_ _ _ _ _ _ _ _ _ _ _ _ _ _ _

a

ATTACHMENT 1<50. Q. What do thoco cnolycso chsw? .

-

A. The fracture analysis demonstrated that.the Shoreham

vessel has a high margin of protection against brittlefracture following a design basis LOCA. For example,

1 1/2 inchesickness flaw,$m agaimst

for a uarter6afd rher

deep and 9 inches long, thegbri[tle fracture - ' / siO' = q' r %^ factor of 2. ,

51. Q. What causes the high brittle fracture margin in the

Shoreham vessel?

A. The Shoreham vessel has significant advantages since

the key f actors required for unstable crack propagation-- radiation embrittlement and high pressure stresses

following thermal shock -- do not occur in a BWR. The

reasons for this are:

(1) The radiation flux level at the wall location inthe Shoreham vessel is not enough to cause appre-

ciable embrittlement of the vessel during its life-

time. 'The low radiation flux level is due to thefact that there is a water annulus between the.core and the vessel, and that there is low power

density in a BWR. Thus the toughness level re-

quired to assure ductile behavior is maintained

throughout the design life.

(2) Unstable crack propagation requires the presence of

pressure stresses following the ttiermal shock.However, in a BWR, when there is emergency

.

8207120169 820707PDR ADOCK 05000322T PDR

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_ _ _ _ _ _ _ _ _ _ .

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s

CERTIFICATE OF SERVICE

In the Matter ofLONG ISLAND LIGHTING COMPANY

(Shoreham Nuclear Power Station, Unit 1)Docket No. 50-322 (OL)

I hereby certify that copies of CORRECTIONS ANDADDITIONS TO THE " TESTIMONY OF PETER S. BARRY, CRAIG K. SEAMANAND SAM RANGANATH ON SUFFOLK COUNTY CONTENTION 25 AND SHOREHAMOPPONENTS COALITION 19 (a) -- PRESERVICE AND INSERVICE

'

INSPECTION PROGRAM AND REACTOR PRESSURE VESSEL INTEGRITY" wereserved upon the following by first-class mail, postage prepaid,on July 7, 1982:

Lawrence Brenner, Esq. Atomic Safety and LicensingAdministrative Judge Appeal Board PanelAtomic Safety and Licensing U.S. Nuclear Regulatory

Board Panel CommissionU.S. Nuclear Regulatory Washington, D.C. 20555Commission

Washington, D.C. 20555 Atomic Safety and LicensingBoard Panel i

'Dr. Peter A. Morris U.S. Nuclear RegulatoryAdministrative Judge Commission |

Atomic Safety and Licensing Washington, D.C. 20555Board Panel

U.S. Nuclear Regulatory Bernard M. Bordenick, Esq.Commission David A. Repka, Esq.

Washington, D.C. 20555 U.S. Nuclear RegulatoryCommission

Dr. James H. Carpenter Washington, D.C. 20555Administrative JudgeAtomic Safety and Licensing David J. Gilmartin, Esq.

Board Panel Attn: Patricia A. Dempsey, Esq.U.S. Nuclear Regulatory County Attorney

Commission Suffolk County Department of LawWashington, D.C. 20555 Veterans Memorial Highway

Hauppauge, New York 11787

| Secretary of the Commission Stephen B. Latham, Esq.U.S. Nuclear Regulatory Twomey, Latham & SheaCommission 33 West Second Street

,Washington, D.C. 20555 P. O. Box 398

| Riverhead, New York 11901

Herbert H. Brown, Esq. Ralph Shapiro, Esq.Lawrence Coe Lanpher, Esq. Cammer and Shapiro, P.C.Karla J. Letsche, Esq. 9 East 40th StreetKirkpatrick, Lockhart, Hill, New York, New York 11901

'

Christopher & Phillips Albany, New York 12223

1

- . - - _ . . _ __

..

6-2-

i

l

8th Floor |

1900 M Street, N.W. Howard L. Blau, Esq. l

Washington, D.C. 20036 217 Newbridge RoadHicksville, New York 11801

Mr. Mark W. GoldsmithEnergy Research Group Matthew J. Kelly, Esq.400-1 Totten Pond Road State of New YorkWaltham, Massachusetts 02154 Department of Public Service

Three Empire State PlazaMHB Technical Associates Albany, New York 122231723 Hamilton AvenueSuite K Mr. Jay DunklebergerSan Jose, California 95125 New York State Energy Office

Agency Building 2Empire State PlazaAlbany, New York 12223

Respectfully submitted,

LONG ISLAND LIGHTING COMPANY

0bh e' Daniel O. FlanaganNJ

Hunton & Williams707 East Main StreetP.O. Box 1535Richmond, Virginia 23212

|

!

|

.

__ , - . . - . _ ,_,

f.

-.

.

LILCO, June 14, 1982*

.

UNITED STATES OF AMERICANUCLEAR REGULATORY COMMISSION

Before the Atomic Safety and Licensing Board

In the Matter of ))

LONG IST,AND LIGHTING COMPANY ) Docket No. 50-322 (OL))

(Shoreham Nuclear Power Station, )Unit 1) )

TESTIMONY OF PETER S. BARRY,CRAIG K. SEAMAN AND SAM RANGANATH

ON SUFFOLK COUNTY CONTENTION 25 ANDSHOREHAM OPPONENTS COALITION 19(a) --

PRESERVICE AND INSERVICE INSPECTIONPROGRAM AND REACTOR PRESSURE VESSEL INTEGRITY

PURPOSE ,

This testimony demonstrates that the preservice inspec- t

tion (PSI) and inservice inspection (ISI) programs comply with

| 10 CFR 50.55a(g). The technology and the methods used in the

f examinations are established by the ASME Code and assure the i

; ,

l effectiveness of the inspection programs. Basic technology

will not undergo significant change so that there will be somei

correlation between the PSI and ISI results.

! Shoreham has taken numerous steps to reduce the number

of noninspectable areas so that a high percentage of the plant,

.

- _\w _.[ m . , _

. _ . _ _ _ . _ _ _ . _ . , . . _ . . __ _. _ _ _ .

.

.

| .i

-2-

.

will be inspected. However, it is not necessary for all areas

of a plant to be inspected. The same controls on the quality

of welding materials, the same welding techniques and quality

assurance programs assure that noninspected welds have the same

quality as inspected welds. For this reason, relief requests

will have no impact on the safety of Shoreham. Exemptions,

similarly, have no impact on safety, because they are permittedwhen suitable alternatives have been performed.

The significant quality control requirements of

Regulatory Guide L.150 have been met or exceeded for the pre-

service inspection of the reactor vessel. The Regulatory Guide

does not establish " travel time," nor does it deal with ALARA

Regulatory Guide 1.2 has been complied with atconcerns.Shoreham and the Shoreham pressure vessel will behave in a

non-brittle manner.

,

, _ _ , - _ __ _ _ - -

.

.

.

*

UNITED STATES OF AMERICA! NUCLEAR REGULATORY COMMISSION!

.Before the Atomic Safety and Licensing Board

| In the Matter of )! )i LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 (OL)| )i (Shoreham Nuclear Power Station, )' Unit 1) )

TESTIMONY OF PETER S. BARRY,CRAIG K. SEAMAN AND SAM PANGANATH

ON SUFFOLK COUNTY CONTENTION 25 ANDSHOREHAM OPPONENTS COALITION 19(a) --

PRESERVICE AND INSERVICE INSPECTION-| PROGRAM AND REACTOR PRESSURE VESSEL INTEGRITY

1. Q. Would you please state your names?|

| A. My name is Peter S. Barry. My business address is!

! Nuclear Energy Services, Shelter Rock Road, Danbury,l

| Connecticut 06810.t

My name is Craig K. Seaman. My business address is

Long Island Lighting Company, Shoreham Nuclear Power

Station, P. O. Box 618, Wading River, New York 11772.

My name is Sam Ranganath. My business address is

General Electric Company, 175 Curtner Avenue, San Jose,

California.

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

I

..

.

-2-

2. Q. Would you please summarize your professional

qualifications?

A. (Barry) I am director of Technical Operations for

Inservice Inspection activities for Nuclear Energy

Services. I hold both ASNT and NES Level III

Certifications in Ultrasonics and have been involved

with all aspects of Inservice Inspection Programs for

Nuclear Power Plants. A complete resume appears on

pages 21-22.

(Seaman) I am Senior Project Engineer for engineering

mechanics, power systems and structural engineering at

the Shoreham Nuclear Power Project. I am responsible

for the development of the Shoreham Pre-Service

Inspection Program. I have worked for the Daniel

International Corporation at the Enrico Fermi II

Nuclear Project in Michigan as an Engineer. A complete

resume appears on pages 23-24.

(Ranganath) I am a Manager of the Fracture Analysis

Unit for General Electric Company. I have been active

in the field of stress analysis for the past 10 years

both as an Adjunct Lecturer at the University of Santa

Clara and as an employee of General Electric Company.

A complete resume appears on pages 25-26.

3. Q. Would you please identify which portion of the testi-

mony you are sponsoring?

.

-3-

A. (Barry) I am co-sponsoring with Mr. Seaman the testi-

mony on SC 25, and am sponsoring the testimony on SOC

19(a)(1) and (2)..

(Seaman) I am co-sponsoring the testimony on SC 25

with Mr. Barry, and am sponsoring the testimony on SOC

19(a)(3).

(Ranganath) I am sponsoring the testimony on SOC

19(a)(4).

4. Q. Would you please summarize your conclusions presented

in this testimony?

A. The preservice inspection (PSI) and inservice inspec-

tion (ISI) programs meet with the applicable ASME Code

and therefore comply with 10 CFR 50.55a(g). In partic-

ular, the technology and the methods by which they are;

| applied in the examinations are established by the ASME

Code and together they assure the effectiveness of the

inspection programs. Because the basic technology will

not undergo significant change, there will be some cor-

relation between the PSI and ISI results.

Shoreham has taken a number of steps to reduce the num-

ber of noninspectable areas and a high percentage of

the plant will be inspected. Moreover, it is not

| necessary for all areas of a plant to be inspected

because results from inspected areas can be extended to!

noninspectable areas. The same controls on the quality

._

.

-4-

of welding materials, the same welding techniques and

quality assurance programs assure that noninspected

welds have the same quality as inspected welds. For

this reason, relief requests will have no impact on the

safety of Shoreham. Exemptions, similarly, have no

impact on safety, because they are permitted when

suitable alternatives have been performed.

On SOC Contention 19(a)(1), the significant quality

control requirements of Regulatory Guide 1.150 for the,

preservice inspection of the reactor vessel have been

met or exceeded. The Regulatory Guide does not estab-

lish " travel time," nor does it deal with ALARA con-

cerns. Finally, Regulatory Guide 1.2 has been complied

with at Shoreham. The Shoreham pressure vessel will r

behave in a non-brittle manner.

5. Q. Are you familiar with SC Contention 25?

A. Yes. The first part of the contention states that

"LILCO has not adequately demonstrated the effective-

ness of the technology and methods available that are

required to satisfy the inspection and tests specified

by 10 CFR 50, Appendix A, GDC 32, 36, 39 and 45."

6. Q. What inspection and tests are specified by GDC 32, 36,

39 and 45?

A. Those GDC's do not establish any inspections or tests

-- they deal with the design of various systems.

-.. - .- _

.

-5-

7. Q. Do the technology and methods used for the preservice

and inservice inspection programs meet applicable

standards?.

A. Yes. These inspections are established by, and conduc-

ted in accordance with the requirements of Section V of

the ASME Code, as required by 10 CFR 50.55(a)(g).

Shoreham usas four basic techniques for its inspection

program: ultrasonic testing, magnetic particle examin-

ation, liquid penetrant testing and visual examination.

This technology and the methods by which it is applied

is effective.

8. Q. What assurance do you have that the PSI examinations

are being conducted in accordance with the ASME_ Code?

A. In addition to LILCO's and NES Quality Assurance

Programs, the ASME Code requires that an independent

third party organization verify that the PSI and ISI

Programs comply with all requirements of the ASME Code.

LILCO has retained the Hartford Steam Boiler Inspection

'

and Insurance Company to perform these functions.

|,

9. Q. The County next contends that the PSI technology cannot:

be correlated to that used in the ISI program. Is the'

County correct?

| A. No. The same technology is planned for use in both the

PSI and ISI programs: liquid penetrant, ultrasonic

i testing, magnetic particle testing and visual examina-

tion. These technologies are not expected to undergo

.

-6-

significant changes. However, future amendments in the

ASME Code may require different tests or methods of

performing those tests.

10. Q. Even if there are different methodologies for conduct-

ing the examinations, will there be some correlation

between PSI and ISI examinations?

A. Yes, there will always be some correlation because the

technology remains essentially the same. Even where

methodologies change, correlation will remain.

11. Q. What is the purpose of having correlation between the

PSI and ISI programs?

A. Results of the PSI can be used to establish the

baseline condition of the welds. By comparing ISI

results to this baseline condition, a determination can

be made as to whether a new indication has appeared or

whether an indication has grown in size.

12. Q. Is it necessary that there be exact correlation between

PSI and ISI inspections?

A. No. Each indication disclosed during an ISI examina-

tion can be evaluated on its own without reference to

the PSI results. For every examination performed the

Code defines acceptance standards, and for this reason,

each ISI examination can be evaluated on its own with-

out reference to the PSI. Indeed, the ASME Code itself

changes, so that methodologies, will be different for

the various ISI inspections.

.

.

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13. Q. Let's move on to the next concern of the County which

is that "results from the inspected areas of the

reactor pressure boundary cannot be extended to.

non-inspectable areas." Is the County correct?

A. No. It is not necessary to inspect each area to have a

reasonable assurance that it does not have cracks.

Based on accepted statistical analyses, you can demon-

strate a very high probability that the non-inspected

areas are of the same quality as the inspected areas.

14. Q. Why does that assurance exist?

A. The results in the very high percentage of inspectable

areas can be extended to non-inopectable areas. This

assurance exists for a number of reasons: all mate-

r'.als, including welding materials, must meet the same

Code material specifications; all welding procedures

and techniques must meet the same Code requirements;

all welders are qualified in accordance with the same

Code Standards; and a common quality assurance program

must be applied to all of this work. All work is per-

formed in accordance with the appropriate Edition of

the ASME Code for construction, thereby establishingi

the " base-line" quality of these components.

Therefore, excellent correlation in terms of quality

exists.

15. Q. Has Shoreham requested exemptions from the inspection

program because the design and piping configurations

pre-date the latest Code?,

^

r~ . s n ..- -

_ 's c g,

i

.

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A. No. The Code establishes exemptions because it has

been determined that.if a plant meets;certain condi-'

- tions,7 1t is unnecessary to inspect the_ exempted areas.:

16. Q. Can you give a'n exanpl'e?

A. Yes. ASME Code XI? Section'IS-121(c), for example,/

| 'provides that ifi certain piping componants, less than 1,

.

| inch in diameter, are usually inspected during hydros-

tatic testing, they do not.nesd to'be: tested using

additional methods.

-

|

17. Q. So, in other words, exemptions from inspections are

provided in the Code, regardless of the date of thet

,

plant's design?

A. Yes.;

- 'i ,

_.

,18. Q. The County contends that the exemptions and' waivers

,vi$1_have an impact on the safety of the plant. Will~

'- thsy?' -

~. ,

A '.~ No. Gy-definition, exempt, ions, as discussed above, arepermitted by the ASME Code when' suitable alternatives

!'

have been met. So exemptions have no impact on safety..

By " waiver," the County probably means relief requests.~

~

Theae requests will be evaluated for th~eir impact on

i ~

- ; !.1i safety. Keep in mind, however, for the areas thurl s,

not be inspected there is a reasonable assurance of

their qaality.i

,

.

19. Jg. What has LILCO done to mitigate the, number of areas for

_

which relief requests will be. sought?

-

~

'I '

' N - . . _

'Y - , ~ r- r- 4 . - 6 e- - ~ = w + .'m

.

.

9

A. First, all welds on the reactor vessel have received a

pre-service inspection to verify the quality of the

'

welds. Additionally, accessibility for ISI was a.

design parameter for Shoreham. For areas where the

nondestructive testing (NDE) specified by the Code can-

not be performed, alternative NDE techniques will be

utilized to the maximum extent possible. While perfor-

ming the PSI, some areas have been identified where

interferences would limit ISI examination. Wherever

possible, corrective action, such as modification of

pipe supports, attachments, or structural steel, has

been taken. Weld profiles that will enhance ISI exa-

minations were specified in construction documents.

Removable insulation was specified for areas anticipa-

ted to be included in the ISI scope.

20. Q. Were any other actions taken?

A. Yes. In recognition of anticipated ISI examination

| requirements, Shoreham installed a fixed inspection

track system in the reactor vessel cavity. The track

system is designed to carry remote ultrasonic examina-

tion scanning equipment, thereby overcoming the limited

access in the vessel area. When ASME Code changes

indicated additional areas would have to be examined

during ISI, LILCO contracted with NES to provide addi-

tional inspection track capabilities. These tracks

serve the same purpose and function as the orginal

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

.

-10-

fixed system, but are installed and magnetically held

in place during outages when ISI is to be performed.

The magnetic tracks have already undergone a field test

program to assure their satisfactory operation. The

combined effect of the two track systems is to maximize

| areas which can be examined.

21. Q. When will the extent of the exemptions and relief

requests be identified?

A. At present, the PSI Program is still ongoing. Upon

completion, it will specify the number of non-

inspectable areas for PSI but all will have been eval-

uated for safety considerations. The exact number of

non-inspectable areas for ISI cannot be identified

until the precise ISI program is finalized. Shoreham

must meet the ASME Code that is in effect twelve months

prior to the date of the issuance of the operating

license, which has not yet been determined. The ISI

Program will be finalized once the applicable Code

Edition and Addenda are defined. It should be noted

that changes in the Code have been monitored and steps,

such as the use of magnetic tracks have been taken, in

anticipation of ISI requirements.

22. Q. Suffolk County contends that Shoreham will not meet 10:

| CFR 50.55(a)(g) because it will not comply with thet

appropriate ASME Code. Will Shoreham comply with the|

applicable Code?'

.

.

-11-

A. Yes, it will for the reasons discussed above.

23. Q. Are you familiar with SOC Contention 19(a)(1)?

* A. Yes. SOC contends that the quality control of the

ultrasonic testing (UT) equipment does not meet the

requirements of Regulatory Guide 1.150 and, thus, is

inadequate to provide reliable and reproducible UT

results.

24. Q. Are Regulatory Guide 1.150 requirements concerning UT

testing equipment applicable to the Shoreham reactor

pressure vessel preservice examination?

A. No. Regulatory Guide 1.150 is applicable to preservice

examinations performed after January 15, 1982.

Shoreham's preservice inspection of the reactor vessel

was completed in 1981. The significant quality control

requirements of Regulatory Guide 1.150 are met or

exceeded for the equipment used to perform the Shoreham

reactor pressure vessel preservice and inservice exam-

inations.

25. Q. What is the first requirement?

A. Paragraph C.1.1 is the most important quality control

requirement of Regulatory Guide 1.150. It requires

that calibration checks of UT equipment for screen

height linearity and amplitude linearity be made within

one day preceding and one day following the examina-

tions.

I

|

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_ _ _ _ - _ _ _ - _

-12-

26. Q. Does Shoreham meet the frequency requirements for

calibration checks?

A. Yes. Shoreham exceeds the frequency requirements for

calibration checks because it performs these checks of

the UT testing equipment on a daily basis throughout

the examination period.

27. Q. What is the next quality control requirement for the UT

testing equipment of Regulatory Guide 1.150?

A. Paragraph C.1.2 of Regulatory Guide 1.150 requires that

screen height linearity of ultrasonic instruments be

determined according to the ASME Code within the time

limits specified above, that is, within a day prior to

and within a day following the examination.

28. Q. Does Shoreham perform the check of screen height lin-

earity in accordance with the manner and time limits

specified by the Regulatory Guide?

A. Yes. Shoreham performs the check of screen height lin-

earity in accordance with the ASME Code and performs

these checks on a daily basis, which is more frequent

than the Regulatory Guide requires.

29. Q. What is the third requirement concerning quality con-

trol of UT equipment?

A. Position C.l.3 requires that amplitude control linear-

ity be determined in accordance with the ASME Code,

1977 edition, within one day preceding and within one

day following the examination.

. _ . . -

_.

.

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30. Q. Does Shoreham comply?,

A. Yes. Shoreham follows the ASME Code for determining

amplitude control linearity, and, as with the other

checks, performs them on a daily basis.

31. Q. Why does Shoreham perform more frequent instrument

checks than required by Regulatory Guide 1.150?

A. Good practice indicates that frequent checking of

equipment is desirable because if an instrument fails

during the course of a preservice or inservice inspec-

tion examinaton, it is not necessary to repeat the

entire examination. Rather, it would only be necessary

to repeat that portion of the examination performed

since the last satisfactory equipment check.

32. Q. What are the next quality control requirements of

Regulatory Guide 1.150 and have they been met at

Shoreham?

A. The Regulatory Guide then requires that photographic

records be obtained for: (i) the frequency amplitude

curve, and (ii) the unloaded initial pulse against a

calibrated time base. These photographic records have

not been obtained for the Shoreham preservice examina-

tion of the reactor pressure vessel.

33. Q. What is the purpose of these photographic records?

A. Although records provide information regarding the spe-

cific characteristics of the frequency amplitude curve

.

.

-14-

and pulse, their specific use is uncertain. Because of

the uncertainty of how this information could be uti-

lized to correlate different examination results, it is

likely that this requirement will be deleted or changed

in the future.

34. Q. Why are such photographic records of doubtful signifi-

cance?

A. They are only supplemental data which have no impact on

the reliability of the examinations and which will not

effect future ISI examinations. The more important

data is the screen height linearity and amplitude con-

trol linearity and, as discussed above, the performance

of these checks meet or exceed the Regulatory Guide

requirements.

35. Q. Are you familiar with SOC Contention 19(a)(2)?

A. Yes. SOC contends that "UT travel time does not meet

Regulatory Guide 1.150 and thus is inadequate to assure

detection of defects of significant length (larger than

the standard calibration holes) or significant depth."

36. Q. What ultrasonic testing examination travel time does

Regulatory Guide 1.150 establish?

A. Regulatory Guide 1.150 does not establish any UT travel

time..

37. Q. What is UT travel time?

A. In simple terms, it is the speed at which the UT

. . - .

- . _ _ _ _ _ _ _ _ _ _ _ _ _ _

,

.

-15-

transducer travels over the inspection or examination

surface.

38. Q. Is Shoreham's UT examination travel time adequate under

other applicable standards?

A. Yes. ASME Code XI, which is applicable to Shoreham,

restricts travel speed of test heads to a maximum of

six inches per second. The Code recognizes that

excessive travel times could result in missing

recordable indications and therefore established the

above travel speed. For Shoreham, examination speeds

were substantially less than this. Typically, travel

times were not in excess of 2 inches per second and in

no cases did the speeds exceed 6 inches per second.

39. Q. Do the regulations require Shoreham to follow ASME Code

XI?

A. Yes.

40. Q. Is the UT examination at Shoreham sufficient to detect

defects of significant length or depth?

A. Yes. NES has conducted tests to verify that the com-

bination of scan speed, pulse rate and data recording

level used at Shoreham will assure the detection of

recordable indications. Furthermore, following the

detection of a recordable indication, specific typing

and sizing of the indication in comparison to the cali-

bration standards is done in a static mode which is

identical to the examination calibration technique.

F<

-16-

41. Q. Are you familiar with SOC Contention 19(a)(3)?

A. Yes. SOC contends that ALARA has not been demonstrated

for examining personnel..

42. Q. Does Regulatory Guide 1.150 or 1.2 establish ALARA

guidelines?

A. No. Regulatory Guide 1.150 does not address ALARA,

therefore, it is outside the scope of this Contention.

43. Q. Are you familiar with SOC 19(a)(4)?

A. Yes. SOC contends that the structural integrity of the

pressure vessel at Shoreham has not been demonstrated

in accordance with Regulatory Guide 1.2.

44. Q. Does Shoreham comply with Regulatory Guide 1.2?

A. Yes, as discussed in the FSAR, Appendix 3B, Sec. 1.2,

Shoreham complies with Regulatory Guide 1.2. The

structural integrity of the Shoreham vessel has been

evaluated to demonstrate that brittle fracture will not

occur in the Shoreham vessel as a result of a

| loss-of-coolant-accident (LOCA).I

l 45. Q. What is the purpose of Regulatory Guide 1.2?

A. The injection of cold water by the emergency core cool-

ing system into a hot reactor vessel after a LOCA acci-

dent raises the concern that a vessel embrittled by.

| radiation could fail by brittle fracture because of theI

high stresses due to the thermal gradient. Regulatory

Guide 1.2 addresses this concern.

i

.

-17-4

46. Q. What are the specific requirements of Regulatory Guide !I

1.2? !

A. The requirements of Regulatory Guide 1.2 are twofold. I

First, it specifies that the Heavy Section Steel

Technology (HSST) program be monitored and that mate-

rial property data developed under the program be col-

lected and analyzed to verify nonbrittle behavior of

the reactor vessel materials. Second, it requires dem-

onstration of an acceptable safety margin against brit-

tle fracture of the vessel due to ECCS operation any

time during the vessel life. If such a margin can not

be demonstrated, Regulatory Guide 1.2 requires a demon-'

stration that an engineering solution, such as anneal-

ing, could be applied to ensure adequate toughness.

47. Q. How is the requirement on data collection and use sat-

isfied in the case of Shoreham?

A. The fracture evaluation of the Shoreham vessel was

based on a material toughness curve that provides a

lower bound to the data developed under the HSST pro-

gram. An industry wide task group, including GE

experts and sponsored by the Pressure Vessel Research

Committee of the Welding Research Council, developed|

| the lower bound curve based on the toughness data gen-|

erated under the HSST program. The validity of this

| curve has been confirmed by new data produced under

several programs including those sponsored by the

Electric Power Research Institute.

|

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.

.

-18-

48. Q. Was a fracture evaluation performed to demonstrate

i brittle fracture margin following a LOCA?!

A. Yes. A generic evaluation was first performed to eval-,

uate the fracture margin in a BWR vessel following a

LOCA as indicated in FSAR, Appendix 3.B. Sec. 1.2.

Subsequent analyses using more recent data have con-

firmed this result. See, for example, " Fracture

Mechanics Evaluation of a Boiling Water Reactor Vessel

Following a Postulated Loss-Of-Coolant-Accident,"

Volume G, Transactions of the 5th International

' Conference on Structural Mechanics in Reactor

Technology, Berlin, Germany, August 1979. Shoreham

specific data confirm these generic conclusions.

49. Q. What factors were examined in the evaluations?

A. The most important factors, as discussed in the FSAR

Appendix 3B, Section 1.2 were:

(1) a comprehensive thermal analysis considering the

effect of blowdown and the low-pressure coolant

injection system reflooding;

(2) a stress analysis considering the effects of pres-

sure, temperature, and residual stresses;

(3) the radiation effect on material toughness;

(4) methods for calculating crack tip stress intensity;

associated with a nonuniform stress field following

the design basis accident.

._. ._- - -

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.

-19-

50. Q. What do these analyses show?

A. The fracture analysis demonstrated that the Shoreham

vessel has a high margin of protection against brittle

fracture following a design basis LOCA. For example,

for a quarter thickness flaw, 1 1/2 inches deep and 9

inches long, the safety margin against brittle fracture

is a factor of 2.

51. Q. What causes the high brittle fracture margin in the

Shoreham vessel?

A. The Shoreham vessel has significant advantages since

the key factors required for unstable crack propagation

-- radiation embrittlement and high pressure stresses

following thermal shock -- do not occur in a BWR. The

reasons for this are:

(1) The radiation flux level at the wall location in

j the Shoreham vessel is not enough to cause appre-

ciable embrittlement of the vessel during its

lifetime. The low rad..ation flux level is due to

the fact that there is a water annulus between the

core and the vessel, and that there is low power

density in a BWR. Thus the toughness level

required to assure ductile behavior is maintained

throughout the design life.

|(2) Unstable crack propagation requires the presence of

pressure stresses following the thermal shock.

However, in a BWR, when there is emergency

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injection of coolant following a LOCA, the vessel

pressure drops automatically as the temperature

drops since the pressure follows the saturation.

curve . The absence of significant pressure,

stresses assures that unstable crack propagation

will not occur in the Shoreham vessel.

54. Q. What are your conclusions on the structural integrity

of the Shoreham vessel?

A. In summary, the structural integrity of the Shoreham

vessel has been demonstrated in accordance with

Regulatory Guide 1.2 by:

(1) showing that crack propagation will not occur in

the vessel following a LOCA;

(2) using lower bound toughness properties based on

extensive materials data collected under several

research programs to verify nonbrittle behavior;

(3) assuring that a high margin of safety against brit-

| tle fracture exists for postulated accident condi-

tions throughout the design life.

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1

PROFESSIONAL QUALIFICATIONS

Peter Stuyvesant Barry

Director of Technical Operations,

Inservice Inspection,

,

Nuclear Energy Services, Inc.4

1

My name is Peter Barry. My business address is Nuclear

Energy Services, Inc., Shelter Rock Road, Danbury, Connecticut.

I am employed by Nuclear Energy Services (NES) as the Director

of Technical Operations for its Inservice Inspection activi-

ties. I have been employed by NES since the company was formed

in 1974.

My college training (Yale University, University of

I Bridgeport) was in Liberal Arts. I have worked in the field of

Nondestructive Testing, specifically ultrasonic testing, since

1966. At that time, I took employment with Branson Instruments

Company in Stamford, a manufacturer of ultrasonic equipment and

became an Applications Engineer in 1967. Subsequently, in

1970, I took a position with Krautkramer Ultrasonics, again as

an Ultrasonic Applications Engineer. This position included

teaching responsibilities as well as field test activities.

In 1973 I joined Sperry Products Division of Automation

Industries, also a manufacturer of ultrasonic and other

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nondestructive testing equipment, again as an Applications

Engineer. It was at this time that I became involved with NDE

Applications related to Nuclear Power Plants. I became Manager

of Nuclear Applications in 1974.

I have been a Level III Examiner in ultrasonics since my.

employment with Krautkramer Ultrasonics. I currently hold both

ASNT and NES Level III Certifications in Ula.rasonics. I am

! also a member of the ASME Section XI Subcommittee's Working

Group on Nondestructive Examinations.

During my employment with NES, I have been involved with

all aspects of Inservice Inspection Programs for Nuclear Power

Plants. Currently I am responsible for establishing technical

policy for the Company's ISI activities.

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.

PROFESSIONAL QUALIFICATIONS

Craig K. Seaman

Senior Assistant Project Engineer

Long Island Lighting Company

My name is Craig K. Seaman. My business address is

Shoreham Nuclear Power Station, P.O. Box 618, Wading River, New

York. I am employed by the Long Island Lighting Company

(LILCO) as a Senior Assistant Project Engineer for engineering

mechanics, power systems and structural engineering on the

Shoreham Nuclear Power Project (Shoreham). I have been

employed by LILCO since 1979 and previously from 1975 to 1978.

I received the degree of Bachelor of Science in Engineering

from Cornell University in 1975 and have taken several graduate

level courses in Nuclear Engineering at Brooklyn Polytechnic

Institute. In 1978, I attended a course in the ASME Code

Section III and, in 1979, attended a course in the ASME Code

Section XI. In 1980, I attended the BWR Design Orientation

course at the General Electric Training Center.

I worked as an Engineer and Construction Supervisor in the

LILCO Construction Division at the Shoreham Nuclear Project

from 1975 through January 1978. From February 1978 through

.

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August 1979, I served as an Engineer for the Daniel

International Corporation at the Enrico Fermi II Nuclear

Project in Michigan.

In August 1979, I rejoined the Long Island Lighting Company

as an Assistant Project Engineer at Shoreham. In the Fall of

1979, I was assigned LILCO responsibility for development of

the Shoreham Pre-Service Inspection Program, and have been con-

tinuously involved in its development since that time. In

December 1981, I was promoted to Senior Assistant Project

Engineer for the Engineering Mechanics, Power Systems and

Structural Engineering Section.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ ._. _ _ . _ ___ ___ _ __ ____________ _

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-25-

Dr. Sam Ranganath

Manager, Stress and Fracture Analysis Unit

General Electric Company.

San Jose, California

My name is Sam Ranganath. My business address is 175

Curtner Avenue, San Jose, California 95125. I am employed by

General Electric Company as Manager of the Stress Fracture

Analysis Unit. I have held this position since April 1978. My

group has the overall responsibility for performing stress

analysis, fracture mechanics and fatigue evaluations for

Boiling Water Reactor (BWR) pressure vessel components. This

includes applications of ASME Code, preparation of Code cer-

tified stress reports, participation in external research con-

tracts involving material behavior and the evaluation of

stress / fracture related problems in operating plants. I have

been employed by General Electric Company since 1974.

I completed my Ph.D. in Engineering at Brown University,

Providence, Rhode Rhode Island in 1971. I was a Post Doctoral

Fellow at Brown University from December 1970 to November 1971.

In 1981, I received a Master of Business Administration degree

from the University of Santa Clara. I have been active in the

field of stress analysis, fracture Sachanics and material

behavior for the past ten years and have published several

technical papers.

I am also an Adjunct Lecturer at the University of Santa

Clara and teach graduate courses in pressure vessel design and

fracture mechanics.

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I am a member of the ASME Section XI Subgroup on Evaluation

and Standards. This group is involved in developing fracture

mechanics procedures to evaluate pressure vessel components. I

am a Registered Professional Engineer in the State of

California.

List of Publications in the Field ofFracture Mechanics and Stress Analysis

1. " Fracture Mechanics Evaluation of a Boiling Water ReactorVessel Following a Loss of Coolant Accident." Proceedingsof the 5th International Conference on Structural Mechanicsin Reactor Technology, Berlin, Germany, August 1979.

2. Environmental Crack Growth Analysis Based on Elastic-Plastic Fracture Me'chanics"(coauthor with H. S. Mehta).Presented at the 1992 Pressure Vessels and PipingConference, Orlando, Florida, June 1982; ASME Paper82-PVP-23.

3. " Residual Stress Analysis of Piping with Pre-ExistingCracks Subjected to the Induction Heating StressImprovement" (coauthor with M. L. Herrera and H. S. Mehta);presented at the 1982 Pressure Vessels and PipingConference, Orlando, Florida, June 1982; ASME Paper82-PVP-60.

4. " Fatigue Behavior of Carbon Steel Components in HighTemperature Water Environments" (with J. N. Kass and J. D.Heald in Low Cycle Fatigue and Life Prediction, ASTM STP770, American Society for Testing and Materials 1982.

5. " Failure Analysis, Testing and Product Improvement of aControl Rod Drive Component from a Boiling Water Reactor"(with J. N. Kass, D. E. Delwiche and D. L. Peterson) inFailure Prevention and Reliability, Proceedings of the 1977Failure Prevention and Reliability Conference, Chicago,Illinois, American Society of Mechanical Engineers.

6. " Elastic-Plastic Stress Analysis and ASME Code Evaluationof a Bottomhead Penetration in a Reactor Pressure Vessel"presented at the 1979 Pressure Vessels and PipingConference, San Francisco, California, June 1979, ASMEPaper 79-PVP-17.

7. " Engineering Methods for the Assessment of Ductile FractureMargin in Nuclear Power Plant Piping" (with H. S. Mehta)Presented at the Second International Symposium on Elastic-Plastic Fracture, Ihiladelphia, Pennsylvania, Octcher 1981.(To appear in the new ASTM Special Technical Publication onElastic Plastic Fracture) American Society for Testing andMaterials.

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