dose plus 1 year DBA-LOCA dose.
TABLE 4 DIESEL GENERATOR DATA Radiation Area Normal Oper. Max. Dose DBA-LOCA1 40 Yr. Dose TID 2 Elevation Zone Rate (R/Hr.) Dose (Rads) (Rads) (Rads) 736 +6 1 4.5 100 3.5E2 4.5E2 2 4.5E2 1.0E4 3.5E2 1.0E4 3 1.8E3 4.0E4 3.5E2 4.0E4 4 2.8E3 6.0E4 3.5E2 6.0E4 5 1.8E3 4.0E4 3.5E2 4.0E4 6 4.5El 1.0E3 3.5E2 1.4E3 52 7 1.4 30 3.5E2 3.8E2 82 8 9.1E2 2.0E4 3.5E2 2.0E4 9 6.8E1 1.5E3 3.5E2 1.9E3 10 4.5El 1.0E3 3.5E2 1.4E3 11 4.5 100 3.5E2 4.5E2 12 4.5 100 3.5E2 4.5E2 13 2.3 500 3.5E2 8.5E2 14 1.4 30 3.5E2 3.8E2 NOTES: 10se Figure 5.0-1 to determine fraction of 1 year dose. 2 TID = Total Integrated Dose. Equal to total of 40 years normal operating dose plus 1 year DBA-LOCA dose.
FROM: NUREG 0588 ATTACHMENT 1 TABLE 5 McGUIRE NUCLEAR STATION - UNITS 1 AND 2 Page 16
SUMMARY
OF ENVIRONMENTAL QUALIFICATION OF CLASS 1E EQUIPMENT Rev. O LOCATED INSIDE CONTAINMENT EQUIPMENT ID: Valve Motor MANUFACTURER: Limitorque MODEL #: SM8 (1) Operators (Upper Containment) ACCIDENT ENVIRONMENT OPERABILITY OPERABILITY ACCURACY ACCURACY' ENVIRONMENT TO WHICH REQUIRED IN DEMONSTRATED REQUIRED DEMONSTRATED l (2) QUALIFIED ACCIDENT (Y OF SPAN) (% OF SPAN) ENVIRONMENT (3) w 5 r-180 F Temp: 340 F 5 min. 30 days N/A N/A (( Temp: Press: 14.8 psig Press: 105 psig post DBE (Notes 7 and 8) RH: 100% RH: 100% Rad: 4X108R Rad: 2X108R Chem Spray: Boric Chem Spray: Boric acid and sodium acid and sodium From the Manufacturer's File Location: tetraborate soln. hydroxide soln. Radiation Zone 23 3000 ppm Boron, Elevation 778+10 10.5 pH t
! QUALIFICATION REPORT (4): Limitorque Test Report: B0058, January 11, 1980 METHOD: Test EL40101T/17 - 4/13/82
FROM: HUREG 0588- , ATTACHMENT 2 TABLE 6 Page 2 McGUIRE NUCLEAR STATION - UNITS 1 AND 2 Rev. O
SUMMARY
OF ENVIRONMENTAL QUALIFICATION OF CLASS 1E EQUIPMENT LOCATED IN THE ANNULUS MODEL #: 1153GA9 MANUFAC1URER: Rosemount EQUIPMENT ID: Transmitter-RCS Pressure (WR) s OPERABILITY ACCURACY ACCURACY ACCIDENT ENVIRONMENT OPERABILITY DEMONSTRATED REQUIRED DEMONSTRATED ENVIRONMENT TO WHICH REQUIRED IN ACCIDENT (2) (% OF SPAN) (% OF SPAN) (1) QUALIFIED ENVIRONMENT 3 i 8% E5 Temp: 350*F 2 weeks 1 year i 10% Temp: 142*F Upper Range 100% post DBE post DBE
' RH: 100% RH:
Rad: 1.2X107R Rad: 4X107R From the Manufacturer's File Radiation Zone: 10 Elevation: 738+3 l QUALIFICATION REPORT: Test Report RMT Report #3788, Rev. A , METHOD: Similarity & Type Test l . l EL40101T/42 - 4/13/82
FROM NUREG 0588 ATTACHMENT 4 TABLE 7 McGUIRE NUCLEAR STATION - UNITS 1 AND 2 Page 4
SUMMARY
OF ENVIRONMENTAL QUALIFICATION OF CLASS 1E EQUIPMENT Rev. 0 LOCATED OUTSIDE CONTAINMENT AND EXPOSED TO THE POST-LOCA RECIRCULATION ENVIRONMENT ' EQUIPMENT ID: Containment Spray Pump Motors MANUFACTURER: Westinghouse, Buffalo MODEL #: 73F56019-1573, 73F56019-(1) 2573, 73F56019-3S73, 73F56019-4573 RECIRCULATION RADIATION RADIATION LEVEL TO WHICH ENVIRONMENT QUALIFIED (TID) (2) (TID) - N 5.2X105 RAD 2X108 RAD R v From the Manufacturer's File Radiation Zone: 3 Elevation: 695
- QUALIFICATION REPORT
- WCAP 8754 Rev. 1, WCAP 7829
; METHOD: Test and Analysis EL40101T/135 - 4/13/82
FROM NUREG 0588 ATTACHMENT 1 TABLE 8 McGUIRE NUCLEAR STATION - UNITS 1 AND 2 Page 2
SUMMARY
OF ENVIRONMENTAL QUALIFICATION OF CLASS 1E EQUIPMENT Rev.'O LOCATED INSIDE CONTAINMENT EQUIPMENT ID: Transmitter - Pressurizer MANUFACTURER: Barton MODEL #: 764 (Lot 2) (1) Level (Lower Containment) (Unit 1) ACCIDENT ENVIRONMENT OPERABILITY OPERABILITY ACCURACY ACCURACY ENVIRONMENT TO WHICH REQUIRED IN DEMONSTRATED REQUIRED DEMONSTRATED - (2) QUALIFIED ACCIDENT (% OF SPAN) (% OF SPAN) ENVIRONMENT (3) , t, . 4 Temp: 327 F Temp: 380*F 2 weeks 4 months i 25% Max. Error 15% o Press: 14.8 psig Press: 75 psig post DBE post DBE RH: 100% RH: 100% Rad: 1.4X10 7R Rad: SX107R Chem Spray: Boric Chem Spray: Boric acid and sodium acid and sodium tetraborate soln. oxide s . From Manufacturer's File: 8.5 pH Radiation Zone: 8a Elevation: 738+3 QUALIFICATION REPORT (4): WCAP 9885 METHOD: Test EL40101T/3 - 4/13/82
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