ML20138F715

Requests Staff Consideration of Relief Request M96-003 Under Above Subparagraph in Lieu of 10CFR50.55a(a)(3)(ii) in Original Request.W/Request for Alternative Exam as Described in 10CFR50.55a(g)(6)(ii)(A)(5)

ML20138F715
Person / Time
Site:	Mcguire
Issue date:	04/29/1997
From:	Barron H DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M95773, NUDOCS 9705060034
Download: ML20138F715 (19)

Text

..

11 1

Duke h>uvr Onnpany H H Baknn kl. Guire Nm lear Generatton Department ike hesiden:

• ' 12k)Haken Ferrylhad(AIGI)nP)

(104)8754SM Huntentitic NC2607M140 (L4)8,~54309 Far

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DUKEPOWER April 29, 1997 Document Control Desk U.S. Nuclear Regulatory Commission 4

Washington, DC 20555 i

Subject McGuire Nuclear Station, Unit 2 Docket No 50-370 Relief Request,96-003 Request for Additional Information i

(TAC No. M95773)

Dear Sir:

L A teleconference on February 25, 1997, among Office of Nuclear Reactor Regulation, Idaho National Engineering l

Laboratory, and Duke Power personnel determined that the criteria of 10CFR50.55a (g) (5) (iii) more accurately

?

l support the basis of Relief Request M96-003.

Therefore, Duke Power requests Staff consideration of Relief Request M96-003 under the above subparagraph in lieu of 10CFR50.55a (a) (3 ) (ii) in the original request.

As confirmed in the teleconference, the information i

submitted with the original request for relief is sufficient to support consideration under 10CFR50,55a (g) (5) (iii).

l As referenced in the docketed correspondence of February l

25, 1997, included is a request for Alternative i

l examination as described 10CFR 50.55a (g) (6) (ii) (A) (5).

l As' requested in the January 9,

1997, Request for Additional Information, a copy of this information will be i

forwarded to Michael T. Anderson at the address supplied.

J d

9705060034 970429 jI b

PDR ADOCK 05000370 I

P PDR 1

]

a on___ _~_

H. H. I. I. l.li.Il.l I.

1 U.S. Nuclear Regulatory Commission

.. April 29, 1997 page 2 Should there be any questions regarding this matter, please contact John M. Washam at (704) 875-4181.

i l

Very truly yours, l

l 1

J?

d H.B.

Barron I

xc.

Mr.

L.A. Reyes l

Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. Victor Nerses, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission l

One White Flint North, Mail Stop 9H3 Washington, DC 20553 1

I l

Mr. Scott M.

Shaeffer, Senior NRC Resident Inspector, McGuire McGuire Nuclear Station I

Michael T.

Anderson INEL Research Center 2151 North Boulevard P.O.

Box 1625 l

Idaho Falls, Idaho 83415-2209 I

l l

l

Duke Power Company Station: McGuire Nuclear Station, Unit 2 10 Year Interval Request For Alternative No.97-001 Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), Duke Power has determined that it is unable to conform with the examination requirements of 10 CFR 50.55a(g)(6)(ii)(A) for McGuire Nuclear Station, Unit 2. Accordingly, information is being submitted in support of this determination, and a request for alternative pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) is being sought from the requirements of 10 CFR 50.55a(g)(6)(ii)(A).

Background:

In response to NRC Information Notice 96-32, " Implementation of 10 CFR 50.55a(g)(6)(ii)(A) " Augmented Examination of Reactor Vessel", Duke Power has reviewed the information contained in this notice for applicability to its facilities and has taken action to avoid or mitigate the effects of limited examinations. These actions are taken to eliminate and/or reduce the concerns expressed in this Information Notice.

Because of concerns regarding the scope of examination of reactor vessels, the NRC issued, in 1992,10 CFR 50.55a(g)(6)(ii)(A), " Augmented Examination of Reactor Vessel", which contains requirements for an augmented examination of reactor vessels.

The rule requires the licensee to implement, before the time required by normal updating of the inservice inspection (ISI) program, provisions in the 1989 Edition of the ASME, Boiler and Pressure Vessel Code,Section XI, to examine " essentially 100%" of the length of all reactor vessel shell welds. " Essentially 100%" examination as used in 10 CFR 50.55a(g)(6)(ii)(A)(2) means more than 90% coverage of the examination volume of each weld, where the reduction in coverage is due to interference by another component, or pan geometry."

In many cases, licensees have determined that the overall average examination coverage for reactor vessel shell welds may be more than 90%. However, the corresponding examination coverage for individual welds may be substantially less than 90%. In these cases, licensees are unable to completely satisfy the requirements for the augmented reactor vessel examination. Therefore, they must propose an alte: native that would provide an acceptable level of quality and safety.

The licensee must expend all efforts using the latest methods and techniques to achieve acceptably adequate examinations during weld inspections. When examinations cannot be completed with 100% coverage, then a request for alternatives from the Code of Federal Regulations must be submitted.

s l.

System / Components For Which Alternative is Requested:

In response to the requirements of 10 CFR 50.55a(g)(6)(ii)(A) relief is requested for two McGuire Unit 2 reactor pressure vessel shell welds specified in Examination Category B-A Item Bl.10.

These welds, listed below, were examined and found to have 90% coverage or less. Attachment 1 provides j

examination data for these welds which were examined during Refueling Outage 8, and for which relief is requested from the requirement for augmented i

examination of the reactor vessel, a.

Reactor Vessel Shell Welds Weld Numbers Item Numbers 2RPV-WO3 B01.011.003 2RPV-WO6 B01.011.004 IL 10CFR50 Requirement:

10 CFR 50.55a(g)(6)(ii)(A) requires that all licensees shall augment their reactor vessel examination requirements at least once for the reactor vessel shell welds specified in Item Bl.10 of Examination Category B-A" Pressure Retaining Welds in Reactor Vessel in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code. 10 CFR 50.55a(g)(6)(ii)(A) further states that "for the purpose of this augmented examination essentially 100% as used in Table IWB-2500-1 means more than 90

% of the examination volume of each weld, where the reduction in coverage is due to interference by another component, or part geometry." The augmented examination, when not deferred in accordance with the provisions of 50.55a(g)(6)(ii)(A)(3), shall be performed in accordance with the related i

procedures specified in the Section XI Edition and Addenda applicable to the inservice inspection interval in effect on September 8,1992, and may be used as a 4

substitute for the reactor vessel shell weld examination scheduled for implementation during the inservice inspection interval in effect on September 8, 1992. 10 CFR 50.55a further states that " Licensees that make a determination i

that they are unable to completely satisfy the requirements for the augmented reactor vessel shell weld examination specified in 10 CFR 50.55a(g)(6)(ii)(A) shall submit information to the Commission to support the determination and shall propose an alternative to the examination requirements that would provide an acceptable level of quality and safety.

I 2

1 111.

. Requirement for which Alternative is Requested:

i An alternative is being requested to the requirement of obtaining " essentially 100%" examination volume of weld coverage for McGuire Unit 2 reactor vessel Bl.10 welds listed in section I and shown in Attachment 1.

l IV.

Basis for requesting Alternative:

)

l If licensees make a determination that they are unable to completely satisfy the

.)

requirements for the augmented reactor vessel shell weld examination specified j

in 10 CFR 50.55a(g)(6)(ii)(A); then 10 CFR 50.55a(g)(6)(ii)(A)(5) requires the licensee to submit information to the Commission to support this determination and to propose an alternative to the examination requirements that would provide an' acceptable level of quality and safety. The licensee may use the proposed l

alternative when authorized by the Director of the Office of NRR.

! provides the calculations documenting the actual amount of Code l

required examination coverage obtained. A combination of multiple angles and ultrasonic techniques was used to obtain the maximum coverage possible. The use of an alternate transducer head provided increased coverage through optimum j

transducer arrangement for scanning close to obstructions. However, during the ultrasonic examination of the welds referenced below and listed in Attachment 1 of this alternative, the greater than 90% coverage required per 10 CFR l

50.55a(g)(6)(ii)(A)(2) could not be obtained due to geometry and actual physical barriers.

Reactor Vessel Lower Shell to Lower Head Weld ( 2RPV-WO3) (item Number B01.011.003):

I This examination was limited to 43.60% aggregate coverage of the required weld volume. The principal limitation for this weld is six core guide lugs welded to the vessel ID just above the weld on the lower shell section, whose presence restricts the scanning surface in that area and limits the examination coverage.

Reactor Vessel Uooer Shell to Nozzle Belt Weld ( 2RPV-WO6) (Item Number B01.011.004h This examination was limited to 48.20% aggregate coverage of the required weld volume. The principal limitation for this weld is the presence of a taper at the ID surface starting at the upper edge of the weld and extending up from the nozzle belt section. The taper causes the scanning fixture to lift-off the vessel sur ace, r

thus disrupting the sound beams which in turn reduces the examination coverage.

The reactor vessel shell welds were examined from the vessel inside surface using automated ultrasonic examination equipment. The examinations were done with 3

4 i

various contact head arrangements to optimize the maximum examination coverage. This allowed each transducer to scan as close as possible to any

[

obstruction around the area examined. Although the coverage requirements of 10 CFR 50.55a(g)(6)(ii)(A) could not be met, the examinations were performed with j

modified equipment and tooling designed to accomplish the maximum coverage i

j possible.

As a result of inspections performed, the 100% requirement would be impractical l

for McGuire Nuclear Station. The reactor vessel welds were examined to the maximum extent practical to the requirements of Section V, Article 4 of the 1980 j

Edition through the Winter 1980 Addenda of the ASME Boiler and Pressure

{

Vessel Code and the additional requirements of Regulatory Guide 1.150. To meet the 10 CFR 50.55a(g)(6)(ii)(A)(2) examination coverage ; requirements, design j

modifications would be necessary to gain access to the welds in order to obtain j

complete coverage. The design modifications are impractical due to the vast j

scope of work that would be required. Imposition of this requirement would i

cause a considerable burden on Duke Power with no commensurate safety benefit realized.

i I

V.

Alternate Examinations:

In addition to the volumetric examination that has been performed on the

)

McGuire reactor vessel, Duke Power has performed a visual examination of the internals and the inside of the reactor vessel a required by ASME Section XI, Table IWB-2500-1. This visual examination did not identify any rejectable conditions per ASME Section XI acceptance standards.

The use of radiography as an alternate volumetric examination method is not feasible due to component thickness and restrictions from physical barriers which l

prohibit access for the placement of source, image quality indicators, film, etc. In addition, the background radiation levels would not allow for a radiographic examination to render meaningful results.

Performing the ultrasonic examination from the outside of the reactor vessel is not a viable option. The design of McGuire's reactor building prohibits access for the i

equipment and personnel from outside the vessel.

Duke Power Company will continue to perform ultrasonic examinations of all i

reactor vessel welds to the maximum extent practical in accordance with the requirements of ASME Section V,~ Article 4,1989 Edition and Regulatory Guide i

1.150, Revisior 1. Appendix A. The application of Code Case N-460 will be i

utilized in all cases where less than 100% but greater than 90% weld coverage is obtained. In cases where weld coverage of less than 90% coverage is obtained, a request for relief from ASME Section XI Code requirements will be submitted.

Duke Power Co. proposes as an alternative to the greater than 90% coverage requirement of 10 CFR 50.55a(g)(6)(ii)(A), that the examination coverage 4

.a obtained on the welds listed in Attachment I be considered to provide an acceptable level of quality and safety.

No additional examinations will be required.

VI.

Justification for Granting Alternative:

10 CFR 50.55a(A)(5) states that " Licensees that make a determination that they j

are unable to completely satisfy the requirements for the augmented reactor vessel shell weld examination specified in 50.55a(g)(6)(ii)(A) shall submit information to the Commission to support the determination and shall propose an alternative to the examination requirements that would provide an acceptable level of quality and safety. 10 CFR 50.55a(a)(3) states that alternatives to the requireraents of paragraph (g) may be used when authorized by the Director of the Office of Nuclear Reactor Regulation. The proposed alternative (s) must demonstrate that j

an acceptable level of quality and safety, or compliance with the specified requirements would result in hardship or unusual difficulty without a

compensating increase in the level of quality and safety.

Examination of 100% of reactor pressure vessel shell is impractical. Examination of the accessible weld volume provides sufficient and reasonable assurance of vessel integrity. The reduction in the expected examination coverage will not endanger life or property or the common defense and security because the reactor coolant system is designed and constructed to have a low probability of gross rupture or significant leakage throughout its design life. Technical Specificationt 3/4.4.6 for McGuire Nuclear Station places conservative limits on the amount of reactor coolant leakage allowed during system operation. Any weld failure would allow additional coolant to leak from the system. The reactor coolant system leakage detection system is in place to detect any variation in the system water inventory. The purpose of the containment building is to retain any such leakage within its boundaries.

If leakage exceeds Technical Specification 3.4.6.2,

procedures are in place to assure safe shutdown of the unit within specified time limits.

Due to the design of the McGuire reactor vessels and location of the physical obstructions, it is impractical to obtain the examination coverage required by 10 CFR 50.55a(g)(6)(ii)(A)(2) without placing undue hardship on Duke Power.

Based on the portions of the required volumetric and visual examination that have been completed, any existing pattern of degradation would have been detected.

Duke Power Company will continue to ultrasonically examine the reactor vessel Bl.10 category welds to the extent practical within the limits of original design and construction. This will provide reasonable assurance of weld / component integrity.

5 J.1 m

.. -.. -. - = - _ -

-.~

Pursuant to 10 CFR 50.55a(g)(6)(i), granting this alternative for the reactor vessel Bl.10 category welds is authorized by law, will provide reasonable assurance of weld / component integrity, and will not endanger life or property or the common defense and security and is other wise in the public interest giving due i

consideration to the burden upon the licensee that could result if the requirement were imposed on the facility.

Vll.

Implementation Schedule:

The reactor vessel welds listed above will be examined during the third period of the second interval. The examinations are c:2rrently scheduled during refueling outage 7.

I Attachments:

1. Listing ofinformation for welds with limited UT examination coverage

2. Detailed drawings of affected welds including calculation methods l

l l

l l

Evaluated By:v Date: 4f[6h7 6 blonn3e.cA i

Reviewed By:

Date: 4 lbN.1 V

i 1

6 1

ASME Class 1 NDE Inservice Inspection Request For Alternative Serial No.97-001 For McGuire Unit 2 Based On ASME XI-1980 Code Through Winter 1980 Addenda Page1of1 Item No.

Exam Category /

System Or Function A en To Be Reason For Request Weld ID No.

lxensee Proposed Figure No.

Component 11amined Al'emate Examination B01.011.003 B-A Reactor Vessel Houses the fuel assembhes, control uwerSimil m Unuted scan due m gmmetric 2RPV-WO3 None IWB-2500-1

'"d5 ""d '*3 i"'***I8' *IS

' **' H**d

• "6 "'"

8 directs the flow of reactor coolant Wtld Actual coverage obtained = 43.60%

B01.011.004 B-A Reactor Vessel Houses the fuel assemblies, control Upper Shell to Unuted scan due to geometric 2RPV-WO6 None gyg,3g g rods, and vessel internals, also Nozzh Belt configuranon.

directs the flow of reactor coolant Weld Actual coverage obtained = 48.20%

J f

A-MCQUIRE UNIT 2 REACTOR VESSEL i

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4 ves SER. NO.97-001 ATTACHMENT 2 REACTOR VESSEL OUTLINE PAGE 1 OF 10 4

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SER. NO.97-001 ATTACHMENT 2 REACTOR VESSEL OUTLINE PAGE 2 OF 10

- ~ - -

4 325.53

.157 NOM f0 FLANGE SURFACE 9 173.8 CLAD 0.465 (TI)

• Z" s 339.28 1

J l

80ff0M HEAD RA01U5 CENTER "Z* = 342.39 I

18.*

45*

I d.

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j "2"

s 349.12 I

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TO FLANGE SURFACE I

4 "Z"

s 352.22 i

• Z' s 353.58 TI/2

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• Z*

• 356.68

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WELD REFERENCE LINE ALPHA 11.09

., D

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• ALPHA 13.17

/2 ALPHA 13.95 ALPHA

• 15.18

=.

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• 1,*.,.

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RADIUS s 88.194 ALPHA : 67.39 7"

7 e:,,k,.. '

(GUIDE LUG LIMIT! ALPHA s 19.4s ALPHA 2g.13

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' 60**

ALPHA s 22.14 ALPHA 24.18 j

.157 NOM 5.591 tT2) l 4

8 M ES' SER. NO.97-001 9 ALL

• 2" O! ENS 10NS REFEMNCE Tm DISTAPCE DETWEN av i.oHlno suRrAce no rm soon 14 mr rm cEntta or vm c ACT mao is arrsEr vEnricAu.y rnos rm ecos ATTACHMENT 2

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ustee im 4-seca arrsEr.

B01.011.003

2. seAcree eErwEn twe inAusoucEns is s. ie* IN AU.

01 RECT 10NS (2.96' A80VE TRANS. 2.81

• BELOW TRANSl.

2RPV-W03

s. MEAsumens ARE in seats Ano oECism. oranErs.

PAGE 3 OF 10 4 UT SCAN PATHS DES 10NATED BY-

. -...... L 1mS.

5.

• X" 15 EXTDOED 2.S* FOR GCANS BETWIN LUGS. Ato 7.5" POR $ CANS DELCM LUGS M

i I4WER SHELL TD IDWER HEAD WELD ITEM NO.: B01.011.003 I.D. NO.: 2RPV-WO3 l

Total Exam Area = 53.32 in (Near Surface + Weld + T/2) s Near Surface Area =

-8.37 ina (Cross-Section)

Weld Area =

6.37 in2 (Cross-section) l T/2 Area = 38.58 in2 (Cross-Section)

BETWEEN LUGS CIRC 70'

lets 6.24 in8 of Near Surface Area (74.6 %)

O' Gets 9.07 in2 of Total Exam Area (17.0 %)

45' & 60' Get 31.79 in' of T/2 Area,.44 in* of Weld Area Total Coverage =

31.79 + 0.44 + 0.44 63.7 %

=

38.58 + 6.37 + 6.37 AXIAL 70' Gets 8.24 in2 of Near Surface Areas however due to the Full-Node Exam, 100 % Coverage of the Near Surface Area is obtained by the 45' & 60'.

45' Gets 100 % of Weld and T/2 Area 60' Gets 100 % of Weld and T/2 Area BELOW LUGS AXIAL Due to the Full-Node Exam, 7.58 in2 of the Near Surface Area is obtained by the 60* (90.6 %).

45' Gets 3.79 in of T/2 Area (0 % Weld Area)

Total Coverage =

_3.79 + 0.00 + 0.00

=

7.4 %

38.58 + 6.37 + 6.37 60' Gets 9.11. int of T/2 Area and 1.23 in* of Weld Area Total Coverage =

9.11 + 1.23 + 0.00

=

20.1 %

SER. NO.97-001 38.58 + 6.37 + 6.37 ATTACHMENT 2 B01.011.003 2RPV-WO3 PAGE 4 OF 10

J LOWER SNRTL TO IDWER HEAD WELD CONTINUED V

There are six (6) segments between lugs, each 31.60' covered by th,e center of the head (O' and 60' Circ).

The outaide transducers each cover un additional 2.06' which results in 35.72* covered by 7 0*, 60' & 45' Axial and 70* & 45' Circ.

There are alto six (6) j segments below lugs, each covering the remaining 24.28' for Axial Scans.

l 4

i O' & 60' Circ Coverage 189.60 x % Between

=

l 100 x 360 i

j 70' & 45' Circ Coverage 214.32 x % Between

=

100 x 360 1

Axial Coverage 214.32 x % Between + 145.68 x % Below

=

100 x 360 1

i i

AXIAL CIRC d

NS

_fdL.

45 70 A

_f.5_

0 96.2 67.7 62.5 44.4 33.5 37.9 9.0 Aggregate Coverage

=

[ 96.2 x 8.37 + (67.7 + 62.5) x (38.58 + 6.37 + 6.37)

+ 44.4 x 8.37 + (33.5 + 37.9) x (38.58 + 6.37 + 6.37)

+

9.0 x 53.32 )/

[ 8.37 x. 2 + (38.58 + 6.37 + 6.37) x 4 + 53.32 ]w SER. NO.97-001 43.6 %

ATTACHMENT 2 Aggregate coverage

=

B01.011.003 2RPV-WO3 PAGE 5 0F 10 j

J

-..... ~. - -.. -.

r

k..

h 60 DEGREE 45 DEGREE AXIAL COVERA 6E i

AXtAL COVERAGE BELOW LU6S I

BELCv LUGS z e ss9.re 1

• 2* a 342.30 - -

CIRC C0VERAGE 70 DEGREE

- N

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(ALL ANGLES)

AXIAL COVERAGE

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ALPItA e 17.39%e t

[ MIN L M Ll#1tl ALPsta e 89.49%

,8 AL.44 e 20.88 %*

s ALPIIA e 22.84 % #,

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TAPER BELOW OUTLET NOZZLES l

l 12s.42 "Z"

= 125.60 ANGE SLHFACE I O. 82 7-

"Z" s 128.70 (TI)

"Z" = 136.76 "Z" = 130.76 g

~ ~ ~ ~ ~ ~ " " "

l T1/2 f

f

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. _ _ _ _ _ /.c.

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"Z"

= 159.00 l

. O 173.0 CLAo

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l 8.465 l

(T2)

NOTES:

1.

ALL "Z" DIMENSIONS REFERENCE THE DISTANCE DETWEE" TW MATING SURFACE APO THE DOOM. TE CENTER OF TtE CONTACT HEAD IS OFFSET VERTICALLY FROM THE BOOM SY I4.94" USING THE 4-INCH OFFSET.

2. SPACING BETwEEN T}E TRANSOUCERS IS 3.10" IN ALL DIRECTIONS.

3. HEASUREMENTS ARE IN INCHES ANO DECIMAL DEGREES.

4 UT SCAN PATHS DESIGNATED BY............ lites.

5. ALL SCANS USE STANDARO HEAD EXCEPT FOR B6.1.2sA-1 & B-l APO B6.2.2sA-2 & B-2 WHICH USE ALTERNATE HEAD.

SER. NO.97-001

~.

ATTACHMENT 2 DO1.011.004 2RPV.WO6 PAGE 7 OF 10 i

i i

UPPER SHELL TO NOZZLE BELT WELD ITEM NO.: B01.011.004 I.D. NO.: 2RPV-WO6-

)

Total Exam Area = 93.65 in' (Near Surface + Weld + T/2)

Near Surface Area = 10.47 in' (Cross-Section)

Weld Area = 11.38 in' (Cross-Section)

T/2 Area = 71.80 in (Cross-Section)'

2 CIRC 70' Gets 3.'82 in* of Near Surface Area (36.5 %)

~

O' Gets 6.08 in2 of Total Exam Area (6.5 %)

45' & 60' Get 28.50' in' of 'T/2 Area (0 % Weld)

Total Coverage =

28.50 +

0.00 +

0.00

=

30.1 %

71.80 + 11.38 + 11.38 AXIAL 70' Gets 5.40 in2 of Near Surface Area (51.6 %)

45' Gets 63.34 in' of T/2 Area f

4 5'-UP Gets 11.10 in8 of Weld Area 4 5'-DOWN Gets 4.74 in8 of Weld Area Total Coverage =

63.34 + 11.10 +

4.74

=

83.7 %

71.80 + 11.38 + 11.38 60' Gets 67.39 in8 of T/2 Area 60*-UP Gets 9.87 in' of Weld Area i

60'-DOWN Gets 9.34 in8 of Weld Area J

Total Coverage =

67.39 +

9.87*+

9.34 91.6 %

=

71.80 + 11.38 + 11.38 SER. NO.97-001 ATTACHMENT 2 B01.011.004 2RPV WO6

~

PAGE 8 0F 10

l UPPER SHELL TO NOZZLE BELT WELD CONTINUED AXIAL CIRC

_' L 60

_ 4 5...

70 60

_45.

0 51.6 91.6 83.7 36.5 30.1 30.1 6.5 Aggregate coverage

=

[ 51.6 x 10.47 + (91.6 + 83.7) x (71.80 + 11.38 + 11.38)

+ 36.5 x 10.47 + (30.1 + 30.1) x (71.80 + 11.38 + 11.38)

+

6.5 x 93.65 ]./

[ 10.47 x 2 + (71.80 + 11.38 + 11.38) x 4 + 94.65 ] =

Aggregate coverage 48.2 %

=

l i

l SER. NO.97-001 ATTACHMENT 2 B01.011.004 2RPV-WO6 PAGE 9 OF 10

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SER. NO.97-001 ATTACilMENT 2 B01.011.004 2RPV-WO6 PAGE 10 OF 10