ML20085H770

Details Info Utils Must Provide for NRC Safety Evaluation on Intergranular Stress Corrosion Cracking Pipe Cracks.Info Due 3 Wks Prior to SER Due Date

ML20085H770
Person / Time
Site:	Peach Bottom, Browns Ferry, 05000000
Issue date:	09/16/1983
From:	Johnston W Office of Nuclear Reactor Regulation
To:	Lainas G Office of Nuclear Reactor Regulation
Shared Package
ML20083L924	List: ML20085H770 ML20085L392 ML20086F743 ML20087N832 ML20087N874
References
FOIA-84-15 NUDOCS 8309260442
Download: ML20085H770 (3)

Text

. - - _ _ - - -

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UNITED STATES 8

' ~,j NUCLEAR REGULATORY COMMISSION s

WASHINGTON, D. C. 20555

...../;E 3

SEP 16 1983 g

k Docket Nos: 50-277/259 MEMORANDUM FOR:

Gus C. Lainas, Assistant Director for Operating Reactors Division of Licensing FROM:

William V. Johnston, Assistant Director Mater.als, Chemical & Environmental Technology Division of Engineerino d

SUBJECT:

SAFETY EVALUATION OF IGSCC PIPE CRACKS IN PEACH BOTTOM UNIT 2 AND BROWNS FERRY UNIT 1 Based on the discussions with the respective project managers of Peach Bottom Unit 2 and Browns Ferry Unit 1, we learned that the Peach Bottom Unit 2 and Browns Ferry Unit 1 plan to restart during the first week and the third week of Octcber 1983, respectively.

So far, we have not received any submittals from Peach Bottom Unit 2 regarding the ongoing inspection and repairing activities.

For Browns Ferry Unit 1, we received some preliminary information from the licensee. Based on the preliminary information, we have already informed the licensee

..~of Browns. Fer.ry Mnit_1.regarding our. concerns on some of the; unrepaired c

large size defective welds and those welds repaired with " mini" overlay.

In order to properly perform the safety evaluation of those two plants, we need to receive the docketed submittals from the licensees at least 3 weeks prior to the SER due date.

In view of the NRC order recently issued to the 5 plants for early inspection requiring 100% UT examina-tion of nonconforming Class 1 austenitic stainless steel welds in Recirculation, RHR, core spray, and RWCU piping' systems, it is con :

.sidered prudent to include those 4 piping systems in the safety evaluation of Peach Bottom Unit 2 and Browns Ferry Unit 1.

Therefore, the licensees' submittals should provide the following information as a minimum:

1.

Provide all the information requested by Bulletin 83-02.

2.

Address uncertainties in the UT sizing, especially the depth measurements and provide an error band with each reported measurement if possible.

s Contacts:

W. Koo and W. Hazelton X-28589 X-28075

^

XA Copy Hos Been Sent to PDR b O b /p d N A [

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Gus C. Lainas SEP I 6 1983 3.

Provide tables to identify all the Class 1 austenitic stainless steel welds in Recirculation, RHR, core spray, and RWCU piping systems. The tables should provide the following information:

(a) Pipe size and weld identification number.

(b) Description of each weld joint such as " cast stainless steel valve to stainless steel pipe."

(c) Description of piping and weld material, especially carbon content.

(d) Stress rule index of each weld joint.

(e) Specify whether each is a shop weld or field weld, and the vendor performing tha fabrication.

(f) Detailed information on solution heat treatment performed on shop weld after welding, including annealing temperature, holding time, and the cooling rate.

(g) Identify which. are conforming welds and nonconforming welds, and the

.m._.-. reasons-for classifying the weld as confohning.

(h) Results of UT examination including the depth, length, and location of all indications (length description like 360*

intermittent is not adequate).

I (i) Indicate if both sides of welds were 100% UT inspected, especially the immediate areas under the crown.

(j) Provide justification for not performing UT examinations on all nonconforming welds.

(k) Provide isometric piping diagrams and identify the conforming and nonconforming welds by different symbols including which weld was UT examined or overlay repaired.

4.

When nonconforming welds in those 4 piping systems are not UT examined, the licensee should provide justification and provide information on the following items:,

(a) Can the pipe line~containing the subject nonconforming weld be isolated?

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Gus C. Lainas SEP 16 1983 (b) The temperature experienced by the subject weld during normal plant operation.

(c) What is the flow rate in the pipe line containing the subject nonconforming weld during normal plant operation?

5.

Identify plans to replace the nonconforming pipe lines.

6.

Identify any compensatory measures such as additional leak detection devices, augmented leakage limits and monitoring, and others to ensure safe operation of the plant.

7.

Provide details in fracture mechanics evaluation of defective welds, especially the crack growth calculations, including additional stresses brought about by adjacent overlays, to allow the staff to confirm the licensees' calculation.

8.

Provide the details of weld overlay design evaluation.

9.

Provide detailed evaluations of the shrinkage stress due to weld overlay at each affected weld and its effect on IGSCC initiation and growth.

Please inform us as soon as possible when the above information pertaining to Peach Bottom Unit 2 and Browns Ferry Unit 1 will become available.

k William V. Joh ston, Assistant Director Materials, Chemical & Environmental Technology Division of Engineering cc:

R. Vollmer D. Eisenhut E. Sullivan D. Vassallo S. Pawli_cki -

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K ecker C. Cheng R. J. Clark G. Gears J. Stolz W. Hazelton W. Koo~

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UNITED STATES

[ p, s,33' j NUCLEAR REGULATORY CO.'.1MISS10N

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August 10, 1983 Docket No'. 50-296 4

MEMORANDUM FOR: Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing FROM:

Dick Clark, Project Manager 4

Operating Reactors Branch #2 Division of Licensing

SUMMARY

OF AUGUST 9,1987 MEETING WITH TENNESSEE VALLEY AUTHORITY (TVA) REGARDING INSPECTION OF LARGE DIAMETER BWD PIPE WELDS -

Re:

Browns Ferry Unit No. 3 A meeting was held on August 9,1983 hetween representatives of the NRC and TVA to discuss the proposed acceleration of inspections of welds in larga 4

diameter piping at Browns Ferry Unit 3 in view of the cracking recently identified at several BWR facilities. Enclosure 1 is a list of meeting attendees. Enclosure 2 is a copy of the material presented by the licensee.

Darrell Eisenhut (NRC) opened the meeting by stating that this was an information gathering meeting and that the NRC position wo~uld not be established until after available information had been received and reviewed.

4 Browns Ferry Unit 3 (BF-3) is scheduled to shutdown on November 11, 1983 for refueling and various NRC required modifications (e.g., completion of the Park I torus modifications). During the outage, the recirculation and t

RHR system piping will be inspected in accordance with IE Bulletin 83-02.

Similar inspections have been performed on Units 1 and 2

~

TVA presented its justification for continued operation of BF-3 for.the next 3 months until the scheduled refueling outage in November, based pri-marily on the following:

o-Leak-Before-Break is a valid criteria;

'o Circumferential flaw instability'is impossible; o Crack propagation rates in large diameter piping is very slow once a crack reaches approximately 30% throughwall, since the metal is in-compression; 1

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, o There has not been any crack detected to date in any BWR that has come close to compromising the structural strength of the pipe; o A crack would have to be 360* around the pipe and 63% throughwall in the entire length of the crack before infringing on the ASME Code safety margine; the asymetry of weld sensitization and bending loads tend to preclude a uniform depth circumferential crack from developing; o 11 welds in the BF-3 recirculation /RHR systems were inspected in November 1981; no cracking was observed.

TVA has implemented for all three operating units compensatory measures such as reducing the shutdown limit on unidentified drywell leaking, reducing the surveillance interval to once per shift, etc.

TVA evaluated the impacts if BF-3 were required to shutdown within 30 or 60 days. TVA stated that they would probably not be able to start any inspec-tions or modification work on Unit 3, even if it were down, until early November 1983 for the following reasons:

)

o Browns Ferry Unit 1 has been down since April 15, 1983 The inspection-and repair of welds has extended the cutage by 25 days. Frojected startup is now almost November 1, 1983.

o Sequoyah Unit 2 is currently down for refueling and inspections, o TVA does not have the support personnel (health physics, security, engineering, craft workers to erect scaffolding and remove insulation, etc.) to support simultaneous outages in two Browns Ferry units. All of the qualified UT examiners are tied up with the inspections on BF-1 and Sequoyah 2.

o All of the recent and upcoming outages involve modifications (e.g.,

torus mods) that require offloading the entire core. There is only one fuel handling crew and one crane for the three Browns Ferry units.

o TVA cannot effectively train, supervise, and manage the approximately 1700 outside craft personnel needed to conduct simultaneous outages in two units in the same building, o Specialized tools and equipment are in short supply, o The machine and prefabrication shops, layout areas, and other support facilities.are tied-u.p with.the BF-1 outage.

At the request of the staff. TVA will submit within the next several days an evaluation of the feasibility of (1) reducing.the allowable out-of-service time for drywell leakage detection systems and ECCS systems, (2) conducting I

f-

_________m____m

I

. visual inspections of the piping between now and the scheduled November 11, 1983 outage, and (3) the desirability of providing special operator training to increase awareness of and prompt response if there is an indication of leakage.

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5./

j.7-j aid J., Clark, Project Manager Operating fleactors Branch #2 Division of Licensing

Enclosures:

As stated cc w/ enclosures:

See next page t

l.

E

I

't Mr. Hugh G. Parris Tennessee Valley Authority Browns Ferry Nuclear Plant, Units I, 2 & 3 i

cc:

H. S. Sanger, Jrl Esquire U. S. Environmental Protecti:n General Counsel Agency Tennessee Valley Authority Region IV Office Regional Radiation Representative 400 Commerce Avenue E 118 33C 345 Courtland Street Enoxville, Tennessee 37902 Atlanta, Georgia 30308 Resident Inspector Mr. Ron Rogers Tennessee Valley Authority U. S. Nuclear Regulatory Ccmmission 400 Chestnut Street, Tcwer II Route 2, Box.311 Cnattanooga, Tennessee 37401 Athens, Alabama 35611 Mr. Charles R. Christopher Mr. Donald L. Williams, Jr.

Chairman, Limestone County Commissicn Tennessee Valley Authority P. O. Box 188 400 West Summit Hill Dr., W10885 Athens, Alabama 35511 Knoxv111st, Tennessee 37902 Ira L. Myers, M.D.

George Jones State Health Officer Tennessee Valley Authority State Department of Public health State Office Buildine P. O. Box 2000 Mont;crery, Alabama 36130 Decatur, Alabama 35502 Mr. Oliver' Havens-Mr. H. N. Culver 2?9A FED U.S. Nuclear. Reculatory Ccrmission 430 C:::erce Avenue Reactor Training Center Tennessee Valley Authority Osborne Office Center, Suite 200 Kn xville, Tennessee 37902 Chattanooga, Tennessee 37411 James P. O'Reilly Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 S

3 A

1 1

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ENCLOSURE 1 ATTENDANCE LIST MEETING WITH TVA August 9,1983 Name Orcanization Name Organization Dick Clark NRR/0L Jim Green TVA Larry Mills TVA Tom Ziegler TVA Dick Parker TVA Joe. D. Ferguson TVA J. D. Wolcott TVA Gary Pitzl TVA Jim Domer TVA T. Tesch ANI Michael P. Ferrante Am NJclear Ins.

Joseph C. Danka EPRI Jim Charnley NUTECri Paul Goldberg NRC/0PE Wm. J. Collins NRC/IE Vince Derr NUTECH l

J. Dietrich CP&L Ambjorn Lindsuog RKS T. Alexion NRC/DL D. L. Radius CBI S. Ranganath GE R. Gamble Impell John A. Zwolinski NRR/DL-J. E. McEwen, Jr T&I John W. Pendlebury GE Victor Benaroya NRR/ CME 3 J. B. Henderson NRC/IE Robert Baer NRC/IE J. J. Blake NRC/R-I James L. Coley NRC/R-II F. S. Cantrell NRC/R-II D. B. Bassallo NRR/DL B. D. Liaw NRC/DL G. C. Lainas NRC/DL R. H. Vollmer NRC/DE J. A. Olshinski NRC/R-II D. Eisenhut NRC/DL E. Case NRC/NRR

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1 ENCLOSURE 2

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AUGUST 9. 1933 1.

JUSTIFICATION FOR CONTINUED OPERATION.

A.

LEAx-BEFORE-BREAx S.

CRACK-GR0wTH Y.00ELS VAllo C.

STRUCTURAL INTEGRITY D.

EXTENSIVE CRACx!NG EXPECTED E.

LESS OPERATING IIME 11.

SUMMARY

OF FREVIOUS INSPECTIONS 111.

DESCRIPTION OF PRIMARY SYSTEM LEAx DETECTION IV.

IMPACT OF UNIT 1 INSPECTION AND REFAIR ACTIVITIES A.

COMPLETED INSPECTIONS - lARGE 01,AMETER PIPING 3.

REPAIR PHILOSOPHY C.

CURRENT INSPECTIONS - SMALL DIAMETER PIPING V.

IMPACT ON OTHER SAFETY-ELATED AC'TIVITIES VI.

AVAILABILITY OF QUALIFIED INSPECTION PERS0h4NEL VII.

IMPA:T cF "R0uNo R031N" RESuLTS e

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JuSTtFICATION FOR r.0MTIMUED GPERATION UNIT 3 j

A.

LEAx BEFORE BREAK CRITERIA STILL VALID STAINLESS STZEL HAS HIGH TEARING MODULUS CIRCUMFERENTIAL FLAW INSTABILITY IMPOSSIBLE (EPRI RE? ORT 9?

si?-2508-LD, AND MP-2251)

ASYMMETRY'CF WELD SEMSITIZATION AMD EENDING LOADS - VE EXTENSIVE :IELD EXPERIENCE 3 GPM LEAK RATE BEFORE PLASTIC COLLAPSE'(EPRI AND GE TE

SUMMARY

, AUGUST 4, 1983) e 5

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CRACK GROWTH ".00ELS VALID INITI ATION TO DETECTAELE SIZE SLOW SMALL DIAMETER PIPING - CRACX PROPAGATION RAPID, SMAt;L THROU LEAXS l_ARGE DIAMETER PIPING - CRACX PROPAGATICM FAIRLY RA 4

VERY' SLOW BEYOND APPROXIMATELY APPROXIMATELY 20-PERCEN' THROUGHWALL:

30-PERCENT THR0uGHWALL j.

CRACKING FOUMD TO DATE : ALLS WITHIM THESE MODELS b

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ANALYSIS FOR STRUCTURAL INTEGRITY-ADEGUATE '4ARGIN C: SA:E ExtSTS CRACx MuST 3E 3600 AROUND THE PIPE, 63-PERCENT THROUGH'<lALL BE:0RE CODE SAFETY MARGINS INFRINGED a

StztNG UNCERTAINTY PROM ROUND ROBIN DOES NOT IN,RINGE ON STRUCTU:AL MARGIN DI AGRAM (EPRI, GE PRESENTATION AUGUST 4,1983)

STRESS RATIOS iESS THAN 0.5 i

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St z!NG BASED ON CRACx DEPTH SH0'elN TO BE VERY CONSERVATIVE (E?R t

PRESENTATION AUGUST 4, 1983) w e e 4

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CRACxtolo IS '!OT SURPRISING IGSCC TAKES A LONG TIME TO It!!TI ATE TO DETECTAELE LEVE l.

MUMBER OF WELDS IN PAST INSPECTIONS VERY SMALL-IE3 EXAMIrlATIO!! TECHNIQUES MORE SEi!SITIVE TO IGSC NUMBER OF WELDS EXAMINED HAS GREATLY I'! CREASED a

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IJN!T 1 UNIT 3 HAS LESS OPERATING IIME THA!!

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1983 - 1739 X2 9 EF?9 AT UNIT 3 SHUTDOWN k!OVEMBER ll,

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13, 1983 - 1877 EFPD AT UNIT 1 SHUTDOWN - APRIL MAJOR PARAMETE:,5 IDENTICAL FOR ALL 3 UNtTS S/ME PIPE FABRICATORS Att0 :ABRICATION PROCEDURES SAME INST ALL ATION PROCEDURES SIMILAR cPERATING PARAaETERS.

CRACKING IN Utili 3 NOT EXPECTED TO BE AS e

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SUMMARY

0? ?REVIOUS '.lELD INS?ECTIO XS COM?LETED ON 'jNt T 3 A.

11 WELDS INS?ECTED IN MOVEMBER 1981 B.

?!O tRA xtNG OsSERVED C.

TECHNIQUE USED NOT AS SENSITIVE TO IGSCC AS IES TECHNtousS SINGLE ELEMENT TRANSDUCER USED

?ULSE-ECHO TECHNIQUE USED 450 SHEAR WAVE LARGER CRYSTAL USED EVALUATION AT NODE-AND-A-HALF-NO TRANSFER TECHNIQUE USED

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IS XI-IE Cc tCII71 RECUIREMENTS IN SDDITION IO CURRENT IECHNICAL $PECIFICATICMS IECHNICAL SFECIFICATIONS St3MITTED MARCH 25.1983 FCR 2 G?M INCREASE LIMITATICNS.

MARCH 3TH TECHNICAL SFECIFICATIONS ACMINISTRATIVELY EN:CRCED JLLY l,1983.

CURRENT AIMINISTRATIVE RECUIREMENTS

1. CCLD SHUTDOWN REQUIRED IF LNIDENTIFIED LEAKAGE INCREASES BY 2 G?M AVERAGED OVER.24 H0uRS.

2. AFPLIES CNLY AFTER REACTOR HAS BEEN IN RUN MODE FCR 24 HOURS AFTER STARTUP.

3. SURVEILLANCE INTERVAL IS ONCE/8 HOURS.

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METHODCFl.EAXAGEMEASUREME'li O!SCHARGE :LCW F:0M FLCOR CRAIN SLN? INTEGRATED.

INT:GRATION READING CCMPARED TO READING TAKEN 24 HCUF.S EAF. LIE'.

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IV, UNIT 1 ItlS?ECT10'! A'!D RE? AIR ACTit!! TIES l

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S'J:'.".ARY U!D 2.CT!0" P.A'l JULY E. 1933 l.

buMMARY OF,iiELo EX AMI:lATIC:!s A.

NEtos E.< AM t.' LED 1.

RECT 9C Rtssas -

40 SwEEPoLETs -

8 LAaGE 01AMETER - H 91 2.

9Fo Su??tY (20") -

7

~ETuan (24'") -

25

32 0_ial 13 1

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R.Esutis cF ExAMinATtons (sEE F suais 1, 2. Ano 3) i WEtos F.Avine CaACxtn.e RECTRC RIsEas -

5 SwEE?cLETs -

4 lARGE OIAMETEa -

21 33 RM9' l.

SUPPLY & RETuaN - B I.

TOTAL 47 l

2.

WEtos TO BE REPAIRED - 30

LARGE DI AETER RECTRCULUiO'l P!P!.'!G WEL9S ALLOWED INITIAt

?EacENT ER:0R INITIAt CRAcx CRACK 5IZE.:0R

.ARGIN.:0R 3.t: E WEto No.

SrzE OME cycle OPE 2 t e'l KR-1-24 25% - 3 In.

71 184 KR-!-43 28% - 3500 50 78 GR-1-55 29,i - 17.5 in.

53 117 33, - 3500 51 55 5

GR-1-54 KR-1-47 23% - 21 In.

53 174 GR-1-27 35% - 14 IN.

55 83 GR-1-57 32% - 3500 51 59 KR-1-45 23% - 3500 41 78 30% - 25 In.

55 120 GR-1-61 KR-1-52 27% - 3500 50 85

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1 TABLE 1 OVFRLAY DESIGNS Steady Pipe Crack Ove rl ay Ove rl ay State Wald Nc.

Diameter Depth Thickness Lencth Stress Cate-ry KR-1-37 22in.

35)

.200in.

4. Sin.

6,099" psi 1

KR-1-15 22 27

.200 4.5 6,099",

1 KR-1-3 23 43

.35 7.0 c,411 2

GR-1-3 29 35

.25 7.0

,145 3

-GR-1-53 28 45

. 35 7.0 7,345 2

GR-1-54 28 45

.35 7.0 8,334 2

GR-1-60 23 36

.25 7.0 7,725 3

KR-1-18 12 35

.125 2.5 15,110 4

KR-l-21 12 35

.125 2.5 12,461 4

KR-1-22 12 35

.125 2.5 11,754 4

KR-1-16 12 35

.125 2.5 13,233 4

GR-1-41 12 12

.200 4.5 35,244 5

. - GR-1-46 12 20

.125 2.5 12,593 4

• v D-RHR-1-17 24 31

.200' 4.5 13,224 6

0-RHR-1-13 24 20

.200 4.5 15,058 6'

t D-RHR-1-15 24 30

.200 4.5 9,461 6

DS-RH R 9 20 29

.200 4.5 13,529 6

DS-RHR-1-53 24 41

.200 4.5 13,427 6

DS-RHR-1-ll 20 24

.200 4.5 12,633 6

DS-RHR-1-10 20 30

.'200 4.5 12,057 6

DS-RH R-1-5 24 31

.200 4.5 11,879 6

D-RHR-1-5 24 36

.25 7.0 11,599 7

DS-RHR-1-4 24 30

.200 4.5 14,071 6

D-RHR-1-20 20 43

.200 4.5 12,450 6

D-RHR-1-8 24 25

.200 4.5 11,396 6

.25 7.0 13,617 7

DS-RHRl-4A

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Pressure stress only - assumed equal to that fo r J

weld number KR-1-3 4 i

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TAlli.E 2 OVEltl.AY CATEGORIZ ATION Initial Source of Wold Applied Crack Pipe Pipe Overlay overlay Cntegory Residual Stress Stress Size

• Diamotor Thicknens Thickness Le ntit h 1

Reference 3 6,099 psi 70%

22" 1.03

.2" 4.5" 2

Computer Itun 8,411, 90 28.51 1.322 35 7.0 3

Compu te r Ru n 8,145 72 28.15 1. 1 311

.25 7.0 4

Reference 4 15,l10 70 12.75

.789

.125 2.5 S

Reference S 35,344**

24 12.75

.579

.20 4.5 6

Compu te r Ru n 15,0511***

11 6 20.

1.031

.20 4.5 7

Computer Itun 13,617 till 20 1.031

.25

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The measured U.T. depth was multiplied by 2.0 for une in the crack growth analysis.

The tabulated crack sizes are in percent of unrepaired pipe thicknens.

• 579 35,344 Crack growth evaluation purf on med uning 26,270 psi a

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15,058 Crack growth, evaluation performed using 12,612 psi 3,g g

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TVA-01-031 a

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i Figure 2 CRACK GROWTH FOR CATEGORY 1

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TIME (MONTHS 1 Figure 4

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CRACK GROWTH FOR CATEGORY 3

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D Figure 8 CRACK GROWTH FOR CATEGORY 4

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20 40 60 30 C0 TIME (MONTHS) l Figure '10 CRACK GROWTH FOR CATEGORY 5 e.

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Figure 12 CRACK GROWTH FOR CATEGORY 6

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VI. INSPECTION SIT"ATION I.

EXPANSION OF IGSCC INSPECTION

- CORE SPRAY (ALL STAINLESS WELOS) 27

- REACTOR '4 ATE 3 CLEANUP (ALL ~4 ELDS 4

INSIDE FIRST ISCLATION VALVE)

- HEAD SPRAY (ALL ~4 ELDS INSIDE THE 21 FIRST ISCLATICN VALVE) l TOTAL 52 II.

INSPECTION CF 3RC'4NS FE33Y UNIT j' REPAIR '4ELOS (36-PLUS REPAIR '4ELOS)

/-

- UT OVERLAY FCS BONDING AND DEFICTS

- UT OVERLAY FOR "UNDERCLA3 CRACEING"

- UT TO ESTABLISH NEW IGSCC 3ASELINE AUGUST 18 - CALI3 RATION ELCCE FABRICATED AUGUST 22 - INSPECTION PRCCEDURES APPRO,VE3 WAITS SAR e,

INSPECTICN SITUATICN

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- CURRENTLY EEHIND SCHECULE CUE TO DRAIN CN PERSCNSEL TO SUPPCRT ERIWNS FERRY UNIT 1 IGSCC INSPECTION SEQUOYAH INSPECTION SITUATICN

- SEQUOYAH UNIT 2 CURRENTLY DO*4N FOR REFUELING GUTAGE AND RCUTINE.

J-IN-SERVICE INSPECTION IN PRCGRESS

- EXPECTED RETURN TO SERVICE SEPTEM3E3'30, 1983 t

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IMPACT OF R0uMD R0slN RESULTS A.

TVA wAS TEAM 5 B.

RESULTS SHOW VERY GOOD STANDARD DEVIATION SA'ME PERSONNEL, EQUIPMENT TECHNIQUE AS USED ON 3R0xNS ERRY C.

UNIT 1 ERRY UN!T l C

RESULTS LEAD TO ADDITIONAL CONSERVATISM ON 3RO D.

GENERAL c0NPIDEncE IN T'/A's SiztNa cA?AsIt!TY SUPP 0U":

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