ML20126F616

Forwards LC Shao Draft Ltr Entitled, RES Position on SG Tube Integrity & J Hopenfeld Entitled, Reply to Request for Comments on Draft RES Position on SG Tube Integrity by LC Shao

ML20126F616
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	12/22/1992
From:	Kokajko L Office of Nuclear Reactor Regulation
To:	Cross J PORTLAND GENERAL ELECTRIC CO.
References
NUDOCS 9212310021
Download: ML20126F616 (15)

Text

_ _ _ _ _ _ - _ _ _ - _ _

I December 22, 1992

?~

Docket No. 50-344 Mr. James E. Cross Vice President and Chief Nuclear Officer Portland General Electric Company 121 S.W. Salmon Street Portland, Oregon 97204

SUBJECT:

INTERIM PLUGGING CRITERIA FOR 1ROJAN NUCLEtR PLANT Enclosed are two documents for your review.

The fir 1t document is a draft memorandum from L. C. Shao to E. S. Beckjord entitled (Enclosure 1)

"RES Position on Steam Generi. tor Tube Integrity." The second document (Enclosure

2) is a memorandum fr a J. Hopenfeld to G. Burdick dated December 9, 1992, entitled, " Reply to your request for comments on Draft 'RES Position on Steam Generator Tube Integrity' by L. C. Shao."

Both of these documents have been placed on the docket file for Trojan Nuclear Plant. As such, both documents have been placed in the Public Document Room.

Sincerely, Original signed b/

Lawrence E. Kokajko, Senior Project Manager Project Directorate V Division of Reactor Projects lil/IV/V Office of Nuclear _ Reactor Regulation

Enclosures:

As stated cc w/ enclosures:

See next page D1513LW110B:

Docket File.

NRC & Local PDRs PDV Reading File

'JRoe MVirgilio TQuay DFoster LKokajko OGC, 15B18 ACRS(10). P-315 PDV Plant File KPerkins, RV SHoffman, 14B20

/

OFFICE LA/POV

,/

BMMDV D/PDV hAME DFoster M

[Kokajko:1h TQuay DATE

/9'/(k92D

\\1//O /92 l>/7V92 OfflCIAL RECORD COPY DOCUMENT NAME:lJIPC.LlR b.

d 240010 N

922231002 921222 fbr i\\

PDR-ADOCK 03000344 P

PDR

r-Q Cih p*

UNITED STATES n

g

.,i NUCLEAR REGULATORY COMMISSION o

WA$HINoTON D.C. 30M6 4,,..., * *#g December 22, 1992 I

Docket No. 50-344 Mr. James E. Cross Vice President and Chief Nuclear Officer Portland Ge'neral Electric Company 121 S.W. Salmon Street Portland, Oregon 97204

SUBJECT:

INTERIM PLUGGING CRITERIA FOR TROJAN NUCLEAR PLANT Enclosed are two documents for your review. The first document-(Enclosure 1) is a draft memorandum from L. C. Shao to E. S. Beckjord entitled. "RES Position on Steam Generator Tube Integrity." The second-document (Enclosure

2) is a memorandum from J. Hopenfeld to G. Burdick dated December 9, 1992, entitled, " Reply to your request for comments on Draft 'RES Position on Steam Generator Tube Integrity' by L. C. Shao."

Both of these documents have been placed on the docket file for Trojan Nuclear Plant. As such, both documents have'been placed in the Public Document Room.

Sincerely.

j Lawrence E. Kokajko, Senior. Project Manager Project Directorate V

' Division ~of Reactor Projects Ill/IV/V Office of. Nuclear. Reactor Regulation-

Enclosures:

As stated cc w/ enclosures:

See next page

t Mr. James E. Cross Portland General Electric Company Trojan Nuclear Plant cc:

Senior Resident Inspector U.S. Nuclear Regulatory Commission Trojan Nuclear Plant Post Office Box 250 Rainier, Oregon 97048 Mr. Michael J. Sykes, Chairman Board of County Commissioners Columbia County St. Helens, Oregon 97501 Mr. David Stewart-Smith Oregon Department of Energy Salem, Oregon 97310 Regional Administrator, Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 Mr. Tom Walt General Manager, Technical Functions Trojan Nuclear Flant 71760 Columbia River Highway Rainier, Oregon 97048 Mr. Lloyd K. Marbet 19142 S.E. Bakers Ferry Road Boring, Oregon '97009 Mr. Jerry Wilson Do It Yourself Committee 570 N.E. 53rd Hillsboro, Oregon 97124 i

Mr. Eugene Rosolie Northwest Environmental Advocates 302 Haseltine Building 133 S.W. 2nd Avenue Portland, Oregon 97204 Mr. Robert Pollard Union of Concenned Scientists 1616 P Street, Suite 310 Washington, DC 20036 r

s e--

ENCCOSURE1 ggy UNITED $TATES

,,pe -

)

([,g NUCLEAR REGULATORY COMMISSION

[h),

t

(

,J' wAuwatos, p. c.rosso

'. l Jd 1

h

//fYL J.4h cJi y MEMORANDUM TORT ERIC S. BEC&TORD, DIRECTOR, OFFICE OF NUCLEAR REGULATORY RESEARCH FROM:

IAWRENCE C. SHAO, DIRECTOR, DIVISION OF ENGINEERING, RES SUBJECTt RES POSITION ON STEAM GENERATOR TUBE INTEGRITY The Division of Engineering has provided a discussion of the key aspects of the rationale used to support steam generator tube alternate plugging criteria (APC), and to provide independent conclusions on its viability as$1eg only t ap interim (one fuel cycle

~

critorion.

The interim APC a

'gracking (ODSCC) pecific ca)se' of o the s outer diamator stress corros intergranular attack (ICA) Cat and-intersectionsinthesteam'g[e)nerator.e support plate (TSP)

The technical rationale possible on data and analyses available from our researc elsewhere in the rtechnical literature, but also represent staff technical experience and opinions. The report endeavors to maintain a clear distinction betwoon staff opinion and published data.

Lawrence c. Shao, Director, Division of Engineering Technology

Enclosure:

As stated Cc*

,y y

n

..w

~

.n

,e--

.smv-n

,.-,,..g.

~

'ee,

.m>..

.~e--

. && ' /

VGb, a

Enclosure Discussion of Technical Rationale for Interim (One Cycle) Alternate Plugging Criteria for Steam erator Tubes The purpose of this report is to provide a discussion of the key technical aspects of the rationale used to support steam generator tube alternate plugging criteria (APC) and to provide independent conclusions on its viability as an interim (one fuel cycle) criterion. A chronology of events relative to this issue is provided as attachment 1.

The interim APC applies only to the specific case of outer diameter stress corrosion cracking (ODSCC) and intergranular attack (IGA) at tube support ' plate (TSP) intersections in the steam generator. The technical rationale presented in this report are based to the greatest extent possible on data and analyses available from NRC research and elsewhere in the technical literature but also represent staff technical enerience and opinions.

The report endeavors to maintain a clear distinction aetween staff opinion and published data.

The rationale presented in this report are based on technical considerations which we believe are adequate to justify APC for one fuel cycle.

Longer term technical considerations, such as reliability and sensitivity of NDE techniques for steam generator tube ins)ection, are the subjects of NRC research which is being coordinated with the Office of Nuclear Reactor Regulation (NRR) as part of an overall APC action plan, j

i (1)

Background:

Steam generator tube structural integrity guidance provided in Regulatory Guide 1.121 has generally translated into a 40% through-wall

" plugging limit" for flaws in steam generator tubes as part of the plant technical specifications.

However,ieddy current inspection techniques routinely applied to detect and size cracks in steam generator tubes are not highly reliable for detecting and sizing the crreks until they are actually beyond 40% through-wall.

Further, it has been argued by the industry that the 40% plugging limit is overly conservative, at least for the case of axial ODSCC/lGA cracks confined to tube support plate intersections. The claim for over-conservatism is based on burst tests of cracked tubes removed from steam generators, e.g., pulled tubes. Detailed examinations of these tubes have revealed short cracks which, when tested,

)roduced correspondingly high burst

pressures even for up to through wall crac(s.'. NRC research results on tubes with idealized flaws (electrical discharge machined slots) support the contention that the tubes retain significant structural integrity even for up to.through wall cracks provided that the cracks are short. From this research, "short" can be defined as less-than 0.5 inches, which is the length of through-wall crack that would be predicted to result in a burst for 7/8 inch diameter, 0.050 inch wall thickness tubing under main steam line break (MSLB) differential pressure * (see attachment 2).

Based on this data and supporting analyses, the industry has proposed an alternative to the Regulatory Guide 1.121 guidance, the so-called alternate plugging criteria (APC), for steam generator tubes.

The APC are based on correlations between the voltage L:

~

j g

- s i

l amplitude recorded during eddy current tube inspections with a bobbin coil and The APC subsequent measurements of the tube burst pressures and leak rates.

are also currently restricted to ODSCC at tube support plate intersections.

1RR) has approved several licensee 1he Office of Nuclear Reactor Regulati requests for steam generator APC o him(onefuelcycle) basis.

(2) Technical Precedent: The NR ctions accepting industry requests for APC are not the first time th NRC has allowed specific exceptions to the Plants with steam 40% plugging limit for flaws in steam generator tubes '

generator tubes pitted beyond the 40% limit have been allowed to continueExcepti operation based on burst test results of pulled tubes.

"The F* criterion plugging criteria have also been based on the F* criterion.

exempts a utility from the 40% plugging limit for that portion of the defective tubing within the tubesheet below a certain dimension (the F*

distance) from the top of the tubesheet or from the top of the hardroll, whichever is lower. The purpose of the F* exemptions was to provide a distance within the tubesheet beyond which it could be reasonably assured that a double-ended break in a tube would not result in the tube leaving the tubesheet.

While there is a small gap between the TSP

-(3) Support Plate Considerations:and the tube, the TSP should provide restrain The contained within the TSP thickness (0.75 inches for the Trojan TSP's).

restraint would be provided by the rigid TSP as the tube expands against it under the influence of a differential pressure such as that caused by a MSLB.

(4) Degradation Mechanisms:

While all of the initiation, growth, environmental sensitivity, and synergistic effects of the various mechanisms of steam generator tube cracking are not completely understood, growth of ODSCC cracks in TSP regions is not expected to be significant during one fuel

cycle, for purposes of this report, significant can be defined as a through-Evidence from wall crack growing on the order of 0.5 inches beyond the TSP.

the Trojan' pulled tube examinations has shown that the OD lengths of the For Trojan, the cracks were cracks ranged almost up to the TSP thickness. Thus, for the Trojan cracks to generally confined within the TSP thickness.

become significant, they would have-to extend on the order of 0.5 inches I

beyond the TSP intersection during one fuel cycle. Upper bound laboratory 0DSCC growth rate data indicate that crack growth of this magnitude would not be expected to occur during one fuel cycle.

I (5) Probability of Initiating Event (s): The key initiating event for SGTR is considered to be a main steam line break (MSLB). The MSLB would cause approximately a 2600 psi pressure differential across the steam generator tubes. An MSLB has never occurred in a U.S. plant. Quoting from the "Under the Evaluation and improvement 'of NDE enclosure to Reference 5, Reliability for Inservice inspection of Licht Water Reactors Proaram sponsored by the NRC, a team of experts estimated the median frequency of an NSLB to be This extrapolates to a frequency estimate 1.7 x 10 per reactor year.....

of 6.8 x 10 per reactor year for a four-loop plant."

(6) Steam Generator Tube Rupture Experience:

U.S. industry experience since-1975 has documented only 6 SGTR's and only two of those were due to ODSCC The more recent of the two was the SGTR at McGuire Unit 1 in 1989.

cracking.

The McGuire SGTR was initiated by ODSCC at a manufacturers burnish marking.

i

g The cracking progressed under the combined influence of residual stresses from the burnish marking and a

• local metallurgical contaminant. Failure was attributed to linking of several axial ODSCC cracks to form a 3.5 inch OD macrocrack on the cold leg side of the generator. The ODSCC cracking was in the free span away from TSP ions.- None of the documented SGTR's were associated with ODSC rsections.

(7) Pulled Tube Burst Test Results ressure test results from the pulled f

Trojan steam generator tubes have shown no leakage under normal operating or HSLB pressure conditions,8. When pressurized to failure, the burst pressures measured for the pulled tubes were in excess of the MSLB pressure by at least a factor of two.'

(8) Steam Generator Tube Rupture Tests and Research - Research results from burst tests of tubes using idealized flaws (unifom thinning and EON slots) has provided a significant body of data on tube integrity.

Equations which have been fitted to the data generated for EDM slot s)ecimens have shown that 0.5 inches would be the length of through-wall flaw t1at would be expected to 9

result in burst at HSLB pressures for 7/8-inch diameter. 0.050 inch wall thickness tubing' (see attachment 2). The equation for the through-wall EOM slot represents an extrapolation from data measured on up to 90% through wall f

slots. NRC sponsored research has shown that "the empirical equation developed from EDM notches provides a realistic estimate of remaining margin to failure for tubes with stress corrosion cracking when bounding flaw dimensions are used. An empirical equation fitted to data from burst tests of uniformly thinned steam generator tubes has also been developed.' This equation is contrasted with the EDM equation in attachment 2.

It can be seen that the two equations are of similar form but that the uniform thinning equation provides more conservative estimates of tube burst pressures for flaw depths greater than a/t of 0.8, where a - flaw depth and t - tube wall thickness. Use of either equation to bound degradation below a/t = 0.8 should yield similar results in terms of burst pressure. However, the EOM equation provides a more accurate representation of stress corrosion cracking and j

should be used for flaw depths greater than a/t 0.8.

(9) Sumary and

Conclusions:

Based on a review of steam generator tube operating experience, on destructive examinations of tubes removed from the Trojan plant, and on expert opinion concerning the frequency of main steam line break, it is concluded that it is rea;onable to continue operation of the Trojan plant for one fuel cycle with flaws greater than 40% through-wall at TSP intersections.

Qi,s conclusion is supported by the information provided a

in the_ preceding paragrapns, which is summartzed below.

Subsequent operation wn i require aoaitiori4i reviiw eiter compieucn of one cycle and will include consideration of information developed at that time.

o Examination of tubes removed from the Trojan steam generator has l

revealed cracks which are generally confined to the TSP thickness.

I Pressure tests on these tubes showed no leakage under normal operating l

or MSLB conditions. The burst pressures were in excess of the HSLB pressure by at least a factor of two.

Further, these tests did not I

account for the additional restraint that could be anticipated from the TSP which should further increase the burst pressure for the service

geometry, l

L L

-c

Based on empirical correlations between burst pressure and crack size, o

validated by both laboratory data and by burst tests of' tubes removed from service, for cracks to be significant, they must extend outside of the TSP intersection on the order of 0.5 inches. However, based on service experience and examination of tubes removedjfr s.the Irojan steam generators, the existence of cracks extejintrappreciably bbnd n

the ISP intersections is unlikely at this timeJased on conservative 1aboratory test data on stress corr 0&lon crack growth rates, it is unlikely that existing cracks within the TS would extend to a critical length drring one fuci cycle.

Itius, cracks that are within the TSP can WLIstand the MSLB pressure, and it is unlikely that such cracks would extend to a critical-length during one fuel cycle. On this basis, it is concluded that operation for one fuel cycle is justified.

1 i

I

- t i

VD

'

• e' if' l

References s

• 1rojan Nuclear I'lant Steam Generator Tubt Repair Criteria for Indications at Tube Support Plates," Westinghouse Energy Systems, WCAP-1-

13129, Revision 1, December,1991, WESTINGHOUSE PROPRIETARY CLASS 2 HUREG/CR-0718, Steam Generator Tube Integrity Program, Phase 1 Report, 2-USNRC, September, 1979.

HUREG/CR-5150, Steam Generator Operating Experience, Update for 1984-3-

1986, USNRC, June, 1988 NUREG CR/5117, Steam Generator Tube Integrity Program / Steam Generator 4-Group Project, Final Project Summary Report, USNRC, May 1990.

Memorandum, From C.J. Heltemes to F.P. Gillespie, GI-163, Hultiple Steam 5-Generator Tube Leakage, September 28, 1992.

HUREG CR/5796, Steam Generator Operating Experience, Update for 1989-6-

1990, USNRC, December, 1991.

HUREG CR/2336, Steam Generator Tube Integrity Program, Phase !! Final 7-Report, USNRC, August, 1988.

~~

%y;y

.n y Chronology of staff related events relative to Steam Generator Tubu Integrity The NRR staff had been reviewing Alternate Plugging Criteria (APC) during 1991, and in mid-December 1991, received a 0

series of requests from Portland General Electric relative to operation of the steam generators of the Trojan Nuclear Plant, wherein use of the industry APC was an important issue.

1991, Mr. J. Hopenfeld filed a Differing On DecembLr'23, Professional Opinion (DPO), noting that "Recent experience O

at the Trojan plant indicates that the present inspection techniques are not sufficiently sensitive to detect steam generator degradation." Mr. Hopenfeld's DPO vent on to note his concern "that a Main Steam Line Break (MSLB) outside containment could trigger a multiple steam gnnerator tube failure which would than (sic) result in a core melt because of depiction of coolant inventory."

A Safety Evaluation Report by NRR was published on February 1992, which allowed operation of the Trojan Nuclear Plant O

(cycle 14) under reduced leakage 5,for one additional cycle allowances, and under repair criteria that include application of APC.

On March 16, 1992, Dr. Joseph Huscara published a Memorandum on " Steam Generator Tube Inspection, Integrity O

and Plugging Issues" wherein he stated his " concerns with generic acceptance of industry proposals for alternate tube plugging criteria which would allow operation of steamtubes."

generators with known through-wall cracked (leaking)

J. Hopenfeld filed another paper titled O

On March 27, 1992, Multiple Steam Generator Leakage."

"A How Generic Issue This document contained an analysis titled " Safety Issue Relating to Continuous Operation with Degraded Steam Generator Tubes" which indicated "a core melt probability frequency of 10 E-4/Ry with containment bypass."

via a RES moved to decide on prioritization of this issue, Memorandum of September 28, 1992 from C. Heltemes to F.

O Gillespic titled "GI-163, Hultiple Steam Generator Tube Leakage;" this prioritization evaluation was revised on November 16, 1992.

1992, NRR replied via a Memorandum from On November 24, "Hultiple O

Gillespio to Heltomes titled Generic Issue 163, Steam Generator Tube Leakage."

RES submitted a subsequent Memorandum on November 30 from Heltomes to Gillespie, on the same subject, stating "we O

I 3dNd#*

' ey yy

,;p i '

agree that the preliminary prioritization of this issue that we sent to you for corraent on September 28, 1992 nay not have accurat.ely reficcted the current Mconsing. position that has been applied to 'frojan and

?r plants."

1,,

I e

/

L-gy.)

' J

~ 't > ;

";thf srrm.~s:- (x}

BURST PRESSURE PARAMETER CURVES s

EDM SLOT AND UNIFORM THINNING SPECIMENS 1 -:-

0.9 -

0.8 -

O.7 -

O 0.6 -

dQ 0.5 -

n.

<C EDM. alt = 0.8 b

0.4 -

t.u O

UNIFORM THIN O'

~

alt = 0.8 t

nsis fg UNIFORM TH EDM. alt =0.9 a/t = 0.9 o,_

11 1 1. ;.

~

0 ECM, alt =1.0 Chrough-wa9

~

^ M ^ ^~*-7 ^ ^ ^=

0 0.4 0.8 1.2 1.6 2

LENGTH (inches)

e..r

.rawe 3

ENCLOSURE 2 IlOTE TO:

G. Durdick TROM :

J.

Hopenfeld

SUBJECT:

Reply to your request for comments on Draf t "RES POSITION ON STEAM GENERATOR TUDE INTEGRITY" by L.C. Shao Based on certain data, discussed in Items 1-8, the document concludes that "it is reasonhble to to continue operation for one fuel cycle with flaws greater than 40) through-wall at TSP intersections" the document further suggests that " subsequent operation will require additional review after completion of one cycle and will include consideration of information developed at that time".

GENERAL COMMENT

The information provided in Items 1-8, of the subject document does not address the main issue concerning steam generator tube integrity which arose from recent operating experience with ODSCC.

The issue in as follows:

Is it safe to operate plants where an accident such as steam or feed line break may open existing but previously undetected cracks, which will result in a significant primary to secondary leakage.

Whether the' leakage is significant or not would depent on whether the operator can stop the leak before the RWST is depleted.

The fact that crack:. within the TSP can withstand the HSLB pressure and that their length will not becomo critical during one fuel cycle is not an indication that they also will not leak. The Trojan burst test results show that three out of the 21 test specimen developed leaks at pressures, of 3300 psi, 7500 psi, and 5500 psi,.

with an average depth of penetrations of 38%,

58% and 72%

respectively.

Item (7) points out that the above specimen "have shown no leakage under normal opr. rating or MSLB pressure conditions". IT FAILS TO POINT OUT, HOWEVER, THAT THERE IS HO DATA WilICH WOULD ALLOW ONE TO g

RELATE THE ABOVE LEAKAGE WITH THE OBSERVED DEGREE OF DEGRADATION.

In other words, if these specimen had more severe wall penetration would these specimen had leaked under MSLB loads. Considering that the 21 specimen represent a sample of a population of 13,000 the conclusions in (7) above are questionable.

The document does not mention the fact that two other tubes which were pulled out of two US plants developed leaks at SLB pressures which were at least an order of magnitude higher than under normal delta ps'. A third tube, trom a Belgian plant indicate a factor of 8

increase in leakage under SLB conditions.

Theoretical considerations, on the other hand, indicate a factor of 1000 increase in leakage under SLB conditions.

i 1

In conclusion, the absence of a deterministic and empirical models for these newly observed cracks precludes the conclusions reached i

in the subject document. The document claims that the conclusions in Item 9 are supported by items 1~8 could be considered valid only if one ignores the theoretical considerations and the available plant data which indicate that substantial LEAKAGE may result at SLB pressures even if the tube does not rupture.

Finally the justifications for any plant operation should not be based on staff opinions or published data on SCC.

SCC is a semi capirical art, in the absence of applicable database other routes of approaching the problem should be considered.

The justification for operating with cracked tubes should be based on what procedures would the operate follow given certain primary to secondary leak and a MSLB between the containment and the MSIV.

These justifications should clearly demonstrate compliance with 10CFR100.

I beleive that the staff can more properly judge operator action than predict localized corrosion behavior.

SPECIPIC COMMENTS:

Item 1.

The EDM initiated groves studies provide some measure of the ability of the tube to resist rupture given certain known wall imperfactions.

It bears little relation to how ODSCC form, propagate and leak in steam generator environments.

Item 4 This definition of "significant" is questionable. It makes no difference whether the cracks extend beyond the tsp if they leak at the gap.

It appears that operators rely on such leakage because they lowered the leakage requirements during normal operations. Unles one can show that the TSP will cause cracks to plug and they will stay plugged under MSLB pressures the above definition may lead to confusion.

The statement that " upper bound laboratory ODSCC....

is not would not be expected to occur during one fuel cycle supported by datau The document should compare and present plant and laboratory data with regard to stress intensities and environments before making such claims.

Item 5 y

e-m

,-~-

m

-or

--s----,

,e-

Not clear where the 2600 psi comes from.

I expect a smaller pres.sure.

The high frequency quoted, 6.8x 10-4 contradicts the statement that

" it is reasonable", Item 9, because this frequency would result in a core melt probability of 6.8x 10-3 with containment bypass as discussed in the March 27 Hemo. The above number is considerably higher than present safety goals.

The statement that the key initiating event for SGTR is MSLB is incorrect when taken in the context of the entire document. Item 6 contradicts this statement.

Item 7 Although this item is correct, as stated, it presents only part of the data. As already discussed, three tubes leaked at Trojan.

and a.'so three tubes from other plants leaked at MSLB pressures.

Rudimentary consideration dictate that leakage increases when delta p across the wall in-increased.

Item 8 The lengthy discussion of uniform thinning only confuses the main issue. There are several ways that the reduction load bearing capabilities of a component due to corrosion can be accounted for there is nothing special about these equations. The ASME code takes this into account. The main problem here is I4CALIZED corrosion with an UNKNOWil ATTACK RATE.

ATTACIDiENT 1 Second Item : Dr. instead of Mr.

or just Hopenfeld The following is missing:

On Sept 1.

11, 1992 J. Hopenfeld filed an addendum to the March 27, 1992 concluding that " strong coupling exists between hot leg mass flow, SG tube leakage and crack propagation. If confirmed, such relation between system behavior and undetected tube defects may cause small leaks to quickly enlarge and results in a MULTIPLE TUDE RUPTURE BEFORE THE RCS IS DEPRESSURIZED BY FAILURE OF THE SURGE LINE.

THE RESULTANT CONTAINMENT BYPASS WILL INCREASE THE SOURCE TERM."

J. Hopenfeld cci P. Norian, Gm. Mazettis, W. Minners i

.