ML20054M252
| ML20054M252 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 06/14/1982 |
| From: | Barry P, Ranganath S, Seaman C GENERAL ELECTRIC CO., LONG ISLAND LIGHTING CO., NUCLEAR ENERGY SERVICES, INC. |
| To: | |
| Shared Package | |
| ML20054M250 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8207120169 | |
| Download: ML20054M252 (3) | |
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< 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 brittle fracture following a design basis LOCA. For example, for a uarter 1 1/2 inches ickness 6afd flaw,$m agaimst deep and 9 inches long, thegbri[tle fracture rher
-'/s i O' = 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 in the 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 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 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 820707 PDR ADOCK 05000322 T PDR o
s CERTIFICATE OF SERVICE In the Matter of LONG ISLAND LIGHTING COMPANY (Shoreham Nuclear Power Station, Unit 1)
Docket No. 50-322 (OL)
I hereby certify that copies of CORRECTIONS AND ADDITIONS TO THE " TESTIMONY OF PETER S. BARRY, CRAIG K. SEAMAN AND SAM RANGANATH ON SUFFOLK COUNTY CONTENTION 25 AND SHOREHAM '
OPPONENTS COALITION 19 (a) -- PRESERVICE AND INSERVICE INSPECTION PROGRAM AND REACTOR PRESSURE VESSEL INTEGRITY" were served upon the following by first-class mail, postage prepaid, on July 7, 1982:
Lawrence Brenner, Esq. Atomic Safety and Licensing Administrative Judge Appeal Board Panel Atomic Safety and Licensing U.S. Nuclear Regulatory Board Panel Commission U.S. Nuclear Regulatory Washington, D.C. 20555 Commission Washington, D.C. 20555 Atomic Safety and Licensing Board Panel i Dr. Peter A. Morris U.S. Nuclear Regulatory Administrative Judge Commission l Atomic Safety and Licensing Washington, D.C. 20555 Board Panel U.S. Nuclear Regulatory Bernard M. Bordenick, Esq.
Commission David A. Repka, Esq.
Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Dr. James H. Carpenter Washington, D.C. 20555 Administrative Judge Atomic 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 Law Washington, D.C. 20555 Veterans Memorial Highway Hauppauge, New York 11787 l Secretary of the Commission Stephen B. Latham, Esq.
U.S. Nuclear Regulatory Twomey, Latham & Shea Commission 33 West Second Street Washington, D.C. 20555 P. O. Box 398 l 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 Street Kirkpatrick, Lockhart, Hill, New York, New York 11901 Christopher & Phillips Albany, New York 12223 1
6 i
l 8th Floor l 1900 M Street, N.W. Howard L. Blau, Esq. l Washington, D.C. 20036 217 Newbridge Road Hicksville, New York 11801 Mr. Mark W. Goldsmith Energy Research Group Matthew J. Kelly, Esq.
400-1 Totten Pond Road State of New York Waltham, Massachusetts 02154 Department of Public Service Three Empire State Plaza MHB Technical Associates Albany, New York 12223 1723 Hamilton Avenue Suite K Mr. Jay Dunkleberger San Jose, California 95125 New York State Energy Office Agency Building 2 Empire State Plaza Albany, New York 12223 Respectfully submitted, LONG ISLAND LIGHTING COMPANY 0bh e
' Daniel O. FlanaganNJ Hunton & Williams 707 East Main Street P.O. Box 1535 Richmond, Virginia 23212 l
l
f LILCO, June 14, 1982 UNITED STATES OF AMERICA NUCLEAR 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 AND SHOREHAM OPPONENTS COALITION 19(a) --
PRESERVICE AND INSERVICE INSPECTION PROGRAM AND REACTOR PRESSURE VESSEL INTEGRITY PURPOSE ,
This testimony demonstrates that the preservice inspec- t tion (PSI) and inservice inspection (ISI) programs comply with l 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 some i
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 ,
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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 permitted when 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 concerns. Regulatory Guide 1.2 has been complied with at 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 l In the Matter of )
! )
i LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 (OL) l )
i (Shoreham Nuclear Power Station, )
Unit 1)
)
TESTIMONY OF PETER S. BARRY, CRAIG K. SEAMAN AND SAM PANGANATH ON SUFFOLK COUNTY CONTENTION 25 AND SHOREHAM OPPONENTS COALITION 19(a) --
PRESERVICE AND INSERVICE INSPECTION-l PROGRAM AND REACTOR PRESSURE VESSEL INTEGRITY
- 1. Q. Would you please state your names?
l l A. My name is Peter S. Barry. My business address is
! Nuclear Energy Services, Shelter Rock Road, Danbury, l
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. 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?
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 l
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 l 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
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.
- 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.
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- 9. Q. The County next contends that the PSI technology cannot be correlated to that used in the ISI program. Is the County correct?
l 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
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.
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.
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.
- 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 establishing i
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?
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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,
/
l , 'provides that ifi certain piping componants, less than 1 l inch in diameter, are usually inspected during hydros-tatic testing, they do not.nesd to'be: tested using additional methods.
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- 17. Q. So, in other words, exemptions from inspections are provided in the Code, regardless of the date of the t
plant's design?
A. Yes.
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,18. Q. The County contends that the exemptions and' waivers
,vi$1_have an impact on the safety of the plant.
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Will thsy?' -
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A '.~ No. Gy-definition, exempt, ions, as discussed above, are permitted 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.
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Theae requests will be evaluated for th~eir impact on i
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i safety. Keep in mind, however, for the areas thur l , s not be inspected there is a reasonable assurance of their qaality.
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- 19. Jg. What has LILCO done to mitigate the, number of areas for which relief requests will be. sought?
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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 l 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
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.
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- 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 l CFR 50.55(a)(g) because it will not comply with the t
appropriate ASME Code. Will Shoreham comply with the l
applicable Code?
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.
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- 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.
- 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
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
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
- 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 l loss-of-coolant-accident (LOCA).
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- 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 l
radiation could fail by brittle fracture because of the I
high stresses due to the thermal gradient. Regulatory Guide 1.2 addresses this concern.
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- 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 l
l the lower bound curve based on the toughness data gen-l erated under the HSST program. The validity of this l
curve has been confirmed by new data produced under several programs including those sponsored by the Electric Power Research Institute.
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- 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.
- 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 l
pressure stresses following the thermal shock.
However, in a BWR, when there is emergency l
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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-l 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
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
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.
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.
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 of Fracture Mechanics and Stress Analysis
- 1. " Fracture Mechanics Evaluation of a Boiling Water Reactor Vessel Following a Loss of Coolant Accident." Proceedings of the 5th International Conference on Structural Mechanics in 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 Piping Conference, Orlando, Florida, June 1982; ASME Paper 82-PVP-23.
- 3. " Residual Stress Analysis of Piping with Pre-Existing Cracks Subjected to the Induction Heating Stress Improvement" (coauthor with M. L. Herrera and H. S. Mehta);
presented at the 1982 Pressure Vessels and Piping Conference, Orlando, Florida, June 1982; ASME Paper 82-PVP-60.
- 4. " Fatigue Behavior of Carbon Steel Components in High Temperature Water Environments" (with J. N. Kass and J. D.
Heald in Low Cycle Fatigue and Life Prediction, ASTM STP 770, American Society for Testing and Materials 1982.
- 5. " Failure Analysis, Testing and Product Improvement of a Control Rod Drive Component from a Boiling Water Reactor" (with J. N. Kass, D. E. Delwiche and D. L. Peterson) in Failure Prevention and Reliability, Proceedings of the 1977 Failure Prevention and Reliability Conference, Chicago, Illinois, American Society of Mechanical Engineers.
- 6. " Elastic-Plastic Stress Analysis and ASME Code Evaluation of a Bottomhead Penetration in a Reactor Pressure Vessel" presented at the 1979 Pressure Vessels and Piping Conference, San Francisco, California, June 1979, ASME Paper 79-PVP-17.
- 7. " Engineering Methods for the Assessment of Ductile Fracture Margin 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 on Elastic Plastic Fracture) American Society for Testing and Materials.
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